Text

License Amendment Request Nos. 310 and 182
	ML050480264
Person / Time
Site:	Beaver Valley
Issue date:	02/11/2005
From:	Mende R; FirstEnergy Nuclear Operating Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	L-05-009
	Download: ML050480264 (158)
	v • d • e

FENOC Beaver Valley Power Station Route 168 P.O. Box 4 FirstEnergy Nuclear Operating Company Shippingport, PA 15077-0004 Richard G. Menzde 724-682-5206 Director, Performance Improvement February 11, 2005 L-05-009 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001

Subject:

Beaver Valley Power Station, Unit No. 1 and No. 2 BV-1 Docket No. 50-334, License No. DPR-66 BV-2 Docket No. 50-412, License No. NPF-73 License Amendment Request Nos. 310 and 182 Pursuant to 10 CFR 50.90, FirstEnergy Nuclear Operating Company (FENOC) requests an amendment to the above licenses in the form of changes to the Beaver Valley Power Station (BVPS) Technical Specifications. This License Amendment Request (LAR) proposes 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, and increase the ability to return to power after a plant trip while still maintaining margin to safety limits under all operating conditions.

The proposed Technical Specification changes are provided in Attachments A-1 and A-2 for Unit Nos. 1 and 2, respectively. The proposed changes to the Technical Specification Bases are provided in Attachments B-1 and B-2 for Unit Nos. 1 and 2, respectively. The proposed changes to the Licensing Requirements Manual (LRM) are provided in Attachments C-1 and C-2 for Unit Nos. 1 and 2, respectively. The Technical Specification Bases and LRM changes are provided for information only.

FENOC requests approval of the proposed amendments by January 2006 in order to support the RAOC analysis which assumes Extended Power Uprate (EPU) conditions, including the Best Estimate Loss of Coolant Accident (BELOCA) methodology. For each unit the RAOC amendments will be implemented concurrent with the applicable unit's EPU and BELOCA amendments. Thus, FENOC requests the following implementation periods.

The Unit No. 1 RAOC amendment shall be implemented prior to the first entry into Mode 4 during plant startup from the IR17 refueling outage planned for the spring of 2006. The Unit No. 2 RAOC amendment shall be implemented prior to the first entry into Mode 4 during plant startup from the 2R12 refueling outage planned for the fall of 2006.

Beaver Valley Power Station, Unit Nos. 1 and 2 License Amendment Request Nos. 310 and 182 L-05-009 Page 2 The Beaver Valley Power Station review committees have reviewed the proposed changes. The changes were determined to be safe and do not involve a significant hazard consideration as defined in 10 CFR 50.92 based on the attached safety analysis and no significant hazard evaluation.

No new commitments are contained in this submittal. If there are any questions concerning this matter, please contact Mr. Henry L Hegrat, Supervisor, Licensing at 330-315-6944.

I declare under penalty of perjury that the foregoing is true and correct. Executed on February WL, 2005.

Sincerely, Richard G. Mende

Enclosure:

FENOC Evaluation of the Proposed Changes Attachments:

A-1 Proposed Unit No. 1 Technical Specification Changes A-2 Proposed Unit No. 2 Technical Specification Changes B-1 Proposed Unit No. I Technical Specification Bases Changes B-2 Proposed Unit No. 2 Technical Specification Bases Changes C-I Proposed Unit No. 1 Licensing Requirements Manual Changes C-2 Proposed Unit No. 2 Licensing Requirements Manual Changes c: Mr. T. G. Colburn, NRR Senior Project Manager Mr. P. C. Cataldo, NRC Senior Resident Inspector Mr. S. J. Collins, NRC Region I Administrator Mr. D. A. Allard, Director BRP/DEP Mr. L. E. Ryan (BRP/DEP)

ENCLOSURE FENOC Evaluation of the Proposed Changes Beaver Valley Power Station License Amendment Requests 310 (Unit No. 1) and 182 (Unit No. 2)

Subject:

Application to Implement the Relaxed Axial Offset Control (RAOC) and Heat Flux Hot Channel Factor (FQ) Surveillance Methodologies Table of Contents Section Title Page

1.0 DESCRIPTION

............................... 1

2.0 PROPOSED CHANGE

S ................................ 1 2.1 Proposed TS Changes ................................ 2 2.2 Proposed TS Bases and LRM Changes ................................ 7

3.0 BACKGROUND

.8 .................. 8

4.0 TECHNICAL ANALYSIS

.................. 9 4.1 Non-LOCA Related Evaluation ........... .................... 11 4.2 LOCA and LOCA-Related Evaluations ............................... 12 4.3 Core Design Evaluation ............................... 12 4.4 Other Areas ............................... 13 4.5 Conclusion ............................... 14 5.0 REGULATORY SAFETY ANALYSIS ............................... 14 5.1 No Significant Hazards Consideration ............................... 14 5.2 Applicable Regulatory Requirements/Criteria .............................. 17

6.0 ENVIRONMENTAL CONSIDERATION

............................... 18

7.0 REFERENCES

............................... 18 Attachments Number Title A-1 Proposed Unit No. 1 Technical Specification Changes A-2 Proposed Unit No. 2 Technical Specification Changes B-1 Proposed Unit No. 1 Technical Specification Bases Changes B-2 Proposed Unit No. 2 Technical Specification Bases Changes C-1 Proposed Unit No. 1 Licensing Requirements Manual Changes C-2 Proposed Unit No. 2 Licensing Requirements Manual Changes i

Beaver Valley Power Station License Amendment Requests 310 (Unit 1) and 182 (Unit 2)

1.0 DESCRIPTION

This License Amendment Request (LAR) is a request to amend Operating Licenses DPR-66 (Beaver Valley Power Station Unit No. 1) and NPF-73 (Beaver Valley Power Station Unit No. 2). The proposed changes will revise the Operating Licenses 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 and to increase the ability to return to power after a plant trip while still maintaining margin to safety limits under all operating conditions.

The changes proposed to Technical Specifications (TS) 3.2.1, AXIAL FLUX DIFFERENCE (AFD), and 3.2.2, HEAT FLUX HOT CHANNEL FACTOR - FQ (Z), are being made to adopt the RAOC calculational procedure of the Standard Technical Specifications (STS), i.e., NUREG-1431, "Standard Westinghouse Technical Specifications Westinghouse Plants" (Reference 1). Changes to the other TS listed in the enclosed table are made to provide consistency with the changes made to TS 3.2.1 and 3.2.2.

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 (Reference 2).

2.0 PROPOSED CHANGE

S The specific changes to the TS that are proposed are shown in Attachments A-I and A-2 for Beaver Valley Power Station (BVPS) Unit Nos. 1 and 2, respectively. Changes to the TS Bases are shown in Attachments B-1 and B-2, respectively. The changes proposed to the Licensing Requirements Manual (LRM) are provided in Attachments C-1 and C-2 for Unit Nos. 1 and 2, respectively.

The proposed Technical Specification Bases and LRM changes do not require NRC approval. The BVPS Technical Specification Bases Control Program controls the review, approval and implementation of Technical Specification Bases changes. The BVPS Licensing Document Control Program controls the review, approval and implementation of LRM changes.

Changes to these two documents are controlled by the 10 CFR 50.59 Page 1

Beaver Valley Power Station License Amendment Requests 310 (Unit 1) and 182 (Unit 2) process. The Technical Specification Bases and LRM changes are provided for information only.

The proposed changes to the Technical Specifications, TS Bases and LRM have been prepared electronically. Deletions are shown with a strike-through and insertions are shown double-underlined. This presentation allows the reviewer to readily identify the information that has been deleted and added.

To meet format requirements, the appropriate Indices, the Technical Specifications, the TS Bases and the LRM pages will be revised and repaginated as necessary to reflect the changes being proposed by this LAR.

Changes to the following Technical Specifications are being proposed in this LAR.

I Affected Technical Specifications l No. Unit 1 Unit 2 Title 1 3.2.1 3.2.1 AXIAL FLUX DIFFERENCE (AFD) 2 3.2.2 3.2.2 HEAT FLUX HOT CHANNEL FACTOR - FQ (Z) 3 3.2.3 3.2.3 NUCLEAR ENTHALPY HOT CHANNEL FACTOR - F1XH 4 3.2.4 3.2.4 QUADRANT POWER TILT RATIO (QPTR) 5 3.3.1 3.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION (Table 4.3-1, Note 3) 6 6.9.5 6.9.5 CORE OPERATING LIMITS REPORT (COLR)

All of the proposed TS and TS Bases changes are consistent with the STS (Reference 1).

2.1 Proposed TS Changes Change Number I This proposed change is a re-write of Technical Specification 3.2.1, "AXIAL FLUX DIFFERENCE (AFD)". The Limiting Condition for Operation (LCO), Actions and Surveillance Requirements are revised to be consistent with the STS RAOC version of this TS. Surveillance Page 2

Beaver Valley Power Station License Amendment Requests 310 (Unit 1)and 182 (Unit 2)

Requirement 4.2.1.2 is revised to be consistent with the RAOC methodology and is included as a footnote to the LCO statement. Surveillance Requirements 4.2.1.3 and 4.2.1.4 are deleted.

Basis for Change Number 1 The proposed changes are being made to be consistent with the RAOC methodology and the STS (Reference 1).

The current Technical Specification 3.2.1, per the core operating limits report (COLR), specifies a target band of +7%, -7% for normal operation from 0% to 100% Rated Thermal Power (RTP). The RAOC methodology allows an Axial Flux Difference (AFD) operating space relaxation to +10%,

-15% Al at 100% RTP and linearly increasing to +24%, -32% Al at 50%

RTP. Penalty minutes are not applicable to the RAOC methodology and therefore reference to them is no longer necessary in the TS or TS Bases.

The necessary LRM changes are discussed in Section 2.2.

For the RAOC methodology, the value of the AFD does not affect the limiting accident consequences with thermal power less than 50% RTP.

Reducing the power range neutron flux high setpoints is not necessary to provide an adequate level of protection. Reducing the power level to less than or equal to 50% RTP maintains the plant in a benign condition since under RAOC methodology there are no AFD limits below 50% RTP. In addition, a rapid rise in power to greater than 50% RTP, with AFD outside limits, does not immediately create an unacceptable situation. Since the transient analysis setpoint calculations for f(AI), input to the Overtemperature AT Trip Function, are based on the same core power distributions used in a reload cycle design, the Overtemperature AT Trip Function provides an acceptable level of protection for such an excursion. It is also noted that the event would be successfully terminated by a trip at the previous setpoint level.

Surveillance Requirement 4.2.1.1 has been revised to be consistent with the STS. The first sentence of Surveillance Requirement has been moved to Note 1, which is applicable to the LCO statement. The remaining portion of this surveillance can be deleted because penalty minutes are not applicable to RAOC. The Note on the Applicability has been designated with a number instead of an asterisk for clarity. This is an editorial change that requires no further justification.

Surveillance Requirements 4.2.1.3 and 4.2.1.4 are deleted because they pertain to the target flux difference, which is not applicable to RAOC. With Page 3

Beaver Valley Power Station License Amendment Requests 310 (Unit 1) and 182 (Unit 2)

RAOC there is no target band. In addition, the activity required by Surveillance Requirement 4.2.1.4 is addressed by Notation (3) of Table 4.3-1.

Change Number 2 This proposed change is a re-write of Technical Specification 3.2.2, "HEAT FLUX HOT CHANNEL FACTOR - FQ (Z)". The LCO is revised to indicate that two limits exist, i.e., FcQ(Z) and F'wQ(Z). The Actions and Surveillance Requirements have also been revised to reflect the FQ(Z) surveillance methodology. Several Notes have also been added to specify Action and Surveillance Requirement conditions.

Basis for Change Number 2 The proposed changes are being made to be consistent with the RAOC methodology and the STS.

Action "a" of TS 3.2.2 is revised to change the time allowed to reduce the Power Range Neutron Flux-High Trip from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . As written, the completion time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to reduce the Power Range Neutron Flux-High Trip setpoints presents an unjustified burden on the operation of the plant. A completion time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months will allow time to perform a second flux map to confirm the results, or determine that the condition was temporary, without implementing an unnecessary trip setpoint change, during which there is increased potential for a plant transient and human error. Following a significant power reduction, at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are required to re-establish steady state xenon prior to taking a flux map, plus additional time to obtain a flux map and analyze the data. A significant potential for human error can be created through requiring the trip setpoints to be reduced within the same time frame that a unit power reduction is taking place within the current completion time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . To account for setpoint adjustments and any necessary initial preparation, a completion time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is proposed. The completion time extension is acceptable because Action a. 1 requires a reduction in power well before the proposed 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and it provides enough time to safely take the necessary steps to determine if a setpoint change is indeed required.

Also, "after each FQ(Z) determination" has been added to Action "a" to define the frequency of the action requirement. The proposed change will require the Thermal Power, the Power Range Neutron Flux-High Trip Setpoints, and the Overpower AT Trip Setpoints reductions be repeated after each subsequent FQ(Z) determination if FQ(Z) is not within limit. This will Page 4

Beaver Valley Power Station License Amendment Requests 310 (Unit 1) and 182 (Unit 2) ensure that Actions are continued until the parameter is within its limit at the current power level.

In Action "a" the requirement to perform the Overtemperature AT Trip Setpoint reduction with the reactor subcritical is being deleted. This change is made to provide consistency with Reference 1.

Action "b", which requires the out of limit condition of Action "a" to be identified and corrected prior to increasing THERMAL POWER, is being replaced by an action that requires the verification that FCQ(Z) and FWQ(Z) are within limits prior to increasing THERMAL POWER above the limits of Action "a". These two actions ensure that core conditions during operation at higher power levels and future operations are consistent with safety analyses assumptions.

Actions a.4 and b.4 have been added to TS 3.2.2. These actions are applicable to the measurement of the peak fuel pellet power within the reactor core at the steady state power, which is known as FcQ(Z), and the measurement of the peak fuel pellet within the reactor core that is adjusted for power distribution transients encountered during normal operation, which is known as FWQ(Z).

Actions a.5 and b.5 have been added to TS 3.2.2 to define the alternative for not meeting Actions "a" or "b". These Actions are more restrictive than the Actions provided in the current BVPS TS. These Actions require that the unit be placed in Mode 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months if THERMAL POWER is not reduced to comply with the Action.

Footnotes (1) and (2) have been added to Actions a and b respectively, to require that verification of FcQ(Z) and FWQ(Z) shall be completed whenever Actions a or b are entered.

Surveillance Requirement 4.2.2.2 and 4.2.2.3 have been revised to incorporate the FQ(Z) surveillance strategy to determine FQ(Z) is within its limits. These revised Surveillance Requirements are also required for FCQ(Z) and FWQ(Z) any time that they exceed their limits prior to increasing power as described in their respective Actions.

Footnote (3), added to each of these Surveillance Requirements, allows power escalation and delays obtaining a power distribution map at the beginning of a cycle until an equilibrium power level is reached.

Footnote (4) is added to Surveillance Requirement 4.2.2.3 to indicate that additional actions are required when the maximum over z of [FcQ(Z)/K(Z)]

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Beaver Valley Power Station License Amendment Requests 310 (Unit 1) and 182 (Unit 2) increases over the previous evaluation of FCQ(Z). This footnote ensures that future operations are consistent with safety analyses assumptions.

The Action and Surveillance Requirement Notes have been designated with numbers instead of asterisks for clarity. Replacing the asterisks with numbers are editorial changes that require no further justification.

Change Number 3 This proposed change consists of two modifications of Technical Specification 3.2.3, NUCLEAR ENTHALPY HOT CHANNEL FACTOR.

The first is to change the nomenclature used for CFDH and PFDH. The second is to insert "RISE" between "ENTHALPY" and "HOT" in the TS title.

Basis for Change Number 3 The proposed changes are being made to make the TS nomenclature consistent with the COLR nomenclature for CFDH and PFDH. The COLR nomenclature is CFAI for CFDH and PFAII for PFDH. The TS title change, which is carried through the TS Index and Bases, results in consistency with the STS and COLR. These changes are editorial and require no further justification.

Change Number 4 This proposed change is a modification of Technical Specification 3.2.4, "QUADRANT POWER TILT RATIO (QPTR)", to reflect the FCQ(Z) and FWQ(Z) surveillance requirements.

Basis for Change Number 4 The proposed change is being made to be consistent with the RAOC methodology and the STS.

Change Number 5 This proposed change is a modification of the Notation to Table 4.3-1 of Technical Specification 3.3.1, "REACTOR TRIP SYSTEM INSTRUMENTATION". Notation 3 of Table 4.3-1 is modified by changing "15%" to "50%" of RATED THERMAL POWER.

Basis for Change Number 5 The proposed change is being made to be consistent with the change to Surveillance Requirement 4.2.1.1.

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Change Number 6 This proposed change is a modification of Technical Specification 6.9.5, "CORE OPERATING LIMITS REPORT (COLR)".

The change consists of revising TS 6.9.5.a by replacing "Constant" with "Relaxed" for consistency with the FQ surveillance methodology.

In addition TS 6.9.5.b is revised:

(a) by adding WCAP-10216-P-A, Relaxation of Constant Axial Offset Control FQ Surveillance Technical Specification (Reference 2) to the list of references, and (b) by deleting WCAP-8385, "POWER DISTRIBUTION CONTROL AND LOAD FOLLOWING PROCEDURES -

TOPICAL REPORT", the T. M. Anderson to K. Kniel letter, and NUREG-0800, Standard Review Plan, U.S. Nuclear Regulatory Commission, Section 4.3, from the list of references.

Basis for Change Number 6 The proposed change is being made to be consistent with the RAOC methodology and the STS. A reference to WCAP-10216-P-A is being added because it is applicable to the RAOC methodology. The references to WCAP-8385, the T. M. Anderson to K. Kniel letter, and NUREG-0800 are being deleted because they are, applicable to the Constant Axial Offset Control (CAOC) methodology, not the RAOC methodology.

2.2 Proposed TS Bases and LRM Changes Since TS 3.2.1 and 3.2.2 are being revised to reflect the STS, it is prudent to also expand the TS Bases to what is contained in the STS Bases. The existing BVPS TS Bases have a single discussion for TS 3.2.2 and TS 3.2.3.

By adopting the expanded STS Bases, it became necessary to provide a separate TS Bases for TS 3.2.2 and TS 3.2.3. The expanded TS Bases are provided for information only in Attachments B-1 and B-2.

The LRM contains the COLR, which will be modified to reflect the changes proposed in this LAR. These changes consist of revising COLR Specifications 3.2.1 and 3.2.2 to reflect the RAOC methodology. The other changes include revising COLR Figure 4.1-2 and deleting COLR Figure 4.1-4 to be consistent with the RAOC and FQ surveillance methodologies, and the addition of tables for the FQ(Z) penalty factor and the W(Z) values.

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Beaver Valley Power Station License Amendment Requests 310 (Unit 1) and 182 (Unit 2)

The COLR also contains various values and figures that are cycle specific.

These values and figures are appropriately identified and labeled. The specific operating values for AFD and FQ(Z) provided in the COLR will be evaluated as part of the reload process using the WCAP-9272-P-A (Reference 3) methodology and RAOC and FQ(Z) methodology analyses to verify these parameters. The COLR will be revised during the cycle-specific reload core design analysis.

Licensing Requirement 3.4, Axial Flux Difference (AFD) Monitor Alarm, and the Bases for Licensing Requirement 3.8, Leading Edge Flow Meter, are being modified in each unit's LRM. The change is made to maintain consistency with the change to TS 3.2.1. The power level referenced in Licensing Requirement Surveillance 3.4.1 and Licensing Requirement Bases B.3.8 is changed from 15% RTP to 50% RTP to be consistent with the Applicability of TS 3.2.1 and the Action. The statements pertaining to penalty minutes and target band are deleted from Licensing Requirement 3.4 and its Bases, because penalty minutes and target band are not applicable to RAOC. The requirement to monitor a restored channel for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is deleted because it relates to penalty minutes. All of these changes are acceptable because they are consistent with the proposed changes to TS 3.2.1.

As previously mentioned, the TS Bases and LRM are provided for information only and can be changed by the 10 CFR 50.59 process. As such no further discussion of the proposed TS Bases and LRM changes is necessary.

3.0 BACKGROUND

Axial power distribution control at BVPS is currently achieved by the Constant Axial Offset Control (CAOC) methodology. This methodology was developed and described in WCAP-8385 and WCAP-8403 (Reference4). This strategy assures peaking factors and departure from nucleate boiling (DNB) remain below the accident analysis limits. The CAOC methodology developed in Reference 4 does this by maintaining the axial power distribution within a band of +7%, -7% Al, for the BVPS units 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.

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Axial Flux Difference (AFD) is a measure of axial power distribution skewing to the top or bottom half of the core. It is very sensitive to core-related parameters such as control bank position, core power level, axial bumup, 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 influences the possible power shapes and could result in violations of the kw/ft limit during Condition II transients. Condition II transients, assumed to begin from within the AFD limits, are used to confirm the adequacy of Overpower AT and Overtemperature AT Trip Setpoints.

The CAOC methodology is presently incorporated into Technical Specification 3.2.1, Axial Flux Difference. The FXY methodology is presently incorporated into Technical Specification 3.2.2 Heat Flux Hot Channel Factor FQ(Z). Technical Specification 3.2.4, Quadrant Power Tilt Ratio (QPTR) refers to the FQ(Z) surveillance requirement. Application of the RAOC and FQ surveillance methodologies requires the alteration of these Technical Specifications. A change to Technical Specification 6.9.5, CORE OPERATING LIMITS REPORT (COLR), is also required to provide the methodology change. In order to provide consistency and to avoid duplicate requirements between the power distribution limits Technical Specifications and the reactor trip system instrumentation Technical Specification, Note 3 of Table 4.3-1 of the TS 3.3.1 also requires modification.

4.0 TECHNICAL ANALYSIS

The implementation of RAOC and FQ surveillance methodologies have been previously developed and approved by the NRC in WCAP-10216-P-A Rev.

1A (Reference 2). 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 Page 9

Beaver Valley Power Station License Amendment Requests 310 (Unit 1) and 182 (Unit 2) direct surveillance of the elevation-dependent heat flux hot channel factor and provides margin compared to Fxy surveillance.

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.

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'F.

The heat flux hot channel factor is measurable but will normally only be determined periodically as specified in the Technical Specifications 3.2.2 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 +12 steps from the group demand position.

(b) Control rod groups are sequenced with overlapping groups as described in Technical Specification 3.1.3.6, "CONTROL ROD INSERTION LIMITS".

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(c) The rod insertion limits of Specifications 3.1.3.5, "SHUTDOWN ROD INSERTION LIMIT" and 3.1.3.6 are maintained.

(d) The axial power distribution, expressed in terms of Axial Flux Difference, is maintained within the limits.

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), to provide assurance that the limit on the heat flux hot channel factor, FQ(Z), is met. The power factor, W(Z),

accounts for the effects of normal operation transients within the AFD band and is determined from expected power control maneuvers over several ranges of burnup 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

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 is not impacted by the change from CAOC to RAOC.

The f(AI) function is generated based on the expected axial power shapes from the various Condition I and II events. Because RAOC allows for more Page 11

Beaver Valley Power Station License Amendment Requests 310 (Unit 1) and 182 (Unit 2) severe power shapes to be generated, it was necessary to move the negative wing of the Overtemperature AT f(AI) penalty to eliminate shapes that may violate the DNB criteria. This will 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. It is concluded that the implementation of RAOC does not adversely affect the results of the non-LOCA analyses and the conclusions made in the UFSAR remain valid.

4.2 LOCA and LOCA-Related Evaluations The change from CAOC and FY surveillance to the RAOC and FQ(Z) surveillance methodologies has beeh evaluated for impact upon the existing LOCA safety analyses and in support of the planned extended power uprate (EPU) program. The LOCA and LOCA-related accident analysis remain valid for the methodology implementation given the above parameter changes and their effect on the safety analysis limits.

The RAOC and FQ(Z) surveillance methodologies do not affect the normal plant operating 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 impacts have been evaluated, and a peak cladding temperature (PCT) penalty of 16 'F has been established for the Unit No. 1 large break LOCA limiting time period. The limiting PCT time period for the Unit No.

2 large break LOCA was determined to not require a PCT penalty. Margin to the 2200'F PCT limit remains for both units.

No core design inputs to the EPU small break LOCA analyses changed due to RAOC operation. Thus, these analyses remain unaffected by RAOC. The small break LOCA analysis is not dependent on the specific axial power distributions associated with the change to RAOC.

4.3 Core Design Evaluation The change from CAOC and FXy surveillance to the RAOC and FQ(Z) surveillance methodologies have been evaluated for impact upon the BVPS core design. Consistent with the approved RAOC methodology, the Page 12

Beaver Valley Power Station License Amendment Requests 310 (Unit 1) and 182 (Unit 2)

Condition I axial power shapes were analyzed to demonstrate compliance with the 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 extended power uprate (EPU) analyses were made based on these evaluations. The negative wing of the Overtemperature AT Trip Setpoints f(AI) function from the EPU analyses was revised based on the limiting Condition II axial power distributions such that the DNB design criterion is met for accidents which are terminated by Overtemperature AT Reactor Trips.

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 rod internal pressure of the fuel rods and the cladding transient stress and transient strain. Westinghouse demonstrated that all fuel performance limits are capable of being met under RAOC operation. Compliance with the safety analysis assumptions will be performed on a cycle-specific basis during core design analysis.

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

4.4 Other Areas A review of the areas listed below has been performed for this evaluation, and it has been determined that they are unaffected by the RAOC and FQ(Z) surveillance methodologies changes.

(a) Emergency Operating Procedures (b) Instrumentation and Control Systems (c) Radiological Analyses (d) Mechanical and Fluid Systems (e) Plant Operability with respect to operating margin Page 13

Beaver Valley Power Station License Amendment Requests 310 (Unit 1) and 182 (Unit 2) 4.5 Conclusion The technical analysis demonstrates that the implementation of the RAOC and FQ(Z) surveillance methodologies does not affect the normal plant operating parameters, protection system actuation, the safeguard system actuation, or any other plant capability important to the mitigation of a Non-LOCA or LOCA accident.

5.0 REGULATORY SAFETY ANALYSIS This License Amendment Request (LAR) requests approval to implement the Relaxed Axial Offset Control (RAOC) and FQ surveillance methodologies for the two Beaver Valley Power Station (BVPS) units.

These methodologies are used to reduce operator action required to maintain conformance with power distribution control Technical Specifications and to increase the ability to return to power after a plant trip while still maintaining margin to safety limits under all operating conditions.

The Constant Axial Offset Control (CAOC) methodology is presently incorporated into BVPS Technical Specification 3.2.1, Axial Flux Difference. The Fo, methodology is presently incorporated into Technical Specification 3.2.2 Heat Flux Hot Channel Factor FQ(Z). Technical Specification 3.2.4, Quadrant Power Tilt Ratio (QPTR) refers to the FQ(Z) surveillance requirement. Application of the Relaxed Axial Offset Control (RAOC) and FQ surveillance methodologies requires the alteration of these Technical Specifications. Changes to Technical Specification 6.9.5, CORE OPERATING LIMITS REPORT (COLR), is also required to provide the methodology change. In order to provide consistency and to avoid duplicate requirements between the power distribution limits Technical Specifications and the reactor trip system instrumentation Technical Specification, Table 4.3-1 of TS 3.3.1 also requires modification. The proposed Technical Specification changes are consistent with NUREG-1431, "Standard Westinghouse Technical Specifications Westinghouse Plants," Revision 3.

5.1 No Significant Hazards Consideration FirstEnergy Nuclear Operating Company (FENOC) 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. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Page 14

Beaver Valley Power Station License Amendment Requests 310 (Unit 1) and 182 (Unit 2)

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 all 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 increase 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 configuration, 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. Does the proposed change 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 Page 15

Beaver Valley Power Station License Amendment Requests 310 (Unit 1) and 182 (Unit 2) 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 all 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. Does the proposed change 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. All of the appropriate acceptance criteria for the various analyses and evaluations will continue to be met.

The impact associated with the implementation of RAOC on peak cladding temperature (PCT) has been evaluated for the planned extended power uprate. This evaluation has determined that implementation of RAOC at the extended power uprate power level will not result in a significant reduction in a margin of safety for either unit.

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

Based on the above, FENOC concludes that the proposed amendments present 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.

Page 16

Beaver Valley Power Station License Amendment Requests 310 (Unit 1) and 182 (Unit 2) 5.2 Applicable Regulatory Requirements/Criteria A review of 10 CFR 50, Appendix A, General Design Criteria for Nuclear Power Plants (Reference 5) 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 The 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 2 events are the uncontrolled bank withdrawal, cooldown and boration/dilution accidents.

The most important (limiting) Condition 3 and 4 events are the loss of flow accident and LOCA, respectively. This is the case for both units.

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 Page 17

Beaver Valley Power Station License Amendment Requests 310 (Unit 1) and 182 (Unit 2) endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the 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

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22 (c)(9). Therefore, pursuant to 10 CFR 51.22 (b),

no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

7.0 REFERENCES

I. NUREG-1431, "Standard Westinghouse Technical Specifications Westinghouse Plants", Revision 3, June 2004.

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

3. WCAP-9272-P-A, "WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY," July 1985 (Westinghouse Proprietary).

4. WCAP-8385 (Proprietary) and WCAP-8403 (Non-proprietary),

"Topical Report Power Distribution Control And Load Follow Procedures", September 1974.

5. 10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants".

Page 18

Attachment A-1 Beaver Valley Power Station, Unit No. 1 Proposed Technical Specification Changes License Amendment Request No. 310 The following is a list of the affected pages:

Page 3/42-1 3/4 2-2 3/4 2-3 3/4 2-5 3/4 2-6 3/4 2-6a 3/4 2-8 3/4 2-9

3/4 2-10 3/4 3-11

3/4 3-12

3/43-12a*

3/4 3-13 6-18 6-19

No change made to this page. Included for information and readability only.

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE (AFD)

LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENG4r(AFD-inA-lnux-diffe-ence units shall be maintained within the t-arget-bandlimitsL'I specified in the CORE OrEr ATINC LIMITS REPORT (COLR}-.

APPLICABILITY: MODE 1 ABGVE-withTHERMALPOWER > 50* RATED-T1BERMA1l PGWERRTP*.

ACTION:

a. With thc indicated AXIAL FLUX DIFFBRENGC outsidc of thc t-arget-band-wthERMAL-.POWER--

1. .b-ve 9Ov ef RAT-ED T_ . P Wutes.

a) Either restere the indateFiwthin-the target band -limits, or b)-Reduee--THERMAL-POWER--to-4ess-than--90% ef RATED THERMAL-POWER--

2.- e-twen -- *nn-- -TAETR Tlll:RMWER-:

n - POWE:R noP1%MTonj mnav contnuc A1oide

1) The indicated AFD has not becn outside of the tage ban f... :.

deviatien -etunya iLduig-h rVitua 24-eurs, and

2) The indicated AFD is witli. _ee-al eperatien limits epecified in thc COLR.

Gt s--e-,eduee TIERMAL-POWE---t ehie - t-an

'0 o f RATED TIIERMAF-POWBR-witlin--4-mtnes and reduee the Pewer-Range Ncutron Flux! Iigh Trip Setpoints te !5 55% ef RATED TIERMIAL POWER within the-net-4-heur-S.

b) Surveillance testing of the Prwer Rangc Ncutren Specificatien 4.3.1.1.1 previded the indicated AFD is maintained within the limits. A total; of hor v: opraionl ayV be lll a-ex muae wit thea testing wijtheut penalty deviaticn.

ith D not withinliAT.POW to < of TP within 3mi nutes.

(1) TheA AD shaIIbe sqdered out limits two or more OPERABLE-excore-channels-indicateAFD-to-be-outsideil-imits-

~IJ2 See Special Test Exception 3.10.2.

BEAVER VALLEY - UNIT 1 3/4 2-1 Amendment No. 2-3-I

POWER DISTRIBUTION LIMITS IMIIN6-ONDI YON-FOR-ERATION{ (Coint ued)

b. TIIERL rGWERosha,4--net be 5nereased above 9O% ef -PTED 9IIERMA-L POWER wiless the indicated :AFD is within the target e- THERAir--POWER-hhall not--be inereased-above- So.-e--RATED THERMA-,eWER-un~ese-sthe-indieat-e aeFD-hs-not--been-out-i-de

_c-h-tare ban fe mer _ n hor_.-. rs.F ei e lative in- th pzv-i--D

-- 24 hurs-7.

SURVEILLANCE REQUIREMENTS 4.2.1.1 AerifvAFD-Thc indicated AXIAL FLUX DIFFERENCE eha1- be determined to be-within its-limits during POWER OPERATION abovc 15' Cf RATED THERMAy-POWER-for each OPERABLE excore channel at least once per 7 days.

hnsdCred outsidc of its target badwz 1 at least 2 Cf 4 er 2 Cf 2 OPRBEecree l r indieating tlc ArD to bc -utsidc the target band. POWER OPERATION e b s b a __ la . , 4n. b a-.----One-mfnut-e-pen by-.4deviation-fer--eah-one--iinute-of-PGWER OPERATION outside Cf thze-t-aget-band-a-t-1THE PAL--GWER 4ePel4- u1E-o or- P gvc A0°-o ^A -?Uv.LDOE. n PGWER OPERATION eutside - of th--ta tband at TIIERMAL POWER lv.s between 15% and 5 Cf RATED THERMAL POWER.

BEAVER VALLEY - UNIT 1 3/4 2-2 Amendment No. 22-5

£Next-page-i s_3 4 52z I

VUCWbtt: tiSelS:'tNLMll'

-SURV73-143NCS-REQIBREMENT- oentnued)-

4.2.1.3 The-

.. taet flux difference ef eae emeare ehannel Pewer Days- $ e esons of Specifieatien are

_4..4 net pl--ab-e.

4.2.1.4 Tbe aet -9cu-derenee shaI4--be-updat-ed-at lcast---onee per- 3-Effete ru I Pcwer ays-by either-det-ermin-g-t-he-ta~rget-f-iiu va~uei The-provi ions of Speeificatien 4.0.4 are not-pplieable.

EANVER VALLEY ;i _T a/4- . 13 A -imln me" d 9_

(Next pagev in m.'4 n._')

POWER DISTRIBUTION LIMITS 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR-F (Z)

LIMITING CONDITION FOR OPERATION 3.2.2 FQ(Z) as-apProximatby FQ(Z)andFV(Z), shall be iimited-by the-folloviw-ng-r-elatonship the ewithin heCQLR_ Rmi.s.s.ecfedin tEK(Z)J fr- P >} .5 r(Z) ' - [R(TZ)]

P rEPLIfITr --

-here, GFQ -The FQ limit at RATED-THERMAL powER p-vided

_ F=PIEt nOPERATING LIMTSl l EmfRr K(Z) - The nermalized FQ(Z) as a function of corc height--GOrevided--in--th CORE -OPErRATING LIMITS REPORT-ard

~P TED TIHERMAL POWER APPLICABILITY: Mode 1.

ACTION:

With F,(o) emeeedn-t b- 17E a.-Reduce THE ATPOWER.t L ea-1foreaeh 14FQ4F(Z) exeeeds t-he- limit-w hin-rinmtres-and-simiy-aray-r-educe-the- Power Range =ete _!= h Tri epo- W .-.. e-nx 4 heuros POwElR OPERATION may preeeed for up to a total of 72 heur-sa s-be-;n POE PRTO .aypees_~..

the-GeOpewer AT Trip Sctpeints havc becn -edueed at least 1°6for cach 1°6FQ Z)ecedth it.TeOepwr AT-

= 7 v ; ~ ~

C~~ v r s 1 -iL-s tl = ;

s e--tical .

ie ef the zut of limit eendition prier to incrcasing TIIERMAL POWER, THERMAL POWER a cAesd rvd F (Z) is demonstrated

a. With F6(Z)o ithin limit 1.eReduce__THERMAL POWER-Ž---.-%RTP for-eachl26 F6 (Z) exceeds the limilwithin i5ruminutes after each_F6(Z) determination :and

2. Re~duce the Poner P anqe__Neutron__Flux-Hiqhig Trip Setpo-ints 1 foreach_19_ F (Z)exceads-the-limit within-72-hourB-after-each.FQc(Z)_determination;_and

3. Reduce-the-Overpo-wer-ATTxripSetpoints? 1Ž for-each

.1%F (Z)_ eds the I within 72_hotrs after each F6 (Z) dtrination; and

4. PerformSurveillanceRequirements_4.22.2_2and._4._2.2.3 p rorto-incxeasinq THERMAL POWER abovy the limit-of Actiona..l..

S. ~~ bei ~ OD~ ~ n~ h _ _t_se 6 ho-urs.

h- With F6(Z) not within limits 2

1. _ReduceAFD Iimits >flt-for-each1lFt( Z)exceeds-limit wi thinA4hours;_and 2 repi e the Powe-r R;nqge Neutron Flux-High Trip se nts > t for Pch-1tat taelmaximuo 1wahle power_-of-theADimitsisJ reduced-within722ihours and

3. Reduce-theOverpowerAT-Trip-Setpoints->_l%-for-each 1k that..the L-aimu i1ob1 Pcwr nf tbAF l~iMiLS is reduced-within 72hours-;and A_4-Perf-ormSurv-e-llnce-Reruir-ements 4_2--2 2and-4A-2--=3.

prior-to increasing-HERMAL-OWER-abovejthe-maximum allowablePowe- ofitheAFD-limits-S. Otherwise he i n M hil Ifbour

-Ii Action&a.Ashall-be-complated-whenever-Action&aisBentered.

A(2)_Action-bA4s hal1 be-completed-whenever-Action-b-is-entered.

BEAVER VALLEY - UNIT 1 3/4 2-5 Amendment No. 1-54 l

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 Fy-shal' be-eva uae -t-d

-i-d - (4Z-4s-wit-hi4t -s-etSemine lmlt-bye-

a. Using the_ le irere etetor to obtai4n-a--pewer distribution-map at any THERMAL POwER greater than So ef sRATED sl1 r POWR-b.-iexsn e--e vr~ur-ed- Pxy c-omponet o tx --oe-r diest-ribution--map-- y 3 °a - - t--aecount f-er-manuf-aetur-ing tole-e~r-aneef--andi--futher Lee-a-sing-the--va~ie y--5 -t-

-aecount-fer-meastreemen-t-uneert-taot-is--

c. Cmeib-h-xycmued- _~-Ex-ebt ine n-33 -above t-;

appr pr-ate-measured--eore-planes--gien--in-e-and--f-below-,and 2 The nehaiohp~

L RTP

[lYr'YXlIi wherc L -i-Iy he-14mit--for--f--aeteonal--THERMAL--POWER

_ _ __ _ _ _ __ __ __ _ _ _ _ _ _R T P eperation expresed as a fnctien oef---F, , rFXY i3 Ph-Peer actr rF ,y-proevIdedtin the CGRE-GOPERATING LIMITS REPGR and-P - l-t-he-fract n-oS R~rEi EERMb-@WERat-w~eh-Fy-wa-9-mea ured-

d. ceareduingt -. Ae followir. schedule:

C RTP

-1.-- When- FX'-y ia greate -than-the- FW --H 3-{ertii aprorite measured corc plane but ecss t.han--the-T*;~

rclationship, additional power distribu£ sn-mape-ehale be-takenand- C -F7 -empa--ed-t FX--and RTP - L-a) Lither within 24 hour3 after exceeding by 20!k of RATED THERMAL PO-ER-- reater, t e- THRMA-L-PGWBR at -w14eh-- wass 1s tt dete-Mfned, or b) At lea3t onec per 31 ErrD, whhichever occurs firTt.

BEAVER VALLEY - UNIT 1 3/4 2-6 Amendment No. 1-4 I

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (continued)

. Wh n the y ixyli power--di-st A'bien--maps shaI4-be-tken--and--cxy-RTP L _ _ _ _ _ _ _ _ _

eempaeed-te-- -- and-j-a-eae- cC p r 31 EPPD.

f'ods-and-e-}-tenroddzd corc p1anes-in--t-he--CRE-iOPERATING LIMIT5 REPORT-f T---he--FT limits ef , above _are nt-ap ieab in te fellowing-Cf es a3-measu r-em-t-he-botlo-o&

the ful..

1. Lower corc region fr er4 te--1-S% nelive

2. U- e eere-regio fem 85 to 100°s, incluaivc.

ceasured from grid eenterline.

inehes) about the bank demand-I pesitien1_ f theC ban I'D" Cerntrel reds.

g. tC f cvaluated-to -detennine if "-(Z) is Wi~thinitsiemit.

4.2.2.3 -Iher--- FQ(Z) i measured pusuant-t>-Speeo w 4.10.2.2,

-enverai--m --- u----FB-)- -halbc -btained

-from-a-power distributien map and increaed by JI tC aeeetflt fer manufacturing toler-ces and F--rth1 i sed by 5e tC aeeen fe m,.asure...n_ tneertaintyr 4.be2verFi (fi to bewithin.-the lit cording lo theIollowing-schedule (

a-. Once-after-each refueling-prior-to-THERMAL-POWER-exceedinq 251lRTP,-and

b. Once-within-12-ho-ur-s-afte-chie-vin-e-uilibrium-conditions after-exceedinq, byŽ> 10_RTP.,_theTHERMAL-POWER-at-which FQ0 (Z)_ewaaLt verified, and c At-leastonc era1Ef c1i-euUP ower-Days therea ter

A-2-2.3 FQ(Z)_shal ye VP fi e td-t-be-within-the-limit (4) accordin to-the-fo1Uowinq-schedule2.L-

a. Once-aftereach refueling-prior-tosT HERMALPMNER-exceedinq J75_RTP-;and b- once within 12 houirs-after ach e~ggu~bim conditions

.after-exceedinq,-byv2 10i%-RTP-,_the_THERMALPOWER-at-which F6M(Z)=wasast verifilec-biad

-c_ At_1east-once-per 31 Effec-tive ull-Po~wer-Davs-, ther-eaft-e-r 9==rDAurg-o e--esal-athion at the-bJehqinning== ofec cycle- THERMAL achiee at _,I Pr _ api, st~ainedL (AA 1 fmeasurements-indicate that-the-maximum-ovexrz of_ [F& (Z) K&=)1 hansincx-easedsincethere oustemaluati:nonF6 (Z)

a. Increas e F6 (Z)=ky thegqreater-of-a-factor-of lD2_or-bvyan APoriate factor specified in t)hCJ andreve.-3fy FW(Z) iswti li its, or b-. Repeat-SuravteillanceuRequirement 4_2-2_3-onceper27 Ef Livej ull PoWer f:aySunti-No AAaabo-v-ie metor

.two-successiveflux-maps-indicate-that-the-maximum-over-z Of-F6( Z)/K(Z)l has-not-increased.

BEAVER VALLEY - UNIT 1 3/4 2-6a Amendment No. b54 l (Next page is 3/4 2-8)

POWER DISTRIBUTION LIMITS 3/4.2.3 NUCLEAR ENTHALPY RI-EAHOT CHANNEL FACTOR - FXH LIMITING CONDITION FOR OPERATION 3.2.3 FXH shall be limited by the following relationship:

FXH

CFADH [1 + PFAEH (1-P)] l where: CFADH = FRH limit at RATED THERMAL POWER provided in the CORE OPERATING LIMITS REPORT, PFADH = The Power Factor multiplier for FXH provided in the CORE OPERATING LIMITS REPORT, and THERMAL POWER

= RATED THERMAL POWER APPLICABILITY: MODE 1.

ACTION:

With FXH exceeding its limits:

a. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 hours and reduce the Power Range Neutron Flux-High Trip Setpoints to < 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b. Demonstrate thru in-core mapping that FXH is within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding the limit or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and

c. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER, subsequent POWER OPERATION may proceed provided that FXH is demonstrated through in-core mapping to be within its limit at a nominal 50% of RATED THERMAL POWER prior to exceeding this THERMAL power, at a nominal 75t of RATED THERMAL POWER prior to exceeding this THERMAL power and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after attaining 95% or greater RATED THERMAL POWER.

BEAVER VALLEY - UNIT 1 3/4 2-8 Amendment No. 1-54 l

I No change proposed. Included for information only.

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.3.1 FXH shall be determined to be within its limit by using moveable incore detectors to obtain a power distribution map:

a. Prior to operation above 75% of RATED THERMAL POWER after each fuel loading, and

b. At least once per 31 Effective Full Power Days.

4.2.3.2 The measured FRH of 4.2.3.1 above, shall be increased by 4t for measurement uncertainty.

BEAVER VALLEY - UNIT 1 3/4 2-9 Amendment No. 73

POWER DISTRIBUTION LIMITS 3/4.2.4 OUADRANT POWER TILT RATIO (OPTR)

LIMITING CONDITION FOR OPERATION 3.2.4 The QUADRANT POWER TILT RATIO shall be less than or equal to 1.02.

APPLICABILITY: MODE 1 greater than 50 percent of RATED THERMAL POWER. (1)

ACTION: With the QPTR not within the limit:

a. Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , reduce THERMAL POWER greater than or equal to 3 percent from RATED THERMAL POWER (RTP) for each 1 percent of QPTR greater than 1.00, and

b. Within 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, perform Surveillance Requirement 4.2.4 and reduce THERMAL POWER greater than or equal to 3 percent from RTP for each 1 percent of QPTR greater than 1.00, and

c. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and once per 7 days thereafter, perform Surveillance Requirements 4.2.2.2,_A_2.L3_, and 4.2.3.1, and

d. Prior to increasing THERMAL POWER above the limit of ACTION a or b above, re-evaluate the safety analyses and confirm the results remain valid for the duration of operation under this condition, and

e. After ACTION d above is completed and prior to increasing THERMAL POWER above the limit of ACTION a or b above, normalize the excore detectors to show a QPTR less than or equal to 1.02, and

f. After ACTION e above is completed and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after reaching RTP or within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after increasing THERMAL POWER above the limit of ACTION a or b above, perform Surveillance Requirements 4.2.2.2_,42-2 and 4.2.3.1.

g. Otherwise, reduce THERMAL POWER to less than or equal to 50 percent RTP within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

(1) See Special Test Exception 3.10.2.

BEAVER VALLEY - UNIT 1 3/4 2-10 Amendment No. 1-92 l

INo change proposed. Included for information only.

TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS Channel Modes in Which Channel Channel Functional Surveillance Functional Unit Check Calibration Test Required

1. Manual Reactor Trip N.A. N.A. S/U(1) N.A.

R (10)

2. Power Range, Neutron Flux

a. High Setpoint S D , M Q 1, 2 and Q( 6)

b. Low Setpoint S R (6) S/U"1) 2

3. Power Range, Neutron Flux, N.A. R(6) Q 1, 2 High Positive Rate

4. Power Range, Neutron Flux, N.A. R (6) Q 1, 2 High Negative Rate

5. Intermediate Range, S R (6) 1 f142) (14) 53(1 4) 4 Neutron Flux (15)

6. Source Range , Neutron Flux

a. With Rod Withdrawal S Q(8 ) 2 3 (14) 4(14)

Capability

b. With All Rods Fully S R 6) Q(8) 3, 4 and 5 Inserted and Without Rod Withdrawal Capability

7. Overtemperature AT S R (6) Q 1, 2

8. Overpower AT S R Q 1, 2

9. Pressurizer Pressure-Low S R Q 1, 2

10. Pressurizer Pressure-High S R Q 1, 2

11. Pressurizer Water S R Q 1, 2 Level-High BEAVER VALLEY - UNIT 1 3/4 3 -11 Amendment-No. 217

No change proposed. Included for information only.

TABLE 4.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS Channel Modes in Which Channel Channel Functional Surveillance Functional Un'it Check Calibration Test Required

12. Loss of Flow - Single Loop S R Q 1

13. Loss of Flow - Two Loops S R Q 1

14. Steam/Generat or Water S R Q 1, 2 Level-Low-Low

15. DELETED I

16. Undervoltage-Reactor Coolant N.A. R M 1 Pumps

17. Underfrequency-Reactor N.A. R M 1 Coolant Pumps

18. Turbine Trip

a. Auto Stop Oil Pressure N.A. N.A. 1, 2

b. Turbine Stop Valve N.A. N.A. S/(1) 1, 2 Closure

19. Safety Injection Input from N.A. N.A. R 1, 2 ESF

20. Reactor Coolant Pump Breaker N.A. N.A. R N.A.

Position Trip

21. Reactor Trip Breaker N.A. N.A. Ma 5,11) 1, 2, 3X14) and S/U() 4(1 4) 5(1 4)

BEAVER VALLEY - UNIT 1 3 /4 3 -12 Amendment No. 240

No change proposed. Included for information only.

TABLE 4.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REOUIREMENTS Channel Modes in Which Channel Channel Functional Surveillance Functional Unit Check Calibration Test Recruired

22. Automatic Trip Logic N.A. N.A. M (5) 1 3 3(14) t143 ' (14) '

23. Reactor Trip System Interlocks A. P-6 N.A. R(6) R 1, 2 B. P-8 N.A. R (6) R 1 C. p-9 N.A. R (6) R 1 D. P-10 N.A. R (6) R 1 E. P-13 N.A. R R 1

24. Reactor Trip Bypass N.A. N.A. M(12)

Breakers SxU( (1)

BEAVER VALLEY - UNIT 1 3/4 3-12a Amendment No. 215

TABLE 4.3-1 (Continued)

NOTATION (1) - If not performed in previous 31 days.

(2) - Heat balance only, above 15 percent of RATED THERMAL POWER.

(3) - At least once every 31 Effective Full Power Days (EFPD) compare incore to excore axial imbalance above 1550 percent of RATED THERMAL POWER. Recalibrate if absolute difference greater than or equal to 3 percent.

(4) - (Not Used)

(5) - Each train tested every other month.

(6) - Neutron detectors may be excluded from CHANNEL CALIBRATION.

(7) - Below P-10.

(8) - Below P-6, not required to be performed for source range instrumentation prior to entering MODE 3 from MODE 2 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entry into MODE 3.

(9) - (Not Used)

(10) - The CHANNEL FUNCTIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip circuits for the Manual Reactor Trip Function. The test shall also verify the OPERABILITY of the Bypass Breaker trip circuit(s).

(11) - The CHANNEL FUNCTIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip attachments of the Reactor Trip Breakers.

(12) - Local manual shunt trip prior to placing breaker in service.

(13) - Automatic undervoltage trip.

(14) - With the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal.

(15) - Surveillance Requirements need not be performed on alternate detectors until connected and required for OPERABILITY.

BEAVER VALLEY - UNIT 1 3/4 3-13 Amendment No. 2 1

ADMINISTRATIVE CONTROLS 6.9.3 ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT

- - - - - - - - - - - - - - - NOTE - - - - - - - - - - - - - - - - -

A single submittal may be made for a multi-unit station. The submittal should combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

The Annual Radioactive Effluent Release Report covering the operation of the unit during the previous year shall be submitted prior to May 1 of each year in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program (PCP) and in conformance with 10 CFR 50.36a and 10 CFR Part 50, Appendix I Section IV.B.1.

6.9.4 MONTHLY OPERATING REPORT Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the pressurizer power operated relief valves or pressurizer safety valves, shall be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report.

6.9.5 CORE OPERATING LIMITS REPORT (COLR)

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

2.1.1 Reactor Core Safety Limits 3.1.3.5 Shutdown Rod Insertion Limits 3.1.3.6 Control Rod Insertion Limits 3.2.1 Axial Flux Difference-Gonsetant-Relaxed Axial Offset Control 3.2.2 Heat Flux Hot Channel Factor-FQ(Z) 3.2.3 Nuclear Enthalpy Rise Hot Channel Factor-FIAH

.3.2.5 DNB Parameters 3.3.1.1 Reactor Trip System Instrumentation -

Overtemperature and Overpower AT Setpoint Parameter Values BEAVER VALLEY - UNIT 1 6-18 Amendment No. 2-5B l

Change proposed in LAR 318 provided for information.

AfMINI STRATIVE CONTROLS COME OPERATING LIMITS REPORT (Continued)

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:

WCAP-9272-P-A, "WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY," July 1985 (Westinghouse Proprietary).

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

WGAP 102CC P A Rev. 2/WGAP 11524 NP A Rev. 2, 'T-h 1981 Version of the Westinghouse EGGS Evaluation Model Using the, 17 A "Power Shape Senaitivity Studies" 12/7 and Addendumn 2 A "BASH Methedelegy imprevements and Reliabilit, Enhaneements" 5/88. WCAP_12945-P-A, Volume 1 (Revision 2) and Volumes 2 through 5 (Revision 1), "Code Qualification Document for Best Estimate LOCA Analysis," March 1998 (Westinghouse Proprietary).

WCAP 8385, "POWER DISTRIBUTION CONTROL SUD LOAD-FOLLOWrNGv PROCEDURES-TOPICAL REPORT." Sept-ember-19 4-West-inghouse Proprietary).-

Safety-Analysis Aspects of a. mpreved Lead Follow Paekage--

-NEGC 0800, -St-andard Rview Pla, U.S. Nue1ear--egu1at-ery oemmseien_-4eet.3en4 , Nucleare---Design, July- 1981. -aBrneh Teehnica4 Eesitio PB 4.-1, WestinghusC oGenstant Axial WCAR-l021f6-P-A ,ReyisionAlA,il"Relaxation-of-ConstantAxial Offset-Control-F xSurvei-llance _Technical-Specification.I FebruarivyilL99A__

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

As described in reference documents listed above, when *an initial assumed power level of 102% of rated thermal power is specified in a previously approved method, 100.6% of rated thermal power may be used when input for reactor thermal power measurement of feedwater flow is by the leading edge flow meter- (LEFM).

Caldon, Inc. Engineering Report-80P, "Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFMP System," Revision 0, March 1997.

BEAVER VALLEY - UNIT 1 6-19 Amendment No. 4-5-G I

Attachment A-2 Beaver Valley Power Station, Unit No. 2 Proposed Technical Specification Changes License Amendment Request No. 182 The following is a list of the affected pages:

No change made to this page. Included for information and readability only.

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE (AFD)

LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENGE-(AFD+ influxdifference unitLashall be maintained within the t-arget-bandlimits(2) specified in the CORE OPERATING LIMITS REPORT (COLR)-.

APPLICABILITY: MODE 1 abeve-Xth XRMAILPQNER_5Os Pcrccn YED IIEr1RMAL rOWERRTP-*-U.

ACTION:

.... oI , _, _: _s

_ I tar-get--band-and-wit-h THERMA--POWER-a) dithcr reatoez the indieat- -AFD te within the taret band-RAT-ED Th1ERMAL PGWER-

'^~~~~- T7\ norr n--., rn-n 2-. Bet-ween-50O-pereent-and-0-g-pe-r-eent c ruT.D TrUErPML POWER:?

a) POWCR OPERATIOPN-may-on t-inue-preov-ided-

1) The indizated AF'D has net been outside ef the arge ..ur penalty deviation-eiini1-at-ivc-daring thz przvus 24-h urs,-and

2) The indicated ArD Is wit -D. e eptable epraien -,4,.~TTnlil Th GOL_

TT. _

Gther wis-. r-euee THEMA POWE to les tha ..

be perceznt et RPA'SD TH ubt ln aO- muetee an- redee_ the Pei cr-Rangc Ncutron -Fux-&&-g. Trip tpcints

_ to

55 rpecent ef--RAT-ED-T-ERMH I -pGWER-wit-hkn-the next-4-heurs--

6urveillanee t b) ~i -e{- Pewzr Range utren

_p_ ifiatien 4.3._.1.1 prevlded the--4ndicate AFD i maintained within th limit. Af "l - '" _ - - - -_ _- A - _- - - - - - . *'I-l - .4 *. *-% 4- -

s 4-he tal o u r s 3. vprav Iy eg lU- t AFD eutihdez of th targc t band during this t-e-t-ng-witheut-penalt-ydeveiat-on.-

Wit F ti iiitreiipTPMLPOE o<b(ko ~

- TheAD sall hsi imi he OPERATBLEexAorep chan indicat AFD to_he otsii eimits

-- 21 See Special Test Exception 3.10.2.

BEAVER VALLEY - UNIT 2 3/4 2-1 Amendment No. 1-11 I

POWER DISTRIBUTION LIMITS LIMITING NDIT-- -FOR-G ATION (Cent-inued-)

AONr --tT I _ed u

SAMTI..1.v. % -~s tin e THERMAL W shal n eabove J9 prnt ef TE~~tf~~wT>~ noT UT ,,1 &rnI7 RAE q41ERMAL PGWE s un - _- -n.di e AF iss within the-,

t-arget-band-and ACTION a. 2. a) 1), above-has-been-stisfied-

c. TIIERMAL O sh notbe- 4eresc a----e 50 porent ef RATrD THERMAL POWER unless t vets tare-ti-tbe-t--ban for-mr .. an e Aurp eenaltyJ devi-aion-cmulatic durne-b4 aevie -hX eve SURVEILLANCE REQUIREMENTS 4.2.1.1 Verify AFD The-indicaed AXIA--L FLUX DIFFERCNCE- haa-1--be det-ermhend-t e-be-within -t- limits dueng-POWER-GPATJGN-abeve-15 pereent efRATED THERMAL PGWER for each OPERABLE excore channel at least once per 7 days.

4.2.1.2 The indizatedx A- shall b e~ ide~ared eu1t1Ze ef its target-banS-when at lca3t 2 of 4 or 2 of 3 OPERABLE cxcorc channels are indicatingth to be utid thctarget band. PG*WER OPERATION out-side-of-t-he-t-arget-band-sha-e-be aeeumulat-ed-on-a-t-ime-basi-s-ef--

a. One-mfneute-penaaltty-- devev at n-f-r--eaeh----minut e- f--PGWER OPERATION eutside of the tar et-ban4--at--THERMAL -POWER

-levels equal to or abbeve 5n pereent of RxATED T4BRnMAL POWER, a _vd I ___ - I _ _1 .

t) . vllre tC penalty iCait+/- teWr cazn +/- tnuue ok rOWER OPERATION eutside of thc t-ergt band at IERMKAL POWER evels bet-we ePOWERERW -

BEAVER VALLEY - UNIT 2 3/4 2-2 Amendment No. 142 (Next- aqe is 3/4 2-4) I

POWER DISTRIBUTION LIMITC SURBEEC-E-REQUIREME*ES-46bntl uedX4 4.2.1.3 The target flux differenee fec OPEERBE ~exse-cbnee shall be determined b pr 92 Effct Full Power Days. -c Lr-v-iens of ZPceificatien 4.0.4 are not epp4ieable.

4.2.1.4 The-target flux differenee-shal4-be--updat-e-at---least-enee per 31 Effecetivc Full Power Days by eit-her-detesmwnrng-the target flux differnce pursuan-t -e42.1.3 boe the cyele life. The -

pr-vi-ion- f S eeifeatie 4.0.4 are net applicable-.-

BEAVE~R VAL3LEY 4N-TR; T- 2

POWER DISTRIBUTION LIMITS HEAT FLUX HOT CHANNEL FACTOR-FQ+/-ZI LIMITING CONDITION FOR OPERATION 3.2.2 FQ (Z),as-approximatedbyvF6(Z)andF (Z). shall be 4nmted by thc following relatinhpsin the

-po~a-y- [(Z K I for r

0.5 wherc. CFQ - Thc rQ limit at RATED ThERMAL POWER-previded in the GGRB -OPERATING LIMITS REPORT, K(Z) - Th- a-lized F -as -a- et-en ef eer height previded--in-the GORE OPERATING LIMITS REPORT, and THERMAL POERv-RATED THERMAL POWER-APPLICABILITY: MODE 1 ACTION:

Wit ,Z) exeeedie. its limit.

a.---- Rcduee--THERMAL}R at les t-as-pe eent--ercaeh-1-peareent-sQZ xee t-h limit witn- -mntes-ad slmila-rry Phwer Range Neut0- F4ux -High Trip Setpoints within. eh next 4- houra, POWER OPERATION may preeeed for up te a total ef 72 hours, sscqunt POWER OPERATION may

_t __~~~~~~~- Iv rmr 4 oott., vrF_.

pree p ;av*§FrevidedT th E -p Ve et *-ave beeX redueed at least I pereent for cach 1 perent F,(Z) emeeeds be-pernfarmed-w-itL te- rccto-suberitical.

b----i-dent-i-fl-and-c r r e et--t. cause--of-the eut- >--imt rer e inresi UA PWER: H may then be incrcaacd previd-ed-s() i de enstated-through incorc-mapping to bc within--i4mft.

a. WithF6(Z) not within 1imitl 1 ):

1. Reduce__THERMAL__OWERŽ-1 2 RTP for-each -l-6 F6 (Z) exceeds-thep imi-twithin =-5-minutes-aftex-each-FQc(Z) determination;_and

2. Reduce the Power Range Neutron F} ux-Hig rip Setpoints >1 -for achI% FC( Z)-expds the limit Within 72_hursafteraeICh_F ((Z) determination;-and tpower AT- Setpoint > 1% for each 1% F6(Z)_exceeds-the-2.imit-within-72-hours-after-each

=

F6 (Z)-determination;_and A___Perf orS-urv~eilancteqecuir-ement-sA-2-2-2_and-4-2-2-3 prior to increaMsAi POWER ove the limit-of Action a.l.

5. OtherwiseJbe-in-MODE-2_within-the-ofllowing-6-hour-.

b. -With-FQw(Z)_no i-thin-limits2

1. Rpdied M limits > 1_ for each 1% FW(Z) Pxceeds limit w-ithin-4houran

2. Reduce-theP-ower _Ranqe__Neutron FluxiHiqh _Trip Setpoints > 1t foxreach- 1gthat-the-maximum-allowable power-of-the-AFFD.imi-ts-is-reduced-withinl72_hours; and

3. Re~duic' the Overpower AT Tri Setpoints > 1% for each 13Athat-the-maximum-allowableipower-of-the-AJ\DF imits is-reduced-withinl72_hours;_and 4_ _ePerformSuryveillance-Reruirementsa4.A2-2.2and.42_2_3.

prior to increasfn THERMAL _________the-maximum O.t erwis e be in MQDE_2 Within-the -fnLowingiahours-.

AlL -Action-a-Ashall-be-complet-ed-whenever-Action-a-is-entered.

(2) Actionbh.4_shall be completed whenever Action b is entered BEAVER VALLEY - UNIT 2 3/4 2-4 Amendment No. l

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 bey-shad evaluated to dete-rmi-ne-4i-f-FQfZ -)--S-W.tLhi--

limit-by

a. Using tle-mevdable incore -detctors to obtain---a-- pewer distiui at ay nRAL PWER grfeater>- than 5 perent ef PATED THmERMoAL POWER--

distribut-i-e-map by 3 pereent to a cut-e--nufacturing t bleraneeo and further inereasi

b. i erpeasing at- measur-ed--F -- e, ete e

______ -t-he-F

ftT- (Fxy_____in_____

1. The F. limits fe -ED THERMAL PGWERn4F4IyRTI ) feE the appr-priatc measured-eorc planes givenFin-ye--i-n -

belew, and bFy) t Ft-SPa-too 3 L D wPihoy orooPr L_ I Vepe i~s exrssed asX aV fuete xf yRT- PX I7S -

the pewer faeter multiplier o x rvie-y-h GORE OPERATING LIMITS RE WORT, and r i3s the fraction of RAT-ED TI1RA OE twie-x-a-esrd

2. he elhati-spOWRahe-~

d.- Re ea eur4ng-4yaccrding to--thc following sehedule:

1-.--When--x C is--grea-er--t-han--t-he--F. RTP mi fer-t-he aproratc measured corc plaeebut leesstta~nleQ@-F Li be-t~enand-x e-eMpa-r-ed-t-eF,,y -and-Fxy a) Hit-her- it~-2F-}ne- e xed~-

2O pereent of RUATED TIIERMAL. POWER or greate~rthe

!RIERMAL PGW*Eowa _hiieh-F.,e was !at- d-eter-mined, b) At least enee per 31 EFPD, whichever eeeurs BEAVER VALLEY - UNIT 2 3/4 2-5 Amendment No.

POWER DISTRIBUcTION LIMIT_

GURVEItLANGE- REQUIREMNENtS-f ned4-

2. When th I is less than or equal te te-Fy-limit fe- the -aprhpretmeaeurccpe addie

- -_ \,__.. -,. - -- I- -I-- - x

- e IY eempa-red-te-F y 8-fnd-FyI-at-ast-on e-per-+/--E FPD c! The rxy-4-iid t -l ed-The-rmal-- ewer (FR-xy ) -hal-- be p !ro!4xi A e A for all:

-eeRa-

ae

_ I _- _ _ I L corc

..---_.- - pleanec cont-ain-ing ba f "D" e ontrol coe I ein te GRE: OPERATINGC LIMIT_ REPORT.

f. The F -it of c, a boc arcnt pieb-i_ e

-llowi-ng-eere-plane-sege nfl-as easur-el-in-per-eent--of-eore hegtEthebebefttem fthe fuel 1-.-Lower corc r ogin from O- te 15 pereent, i-neusive.

2. pe er rgen rmc et 0 en nl-------

3. Gr4--pla-f!em -i-ere t-er4e4e-ght (+--2.88---inehes+

mea-sur grweid esntterline-.

1. Corc plane rgir with + 'percent of- corce

(+ 2.88 inehes) -Latiet -tiaban kriIo Ibh9 3f han~r 'In" eotrlrods.

be With-F., -eeeeednvy 4Sll beeaiaedt-et-erm-inci-f-04S-2"s -w-it-kim4t--4-in-t-.

4 2 2.3---When-FQ4Ze) -- measu ed-pursuant-to-Speeei-f-eat-io 4..10.2.2, an everala-meaeuRed-3 bc ebtained from-e-pewer owahall distribution aunhnp ent tro anufanturing 4_2v2e 2 Ft(Z) _ to b_ wt.__ 1nmit acoi CT-to

.the-followingqschedule2 (3.

a- Once-after-each-refue~ing-prior-to-THERMAL-POWER-exceeding 355RIZTPjand Onne witbin 12 hr-after achievting qu;librium pcnditionS after exceeding.qby. n PTp, theTHEPmAL POWE at which FQ(Z) -asilastve rifed-and

,c-. At-least-once per-31-EffectiveFullPower-Days-thereafter.

.42.2.3 F6(Z)-shall.-be-verified-to-be-within-the-limit(4) accordinq to-the-followings chedule (3)

a. Onc-aefter-each-refuelinqnrtior-to-THERMAL-POWER-exceeding 75l-RTP;-and b- Once within 12 r fter hienqgeuibri onditions after-exceedinq. bYŽ2 10%_RTP._theIHERMAL-POWER-at-which F6(Z)waslaast verified? and c<___At least-once-per-31-Ef fetive-Full-Power D _thereafter (3) puringpower e-ca ation at the beginning.on c1hcycle, THRPmAT POW incrrea hpmayhe _a__ ha_________

achieved,_at whicha power distribution w A4A If measurements-indicate-that-the-maximum-ov~er z-ofj F(c(Z)L&LZJ.

haslincxeased~since the sriusevaluationnF (Z)=

w=

a lccreasQF (Z) by the qratpr of afactor of 1-02 or hy an anropriate fnctor specifiedlion- the rQTlE and rpvpri f Fw(

is within-limit o b.___Repeat-Surveillanc.R Reruirement 4-2.223-onceper7 EffectimeElull PomerDays until-Note__fAi)a aboveismet-or two-successsive lux-maPs-indicate-that-the-maximum-over-z

-ofl FQ(Z)l-K-(Z )lhas-not-increased..

BEAVER VALLEY - UNIT 2 3/4 2-6 Amendment No. l

POWER DISTRIBUTION LIMITS NUCLEAR ENTHALPY RISE-HOT CHANNEL FACTOR - FNH LIMITING CONDITION FOR OPERATION 3.2.3 FRH shall be limited by the following relationship:

FRH

CFADH [1 + PFADH (I-P)] l where: CFADH = The FNH limit at RATED THERMAL POWER provided in the CORE OPERATING LIMITS REPORT, PFADH = The Power Factor multiplier for FNH provided in the CORE OPERATING LIMITS REPORT, and THERMAL POWER P =

RATED THERMAL POWER APPLICABILITY: MODE 1 ACTION:

With FNH exceeding its limit:

a. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 hours and reduce the Power Range Neutron Plux-High Trip Setpoints to < 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b. Demonstrate through in-core mapping that FRH is within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding the limit or reduce THERMAL POWER to less than 5 percent of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and

c. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER, subsequent POWER OPERATION may proceed provided that FN is demonstrated through in-core mapping to be within its limit at a nominal 50 percent of RATED THERMAL POWER prior to exceeding this THERMAL power, at a nominal 75 percent of RATED THERMAL POWER prior to exceeding this THERMAL power and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after attaining 95 percent or greater RATED THERMAL POWER.

BEAVER VALLEY - UNIT 2 3/4 2-7 Amendment No. hi, 1

No change proposed. Included for information only.

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.3.1 FRH shall be determined to be within its limit by using movable incore detectors to obtain a power distribution map:

a. Prior to operation above 75 percent of RATED THERMAL POWER after each fuel loading, and

b. At least once per 31 Effective Full Power Days.

4.2.3.2 The measured FNH of 4.2.3.1 above, shall be increased by 4% for measurement uncertainty.

BEAVER VALLEY - UNIT 2 3/4 2-8 Amendment No. 31

POWER DISTRIBUTION LIMITS QUADRANT POWER TILT RATIO (OPTR)

LIMITING CONDITION FOR OPERATION 3.2.4 The QUADRANT POWER TILT RATIO shall be less than or equal to 1.02.

APPLICABILITY: MODE 1 greater than 50 percent of RATED THERMAL POWER. (1)

ACTION: With the QPTR not within the limit:

a. Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , reduce THERMAL POWER greater than or equal to 3 percent from RATED THERMAL POWER (RTP) for each 1 percent of QPTR greater than 1.00, and

b. Within 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, perform Surveillance Requirement 4.2.4 and reduce THERMAL POWER greater than or equal to 3 percent from RTP for each 1 percent of QPTR greater than 1.00, and

c. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and once per 7 days thereafter, perform Surveillance Requirements 4.2.2.244_2_2_3, and 4.2.3.1, and

d. Prior to increasing THERMAL POWER above the limit of ACTION a or b above, re-evaluate the safety analyses and confirm the results remain valid for the duration of operation under this condition, and

e. After ACTION d above is completed and prior-,to increasing THERMAL POWER above the limit of ACTION a or b above, normalize the excore detectors to show a QPTR less than or equal to 1.02, and

f. After ACTION e above is completed and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after reaching RTP or within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after increasing THERMAL POWER above the limit of ACTION a or b above, perform Surveillance Requirements 4.2.2.2A 4.2 2 3 and 4.2.3.1.

g. Otherwise, reduce THERMAL POWER to less than or equal to 50 percent RTP within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

(1) See Special Test Exception 3.10.2.

BEAVER VALLEY - UNIT 2 3/4 2-9 Amendment No. As Ocrreoted by Letter dated Geteber 19, 1995

No change proposed. Included for information only.

TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS Channel Modes in Which Channel Channel Functional Surveillance Functional Unit Check Calibration Test Required

1. Manual Reactor Trip N.A. N.A. S (/U (l ) 1 2 3(14),

4 t143 (14)

2. Power Range, Neutron Flux

a. High Setpoint S D , N3 Q 1, 2 and Q

b. Low Setpoint S R (6) 1(7) 2

3. Power Range, Neutron Flux, N.A. R (6) Q 1, 2 High Positive Rate

4. Power Range, Neutron Flux, N.A. R (6) Q 1, 2 High Negative Rate

5. Intermediate Range, Neutron S R (6) 41 3 (14)

Flux

6. Source Range(1 5 ), Neutron Flux

a. With Rod Withdrawal S R(6) Q(8 ) 2, 3 (14) 4 (14)

Capability and 5 (145

b. With All Rods Inserted S R (6 Q( 8 ) 3, 4 and 5 and Without Rod Withdrawal Capability

7. Overtemperature AT S R(6) Q 1, 2

8. Overpower AT S R Q 1., 2

9. Pressurizer Pressure-Low S R Q 1, 2 (Above P-7)

10. Pressurizer Pressure-High S R Q 1, 2

11. Pressurizer Water Level-High S R Q 1, 2 (Above P-7)

BEAVER VALLEY - UNIT 2 3/4 3 -10 Amendment No. 94

No change proposed. Included for information only.

TABLE 4.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS Channel Modes in Which Channel Channel Functional Surveillance Functional Unit Check Calibration Test Required

12. Loss of Flow - Single Loop S R Q 1 (Above P-8)

13. Loss of Flow - Two Loop S R Q 1 (Above P-7 and Below P-8)

14. Steam/Generator Water Level- S R Q 1, 2 Low-Low

15. DELETED.

16. Undervoltage-Reactor Coolant N.A. R M 1 Pumps (Above P-7)

17. Underfrequency-Reactor N.A. R M 1 Coolant Pumps (Above P-7)

18. Turbine Trip (Above P-9)

A. Emergency Trip Header N.A. R S/U( 1 ) 1, 2 Low Pressure B. Turbine Stop Valve N.A. R S/U<') 1, 2 Closure

19. Safety Injection Input from N.A. N.A. R 1, 2 ESF

20. Reactor Coolant Pump Breaker N.A. N.A. R N.A.

Position Trip (Above P-7)

21. Reactor Trip Breaker N.A. N.A. M( 5 11) 1, 2, 3(14),

1 and S/U( 4( ), 5 BEAVER VALLEY - UNIT 2 3/4 3-11 Amendment No. 61

No change proposed. Included for information only.

TABLE 4.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS Channel Modes in Which Channel Channel Functional Surveillance Functional Unit Check Calibration Test Required

22. Automatic Trip Logic N.A. N.A. M(5) 1, 2, 3(14),

4(14), 5(1 4)

23. Reactor Trip System Interlocks A. Intermediate Range N.A. R( 6 ) R 1, 2 Neutron Flux, P-6 B. Power Range N.A. R (6) R 1 Neutron Flux, P-8 C. Power Range N.A. R(6) R 1 Neutron Flux, P-9 D. Power Range N.A. R( 6 ) R 1, 2 Neutron Flux, P-10 E. Turbine First Stage N.A. R R 1 Pressure, P-13

24. Reactor Trip Bypass Breakers N.A. N.A. 1, 2, 34 4 (14), 5(14)

BEAVER VALLEY - UNIT 2 3/4 3-12 Amendment No. 132

TABLE 4.3-1 (Continued)

TABLE NOTATION (1) - If not performed in previous 31 days.

(2) - Heat balance only, above 15 percent of RATED THERMAL POWER.

(3) - At least once every 31 Effective Full Power Days (EFPD) compare incore to excore axial imbalance above 1-550 percent of RATED THERMAL POWER. Recalibrate if absolute difference greater than or equal to 3 percent.

(4) - (Not Used).

(5) - Each train tested every other month on a STAGGERED TEST BASIS.

(6) - Neutron detectors may be excluded from CHANNEL CALIBRATION.

(7) - Below P-10.

(8) - Below P-6, not required to be performed for source range instrumentation prior to entering MODE 3 from MODE 2 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entry into MODE 3.

(9) - (Not Used)

(10) - The CHANNEL FUNCTIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip circuits for the Manual Reactor Trip Function. The test shall also verify the OPERABILITY of the Bypass Breaker trip circuit(s).

(11) - The CHANNEL FUNCTIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip attachments of the Reactor Trip Breakers.

(12) - Local manual shunt trip prior to placing breaker in service.

(13) - Automatic undervoltage trip.

(14) - With the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal.

(15) - Surveillance Requirements need not be performed on alternate detectors until connected and required for OPERABILITY.

BEAVER VALLEY - UNIT 2 3/4 3-13 Amendment No. 94 l

ADMINISTRATIVE CONTROLS REPORTING REQUIREMENTS (Continued)

The Annual Radioactive Effluent Release Report covering the operation of the unit during the previous year shall be submitted prior to May 1 of each year in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program (PCP) and in conformance with 10 CFR 50.36a and 10 CFR Part 50, Appendix I Section IV.B.1.

6.9.4 MONTHLY OPERATING REPORT Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the pressurizer power operated relief valves or pressurizer safety valves, shall be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report.

6.9.5 CORE OPERATING LIMITS REPORT (COLR)

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

2.1.1 Reactor Core Safety Limits 3.1.3.5 Shutdown Rod Insertion Limits 3.1.3.6 Control Rod Insertion Limits 3.2.1 Axial Flux Difference-CenstantRelaxed Axial Offset Control 3.2.2 Heat Flux Hot Channel Factor-F,(Z) 3.2.3 Nuclear Enthalpy Rise Hot Channel Factor-FPAH 3.2.5 DNB Parameter 3.3.1.1 Reactor Trip System Instrumentation -

Overtemperature and Overpower AT setpoint parameter values

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:

WCAP-9272-P-A, "WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY," July 1985 (Westinghouse Proprietary).

BEAVER VALLEY - UNIT 2 6-19 Amendment No. 130 l

Change proposed in LAR 191 provided for information.

AD INISTRATIVE CONTROLS REFORTING REOUIREMENTS (Continued)

WCAP-8745-P-A, "Design Bases for the Thermal Overtemperature AT and Thermal Overpower AT Trip Functions," September 1986.

WCAP 10266 P A Rev. 2/WCAP 11524 SiP A Rev. 2, "Thc 1981 Versien ef the Westingheu e ECGS Evaluatien Med_ l Using the BASHI Ccde, Kabadi, J. N.,- March 1987; including Addendum 1 A "Power Shape Sensitivity Studies" 12/87 and Addendum 2 A "BASH Mcthedelegy Improvements and Reliability EIhatnemeznts" s5P/0. WCAP_12945-P-A, Volume 1 (Revision 2) and Volumes 2 through 5 (Revision 1), "Code Qualification Document for Best Estimate LOCA Analysis," March 1998 (Westinghouse Proprietary).

y.,p.A1 o"or I.tnrTD r-M'Y'nTT 'I"T('ThffT TT1wxl"

. . us -c v _ v, X v, ag x- - - -x - - - -1. _- 1xv- Avi A, PROCEDURE- TOPICAL REPORT. " SCpt-eb- - (WJel-ngheues Prorietary) .

-T.--W--T.M. -- ern- n-t-- K.

e iefpre! -f Core Per-formanee Brane-NRA0--an e ry H,- - 9---cA-t-t-aLehment-.p Lperaat-i-on-and Safety Analysis Aspecs Wf -a - - Loa

- - - Fellow Paekaqe-.

NvRE 0800, Standard Review Plan, U.S. Nuelear Regulater Czmmissien, Zcectien 4.2, Nuelear De 1-g~n, July 1981. Braneh-Teehnial Pes4itien CPB 4.3 1, West-inghcuse Constant Axial Gffset Centrol (GAOC), Rev. 2, July 1981.

WflAP-10--P-A, Rev io 1 "~ t!ion a of Constant-Axial Offeet gontro1-F0_ SU3vj11larice Technical necificatin February. 99A4 WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report," April 1995 (Westinghouse Proprietary).

As described in reference documents listed above, when an initial assumed power level of 102% of rated thermal power is specified in a previously approved method, 100.6% of rated thermal power may be used when input for reactor thermal power measurement of feedwater flow is by the leading edge flow meter (LEFM).

Caldon, Inc. Engineering Report-80P, "Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFM4- System," Revision 0, March 1997.

Caldon, Inc. Engineering Report-160P, "Supplement to Topical Report ER-80P: Basis for a Power Uprate With the LEFM4P"System," Revision 0, May 2000.

BEAVER VALLEY - UNIT 2 6-20 Amendment No. G I

Attachment B-1 Beaver Valley Power Station, Unit No. 1 Proposed Technical Specification Bases Changes License Amendment Request No. 310 The following is a list of the affected pages:

Page B-I B-VI B 3/4 2-1 B 3/4 2-2 B 3/4 2-3 B 3/4 2-4 B 3/4 2-5 B 3/4 3-Vj

I Providedfor IWformation Only. l TECHNICAL SPECIFICATION BASES INDEX BASES SECTION PAGE 2.1SAFETY LIMITS 2.1.1 Reactor Core .................................... B 2-1 2.1.2 Reactor Coolant System Pressure ................. B 2-2 3/4.0 APPLICABILITY.................................. B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL......................... B 3/4 1-1 3/4.1.2 BORATION SYSTEMS......................... B 3/4 1-2 3/4.1.3 MOVABLE CONTROL ASSEMBLIES. B 3/4 1-3 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE B 3/4 2-1 A 3/4.2.B

-/4.2.2 HEAT FLUX AND NUCLEAR ENTIaAnPY HOT A EL A TOR .......................

3/4A2.2 HEAT-FLUX-HOT-CHANNEL-FACTOR FQjZ I ........ B 3 412-4 314 2.3 ARNHCLT-y O I *-CHANNEL-FACTOR RTS

_H .......................................

3/4.2.4 QUADRANT POWER TILT RATIO. B 3/4 2-5 3/4.2.5 DNB PARAMETERS. B 3/4 2-11 3/4.3 INSTRUMENTATION 3/4.3. 3 l AND 3/4.3.2 REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION........................... B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION................ B 3/4 3-2 3/4.3.3.:1 Radiation Monitoring Instrumentation ...... B 3/4 3-2 3/4.3.3.!5 Remote Shutdown Instrumentation........... B 3/4 3-3 3/4.3.3.~B Accident Monitoring Instrumentation....... B 3/4 3-3 BEAVER VALLEY - UNIT 1 B-I Change No. 1-0132A I

[IProvided for Information Only.

TECHNICAL SPECIFICATION BASES INDEX BASES TECHNICAL SPECIFICATION BASES FIGURE INDEX FIGURE TITLE PAGE B 3/4 2-1 Typical Indicated Axial Flux Differences Limits as a Function of % RATED THERMAL B 3/4 2-3 POWER for RAOCVersus Thermal Pewer at BGL B 3/4 2-2 pTicaoFm zmael iPed-open tinv-Enel-oPe, B-3A 2-4C BEAVER VALLEY - UNIT 1 B-VI Change No. 1-01-424 l

4 PE . .RD -.IBUTO I,,d IFor .

LI-I - -to O an;y.

,Providedfor Information Only.l 3/4.2 POWER-DISTRIBUTION LIMITS BASES The peccifications of this scetion provide aesurance of fuel integrity during Condition I (Normal Operatioen .)and- - I- neidents oe Moderatc Frequeney) evensetsby-: (a) ma naining-the- nm-DNIBR in the eerer the design DNBR limit duri g-ermao- peratioe and in short tc-rm--t-r-a-nsflen~t-s,and Hbb ___ i in-t-e---f-i-s-iorr-gas--r-el-e~e,-fueeT e l<t tempr-c e_ A nd t. elzAdzA= .. A t:1'At

~~~~_ _'_tea AssAt_ in ddti

_ __1 e__ , .,mtin-h _e dnit durngCodition I events provides ass ranee tht the initial enditions ass ed for the LOCA-analyses ae- met and the EGG

-aeetac - r a limit£ of 22 F is net -AAAeede

-The-def-imi-tit-one of---hot---ehanrtel---f-aect-er-s a--used---4}--t-hese specifications arc as follows-?

F (,A,) IHcat Flux Iot Channcl Factor, is defined as the maximum local eoat flux e the surfaee ef a fuel red at core elevation Z divid.d by the average ful red heat flux, allewivng -fe manuf-aet-r-ing-t-oleeanees-enf-uel-pel4et-s-and-rods--

AH-Nucleaer--Ent-halpy-Rise-Het--Channel--a-ter--,--i--defi edas---ta e ratio of the integral of linear power along the red with the highs cnt ratd powertoh-a erdpw.

3/4.2.1 AXIAL FLUX DIFFERENCE (AFD)

BACKGROUND

.The-Ipurpose_ of-thisLCQ-isRto-establish-limiLts-on-the-valueEL-of-the AFD-in-order to-limit-the-amount-of-axial-power-distribution-skewing to-either-the-toP-or-bottom-of-the-core. BYvjlimiting-the-amount-of power-dis.tribution skewing.-core-peaking-factors-are-consistent-with theass mptionS usted in the qfettyanayIsSlimLim ing Power rci strihution ewi __ _ ___ exenon di~rribution

= ntrol-Relaxed-AxiaLfffset Cont o(RAOC) isacalculational-Drocedure-that defines-the-allowed-operational-space-of-the-AFDveersus__THERMAL POWER, TheAEDliimitsare qelected-by-considering-a-range-of-axial xenon distributions thab may occur a resaltt of large variations of the AFXD, sequen9y. power peaking factor; andu areexanind to gnsure-that the-1oa ofcosn cieetLO~~

f flfo1waccident. and__anticipated.-transientli-ts-are me .t violation _o~fthe__AFD__limits lnvalidate theconclusions__of the accidentand _transientanaLyses _with regardto fiueLcladdinq integrityv The imn so I +/-Ab FLU TT DIFFERENG assu that the- %T& " _s._

W -\..

envlep

._,.. v_ s U - - <...;4, alzedaxil VA eak-t-n--c-nr-+/-

-\F>v is-t-t--per edoued

~e te-oht durlng----eIYY. e noroa pr-- -- e e t f en reditribtionfollowing peweir changes.

Tar et flux differecne ia detciee Z-A eqyiibrium xIo-eni t=one withr thete __epective insertion 1 ... - ¢H- h -d be inserted n ea their pfr steady state operation- at-high-pewer zv~ 1.

l13 tevale-&f-the-t Thc lget f 3ux-di-f-f-er-enee-obtained-under-t-hese eendiAH~ons-diheided-by--t-he-faet4o-efe freatonEf RATD M -PGWERia t-he rg cont Target flux diferces for other THERMAL POWER

-level arc BEAVER VAL3EBY UNIT 1 B 3/4 2 1 Amendment Ne. 1'54

,-- ,- m -I I n__I ~.. - .~-,. ,-.-~ , W-~~~ _~ 7 ,

PGWER DICTRIBUTIOII LIMIT£

[I Providedfor Information Only. I BASES 3/4.2.1 AXIAL FLUX DIFFERENCE (ArD) (Cznt-inued V - 4 U..,

_x 4---I -L-Y

~1__ -Lj _.. PTTPvsarB D TH~.T T En "^T._ ""

va5u

- 1-

_,v

~x

- JlSvWy4 apprate fraetie-lTEMLPWcl "ve.Teprei -j_

t-he-t-arge . lif-f-erence-value i neesr-nup eonsidexat-ns--

Althoug1 it ja 4 i-ntended that--the--p Itnt w4-4-It -eprated with the AXIAL FLUX DIFFERENCE within the target-band about the target flux diffErence, during rapid plant THERMAL POWER-oeduete e, eentel xed wso vR! xes_ - _11--u^ _s w - _7_ A r3a-r - A 4:._C -L -- __ WS:

r__!M-

- - -IV_- , - - - - - - - - - _ - . - 1 a. Z- _ 1.1

__ .3.z7A-P - 1

-sr i1 J _3 ftllY. . -- _IZ rs ! _ - - _. I I -S_ _Fi m i~ --

xsenonre twbt-i-n-s -ficiently-t-o-ehange-t-he env eoetf<ekn f-aeters-whieh-may-be--reaehed-en a -ewsequent-r-et urn-to RATED TIIERMAS POWER- wt-hthe- AFD within--the--t-aret--band--prveIded--t-he-t4me devition erfflat---liit he pevius 2 hers is previded for operation utAeid pcecified in the CORE GPERATIN T LIMITS REPORT fer THERMA TILPGWF rv. IV. e VTi_

r,< q h~r%

mr.M rn a

7rr~ e -_

T -.. rjlTE 2W

.1.

=

'I

=ee1 Uew~

_ o-z

_ .. S 1D ro. _A-ana rno._

_ U-

_.C U

01 I11--l-11t

~TTnllT TT.?T"n

"^~~~ A __

-X'vWLM$

- v __ _.F U=lelnU1-

-I ArD-out--de ef-c tege -and aes less lwgnlaf-eant. T. -enalty of-Ihniurs-M o~tmln 1=4M-Ate WC}~seleimi- s fiane

_ __ 4 __ _ _*_ .

L.A.. LJSSI L.SA kJ.LLAACL. frI.L ILLLI LI S.LIiifl+/-J LAL.L4. L.S*4. V.ILASA SS 04. 4.1 *%.IAA.t .. 1/44.. fl.&LA.L flit. 4. SAL..

... pute det~s --- ute aver-ageXy ef eaeh; ofteGPR exeer deteeter-utput aand provides an-a arm!-esediatey if out-i-de-t-he-t-arget-bhadnd-dathTHERALE-HRUPOR-4 e--g-r-eat-f---t-han----gG%--etf RNED q4E UP.-i EW'ER-F----D -rn epiea t- -J at----hERAJrl--PGW.-%-4eve-159 bet-wee-0-%--and--950%-and--between 15%-aend-SID4 RATED TIIERM!x-POWERr-the utereututsan lar L.--- =- hen- . the- penalty deviatie~n aeeumulates bzyey,zs t, ef I ads, respeveWy Figurc B 3/4 2 1 ehow3 a typical monthly tar-get bandnear th-e beginning of eorc life.

The F itore on tomatiCn ha Susin the 1litproces comnuter. which has anAFD monitor alarm The comnuter determines the minuteveraveeof-eachoftheOPERABLEexcoredetector-outpt;s andpxov san-alarmmessageimmediatelyf theAEDfortworormQ nDPERABLE exconre channels~ is: ouitside itsm msmeeifiPA limitsm Tf the AFD monitor-is-out-of-service,_indicated-AF-D_ for-eachOPERABLE-excore channel-is-manually-monitored-in-accordance-with-the-requirements specified-in-the-Li censingRecruirements-Manual..

BEAVER VALLEY - UNIT 1 B 3/4 2-1 Amendment-Change-No. 441-024

Po. - -vId for - Infr i'o n. ng. .

IProvided for Information Only.

POWER-DISTRIBUTION-LIMITS BASES 3JA_2_1AXIAhFLUX-DIFERENCEU (AED) (Cont-inuedL BACKGROJNDlIContinued)c Althoughte-RAOC defines limits t atM usslemet to Stlsfyafe nalse s. typicpll- anr oppratin schereC- nstant Axia -p-ffst-t Control.ACAQC)__isuse-d5to-control-axial-power-distributionin-day_-tQ d ayperatLon See__WCAP43 (nonprprieta r PowDe istribution Control_ and-Load Followinq__Px oedures,2LlWestinghouseEl-ectric Corporation,-September-192AL____CAOC__recuires _that-theAED_ be controlled-within-a-narrow-tolerance-band-around-a-burnup-dependent target-to-minimize-theyvariation-of-axialpeakinq factors-and-axial x~nn lisr~lltio ilrivp4a ivnit- rmvnoivirgs:~

The CAOCrC onerat-i na srace is t-%rnito I I% smal lev anA 1 4 =c uy4 I-hTin i-Ih D PtOf P ra 4- - - - - -a fl-- 4-- -- 'I .. " 4-1, 4 - 4-- 1, - fN A r% t &- 4 --

a~1s_59 -- o WU N l L- iUl~ Q EpCe

,constrains-the variation-of-axial-xenon-distributions-and-axialpo-wer

-dis~thiutions~ _RAOC-calculations-ass-ume-a-wide-ranqe _ofxenon distributionsaand-then-confirm-that-the-resultinqcpower-distributions satisfy the reauirements of the accident anIlvyes.

SAFTY

_PPLICABTE SES The__AED-is-ameas-ure-ofAthe-axia-power-distriibuti-on-skewinqjto ei-ther-the-top-or-bottom-half-of the-core .- The-AFD-is-sensitiverto many-coie related-parameters such-as-cont o ank-positions-, core power level,._axial burnu -axial-xenon-distribution, and. to-a Ilesser extent. reactor coolant temperature and boron concentration The allowed ranae-of-the AFD is "sed in thebnuiclear dgmin Croess to con f i rm---.eato ihn hs prnfhmes ore,=pfaki tors adthateet safetvanalysia reque emnts, TheRAQC-methodologqyLSeeWCAP--021S9=PA._Revision_1A.,2 "Relaxationa ronstant Ax~ial Offset-Control-F0 Surveillance Technical Specifications. Februa-y l994) establishes a xenon distribution libr;;rv~~ tettml Wit ieAT iis n ensi-o al-axi-al power distribution cElculations are thenPrformed nstae

,that-nornal-operation-power shap~esar-Pacce tablefor-theLOCA-and-loss__of flowaccident. and for initial conditions of anticinated

.si-etvans ienThe-tentatuivelimits-are-adnssd-asn

,the-safet-y-analysis-r-eauir-ements-BEAVER VALLEY -AUNITBn B-31-42-22 Chanqe-No._-1024

ROWER-DI STRIBUTION-jIMITS LiProvided for Information Only.

3IA4-2--AXIAL-FLUX-DEEERENOEA(AF-DE+/-Cont-inuad}

AP-PLICABLE-SAEEYM-ANALYSESU-ContitnuedI Th lmtsonte F) nsr t~t-1__ i t on tie-Hat Plux Hpt Channnel PFar'tn-r - dZ7. A~re nnot p~rPPHPHe rtitrincy Pi t-hpr -nr-jrmn1 operatiLon-orin-thev nt-oL-xenonrxediLs txibution-follo-wing-p -wex

.changes--The-limits-on-the-AFD-alsox-e strictthe-ran e-of-power distributions-that-ar-eused-as-nit-ial-condi-tionsin-the-anal-vses-of

-Condit-ion-2-,3-,or-4e-vents-- -This-ensure s that-the-fueL-cl adding integr-ityis-maintained-for-these-postulated-acc-idents--The-most limnit-in ~ni ti 4 event- With rspecrt ito- the A1FT) limits -is -the T.OCA. -The-mpstJ ht rcndlitionn 'A Pnt with rP-pect- to -the-AMT mjjil`aiqst~he lpi Sof flo aciet hms lmtnition2 emets-ith pptto hpAMlimits icuide-the-uncontro~lpld RCCA

~ank-withdrawalat-power., dr-opped-RCCAs , and-boron-dilution accidents. Condition 2 accidents simulated to beain from within the AFD--imits-ar-eused-t-o-c-nf-ixnthe-ad-cracyvof t-h ~~pwr-AT-and The limits on the AFD satisfy Criterion 2 of 10 CFR 50. 36c)(2)_(ii),

LCQ The-shape-of-the-pnweprof i1 ei ntheaxia1Lf e ethe~exica1y direction-is-largely-under-the-control-of-the-operator-throuqh-the manual oseration of the control banks or automatic motion of control banks. T aomthe Ltic mOtion of the control banks isi osp to mperaturedeviations resulting fromp-manualoperation of the chendcaL~and. VQu Crontrl system to chanqe boron concentration or Iromnu we-l-evel changers Signals _are available to .the operator from__the Nuclear Instrumentation.System INIS)__ excore _neutron-detectors ((UFSAR.

Chapter 2J--_Separate-siqnals-are-taken-from-the-top-and-bottom detectors. TheAFD-is-defined-asthe-diffeerence-in-normalized-flux weLqnl F c nhe-tn I . otm ieteto flux difference nitsc fpresue a ifferencenAg annd to ld s flux orAeLA The-AFD-limits-are-provided-in-the-COLR. Ficrure-B-3/lA2--Ishows typicaL RAOC-AFD l imits-.TheAFD imits-for-RAOC-do-not-dependon the-target-flux-difference___However, the-target-flux-differencemav be-used to minimize chan es in the axial _ i__tiut__inn_

vtiolat-ing-thimsLCO-onthp AFn cr-u11a d abe o e if n rnnritio-n

'A- or 4 event-o'rrnvr whil-

-- th-h AFTM is oiut-iAe its

-- , - W specifted-limits-~

BEAVERNVALLEY--AJ-NTT 1 Cag~~~

-Change-No---1--D2A

POWERDDISTRIBUTIONLIMITS BASES 3L4.2 1llAXIAL-FLUX-DIFFERENCEIlAFD)i(Continued)

APPLICABILITY h r i scable iemt in than ssual to Sot RTP when theci mzInation of THqPMAL POWER and rnrp aking factors are of primaryjimportanc in ..efe t lanlysas.

ForAFD-limitsdeveloped usincRAoC-method-ylthe ue-ofLtheAED does-not-aff-ect-the-limitinc-accident-conseauences-with-THERMAL2POWER c_50_RTP and for-lower-operatinqcPower-MODES.

ACTION AR an a;Iternt-ive to, restorin t.Jhp A~n to within it-, epecifipcl imt. e-eMNxmue -~-TRt i-on to -< 0 k RTP-a nr ih th-au of th A~

not-impor-tantintheapplicableasaf-etvyanalyses. A-completion-time of-3-0minutes is-reasonablehbased-on-operating-expexience,--to reach S D RTP-without-challenging-plant-systems.

SU1yvEILLANCE RECIEMENTS (SR)

SR 4.2.1 .1 ThisSurveillanceerifies-that-the-AFD,_as-indicated-by _the-NIS excore-channel1,is-within its-specifi-edlimits. The-survei1ance interval-of 7 daYs-is-adequate-considerinq-that-theAFDiAs-monitored by-a-computer-and-any-deviation-from-ecruirements-is-alarmed-orithe indicated-AFD-is--manuall-ymonitored._as__required-by-thejLicensing ReguirlementsManual_

BEAVER-MALLEYz.A3NT _N B 34_2-2b 0hange No. 1_0-24

FIGURE B 3/4 2-1 TYPICAL INDICATED AXIAL FLUX DIFFERENCE VBRSUS TIIERML P WER AT B_

iits-as-a-Hunction-o" RATEDTHERMALPOWER-for-RAOC BEAVER VALLEY - UNIT 1 B 3/4 2-3 Rc i ssued-MaT-2ChanqeNo 1- 024

[I - ". r-S ProvidedforInformation Only.

Insert B.1-1.

2.20 2

90 UNACCEPTABLE _ t _ _ _ UNACCEPTABLE s4 a) OPERATION (\_ OPERATION 0 __ ACCEPT BLE __ _

80 OPERATION_ _-

8C (-_3250)-

0 c4 30 - - - = =-l = = -r - - - Fr 3C_-1

_ D1ISAFTG _ _OR _ __

ILLUSTRA~TIONONLY_.. l 1_ 1 1 20c_ DOAIQT.JSE..FOR _ __

10 1 1__

40_

-60 -50 -40 -30 -20 -10 0 10 20 30 40 50 60 Flux Difference (Delta I)%

- - T :"* s r t- r. - .: - :. 74w Providedfor Information Only.

POWER DISTRIBUTION LIMITS BASES

, A.

iJ ' t.*

z.

-9~

M.:. :

/.

,lFI - 11-t!,

,,t...

"-1 haUf ftfl

-tr^

i hP V tl tI.J ittua,p%

\ UTn t.,s..J....A .. *

^

-TTT'T af~ 13.

31-4-22-HAITIUX-HOT-CHANNEL-FACTOR--Faoz ensurc that 1) thc din peak leeal power-densietyand NB ar no emeede an 2) n_ . AAve A'-,\_& f A _t ApTnne 'AA -

1a fucl clad temperatrc will net emeeed the-NEGGSaece.ptanec critcria Baeh of--these hot channcl factors are meas s ble W~t-1-1 ne-rmal-ly ly be de Mined periedieally as specified in Specificatizn-s 4.2.2 and Specificatie 4.2.3. This periedic strt-veillanec i3 sufficient to

a. Ccntrel reds in a- -sing-c gz p-meve-tegether with--no indivdAA-d lred7-i-nsecrtien-cli--erng--by-more-t-han +/-92 steps frem-t-he-grvup-demand-positien-.

b. Gentrel red greups arc equened wth -everlappingrgroups as deseribed in Cpecificatien 3.1.3.'.

and 3.1.3.5 are maintained-.

d-.-aThe axia-pow r sdistributien-r--expreseed-i4n-terms of AXIAL FLUX DIFFERENCE is-maeained-wAt4An-t-he-1i4mt-s.

The-r-e~exatin4 as a f4unct ien f-T-HERMAI--POWER-a14-ows-ehanges in thc radial pwe-eshae F~-A1 I -- R dbios It rO3 t-hrugh-d-abve-ar-e-ma ntained.

manufaetur-ing teleranee must be allowed for. 5% is th ap~piate ehprmna rror allewanee for a full corc map--takcen with the incorc deie the aplrepriate The -peei ef----jH -- ont-a4ns---n 8---%al4ewance-for-unecrtainticswhich mans that normal, full er, e loop Gperatien wilT rePtGR--bhe-des-gn i-mA- peeified in-the CORE 50D~-G LMTS OS- O_

BEAVER VALLEY UIT 1 B 3/4 2 4 Amend et Ne.-54

Il

-- -- S..;. 2 - - - . q Providedfor Information Only. I VuwL1t D+/-IlsliUTIurI LIMITS BASES 3/4.2.2 AN -3/4.2-.3 HEAT- FLUX MND NUCLEAR ENTIM13PY IHGT GLANNEL rACTORS 0 a n f Fuol od be4-n~= eduo thoe Value of-t-he-D . ratio. marc~4-n--as been mainti-a-ned-between-the-DNBI value used-4n-the-saf-et-yv-analyses(1.33-and-t-he-deaeganlimift L'_2}-- e-efet-t-he rod-bow-pena-tv-and-ethe?

penalties-whieh-may-apply--

Thc-radial veakcine factor- Fa oial ene-d assscw1rtanee-t~v1fa tehe _hne fap PS.: re inswti t 1X

<4V r u , W 1D W 1

V s . 1- R l T.ee-The .- _

CORE OPERATINC LIMITt RPEoRT w frdm-e

_peted powv r urnup conditines in the cerc.

BACKGROIM The nrp of the lmits on the values of F Z iq to limit the local -ell-et)_peak-pgxae dens healuo alonqgthe-axial-heiqhti(Z)_of-the-core.

E(Z)_is-defined-as-themaximum-local-fuel-rod-linear-power-density divided-by-the-averaqe__fuel rod-linear-power-densityI. assuming nominal-fuel-pellet-and fuel rod-dimensions,.__Therefore,_Q.(Z)-is-a measur-e ofthe-peak-fuel -pellet p-ow-er-wi~thin-the-reactor-cor-e-During-powex-operation,_the-globalpow-er-distributi.on-is-l-itedmby TIM 1.2-.1 "AXTAT, FTTIjx nTPFFRENCE -(AED)," and LCO 2-2-4 2QUADRANT POW1 TTT.T PAT-T(5ilPPTR). whic-h Arp directly and nnnti nuo slv measuredprcs variab 9rbprL-es T.COs, along With T.C'O I6,3-'

LControl-_Rod-Ins~ertion- imits,2" maintain-the-core- imit-s-on-power

,distributions-on-a-continuous-basis-En(Z) vari wit ful IoA ing patt control bank i sertion, fuel burnup,.and-changes-in-axialpo-we~rdistribution..

FQo(Z)is-measured-periodicalv-yusinq-the-incore-detector-system.

IThsemeaaurementsa=rg Jh wenhtra equilibrium-conditions.-

Using-the-measured-three-dimensional-power-distributions-,it is Possible-to-derive-a-measured-value-for-FQ(Z). However-because-this Yalue repr sents an-equilibrium ondition,_it doesnotLinclude-the rari at ionsinl the value of--axe- Ye__psentjduring spiwtch__-

nonequilibrium-situations-such-asload-followingqor-power-ascension-RPAVEP VAT.T.PY - UNT1e -Chang o-_1-C24

PQWER-DI SIRRBU-TON -LIMITS Prouided for Information Only.

s-I BASES 31--2-2HEAT-ELUXHOTCHANNELFACTORmEo{Z (ContinuedL BACKGROUNDR.IContinuedl Tn nrolint fnor tlesgo{ssible variation t-the= cuilibrium yvalbipnf Fn-(Z)

- -(

isz nrldiqusted as.-P4(7. I - I bv aneeain

=

tonnendeint- fac-tonr t-hat Accirnvts for the calcu ated worst case transient-condit-ions Core-monitoringaand-controlunder-non=-equilibrium._conditions-are crcnmn~liphp hv cirperptina t-he conre within the I imitsq nf the approPriateALCOs, includinthe-limits-on-AFD._QP-TRand-control-rod insertionr APPTLICARLE SAFRTY ANALYSES This LCOprecludes coirepower distr-ibutions that violate-the fo1oŽwinguel-des igncriteria-a.. Duringqa-largesor-small-break-loss-of-coolant-accident (LOCAL _the-peakcladdingqtemperature-must-notexceed 22D-OF,_as-specified-in1l0_CFR._50.46,_lL974..

blDurinqgaJloss-of_2fonvedtxeactor-coolantfow-accident, theremust-be at least-951p prsbability atthe_951 confidence lvel(the95195DNB-criterion)that the-hot fuel-rod-in-the-core-does-not-experience-a-departur~e-from nucleate-boilinqgjDNBL.condition,.

M..

Thirinr crx

-eicted the enercy depositiont rod accident, the fuel mus~ nt exceed 280 cal gms pec~ified iiin Recrulatory G ide -.

77, Rev. n0.My 1974 and d.___The-contrDolxods-must-be-capable-of-shuttingdown-the reactor-with-a-minimumrxeauiredSDM-with-thehi-ghest-worth

.controlrxod s-tuck ffullvywithdrawnas_ specified in 10 CEFR Appendix.GDC2E Limitts-onF({Z)_ensure-that-the value-of-the-initiaL-totalpeaking fact.or-assumed-inthe-accidentanalyses-remains-valid. _Other

-criteria-must-al-so-be-metle qmaximum-claddingqoxidation,_maximum hyd en-qeneration. rcool hlp qe ometry,_la onqc encm rinqL Hwever, th tpclt-raturei,s_ if-ally most 1u j q-FQI(Z)_satisfiesCriterion_2_of lDOCFR_50.36(c.)_(2)_(ii)..

BEAVERVALLEY VUNIT1Al B3j-4 2 4a gh-anqe NOJ -D=2A

PQWER-DISTRIBUTION-LIMITS Providedfor Information Only.

1

.1 BASES 31-4 2-2_HEAT _ELUX-HQT-CHANNET, PACTOIZQ V -Z31Continued}.

LCO The Heat FJ-ux{Hot-Channel__EactorxFG{Z3s hallb==eli mitedbythe followinq relationships-;

EQ ( Z)liACFQ / pljffK ({) for_ <_5 Fo-( Z )L[CFQ-JIIL5iK+/-KI(Z)-f-or-P-<O 5 wherpe= rpFis th V limit a rovidedinhe K(ZAis-the-normalizedFQ (Z)-as-a function-oLfcore-height provided-in-the-COLRand PTHERMAT P=W The-actual VA1ueg of.CFQ and Kv(Z) are ien in-the QCOR Dhwever. CFO isn numer onh p er of 2Q0andaKiZ) is a unction that lookslikeLtheoneprovided inFiqureB3422FiqureB_3 A 2-z2isforil-lust-ratiionopnuronesolnLy.The5CQLR-actual-unitspecific figures-are-contained-in-theCOLR..

Fsor-Relaxed_-Axia-l Offse~t-Control-op-eration,- j,.Z)s-appr-oxima-ted~b C Z a Z(Z) Thus, F and W..... meet theprec limits-onEs1 I-An C(Z))evaluation-requires-obtaining-an-incore-flux-map-An-MODE l.

Frmthe incore flux Maresults we nbtain-thie-e-aurgd valueF(Z)l

-of--IZL--Thepj, E c (Z)f- FZ

t=

1.0 15 Whe-r ! 1 . OR15 iR a2 fac-torr t-ha=t- aerni n t- c for fiuel manu'ifac-turinoT tolerances _and__fluxDLmap__measurement uncertaintyz as

.pecified in WfAP-7308-L-P-A, "Evaluation of Nuclearr Hot Crhann&l Factor TJncrtainf es. June 19RR C

FM() is an excellIent approximation for FIZ) whn the reactor i a t-t-he st-eadu stnate nnwer at- which the int-e-n- fluxw- -- -

man was t-aken

- - - -t BEAVER NALLEY__JUNTTI1 B; 3/4 2-4b -Change-No---l--24

I Pr id _

for I form a v - -o l lProiddfor Information Only.l 1.2 1.0 0.8 N0.6 I--

0.4 0.2 0.0 0 2 4 6 8 10 12 Core Height (feet) aurDBo3IA 2-2 Typical-F T~ormalizedOPerating~nvelope ,K(Z)

BEAVERALLEY - UNIL1a By31-2 4A Change-No-l024

I. _ r - z I,- -': I I 1.1

- rS ~

X 3? M t - r- -

P-OWER-DISTRIBUTION-jIMITS Pl ProvidedforInformation Only. I BASES 3IA.2.2HEATFLUX-HOT-CHANNE LACTOR-F-ZWl-lContinuedL LcO-CContinuedl Ee(Z)J=

s Q(Z) where. YIJ s cZ) -vacyledependent-function-thataccountsfor.nwex distribution _ transients encountered during normal pperation. WJZL)is-included-in-theCOLR. TheE Q(Z) is calculated-at-eauilibrium-conditions.

The-F(Z)_limits-define-limitinq-values-for-core-power-peaking that precludes-peak-cladding-tem-eratures-above-220 0F-during-either-a large or-small break LOCA.

This LC O rteguires opgratinL withinthe bound e n h safmetLrl<=gy analyses Calculations-areep-errmed-in-the-c-re-desiqnpr-oc-e-ssto

.confinm-that-thegcore-canbe controlled-in-such-a-manner-during pperation that it_canstawwith the LoCAF I)limi tIfE cQ(Z) cannot-be maintained-within the LcOTimitsrr-eduction of.the xoe nower is_ reauired-and-if Fw(Z) cannot-be-maintained-within the LCO-limits. reduction-of theA ED limits issrecuired.ilNote-that sufficientrxeduc.tionof-the__AFEDlimitswill-also-resuLt-in _a redction of orp Violatina the LCO limits for F_(Z) nroduces unaccePtable consequences ifia--design-basis-event-occurs whileF 0 Z otsideitsspecified limits!

ARPLICABILITY ThS () limits must P maintain i oD I to nreVent cre ner distributions fromexceedinga the limitsassumedinthe safetY analvses- Annlicahilitv in other moR5s ic not rentjred beratise there i e ither insufficient storef enr inthe fuel or insuffirignt Afe1 erretino the reactor co ;;n requirea Jinit__n the odistributionnf conrpower BEAVER VALLEY - UNITd1 B-31A-2--4d .Chan~geNo-__1=D2A

PONER-DISTRIBUTION-LIMITS Providedfor Information Only.

BASES 31-42-2-HEAIEhUX-HOT-CHANNEL-FACT-OR=FtZ)_I- Cpnt inuedL ACTIONS a,1_ReducinqgTHERMAL POEER-bY-2-by-RTP-for-each-lt-by-which E (Z) exisaintains an accm table_________

Q nower densitv. F4ZM is M F~dZ multiolied bv a factor accountinq__for-manufacturingtolerances-and-_measurement uncertainties. E M(Z)_is-the-measured-value-o _F z).-The comPle tion-time-of l15_minutes-provides-an-acceptable-time toreduce-power-in-an-orderly-manner-and-without-allowing

.the-plant-to-remain in-an-unacceptable-condition__foxran extef-nded oeAriod of time. Trhe max-i~mum allo-wahlej oowe~r le~vel init iall y deetermi ned byeaAC IfN_ a ffectecLtd.by reductions within 15 minutes of Ltheb C Z)determinatiop.,f r ;y r necest wi--thtocdhe-dersxease-daximuuWable nnwpr levpl .

Decreases in F'2(Z) would allow increaRina the maximum-allowabhlp w:_leveLand-icncreas in po-weruup-.t this revised-limit, a 2_A reduction-ofthePowerRange-Neutron-Flux- HiHgh-?ip Setpoints-by > 1% foxreach-1--by-whichE (Z) exceeds-i ts

_l it, ir a onservative gction forprotection against the pn-Mre *fcivr trmintg with *- gmze oower distributions - The comoletion time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is sufficient-considering the-sma~lllikelihood-of a severe

,transient in-this-lime -eriod-andcL thgepre~cedincgpxompt reduction-in-THERMALMSOWER-inaccordance-with-ACTlON-a I.

The-maximum-allowable Power-Range-NeutronFluxziHigh-Trip Setpoints _initiallyedetermined _bvAACTIONa.2_mav be affected-by-subsecuent-determinations-of Fc(Z)_and-would rCir _o _ irnFlx-Hgh T-r'p RetBoint reductions within 72 houmrs of theFg(Z) determination, if ly with the decreasedg maxlmu' allowable Power Panae Netiron Fluxw - Hiaih Trin Setonints Decrreases jn42EC(Z) would allow increasingthe~maximUmna1Iswableower Range-Neutro'lx - Hiqh-Trip-Segtpoints-BRAVER VAT.TP.y - IUT14 -Chaliqe No 1_024

- - _ _ - W a POWER-DISTRIBUTIONALIMItS Provided for Information Only.j BASES 3A.2HEAT ELUXJ1H CHANNELTEACTQR-F tZ)L ontinuedL ACT=ONS-AContinuedl a.3 _Reduction-in-the.O vrpower-AT-Trip-Setpointsl Yalue-ofiKl41 by > Vk for each i by which FQ(Z) exceeds its limit. is a consermatiyeactionf roprte-ction ag inst-the-conseqruences

-f sev e t ranssientswith-unana lyedower-distributions The-completion-timeoQfL7.2Thours-is-sufficient-considering the-smalljlikelihood-of-a-severe-transientilin-this-time period,_and-the-preededing-promPt-reductioninETHERMAL-P-OWER in__accmrd itbACTTNa-. The___maximu allowable ep ymowr d etermin s edbyCTION a_3_may be-affeectedLyhsubsecuent deteroinatons ofEQ( Z)

~Q and-wmoul-d-reqruir Oxrower-AT -TripSe-tp~oint -reductions within 72_hours-of the.FY(Z)determination,_ifnecessarvyto compl-ywith-the-decxeased-maximum-allowableOvierpower-AT TrpStans prae nF Z-3=c allow inqreai-dnq the~~ maiupsetpoints.

aAVerification-thatLF(Z)-andF W(Z )_have been-restored-to withinitts-limit,_by-performing-SRA4_2_2_2_andBSR4.4_2_2_3 prior-to-increasinqgTHERMAL-POWER-above-the-limit-impos-d bv _ACTION-a.I _ ensures that coreconditionsduring rationLat hi~qhe s levp-ls-md--f-uturje -oerationna~re consistent with safety aalyses tions Action a is m odifIxy te-l-that re~q~4re CTI N a.to be-performed-wheneveer-ACTION-a-is.entered . This-ensures

-thatSLA-2-2-2-and SR 4.2_2L3_will-be-performedDrionrt5o increasing-THERMAL POWER-above-thelimit-of__ACTION-a.l, ev-enwhen-ACTION-a-i ~ exitedPrior toe er-forminc ACTION a.4. Performance of R 4.2.2.2 and SR 4.2.2.3 are necessary-t-o-assure -FOL-is-pr-Operl Luat-ed-P increasainqTHERMAILPDWER-a.5 If ACTIONS-a-L-thr-ough-a__ ar-enot met__within _their associated-completion-times-,the-plant-must-be-placed-in-a MODE-or-condition-in-which-theLCO-reruirements-are-not applicablelThi s-isdone-by-placing-thePlant-in-at-least MODE z within hoirs-Thiscompletion tim i reasonable-hasd on gpratig gxperien -e-gagdinq-thgeamouS MODE-2-from-ful2-power operation-in-an-orderlvjmanner-and without cEhallenaina lant Cvntems2 BEAVERVALLEY -_-UNIT_1 B_31-4_2_4f Change-No._1I-024

BQ WE.; D1ITIIBUTIONJIMIT for nfo---r i Inl y- -y.

POWER-D-ISTRIRUTIDN-LIMITS -lProvidedfor Information Only.l

-. 1 --

3142A2HEATHELU HOIHANEL-FACTOREF 0 IZ-Acontinue&d ACTIONS-CContinuedl b ltIf T s

>is foonund-themxie-maximumclcula-ted yalue-ofL.J? 0j that-can-occur-dur-inq normal-maneuv-ers .W( Z )J _exceeds.its specified-limitsthere-exists-a-potential__foric MZ )to become-excessivel-vhigh-if-a-normal-operational-transient pccurs. Reduc-inqthe-AFD limitsby 2 *oreach1%by Wich-y(Z) exreedf its limit within-thp allowed completion ime of 4hours,_reestrictsthe-axial flux-distxibution such that-even-if-a-transient-occurred ,core-peaking-factorsBare not-exceded.

The implicit-ass-umption nistha ifWLZ alues_were recalculated (consistent wit te-ed AED li its), then 9cZ tIimes -the recalculated W~) vralues ud I meet the F 0 Z)_limit. Note-that-complying-with-this-action-_lof redUCinqAFD limitsl_MaY- LreuItiin __

Pw

__=

er __ = _

ction.

the =enrp for ACTTONR nppd bh2. h-l and hA.

b .2 Areductionof-thePower_ Range-Neutron-Flux-Hiqh Trrip Setpoints.-_byv2 %forifor-_eachl hby-which-_themaximum allowablp power is reduced, is a conservative action for orotection acaqinst sl+/-1 szcfequences of severe transients aistiutns Tyec letion time of 72_hours-is-suffficient-considering-the-small-likelihood of a sever transientin this-time-period-and-the-Pjecding prompt-reduction-inTHERMALPOWER-as-a-result-of-reducing AED-limits-in-acrdance-with-ACTIONJk.L.

b,3__Reduction-in-the-Overpower-AT-Trip-Setpoints livalue-ofK4)

> 1% for Pach 11 by which-themaximum a -lowable power is reducedis.a-conseryativreaction for-protection-against the-conseauences-of-severe-transients-with-unanalvzed-pomer distributions Thecompletion time_ of.72-_hours _is suiffic ient -cons-iderirsthe-sm trnient in thi time ppg ropt rPlimitsWERinaa anof wi h ACTIQ a --

JlMits in arnoreiancwith ACTLON-b-I.

BEAVER VALLEYN_ .haUNITn1 -ChancieNo. 1-024

POWER-DISTRIBUIIONALIMITS I Providedfor Information Only. I BASES 31M4.2 *2JHEATFLUXHOT-CHANNELEACTOR=FZ.)CContinuedL ACTlONSil(ontinued1

.4 Verification that F (Z) and Fo(Z)have been restored to Within itslmibefomg SR 4.2.52.2 and SR 4.2-9-3 MAT PQWE above tmaxinum

_ lepower I imit mpns-ed by ACT N1 nsuesu that

-core-conditions-during-operation-at-higher-powerJlevels-and future-operatioin are_consistent with safetvyanaLyse-assumptions.-

Action b is modified by Note 2 t gatre-ire~sACTION b.4to be nerformed whenever rArTTON b is ent-ere1 Thi R ensilres S 42..2and SR 422. ill beperformd prior to increasingTHERMAIWERabovethe-limit-ofACTTION h,.

even-whenACTION b is exit-d-or-iorttpoperorminq ACTION b.4 Performance of SR 4.2.2.2 and SR 4.2.2.3 are n s A Ire QI r PrI evaljupr31o Py "'r tno increasing-EHERMALAOPiER.

b.~5ITf- ACTITONS- --I throtiqli b4 are not met within their nnDociatnd otion t Rt hp plarin a MQ? orE xconsdition in wich Phe rTo, s~ mn ts are not anPlicable. This is done byv lacina the niant in at least M ODE 2within-L-hours-This-completion__time is reasonable-based-on operating experience Xeacrding the amount of time it takes to reach mony 2 from fiiiL- -pration ina n orderly maner and ith-uchalleng ngp _ant sytes SURVEIL2 CE2REQU1REMENTS SRA4-.2..2-1 Theprovi sions-of-Specification-4-0.A4are-not-applicableAbecause-all thi rvedIlances must be eom in MQDE_.1 4R24-22.- ad SR 4-2-2-A are modified b t 3 The-Noteapplies du3Zing the fit o _ox-ascenqln ft r a retupling It states that THERMAL-POWER-may-be-increased-until-an-equilibrium-power-level-has been achieved at which a nower distribution man can be obtained Thisallowance-is-modified._however._byvone-ofthe-surveillance interval-conditionsi. _ 4.2.2.2.band _4.2:3.bJ thatoeqire Yerification-that FM(Z) and WZM)ar-withinltheirspecifie LLimits after a power ri se:of more than 10%PRTP over the 1TRAT. -POWER-at 1064hIr thovmx wcbro, lnat- vgrifiv-Fa t-f Vhm with umt-if~crl l4mit-z BEAVARLYLALLEYX=I 1 R 314 2-4h 1apNTT I - -----

~-~-

[

I -. - -1 ar.

POWER-DIS-TRIBUTQNALIMITS ProvidedforInformation Only. j BASES 31422HEA.FLWLHQOT-CHANNELkYACTDRmF (Z)XACpntinuedI SURVEILLANCE-REQUIREMENTS&jlhntinuedI.

Babeen

-aue (Z). ZaXw(Z) could n_ l*nlybe mesured in this reloAdl core. ther-e is anothesuv ilcinralon-itn

2. 2 -a An 4-2-2-la auplicab'lje ol for -reloadl cores. that 2~.

me-ter-s1 7PRTP-7his-ensures thatsQmedeteminatinnofiE (Z)_an.E w(Z) -ar-eadeat-a lowerpowqe-rleynl-at-which-adeuate-margin-is-availab beforeoinq to1.0OORTP ___Also, this surveillance-interval-condition,_together

.wi.th-the-surveillance-interyal-condition-reruirinqgverification-of F (Z))andEw(Z)_followinq-a-pawer increase-of-more-than Lot. ensures that re n as RTP (or ny other level for

_______ ____is _achieved ____ th ahspncp of these gur - -an- ineS ti r eoWer to RTP and Prme for 1 lsit verificat-on of F (Z).=wndEQ(Z)d.

The surveillance interval ondpt-ion__is =not=jintended p requre yer-ification-of-these-parameters-after-everyvlfolAincrease-in-power lev-el aboyp-thelast veri ficati on. It onlyr retcuiresr erification af ter_.a-power levelAis-achievedf or-extended-operationthatAis lhigher than powei at which E(Zh=si laast measuredL SR 4.2.2.2 Veri:icationthatF _ (Z) iJs-within__its-sP-ec-ifiedUl-imits-involves

.incr-easing-.Y -(oZa)Itqwlfooxmanufac-turino -toer-n-e--andmeas-urement uncertainties-in-order-to-obtainiFlc(Z). Specifically, Fm(Z)_is-the measuredyalue-L FLZ Mobtained-froo incore 1uxma presultsand Fc(Z) M .8(Z)i.oa5. CZ) is then compare to itspe ified limits.

The-limiLt withwhi Fc(Z) s ompared_yariejt-- i r ecyrifhpowerj above 509._RTP-and-directlyv ith a function-call-d-=Z)_provided in

.the-COLR,.

Performing-this-sur veillance.iAn-MODE 1-prior-to-exceedinq 751_RTP

.ensures-that-theE (Z) -limitiis-met-whenRTP-is-achievied.,_be-cause peaking factorseneral ly drecreasep ap Per Ievel is incr-ased.

BEAVER YALLEY - UNIT-1 .Change No-__1:02A

POWERD-ISTRIBUTION-LIMITS Proided for Informaton Only.

BASES

~.

3-/.4-fi2-HEAT-F-LUX-HOT-CHANNEL-EACTORz-F (21 SURVEILLANCE-REQUIREMENTS tontinuedY.

If__THERMALPOWER-hasbeen-increasedbyv 1f0lRTiPsince-the__last de-te-mingt--i op of..EQZ)another evalu1atio in f~anatgr jsreg-pijrp 12-hoursafterxachie-vinqquilibr-umconditions-atthishe rpower lee1(toensure-thatLF (Z)) aluesaelbeingreduced suffic3ently with-power-increase tostaywithinth heALCO-imitsiL.

The surveillance-_teryal cM L31_EEPD-is-adecruatetomonitor-the l opower distribution with core ause such change~s are slow_and well controlled when the plant ispn _t in _accoxdance

_iththeTechnical..Specificationg (TLS SR42_4.2_3 The-nucleardesignprocessjincludes-calculationsperfformedto determinetLhat the ca

- it-bin e (z) limitsc

_fluxMa are taken in steady state conditinns.-the yariatio npol Pr ristributmion resulting f Prational maneuers ar nea_ data These variations are ho ever coner atiey lculate nidein id g oof-unit-maneuverBsin-normal-operation. The-maximum-peakin-f actor

,incz-eas vr-stead stat valule-scalcul-ated-as-a-Lunctiton-of crx~e elev-ation,.Z,_is-calledW(Z)-.-Multiplying-the-measured-total-peaking factor--FC(Z).,-bvYZ)_qives-the-maximum-FQ(Z)_calculated-to-occur-in normaL.operation, F Q(Z)

The-limit-with-which _ (Z) is-compar-edvarie-s-i h-power abo-vxe5Q9__RTP-and-clde-Lv-ywith-theu nuncti3nR(ZLj pro3vded&An-the COLR..

The W(Z) curve is provided in the COLR for discretecore9levations.

Flux ap ata are typicallM taken fdor-3 to 7S core elevations.

EW(Z) evalu areinot a-f

_ti te following aXia;l core rgins.measurecd-inpercent of core heightL

a. Lower-core-region, .froom-0-to_155inclusi veand

b. Upperscore-region,-from-85.AtoXL00%-inclusiv.e.

BEAVER VALLEY - UNIT 1 B_3-a/42-4 - lOn RTP above -the THRRMATV POWER of-it-s-last verification. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after achievi ng_ equilibrium conditions-to-ensure-thatFIZiL_iswithinits imiftathijqher power levels-BEAVERIVALLEY - UNIT_1 B_314 2-4k Change-No__l-024

1 , . .,- ;7 ~~- -' ~A POWER-DISTRIBUTIONA 1 IMITS Provided for Information Only.

BASES 3.4.2_2_HEAT-FLUX-HO-T-CHANNEIL ACTOR.FQI1o SIJRVEILANCEREQUIREMENTS ontinuedL The _ Pryeil hnnc interval of31_PD athe shapg _ _ _ ___ _d__' _ p. The surveill ce may be donemore frequentl-yl f__reqruiredby thexresuLts-ofFQ-IZ) eyaluations-The-survei2ance interval-ofr31EFPDis-adeQuate-to-monitor-the ghange-of-power-dis tribution-because-such-a-change-is-sufficiently slow, w ht is onerated inaccrdance With thre TS, tn precldre adverse peaking factors between 31 day surveillance 314A23_NUCLEARENTHALPYRISEAT CHANNEL-FAC-T-ORQFH BACKGROUND Thet a 7purposeo 1i -tismiLC'O t is son to thepaN;tlit est-ablish 1j~jntson thepwr-detnsit at~ay pointn=the _core o thatltheuELue._deslascri tera arge nt exceededand-the-accident-analysis-assumptions-remain-valid. The desiqnlimitz-onlolcalAp lletl and-integrated fuel rod-peak-pnwer densit-yare-expressed-in-terms-of-hot-channel-factors. ControL-of the distributionwith respect-to-these_]factoxs-ensures roreowe that local conditions in the fuel rods-and coolant -channel do -not chal e Corp n tp- tany loation__during either nrmal poeration or apos tula ed accident edinh lyses E NH is-defined-as-the ratio-of__the-integral-of-thel inear-powe a3nq_5hc_lfug radwihth d t- int-c -ower to the aver>g intefeg _a rnmeasre of the maxiliu total powe rprodu red in a fuelrod.

F HNis-sensitive-to fuel-loadincrpatterns ,bank-insertion,_and-fuel bnupb=N _ l r Iical ank-nrtion and yica1lydecreases-withfuel burnup, F NH __i noLt_directly_measurable but_is_inferred from-a-p-wer dis~tribution-map-obtained-with-the-movable-incore-detector-system.

SpAcificalL1vthe-resul-tsof-the-three-dimensional-power-distribution N

mpaareanaLyzedbvya_co piitrdetermine-FAH,_Thisfactor-is calculated.at-least-evervy31lEFPD___However-,during-power-operation, thegqLobal-power-distribution is-monitored-bYvLCOQ3-2Ml2_AXIAL.FLUX DIFFERENCE_(AFD)" andLCO_3.2-A, "QUADRANT-POWER.TILT-RATIO-(Q2JTR) which-address-directly-and-continuouslyjmeasur-ed-processyariables.

BFA V-_R VAT.T.EY - TM h T 1 Changeeo.1024 C a g o - 2

I

_OWEP DISTRIBUIONA-S

.o _ d.fo [ Ifo'_,r ." ;r - o I O P-OWER-DLS-TRIBU=LN-LIMITS ProudedforInforrnation Only.

Y -

BASES 1/4.2 3- NfTc!IR 3 E LPY RT-qH T MANNF-T3 _ATOR T; _Cont-inled BACKGROUND-IContinued)

The-COLRProvides-peakinq gfactor-imitsthat-ensure-that-the-des iqn basis value of tlieparturep rom-mcleate roilinm ( ) ims met fo nonaL-Peration. operational transients, and t it ionditio a-risi~qna f-rom

- z.Y everntsc

.__.I - ,M of modraArte,----f-reouenc- The TDNB escirrn hasis up iensres _N~pjnA~ty pobab4tih itih fuel-rod-is-at-least_9556at-a-951-confidence-lev-el___ThislIs_met-by limiting-the-minimum DNBRIo--the-95./95 DNB-crit-eri-on-ofLL22--for typical-and-thimble-cells.using-the-WRB-2MCritical-HeatFlux_(CHE) coxrelation -and-1-23-fortthetypi-cal-cel1and-l-22 for-thethimble

,.ll I1 tIP.ca V tIrol WtD M.- 1 r'ULTI r=nrr *4- AllI 'nKTR m +-gmil f -r- n via; gn euvnt-s are asuimed t-o heain with an . value t-hat- satisfie the LCO-recuirements-Operation-outside-the LC( Iimitsjmay. produce unacceptable reventces if a DNB 1 ntn-bapsi enuMres that there is noioverheating of the fuel th - r--eults in possible claddrngo tr _th aa fision ducts to the reactor-coolant.

AP-LICABLE-SAFETY3ANALYSES Limits AH preclude ore power distxlbuionstlat exceed the followincafuel-desi-cinAimitso a.There-mus~tbe-at-l-eastL 95% probabilityvat__the-59I.

confidence -evelItthe-95 /95DNB-criterion)-that-thehottest fuel rodin-the-core-does-not-experience-ajDNB-condition,

b. Dur-inqajarqe ora smll break _asofceolaraccident clapiri ek _.LO(A) tem erature (PCT) mus-t- not excred 220.6fin as ecifiedF in 10 CER 50-46. 1q74.t C~ rhurina an eiected rod accidpnt, the Pnerav denosition to the fuel must not exceed 280 cal/am as snecified in Reculatorv-Guide-l-I2Rev--O-Mav l91A,-and d_ pFlel AdesMi rn imi 4ts C -rreaui red- h 10.
CF y t v
50. Ann"Cndixs - -
A
_nC 26 fnr the condition e rol rods must abl of shut-t-ndomnathe__pac-rithoawminimh umrqu irrdSDe M with the6highest-worth-control-rod-stuck-ful-ly-withdrawn..
BEAVER.VALLEY - UNI1 B B_3,14_2 -4m §;hapge _Nq-_1>02A
POWER-DISTRIBUTION-LIMITS .Prodedfor Innfrmation Only.
BASES 314.2.3 R-WHAT.Py T H EL-EA =C (Cantinue APPLICABLE-SAEETY.ANALYSES-(Continued)
For-transients-that-maY-beDNBilUmited,_theReactorCoolantSystem fl~m andFNr~e tecrpaaee ost-inp-r-tanCPe.Telirt pn N -Hensure-that-theDNB-design.basis-is-me-tfor-normal-operation,.
Itpertransie q. a anv trans iiesievents of modra-Ph f r-cpi-n The 8 sis ensgnre3 prnhahilify thatDNB-will-not-occur-on-the-most-limi nfuel rod-is-atleast_9.51 at a_5 onfidence_1eveLp.__]This--is-em b lmi-t ngthe-minimum DNBR
.to-the_5195_DNB cxitrionL22fortpicalandthimble-ells using-theWRB-2MCHF-correlation,-andL- 23 for-the-typicaL-cell-and 1-22-for-the-thimble-cell-using-the-WRB.L1CHF-correlation_ These yalues-Provide-a-high-degree-ofas surance that-the-hottes t fuel-rod in the core does not expp ience a DNBN The allowable N limit increases with decreasing poer level. This f-unctionali-ty-in- NH _iS included-in-the-analyses-thatprovide-the Reacto~r-_CxreSaf e tYimit SCSls )of-SL-2-1_1___Ther-efore,-NEB-eents in which the core limits are modele lici1~.tlyuse this variable yaLue-of.FNH _in-the-analyses.-LikewiseLaJlltransients-that-may-e DNB imited-are-assumed-to-begin-w-ith-an-initial-i NH as-a-function rfpOwer-lev~eldefined-by-theCOLR-limit-ecruation.
T~he TLCA saf etvaalvs is indirectlv Nles ~3 P a i n=l parameter--The-Nuc-ear-HeatEYlux-HotChannel-Fact-or-, (Z)-and-the axial-peaking-factors-are-also-indirectl-ymodeled-in-the-LOCA-safety anaLys-es-that verif-the-acceptabilit v-ofthe-resuLting-peak-claddinq temperature.
jhe= fel is protected in part by Technical Specifications which ensure thr~gbe i t ions as d in th- a-fety -and a cci ent ana srem eolapwig-LCQs ensure thisL LCO 3-1 3.6, LControl-Rod__Insertion-Limits2LLCO-3 2I,21_AXIAL-FLUX-DIFFERENCE A LCOo -L22--CHeatOFlux-AHo-tChannel Eac-tor (ZOL } 3LC0 2-3
-"Nucl ear Rn In Rise1HotChanne a5 and T1QO 3 2D4 1
!LQUADPJU-20WER TILT RA PTRj I BEAVERYALLEY - UNIT_1n B. 314_2--4n ,Chanqe No_1:-D24
POWER-DISTRIBU2ITON-LIMITS Provided for Information Only.
BASES 3 2 _ _H (ContinuedN APPLICABLE-SAFETY-ANALYSES-(lContinued)
FNH and Z are meaFured perio us-irm-the-moyablp ircore detertor BYtem_-Measurementsa reenerallYtaken-with-thescore-at,.
-oxnear,_stead>state-nition ns.-Coremonitoringa and-controol under transient _conditions (Condition1-ev-ents)-are-accomplishedbyv operating-the-core-within-the l imits-of-the-LCOs-onAAFD,_QPTR, and Bank-lnsertionLimits.
N EAH _satisfies-Criterion-2-of 10CFR_50-336A(c)-(2)-(ii)-
LCO ANH _shall be-maintainedwithintheilimits _of-thexrelationship provided in the COMIR Th NH limit-identifies-the-coolant-flow-channel-with-the-maximum enthap riseThis-channel-has-the-hi-hest-probabilitv -aDNB Theimktinq-vale9f-FAN Jdescribedbythe-ecuation-cntained.Jn the-COLRis-a-desigqnradial-eakinglfactor-lnuclearenthalpyrise hot-channel-facto£._used-in-the-unit-safetv-analy.ses_
A power m ltication fantor in 1 i d 4 . s i anl additional qginfb;gher radial peaing fromredced thfema1 feedback and ratr contrl rod insertion at l _ __ __ __ _
voaluloed to increasp Jh value for PP p2Cified inmtheCOLR or eye rTVkReTRreyionnTioLHETERMALPi=WER APPLICABILITY e iu lmitsest be maiitainp i ODEItopre ude pp di tributions exfuel-desin fP Applicabilitvyin-other-MODES-is-notLxecruired-because-thereiseither insufficient-stored-enerqvin uthe ofuer-or-insufficci-enenerrqqyheinq transferred._to-the-coolant-to-require-a-limit-on-the-distribution-of coxempowerSpecificaU1yWthe-desiqn-bases-events-that-are-sensitive KtQEaH-in-otherUMODESl(MODES2-through_5ihave-siqnificant-marginto the INBRli. there fre. bherei snoneed to re t rict_ FNH=j theseMODES-BF1ERVALL.XY --- UT
I -
Pro . : nfo--i- on -- ._On- y..
POWERJDISTRIB=TIONTLI.MTS rvdedfor InforrnationOnly.
BASES 314.2.3 NUTCTF.P~. RNTHA.PY RIEF HOT cH-LEnti-nuedJt ACTIONS aW ith excedina its limit . reduce THERMAL POWER to
<_5SDRTP and-reduce-the Power-RanQe-Neutron-F-lux k s. M..
Trios Setnoints to < SS RTP in accordance with ACTION a.
Reducing RTP to *<5D0_RTP-increaseesthe DNB-margin-and-does state operatiorn I
_h_ ______ti______t__points ensures tha~t q nftinifl-nflVLQueratiofL r-e!Lains at-, an ttabe Il Iow ne-%wper level with adeouate DNRp marrin The allowed Pcompletiontimeof-2-hour5tro reduce_-THERMALP-OWERprovides an-acceptableatime-ts reach-the-required-powe levyefxoxn full-power-operation-without-allowing-the-plant-to-remain in-an-unacceptablecondition for-an-extended-period-of
.time._The-allowed-completion-time-of-4_hours-to-rese~tthe trip setpoints recognizes that Poncepower is reduced, the safetyvana1ysis assumptions are satisfied andthere is no _
iurnant neepd to rArbinm the tri n set-noi nts e _ Jo * -*._..*
Prote-cti~onSy-stem..
b_ Once-the-power-level has-been-reduced-to_<_50kRT-P-per ACTIONa,_anincoreflux-map-I SR-4.2.3.1) mustbhe-obtained andthe-measuredvalue-oLE Nverified-not-to-exceed-the allowed limit at the lower nower level The unit is pxovided 22-additional-hours-top eerf orm.thistaskoverand above-the 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months allowed-bv ACTION-a .- The-comuletion-time of-24-hours-is-acceptablie-because-of-the.increase3in-the
_NSjmargin which
- is obtained t lowerower lev the low proability of havinq aD lisiting event within t-hi4 *79A hn-p-r vNgmv4rnn1 f 4n=
'Arl;li I-t -- + eTn"nr gmrn=
hasaindicated-that-thiscomnple.tion-time is-sufficientt_
Qbtain_ the__njcnre _ p rf-orm-the- _eqired calculations.and-evaluate H Should as tsfactory=.js coremapnothomplew Zithin thei e euiredComp1etion Time,_the-plant-must-beplaced-in-a-mode-in-which-theLCO requirements-are-not app~licable. This-is-done-by reducin RTP to less than 51, i.e., aciigtheplnt in at least 2 lh_ _pur isreasonab3 e, q.peratj n xperience regardinc-the-time-required to-reac 2 from-full-power nonditions nano derLyanner-and-without-chalengig plant systems.-
REAVER VALLEY - UNIT 1 R. 314_2-4p. Mha~nge=No._l-0-24
.' , - I r-I . 1- ? - i. I . rmatio- .- 7. Z POWERDISTRIBfUT1ON- IMITS ,Prouidedfor InforrnationOnly.
BASES 14 3 NTTCAR PN7P1ATPY RTS H T MANNEL PACOR . (Continuexdl ACTIONS_(Continued).
Tdent--j~fjratonand correcgtion of thecause of an oit of limit-condition-and verification-that- FH .. is.withinAits specifiedlimi-ts-pxior to-increasing-THERMAP OWER-after-an out-oflimit-occurrence,_ensures-that5the-cause5that-led-to th F
-_ e Hnq its _limit is orrecqted.L anddthat subseauent-operation-proceeds-within-theLC9_limit___This action-demonstratexathat the- NH 1 Jit-is-within-the LCQ-limits~ro to exceeding 50TuJE. aqain__pxioxrto exceedinqg 5%_RT-,_and-within-24-hours-afterTHERMAL-POWER is 2 95_RTPT SURVEILLANCE-REQUJ.REMENTS SR 4-2-a-id Th e va 1u e o fi AH- Asde t-erminedAbvu s ing the-movable-Incor-edet eat ox svstem to obtain a flux distribution mar A data reduction comnuter P-ogramtla2hen-calculatesthemaximumvalueofp AFHfromthe-meas.red fljxddi stributions-After-each refuelinq, NF -Hmustbe determinedinMODE-lPriorm--t excding RTPe This tharirnpet ens zStat-FaH iit-are-mat at-the-beginninq of each-fuel c1 e..
.The-31-EF-PD-sury-e-llnce-3interval-_is-acce table-be-qause the-power
,distribution-chanqes relativly-e slowlY-over-this-amount-ofi.fuel burnup__Accordinqglythis suryeillance-interval isshort-enough-that the NH limit-cannot-be-exceededf or-any-s ignificant-period-of SR 4-2_3.2 The-measured valueoofLF. l must-be-multiplied-by-l.04-to-account-for me asur-e ment-ucer-t a int-y-bef-ore.Anaking -cpmnpa nsont~o-he N A4 T-7mil--
BF.AVRE VATLTEY - UNIT 1 Chan-geNo. 1-Q2-4 I
POWER DISTRIBUTION LIMITS Provided-for Information Only.
BASES 2/4.2.2 AND 3/4.2.3 HEAT FLUX 2VI MiC-LEA Bi~bblpy HOT_ CH4.1ai r-AGT-ORS-F-n =I! H(Contsne&
Fuel red hasbeen redWE rthe value intuene ynalyse (1.33) and-the-des gn-14mit (1.21) >ef t-et-t-he rd-bow-penat-ty-and-ot-her penaltle -whieh-may-appyY--
The-r-adial aeak airFy49-Ee<-F)i-s mea ur-ed-per4oically tco -pvide assuranee that the hot channel factor, %-If 7 -emains - - hin-n ite axmi-uA----heuy- limit for RATED THERMAL POWER -FXY provided in the v-RE OPERATING LIMITS RPORT determined-frm -eMptz p eere.~
3/4.2.4 QUADRANT POWER TILT RATIO (OPTR)
BACKGROUND The Quadrant Power Tilt Ratio 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. The QPTR is routinely determined using the power range channel input which is part of the power range nuclear instrumentation (NI). The power range channel provides a protection function and has operability requirements in LCO 3.3.1. While part of the NI channel, the power range channel input to QPTR functions independently of the power range channel in monitoring radial power distribution. For this reason, if the power range channel output is inoperable, the power range channel input to QPTR may be unaffected and capable of monitoring for the QPTR.
The power density at any point in the core must be limited so that the fuel design criteria are maintained. Together, LCO 3.2.1, "AXIAL FLUX DIFFERENCE (AFD)," LCO 3.2.4, "QUADRANTP-OWER-TILT-RATI-0 CQpn and LCO 3.1.3.6, "Control Rod 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 design criteria and that the power distribution remains within the bounds used in the safety analyses.
BEAVER VALLEY - UNIT I B 3/4 2-5 Amendment-Change No. 19 D21-0A4 l
Providedfor Information Only.
INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued)
Table 3.3-1 Action 2 has been modified by two notes. Note (4) allows placing the inoperable channel in the bypass condition for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months while performing: a) routine surveillance testing of other channels, and b) setpoint adjustments of other channels when required to reduce the setpoint in accordance with other technical specifications. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months time limit is justified in accordance with WCAP-10271-P-A, Supplement 2, Revision 1, June 1990. Note (5) only requires SR 4.2.4 to be performed if a Power Range High Neutron Flux channel input to QPTR becomes inoperable. Failure of a component in the Power Range High Neutron Flux channel which renders the High Neutron Flux trip function inoperable may not affect the capability to monitor QPTR. As such, determining QPTR using the movable incore detectors once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months may not be necessary.
The following discussion pertains to Table 3.3-3, Functional Units 6.b and 6.c and the associated ACTION 34. The degraded voltage protection instrumentation system will automatically initiate the separation of the offsite power sources from the emergency buses.
This action results in an automatic diesel generator start signal being generated as a direct result of the supply breakers opening between the normal and emergency buses. The failure of the degraded voltage protection system results in a loss of one of the automatic start signals for the diesel generator. Therefore, the ACTION statement requires the affected diesel generator to be declared inoperable if the required actions cannot be met within the specified time period.
The instrumentation functions that receive input from neutron detectors are modified by a note stating that neutron detectors are excluded from the CHANNEL CALIBRATION. The CHANNEL CALIBRATION for the power range neutron detectors consists of a normalization of the detectors based on a power calorimetric and flux map performed above I5iG% RATED THERMAL POWER. The power range neutron detector CHANNEL CALIBRATION is performed every 18 months but is not required for entry into MODE 2 or 1 on unit startup because the unit must be in at least MODE 1 to perform the test. The neutron detector CHANNEL CALIBRATION for the source range and intermediate range detectors consists of obtaining detector characteristics and performing an engineering evaluation of those characteristics. The intermediate range neutron detector CHANNEL CALIBRATION is performed every 18 months but is not required for entry into MODE 2 on unit startup because the unit must be in at least MODE 2 to perform the test. The source range neutron detector CHANNEL CALIBRATION is performed BEAVER VALLEY - UNIT I B 3/4 3-Ij Amendmenrthange No. 2-91-024 I
Attachment B-2 Beaver Valley Power Station, Unit No. 2 Proposed Technical Specification Bases Changes License Amendment Request No. 182 The following is a list of the affected pages:
Page B-I B 3/4 2-1 B 3/4 2-2 B 3/4 2-3 B 3/4 2-4 B 3/4 2-5 B 3/4 3-4
Ii Providedfor Information Only.
TECHNICAL SPECIFICATION BASES INDEX BASES SECTION PAGE 2.1 SAFETY LIMITS 2.1.1 REACTOR CORE ................................. B 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE .... ......... B 2-2 3/4.0 APPLICABILITY ....................................... B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL ............................. B 3/4 1-1 3/4.1.2 BORATION SYSTEMS ............................. B 3/4 1-2 3/4.1.3 MOVABLE CONTROL ASSEMBLIES ................... B 3/4 1-4 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE (AFD) .B 3/4 2-1 3/4.2.2 AND 2/4.2.3 H&AT FLUX AND UJC-LRAR RNTILALPY HOT ChANNEL rACTORS r-Q(Z) AND FR. B 3/4 2 a 31A42..2 AHEAT-FLUX-HOT-CHANNELEACTORmFEiZ ._ 32-4 3/4 2-3 NUTCLEA 1NTHATLPY RTSE HOT CH7ANNEL FAC!T0R xRH-B3-4-2.:Ak 3/4.2.4 QUADRANT POWER TILT RATIO .B 3/4 2-5 3/4.2.5 DNB PARAMETERS .B 3/4 2-11 3/4.3 INSTRUMENTATION 3/4.3.1 AND 3/4.3.2 REACTOR TRIP SYSTEM INSTRUMENTATION AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION .B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION .B 3/4 3-10 3/4.3.3.1 Radiation Monitoring Instrumentation .B 3/4 3-10 3/4.3.3.5 Remote Shutdown Instrumentation .B 3/4 3-11 3/4.3.3.8 Accident Monitoring Instrumentation .B 3/4 3-11 BEAVER VALLEY - UNIT 2 B-1 Change No. 2-04428. l
I 4Provided for Information Only.
3/4.2 POWER DISTRIBUTION LIMITS BASES The peei-fi-ations-ao ths 3cet-i-n-pravi-de as-ur-anee-f uc Moderate Frequeney4-event--by -amnta-a-t-ning-t-he-m-nimum-DNBR-in the cere . .e -design-M-BRlimit during--innral orti e Inis heet tem1l tras ens and- (b liitn th*-le__Va_ fSse gasll re-- se_, _ekl1:_F .s
_ An g _a- r-_--__
asued desig ritra.s in aditen 11-1us44l- V,;th peale ow=
density-during-Xend-it o I events prevides ass anee that the initial
-erUl-ed a the c LOA analy2se0-- at deE- e andtheee
,aceept-ane -- _r.o 20Fi ot-ecxeeeded-.
The-def-iat-ions---f-hot---channel--f-a-et-or-ea s---used--in---these epeeifieatiens are a3 follows; Hlaet rlur de local-heat flux en the surfaee of a fuel-rd at er_
elevatioe 2 divided by-the-ave-rage f el--red luxhetflu alle 9 tolernces en fe peles n
-H lt;uelear r,-.ap Rs llt Gh nn Faeter, is defined as the i-n-tegral- of lincar pewer alon terdwthtehge-integrtte the averae red vpewer.
3/4.2.1 AXIAL FLUX DIFFERENCE (AFD)
BACKGROUND o __i- POi-qt _r~ _ it __ he ynlut-s-of -the AFD in order to limit the Qu~int of nxapWL-r dis-trihbltion kwg to-eitherithe-top~or-bottom-of-thecore. Byl Iimitinq5the-amount-of power distribution-ske-winC conlieakinq facrs-are-consis-tentwith
-the_assumptionsuse-din-the-safeLtvanalvyes-.Limit-inqgpower distribution-skewinqgover-time-also-minimizes-the-xenon-distribution skewinq._which-is-a siqnificant -actorjin-axiaL-power-distribution control-Relaxed Axial Offset Contro1 (RAOC) is a calculational prpocdure that diefie e ll ep eraina P pf tbe AF)lerasu THERMM EWEROTM eAED Jimtsarselectedbyconsiderin -aangeq of aXial xenon-distributions-that-mav-occur-as-a-result-of-larqe variations-of theAFD. SubsecuentLy paower-eaking-factors-and-power-distributions are-examined-to-ensure-thath-theloss-of-coolant-accidentL(LOCA)-..oss nf l-ow _ accident, and _antic.iataedtransient limits _ar-emet-Vilation of li conclusions of the acncieipnt__and tranaient, an sthr.gard to fuelcladding i~ntg~rity Thc 9- lits on AXIAL FLUX DIrFFERFENE assure that thc rQS pper biund-eenvelope-t-me- the -neraled--axi-a--peaking-f-a-et-or--is--net-exeeeded-dur-ng-ei-t-her--normal-o-peratien-o&r-in-t-he-event--of--xenen redbstruten fepew eehanges.
Tar et flux diffcrcnce is de-cMined at cqulibrium=xce eondThee q -full length reds may bc pesitiened-within the corc in accrdance with their respectiLv inaertion limits and sheuld be hinpewer levels. The value ef the target flux differene~ bta4nedI
-r t -_ divided by the fractie zf RATED TIIERMAL DOWER
-i9 the taret- flux differenee at RAT-ED TIIERMAL POWER for- the
. _ _ :1 I_ It _ _- n._---- :c _ _ _ __ _ o asseciatcd core auiiup-.i ns. ret flum differeces Ear otner TIIERMAL POWER levvls arc zbtaind by m1 ultiplying-thc RATED THERMAL PGWER-value-by-t-he--appr-epr4ete-aEfeteinai THERM -POWER-4evel-.-The periodie-updat4ng-of-the target flux di f-ferenee-vtue iis nece sary to reflect corc burnup-eensideratiens.
Although-t-4 int-ended-that-the-plant--aw-1-be-oper tted-ith-the AXIAL FLUX DIFFERENCE wi-thin the target-band-about- t- t-rget- flux di-ter-ence-,-dur-ing--api-d-planbTHERl F ER-redueti-ons eotrl rod rdued TIIHIERMAL OWER levels. This deviat-e+/-. will net affeCt the
_ _-_ _ ____ ___ -if_ r%'_Vv--1 =3_4 - inn -n l J1-C%* t r.4 '. S.
v _ .
t Xr;,
U;.J. h L./ t; vl
. n v ff; La..
.4 .4 a.A;r sw~s-v
.. t n v s
t. fl.
u s h
y A
f
_-v . v_ , X factora which-may be rcached on a subsqcuent return to RATED THERMAL PGWEJR (with the AFD within-the target bevicled t-he-time BEAVER VALLEY - UNIT 2 B 3/4 2-1 Amendment- Ohan-geNo. 462-.
28a
L 'I I-- -f I . :'F; r " I ' -.f ' t. .. O
.Prouidedfor Inforrnation Only. I POWER DiCTRiDIJTiON LiMiTEs BASES AXIAL FLUX DIFFERENGr: (AFD)~ (Gentinued)-
penagt-y-deviat-ion-44mit-eululative during-the-pr-evIeous-24-hoeur-sa-is previded-fer-epe ation-_ut a idC c-f t-he--t-arget--band-but--wtth-ii-t-he limita _sp cified in thc CORE OPERATINC LIMITS REORT for TIIERMA.L POWER levels between 50% and 90 ef RATED TIIERMAL POWER. FPr THERMAL PGWER levels betwcen 15'° and 50% ef RATED TIIERMAL POWER, edviations of the AFD o e thetarg at h-e nienl*t-;rnf!;9:hnwrn acua timc,~ eue infj6can0c jelc3ti Pr-evisons--fe-r--Monoitereng--tin e- AF-on -n-autoemat i ebaaiea ar-e derived f rm-t-he- pant prI e e comput-ar-th rugh- t. Ar-D Moniter A-arm--T-he--emput-det-mines-t-heone--m-nut-ever-a eof eae uiE1tLSLB emearc aectector outputs and proviaea an- alarm -meacssa immcdiatcly if the AFD for at lcast 2 of 4 or 2 of 2 OPERABLE cxcorc ehannels are- utside the tarhet band-and-tc THIERMAL POWER i greater R h mp eratieion a rnalt th r%
PGWte _eputcr oupt an -al _ _ ntepn deviaton-aeceumulat-e--beyend- the 1 imia-e---1-hour-and 2 -- houra, respeet-ively-.
.. I I __I
-. 1Cr- - - .. _._. __r _j
- -- i - - -1___
- - __ - - - - - - -I__
"I----
Vv~W_
2/4 .2 .2 and 2/4.2 . HEAT- FLUX AND NTILPY HOT CIMAN-EL FAC-TOGRG nRSn The mits en heat flux and--nuclca. c- a-enueta we densignlmts npa laa~wdns- d min-imum-DIM.- arc not execeeded and 2)--in the event of a LOCA the peak fu: e la+- - - A. Meatr AwAill Ao eee the EGS aenta eirA A4 it0 ° F .
_ IL _C- 1 I_
. I - -- r I I _
pocifiod
. J-in I
Il aly el e dtrie eldcly a pecified in Speei-feat-ns 4.2.2 and 4.2.3. This--periodic surve4illancc is suffmietned-? - t the hot-channel factor--- its are maent-a-ned-provIded-
a. Coent-rol-reds--b-in-a sing -- group-meve-together--wit-h--no
-ndivAdua' red-ins-i an -+/--i -seps f-em-t-hegreupdemand-pos-i.t-en--
b. Geentel red greups are sequeneed with-everlpping groupa as deser.bed-in-speeAefiCat4 o- 3.1.3.6.
BEAVER VALLEY - UNIT 2 B 3/4 2-2 Amendment Change No. 028.
P1OWER-DISTRIBTUIIONLIMITS Provided for Information Only.
BASES 3/ 4I.2-1AXIAL-FLUX-DIFFERENCEJIAFD)_(Continued)
BACKGROUND_(Continued)=
The AED ismonitored on an au O's u p rp q f
=juter, whih ban as AsFDL monitorA> lArm- The computeizpeterminen th m of each of the OPERABLE excore detecootps and-provides-an-alarm-message-immediately itheAFDfortwo-or-more PERABLE-excorechanne1s-isotsidets specified limits.. -IfUitheAF monitor is-outof- serYice.indicated-AFD for-each-OPERABLE-excore
.channel-is-manually-monitoredin-accordance-with-the reqruirements csnrcifiedA in then Li-~-cjnsin Reoui-rement- Man-mim Alt-houianh the PRAnr defines limitns tbhatTmust bh met tn satisfv safetv analysest-ica a opial Consta Offset Control (labl qu ed to control iistrhibtion in da aoay= O -03opr etary1, Pr st ributi-on Control and Load-ollowinqg Procedures ,ILwestinghouse-Electric Corporat in-September 197A4 - - CAQCrecuires that theAED-be
.controlled-within-a-narrow-tolerance-band-around-a-burnup-dependent targettominimize-theyari toLaxial-peakinq acsorsand-axial xenon distribution during unit maneuvers.
TheCAQS nper ting __is t icallsaller and- lies Withinthe RAOC operating-space. Contro~lwithintheAOC!operatincg space constrains-the-variation-oL.axiaLxenon-distributionsandaxialpower distibutions_ RAOCcalcslationsassume-a-wide-ranqeofxenon distributions-and-then-confirm-that-the-resulting-power-distributions sat is fyvthe-requ irement sof the-accident-analyses..
APPLTICARThE SAFFTY ANATUS The AFT) is _ ueasure of the ax L ____rdist_ _ _ution_______ to_
eithr ttto mahalfof he-cep heAD is sensitive to manycoreelated-parameter-suchias-control-bankpositions, core P-ower-le-yvL-axial-burnup__axial xenon-distributionand, to a-lesser extent4.-reactor-coolant-temperature-and-boron-concentration.-
Te alloWed range of the AWn isused in the nclear desinn2pxocess to confixr toperation within these- li g roduces core factors nd alial power distrihutign Pt-safta recruixementsa TheRAOC-methodolog ylSee-W-CAPz12136P=A-iRevision1A. -L Relaxation-of Constant I Of fset ontrol- Surypillanng Technical Specification,-i ebruary,I99AL__establishes a _xenon distribution ibrarywith tentativelvwjtidel JlAE lmitr one-dimensional-axial 1 ; noW has e s -are acceperablt_fhr eQCA-And I s soflow-apccident, and- for iinitial-=Londitions-of-antiipaed
,transients- The-tentative-limits.-are-adiusted-as-necessar-yto-meet the-safety-analysis-recuirements..
BEAVER VATPEY - LUNIT2 B 3L42-2-a ghange Wo. 2-028
Providedfor Information Only.
POWER-DISTRIBUTION-LIMITS BASES 3LA42-1AXIALkELUX-DIFFERENCE-lAFD)f(ContinuedL APPLICABLE-SAEETY.ANALYSESilContinuedI The limits on the Arn ensure that the limits on the Heat Flux-Ho-t rh~nk --
LlidlillC 1 dU f-IVi_.
rLI:-I .^aA V- I rl95
- are not exceeded durina either normal operation-or inthe-evfentof eon-redistribution f-ollowingQower ghanges.. The-limits-on the-AFDal.so-restrict-the ranqe-ofpower distributions-that-are-used-as-initial-conditions in-the-analyses-of
.Condition_2-, ,_or-A.-events. This-ensures-that-the-fuel-cladding intearitvAis-maintained-for-these-postulated-accidents_ _ The-most limiting Condition 4 event with resruect to the AFT) limits is the IQSS The11m st limiting pnditijon 3 event with respect to thie-AFD ements thelgs sof Acclimid t-s limits c lde-nAFD td dCCA Cncon bank-withdrawal-at power, droppedRCCAs, andboron _dilution accidents. Condiion-2-accidenits-simulatedeto-babinfXirm-wi-thir-the AFD imi-ts-areused-to-conf imrthe-adeq-acyof-theiOverpower-AT-and Overtemprature AT tr setpoints.
The Iimits5 onjthe AMF sat-is-fy C!riteri on 2 of 10 CFR SO .36 (c) (2) (ii)~
LCQ The-shapeofL-the-P eilrofile i n theilaxia1-Ai-e.-the yerticall direction-is-largel-yunder-the-control-of-the-operator-through-the manual-operation-of-the-controlbanksor automaticm moti-on-f--ontr-O1 banks. The automatic motion of the control banks is in rqpnnse to
_rat1fro devi ati on PEm riilting ra,-,ar ___ _a___ofthe ghemu-a1L and V ue Control System to change boron concentration or from-po-wex-1 eyel-changes-Signalsare-available___to the oprator-from the-Nuclear InstrumentationSvstem_(NIS] excore _neutron-_detectors (UESAR$.
Chapteri7L). Separate-siqnals-are-taken-from-the-top-and-bottom
.detectors-. TheAFD-is-defined-as-the-difference-in-normalized. flux siqgnals -between the-top and bo~ttom eorreretectors in each detector well,- For convenience, this flux difference is converted to prvde flux difference Units eXprpessed as, a pprr tae and laheled as %-A flux-or_%AIL.
The-AFD-limits-areprovided-in-the-COLR. _Fiqure-B-3/4 I2zlshows typicalRAOC-AFD-limits_ TheAED-limits-forRAOC-do-not-depend-on the-target-flux-difference. However, the-target-flux difference-may heused to minimizeh es-in-the axial e istribution-Violating this LCQ.nn the AFD could oduc una=etabl nosequences ift a Cosin. _2 o4vrs ile-thDitideits Bpecified-limits-3EAVER VAT.TEY - UNIIT 2 A 314 2-2.b -CAW 202
idforInformation Only.
POWERDISTRIBUTIONLIMITS BASES 3L4_2_1__AXIAL-FLUX-DITFERENCEAAFED) (ContinuedL APPLICABLI-IY Thp AFn uirpmentsa reapaIrcable in MODE 1 reater than gr gqmal to o RTP whenthecombinatignnffHTRMAL-OWER anQ rorze-eakincq fla t-ors-ar-e-of-primar-Y-3mpor-tanc-ein-s afey nalyks-s F-or-AE-D-imiLt-s-de-veloped-usinq_.RAOC-me-thodolo y-the value-.o f the_-AED does.not-affect-the-limitinn-ac-cidentconsenuences-withTRHERMAL-POWER 50tRTP-and-for-lower-operatinqcPower-MODES.
S<
ACTI-ON As an altern tyve to restoring the AFnf to withinxitasneified himitp tphe ACTI ea ndts a i oTHERhAL ihER-r-edu ft h5AD< RTP.
This-places-the-cor-ein-a-condi-tion-for-which-the v-al-ue-of-the-AFD-is not imnnrtant in thp annlicahlp safetv analvses A conmnltinn timp of-3-0minutes-is-reasonable,_based-on-operating-experience,-to-reach 501RTP-without-challengingqplant-systems-SURVETTLTANCER1 C TRRMNTS (.)
SR_4-2--1 ThisSuryv ance ifies__thatthe-AE -as-indicated bvytheLNIS ooor nns> .3.p
.tA X LA ' l ,
.........A.
vo TV 4tsc-w;f-k4n U .......- .
no,"e-MW . a F4oA= '
l4m4+
a I 3 hC.
Tho6 . ilrg;
~
- JY 2~ l 11 Mnrog L
intervaL-of-7-days-is-adeQuate-considering-that-the-AF-D-ismonitored by-a-computer-and-any-deviation-from-reauirements-is-alarmed-or-the inncatt i AM i manul Mnit are qudii by heLiensing gnual, BEAVER VALLEY -AJNIT-2 B-=31--- 2-2c -Chancie-No. 2--028
7% 7%
I1 Replace with Insert B.2-1.
A ,- T- n Jt A f .9 Yed jUI & d-
-TYPIGAL INDICATM EDMTV""-Mrel3
=~~e AX AL FL X DI-F- R-RENGE- - VA VR -B B I ERM L FOR --T -BGII Axi-al-F-lux-Dife-ereiLmOts-as-a- ction-oA ED
,THERMAL-"-IERjtoxrRAO-C BEAVER VALLEY - UNIT 2 B 3/4 2-3 .Cb-7uNg I
- - - r . . . I . . . . . . . . . . . . . . . . . .
I I I I I I I I I I
I 90 _ _ACCEPTABLE UNACCEPTALE
'.4
01) OPERATION L OPERATION 3:
0 __ ACCEPT AL 1 1
-1 so OPERATION
___ __ I - _ __111
()
Ed 0) a)
'4.
E0 C:
r4 44 o0 +24 50
_ -312 150) r_ IIT ,(+2,4, 50,)r-40-2ILLUSTRATI-ON-ONLY.
20o _ _ _ 12QAlSFOR -tlll OPERATION 1c 0
- _ _ I L I I II II _ I lI1
-60 -50 -40 -30 -20 -10 0 10 20 30 40 50 60 Flux Difference (Delta I)%
Pr- . ov .i . -. nf _ .r.-_ a T!ionMw nll y.M
-_^.FTT -- - T- - --- T E T
-ETTf\ti lProuidedforInforrnation Only.l r~~~"vj--. _ v _
w F-BASRS FACTOR-01-F 4 1 1-f Centinued4 c ee}-ien4fo-eim nThe--contsereol-n o-S pefifttsons --- - .1.3.5 and 3.1.3.6 ar-e--maintained.
d. The axial power distributizoIexpressed in termef AI FLUX DIFFEREIC is maintained within fie 1-hts.
Thc relaxatian in F as a f m -n. of THE.
ehanges in the radial pewer shape for all permisible red inscrtion lmt. will be mintained -wit.An its} ... its provided eendlti- nsat uh-d-,abo ej-are-ma-intalned.7 When-an--FQ-meesurementme is t l botie-exper-rnent-alc--er-ror--and
t L
A. . A. .. S_.. LCJ S. 4 . A_ _
l t. O ..
~~~.
a_ . vt. . L' . W Vl .
M -- -0
. J _
a-As*_vA_I S.. . S.
5.A-;__
L~ r , J l J JA . tt c .
ere alewne fe a. fu_ eere ma tl-n __t the incorc detceter flux t%
tte and is thet pprepriate ealewaneef- r-mntaetulrng- tdeeranee-.
Thc specified limit ef F1B4 co-sa %aleac e uncertainties w iehar, thrbee eep eperatien will _-s.. lii speecified-in-t-he- GRE-GP r2TINC LIMITC REPORT-F-elrod-bowi-ng-reduces-t-he-vaeue-f-DNB-r-at-i-o.- a-rrgn-has-been mnaintain h i- -chc nlR value ____
u1cd in thclAfpt4 analvcs - __
adhc
- :. __. . I :2 - - -s __-'- .3 I- - - - - - - - n __ - - - I.- - __ - - - - n - 1 -.. ,tI -I-
- m--__-_ - -_ - .. -- v- -_
mayy a
.P I_ I. n . _ _ _ t r -J- - _--___- .V -~f _ - - - -- _-
providc assurance that thc hot channcl factor, FQ-F- .. .- . .. _ I wemow--wiln -
liit. _ts _ - The-F .).y limt f _
-I __ przeih.d in pY+rvided the GORE OrESATIeI LIMITS REGRT-was de {xx
_- xpeele pewe eont-r~ol-mane erevr-~ ful4=-range-ef-bur-nup-eend-it-i-ns--ih-t-he corc.
STET T DV gDUtSTUTW D:iS *4 43NIP-Z B 3/4 2- 4 Amendment-No. 46
I . R.
Pr [ .o for Info71 n Only.
POWER-DIS.TRIBUTONLIMITS Provridedfor Information Only.
31422_HEAT-FLUXHOTCHMNNElIEACTDR--FoIZi BACKGROUND Te prose-of-the-limits-on-the-values-of-F ZL+/-)is_tsloimitthe loca 1i .pei tnpakpower s iqit y.e value ofFnLe-y-aries alonatEhiaxjal heiLqht (Z) of the core F tZL)isdefined-as-the-maximumlo-cal fuel rod-linear-power-density dividedby_5the-average fuel__rod-linear-power-density.,assuming nominal fuel np1let and fieil Yud dimenrionns L Therefore. F-(Z) is a measure of the neak fuel nellet nower within the reactor core nweroperation. the glba powPr distribution is limited by TLCO I 2> 1 "AYTAT.L FT.TTY DFFER.PNC'E TLU (AF)i a qnd TLCO 'A " OADRANT POWER TILT___RATIOCLQPTRL.2_lwhich aredirectly__and _continuously measured-process-variables- These LCOs,_aIonwwjithC-CO 3.-.13_6-L Contol Rod-nstionLimits2 aintainthecore lmits power distributions-on-a-continuous-basis.
F-(Z))varies-withfuel-loadincjpatterns., control-bank-insertion._fuel burnup, and-changes-in-axial wexdistribution_
iu_ -___ lgb the incore detector system.
These measurements are pr _____ with the____ ___________
gpilibhrium cn nitions-Ising-the-measuredxthre-edimensional-power-distributionsJ it possible-t-o-deriv ea-measured alue-forAQ( Z)-__However, because-this value represents-an-ecuilibrium-condition,_it-does-not-include-the variations in-thevalue._ .FQZ)of whichare -resent-durinc nonequilib ations such-as loadfo-llowinq-orpc-er as-censlon RVAVEW VAT.T'F.y - UNIT 2 ~ag ghan-_N o 2-028
-2
PQWER_DISTRIBU=IN_LMIMTS E ProuldedforInformation Only.
BASES 3EL )L-lContinuedL BACKGRQUNDAContinuedi To ;;ccotnt for these anosgible- variations, the ecnxii~ihrium yplueof F%(Z) is adjusted as F~M by an elevation denendent factor that accont slox-thhe calculated worst cag e nrpnsjientnditjons.
.Core-monit.oringqand-control.under-non-eq uilibriumconditionsare accomplished _by_ operatnqthe core within-the limits-of-_the appropriate-LCOs-,,including-the-limits-on-AFD,_QPTRand-controlxrod insext-ion.
APPT-.ITC'A1T-P qAFETT ANAT.YgES This LCOprecldes-core pwe dis tributions-that violate-the followinq fuel-designciteria:
a. _ Durinqca-large-or-smallbreak-loss-ofcoolant-accident lLOCAl,_the-peak-claddinqtemperaturenmust-not-exceed 220-0F _as.specifiedin_10_CFR 50-.46._l9Z4..
b. Durinq__a1os sof force reactor__coolant flow__accident, there-mustb at lea st.95_ probabii--y_a ttheS95k confidence level--"the-95I95-DNB-criterion)-that-the-hot fuel-rod-in-the-coxe-does-not-experience-a-departure-from nucleate-boilinq_(DNB)_condition,
c. Durinp on eijected rod accident. the enera deposition to
_p f _ mtntspecified in ResqulatorySu1ide 1.77, Rev. O. May 1974and
d. The-control rods-mus-tbe-capable-of shuttingdown-the reactor itl inimumrquixre-dSDMwi-th-the-highest-worth
,control rodstuck _fully withdrawnaas-specifiedin lOCLR5 Aspendix 2GDC26.
Limits-onFE(Z)-ensure-that the-value-of-the-initial-total-peakinq factorassumed-inthe-accidentanalysesremains valid. Other criteria-must-also-be-met I e g.-maximum-cladding-oxidation,_maximum hydrogen-qen ration, coolable reomelxY, cool andlonq tr HTowever, the peak claddi eurtr s vi~l ot miting-FQ1 Z)Lsatisfies-Criterion-2-of-1-0CFRR50..36A(c) (2)_(ii).,
BEAVER3VALLEY_ UNIT-2 B_3/4 2_-4a ,Chanqe_No__2_-02_
V'- PoIded - forI , Il.nfo7,m-o DOWER-DISTRIBUTION-LIMITS .ProtddedforInformation Only.
BASES
,31--2A22HE.AT--LUX-HOT-CHANNEL_-EACTQR Z ontinuedl LCO The Heat Flux Hot Channel Factor P~z) shall be limited by the followina-relationshins:
P-t7Zi , r rpn I P1
vKI) fo-r P , q.5 F^(Z) < r -CFQO / g-s
K(ZL for P <_0.5 w rem*Q is the limit at- RP provided in the COTLR.
K(Z)_is-the-normalized-FQiZ)-as-a-func-tion-oLfcore-heeight provided-in.-the-COLR,-and p = THERMALPOEWR The-apual-val-ups of CEQ-a 8 W re given in the COJTi.however. C is no-nall-a-nuer on tho rdler L.f~n Zis a finction that-looks like-the-one-provided in FiqureWB.31.4_2-2
~ r- _ R '
s ^___ J 2z2-is for-illu str&tionfPurplsesonlly.The-COLR-actuaL-unit-specific figures-are-contained-in-theCOLR.
For Relaxed Axial Offset Control oneration. Fn(Z) is approximated by
.EC(Z) ancdL F(Z)__ Thua bothY E(Z) andE Q (Z) -must mee=Jzpreceding limits onFQilL AnF Q(Z))evaluation-reQuires-obtaining-an-incore-flux-map-in-MODE j_
Fromtthe incre fEIxmap resu1Ls 2Web 1bainhe-measureL value. FMMZI)
.ofFQZLThen,-
EC(Z ) M ).Q1
_ Q ___
Where:l-.D0 815_is-af actorthat-accounts f or-fuel-manufacturing tolerancesand flux map-measur-ementunceztainty-_as 4=vncr; f;
-F4 e i n WIA ID- 7,an a-T. - D-2 11L n-vnlt>F-I ort rf Miir1Pl mm- 7rF-Channel Factor Uncertainties." June 1988.
EC(Z) is an excellent apnroximation for T Z)when the reactor is thesteady dwerat staae which the incore flux~mapqaptakeen.
BRAVER VATTE:Y - 2112T02 B 314 2-4b ChLn~qe-No- 9.2 -02=R
1.2 1.0 0.8 N- 0.6 N
0.4 0.2 0.0 0 2 4 6 8 10 12 Core Height (feet)
.TpicalFT-Normalized-Operating-Env-elope ,-K(Zi BEAVERSVALLEY-1JNIL2 B-31-4-2-4c Change-No_- 2=028
POWER-DISTRIBUTIONJLMITS Providedfor Information Only.
BASES IA32L2-HEATPTLUXYHO-TCHANNELYACT REIZLContinuedY LCOJlContinuedl The expression for Fw(Z)=is; 9w(Z) = E C(Z)
w (Z) whe~re: --- Wl(ZLis-a-c-vc-le depDendent-f-unct-ion-that-ac-c.ounts -for_-pomer distribution t-ansients encountered _durinqg norma1 onpyrtion W(Z) is, inc-ludedi- in t-he C'OLR s The is
.calculated-at-eruilibrium-conditions_
The Fo((Z )limits-define-limitinqgv.alues-Ior-core-powerPeakincithat precludes-peak-cladding-temperatures-abov&e22 0Q0Fduringqeither-a la o 1 rk LOCA.
This LCOr
._ res operation within the boundsassu e the s fety analyses calculations are nerformcd in the nore desian nrocess to
-confirmin-that-the_ nor-ecan obe-controldend suuch-jamanner-durinq poperation-that it-can-stavwithin-theeLOCA-F 0 Z) limits - .fPC(Z) cannot-be-emaintained-within-theLCOlimitsx,.reduction-of the coxe power-i s requir d-and iAfrF w(Zcannot--be maintained-wi-thin--the LCO limits. reductiion of the AnFD limits is reauired Note that sufficient-reduct ionof theAFDilimits-wil.1also result-in-a
= _ion of the core_
Yiolatlng the-LC~iits fo-rEZ Dro-duces--unar-cep-table-c-onseauence if-a desiqnbasi-sev-ent-occurs-whil eQ AZ)_isBoutsidelt s-specified APPLICABILITY The FP (Z) imits ust hemaintained-in MnDE 1 to Prevent ore nwer distributions _from-exceedingthelimits_ assumed _inthe_ safetv anjy s-s Applinability in other MQDES-is noreireaoin ebecause there is either insufficient stored ener in the Uel or insufficient energybeinq transf*rr*ed to the *-act~rcoant tn rexpiLr -limit on the distriuonLcor p1wup BEAVER
__ VALLEY - UNITT
__ 2. -3314_2--4d IChanqeNo_ 2-9028
PO sER YMN <;-DISTRIB S is
-QLEILDISTRIBUIIONALIMITS Provided forInfoaton On BASES 314.2.2 HEAT-FLUX-HOT-CHANNEL PACTDR:EZ )(Continuedt ACTIONS a1lReducinqgqHERMALPQOWER-by-2vŽIRTP for-eachil _bvylwhich FQ( )xc-eedsl tsa-imit,_maintains-an-accep able-absolute pnwer_ densiy FQ(Z) is FQ( ) multipliedJy= factr accountinq for-manufacturinqtolerances-and-measurement uncextain ti-es FQ(Z) is-the-measured value-ofEFQt12LThe 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 exte Per id ofme maxi mu o llower l evel int-1-dtrindh ATO ffectpdby qI hs geLete=nnLin s of FQ({Z)a p c
,ediuctionswithin 15mnutps nf the FQ() dme'minotitn, if n e resqAryt cml=wtbt-he-decrea-sed maxinmu allpwabl&
c~z power le-el De-reases inF-QZ) would-allow-increasing-the maximum-allawabl-e-po-wer-lev~el-and--increasinq_p~ow-er Up_2tQ thisxrevised-limit-a.2_Axeduction-oL thePower-Range-Neutron-E-lux- High-Trip Setpoints bvy 2 1% fomeach.-l--by-whichF ( Z)_exceeds i ts limit, i~ cnnnprvat-ive act-ion foprotection -- a=n~h n inhB withunanalyzed power distmibutions__ - The__compl-etio_ tieo _f 732 ho~u rs r:s sufficient-consideringqthe-small=I=ikelihoodoQfa_aev~ere t-ransient-intitis-meerldandt heprec-ding-prompt reduction-inTfHERMAL-POWER-in-accordance-withACTION-al The-maximum-allowabl-e-PowerRangeNeutronFIux- -Hi-gh-rip Setpoints initiallydeterminedby __ACTION a_2 maYv be aff ected-by-subsequent-determinations-of FcI(Z)and-would yire Power _PRnqge Neutron Flx.z , Setpoi rection within 72 ofth (Z)eterinati i
= to onA:_withthdecreas-d_ maxim'nn aI1oahble Rpwer=Range_ Neutrcn l mc - NighTxipSetpoints. Tfecreasres in F (Z) ould allow increasing-thehmaximum-allowableP e r Range-NeutronEluxj{_H h TjprPSgetpoints_
BEAVER-YALLEY - UNIT-2 fchiange No_-0
-V xire~ .1 '. .I - I It w;
-l'T

71"
. T- TIM MM OWER-DISTRIB.UON-LIMITS lProided for Information Only.
BASES 314.2.2HEAT FLUHT5CHANNELEACTORRF0 Zj_(Continued)
ACTIONS-Continued) a,3__Reduction-in-theO-vexpower-ATTrip-Setpoints-Ivalue-ofAL4 by > 1% for pag1Ph bwlicch FQ(Z) x ds its limit. is a cons ervative-actionILorxprotectionagainst-the-conseauences of-sev~er-etransients-with-unanaly edtpower-distributions-The-completi-on-time-ofLJ2-hours-is-sufficient-considering the-small-likelihood-ofa-severe-transient-in-this-time period,_and-the-preceding-prompt-xeduction-inTHERMALRPOWER in acrandce with __CTI9Na .l_ The laximumL allowahl e pyower AT Trp Sep oints initia-Uly determinedJb ACTION a-3_maybeaffectedby subse uentdeterminations-of ( Z)
,and~wuldeeqire Oerpoewe ATripSatpoint-reductions within 72_hours-of-theF C(Z) determination,_jifnecessarY-to comply-with-the-dec-rased-maximum-allowable-overpower-AT the-m ahlpC pSe tpr w TTintg.
a.-AVerification-that E((Z)-and-F (Z)_have..been-restored-to wi-thin-its-limit._byexforming-SR_4_2-.-2.-2_and-SR_4-2-2_3.
prior-to-increasinq-THERMAL-POWER-above-the-l-imit-imposed bY--ACTION-al, ensures that core conditions during pprn at..highpr power ation are r-onsistent with sfetz anseaassumptions.
Acqlipni Lm° ifi ed byr te 1 :at.rquiresAcTIONta.4-to be-performed-whenev-erACTION-a-is-entered-.This-ensures that-SR-42-2-2-and SR 4..2.-3-will be-perfmed-prior-o increasing-THERMAL POWER-aboivethe-limit-ofACTION-a.1, evnwhenCTINiaisexitedprorto erfermingACTION a.4. Performanclp of SR 4.2.2.2 and R 4.2-2-3 are necess arytoa sure-F L-isp evluated- rior-to increasinqTHERMAL P2WER_
a-S IJf ACTL-ONS-a-l--throuqh-a-Aar-e-no-t _met__within_their associated-completiom-times,_the-plant-must-be-placed-in-a MODE-or-condition in-which-theLCO-reauire ents-are-not applicable_ This-isAdone-by-placingqthejplant-in-at-least mODE 2 within r hours-This i osonable-bagpd on oprat~in experience-regardingi amount-of-time-ilattkes oxseah MODE_2_ from-full1poweorsperation-in-an-orderly-manner-and withoutchallenging-plant-systems-.
BEAVER VALLEY - UNIT 2 B 234 2-f ghaage No. 2-028
DQWER-DISTRIB.U=LNLIMITES P[ ProvidedforInformation Only..
- N BASES 3/4.2.2 HEAT-FLUX-HOT-CHANNEL-EACTDR=.Fp ZLlContinuedL ACTIONSAContinuedL b_1_ If-it-is-found-thshemaximumcalculated-va ue ofrjFO that-can-occur-durinc-normal-maneuvers,-F W(Z),_exceeds&its specified-limits,_there-exists-a-Potential-forFQ(Z)_to be.come-excessivel-yhigqh if a-normal-operational-transient o-urs. Reducinc -the-AF-D-limitsby Ž%-_for eachlThv iachfE ee i lii witinletion oIhors.
f 4 restricts t~e-axial flu td~stib on-such
,thateven-if-a-transient-occurred,-core-peaking-factors-are nonexceeded.
The implicit assumption is _thatLil W}Z) values _were recalculated (cosistent with the rp uced.AF li It-q),bt
'PC(7) t-i moms- t-hep recalcula-~ted W(z)1 vaues i#, wouild meet- t-he
_ i i Note that complving with this action (of reducinqAFD liMits) y I so resultin poweireuction.t Hence-the-needfor-ACTIQNSb_2.b-3.andJb-.A_
b..2_A _reduction-of-the__Fower--Range Neutr-onFlux<High -Trip Setpoint s-bvy 2 fIorxeachil16_bywhich _the _maximum allowahlepower is re ced, is qpcrvat-ive action for protection against the rponseeuences of sepver transients with-uana >wer distribut ons. Thecompeton time
_f72-hours-is-sufficient-considering-the-small-likelihood f a Sterestransjntni mime-period-and-the-preCeding prompt-reduction-inETHERMAL-POWER-as-a-result-of xeducinq AEn limitsi n-accardancewith-ACTIONMb.1.
b-3__Reduc tUon-in-theOv-erpawer-AT-Trip-Setpointsi(value-ofK4L hy > 1n for each 1 by which t iR reduced. is-a-conservativeaaction LorProtection-aainst the-cons equencesof -s-evere-transientswith-unanalYz ed-dpower i -f str; hzt- nns r Trlh gmfmy I nn tim- n'ef 72 hris 4a I suf ficient considering the smelI likeihoo of i; setvere trainsient in this time pprind, Cn theppreced'n pomt reucio in M]EIRMAT POWR a q ; -re-sul t of r dui~A)
.1imA- I d _b-N7C-IID BEAVERYALLEY J-UNIT-2 gliange-No 2--02.8
[;_w
_ .- -- .:_!
'.-:- - 7 WIProvddedfor Information Only.
P-QWER DlSl=EU-T-LON-LIMITjS I
i BASES 314_22_ T T HT_-CHANN-ET, F'ACDR--F )Continued)
ACT ONS-Continuedl b.A-Verif ication that been rest red t_
within_ ts I it-. by pgSP, 4.2-2.2 and-SR 4.2.2.3 or t i T T.aemaximum loalpor HLlmp~sdlyAT NLhLesures--that core-conditions-during-operation-at-higher-power-levels-and futur-e-opeeatio-'a are consisten withsafetYanalyses assumptions-Action b is modifiebyd Note thALTeTuOresb_4Iflto>_Ai r2 he nerformed whenpvpr ACTTON h iR entered. Thi R ensres tat SR 4.2-2-2 and SR 4.22-3wilbe erormprior to lncreasing -THERMALL-PWERaboEvaethe-limi-tof -ACTION-b-t ev-en-_when__AC 10N b is exi ed-Pr r-.to performing ACTION b.A4_ P-rformance of-SR P42.2 2 andSR_22_3_are neel arvp slqureF) Qz is rlrspe rpl erprior to increasingTHERMAL-POWER-hb5 Tf ACDTIONS bh1 thrugh h -4 are not met within thbir assoiatedcompletion times,-the n st 1- Placed in a MnTnP-evor eyndit-inn i n wh icrh t-he= T.rO rim~eyjmmgnnfc mrge -not i
an-rl nlicahle This is done 'huv nDe-ainn t-henlant- in at- lest-MOD112-within-6-hours.
This-c-mpletion-timei s__reasonable-based-on opexatin xperience regadinga the amoun f L.time it takes to reach MODE 2 from f ll powpr oPerationjarl nrdltry manner and withqut cbaU-ng,,plant -yQt-m SURVEILLANCEREQUIREMENTS SR-_ 2_2-2 TheProvisions-of-Specification4_D-__ar-e-not-applicable.because-alI the follb veillancs et Prfoed hp in MODE-1 SR 4.2.2.2 and SR 4.2 2.3 are modifiedry ote 3. The Note applies duringthe firstpwer a cenlon after a refueligc. -Its tatesthat THERMAL POWER-may-be-increased-until-an-equilibrium-power.level-has beenachiev-ed-at-which p r edistribution-maPcan-be-obtained.
This allowancei s-modif ied. however. by-one-of-the-surveillance interval conditions-,-A, 4-2-2-2.b-andA42223-bJ thatxreruires yerifcation-thatFc(Z) anF W(Z)_are within their-specified-limits after *-IDwer rise of-moxore thanl RTP overthe TH POWER at which they were last verified to be within specified limit.s BEAVER.VALLEY - UNIT-2 B. 31__42-4Ah Chanqe_-No__-02,8
POWERDISTRIBUTIONLIMITS .Provdedfor Inforaton Only.l BASES 4 ZE(z)ContinuedL SURIEILLANCE-REQUIREM NS (Continuedi.
BcausEQ(Z)ad o t havepreiously been measured-in this reloard c*re, there i another -urvepi1ance interval condition, i.e., 4. 2 2.2.a and 4-.2.2.3.a. alicale o for rea cres, that rgurgs=et-e= iati nn Of=Jas ramebers feocaeexceading 7s5k RTE-T sensurthatorne etermination o (Z) andLEQ(Z) are made-at-a lower-power-levJelat-which-aderuate-mar in isavyailable-before-going to-lO0%-RTP .__Also ,this-surveillance-interv-alcondition._together with-the-surveill ance-intervalcondition-requiring-verification-of F=(Z)_andFw(Z))followinq-a-power-inc-ease-ofmore-than 1O0,_ensures t-hat- thev Ire verified as soon as RTP (or anv other level for exte____________ _____t nben of th-ee surveillance interval convitions, it- 1/2sossble to increase ower to RTP and operate for 31 days without verification of FQ(Z)_ancIEm(Z)
The surypi I ceinterval condcitionjis--no-t-intendedt-p r.-uire verification-of-the s eparameters-aft er-every-lO0Aincrease-in-power levellabove the-last verification.Itonlyrequiresverification after-a-power-levelis-achieved-for-extended-operation.-thatiis 10?6 hi her- than a power at which GaWSlfiLiSrC SR 4.2.2.2 Verification thatEc (Z) is__within-its-specified-limits-involyes increasincE m(Z)_to allow for-manufacturinq tolerance-and-measurement unce-tainties-in-ordex-to-obtainr Ec(Z)_Z) SpecificallyllF-m(Z))is-the meas-ur-ed-value-of -F~ ) ob~tained__fnmm_,inconeEuxap rmesults-and EC(Z) =F m(Z) 1.0815. F is then Omparep to its Recified limitB.
Th_ li_ ___ Q(Z)Z )1 iacoropared var es inv Iery wit Ipwe abo-ve_5D-%-RTP-and-di rpcl uih-a-func-tion-call-ed-YIZL-iprov-ided-in theCOLR.
Performing-this-surveillance-in-MODE-1-prior-to-exceeding -7. 5 _RTP ensures-that-theFQ(Z))Limit is-met-whenRTP is-achiev-ed, because peakin- engrllyAIcreae BEAVER2VALLEY - UNIT_2 B_3-/4 2-4Ai Change-No._2=028
L; I .- 1 1 - -- -- T T 6 a_
POWER-DISTRIBUTION-LIMITS Provided for Information Only.
BASES 314 .2.2 HRAT -ELUX-HOT-CHANNEL-F-AC-TR=FQ=(Z SURVEILLANCE-REQUIREMENTh Coninued)
ILfTHERNAL POWER-has-been-increased-bYv 2 0-iRTP-since-the-last determination of Q 0 (Z),another gypluation othisfactoriarequire 12-hours-after-achievincgequilibrium-conditions-at-this-hi hrpower lex-el toens.ure-that_ c(Z) Zvalues-are-beinq reducd-suffi-iently with-pomr-ncrease t stfwithintheLCOlimits).
Thesunrvei11ance-interalof__31 EEEPDis-adequatet9tomonitorjthe chan-q owhrnt _ u oecause- suchc
_re sloW an wll controlled when the ppant is in aprdanz with theTe hiical.snecifications (TS).
SR_--2t.23.2 Thenucleardesignprocess_ includescalculationsperformed to determine-t~hat the core can be P e-ihnth (Z iis Because flux -ma -_sasr_ taken in stead-y state conditions. ahe variations in power distrihution rts in-g from norm3L operational mineuvers are -not.presnt in the flM TesfL variations a wevi1p_er, conservatively c lculated byonsidering aidersange of-uni-t-maneuvers-in-norrnaLoperation. The-maximum-peakingqfactor increa e-serteady-stat-eaiescalculated-as-a-functionoLfcore elevation._Z,_is-called-WAZL__Mull.tiplying-the-measured-total-peakinq factorw--F(z).-bv-W(Z)-qives-the-maximum-FQ.(Z)-calculated-to-occur-in normaLoperation. PQ(Z).
The-limit-with-which F(Z))=iscompared-va ies-inversely withpower aboovel_ RTP-and-diratlY zi the-functionKIZ)-Provided in-the COLR..
The w(z) curve is provided in the COLR for discrete core elevations.
FluxFiancmpdata re- ty-pically taken -fopr 30Qto- oeeevto man cl;td___~~eW--ya~
E (Z)evaluations are not applicable folr the follo ing axial core r:e-qions measu~rpedi-j-percent of cor heigqht:
a. Lower-core-region, from-0-to-15-tinclusive-and
b. Upper-core-reqion.-from-85-to 100i.-inclusive..
BEAVERAVALLEY AB 2 AagUNIT B-3.14_2-=4 i3 .Chan~geNs-Q28
POWER-DISMRIBUTION-LIMITS PrvidedforInformation Only.
BASES 3I4 .2EATFL-UXHOTCHANNEL-FACTQRz..EQI-2 SARVEIL.LANCE-REQU1REMEN TS oninued)T hecause of the _p> oah lIi t hat-thesepreg on __ ore limithgnq-in-the-safet-yanaLy-s-and-because-of-the-dif-i-cuL.vof akinajgpxerecisemeas em entinthes e__reqns This-sur-ei11ance-has-beenmodified-bvyNote_(4)_that-specifies-in part "Tf measurements..indicate__that eth maximiu ver _z tf JF£(Z)IK(Z)Lhasincreased ... T.
This-statement-refers-to-thef act
.that-both-Ec(Z) and-K-are functions-of-the-axiaL-heightL.At-each applicable-core-elevation-the-ratio-ofF(ZZ ) LKR(Z)Liscalculated-to determine the mgax um ratio (maximum ovef this maximumratio h i ncreasd Jrncethe last set of na _ _atons, then Noth may reui-ire tha ___ _________
_ _ I(Z)_
ezlPeyna4ai, a~~tion ofthexrsine rquL o account-for-anyincr-ease-tno M(Z))that-mayoccur-and-cause~The F~q limit toheexc~eedpd before the reurcFIj3a1atjlon- vex Lf-the-two-most-recent-F iZ)-evaluations-show-an-increase-in-the expression-maximum-over zLof i (Z)-/ZK) L(Z ,_it-is-requiredto-mee-t theEg(Z) limit. ith the L (z) eAspd h qreater of a fart-or of 10 r byan pprpiate-ac rpecifilieCQLR LSee__WCAP 02.lfl A,__RQ CZE-l. "Re-la i~onofConstant-_Axial Off set Control-FMSurveillance TechnicaLSpecitication,2fEebruarvy,.
99~4) or toC evalua;te.F(Z) more faaietv ea 7 VFPD) Thaee alternativereeuirements-prevent.F(Z)_fromexceedinqits._limit_for anywsignificantperio of timewithountdetection-Perf ormin-q-he-s~ur-veilance -in-MOD~E-L-ri) r to-exceedinqSE91RTP ensures__thattheFQ(ZL-limit-isjnmet-whenrRTP-is-achieved, becaus-e peaking_facto rs_reenerallyvdpreaseda power level is increased, En(Z)Iiseyrified at nowgr levelcs > lnt RTP above the THRRMAT. POWPFR of its laSt eifcto. 2hor after achPyiev e 1 ibriuim gonditions-toenasure-thatF I iswithinl ts 1imita thigher power levels.
REAVER VALLTLFUN-TT N
-Chanqce_Ng___2=02n
P-OWER-DJSTRIBU=ON 1 IMIIS lProvded for Information Only.
BASES 3L4-2.2HEAT-LUXHOTCHANNEL-ACTOR-.o0 ZI SURVEILLANCE-REQUIREMENTS (Cbninued)
The surveillance *intervral of 31 EPP iuate to on ita the
,-.h n" n E=f n ow r Ar14 a t hi if- " .,4t-h ev-for9 ,vr,'IiI N Trho 9lrm P 1 1 ;nnf- m=Av be-donemorefrecruentlyvif requiredby- the-resultsof_ FQ1(ZLI evaluations.
The-surveillance-interrvalof-31-EFPD-is-adeTuate-to-monitor-the change-of-powe-r-distribution-because-such-a-change-is-sufficiently slow, when the niant is onerated in accordance with the TS to 2
nrecludeadverse neakina factors between 31 day surveillances 3/4t2t3_NUCLEAR-ENTHALPYRISEHOQTCHANNEL-FACTORFT4 BACKGROUND The r of thi Cabisiiw totL eni at-arv r-oixt in thecoresnthat-thefuel-esicr" criaeot exceeded-and-the-accident-analvs.is-assumintions remain-valid ---The desiqn imt-son-localJpellet)and-inteqratedLfueLrod-peapwer densitv-are-expressed.-in-terms-of-hot.-channel-factors. Control-of the coxep rdistribution-withresp cttothese-factorsensures that local conditions in the fuel rods-and coolant channel- do-not Challenge core in ritt yat, an,, Iatinon] durin geithernormal oerati postulated accident ayzed in the safty anaLyses.
EAH is-defined-as-the-ratio-offthe-integral-hof-the-linearpower thr fuel alg rod with the highest integower to the averaqe jmmgrater fuel rodOwer. Thee forea of the maximum total1powepr dired i n a fuel rod-N EAT4is-sensitiv-et-o-fue2Lladin -pat-terns, bank-insertion,-and-fueI hIurnuIn N_ t-vnic-allv in craQse w it-h corint-vrol hankl insert-ionn and aa 1Y dereases-with uelubnrnupD FNH _ isnotdirectLvymeasurabhlebutis inferred from__apower 3distribution.-map-obtained-with-the-movable-Ancore-detector-system.
Spe;if icalLy-the-results-of-the-thr-ee-dimensionaLp-ower-distr-ibution N
map.a an ed-bv acomputetotdeternine-NH his-factor-is calculated-at-least-everyv31EFPD-.JHowever-,durinqgpower-operation..
-the-q3-obal-po-wer-distr-ibution is-mo-nitored-by LCO-3-2--1__NEXIAL-F-LUX DIFFERENCE-(AFD)2 L.andLCO_3.2.A "QUADRANT-POWER-TILT-RATIO_(QPITR) 2' which-address-directl-yand-continuousl-ymeasured-process variabLes-j B1.AVFPR VAT.T.'F. - UlNIT 2 ~n e 2 X 028
-Changce-No.-2 2
P-OWER-D-ISTRIBUTION-LIMIiS Ll Provided for Information Only. I BASES 3/4.2.3 NUCLEAR ENTHALPY RTSE HOT CHANNEL FACTOR F4,IC ntinued)
BACKGRO.UND-AContinued)
TheCOLR-provide speakingcactorxJimit sthat-ensure-that-the-de sign basin-valu nLf the-depa rture-from nicleate-boiinq (vDNB) is met for norm eration, overational transients, -anvtxansient con"
aricinci from eve-nt-sF of moderate fr-mipncYv. The 1111 ciexin hbcsiq
-ensur~es th na N sl o sunt e-bblt mos~t- Iimjiting fuel-rod-is-at-l eas-t-9-5at-a-95-Wconfidence-level. This-is-met-by 2imit.ngthemininmumDNBR-to1_the 9-5I95_DNB criterion-of__1-22-fEor tvrical and-thimble-cells usinattheWRB-2M-Critical Heat Flux ICHF) morclsiuz__nd_1_23-tox-the-l-pi-caL-cell-and_1-22--fox-the-thimb-le
_1using the WB-1 CHF correlation- All DNB ijnmitedtransient events are-assumed-to begin-with-anF-NE -value-thaLtsatisfies-the LCO-recuirements-operation oautside-the LCO limits-mav_-p-roduce unaccentab-le cons eo if a DNR liitiniq event I rsU. The nNB dai a ensures thae there i overhhtatim oeLiftitful tno tha rult- in nosgihlr cladeiina nerforation with the release of fission nroeiuict-s to the-reactor-coolant-APPLICABLESAFETZYLANALYSES Limits=_ on F prec11de, corspon r dJtriutJons that Pxceed the Th1owinq fuel desian lJi it-s ahThemus brat lastt 95. probability zat__the_ 95%t
,confidence-level-tthe-95I955DNB-cri.terion)_that-the-hottest fuel-rod-in-the-core-does-not-experience-a-DNB-condition,.
h.nirngail ar-e or small break Igqq of ant accident L°CA) neak Ciaddingtemperature yCT) mist not exceed 2200 0 F, a. ,ecified in1QCPR5O.46974
c. Duringqan-e-ected-rod-accident,_the-enerq-ydeposition-to the-Xuel._ must not-exceed2 80_calLqmas-specified-in RegulatorYvGuide 21.77.Rev-_O,_MaY-9-74-,and d= Fluel design limits r iegirped by 10 CFR_50, Appendix A...
GD-C-26 for threcn dctroli iw rodq mum'st-he-capah of s hut tincadown the r-eactor-with-a inimure aui red STi with the-hiahestworthontthdrawn BEAVERYVALLEY_ _gUNIJ-2 B3 3/A-LZ -4M -Chae-Ns__2--02-8
PO , :X vicedfo E o- I S T,_ R U- I. rT LI- -nformation On"I PQWER-D-IS-TRIBU=IN -LIMITS ProvidedforInformation Only.
BASES 3142.3M~rV.R RTHA-PyPTR HO MN L F iniued)L APPLICABLESAFETYANALYSESAContinuedL For-transients-that-mavbeDNB-limited,_theReactor-Coolant-System flomwand-E Nare the coreramersof most importance. The limits on F.NAensure that-the DNB ~desicrn basisA ss-me tfor normnaloperation-joner-a-tiomit andi.any tran n t sais fr Ptensof moeerate fremiencv_ The NB tIesiayn hasis ensuires the nronhailitv that DNB will-no~toccur-on-the-mo-st-imiinqfuel-rod-is-aLtLeast-951 at-a-95}_confidence l evel._ huismet-bylimitinq,_theminimumDNBR to-the9-5L95_DNB criterion of 13.22-orf ical-and-thimblfecens usingqtheWRB-2MCHF-correlation ,and&1.23-for-the-ty-picaL-cell-and 1.22 for the thimble cell usina the WRR-1 CHF correlation .
L These 1AAGI; ya3uespr-ovide-ahihqh-degree-of-assurance-that-the-hottest-fuel..rod inthecre rjogg-n Pri ere a The allowable F N limit increases with decreaaixgpoWer level- This functionality-in-FAH - s.ncluded-in-the-analyses-thatprzovide-the Reactor Core Safetv Limits (STqs of SL 2.1 l Therefore. DNm events in which the core limits are modeled imnlicitlv use this variable vralu1e of FN in the analuvses Likewise-.alltransients-that-mav-be
- ~Lif-t AU^*
n - ^-^*^^-
N DNB limited-are-assumed-to-begin-with-an-initial-AH ^_as-a-function gopower lev~eldefined-b-ythe-COLR limit-equation.
The LOCA gafetan lysip indirectly-modeq NAH a a init-H>=1 OuliIC tC_ The Nuclear Heat Flux Not Channel Factor. FE(Z). and the axial-peakingqfactors-are-also-indirectl ymodeled-in-the-LOCA-safety analys-es-that verify-the-a-ceptability-of-the-resulting-peak-claddinc temperature.
The fuel iprotected i art h Technical Specifications which en i -i inta conn 8it-i ong-as-sum, in the s fry nd accident, analsesrsmanin vanlid The followina T.LCos ensure this: T.CO 'A1.3.
L'ControlRod nsertion-LimitsThnLCO_3_32-.l,_AX IA-FLUX DIFEERENCE (AFD)" LOT.) .92. INeat Flux H-tot-i Channel Factor FV(7)ItLCO 3.2.3
. __ .S... .... _r __
"Nuclear aRise Hot Channel FIno 3_2. a ,
,QUADRANT-pO1Rg TILT gTTo(QPTR)
BEAVERVALLEY - UN1T_2 B. 314_2--4n 9M-an199=X-Q--2 -- O2 B
PxOWE R-D-TSTRIBUTTON-LIMI-TS [ ProvidedforInformation Only.
I BASES 14RContinue4l APPLICABLE-SAFETYJANALYSES-AContinued).
H--anrd P0 (Z) are measured neriodicall usinqgthemovahle i-core Pd;Eec tor-s vs em-__Me as-u ements-ar-e- qene r 1ytaken-with-the-cor-e-at..
or-near.steady-state-conditions. Core-monitorin~g.andcontrol under
.transient conditions (Condition l1events)-are accomplished__by operating-the-core-within-the-limits-of-theLCOs-on-AFD, QPTR ,and BankIlnseritionLimits-FAH_satisfiesCriterion-2-of _10CER_5-0-36-c6(2)_i LCO E NH shall be maintainedwithintheilimitsofthe-relationshiP prnyided in the COL R.
The.N -H limit-identifies-the-coolant-flow-channelwitth-the-maximum enthalp riseThischannel-hasithe-highest-probabilty for a DNB.
The__limitinq_1alueofEF dcribedbvyth eequation-ontainedlin theCOLR_ aadesiqn radiaIakinqfactor-Anuclear-enthalpyurise hot-channel-factor)-used-in-the-unit-safety-analyses.
A power muliplication factor in this e-additional mar-gin for hi=her radipl Deakina from redued thermal feedback and areatetr = ___
cornt-rol _ _t t_.
-rod i nser-t-io _ n at- low m_ _lezvemlsc nowemr .= rrTh 1 imi tina f F1Nis alowed to innyeaqe bj vatwlue for Prspecfifed
,in the__=_ifor et-ve- tionR _e T. ._.-Pwp APPLICABTLIY The_FH Jimimust-be mainta-inedDEn EL to lu _r_
disatmiutions F mexceedinthefueldesign limits ft DRan PCT.
APplicabil tin-other-MODESisnot-required-because-there-is-either insufficient eenerrqY-y -the-fuelor..insiuff-icient-enerqY-being transferred-to-the-coolant-to reauire-a-limit-on-the-distribution-of core-Pqwer<< Sthe desiqn-bases-events-that-are-sensitixve tOENH in-other-MODES-MODES 2-through-5)-have-sinficant-marqin-to the D1BR limit andLtherefore, there is p-neee to restrict = N these-MODES-REAVERVALLY-UNIT-2
L I,_ .. , .6i- ; _ P !
POWER-DISTRIBUiIONJLIMLTS Provided for Information Only.]
BASES P.TAP~
314.2.3-MM-PAR ~TE ~
O ~ Tntinuned)
ACTIONS
a. With FNM ex(OWpdin-P it .q_ imit, tnufl pn R to s_-5-0°6RTP and-reduce--the-P-ower-Ranqe-Ne-ut-ron-Flux-- HLqh TripsSetpointsto <55°sRTPPi-n-accordance-withAC TION-a, ReducincLTRTPto s 5% TP inccrease-s-the DNB-margin-and-does not likely ls heDB ttob iltd nse-y sQ-tate ooerat-ion The rdriirct-icn in trin setnoints ensures that continuinq peratiOn remairin at an acqpt- Iae low power Ievelwith-ade uateDNBRmarqin- The allowed completion timeof-2ho-urt to-reduce-THERMAL-POWER-prDvides an-acceptable--time to-reach-theer uired-powerzlevelfLrom full-power-operAtioin-without-allowing-the-plant-to-remain in-an-unacceptable _condition-for-an-extended-perriod-of time_ _ The-allowed completion-time-of-4_hours-torese-t-the trip setpnints r_ oanizes that, once power is rpduced. the saf tyoanala da therp in no iraent nedr to reduie the trin setnoints ,.
Th; cm I 11 l ._ 10 sa M sens i ti-e oerati th myina ertently tripDte-Reactor PxotectionBSvstem-
b. Once-the-power-leve-ehas-been reduced-to_<_50°s._RTP-per ACTION-a,-an-incorertluxrmap_[SR 4-2.3.1) mustbe-obtained and-the-measured-vajupof-EAl N v-eri-fiednot-to-exceed-the allowed limit at the -lower nower l~evel _The-uniLtis pmvided-22-additional-hours-t Pefoxthistaskoyserand above-the-2-hours-allowed-bv ACTION-a, ._The-comple-tion-time of 2A4ho-ursis acceptablehbecause-of-the-increase-in-the DNmargin, which is obta at _ In rower levels, and the- n probability of havin. -aDN lnimiting-event within t-his 24
) 'hour -neriod. Additionallv -nprat-inr Pynperienre has indicat mhat-this-completi-on-time-isssuffici nt-to pbtain___the -incgze flx _map, p r omte erequir-ed canculationsL-and-evaluate AH Shoul-da-satisfactry.Anz coxe-map-notbe hcomplee within-the reuired-Completion Time, the-plant-must-be-placed-in-a-mode-in-which-the-LCO re-uirements-are-not-applicable. This-is-done-byr-educing RTP to less than St gm thepi~t in at 1east MODE 2. Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . The allowed Completion Timef 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ig reanable basQ on operat ngexperience regarig5he5mereuired-to-reach-MODE_2 fxom-fullpower
-onditions-in-an-orderlvymannerand-without-chaUlenginq plant stems-BEAVER VALLEY - UhNTT 2
P r I. ST -RIU_:N-e- fr1 L Informtion On. '%
POWERDISTRIBUTIONLIMITS ; PrutddforlInformationOnly.
BASES 314.2.3 NU-CLEAR-ENT'AILPYrD RTSR HOT SHThTVLEA FACntiuedl ,
ACTIONS&JContinued)-
c. Identification aind rrmctLion of the cause of ansAutof 2imi-tcondition-and erifi.cation-thatFJ NH_ swithin-its specified-limits-prior-to-increasing-THERMAL-POWER-after-an out-of-limit-occurrenceensures-that-the-cause-that-led-to teF NH exceeai imt iq corrected. and that subs eauent-operatioi-pDroceeds-within-the LCO&limit. This N
ac-tion-demons-txate~s -t~ha te-l A MitiW-S-wlthin the LCO limits nrior to exceedina 50% RTP _\4 ne-in csw vi n nr;-- nrv tn exceedina 7596 RTP. and -within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months af ter THERMAL-POWER is >_951_RTM.
SURVEILLANCE-REQUIREMENTS SR 42 -3.1 The-value-ofFNH N- 1Sdetermine dbyusinq the-moyable-incore det ector svstem to obtain a flux distribution man A data reduction comnuter prsogram-then-calculates-themaximum-value-of-E AH fromlthe-measured f1uxdiistributi ons Afterxeachrae-fueling.-F NA must-be-de-termined-in-MODE-1--prior-to xteedinq 75%TP. 9f This rp ent e itsare met actthe-bpginning-of each-f uecc The--3L.EFPD-surveillance tterYal-isaccePtable-because-thLe-power distribution-changes-relativelyvslowly-over-this-amount-of fuel burnup__Accordinqly._this-surveillance-intervaL-is-shor-tenoughWthat N
the_ AH limit-cannot-be-exceeded for-any-significant-per-od-of pp4ra2io.3 SR 4.273.2 The-measured-value-of-F H_must be-multiplied-by l.O4Ato-account-Ior measurement-uncertaint-Ybefore-makiny-compgariskontoitheE 2mitt
_AH BEAVER VALLE - NIT 2 _Chaq No. g02
ProvidedforInformationOnly.
POWER DISTRIBUTION LIMITS BASES 3/4.2.4 OUADRANT POWER TILT RATIO (OPTR)
BACKGROUND The Quadrant Power Tilt Ratio 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. The QPTR is routinely determined using the power range channel input which is part of the power range nuclear instrumentation (NI). The power range channel provides a protection function and has operability requirements in LCO 3.3.1. While part of the NI channel, the power range channel input to QPTR functions independently of the power range channel in monitoring radial power distribution. For this reason, if the power range channel output is inoperable, the power range channel input to QPTR may be unaffected and capable of monitoring for the QPTR.
The power density at any point in the core must be limited so that the fuel design criteria are maintained. Together, LCO 3.2.1, "AXIAL FLUX DIFFERENCE (AFD)," LCO 3.2.4, -QtAT)PA1 T ER TIT.T PATIO (QPTR), "and LCO 3.1.3.6, "Control Rod 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 design criteria and that the power distribution remains within the bounds used in the safety analyses.
APPLICABLE SAFETY ANALYSES This LCO precludes core power distributions that violate the following fuel design criteria:
a. During a large break loss of coolant accident, the peak cladding temperature must not exceed 22000 F in accordance with 10 CFR 50.46;
b. During a loss of forced reactor coolant flow accident, there must be at least 95 percent probability at the 95 percent confidence level (the 95/95 departure from nucleate boiling (DNB) criterion) that the hot fuel rod in the core does not experience a DNB condition;
c. During an ejected rod accident, the fission energy input to the fuel must not exceed 280 cal/gm in accordance with the indicated failure threshold from the TREAT results (UFSAR 15.4.8), and BEAVER VALLEY - UNIT 2 B 3/4 2-5 AmendmentChange No. 42=028 I
Providedfor Information Only.
3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued)
Table 3.3-1 Action 2 has been modified by two notes. Note (4) allows placing the inoperable channel in the bypass condition for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months while performing: a) routine surveillance testing of other channels, and b) setpoint adjustments of other channels when required to reduce the setpoint in accordance with other technical specifications. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months time limit is justified in accordance with WCAP-10271-P-A, Supplement 2, Revision 1, June 1990. Note (5) only requires SR 4.2.4 to be performed if a Power Range High Neutron Flux channel input to QPTR becomes inoperable. Failure of a component in the Power Range High Neutron Flux channel which renders the High Neutron Flux trip function inoperable may not affect the capability to monitor QPTR. As such, determining QPTR using the movable incore detectors once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months may not be necessary.
The following discussion pertains to Table 3.3-3, Functional Units 6.b and 6.c and the associated ACTION 34. The degraded voltage protection instrumentation system will automatically initiate the separation of the offsite power sources from the emergency buses.
This action results in an automatic diesel generator start signal being generated as a direct result of the supply breakers opening between the normal and emergency buses. The failure of the degraded voltage protection system results in a loss of one of the automatic start signals for the diesel generator. Therefore, the ACTION statement requires the affected diesel generator to be declared inoperable if the required actions cannot be met within the specified time period.
The instrumentation functions that receive input from neutron detectors are modified by a note stating that neutron detectors are excluded from the CHANNEL CALIBRATION. The CHANNEL CALIBRATION for the power range neutron detectors consists of a normalization of the detectors based on a power calorimetric and flux map performed above 155-0f RATED THERMAL POWER. The power range neutron detector CHANNEL CALIBRATION is performed every 18 months but is not required for entry into MODE 2 or 1 on unit startup because the unit must be in at least MODE 1 to perform the test. The neutron detector CHANNEL CALIBRATION for the source range and intermediate range detectors consists of obtaining detector characteristics and performing an engineering evaluation of those characteristics. The intermediate range neutron detector CHANNEL CALIBRATION is performed every 18 months but is not required for entry into MODE 2 on unit startup because the unit must be in at least MODE 2 to perform the test. The source range neutron detector CHANNEL CALIBRATION is performed every 18 months but is not required for entry into MODE 2 or 3 on unit BEAVER VALLEY - UNIT 2 B 3/4 3-4 Amendment-Changq No. _22-_A_
Attachment C-1 Beaver Valley Power Station, Unit No. 1 Proposed Licensing Requirements Manual Changes License Amendment Request No. 310 The following is a list of the affected pages:
BVPS-1 Provided for Information Only.
LICENSING REQUIREMENTS MANUAL (LRM)
INDEX ADMiNISTRATIVE rlONTROLT SECFTION PAGEF.
A.1 Procedure Review and Approval A-1 A.2 Record Retention A-2 TABLES TITLE PAGE Table 3.1-1 Reactor Trip System Instrumentation Response Times 3.1-2 Table 3.1-l.a Combined Overtemperature Delta-T and Overpower Delta-T Response Times 3.1-3 Table 3.2-1 Engineered Safety Features Response Times 3.2-2 Table 3.3-1 Meteorological Monitoring Instrumentation 3.3-2 Table 3.3-2 Meteorological Monitoring Instru-mentation Surveillance Requirements 3.3-3 Table 3.6-1 Seismic Monitoring Instrumentation 3.6-2 Table 3.6-2 Seismic Monitoring Instrumentation Surveillance Requirements 3.6-3 Table 3.9-1 Reactor Trip System Instrumentation Trip Setpoints 3.9-2 Table 3.9-2 Engineered Safety Feature Actuation System Instrumentation Trip Setpoints 3.9-4 Table 3.1 1-1 Explosive Gas Monitoring Instrumentation 3.11-2 Table 3.11-2 Explosive Gas Monitoring Instrumentation Surveillance Requirements 3.11-3 able.1 -1 )YZ) Values =
4.1A-l_
Table_;2- FQ(Z)YPenaltyFactor =
A.412 Table 4.2-1 Heatup Curve Data Points for 22 EFPY (TS 3.4.9.1) 4.2-9 Table 4.2-2 Cooldown Curve Data Points for 22 EFPY (TS 3.4.9.1) 4.2-10 Table 4.2-3 Overpressure Protection System (OPPS) Setpoints (TS 3.4.9.3) 4.2-12 Table 4.2-4 Calculation of Chemistry Factors Using Surveillance Capsule Data 4.2-13 Table 4.2-4a Calculation of Chemistry Factors (Based on St. Lucie and Fort Calhoun Surveillance Capsule Data) 4.2-14 Table 4.2-4b St. Lucie and Fort Calhoun Surveillance Weld Data 4.2-15 iv Revision 35 I
_- -,=s
, _" . I -" - eS - '..- .; .. : a BVPS-1 Provided for Information Only.
LICENSING REQUIREMENTS MANUAL (LRM)
INDEX TABLES TITLE PAGE Table 4.2-5 Reactor Vessel Beitline Material Properties 4.2-16 Table 4.2-6 Summary of Adjusted Reference Temperatures (ARTs) for 22 EFPY 4.2-17 Table 4.2-7 Calculation of Adjusted Reference Temperatures (ARTs) for 22 EFPY 4.2-18 Table 4.2-8 Reactor Vessel Toughness Data (Unirradiated) 4.2-19 Table 4.2-9 RTpr Calculation for Beltline Region Material at EOL (28 EFPY) 4.2-20 Table 4.2-10 RTvr Calculation for Beltline Region Material at Life Extension (45 EFPY) 4.2-21 Table 5.1-1 Containment Penetrations 5.1-1 Table 6.5-1 Snubber Visual Inspection Interval 6.5-6 Table 8.3-1 Reactor Coolant System Chemistry Limits 8.3-2 Table 8.3-2 Reactor Coolant System Chemistry Limits Surveillance Requirements 8.3-3 FIGURES TITLE PAGE 4.1-1 Control Rod Insertion Limits 4.1-3 4.1-2 Axial Flux Difference Limits as a Function of Percent-of Rated Thermal Powerfor-RAOC 4.1-4 4.1-3 FT Normalized Operating Envelope, K(Z) 4.1-5 4.1-4 Dp1etqdMaXifeUtc AA F i)Rvs-Axi. ..eight-During
-ANOperx tie n 4.1-6 4.1-5 Reactor Core Safety Limit Three Loop Operation (Technical Specification Safety Limit 2.1.1) 4.1-9 4.2-1 Reactor Coolant System Heatup Limitations Applicable for the First 22 EFPY (TS 3.4.9.1) 4.2-6 4.2-2 Reactor Coolant System Cooldown Limitations Applicable for the First 22 EFPY (TS 3.4.9.1) 4.2-7 4.2-3 Isolated Loop Pressure - Temperature Limit Curve (TS 3.4.9.1) 4.2-8 v Revision 40 I
BVPS-1 ProvidedforInformtation Only.
LICENSING REQUIREMENTS MANUAL 3.4 Axial Flux Difference (AFD) Monitor Alarm LICENSING REQUIREMENT SURVEILLANCES LRS 3.4.1 This surveillance is only required to be performed when the AFD monitor alarm is inoperable and power is above 4-550% RATED THERMAL POWER. Assumelogged alues FDheA exist-dwu ng-he-preceding-time-interval.--Monitor and log the indicated AFD for each OPERABLE channel at least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least once per 30 minutes thereafter. =-Monitor the-indieated-AFD or OPERABLE-excore-channel-at-east-onee-per-hour-for-the-first-24-hours-after-restoring the A:FD-monitor-alarm-te-OPERABLEstatus-.LRS 1.2.3 is not applicable.
3.4^1 Revision 9 l
BVPS-1 l Provided for Information Only.l LICENSING REQUIREMENTS MANUAL 4.1 CORE OPERATING LIMITS REPORT This Core Operating Limits Report provides the cycle specific parameter limits developed in accordance with the NRC approved methodologies specified in Technical Specification Administrative Control 6.9.5.
Specification 3.1.3.5 Shutdown Rod Insertion Limits The shutdown rods shall be withdrawn to at least 225 steps.*
Specification 3.1.3.6 Control Rod Insertion Limits Control Banks A and B shall be withdrawn to at least 225 steps.*
Control Banks C and D shall be limited in physical insertion as shown in Figure 4.1-1.*
Specification 3.2.1 Axial Flux Difference TheAxialEluxDifferenceJAFDIacceptable-operationlimits-are-pro-idedinFigure41--2.
- -The-target-band-is-47% -about-4he-target-flux-fron-O%-to-4I RATED THEP.ML POW~ERl Th-.idieated-Axia1-xux-Differeneer
a. -Above 903; RATED THERMAL POWER-shall-be-maintained w thin-the-+/-17%
target-band-about the target flux diffefenee
b. Between-50%-and-90 9RRATED4HERMAL-POEsitis-within-the-imits-shown-on Figure-44-2.
c. Bel 3; RATED THERMAL POWER say-deviate-outsi e-he arget-band Specification 3.2.2 HeatElux-Hot-ChannelFactor Fn(Z) and:F rLimits The-HeatFluxHot ChanneLEactor -1EaeQ in F CQ] Kfor - P > 0,5 0 ] K(Z)
CQ)[ iory-0.5 Where: C-I2~= SS = RATED THERMAL POWER I
K(Z) = the function obtained from Figure 4.1-3.
As indicated by the group demand counter BEAVER VALLEY - UNIT 1 4.1-1 COLR 4-7 Revision 4*1-
BVPS-I Providedfor.Infomation Only.
LICENSING REQUIREMENTS MANUAL T-Fyirits-f.TYI A-DTHiRntelPGee if-peife-eere-planes-shallb-Vere: Fof-eere-pianes3-ontaining-D-BankI
-For-unrodded-eore-planes-
-y(RTP)- 1-.68-from4".elevationtleaion
-F~,(UGP)-< -r3- 75 2.3 flt-eevation to 3.7 ft. clevation 4xy(RTP)-b1.79-fromff 4.-elevation .58ft-elevation
-Fxy(RTP)51.81-fre 8 e ev on
-y(RTP) 1.74-from44-elevationwe-&.f1.-elevation
-FRFP) -1.-60-frem 8.9 efevation to 10.2 ft. clevation
-1W .O-2
-P--THHRMAL-PWTER-RATED T4hERMA.AL POWER reA prevs themaimum4oaI-tor-timesrrelti-powelt-(FQ tser-a funetion-of-axial-core-beight-during-nornal-eore-operation.
F6 (Z) F6 (Z) *101 FQ (Z) )FQ ( Z) *-w(Z)
The eatElux HotChannelEactor Es2Q(Z)jiui efinedhv:
FQ(Z)< [P CFQ I* K(Z) for P > 0.5 FQ(Z)- [5 CFQ 1 K(Z) <.5 for P(0
[o.5* W(z)J 3WZ esalue-rovided n TableA4.-tL The-Q(Z penalitvnction.applied-hen-he-analvtici FQ(Z)fiunction-changes-by more-than2Oin-a month.is-pro-idedinTable4 U-.
Specification 3.2.3 Nuclear-EnthalpyvRise-HotChannel-Factor - FN Ul FNi,& < CFm * (I + PFA~I (I-P))
Where: CFAI, = 1.62 PFAH = 0.3 P = THERMAL POWER RATED THERMAL POWER BEAVER VALLEY - UNIT 1 4.1-2 COLR 4- l Revision 44
I ProvidedforInformation Only.
FIGURE 4.1-2 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF PERCENT OF RATED THERMAL POWER FOR RAOC I BEAVER VALLEY - UNIT I 4.1-4 COLR 4-Revision 4-1 II
Providedfor Information Only.
Insert Cl-l.
120 11( lmslol-f(+Olre
_0 I 10~
T I _ 1001__
p90 - UNACCEPTABLE4UACPALE OPERATION l l OPERATION
__ ACCEPTABLE I_
D 8 -OPER-ATI ON 0s ___-- \_t t_
4-l 4-4 50 0
~~77n1
-60 -50 -40 -30 -20 -10 0 10 20 30 40 50 60 Flux Difference (Delta I)%
BVPS i LIGENSNG REQUIREMENQSMANUA6 FIGURE-4.4 MAXIM Ur *PREpL) BUS An DURAGN OMAL-OP-ERA-TRION This nage intentionally left blank.
BEAVER VALLEY - UNIT 1 4.1-6 COLR -7 Revision 4-1
I. - . "e A, .: - 3.:' - - '. ': - *: aS This page contains changes, shown double Providedfor Information Only underlined, associated with LAR 302. BVPS-1 i O LICENSING REQUIREMENTS MANUAL -
Specification 3.3.1.1 Reactor Trip System Instrumentation Setpoints. Table 3.3-1 Table Notations A and B Overtemperature AT Setpoint Parameter Values:
Parameter Value Overtemperature AT reactor trip setpoint KI 4-2S 91.242 I Overtemperature AT reactor trip setpoint Tavg coefficient K2 2 0.014655Q.0/F Overtemperature AT reactor trip setpoint pressure coefficient K3 2 GO 1/040M/psia Tavg at RATED THERMAL POWER T' 5*.25-80. 0F Nominal Pressurizer Pressure P'2 2250 psia Measured reactor vessel average temperature lead/lag time constants Tj 2 30 secs T2
4 secs Measured reactor vesset.llagtime-constant 14x;:secs Measured-reactorsvessel-averageiemperature-lagAime-constant 5*2-secs f (AI) is a function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:
(i) for q, - qb between-fl percent and -l4lQ9percent, f (Al) = 0 (where qt and qb are percent l RRMAL POWER in the top and bottom halves of the core respectively, and qt +
qb is total THERMAL POWER in percent of RATED THERMAL POWER). 37 (ii) for each percent that the magnitude of (qt - qb) exceeds 48 p , the AT trip setpoint shall be automatically reduced by 2) percent of its value at RATED THERMAL POWER. 5 2.52 tiii) for each percent that the magnitude of (q, - qb) exceeds r the AT trip setpoint I shall be automatically reduced by 4-.59)A. percent of its value at RATED THERMAL l POWER.
Overpower AT Setpoint Parameter Values:
Parameter Value.
Overpower AT reactor trip setpoint K4 S 4-.094-61.5 I Overpower AT reactor trip setpoint Tavg rate/lag coefficient K5 2 0.02/°F for increasing average temperature K5 = QPFfboad asing averageJemperature BEAVER VALLEY - UNIT 1 4.1-7 COLR 4 l Revision 4-1 I
BVPS-l 0I ICCEPT BLE 0 ATION 650 640 630
¢ 620 -
610 600 //2< i
/EPTABLE OPERATION 0 0.2 0.4 0.6 0.8 1 1.21.
FRACTION OF RATED THERMAL POWER Figure 4.1-5 REACTOR CORE SAFETY LIMIT THREE LOOP OPERATION (Technical Specification Safety Limit 2.1.1)
BEAVER VALLEY - UNIT 1 4.1-9 COLR I-Revision 44
1l I a l -,-X;t-e .- M.
I -*.
Providedfor Information Only.
Insert Cl-2.
670*
660 -sSANCETBEOE' 600 PAUNACCEPTABLE OPERATION 640 630 I- 620 610 600 ACCEPTABLE OPERATION 580 _ _ _ _ _ _ -
0 0.2 0.4 0.6 0.6 1 1.2 1.4 FRACTION OF RATED THERMAL POWER
BY-PS1 ProuidedforInformation Only.
LICFNSNICG REOTITREMENTS MANUALI
.l TableAALfPage-Iof2)
W(Z) VaLues Exclusion Axial Elation II 3000 100 1800 Zone Point LE( MWRJMflI MWDLM\ITU MWD/MTU M1WDM1U x f1 1,0010 ,l.Q00Q LQQ_
fil x 60 020 11.0000 LOOOO 1,0000 L.QQ x 59 10000 1-0000 x 0.60 1.0000 1.0000 1O-000.0
,1.0000 x 51 0.80 LQOO 1.M00 1.0000) x 5-6 LQQ 1.0000 1O090. LQOM 1.000 x 55 J,2Q 1.000 1.0000 x 5A lAO 1M000 1.000Q 1.000 1.0.000 x 53 L£Q 11QQQQ IQQQQI 1.000 52 1.80 L3182 1.3227 1.2640 .2794 51 2Q0 L3Q23 L2M1 L2616 5-0 2.20 LA3296 1.2829 1.2267 1.2430
-42 2A0 1L3Q26 L263Q L20174 1.2231 48 2.60 LQ2 12425 L204-1 42 2.80 L2220 11683 IM4Q 3.00 L2562 1.2032 1.1518 11646 45 3.20 L2449 11904 1-1423 1j586
-44 3.40 1.2334 1.1821 1.1385 1.1576
43. 3.60 1.2240 1.1735 1.1353 1.1553 42 3.80 L2163 1-1640 1.1315 11581 41 111513 1.1215 11638 4.20 L2IS 11540 11246 L1682 39 4AQ 1L491 11234 11718 38 4.60 1JA9 1.1445 11221 1.1742 31 1.1389 J1201 11L756 36 5QQ .1327 1.116 1.157 35 5.20 11619 1.1264 1.1134 1.1740 34 SAQ ,11J26 11135 1.1736 33 5.60 LL432 1.1113 1.1218 11796 32 50 11514 11355 1.1947 31 6f0O L1608 1.1293 11524 L2096 Note:LIop andBottom15%oExcluded BEAVER VALT EY - UNIT I 41-10Q _COLR Revision
-- - z: - --l . -t, BYPS L ProvidedforInformation Only.
I LICENSING REQUIREMENTS MANUAL Table- lUage2-of2)
W(Z)Values Exclusion Axial Elevation 150 300 10000 1 8000 Zone Point 0) M 1LMflm MMWDI 4)LMT1MWDL/MTU 30 fi.20 11-669 J.141-7 1 1697 1.2215 29 6A0 11527 12319 28 6.60 11626 Li 12402 27 112U4 1s2129 1.2468 67S0 26 :.6oo 37M 1 2242 12526 25 .1841 12336 12621 24 1.182 11r90 !241_ L,2621 23 71.60 18221 119-63 12462 L2632 22 AQ L200 312491 2621 21 8.00 11180 1.2024 1.2496 1.2585 20 8.20 L203 3-2476 L2526 12 8A0 1.2015 1.2429 1-2443 18 1J4 1L918 L2356 1.23327 12 8.80 1.1971 L2268 12213 12035 3L2259 L2M 15 9.20 1.2163 1.2348 11926 14 9A0 12296 1J2410 13 9.60 L1622 1.2414 1.2504 1.2087 12 9.80 1.1680 LO 1.2528 1J2653 1.2184 11 1OO0 1.2648 1.2817 1.2278 1.1726 10 10.20 12753 129-65 L2369 x 9 HDA0 10000 L0000 1"000 x 8 10-60 ,1.0000 10000 10000 1-0000.
1-000 x 2 10.80 LOOOO 1.0000 10000 1.0000 x 6 1100 LOOO0 1.0000 l0000 1.0000 x 5 1L2Q 1 0000 10000 10000 1.0000 x 11.a0 1.0000 10000 1n000 1 0000 x 3 MO 1Q0n L0000 1.0000 x 2 11.80 1.0000 10000 LO0OO x I 12-00 1 0000 10000 1000 Lnoo Note: Ton and Bottom 155% Excluded TVPAVFPR VAT1TV-. - ITTIT 1 A 1-11 C()T P Revision
FI - I -1 ; - - i. rl. - -, . -1S-I - -- -
IBS I Provided for Information Only.
TiCENSING ROITTRMFMPNT5R MANUTAT.
.I TableA4J1.2 OM ealyfactor Note The-enaltvDactor-johc-apuole _(ZQin(accn-dacanceiwiL echnicaLSp-eification Sumiliancequirement .21.2-1 maimum-fac-tg vby ichEqZ) pccted-increase-over-a39EffectiveFullPowerDayiEFPD interval(suriveillanceinteryal-of3-LEFPD plus-the-naximum-allowable-extensio-not-to-exceed-5%oofithe-surveillanceinterivalper C~khicaLcificaton Surveillac ueqiinen t 4.02 It ftig~mtheum ,awhcte EQLZ.asdetermined F'PAVPR- VAT
.-* PIYV- vTKNT I--. As
- sA1 A.. 1-12 'COTP X V's.o Re-vision
BVPS-1 [I Providedfor Information Only.
LICENSING REQUIREMENTS MANUAL B.3.3 METEOROLOGICAL MONITORING INSTRUMENTATION The OPERABILITY of the meteorological instrumentation ensures that sufficient meteorological data is available for estimating potential radiation doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere. This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public and is consistent with the recommendations of Regulatory Guide 1.23, "Onsite Meteorological Programs."
B.3.4 AXIAL FLUX DIFFERENCE (AFD) MONITOR ALARM Surveillance of the AFD verifies that the AFD, as indicated by the Nuclear Instrumentation System (NIS) excore channels, is within its limits. During operation above 4550% RATED THERMAL POWER, when the AFD monitor alarm is inoperable, additional surveillance criteria is required by the Licensing Requirements Manual beyond the surveillance criteria required by the Technical Specifications to detect operation outside of the limits tamet-band-nd4o-eomputethe-penaltydeviation time-bore-eerreetive-etion-eq ed. The lgged vauesf4he-AFD--are-assumed-to-.exist-feFr-he preeeding4ime interval-in-order- eopeate-eute4he-cumulative-penaltydeviation-time B.3.5 QUADRANT POWER TILT RATIO (QPTR) MONITOR ALARM Surveillance of the QPTR verifies that the QPTR, as indicated by the Nuclear Instrumentation System (NIS) excore channels, is within its limits. During operation above 50% RATED THERMAL POWER, when the QPTR monitor alarm is inoperable, additional surveillance criteria is required by the Licensing Requirements Manual beyond the surveillance criteria required by the Technical Specifications to detect any relatively slow changes in QPTR. For those causes of core power tilt that occur quickly (e.g., a dropped rod), there are other indications of abnormality that prompt a verification of core power tilt.
B.3-1 Revision 9 l
BVPS-1 Provided for Information Only.
LICENSING REQUIREMENTS MANUAL BASES B.3.8 LEADING EDGE FLOW METER (Continued)
This surveillance is performed every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when power is above 4-5A0)%. The NIS excore power range channel indications are renormalized if they are not found to be within +2% of the calorimetric measurement. This +2% requirement for renormalization is distinct from the allowance for calorimetric uncertainty, and these allowances are handled as independent contributions to determine the maximum power assumed in design basis accident analyses.
The plant may then be run for the next 24-hour period using this normalized NIS indication. Although calorimetric power indication may be monitored continuously, it is not required to be consulted again until the required daily calorimetric comparisons of NIS indication are performed.
The surveillance requirement to perform planned maintenance and inspections every 18 months is based upon the manufacturer's recommendations, and is consistent with the surveillance intervals specified for similar electronic apparatus.
Additional guidance for determining steady-state THERMAL POWER is taken from the NRC Inspection Manual; Inspection Procedure 61706; C/N 86-036, 07/14/1986; "Core Thermal Power Evaluation"; step 03.02.d, and is described in the BVPS Operating Manual.
B.3-4 Revision 22 l
Attachment C-2 Beaver Valley Power Station, Unit No. 2 Proposed Licensing Requirements Manual Changes License Amendment Request No. 182 The following is a list of the affected pages:
Page V
3.4-1 4.1-1 4.1-2 4.1-4 4.1-6 4.1-7 4.1-10 B.3-1 B.3-3
BVPS-2 Provided for Information Only.
LICENSING REOUIREMENTS MANUAL (LRM)
INDEX TABLES TITLE PAGE Table 3.12-1 Explosive Gas Monitoring Instrumentation 3.12-2 Table 3.12-2 Explosive Gas Monitoring Instrumentation Surveillance Requirements 3.12-3 Table41-l M(Z)NValues .4.1A1 Table-4.L-2. YQ(Z)VPenaltyW1actor 4.123 Table 4.2-1 Heatup Curve Data Points for 14 EFPY (TS 3.4.9.1) 4.2-14 Table 4.2-2 Cooldown Curve Data Points for 14 EFPY (TS 3.4.9.1) 4.2-15 Table 4.2-3 Overpressure Protection System (OPPS) Setpoints (TS 3.4.9.3) 4.2-16 Table 4.2-4 Reactor Coolant Pump Restrictions 4.2-17 Table 4.2-5 Calculation of Chemistry Factors Using Surveillance Capsule Data 4.2-18 Table 4.2-6 Reactor Vessel Beltline Material Properties 4.2-19 Table 4.2-7 Summary of Adjusted Reference Temperatures (ARTs) for 15 EFPY 4.2-20 Table 4.2-8 Calculation of Adjusted Reference Temperatures (ARTs) for 15 EFPY 4.2-21 Table 4.2-9 Reactor Vessel Toughness Data (Unirradiated) 4.2-22 Table 4.2-10 RTpMs Calculation for Beltline Region Material at EOL (32 EFPY) 4.2-23 Table 5.1-1 Containment Penetrations 5.1-1 Table 6.5-1 Snubber Visual Inspection Interval 6.5-5 Table 8.3-1 Reactor Coolant System Chemistry Limits 8.3-2 Table 8.3-2 Reactor Coolant System Chemistry Limits Surveillance Requirements 8.3-3 FIGURES TITLE PAGE 4.1-1 Control Rod Insertion Limits as a Function of Percent-ofRated Power Level or RAOC 4.1-3 4.1-2 Axial Flux Difference Limits as a Function of Rated Thermal Power 4.1-4 4.1-3 FT Normalized Operating Envelope, K(Z) 4.1-5 v Revision 3 l
BVPS-2 Providedfor Information Only.
LICENSING REQUIREMENTS MANUAL (LRM)
INDEX FIGURES TITLE PAGE 4.1-4 D-eletedMaximnuim(+/-PRad- -AxiaVGore-Height-During Normal-Operation 4.1-6 4.1-5 Reactor Core Safety Limit Three Loop Operation (Technical Specification Safety Limit 2.1.1) 4.1-10 4.2-1 Reactor Coolant System Heatup Limitations Applicable for the First 14 EFPY (TS 3.4.9.1) 4.2-6 4.2-2 Reactor Coolant System Cooldown (up to 00F/Hr.)
Limitations Applicable for the First 14 EFPY (TS 3.4.9.1) 4.2-7 4.2-3 Reactor Coolant System Cooldown (up to 200 F/Hr.)
Limitations Applicable for the First 14 EFPY (TS 3.4.9.1) 4.2-8 4.2-4 Reactor Coolant System Cooldown (up to 40F/lHr.)
Limitations Applicable for the First 14 EFPY (TS 3.4.9.1) 4.2-9 4.2-5 Reactor Coolant System Cooldown (up to 60 °F/Hr.)
Limitations Applicable for the First 14 EFPY (TS 3.4.9.1) 4.2-10 4.2-6 Reactor Coolant System Cooldown (up to 1000 F/Hr.)
Limitations Applicable for the First 14 EFPY (TS 3.4.9.1) 4.2-11 4.2-7 Isolated Loop Pressure - Temperature Limit Curve (TS 3.4.9.1) 4.2-12 4.2-8 Maximum Allowable Nominal PORV Setpoint for the Overpressure Protection System (TS 3.4.9.3) 4.2-13 vi Revision 40 l
BVPS-2t Provided for Information Only.
LICENSING REQUIREMENTS MANUAL 3.4 Axial Flux Difference (AFD) Monitor Alarm LICENSING REQUIREMENT SURVEILLANCES LRS 3.4.1 This surveillance is only required to be performed when the AFD monitor alarm is inoperable and power is above 4550% RATED THERMAL POWER. Assumelogged l values of-the-AFD-exist-during-the-preceding-time-interva.--Monitor and log the indicated AFD for each OPERABLE channel at least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least once per 30 minutes thereafter. M r-he-ndicated AFD fcr each OPER emeore nE eh elat leastenee-per-h our-frten--heus-fter-festoring theAFDmonitera tPERABLEstatuB. LRS 1.2.3 is not applicable.
3.4-1 Revision ;7 l
BVPS-2 ProvidedforInformationOnly.
LICENSING REQUIREMENTS MANUAL 4.1 CORE OPERATING LIMITS REPORT This Core Operating Limits Report provides the cycle specific parameter limits developed in accordance with the NRC approved methodologies specified in Technical Specification Administrative Control 6.9.5.
Specification 3.1.3.5 Shutdown Rod Insertion Limits The Shutdown rods shall be withdrawn to at least 225 steps.*
Specification 3.1.3.6 Control Rod Insertion Limits Control Banks A and B shall be withdrawn to at least 225 steps.*
Control Banks C and D shall be limited in physical insertion as shown in Figure 4.1-1.*
Specification 3.2.1 Axial Flux Difference
__tialEluxl eence D))aceep~lu ftin limitsided n Figure 4.1-2 NOTE: The4arget-bandis +% ab tuarget-te4UXfm-0%4o400% RATED-THERMAL POWER Thie-ndk~ieae xab-F~ux-Diff~efenee:
a Ao903 RATED FH RMA6POER shallbe-maintained-vithin_7he-°- target-band about4heafget-fl enee
--br- Between ;EM e50%,a9-and-9 POWER-is-withinhimits-shown-on-Figure 4A-2.
Bo ;ATEDPd eoutidee-argeFband Specification 3.2.2 Heat Euxfthannel Factor -Q(Z) and-Fy-Limits The-Heat-FluxHot-ChanneLFactor - FQ(ZY..ariablei&definediv; FQ(Z), (C<*K(Z) for P->.5 for P 0.5 FQC) 4*KZ L 0.5 ]fr forKP(Z.5 THERMAL POWER Where: -F (Z) = 232.4 P RATED THERMAL POWER
K(Z) = the function obtained from Figure 4.1-3.
As indicated by the group demand counter COLR44 BEAVER VALLEY - UNIT 2 4.1-1 Revision 34
BVPS-2 . Providedfor Information Only.
LICENSING REQUIREMENTS MANUAL
-Te-Fyqiitsfxy,7oF RATED 4HERMALPOWER -wthin- speeifie-eore-panes-shall-be-FxyO4=-F J s xyy* J s) J A~ere.--or-al oref anes-eontaining D-Bank:
p{D TD .
F-r-twodded-eore-planes.
yffIP!5146 o .fl.eeva. atin

Fy(RTPYP)-h.80from -fbelevationeto3.7f1.elevation F ye83 3.7 ft. ele vat tion

-F yqRxy5*-- 84 8televat U A o ft. evation
--- -TP&yIo-44 e1emv7.4fti.levatio-norelevation
--FFM(RP )- 1.72-fro m.
ft. elevatin to 10.2 R. elevation PFXY .0 EHERMA 4u.m trotal peunng fheter tmes-relative-power -F4T&tas2e-of dal o heightduringmabeore-operatiem FQ3 (Z) F(Z) *LO815 FQ ( Z )= F ( Z) *WfZ)
= FQ(Z) -
,TheAHeat-Eux-liotChannellactorE=FQ(Z)1WiUs-defined-by.: deC Q(Z) [ptwz)] K()- forP > 0.5 Q P*W(Z) *K(Z)
FQ(Z)*5 [0.5FQ(Z K(Z) for P < 0.5 3Y(Z-) vlesrprdsinleA.I=1.
ThePqF(Z) penaltvhnction-appied-when-heanalvtic Q(Z)functiomchanges-bynore-thana2%in-a month.ji proidedinj-ble 412.
Specification 3.2.3 Nuclear-nthalpy=iseHol-Channeel-catorzfI I N
FeTl < CF Al * (I + PF IA(I - P))
Where: CFml = 1.62 Mm = 0.3 THERMAL POWER P RATED THERMAL POWER COLR4I BEAVER VALLEY - UNIT 2 4.1-2 Revision -34
F-----l-- -.-- -.-- -- - T-Providedfor Information Only.
FIGURE 4.1-2 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF PERCENT OF RATED THERMAL POWER FOR RAOC I COLR5I-1 BEAVER VALLEY - UNIT 2 4.1-4 Revision 34
XII - -b - -. . w wc I ProvidedforInformation Only.
Insert C2-1.
120 110_______________ F____
1 00 __D}-<-L,
_____{_
90 UNACCEP -- UNACCEPTABLE
__OPERATION l OPERATION 2 , 8 0 _ _]l ACCEPTABLE 1 _ 1ACE-_:>_
1 i l
° - OPERATIAN E-4 _ _
0
-60 -50 30 -20 -10 0 10 20 30 40 50 60 Flux Difference (Delta I)k
-, , =i - -! - %4 -
2 a- 4 l BVPS-2 Providedfor InformationOnly.
LICENSING REQUIREMENTS MANUAL 14GURE-4.-4 MAXIMUM-(FqTPrelI3-V-AXI1A-CORE-H;EGHT DAa _ ON lTis-pageintentionally-lefl-blanlc I COLR 11 BEAVER VALLEY - UNIT 2 4.1-6 Revision 34 I
This page contains changes, shown double l Providedfor Information Only.
underlined, associated with LAR 173. BVPS-2 l LICENSING REQUIREMENTS MANUAL Specification 3.3.1.1 Reactor Trip System Instrumentation Setpoints. Table 3.3-1 Table Notations A and B Overtemperature AT Setpoint Parameter Values:
Parameter Value Overtemperature AT reactor trip setpoint KI 4414-1239 Overtemperature AT reactor trip setpoint Tavg coefficient K2 2 0.0 I83/0 F Overtemperature AT reactor trip setpoint pressure coefficient K3 2 0000820fO1/psia Tavg at RATED THERMAL POWER T`< 576.25SQ0 0 F Nominal pressurizer pressure P' 2 2250 psia Measured reactor vessel AT lead/lag time constants Tl Ž-8 !!Qsec_
l(he-response-timeisAoggledoff-to-meet-the-analysisyalueofzero. t2 *4-20sec_
Measured reactor vessel AT lag time constant T3
0-sec Measured reactor vessel average temperature lead/lag time constants r4 2 30 sec
-r5 <4 sec Measured reactor vessel average temperature lag time constant T6
0-2sec f (Al) is a function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant r---i
--1&tartup tests such that: [~
(i) For q, - qb between _% and +qb1f(AI)= 0, where qt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt + qb is tota THERMAL POWER in percent of RATED THERMAL POWER; 37 (ii) For each percent that the magnitude of qt - qb exceeds Setpoint shall be automatically reduced by . % of its value at RATED THERMAL POR (iii) For each percent that the magnitude of qt - qb exceeds Trip Setpoint shall be automatically reduced by 44611.4% of its value at RATED THERMAL POWER.
COLR X4 BEAVER VALLEY - UNIT 2 4.1-7 Revision 34
BVPS-2 Providedfor Information Only.
LEQUIREM
.Figure 4.1-5 REACTOR CORE SAFETY LIMIT THREE LOOP OPERATION (Technical Specification Safety Limit 2.1.1)
COLR I4 BEAVER VALLEY - UNIT 2 4.1-10 Revision 34
[I ProidedforInformation Only.
II Insert C72-27 670 660 l > \ 435 PSIA U N C CEPTAB LE O P ElkO 650
. 2250 PSI l 640 630
<~0 P SIA 1! 620 1920 PIJb-610 600
.ACC EPTABLE OPERATION<
590 580 0 0.2 0.4 0.6 0.8 1 1.2 1.4 FRACTION OF RATED THERMAL POWER
I7, -, S-- 7e-.; i: S. r B-PS L tf2PidedforInfoaton Only. l TTCENSING REOUTIRPMENTS MANUTAT.
TablegAj1APage1-f 2)
Vales W(Z)
Exclusion Axial Eleva tionI 30 18000 Zone Point I l) I MWQ8l I MWD/MTU I WD1 MWMIM U x 61 .Q00 1.0000 0OOOO L1000 x 60 020 10000 10000 .0000 Lnn x 5- PAD 10OO0 1.0000 I1sOOOD x 58 0.60 .1.0000 LOOOO 1.0000 1.0000 x 52 080 1.0000 10O00 1.0000 x 56 1.0 10000 x 55 I0000 1000 1.000 L20 1.000 LOQQ x LO 1.0000 noonon 1.0000 x 53 160 10000 1.0000 52 1.80 13296 1.3227 1.2640 1.2794 51 200 13023 L2452 1.2616 50 2.20 1.2829 1.2267 .L2430 49- 2.4Q I802 1.2630 1.2074 L2231 2.60 1.2425 12041 LH17 2.80 12220 LO 3.00 ,L2562 1.2032 1.1518 11646 3.20 ,L2449 11904 .11-423 1.1586 3.40 1.233 1.1821 1.1385 1.1516 3.60 1.1735 1.1353 11553 42 3.80 L.2163 1.1640 1.1315 1s1581 41 36 4.0 1209.8 1.1513 1.1275 11638 4.20 .1,2037 11540 1246 11682 344 39 4AQ 1L149 1s1234 38 4.60 12890 l.1445 11221 11242 4.80 1.1389 1.1201 1.1156 5.0 1.L713 LL2 35 5.20 1.161,9 1.1264 1.1134 1.1740 5AD 1.1504 U1176 1.1135 1.1236 33 5.60 11432 11113 1.1218 1.1796 32 580 11514 1LM 1.1355 1.1942 31 6.o0 IIEQ8 11293 11524 1.209 Note:lop-and-Bottoml5%oExcluded CQLR BEAVERYALLEYXJIUNIT-2 4.111 Revision
". ' - -1 3a -~- r
.i' J- _.4- c E.' - sh -z '! '
BVYR2 I IProvided for Information Only.
LLCENSINGREQ!1 EEMUSIMANUAL
~abkbL(Page-2-of-2)
WM-Y alues Exclusion I Axial I Elain I INQ I 3000 I 10MM Zone Po I MwYl 't" M AVDLMflI M0) , l1T 30 6.20 1669 .1417 11.697 1.2215 29 6.40 1.26 11527 11855 1.2319 28 6.60 .1769 11626 19992 12402 27 6.80 .1792 7114 1.2122 1.2468 26 200 1.1812 12.86 12242 L2 25 21. 11813 1841. 1.2336 1.2583 24 2.40 11822 119-04 12410 12621 23 760 LL821 11J963 1.2462 12632 22 8M 1.2003 1.2491 122 21 8.00 1A7IQ 1.2024 1.2496 1.2585 20 8.20 1 12031 12476 12526 19 8A40 1166 1.2015 12429 L2443 18 8.60 11583 11928 L2356 L2331 17 8.80 1.1501 .191 L2268 L2213 1 921L0 1.145 12035 L2259 1205_
15 , 9.20 .1544 1.2163 1.2348 J.1926 90 9L 1J.1580 1.22.96 L2410 11976 13 9.60 L1622 1.2414 1.2504 1.2087 12 9.80 L1680 1.2528 1.2653 1.2184-4
= 11 10.00 A1126 1.2648 1.2812 1.2228 1.0 10.20 1l736 1.2153 129-65 1.2362 x 9 "4A r10000 l.O00 1Q000 10ODD x S O.£60 L0000 1.0000 LQOOQ LO00Q x 27 10.80 10000 1.0000 1.0000 1.0000 x 6 11.00 LOOQO 1.0000 1.0000 1.0000 x 5 1L20 LOOO 1.0000 LQOQ 1.0000I x 4 J1140 LOOO 1.0000 1.0000 1.0000 x 3 1160 1ooo 10000 LQQQQ 1.0000 I x 2 1180 10000 1.0000 1.0000 l0000 x 1 12.00 QQ !n 10000 1o0000 1.0000 Nnte: op and3Bnttom 15%.Excluded COLR BEAVER VA RY-I= 4&j2 eision
gt- - -- s 3 B_____ Provided for Information Only.
L[CENSIN~ MNSAAIA TableA4.1-2 mQ(Z) PenaltyFctor CycleBurnup -(MWDfLt FO(Z) enalt ractor AllBurnups 1.02 Note ThnLamltyacluor tobQp~idtoFiz) n- acc ord ance-w-ith-Tezhnic alpcification SureillanceRequiremef 4.2.2.3. is the u actor whichp tex increase-over-a.39Effectivefull-ow-erDaylEff D- intervali(surveillanceinterval-of31lEF-D plus-the-maximum-allowable-extension-not-to-exceed-25%Y of-the-surveillance-intervalper lh aLjeterSii ei BFAVFR VAT .T.EY-1UMT2 4A113 RevisiQn
BVPS-2 Provided for Informnation Only.
LICENSING REQUIREMENTS MANUAL B.3.3 METEOROLOGICAL MONITORING INSTRUMENTATION The OPERABILITY of the meteorological instrumentation ensures that sufficient meteorological data is available for estimating potential radiation doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere. This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public and is consistent with the recommendations of Regulatory Guide 1.23, "Onsite Meteorological Programs."
B.3.4 AXIAL FLUX DIFFERENCE (AFD) MONITOR ALARM Surveillance of the AFD verifies that the AFD, as indicated by the Nuclear Instrumentation System (NIS) excore channels, is within its limits. During operation above 4550% RATED THERMAL POWER, when the AFD monitor alarm is inoperable, additional surveillance criteria is required by the Licensing Requirements Manual beyond the surveillance criteria required by the Technical Specifications to detect operation outside of the limjijtarget-band-and-tocompute-the-penalty-deviation timeore-erreetive- ation-re T d es-efthe-AFD-are assuned-to-exist'e4-he preeeding4ime-nterval-inorderfor4hepertoeeompute-the-eumulative-penalty-deviationtimne:
B.3.5 QUADRANT POWER TILT RATIO (QPTR) MONITOR ALARM Surveillance of the QPTR verifies that the QPTR, as indicated by the Nuclear Instrumentation System (NIS) excore channels, is within its limits. During operation above 50% RATED THERMAL POWER, when the QPTR monitor alarm is inoperable, additional surveillance criteria is required by the Licensing Requirements Manual beyond the surveillance criteria required by the Technical Specifications to detect any relatively slow changes in QPTR. For those causes of core power tilt that occur quickly (e.g., a dropped rod), there are other indications of abnormality that prompt a verification of core power tilt.
B.3-1 Revision -7 l
- - 5 BVPS-2 LICENSING REQUIREMENTS MANUAL BASES B.3.8 LEADING EDGE FLOW METER (Continued)
The Applicability Statement applies when performing calorimetric power measurements during MODE 1 operations at steady-state conditions above 2652 MWt. The Operating License limits the maximum steady state power to 2689 MWt when calorimetric heat balance measurements are made daily using the LEFM.
If the LEFM is not OPERABLE during the interval between required calorimetric heat balance measurements, plant operation may continue at
2689 MWt steady-state, using the existing Nuclear Instrumentation System (NIS) indication until the next required performance of the daily power calorimetric surveillance is due.
If the LEFM remains inoperable at the time that the next required calorimetric heat balance measurement is due, plant operation may continue at
2652 MWt steady-state, by making calorimetric measurements using feedwater flow venturis and Resistance Temperature Detector (RTD) indications.
The requirement to reduce power within one hour is based upon comparison to similar action statements in the technical specifications. The increase in likelihood that the NIS will need renormalizing after 25 hours2.893519e-4 days 0.00694 hours 4.133598e-5 weeks 9.5125e-6 months compared to after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is considered negligible. A Note, designated by "*", is added to the Licensing Requirement to denote a difference between power measurements obtained when using the feedwater flow venturis and the LEFM. An indication of 2652 MWt from the LEFM is equivalent to an indication of 2612 MWt from the feedwater flow venturis.
It is preferable that the daily heat balance calculations be made using the subroutine on the plant computer system (PCS). If the PCS is unavailable, a manual calculation that accounts for steam generator blowdown is acceptable, and may be performed in lieu of using the PCS.
This surveillance is performed every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when power is above 4-550%. The NIS excore power range channel indications are renormalized if they are not found to be within +/-2% of the calorimetric measurement. This +2% requirement for renormalization is distinct from the allowance for calorimetric uncertainty, and these allowances are handled as independent contributions to determine the maximum power assumed in design basis accident analyses.
The plant may then be run for the next 24-hour period using this normalized NIS indication. Although calorimetric power indication may be monitored continuously, it is not required to be consulted again until the required daily calorimetric comparisons of NIS indication are performed.
The surveillance requirement to perform planned maintenance and inspections every 18 months is based upon the manufacturer's recommendations, and is consistent with the surveillance intervals specified for similar electronic apparatus.
Additional guidance for determining steady-state THERMAL POWER is taken from the NRC Inspection Manual; Inspection Procedure 61706; CIN 86-036, 07/14/1986; "Core Thermal Power Evaluation"; step 03.02.d, and is described in the BVPS Operating Manual.
B.3-3 Revision ;4 l

L-05-009, License Amendment Request Nos. 310 and 182

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Subject:

Enclosure:

Subject:

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

6.0 ENVIRONMENTAL CONSIDERATION

7.0 REFERENCES

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

6.0 ENVIRONMENTAL CONSIDERATION

7.0 REFERENCES

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L-05-009, License Amendment Request Nos. 310 and 182

Text

Subject:

Enclosure:

Subject:

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

6.0 ENVIRONMENTAL CONSIDERATION

7.0 REFERENCES

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

6.0 ENVIRONMENTAL CONSIDERATION

7.0 REFERENCES

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