Text

License Amendment Request: Technical Specification Revision to Adopt TSTF-339-A, Relocate Technical Specification Parameters to the COLR Consistent with WCAP-14483
	ML23101A159
Person / Time
Site:	Vogtle
Issue date:	04/11/2023
From:	Brown R; Southern Nuclear Operating Co
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
References
	NL-23-0152, WCAP-14483, TSTF-339-A
	Download: ML23101A159 (1)
	v • d • e

Regulatory Affairs 3535 Colonnade Parkway Birmingham, AL 35243

205-992-5000

April, 2023

NL-23-0152 Docket Nos.: 50-424 10 CFR 50.90 50-425

U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001

Vogtle Electric Generating Plant - Units 1 and 2 License Amendment Request: Technical Specification Revision to Adopt TSTF-339-A, Relocate Technical Specification parameters to the COLR consistent with WCAP-14483

Pursuant to 10 CFR 50.90, Southern Nuclear Operating Company hereby requests a license amendment to Vogtle Electric Generating Plant (VEGP) Units 1 and 2 renewed facility operating licenses NPF-68 and NPF-81, respectively.

The proposed change would revise Technical Specifications (TS) 2.1.1, Reactor Coolant Safety Limits, (TS) 3.3.1, Reactor Trip System (RTS) Instrumentation, TS 3.4.1, Reactor Coolant System (RCS) Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits, and TS 5.6.5, Core Operating Limits Report (COLR), to adopt most of the TS and COLR changes described in Appendix A and Appendix B of Westinghouse topical report WCAP-14483-A, to relocate several cycle-specific parameter limits from the TS to the COLR. The proposed change is done following the guidance of technical specification task force change traveler TSTF-339-A Revision 2. Along with the parameter relocations, this license amendment request (LAR) modifies the VEGP 1&2 TS 5.6.5, Core Operating Limits Report (COLR), to include WCAP-8745-P-A and WCAP-11397-P-A, and to revise the Specification applicability for the WCAP-9272-P-A, in the list of the NRC approved methodologies used to develop the cycle specific COLR. An administrative revision is also proposed to the TS 3.3.1 depiction of an equation.

The Enclosure provides the description, technical evaluation, regulatory evaluation (including the Significant Hazards Consideration Determination) and environmental considerations for the proposed changes. Attachments 1 and 2 provide marked-up pages depicting the requested changes and revised pages, respectively, to the VEGP Units 1 and 2 TS. Attachment 3 provides associated marked-up pages to the VEGP Units 1 and 2 TS Bases document for information. provides associated marked-up pages of the VEGP Units 1 and 2 Core Operating Limits Reports for information.

This proposed change was discussed during a public meeting held on February 13, 2023, as provided in ADAMS [Accession No. ML23037A921].

Approval of the proposed amendment is requested within one year from the date of this submittal.

Due to the core design and safety analysis evaluation needed to support each core design using the methodology in the WCAP, implementation of this amendment for each unit will coincide with U. S. Nuclear Regulatory Commission NL-23-0152 Page 2

the start of the respective fuel cycles for each unit allowing for the necessary implementation prior to the beginning of the fuel cycle. These fuel cycles are currently anticipated to begin Fall 2024 for VEGP Unit 1 and Spring 2025 for VEGP Unit 2.

In accordance with 10 CFR 50.91, SNC is notifying the State of Georgia and State of Alabama of this license amendment request by transmitting a copy of this letter and enclosures to the designated State Official. This letter contains no NRC commitments.

This letter contains no regulatory commitments. If you have any questions, please contact Amy Chamberlain at 205.992.6361.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the th day of April 2023.

Respectfully submitted,

R.Keith Brown Director, Regulatory Affairs Southern Nuclear Operating Company

5 KB/dsp/cbg U. S. Nuclear Regulatory Commission NL-23-0152 Page 3

Enclosure:

Evaluation of Proposed Change

Attachments: 1. Vogtle Electric Generating Plant - Units 1 and 2 Marked-up TS Pages

2. Vogtle Electric Generating Plant - Units 1 and 2 Revised TS Pages

3. Vogtle Electric Generating Plant - Units 1 and 2 Marked-up TS Bases Pages (for information only)

4. Vogtle Electric Generating Plant - Units 1 and 2 Markups of Core Operating Limits Report Pages (typical, for information only)

cc: Regional Administrator, Region ll NRR Project Manager - Vogtle 1 & 2 Senior Resident Inspector - Vogtle 1 & 2 State of Georgia Environmental Protection Division RType: CVC7000 ENCLOSURE Evaluation of Proposed Change

License Amendment Request: Technical Specification Revision to Adopt TSTF-339-A, Relocate Technical Specification parameters to the COLR consistent with WCAP-14483

1.

SUMMARY

DESCRIPTION

2. DETAILED DESCRIPTION 2.1 System Design and Operation 2.2 Current Technical Specification Requirements 2.3 Reason for Proposed Change 2.4 Description of Proposed Change

3. TECHNICAL EVALUATION 3.1 Basis for the Changes 3.2 TSTF Change Traveler Adoption and Variations 3.3 Applicability of WCAP-14483-P-A, Safety Evaluation and Variations

4. REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 Significant Hazards Consideration 4.4 Conclusions

5. ENVIRONMENTAL CONSIDERATION

6. REFERENCES

ATTACHMENTS:

1. Vogtle Electric Generating Plant - Units 1 and 2 Marked-up TS Pages

2. Vogtle Electric Generating Plant - Units 1 and 2 Revised TS Pages

3. Vogtle Electric Generating Plant - Units 1 and 2 Marked-up TS Bases Pages (for information only)

4. Vogtle Electric Generating Plant - Units 1 and 2 Markups of Core Operating Limits Report Pages (typical, for information only)

Enclosure to NL-23-0152 Evaluation of Proposed Change

1.

SUMMARY

DESCRIPTION

The proposed amendment would revise the Technical Specifications for Vogtle Electric Generating Plant (VEGP) Units 1 and 2 renewed facility operating licenses NPF-68 and NPF-81, respectively.

The proposed change would revise Technical Specifications (TS) 2.1.1, Reactor Core Safety Limits, (TS) 3.3.1, Reactor Trip System (RTS) Instrumentation, TS 3.4.1, Reactor Coolant System (RCS) Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)

Limits, and TS 5.6.5, Core Operating Limits Report (COLR), to adopt most of the TS and COLR changes described in Appendix A and Appendix B of Westinghouse topical report WCAP-14483-A, to relocate several cycle-specific parameter limits from the TS to the COLR.

The proposed change is done following the guidance of technical specification task force change traveler TSTF-339-A, Revision 2. Along with the parameter relocations, this license amendment request (LAR) modifies the VEGP 1&2 TS 5.6.5, Core Operating Limits Report (COLR), to include WCAP-8745-P-A and WCAP-11397-P-A, and to revise the Specification applicability for WCAP-9272-P-A, in the list of the NRC approved methodologies used to develop the cycle specific COLR. An administrative revision is also proposed to the TS 3.3.1 depiction of an equation.

Approval of the proposed amendment is requested within one year from the date of this submittal. Due to the core design and safety analysis evaluation needed to support each core design using the methodology in the WCAP, implementation of this amendment for each unit will coincide with the start of the respective fuel cycles for each unit allowing for the necessary implementation prior to the beginning of the fuel cycle.

Four attachments are provided with this enclosure. Attachments 1 and 2 provide the marked-up TS pages and revised TS pages, respectively, for VEGP 1&2. Attachment 3 contains the associated marked-up TS Bases pages, showing the associated changes to be implemented based on Appendix A of WCAP-14483-A, for information only. Finally, Attachment 4 contains the associated marked-up Core Operating Limits Report, showing typical associated changes based on Appendix B of WCAP-14483-A, for information only.

Internally controlled licensing basis documents such as the COLR, the TS Bases and the FSAR will be updated to reflect these changes at the time of implementation of the LAR.

E-1 Enclosure to NL-23-0152 Evaluation of Proposed Change

2. DETAILED DESCRIPTION

2.1 System Design and Operation

General Design Criterion 10 of 10 CFR Part 50 Appendix A requires that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs).

The Reactor Coolant Safety Limits support th e above requirement by having a departure from nucleate boiling (DNB) design basis, which corresponds to a 95% probability at a 95% confidence level (the 95/95 DNB criterion) that DNB will not occur on the limiting fuel rods and by requiring that fuel centerline temperature stays below the melting temperature. The restrictions of the Safety Limits prevent overheating of the fuel and cladding, as well as possible cladding perforation, that would result in the release of fission products to the reactor coolant. Operation above the boundary of the nucleate boiling regime could result in excessive cladding temperature because of the onset of DNB and the resultant sharp reduction in heat transfer coefficient.

The Reactor Trip System (RTS) initiates a unit shutdown, based on the values of selected unit parameters, to protect against violating the core fuel design limits and RCS pressure boundary during AOOs and to assist the Engineered Safety Features (ESF) Systems in mitigating accidents.

The Overtemperature T trip Function is provided to maintain DNB ratio within the design limit. This trip Function also limits the range over which the Overpower T trip Function must provide protection. Protection from violating the DNB ratio limit is assured for those transients that are slow with respect to delays from the core parameters to the measurement system. The Function monitors both variation in power and flow since a decrease in flow has the same effect on T as a power increase.

The Overpower T trip Function provides protection for the integrity of the fuel under overpower conditions. This trip Function also limits the required range of the Overtemperature T trip Function and provides a backup to the Power Range Neutron Flux High Setpoint trip. The Overpower T trip Function prevents exceeding the allowable heat generation rate (kW/ft) of the fuel.

The RCS is required to operate within certain parameter limits (for pressurizer pressure, RCS average temperature, and RCS total flow rate) that are assumed in the safety analyses to maintain the DNB within the design criterion in the event of a DNB limited transient.

E-2 Enclosure to NL-23-0152 Evaluation of Proposed Change

2.2 Current Technical Specification Requirements

Technical Specification (TS) 2.1.1, Reactor Core Safety Limits, provides limits for a combination of THERMAL POWER, RCS highest loop average temperature, and pressurizer pressure for operation in MODES 1 and 2.

TS 3.3.1, Reactor Trip System (RTS) Instrumentation, includes several cycle-specific input values for the Overtemperature Delta-T (OTT) Function 6 of Table 3.3.1-1 and the Overpower Delta-T (OPT) Function 7 of Table 3.3.1-1 in Notes 1 and 2 of Table 3.3.1-1, respectively, for operation in MODES 1 and 2. The Notes include depiction of the equation used for each function and defines the terms in the equation.

Similarly, TS 3.4.1, Reactor Coolant System (RCS) Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits, also include several cycle-specific parameter limits for MODE 1 operation.

TS 5.6.5.b lists specific NRC approved documents describing analytical methods used to determine the core operating limits and includes parenthetical information identifying which Specifications are supported by the methodology. These currently include core operating limits associated with TS 3.1.1 - "SHUTDOWN MARGIN," TS 3.1.3 - "Moderator Temperature Coefficient," TS 3.1.5 - "Shutdown Bank Insertion Limits," TS 3.1.6 - "Control Bank Insertion Limits," TS 3.2.1 - "Heat Flux Hot Channel Factor," TS 3.2.2 - "Nuclear Enthalpy Rise Hot Channel Factor," TS 3.2.3 - "Axial Flux Difference," and TS 3.9.1 -

"Boron Concentration," as listed in TS 5.6.5.a.

2.3 Reason for Proposed Change

NRC Generic Letter 88-16 allows licensees to remove cycle-dependent variables from TS provided that the values of these variables are included in a COLR and are determined with NRC-approved methodology which is referenced in the TS. These variables are moved from TS to the COLR to avoid the need for frequent revision of TS to change the value of those operating limits which cannot be specified to reasonably bound several operating cycles without significant loss of operating flexibility. In the Generic Letter, the NRC Staff notes that The processing of changes to TS that are developed using an NRC-approved methodology is an unnecessary burden on licensee and NRC resources. The NRC Staff also notes Licensees and applicants are encouraged to propose changes to TS that are consistent with the guidance.

A number of Westinghouse-designed plants have been permitted to remove the cycle-specific values in certain limiting conditions for operation (LCOs) from plant TS and to place them in a COLR, including values for the parameters addressed in TSTF-339.

As noted in the NRC approved WCAP-14483-P-A, Generic Methodology for Expanded Core Operating Limits Report, this philosophy and NRC approval of relocating the limiting values to the COLR is extended to the cycle-dependent Reactor Core Safety Limits parameters, the OTT and OPT setpoint parameters and to the limits on the DNB parameters of pressurizer pressure, RCS flow, and the RCS T-avg. NRC approval of TSTF-339, Revision 2, re-affirmed the NRC Staff agreement that these parameters are considered cycle-dependent and could be relocated to the COLR.

E-3 Enclosure to NL-23-0152 Evaluation of Proposed Change

Section 4.3 of WCAP-14483 provides the benefits of replacing the reactor core safety limits figure. Section 2.3 of WCAP-14483 provides the benefits of moving the DNB parameter limit values. Section 3.3 of WCAP-14483 provides the benefits of moving the OTT and OPT setpoint parameter values.

This license amendment request proposes to move these cycle-dependent Reactor Core Safety Limits parameters, the OTT and OPT setpoint parameters and to the limits on the DNB parameters of pressurizer pressure, RCS flow, and the RCS T-avg from the TS to the COLR, and to identify the pertinen t methodologies in the COLR TS 5.6.5.

During the review for the above changes, it was noted that the equation for the OPT Function 7 in Note 2 of Table 3.3.1-1 mis-identifies one of the terms. An administrative change is proposed to depict that term in the equation as it is used in the determination of the function setpoint.

2.4 Description of Proposed Change

The proposed change would modify the following for VEGP 1&2:

TS 2.1.1 Changes

x The Figure showing the limiting combinations of THERMAL POWER, RCS highest loop average temperature, and pressurizer pressure is moved to the COLR.

x Safety Limits are added for the DNB ratio and for peak fuel centerline temperature.

TS 3.3.1 Changes

x Table 3.3.1-1 Note 1 related to the OTT Function 6 of Table 3.3.1-1

Several parameter values are moved from the TS to the COLR including:

W1, W2 time constants utilized in lead-lag compensator for differential temperature:

W1 = 0 seconds, W2 = 0 seconds

W3 time constant utilized in lag compensator for differential temperature, d 6 seconds

K1 fundamental setpoint, d 114.9% RTP

K2 modifier for temperature, = 2.24% RTP per degree F

W4, W5 time constants utilized in lead-lag compensator for temperature compensation:

W4 t 28 seconds, W5 d 4 seconds

W6 time constant utilized in lag compensator for average temperature, d6 seconds

Tc indicated loop specific RCS average temperature at RTP, d588.4 degrees F

K3 modifier for pressure, = 0.177% RTP per psig

Pc reference pressure, t 2235 psig

E-4 Enclosure to NL-23-0152 Evaluation of Proposed Change

f1(AFD) modifier for Axial Flux Difference (AFD):

1. for AFD between -23% and +10%, = 0% RTP

2. for each % AFD is below -23%, the trip setpoint shall be reduced by 3.3% RTP

3. for each % AFD is above +10%, the trip setpoint shall be reduced by 1.95% RTP

(p) The compensated temperature difference [equation] shall be no more negative than 3 degrees F

The values for each of these parameters is replaced with an asterisk [*] and a Note is included following the listing of parameters identifying that the asterisk denotes the value is specified in the COLR.

x Table 3.3.1-1 Note 2 related to the OPT Function 7 of Table 3.3.1-1

Several parameter values are moved from the TS to the COLR including:

W1, W2 time constants utilized in lead-lag compensator for differential temperature: W1 = 0 seconds, W2 = 0 seconds

W3 time constant utilized in lag compensator for differential temperature, d 6 seconds

K4 fundamental setpoint, d 110% RTP

K5 modifier for temperature change, t 2% RTP per degree F for increasing temperature, t 0%

RTP per degree F for decreasing temperature

W7 time constant utilized in rate-lag compensator for temperature compensation, t 10 seconds

W6 time constant utilized in lag compensator for average temperature, d6 seconds

K6 modifier for temperature: 0.244% RTP per degree F for T > T, = 0% RTP for T T

T indicated loop specific RCS average temperature at RTP, d588.4 degrees F

f2(AFD) modifier for Axial Flux Difference (AFD), = 0% RTP for all AFD

The values for each of these parameters is replaced with an asterisk [*] and a Note is included following the listing of parameters identifying that the asterisk denotes the value is specified in the COLR.

Additionally, the denominator of the modifier for the K 6 term in the equation for the OPT Function 7 in Note 2 of Table 3.3.1-1 is revised to include the Laplace transform variable s as a multiplier for the W6 time constant utilized in lag compensator for average temperature.

E-5 Enclosure to NL-23-0152 Evaluation of Proposed Change

TS 3.4.1 Changes

x LCO 3.4.1

Several parameter values are moved from the TS Limiting Condition for Operation (LCO) and from the Surveillance Requirements (SRs) to the COLR including:

a. Pressurizer pressure 2199 psig;

b. RCS average temperature 592.5°F; and

c. RCS total flow rate 384,509 gpm.

To accomplish this revision, the LCO is changed to read:

LCO 3.4.1 RCS DNB parameters for pressurizer pressure, RCS average temperature, and RCS total flow rate shall be within the limits specified in the COLR.

x SR 3.4.1.1, SR 3.4.1.2, and SR 3.4.1.4

To accomplish this revision, the SRs are changed to read:

For SR 3.4.1.1, Verify pressurizer pressure is within the limits specified in the COLR.

For SR 3.4.1.2, Verify RCS average temperature is within the limits specified in the COLR.

And for SR 3.4.1.4, Verify by precision heat balance that RCS total flow rate is within the limits specified in the COLR.

TS 5.6.5 Changes

The listing of Specifications in TS 5.6.5.a is revised to include the following:

SL 2.1.1 Reactor Core Safety Limits LCO 3.3.1 Reactor Trip System (RTS) Instrumentation LCO 3.4.1 Reactor Coolant System (RCS) Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits

The WCAP-9272-P-A entry in TS 5.6.5.b is revised to identify that it also addresses the Methodology for the above Specifications by adding:

, Reactor Trip System Instrumentation, and Reactor Coolant System Pressure, Temperature, and Flow Departure from Nucleate Boiling Limits

The WCAPs listed in TS 5.6.5.b is revised to include:

E-6 Enclosure to NL-23-0152 Evaluation of Proposed Change

WCAP-8745-P-A, Design Basis for the Thermal Overpower T and Thermal Overtemperature T Trip Functions, September 1986 ( W Proprietary). (Methodology for Reactor Trip System Instrumentation.)

WCAP-11397-P-A, "Revised Thermal Design Procedure," April 1989 (W Proprietary).

(Methodology for Reactor Core Safety Limits and RCS Pressure, Temperature and Flow Departure from Nucleate Boiling Limits.)

TS Bases Changes (for information only)

Corresponding TS Bases changes are identified for TS 2.1.1, TS 3.3.1, and TS 3.4.1 that will be implemented in accordance with the TS Bases Change Program coincidently with the above TS changes. Thus, the Bases changes are provided for information only.

COLR Changes (for information only)

Corresponding COLR changes are identified for TS 2.1.1, TS 3.3.1, and TS 3.4.1 that will be implemented coincidently with the above TS changes. The COLR changes are cycle specific and thus cannot be developed until the cycle for implementation is known. Thus, these COLR markups represent typical expected changes and are provided for information only.

3. TECHNICAL EVALUATION

3.1 Basis for the Changes Southern Nuclear Company has reviewed TSTF-339, Rev. 2, and has concluded that the information in WCAP-14483-A-P and TSTF-339 is applicable to VEGP - Units 1 and 2 and justifies the proposed amendment for the incorporation of the changes to the VEGP

- Units 1 and 2 Technical Specifications, as discussed below.

Section 4.2 of WCAP-14483, which has been approved by the NRC, provides the basis for replacing the reactor core safety limits figure. Section 2.2 of WCAP-14483 provides the benefits of moving the DNB parameter limit values. Section 3.2 of WCAP-14483 provides the benefits of moving the OTT and OPT setpoint parameter values.

The administrative change to the equation for the Table 3.3.1-1 OPT Function 7 in Note 2 is consistent with the term as identified in Farley TS, in the TSTF-339-A, Revision 2, markups, in the Westinghouse Standard Technical Specifications NUREG-1431, Revision 5, and consistent with the actual use in the plant.

3.2 TSTF Change Traveler Adoption and Variations

TSTF-339-A Revision 2 was approved by the NRC on June 13, 2000. NUREG-1431, SL 2.1.1, LCO 3.3.1, Table 3.3.1-1, Notes 1 and 2, and LCO 3.4.1 and TS 3.4.1 Surveillance Requirements were subsequently revised to move the corresponding specific parameter values from the TS to the COLR.

E-7 Enclosure to NL-23-0152 Evaluation of Proposed Change

SNC seeks to adopt Revision 2 of TSTF-339-A with the following VEGP specific variations:

x TS 2.1.1 is proposed to be revised as identified in the TSTF with the exception that the 95/95 DNB criterion correlations are proposed to be determined per the methodologies identified in the COLR Specification 5.6.5 rather than including the specific value in the Safety Limits Specification. The DNB criterion correlations change with fuel type and design as noted in the current Bases where it identifies that the Vantage 5 fuel is analyzed using the WRB-2 correlation while the Lopar fuel is analyzed using the WRB-1 correlation, resulting in separate design limit DNBR values.

The Bases also note that separate design limit DNBR values are identified for the typical and thimble cells. Since the fuel type and design are subject to change with each future cycle, the DNB criterion correlation is subject to change with each future cycle. The variation is consistent with previous approvals for the Farley, Comanche Peak and North Anna operating plants. The Safety Evaluation of the Farley amendment states the staff finds that the proposed TS changes are acceptable, because existing staff guidance in this area has been satisfied and NRC approved methodologies have been applied.

x TS 3.3.1 Overtemperature Delta-T and Overpower Delta-T functions and terms are defined differently in the TSTF than in the previously approved TS 3.3.1 Note 1 and Note 2 for Table 3.3.1-1. However, both include terms with values that may vary by fuel cycle. Thus, the cycle-dependent terms are proposed to be moved to the COLR consistent with the TSTF.

x The administrative change to the equation for the Table 3.3.1-1 OPT Function 7 in Note 2 identified above is in addition to the changes identified in the TSTF.

x TS 3.4.1 is reformatted to simply identify that the specified parameters shall be with the limits specified in the COLR. Repeating the specified parameters in the LCO and identifying the values must be greater than or less than the limits (in the LCO and in the Surveillance Requirements) is unnecessary detail. This variation is consistent with the original guidance of Generic Letter 88-16 allowing relocation of limits that are developed using an NRC-approved methodology. SNC has noted that during review of the WCAP, the NRC Staff recommend that if RCS flow rate is relocated to the COLR, the minimum limit for RCS total flow based on a staff approved analysis (e.g.,

maximum tube plugging) should be retained in the Technical Specifications sim ilar to what is done for the positive limit on moderator temperature coefficient. However, the TSTF implementation was not done consistently with Generic Letter 88-16 guidance. First, the TSTF shows there are now two limits on RCS total flow, one retained in the TS and one included in the COLR. This effectively adds a second limit to the TS which is not consistent with the intent of Generic Letter 88-16, and it is not the wish or intent of SNC to add a second limit. Secondly, the WCAP recommendation appears to result from a concern for changes that would not be consistent with a staff approved analysis (e.g., maximum tube plugging) However, this concern was alternatively addressed with the updated 10 CFR 50.59 rule for evaluating changes that can be made without NRC prior approval. Since the time of this recommendation in the 1990s, Regulatory Guide 1.187, Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments, was issued in 2000 (endorsing NEI guidance) and revised in 2019, 2020, and 2021. Additionally, consistent with TSTF-449-A, Revision 4

E-8 Enclosure to NL-23-0152 Evaluation of Proposed Change

(approved by NRC in 2005), additional controls have been added to the Technical Specifications as shown in NUREG-1431, Revision 5. For Vogtle Units 1 and 2, these additional controls are implemented in TS 5.5.9, Steam Generator (SG) Program, and TS 5.6.10, Steam Generator Tube Inspection Report. TS 5.6.10.d requires An analysis summary of the tube integrity conditions predicted to exist at the next scheduled inspection (the forward-looking tube integrity assessment) relative to the applicable performance criteria, including the analysis methodology, inputs, and results. This analysis requires consideration of the methodology relative to the applicable performance criteria which would require consideration of the RCS flow input. The above requirements preclude revisions which would be inconsistent with approved methodologies impacting the minimum RCS total flow rate in the TS. Thus, consistent with the original intent of the Generic Letter and in consideration of improved implementation guidance, SNC requests to move the limit for RCS total flow from the TS to the COLR.

x TS 5.6.5 is revised. Specific revisions to TS 5.6.5 were not identified in the TSTF and only partially identified in WCAP-14483-P-A. However, with the affected TS 2.1.1, TS 3.3.1, and TS 3.4.1 would need to be listed in TS 5.6.5.a as requiring core operating limits to be established and documented in the COLR. In addition, the WCAP-14483 (Section 5.0) identifies that: This approach is supported by the existing NRC-approved reload methodologies, such as the Westinghouse reload methodology described in Reference 5 [WCAP-9272], which examines each of the DNB Parameter TS limits of RCS T-avg, RCS total flow rate and pressurizer pressure as well as the OTT and OPT setpoints and the supporting bases for the setpoints, on a cycle-specific basis. Thus, Specifications 3.3.1 and 3.4.1 would also need to be identified under the listing of WCAP-9272-P-A in TS 5.6.5.b. In addition, the WCAP does identify that WCAP-8745-P-A should also be added to the COLR listing of analytical methodologies used to determine the core operating limits in support of the OTTand the OPT Function input values to be included in the COLR. Finally, the Revised Thermal Design Procedure addressed in WCAP-11397-P-A is added to support the DNB methodology as identified in FSAR Subsection 4.4.1.1.2 for both TS 2.1.1 and TS 3.4.1.

These plant-specific variations have no adverse impact that precludes adoption of this change traveler TSTF-339-A Revision 2 for TS 2.1.1, TS 3.3.1, or TS 3.4.1.

3.3 Applicability of WCAP-14483-P-A, Safety Evaluation and Variations

TSTF-339-A Revision 2 is based on WCAP-14483-P-A. The Safety Evaluation (SE) provided by the NRC is applicable to the changes proposed in the license amendment request. The only variations from the WCAP are those proposed and approved in TSTF-339-A Revision 2 and the variations from the TSTF identified in the previous section.

The variations are consistent with the guidance of Generic Letter 88-16 and do not invalidate the SE discussions. There are no conditions or limitations identified in the SE.

E-9 Enclosure to NL-23-0152 Evaluation of Proposed Change

4 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria

SNC has reviewed TSTF-339-A and the NRC SE provided in the NRC letter to Westinghouse Electric Company dated January 19, 1999 (included in the WCAP-14483-P-A) that supported approval of WCAP-14483-P. The NRC concluded in the SE that the modified TS contained in WCAP-14483 are acceptable, and subsequently approved minor modifications to the WCAP changes via the TSTF revisions. Thus, the proposed change is consistent with the regulatory requirements identified in the SE and the subsequent supporting TSTF.

4.2 Precedent

Other plants with approved Specifications with similar cycle-dependent parameters relocated to a licensee-controlled COLR include:

x H.B. Robinson Steam Electric Plant [DPR-23; SER ML17039A153]

x Shearon Harris Nuclear Power Plant [NPF-63; SER ML17250A202]

x Comanche Peak Nuclear Power Plant [NPF-87; TS ML053180521]

x North Anna Power Station [NPF-4; TS ML052990145]

x Farley Nuclear Plant [NPF-2; SER ML013400451]

In addition, approval of the license amendment request will make the Vogtle Units 1 and 2 Technical Specifications more consistent with the corresponding Vogtle Units 3 and 4 Technical Specifications and Farley Units 1 and 2 Technical Specifications which include cycle dependent parameters in licensee-controlled documents. Similar fleet requirements tend to improve the consistency of human performance.

4.3 Significant Hazards Consideration

Pursuant to the provisions of Section 50.90 of Title 10 of the Code of Federal Regulations (10 CFR), Southern Nuclear Operating Company (SNC) hereby requests an amendment to Vogtle Electric Generating Plant (VEGP) Units 1 and 2 renewed facility operating licenses NPF-68 and NPF-81, respectively.

The proposed change would revise Technical Specifications (TS) 2.1.1, Reactor Core Safety Limits, (TS) 3.3.1, Reactor Trip System (RTS) Instrumentation, and TS 3.4.1, Reactor Coolant System (RCS) Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits, to move several cycle -specific parameter limits from the TS to the Core Operating Limits Report (COLR), and to administratively revise the equation for OPT Function 7 in Note 2 of Table 3.3.1-1.

Additionally, the proposed change modifies TS 5.6.5, Core Operating Limits Report (COLR), to include changes reflecting the above application of WCAP-14483-P-A by

E-10 Enclosure to NL-23-0152 Evaluation of Proposed Change

appropriately modifying the list of the NRC approved methodologies used to develop the cycle specific COLR.

SNC 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 amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No

The proposed change moves some parameter limit values from the Technical Specifications (TS) to a cycle-specific Core Operating Limits Report (COLR) but does not remove the required designation of the limits. Operation within the limits continues to maintain the assumptions for initial conditions of key parameter values in the safety analyses as valid and does not result in actions that would increase the probability or consequences of any accident previously evaluated.

Therefore, the proposed change does 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 change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed). Operation in accordance with the revised TS and the required parameter limits precludes new challenges to systems, structures, or components that might introduce a new type of accident. Applicable design and performance criteria will continue to be met and no new single failure mechanisms will be created. The proposed change does not involve the alteration of plant equipment or introduce unique operational modes or accident precursors.

Therefore, the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The margin of safety is related to the ability of the fission product barriers to perform their design functions during and following an accident. These barriers include the fuel cladding, the reactor coolant system, and th e containment. The performance of these fission product barriers is not adversely affected by the proposed change.

The proposed change moves some parameter limit values from the Technical Specifications (TS) to a cycle-specific Core Operating Limits Report (COLR) but does

E-11 Enclosure to NL-23-0152 Evaluation of Proposed Change

not remove the required designation of the limits. The limits are not changed and the methodology to determine the limits is not changed. Operation in accordance with the revised TS maintains the assumptions for initial conditions of key parameter values in the safety analyses as valid. This confirms applicable design and performance criteria associated with the safety analysis will continue to be met and that the margin of safety is not adversely affected.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, SNC concludes that the proposed amendment does not involve a 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.

4.4 Conclusions

In conclusion, based on the considerations discussed herein, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions 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.

5 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 Part 20, and would change an inspection or surveillance requirement.

However, the proposed change 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 off site, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criteria 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.

6 REFERENCES

1. Westinghouse Topical Report WCAP-14483-P-A, Generic Methodology for Expanded Core Operating Limits Report, Approved January 1999 [ML020430092]

2. Technical Specification Task Force (TSTF) Traveler 339, Revision 2, Approved June 2000

[ML003723269]

E-12 to NL-23-0152 VEGP - 1 and 2 Marked-up TS Pages

Attachment 1

Vogtle Electric Generating Plant - Units 1 and 2 Marked-up TS Pages SLs 2.0

2.0 SAFETY LIMITS (SLs)

2.1 SLs

2.1.1 Reactor Core SLs

In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the SLs specified in Figure 2.1.1-1.

INSERT 2.1.1-1 2.1.2 RCS Pressure SLlimits

In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained 2735 psig.

2.2 SL Violations

2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2 If SL 2.1.2 is violated:

2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.

Vogtle Units 1 and 2 2.0-1 Amendment No. 161 (Unit 1)

Amendment No. 143 (Unit 2)

INSERT 2.1.1-1 the COLR; and the following SLs shall not be exceeded:

2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained greater than or equal to the 95/95 DNBR criterion for the DNB criterion correlations and methodologies specified in Specification 5.6.5. [1.17 for the WRB-l/WRB-2 DNB correlations].

2.1.1.2 The peak fuel centerline temperature shall be maintained < [5080'F, decreasing by 58°F per 10,000 MWD/MTU of burnup].

Markup shows variation from TSTF-339-A Rev. 2 SLs Remove this figure. 2.0

Figure 2.1.1-1 Reactor Core Safety Limits

Vogtle Units 1 and 2 2.0-2 Amendment No. 128 (Unit 1)

Amendment No. 106 (Unit 2)

RTS Instrumentation 3.3.1

Table 3.3.1-1 (page 7 of 9)

Reactor Trip System Instrumentation

Note 1: Overtemperature Delta-T

The Allowable Value of each input to the Overtemperature Delta-T function as defined by the equation below shall not exceed its as-left value by more than the following:

(1) 0.5% T span for the T channel (2) 0.5% T span for the Tavg channel (3) 0.5% T span for the pressurizer pressure channel (4) 0.5% T span for the f1(AFD) channel

T{1+ 1s}1º{1+ s}1º(p)º

<<100 4>> <<K1-K 2<<T-T>>-K 3{P-P}-f 1(AFD)>>

¬ T0 {1 + 2s} {1+ 3s}1/4 ¬ {1 + 5s} ¬ {1+ 6s} 1/4 1/4

Where: T measured loop specific RCS differential temperature, degrees F

T0 indicated loop specific RCS differential at RTP, degrees F

1+ 1s lead-lag compensator on measured differential temperature 1+ 2s

1, 2 time constants utilized in lead-lag compensator for differential temperature: 1 = 0 seconds, 2 = 0 seconds

1

1+ 3s lag compensator on measured differential temperature

3 time constant utilized in lag compensator for differential temperature, 6 seconds K1 fundamental setpoint, 114.9% RTP *

K2 modifier for temperature, = 2.24% RTP per degree F 1+ s

4 1+ 5s lead-lag compensator on dynamic temperature compensation

4, 5 time constants utilized in lead-lag compensator for temperature compensation: 4 28 seconds, 5 4 seconds

T measured loop specific RCS average temperature, degrees F

1 1+ 6s lag compensator on measured average temperature 6 time constant utilized in lag compensator for average temperature, 6 seconds

T indicated loop specific RCS average temperature at RTP, 588.4 degrees F K3 modifier for pressure, = 0.177% RTP per psig

P measured RCS pressurizer pressure, psig

P reference pressure, 2235 psig s Laplace transform variable, inverse seconds

Vogtle Units 1 and 2 3.3.1-20 Amendment No. 165 (Unit 1)

Amendment No. 147 (Unit 2)

RTS Instrumentation 3.3.1

Table 3.3.1-1 (page 8 of 9)

Reactor Trip System Instrumentation

Note 1: Overtemperature Delta-T (continued)

f1(AFD) modifier for Axial Flux Difference (AFD):

1. for AFD between -23% and +10%, = 0% RTP* * *

2. for each % AFD is below -23%, the trip setpoint shall be reduced by 3.3% RTP

3. for each % AFD is above +10%, the trip setpoint shall be reduced by 1.95% RTP

{ 4 >>1 + s} 1º (p) The compensated temperature difference <<T - T shall be no more negative than 3 degrees F. 3

{1 + 5 s} ¬ {1 + 6 s} 1/4 The values denoted with [*] are specified in the COLR.

Note 2: Overpower Delta-T

The Allowable Value of each input to the Overpower Delta-T function as defined by the equation below shall not exceed its as-le ft value by more than the following:

(1) 0.5% T span for the T channel (2) 0.5% T span for the Tavg channel

T{1+ 1s}1º{s}1º1ºº

<<100 >> <<K 4-<<K 57T>>-K 6<<T-T>>-f 2(AFD)>>

¬ T0 {1 + 2 s} {1+ 3s}1/4 << << {1 + s} {1+ s} >> << {1 + } >> >>

¬ ¬ 7 6 1/4 ¬ 6 1/4 1/4

Where: T measured loop specific RCS differential temperature, degrees F s

T0 indicated loop specific RCS differential at RTP, degrees F

1+ 1s lead-lag compensator on measured differential temperature 1+ 2s

1, 2 time constants utilized in lead-lag compensator for differential temperature: 1 = 0 seconds, 2 = 0 seconds

1

1+ 3s lag compensator on measured differential temperature

3 time constant utilized in lag compensator for differential temperature, 6 seconds K4 fundamental setpoint, 110% RTP *

K5 modifier for temperature change: 2% RTP per degree F for increasing temperature, 0% RTP per degree F for decreasing temperature

7s 1+ 7s rate-lag compensator on dynamic temperature compensation

7 time constant utilized in rate-lag compensator for temperature compensation, 10 seconds

T measured loop specific RCS average temperature, degrees F

1 1+ 6s lag compensator on measured average temperature

Vogtle Units 1 and 2 3.3.1-21 Amendment No. 165 (Unit 1)

Amendment No. 147 (Unit 2)

RTS Instrumentation 3.3.1

Table 3.3.1-1 (page 9 of 9)

Reactor Trip System Instrumentation

Note 2: Overpower Delta-T (continued) 6 time constant utilized in lag compensator for average temperature, 6 seconds

K6 modifier for temperature: 0.244% RTP per degree F for T > T, = 0% RTP for T T T indicated loop specific RCS average temperature at RTP, 588.4 degrees F s Laplace transform variable, inverse seconds

f2(AFD) modifier for Axial Flux Difference (AFD), = 0% RTP for all AFD

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

Vogtle Units 1 and 2 3.3.1-22 Amendment No. 165 (Unit 1)

Amendment No. 147 (Unit 2)

RCS Pressure, Temperature, and Flow DNB Limits 3.4.1

3.4 REACTOR COOLANT SYSTEM (RCS)

3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits

LCO 3.4.1 RCS DNB parameters for pressurizer pressure, RCS average temperature, and RCS total flow rate shall be within the limits specified below:

a. Pressurizer pressure 2199 psig; in the COLR.

b. RCS average temperature 592.5°F; and

c. RCS total flow rate 384,509 gpm.

APPLICABILITY: MODE 1.

NOT E---------------------------------------------------

Pressurizer pressure limit does not apply during:

a. THERMAL POWER ramp > 5% RTP per minute; or

b. THERMAL POWER step > 10% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME

A. One or more RCS DNB A.1 Restore RCS DNB 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months parameters not within parameter(s) to within limits. limit.

B. RCS total flow rate B.1. Perform SR 3.4.1.4. 7 days degraded.

C. Required Action and C.1 Be in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met.

Vogtle Units 1 and 2 3.4.1-1 Amendment No. 96 (Unit 1)

Amendment No. 74 (Unit 2)

RCS Pressure, Temperature, and Flow DNB Limits 3.4.1

SURVEILLANCE REQUIREMENTS

SURVEILLANCE FREQUENCY

SR 3.4.1.1 Verify pressurizer pressure is 2199 psig. In accordance with the Surveillance within the limits Frequency Control specified in the COLR. Program

SR 3.4.1.2 Verify RCS average temperature is 592.5°F. In accordance with the Surveillance within the limits Frequency Control specified in the COLR. Program

SR 3.4.1.3 Monitor RCS total flow rate for degradation. In accordance with the Surveillance Frequency Control Program

SR 3.4.1.4 ---------------------------NOTE-------------------------------

Not required to be performed until 7 days after 90% RTP.

Verify by precision heat balance that RCS total In accordance with flow rate is 384,509 gpm. the Surveillance Frequency Control within the limits specified in the COLR. Program

Vogtle Units 1 and 2 3.4.1-2 Amendment No. 158 (Unit 1)

Amendment No. 140 (Unit 2)

Reporting Requirements 5.6

5.6 Reporting Requirements (continued)

5.6.5 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:

LCO 3.1.1 "SHUTDOWN MARGIN" Insert 5.6.5-1 LCO 3.1.3 "Moderator Temperature Coefficient" LCO 3.1.5 "Shutdown Bank Insertion Limits" LCO 3.1.6 "Control Bank Insertion Limits" LCO 3.2.1 "Heat Flux Hot Channel Factor" LCO 3.2.2 "Nuclear Enthalpy Rise Hot Channel Factor" LCO 3.2.3 "Axial Flux Difference" Insert 5.6.5-2 LCO 3.9.1 "Boron Concentration"

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

WCAP-9272-P-A, "WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY," July 1985 (W Proprietary). (Methodology for Moderator Temperature Coefficient, Shutdown Bank Insertion Limit, Control Bank Insertion Limits, and Nuclear Enthalpy Rise Hot Channel Factor.) Insert 5.6.5-3

WCAP-10216-P-A, Revision 1A, "RELAXATION OF CONSTANT AXIAL OFFSET CONTROL FQ SURVEILLANCE TECHNICAL SPECIFICATION," February, 1994 (W Proprietary). (Methodology for Axial Flux Difference (Relaxed Axial Offset Control) and Heat Flux Hot Channel Factor (W(Z) surveillance requirements for F Q Methodology).)

WCAP-10266-P-A, Revision 2, "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code," March 1987.

(W Proprietary) (Methodology for Axial Flux Difference (Relaxed Axial Offset Control) and Heat Flux Hot Channel Factor (W(Z) surveillance requirements for F Q Methodology).)

WCAP-13749-P-A, Safety Evaluation Supporting the Conditional Exemption of the Most Negative EOL Moderator Temperature Coefficient Measurement, March 1997.

WCAP-16045-P-A, Qualification of the Two-Dimensional Transport Code PARAGON, August 2004 (Methodology for Moderator Temperature Coefficient.)

WCAP-16045-P-A, Addendum 1-A, Qualification of the NEXUS Nuclear Data Methodology, August 2007 (Methodology for Moderator Temperature Coefficient.)

(continued)

Vogtle Units 1 and 2 5.6-3 Amendment No. 182 (Unit 1)

Amendment No. 163 (Unit 2)

INSERT 5.6.5-1 6 L2.1.1 Reactor Core Safety Limits

INSERT 5.6.5-2 LCO 3.3.1 Reactor Trip System 576Instrumentation LCO 3.4.1 Reactor Coolant System (RCS) Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits

INSERT 5.6.5-3

, Reactor Trip System Instrumentation, and Reactor Coolant System Pressure, Temperature, and Flow Departure from Nucleate Boiling Limits.)

Reporting Requirements 5.6

5.6 Reporting Requirements (continued)

5.6.5 Core Operating Limits Report (COLR) (continued)

WCAP-12610-P-A, VANTAGE+ Fuel Assembly Reference Core Report, April 1995 (Westinghouse Proprietary). (Methodology for Axial Flux Difference (Relaxed Axial Offset Control) and Heat Flux Hot Channel Factor (W(Z) surveillance requirements for F Q Methodology).)

WCAP-12610-P-A & CENP D-404-P-A, Addendum 1-A, Optimized ZIRLO TM, July 2006 (Westinghouse Pro prietary). (Methodology for Axial Flux Difference (Relaxed Axial Offset Control) and Heat Flux Hot Channel Insert 5.6.5-4 Factor (W(Z) surveillance requirements for F Q Methodology).)

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

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

5.6.6 Reactor Coolant System (RCS) PR ESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

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

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

b. The power operated relief valve lift settings required to support the Cold Overpressure Protection Systems (COPS) and the COPS arming temperature shall be established and documented in the PTLR for the following:

LCO 3.4.12 "Cold Overpressure Protection Systems"

c. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1. WCAP-14040-A, Rev. 4, Methodology Used to Develop Cold Ove rpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves.

(continued)

Vogtle Units 1 and 2 5.6-4 Amendment No. 182 (Unit 1)

Amendment No. 163 (Unit 2)

INSERT 5.6.5-4 WCAP-8745-P-A, Design Bases for the Thermal Overpower T and Thermal Overtemperature T Trip Functions, September 1986 (W Proprietary). (Methodology for Reactor Trip System Instrumentation.)

WCAP-11397-P-A, "Revised Thermal Design Procedure," April 1989 (W Proprietary).

(Methodology for Reactor Core Safety Limits and RCS Pressure, Temperature and Flow Departure from Nucleate Boiling Limits.) to NL-23-0152 VEGP - 1 and 2 Revised TS Pages

Attachment 2

Vogtle Electric Generating Plant - Units 1 and 2 Revised TS Pages SLs 2.0 2.0 SAFETY LIMITS (SLs)

2.1 SLs

2.1.1 Reactor Core SLs

In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the limits specified in the COLR; and the following SLs shall not be exceeded:

2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained greater than or equal to the 95/95 DNBR criterion for the DNB criterion correlations and methodologies specified in Specification 5.6.5.

2.1.1.2 The peak fuel centerline temperature shall be maintained < 5080°F, decreasing by 58°F per 10,000 MWD/MTU of burnup.

2.1.2 RCS Pressure SL

In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained d 2735 psig.

2.2 SL Violations

2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2 If SL 2.1.2 is violated:

2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.

Vogtle Units 1 and 2 2.0-1 Amendment No. (Unit 1)

Amendment No. (Unit 2)

RTS Instrumentation 3.3.1

Table 3.3.1-1 (page 7 of 9)

Reactor Trip System Instrumentation

Note 1: Overtemperature Delta-T

The Allowable Value of each input to the Overtemperature Delta-T function as defined by the equation below shall not exceed its as-left value by more than the following:

(1) 0.5% ' T span for the ' T channel (2) 0.5% ' T span for the Tavg channel (3) 0.5% ' T span for the pressurizer pressure channel (4) 0.5% ' T span for the f1(AFD) channel c T{1+ 1s}1º{1+ s}1º(p)º

<<100 4>>d <<K1-K 2<<T-Tc>>-K 3{P-P}-f 1(AFD)>>

¬ T0 {1 + 2s} {1+ 3s}1/4 ¬ {1 + 5s} ¬ {1+ 6s} 1/4 1/4

Where: ' T measured loop specific RCS differential temperature, degrees F

' T0 indicated loop specific RCS differential at RTP, degrees F

1+W1s lead-lag compensator on measured differential temperature 1+W2s

W1, W2 time constants utilized in lead-lag compensator for differential temperature: W1 =

seconds, W2 =

seconds

1

1+W3s lag compensator on measured differential temperature

W3 time constant utilized in lag compensator for differential temperature, d

seconds

K1 fundamental setpoint, d *% RTP

K2 modifier for temperature, = *% RTP per degree F

1+W4s 1+W5s lead-lag compensator on dynamic temperature compensation

W4, W5 time constants utilized in lead-lag compensator for temperature compensation: W4 t

seconds, W5 d

seconds

T measured loop specific RCS average temperature, degrees F

1

1+W6s lag compensator on measured average temperature

W6 time constant utilized in lag compensator for average temperature, d

seconds

Tc indicated loop specific RCS average temperature at RTP, d

degrees F

K3 modifier for pressure, = *% RTP per psig

P measured RCS pressurizer pressure, psig

Pc reference pressure, t

psig

s Laplace transform variable, inverse seconds

Vogtle Units 1 and 2 3.3.1-20 Amendment No. (Unit 1)

Amendment No. (Unit 2)

RTS Instrumentation 3.3.1

Table 3.3.1-1 (page 8 of 9)

Reactor Trip System Instrumentation

Note 1: Overtemperature Delta-T (continued)

f1(AFD) modifier for Axial Flux Difference (AFD):

1. for AFD between *% and *%, = *% RTP

2. for each % AFD is below *%, the trip setpoint shall be reduced by *% RTP

3. for each % AFD is above *%, the trip setpoint shall be reduced by *% RTP

{ 4 >>1 + Ws} c1º (p) The compensated temperature difference <<T - T shall be no more negative than

degrees F.

{1 + W5 s} ¬ {1 + W6 s} 1/4

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

Note 2: Overpower Delta-T

The Allowable Value of each input to the Overpower Delta-T function as defined by the equation below shall not exceed its as-left value by more than the following:

(1) 0.5% ' T span for the ' T channel (2) 0.5% ' T span for the Tavg channel 100 ' T 1 W s 1 K K W 1 K 1 " f

' T 1 W s1 W s 1 W 1 W s 1 W

Where: ' T measured loop specific RCS differential temperature, degrees F

' T0 indicated loop specific RCS differential at RTP, degrees F

1+W1s lead-lag compensator on measured differential temperature 1+W2s

W1, W2 time constants utilized in lead-lag compensator for differential temperature: W1 =

seconds, W2 =

seconds

1

1+W3s lag compensator on measured differential temperature

W3 time constant utilized in lag compensator for differential temperature, d

seconds

K4 fundamental setpoint, d *% RTP

K5 modifier for temperature change: t *% RTP per degree F for increasing temperature, t *% RTP per degree F for decreasing temperature

W7s

1+W7s rate-lag compensator on dynamic temperature compensation

W7 time constant utilized in rate-lag compensator for temperature compensation, t

seconds

T measured loop specific RCS average temperature, degrees F

1

1+W6s lag compensator on measured average temperature

Vogtle Units 1 and 2 3.3.1-21 Amendment No. (Unit 1)

Amendment No. (Unit 2)

RTS Instrumentation 3.3.1

Table 3.3.1-1 (page 9 of 9)

Reactor Trip System Instrumentation

Note 2: Overpower Delta-T (continued)

W6 time constant utilized in lag compensator for average temperature, d

seconds

K6 modifier for temperature: t *% RTP per degree F for T > Ts, = *% RTP for T d Ts

Ts indicated loop specific RCS average temperature at RTP, d

degrees F

s Laplace transform variable, inverse seconds

f2(AFD) modifier for Axial Flux Difference (AFD), =

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

Vogtle Units 1 and 2 3.3.1-22 Amendment No. (Unit 1)

Amendment No. (Unit 2)

RCS Pressure, Temperature, and Flow DNB Limits 3.4.1

3.4 REACTOR COOLANT SYSTEM (RCS)

3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits

LCO 3.4.1 RCS DNB parameters for pressurizer pressure, RCS average temperature, and RCS total flow rate shall be within the limits specified in the COLR.

APPLICABILITY: MODE 1.

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

Pressurizer pressure limit does not apply during:

a. THERMAL POWER ramp > 5% RTP per minute; or

b. THERMAL POWER step > 10% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME

A. One or more RCS DNB A.1 Restore RCS DNB 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months parameters not within parameter(s) to within limits. limit.

B. RCS total flow rate B.1. Perform SR 3.4.1.4. 7 days degraded.

C. Required Action and C.1 Be in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met.

Vogtle Units 1 and 2 3.4.1-1 Amendment No. (Unit 1)

Amendment No. (Unit 2)

RCS Pressure, Temperature, and Flow DNB Limits 3.4.1

SURVEILLANCE REQUIREMENTS

SURVEILLANCE FREQUENCY

SR 3.4.1.1 Verify pressurizer pressure is within the limits In accordance with specified in the COLR. the Surveillance Frequency Control Program

SR 3.4.1.2 Verify RCS average temperature is within the In accordance with limits specified in the COLR. the Surveillance Frequency Control Program

SR 3.4.1.3 Monitor RCS total flow rate for degradation. In accordance with the Surveillance Frequency Control Program

SR 3.4.1.4 ---------------------------NOTE-------------------------------

Not required to be performed until 7 days after t 90% RTP.

Verify by precision heat balance that RCS total In accordance with flow rate is within the limits specified in the COLR. the Surveillance Frequency Control Program

Vogtle Units 1 and 2 3.4.1-2 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Reporting Requirements 5.6

5.6 Reporting Requirements (continued)

5.6.5 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:

SL 2.1.1 Reactor Core Safety Limits LCO 3.1.1 "SHUTDOWN MARGIN" LCO 3.1.3 "Moderator Temperature Coefficient" LCO 3.1.5 "Shutdown Bank Insertion Limits "

LCO 3.1.6 "Control Bank Insertion Limi ts" LCO 3.2.1 "Heat Flux Hot Channel Facto r" LCO 3.2.2 "Nuclear Enthalpy Rise Hot Channel Facto r" LCO 3.2.3 "Axial Flux Difference" LCO 3.3.1 Reactor Trip System (RTS) Instrumentation LCO 3.4.1 Reactor Coolant System (RCS) Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits LCO 3.9.1 "Boron Concentration"

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

WCAP-9272-P-A, "WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY," July 1985 (W Proprietary). (Methodology for Moderator Temperature Coefficient, Shutdown Bank Insertion Limit, Control Bank Insertion Limits, Nuclear Enthalpy Rise Hot Channel Factor, Reactor Trip System Instrumentation, and Reactor Coolant System Pressure, Temperature, and Flow Departure from Nucleate Boiling Limits.)

WCAP-10216-P-A, Revision 1A, "R ELAXATION OF CONSTANT AXIAL OFFSET CONTROL FQ SURVEILLANCE TECHNICAL SPECIFICATION," February, 1994 (W Proprietary). (Methodology for Axial Flux Difference (Relaxed Axial Offset Control) and Heat Flux Hot Channel Factor (W(Z) surveillance requirements for F Q Methodology).)

WCAP-10266-P-A, Revision 2, "The 19 81 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code," March 1987.

(W Proprietary) (Methodology for Axial Flux Difference (Relaxed Axial Offset Control) and Heat Flux Hot Channel Factor (W(Z) surveillance requirements for F Q Methodology).)

WCAP-13749-P-A, Safety Evaluation Supporting the Conditional Exemption of the Most Negative EOL Moderator Temperature Coe fficient Measurement, March 1997.

(continued)

Vogtle Units 1 and 2 5.6-3 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Reporting Requirements 5.6

5.6 Reporting Requirements

5.6.5 Core Operating Limits Report (COLR) (continued)

WCAP-16045-P-A, Qualification of the Two-Dimensional Transport Code PARAGON, August 2004 (Methodology for Moderator Temperature Coefficient.)

WCAP-16045-P-A, Addendum 1-A, Qualific ation of the NEXUS Nuclear Data Methodology, August 2007 (Methodology for Moderator Temperature Coefficient.)

WCAP-12610-P-A, VANTAGE+ Fuel A ssembly Reference Core Report, April 1995 (Westinghouse Proprietary). (Methodology for Axial Flux Difference (Relaxed Axial Offset Control) and Heat Flux Hot Channel Factor (W(Z) surveillance requirements for F Q Methodology).)

WCAP-12610-P-A & CENP D-404-P-A, Addendum 1-A, Optimized ZIRLO TM, July 2006 (Westinghouse Pro prietary). (Meth odology for Axial Flux Difference (Relaxed Axial Offset Control) and Heat Flux Hot Channel Factor (W(Z) surveillance requirements for F Q Methodology).)

WCAP-8745-P-A, Design Bases for the Thermal Overpower T and Thermal Overtemperature T Trip Functions, September 1986 (W Proprietary). (Methodology for Reactor Trip System Instrumentation.)

WCAP-11397-P-A, Revised Thermal Design Procedure, April 1989 (W Proprietary). (Methodology for Reactor Core Safety Limits and RCS Pressure, Temperature and Flow Departure from Nucleate Boiling Limits

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits,

Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safe ty analysis are met.

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

5.6.6 Reactor Coolant System (RCS) PRESS URE AND TEMPERATURE LIMITS REPORT (PTLR)

a. RCS pressure and temperature limits for heatup, cooldown, operation, criticality, and hydrostatic testing as well as heatup and cooldown rate s shall b e established and documented in the PTLR for the following :

(continued)

Vogtle Units 1 and 2 5.6-4 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Reporting Requirements 5.6

5.6 Reporting Requirements

5.6.6 Reactor Coolant System (RCS) PR ESSURE AND TEMPERATURE LIMITS REPORT (PTLR) (continued)

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

b. The power operated relief valve lift settings required to support the Cold Ove rpressure Protection Systems (COPS) and the COPS arming temperature shall be established and documented in the PTLR for the following:

LCO 3.4.12 "Cold Overpressure Protection Systems"

c. The analytical methods used to determine the RCS pres sure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1. WCAP-14040-A, Rev. 4, Methodology Used to Develop Cold Ove rpressure Mitigating System Setpoints and RCS Heatup an d Cool down Limit Curves.

2. WCAP-16142-P, Rev. 1, Reactor Vessel Closure Head/ Vessel Flang e Requirements Evaluation for Vogtle Units 1 and 2.

3. The PTLR will contain the complete identification for each of the TS reference Topical Reports used to prepare the PTLR (i.e.,

report number, title, revision, date, and any supplements).

d. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fl uence period and for any revision or supplement thereto.

5.6.7 EDG Failure Report

If an individual emergency diesel generator (EDG) experiences four or more valid failures in the last 25 demands, these failures and any nonvalid failures experienced by that EDG in that time period shall be reported within 30 days.

Reports on EDG failures shall include the information recommended in Regulatory Guide 1.9, Revision 3, Regulatory Position C.4, or existing Regulatory Guide 1.108 reporting requirement.

(continued)

Vogtle Units 1 and 2 5.6-5 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Reporting Requirements 5.6

5.6 Reporting Requirements (continued)

5.6.8 PAM Report

When a Report is required by Condition G or J of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.9 Dele ted.

5.6.10 Steam Generator Tube Inspection Report

A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.9, Steam Generator (SG) Program. The report shall include:

a. The scope of inspections performed on each SG:

b. The nondestructive examination techniques utilized for tubes with increased degrada tion susceptibility;

c. For each degradation mechanism found:

1. The nondestructive examination techniques utilized;

2. The location, orientation (if linear), measured size (if available), and voltage response for each indication. For tube wear at support structures less than 20 percent through-wall, only the total number of indications needs to be reported;

3. A description of the condition monitoring assessment and result s, including the margin to the tube integrity performance criteria and comparison with the margin predicted to exist at the inspection by the previous forward-looking tube integrity assessment; and

4. The number of tubes plugged during the inspection outage.

d. An analysis summary of the tube integrity conditions predicted to exist at the next scheduled inspection (the forward-looking tube integrity assess ment) relative to the applicable performance criteria, including the analysis methodology, inputs, and results;

e. The number and percentage of tubes plugged to date, and the effective plugging percentage in each SG (continued)

Vogtle Units 1 and 2 5.6-6 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Reporting Requirements 5.6

5.6 Reporting Requirements

5.6.10 Steam Generator Tube Inspection Report (continued)

f. The results of any SG secondary side inspections;

g. The primary to secondary LEAKAGE rate observed in each SG (if it is not practical to assign the LEAKAGE to an individual SG, the entire primary to secondary LEAKAGE should be conservatively assumed t o be from one SG) during the cycle preceding t he inspection which is the subject of the report;

h. The calculated accident induced leakage rate from the portion of the tubes below 15.2 inches from the top of the tubesheet for the most limiting SG. In addition, if the calculated accident induced leakage rate from the most limiting accident is less than 2.48 times the maximum operational primary to secondary leakage rate, the report should describe how it was determined; and

i. the results of monitoring for tube axial displacement (slippage). If slippage is discovered,

the implications of the discover and corrective action shall be provided.

Vogtle Units 1 and 2 5.6-7 Amendment No. (Unit 1)

Amendment No. (Unit 2) to NL-23-0152 VEGP - 1 and 2 Marked-up TS Bases Pages (for information Only)

Attachment 3

Vogtle Electric Generating Plant - Units 1 and 2 Marked-up TS Bases Pages (for information only)

Reactor Core SLs B 2.1.1

B 2.0 SAFETY LIMITS (SLs)

B 2.1.1 Reactor Core SLs

BASES

BACKGROUND GDC 10 (Ref. 1) requires that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs). This is accomplished by having a departure from nucleate boiling (DNB) design basis, which corresponds to a 95% probability at a 95%

confidence level (the 95/95 DNB criterion) that DNB will not occur on the limiting fuel rods and by requiring that fuel centerline temperature stays below the melting temperature.

The restrictions of this SL prevent overheating of the fuel and cladding, as well as possible cladding perforation, that would result in the release of fission products to the reactor coolant. Overheating of the fuel is prevented by maintaining the steady state peak linear heat rate (LHR) below the level at which fuel centerline melting occurs. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime, where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Fuel centerline melting occurs when the local LHR, or power peaking, in a region of the fuel is high enough to cause the fuel centerline temperature to reach the melting point of the fuel. Expansion of the pellet upon centerline melting may cause the pellet to stress the cladding to the point of failure, allowing an uncontrolled release of activity to the reactor coolant.

Operation above the boundary of the nucleate boiling regime could result in excessive cladding temperature because of the onset of DNB and the resultant sharp reduction in heat transfer coefficient. Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.

(continued)

Vogtle Units 1 and 2 B 2.1.1-1 Revision No. 0 Reactor Core SLs B 2.1.1

BASES

BACKGROUND The proper functioning of the Reactor Protection System (continued) (RPS) and main steam safety valves prevents violation of the reactor core SLs.

APPLICABLE The fuel cladding must not sustain damage as a result of normal SAFETY ANALYSES operation and AOOs. The reactor core SLs are established to preclude violation of the following fuel design criteria:

a. The hot fuel pellet in the core must not experience centerline fuel melting; and

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

In meeting the DNB design criterion, uncertainties in plant operating parameters, nuclear and thermal parameters, fuel fabrication parameters, and computer codes must be considered. As described in the FSAR, the effects of these uncertainties have been statistically combined with the correlation uncertainty to determine design limit DNBR values that satisfy the DNB design criterion. The Vantage 5 fuel is analyzed using the W RB-2 correlation with design limit DNBR values of 1.24 and 1.23 for the typical and thimble cells, respectively.

The Lopar fuel is analyzed using the W RB-1 correlation with design limit DNBR values of 1.23 and 1.22 for the typical and thimble cells, respectively.

Additional DNBR margin is maintained by performing the safety analyses to a higher DNB limit. This margin between the design and safety analysis limit DNBR values is used to offset known DNBR penalties (e.g., rod bow and transition core) and to provide DNBR margin for operating and design flexibility.

The Reactor Trip System setpoints (Ref. 2), in combination with all the LCOs, are designed to prevent any anticipated combination of transient conditions for Reactor Coolant System (RCS) temperature, pressure, and THERMAL POW ER level that would result in a departure from nucleate boiling ratio

RCS flow, Axial Flux Difference

(continued)

Vogtle Units 1 and 2 B 2.1.1-2 Revision No. 0 Reactor Core SLs B 2.1.1

BASES

APPLICABLE (DNBR) of less than the DNBR limit and preclude the SAFETY ANALYSES existence of flow instabilities.

(continued)

Automatic enforcement of these reactor core SLs is provided by the following functions:

a. High pressurizer pressure trip;appropriate operation of the RPS and main

b. Low pressurizer pressure trip;steam safety valves.

c. Overtemperature T trip;

d. Overpower T trip;

e. Power Range Neutron Flux trip;

f. Reactor Coolant Flow trips (including undervoltage and underfrequency of the reactor coolant pump buses); and

g. Main steam safety valves.

The limitation that the average enthalpy in the hot leg be less than or equal to the enthalpy of saturated liquid also ensures that the T measured by instrumentation, used in the RPS design as a measure of core power, is proportional to core power.

The SLs represent a design requirement for establishing the RPS trip setpoints identified previously. LCO 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits," or the assumed initial conditions of the safety analyses (as indicated in the FSAR, Ref. 2) provide more restrictive limits to ensure that the SLs are not exceeded.

figure the COLR shows

SAFETY LIMITS The curves provided in Figure B 2.1.1-1 show the loci of points of THERMAL POW ER (NI-0041, NI-0042, NI-0043, NI-0044, TDI-0411A, TDI-0421A, TDI-0431A, TDI-0441A), RCS Pressure (PI-0455A, B, and C, PI-0456, PI-0456A, PI-0457, PI-0457A, PI-0458, and PI-0458A), and average temperature (TI-0412, TI-0422, TI-0432, TI-0442) for which the minimum DNBR is not less than the safety analyses limit, that fuel

(continued)

Vogtle Units 1 and 2 B 2.1.1-3 Revision No. 0 Reactor Core SLs B 2.1.1

BASES

SAFETY LIMITS centerline temperature remains below melting, that the average (continued) enthalpy in the hot leg is less than or equal to the enthalpy of saturated liquid, or that the exit quality is within the limits defined by the DNBR correlation.

Bases Insert 2.0-1 The curves are based on enthalpy hot channel factor limits provided in the COLR. The dashed line of Figure B 2.1.1-1 shows an example of a limit curve at 2235 psig. In addition, it illustrates the various RPS functions that are designed to prevent the unit from reaching the limit.

The SL is higher than the limit calculated when the AFD is within the limits of the F1( I) function of the overtemperature T reactor trip.

W hen the AFD is not within the tolerance, the AFD effect on the overtemperature T reactor trips will reduce the setpoints to provide protection consistent with the reactor core SLs (Refs. 3 and 4).

APPLICABILITY SL 2.1.1 only applies in MODES 1 and 2 because these are the only MODES in which the reactor is critical. Automatic protection functions are required to be OPERABLE during MODES 1 and 2 to ensure operation within the reactor core SLs. The main steam safety valves or automatic protection actions serve to prevent RCS heatup to the reactor core SL conditions or to initiate a reactor trip function, which forces the unit into MODE 3. Setpoints for the reactor trip functions are specified in LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation." In MODES 3, 4, 5, and 6, Applicability is not required since the reactor is not generating significant THERMAL POW ER.

SAFETY LIMIT Section 2.2, SL Violations, provides the Required Actions to be taken VIOLATIONS in response to a violation of Safety Limits. The bases for the Required Actions of Section 2.2 applicable to a violation of the reactor core SLs are discussed below.

(continued)

Vogtle Units 1 and 2 B 2.1.1-4 Revision No. 0 BASES INSERT 2.0-1 The reactor core SLs are established to preclude violation of the following fuel design criteria:

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

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

The reactor core SLs are used to define the various RPS functions such that the above criteria are satisfied during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs). To ensure that the RPS precludes the violation of the above criteria, additional criteria are applied to the Overtemperature and Overpower ¨7 reactor trip functions. That is, it must be demonstrated that the average enthalpy in the hot leg is less than or equal to the saturation enthalpy and that the core exit quality is within the limits defined by the DNBR correlation. Appropriate functioning of the RPS ensures that for variations in the THERMAL POWER, RCS Pressure, RCS average temperature, RCS flow rate, and AXIAL FLUX DIFFERENCE (AFD) l that the reactor core SLs will be satisfied during steady state operation, normal operational transients, and AOOs.

Reactor Core SLs B 2.1.1

BASES

SAFETY LIMIT 2.2.1 VIOLATIONS (continued) If the reactor core SL 2.1.1 is violated, the requirement to go to MODE 3 places the unit in a MODE in which this SL is not applicable.

The allowed Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months recognizes the importance of bringing the unit to a MODE of operation where this SL is not applicable, and reduces the probability of fuel damage.

(continued)

Vogtle Units 1 and 2 B 2.1.1-5 REVISION 16 Reactor Core SLs B 2.1.1

BASES (continued)

REFERENCES 1. 10 CFR 50, Appendix A, GDC 10.

2. FSAR, Section 7.2.

3. WCAP-8746-A, March 1977.

4. WCAP-9272-P-A, July 1985.

Vogtle Units 1 and 2 B 2.1.1-6 REVISION 16 Reactor Core SLs B 2.1.1

Remove this figure.

Figure B 2.1.1-1 (page 1 of 1)

REACTOR CORE SAFETY LIMITS VS. BOUNDARY OF PROTECTION

Vogtle Units 1 and 2 B 2.1.1-7 Rev. 1-6/03 RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1

B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits

BASES

BACKGROUND These Bases address requirements for maintaining RCS pressure, temperature, and flow rate within limits assumed in the safety analyses. The safety analyses (Ref. 1) of normal operating conditions and anticipated operational occurrences assume initial conditions within the normal steady state envelope. The limits placed on RCS pressure, temperature, and flow rate ensure that the DNB design criterion will be met for each of the transients analyzed.

The design method employed to meet the DNB design criterion for the VANTAGE 5 and LOPAR fuel assemblies is the Revised Thermal Design Procedure (RTDP). W ith the RTDP methodology, uncertainties in plant operating parameters, nuclear and thermal parameters, fuel fabrication parameters, computer codes, and DNB correlation predictions are considered statistically to obtain DNB uncertainty factors. Based on the DNB uncertainty factors, RTDP design limit DNBR values are determined in order to meet the DNB design criterion.

The RTDP design limit DNBR values are 1.24 and 1.23 for the typical and thimble cells, respectively, for the VANTAGE 5 fuel analyses with the W RB-2 correlation. For the LOPAR fuel analyses, the RTDP design limit DNBR values are 1.23 and 1.22 for the typical and thimble cells, respectively.

Additional DNBR margin is maintained by performing the safety analyses to DNBR limits higher than the design limit DNBR values.

This margin between the design and safety analysis limit DNBR values is used to offset known DNBR penalties (e.g., rod bow and transition core) and to provide DNBR margin for operating and design flexibility.

The Standard Thermal Design Procedure (STDP) is used for those analyses where RTDP is not applicable. In the STDP method the parameters used in analysis are treated in a conservative way from a DNBR standpoint. The parameter uncertainties are applied directly to the safety analyses input values to give the lowest minimum DNBR. The design

(continued)

Vogtle Units 1 and 2 B 3.4.1-1 Revision No. 0 RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1

BASES

BACKGROUND limit DNBR for STDP is the 95/95 limit for the appropriate DNB (continued) correlation. Again, additional DNBR margin is maintained in the safety analyses to offset the applicable DNBR penalties.

For both the W RB-1 and the W RB-2 correlations, the 95/95 DNBR correlation limit is 1.17. The W -3 DNB correlation is used for both fuel types where the primary DNBR correlations are not applicable. The WRB-1 and W RB-2 correlations were developed based on mixing vane data and therefore are only applicable in the heated rod spans above the first mixing vane grid. The W -3 correlation, which does not take credit for mixing vane grids, is used to calculate DNBR values in the heated region below the first mixing vane grid. In addition, the W -3 correlation is applied in the analysis of accident conditions where the system pressure is below the range of the primary correlations.

For system pressures in the range of 500 to 1000 psia, the W -3 correlation limit is 1.45. For system pressures greater than 1000 psia, the W -3 correlation limit is 1.30.

The RCS pressure and average temperature limits are consistent with operation within the nominal operational envelope. A lower pressure and/or higher average temperature will cause the reactor core to approach DNB limits.

The RCS flow rate normally remains constant during an operational fuel cycle with all pumps running. The minimum RCS flow limit corresponds to that assumed for DNB analyses. A lower RCS flow will cause the core to approach DNB limits.

Operation for significant periods of time outside these DNB limits increases the likelihood of a fuel cladding failure in a DNB limited event.

APPLICABLE The requirements of this LCO represent the initial conditions for DNB SAFETY ANALYSES limited transients analyzed in the plant safety analyses (Ref. 1).

The safety analyses have shown that transients initiated from the limits of this LCO will result in meeting the DNB design criterion.

Changes to the unit that could impact these parameters must be assessed for their impact on the DNB design criterion. The

(continued)

Vogtle Units 1 and 2 B 3.4.1-2 Revision No. 0 RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1

BASES

APPLICABLE transients analyzed include loss of coolant flow events and dropped SAFETY ANALYSES or stuck rod events. A key assumption for the analysis of these (continued) events is that the core power distribution is within the limits of LCO 3.1.6, "Control Bank Insertion Limits"; LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)"; and LCO 3.2.4, "QUADRANT POW ER TILT RATIO (QPTR)." specified in the COLR

The pressurizer pressure limit of 2199 psig and the RCS average temperature limit of 592.5qF correspond to analytical limits of 2185 psig and 594.4qF used in the safety analyses, with allowance for measurement uncertainty. the specified in the COLR

The indicated RCS flow value of 384,509 gpm corresponds to an analytical value of 374,400 gpm with allowance for measurement and indication uncertainties.

The RCS DNB parameters satisfy Criterion 2 of 10 CFR 50.36 (c)(2)(ii).

LCO This LCO specifies limits on the monitored process variables pressurizer pressure (PI-0455A,B&C, PI-0456 & PI-0456A, PI-0457 &

PI-0457A, and PI-0458 & PI-0458A), RCS average temperature (TI-0412, TI-0422, TI-0432, and TI-0442), and RCS total flow rate (FI-0414, FI-0415, FI-0416, FI-0424, FI-0425, FI-0426, FI-0434, FI-0435, FI-0436, FI-0444, FI-0445, FI-0446) to ensure the core operates within the limits assumed in the safety analyses. Operating within these limits will result in meeting the DNB design criterion in the These variables are event of a DNB limited transient.

contained in the COLRRCS total flow rate contains a measurement error of 2.7% based on to provide operating performing a precision heat balance above 90% RTP and using the and analysis flexibility result to calibrate the RCS flow rate indicators. This measurement from cycle to cycle. uncertainty includes a 0.1% penalty to account for potential fouling of the feedwater venturi, which might not be detected and could bias the result from the precision heat balance in a nonconservative manner.

Any fouling that might bias the flow rate measurement greater than 0.1% can be detected by monitoring and trending various plant performance parameters. If detected, either the effect of the fouling shall be quantified and the penalty for undetected fouling of the feedwater venturi

(continued)

Vogtle Units 1 and 2 B 3.4.1-3 Rev. 1-10/01 RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1

BASES

LCO compensated for in the RCS flow rate measurement or the venturi (continued) shall be cleaned to eliminate the fouling.

APPLICABILITY In MODE 1, the limits on pressurizer pressure, RCS coolant average temperature, and RCS flow rate must be maintained during steady state operation in order to ensure DNB design criterion will be met in the event of an unplanned loss of forced coolant flow or other DNB limited transient. In all other MODES, the power level is low enough that DNB is not a concern.

A Note has been added to indicate the limit on pressurizer pressure is not applicable during short term operational transients such as a THERMAL POW ER ramp > 5% RTP per minute or a THERMAL POW ER step > 10% RTP. These conditions represent short term perturbations where actions to control pressure variations might be counterproductive. Also, since they represent transients initiated from power levels < 100% RTP, an increased DNBR margin exists to offset The DNBR limit the temporary pressure variations.

Another set of limits on DNB related parameters is provided in SL 2.1.1, "Reactor Core SLs." Those limits are less restrictive than the limits of this LCO, but violation of a Safety Limit (SL) merits a stricter, more severe Required Action.

The conditions which define the DNBR limit

ACTIONS A.1

RCS pressure and RCS average temperature are controllable and measurable parameters. W ith one or both of these parameters not within LCO limits, action must be taken to restore parameter(s).

RCS total flow rate is not a controllable parameter and is not expected to vary during steady state operation. If the indicated RCS flow rate exhibits degradation and the actual total RCS flow rate is below the LCO limit as determined by precision heat balance, power must be reduced, as required by Required Action C.1, to restore DNB margin and eliminate the potential for violation of the accident analysis bounds.

(continued)

Vogtle Units 1 and 2 B 3.4.1-4 Revision No. 0 RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1

BASES

ACTIONS A.1 (continued)

The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time for restoration of the parameters is based on plant operating experience and provides sufficient time to adjust plant parameters, to determine the cause for the off normal condition, and to restore the readings within limits.

B.1

If degradation in RCS total flow rate is detected via the flow rate indicators, a precision calorimetric heat balance must be performed within 7 days of detection of the degradation. The precision heat balance will positively verify actual RCS total flow rate. The 7-day Completion Time is adequate to allow for the setup necessary for this measurement and is acceptable since the RCS low flow reactor trips will protect the reactor against actual low flow conditions.

C.1

If Required Actions A.1 or B.1 are not met within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . In MODE 2, the reduced power condition eliminates the potential for violation of the accident analysis bounds. The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable to reach the required plant conditions in an orderly manner.

SURVEILLANCE SR 3.4.1.1 REQUIREMENTS The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

This surveillance demonstrates that the pressurizer pressure remains greater than or equal to the limit specified in the COLR.

(continued)

Vogtle Units 1 and 2 B 3.4.1-5 REVISION 14 RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1

BASES

SURVEILLANCE SR 3.4.1.2 REQUIREMENTS (continued) The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

This surveillance demonstrates that the average SR 3.4.1.3 RCS temperature remains less than or equal to the limit specified in the COLR.

The RCS flow instrumentation indicates from 0% to 120% as opposed to actual flow in gallons per minute. Therefore, the flow instrumentation is used to detect degradation in flow rather than as a comparison against the actual limit in gallons per minute.

Degradation is defined as a change in indicated percent flow which is greater than the instrument channel inaccuracies and parallax errors.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.4.1.4

Measurement of RCS total flow rate by performance of a precision calorimetric heat balance allows the installed RCS flow instrumentation to be correlated with the precision flow measurement and verifies the actual RCS flow rate is greater than or equal to the minimum required RCS flow rate. In addition, in order to ensure that the measurement uncertainty assumed in the limit for RCS total flow rate is maintained, the instrumentation used for the precision calorimetric heat balance will be calibrated within 30 days prior to the precision calorimetric.

specified in the COLR.

(continued)

Vogtle Units 1 and 2 B 3.4.1-6 REVISION 14 RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1

BASES

SURVEILLANCE SR 3.4.1.4 (continued)

REQUIREMENTS The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

This SR is modified by a Note that allows entry into MODE 1, without having performed the SR, and placement of the unit in the best condition for performing the SR. The Note states that the SR is not required to be performed until 7 days after 90% RTP. This exception is appropriate since the heat balance requires the plant to be at a minimum of 90% RTP to obtain the stated RCS flow accuracies. The Surveillance shall be performed within 7 days after reaching 90% RTP.

REFERENCES 1. FSAR, Chapter 15.

Vogtle Units 1 and 2 B 3.4.1-7 REVISION 14 to NL-23-0152 VEGP - 1 and 2 Markups of COLR Pages (typical, for information R nly)

Attachment 4

Vogtle Electric Generating Plant - Units 1 and 2 Markups of Core Operating Limits Report (COLR) Pages (typical, for information only)

CORE OPERATING LIMITS REPORT, VEGP UNIT 1 CYCLE 24 JUNE 2021

1.0 CORE OPERATING LIMITS REPORT

This Core Operating Limits Report (COLR) for VEGP Unit 1 Cycle 24 has been prepared in accordance with the requirements of Technical Specification 5.6.5.

The Technical Requirement affected by this report is listed below:

13.1.1 SHUTDOWN MARGIN - MODES 1 and 2

The Technical Specifications affected by this report are listed below: Insert 1

3.1.1 SHUTDOWN MARGIN - MODES 3, 4 and 5 3.1.3 Moderator Temperature Coefficient 3.1.5 Shutdown Bank Insertion Limits 3.1.6 Control Bank Insertion Limits 3.2.1 Heat Flux Hot Channel Factor - FQ(Z) 3.2.2 Nuclear Enthalpy Rise Hot Channnel Factor - FN¨+

3.2.3 Axial Flux Difference Insert 2 3.9.1 Boron Concentration

Page 1 of 13 COLR Section 1.0 INSERT 1 2.1.1 Reactor Core Safety Limits for THERMAL POWER

COLR Section 1.0 INSERT 2 3.3.1 Reactor Trip System Instrumentation 3.4.1 Reactor Coolant System Pressure, Temperature, and Flow Departure from Nucleate Boiling Limits CORE OPERATING LIMITS REPORT, VEGP UNIT 1 CYCLE 24 JUNE 2021

2.7 Nuclear Enthalpy Rise Hot Channel Factor - FN¨+ (Specification 3.2.2)

2.7.1 F HNd 'FRTP1 PFP

' 1H 'H

where: P THERMAL POWER RATED THERMAL POWER

2.7.2 65FRTP1.

'H

2.7.3 3PF' H0.

2.8 Axial Flux Difference (Specification 3.2.3)

2.8.1 The Axial Flux Difference (AFD) Acceptable Operation Limits are provided in Figure 5.

2.9 Boron Concentration (Specification 3.9.1)

2.9.1 The boron concentration shall be greater than or equal to 2000 ppm. 2

Insert 3

2 This concentration bounds the condition of keff d 0.95 (all rods in less the most reactive rod) and subcriticality (all rods out) over the entire cycle. This concentration includes additional boron to address uncertainties and B10 depletion.

Page 5 of 13 INSERT 3 for new COLR Sections 2.10, 2.11 and 2.12 2.10 Reactor Core Safety Limits for THERMAL POWER (Specification 2.1.1) 2.10.1 In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the safety limits specified in Figure 7.

2.11 Reactor Trip System Instrumentation (Specification 3.3.1) 2.11.1 Reactor Trip System Instrumentation Overtemperature T (OTT) and Overpower T (OPT) setpoint parameter values for TS Table 3.3.1-1 are listed in COLR Tables 3 and 4.

2.12 Reactor Coolant System (RCS) Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits (Specification 3.4.1) 2.12.1 RCS DNB parameters for pressurizer pressure, RCS average temperature, and RCS total flow rate shall be within the limits specified below:

a. Pressurizer pressure 2199 psig;

b. RCS average temperature 592.5°F; and

c. The minimum RCS total flow rate shall be 384,509 GPM when using the precision heat balance method.

CORE OPERATING LIMITS REPORT, VEGP UNIT 1 CYCLE 24 JUNE 2021

Table 2 RAOC W(Z)

Axial Elevation 150 4000 8000 12000 16000 20000 Point (feet) MWD/MTU MWD/MTU MWD/MTU MWD/MTU MWD/MTU MWD/MTU

1-5 12.072 - 11.267 1.0000 1.0000 1.0000 1.0000 1.0000 1.0000 6 11.066 1.3733 1.4042 1.2807 1.2524 1.2279 1.2288 7 10.865 1.3592 1.3892 1.2714 1.2511 1.2252 1.2237 8 10.664 1.3419 1.3696 1.2569 1.2461 1.2119 1.2118 9 10.462 1.3243 1.3477 1.2420 1.2392 1.1946 1.2012 10 10.261 1.3032 1.3235 1.2331 1.2382 1.1967 1.1895 11 10.060 1.2799 1.2969 1.2245 1.2395 1.2029 1.1869 12 9.859 1.2544 1.2649 1.2132 1.2404 1.2071 1.1850 13 9.658 1.2448 1.2309 1.2046 1.2287 1.2063 1.1876 14 9.456 1.2433 1.2035 1.1971 1.2238 1.2078 1.1907 15 9.255 1.2406 1.1870 1.1866 1.2224 1.2059 1.1962 16 9.054 1.2314 1.1714 1.1819 1.2178 1.2049 1.2024 17 8.853 1.2250 1.1656 1.1769 1.2031 1.2024 1.2048 18 8.652 1.2208 1.1660 1.1767 1.1947 1.1963 1.2063 19 8.450 1.2213 1.1704 1.1829 1.1974 1.1982 1.2120 20 8.249 1.2174 1.1728 1.1860 1.2010 1.2074 1.2250 21 8.048 1.2228 1.1734 1.1872 1.2093 1.2212 1.2399 22 7.847 1.2252 1.1725 1.1868 1.2150 1.2324 1.2523 23 7.646 1.2247 1.1694 1.1842 1.2185 1.2412 1.2623 24 7.444 1.2221 1.1653 1.1805 1.2204 1.2483 1.2710 25 7.243 1.2164 1.1589 1.1740 1.2184 1.2510 1.2747 26 7.042 1.2087 1.1509 1.1659 1.2138 1.2505 1.2750 27 6.841 1.1995 1.1434 1.1568 1.2080 1.2483 1.2738 28 6.640 1.1883 1.1364 1.1493 1.2003 1.2439 1.2705 29 6.438 1.1766 1.1299 1.1417 1.1910 1.2372 1.2676 30 6.237 1.1653 1.1232 1.1343 1.1814 1.2294 1.2627 31 6.036 1.1534 1.1207 1.1260 1.1709 1.2204 1.2567 32 5.835 1.1452 1.1179 1.1215 1.1597 1.2100 1.2491 33 5.634 1.1390 1.1197 1.1196 1.1583 1.2034 1.2398 34 5.432 1.1386 1.1326 1.1259 1.1642 1.2035 1.2355 35 5.231 1.1448 1.1440 1.1366 1.1686 1.2082 1.2367 36 5.030 1.1521 1.1550 1.1463 1.1727 1.2118 1.2391 37 4.829 1.1589 1.1653 1.1557 1.1777 1.2138 1.2395 38 4.628 1.1651 1.1751 1.1643 1.1813 1.2141 1.2379 39 4.426 1.1706 1.1842 1.1724 1.1839 1.2128 1.2342 40 4.225 1.1750 1.1922 1.1794 1.1851 1.2100 1.2289 41 4.024 1.1782 1.1994 1.1858 1.1863 1.2060 1.2220 42 3.823 1.1823 1.2056 1.1910 1.1870 1.1995 1.2120 43 3.622 1.1888 1.2108 1.1949 1.1857 1.1912 1.1995 44 3.420 1.1950 1.2151 1.1982 1.1841 1.1822 1.1864 45 3.219 1.1992 1.2187 1.2002 1.1816 1.1741 1.1738 46 3.018 1.2056 1.2318 1.2057 1.1867 1.1750 1.1743 47 2.817 1.2141 1.2489 1.2211 1.1969 1.1846 1.1842 48 2.616 1.2276 1.2646 1.2408 1.2060 1.1952 1.1955 49 2.414 1.2436 1.2847 1.2606 1.2150 1.2057 1.2074 50 2.213 1.2608 1.3090 1.2806 1.2266 1.2162 1.2192 51 2.012 1.2778 1.3332 1.2999 1.2377 1.2253 1.2293 52 1.811 1.2946 1.3570 1.3186 1.2484 1.2340 1.2388 53 1.610 1.3103 1.3795 1.3366 1.2589 1.2431 1.2490 54 1.408 1.3253 1.4006 1.3535 1.2693 1.2525 1.2598 55 1.207 1.3395 1.4202 1.3693 1.2792 1.2619 1.2708 56 1.006 1.3526 1.4379 1.3837 1.2888 1.2714 1.2823

57-61 0.805 - 0.000 1.0000 1.0000 1.0000 1.0000 1.0000 1.0000

Top and Bottom 5 Points Excluded per Technical Specification B3.2.1.

These W(Z) values are consistent with Figure 5, and are valid over the HFP T avg temperature range from 580.0 to 587.0°F.

Insert 4. Remaining pages to be renumbered.

Page 7 of 13 INSERT 4 for new COLR Table 3 and Table 4 Table 3 Reactor Trip System Instrumentation - Overtemperature T (OTT)

Setpoint Parameter Values

Parameter Value time constants utilized in lead-lag compensator for differential temperature 1 = 0 seconds, 2 = 0 seconds time constant utilized in lag compensator for differential temperature 3 6 seconds fundamental setpoint K1 114.9% RTP modifier for temperature K2 = 2.24% RTP per °F time constants utilized in lead-lag compensator for temperature compensation 4 28 seconds, 5 4 seconds time constant utilized in lag compensator for average temperature 6 6 seconds indicated loop specific RCS average temperature at RTP T 588.4 °F modifier for pressure K3 = 0.177% RTP per psig reference pressure P 2235 psig f1(AFD) modifier for Axial Flux Difference (AFD):

1. for AFD between -23% and +10%, = 0% RTP

2. for each % AFD is below -23%, the trip setpoint shall be reduced by 3.3% RTP

3. for each % AFD is above +10%, the trip setpoint shall be reduced by 1.95% RTP (p) The compensated temperature difference shall be no more negative than 3 degrees F.

Table 4 Reactor Trip System Instrumentation - Overpower T (OPT)

Setpoint Parameter Values

Parameter Value time constants utilized in lead-lag compensator for differential temperature 1 = 0 seconds, 2 = 0 seconds time constant utilized in lag compensator for differential temperature 3 6 seconds fundamental setpoint K4 110% RTP modifier for temperature change, for increasing temperature K5 2% RTP per °F modifier for temperature change, for decreasing temperature K5 0% RTP per °F time constant utilized in rate-lag compensator for temperature compensation 7 10 seconds time constant utilized in lag compensator for average temperature 6 6 seconds modifier for temperature, for T > T K6 0.244% RTP per ° modifier for temperature, for T T K6 = 0% RTP per °F LQGLFDWHGORRSVSHFLILF5&6DYHUDJHWHPSHUDWXUHDW573 7 )

modifier for Axial Flux Difference (AFD), for all AFD f2(AFD) = 0% RTP INSERT 5 for new COLR Figure 7

FIGURE 7 REACTOR CORE SAFETY LIMITS

&25(23(5$7,1*/,0,765(3257 9(*381,7 &<&/( 129(0%(5

&25(23(5$7,1*/,0,765(3257

7KLV &RUH 2SHUDWLQJ /LPLWV 5HSRUW &2/5 IRU 9(*3 8QLW &\\FOH KDV EHHQ SUHSDUHG LQ DFFRUGDQFH

ZLWKWKHUHTXLUHPHQWVRI7HFKQLFDO6SHFLILFDWLRQ

7KH7HFKQLFDO5HTXLUHPHQWDIIHFWHGE\\WKLVUHSRUWLVOLVWHGEHORZ

6+87'2:10$5*,1 02'(6 DQG

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Insert 1

6+87'2:10$5*,1 02'(6 DQG

0RGHUDWRU7HPSHUDWXUH&RHIILFLHQW

6KXWGRZQ%DQN,QVHUWLRQ/LPLWV

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$[LDO)OX['LIIHUHQFHInsert 2

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Page 1 of 13 COLR Section 1.0 INSERT 1 2.1.1 Reactor Core Safety Limits for THERMAL POWER

COLR Section 1.0 INSERT 2 3.3.1 Reactor Trip System Instrumentation 3.4.1 Reactor Coolant System Pressure, Temperature, and Flow Departure from Nucleate Boiling Limits

&25(23(5$7,1*/,0,765(3257 9(*381,7 &<&/( 129(0%(5

1XFOHDU(QWKDOS\\5LVH+RW&KDQQHO)DFWRU )1 ¨+ 6SHFLILFDWLRQ

FNd 'FRTP PFP

' H 'HH

ZKHUH P THERMALPOWER

RATEDTHERMALPOWER

FRTP

'H

PF 'H

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7KH$[LDO)OX['LIIHUHQFH$)'$FFHSWDEOH2SHUDWLRQ/LPLWVDUHSURYLGHGLQ)LJXUH

%RURQ&RQFHQWUDWLRQ6SHFLILFDWLRQ

7KHERURQFRQFHQWUDWLRQVKDOOEHJUHDWHUWKDQRUHTXDOWR SSP

Insert 3

7KLVFRQFHQWUDWLRQERXQGVWKHFRQGLWLRQRINHIId DOOURGVLQOHVVWKHPRVWUHDFWLYHURGDQG

VXEFULWLFDOLW\\DOOURGVRXWRYHUWKHHQWLUHF\\FOH 7KLVFRQFHQWUDWLRQLQFOXGHVDGGLWLRQDOERURQWR

DGGUHVVXQFHUWDLQWLHVDQG% GHSOHWLRQ

Page 5 of 13 INSERT 3 for new COLR Sections 2.10, 2.11 and 2.12 2.10 Reactor Core Safety Limits for THERMAL POWER (Specification 2.1.1) 2.10.1 In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the safety limits specified in Figure 7.

2.11 Reactor Trip System Instrumentation (Specification 3.3.1) 2.11.1 Reactor Trip System Instrumentation Overtemperature T (OTT) and Overpower T (OPT) setpoint parameter values for TS Table 3.3.1-1 are listed in COLR Tables 3 and 4.

2.12 Reactor Coolant System (RCS) Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits (Specification 3.4.1) 2.12.1 RCS DNB parameters for pressurizer pressure, RCS average temperature, and RCS total flow rate shall be within the limits specified below:

a. Pressurizer pressure 2199 psig;

b. RCS average temperature 592.5°F; and

c. The minimum RCS total flow rate shall be 384,509 GPM when using the precision heat balance method.

&25(23(5$7,1*/,0,765(3257 9(*381,7 &<&/( 129(0%(5

Table 2 RAOC W(Z)

Axial Elevation 150 3000 8000 12000 16000 20000 Point (feet) MWD/MTU MWD/MTU MWD/MTU MWD/MTU MWD/MTU MWD/MTU

1-5 12.072 - 11.267

6 11.066

7 10.865

8 10.664

9 10.462

10.261

10.060

9.859

9.658

9.456

9.255

9.054

8.853

8.652

8.450

8.249

8.048

7.847

7.646

7.444

7.243

7.042

6.841

6.640

6.438

6.237

6.036

5.835

5.634

5.432

5.231

5.030

4.829

4.628

4.426

4.225

4.024

3.823

3.622

3.420

3.219

3.018

2.817

2.616

2.414

2.213

2.012

1.811

1.610

1.408

1.207

1.006

57-61 0.80- 0.000

Top and Bottom x per Technical Specification B3.2.1.

7KHVH:=YDOXHVDUHFRQVLVWHQWZLWK)LJXUH DQGDUHYDOLGRYHUWKH+)37DYJWHPSHUDWXUHUDQJHIURP

WR )

Insert 4. Remaining pages to be renumbered.

Page 7 of 13 INSERT 4 for new COLR Table 3 and Table 4 Table 3 Reactor Trip System Instrumentation - Overtemperature T (OTT)

Setpoint Parameter Values

Parameter Value time constants utilized in lead-lag compensator for differential temperature 1 = 0 seconds, 2 = 0 seconds time constant utilized in lag compensator for differential temperature 3 6 seconds fundamental setpoint K1 114.9% RTP modifier for temperature K2 = 2.24% RTP per °F time constants utilized in lead-lag compensator for temperature compensation 4 28 seconds, 5 4 seconds time constant utilized in lag compensator for average temperature 6 6 seconds indicated loop specific RCS average temperature at RTP T 588.4 °F modifier for pressure K3 = 0.177% RTP per psig reference pressure P 2235 psig f1(AFD) modifier for Axial Flux Difference (AFD):

1. for AFD between -23% and +10%, = 0% RTP

2. for each % AFD is below -23%, the trip setpoint shall be reduced by 3.3% RTP

3. for each % AFD is above +10%, the trip setpoint shall be reduced by 1.95% RTP (p) The compensated temperature difference shall be no more negative than 3 degrees F.

Table 4 Reactor Trip System Instrumentation - Overpower T (OPT)

Setpoint Parameter Values

Parameter Value time constants utilized in lead-lag compensator for differential temperature 1 = 0 seconds, 2 = 0 seconds time constant utilized in lag compensator for differential temperature 3 6 seconds fundamental setpoint K4 110% RTP modifier for temperature change, for increasing temperature K5 2% RTP per °F modifier for temperature change, for decreasing temperature K5 0% RTP per °F time constant utilized in rate-lag compensator for temperature compensation 7 10 seconds time constant utilized in lag compensator for average temperature 6 6 seconds modifier for temperature, for T > T K6 0.244% RTP per ° modifier for temperature, for T T K6 = 0% RTP per °F LQGLFDWHGORRSVSHFLILF5&6DYHUDJHWHPSHUDWXUHDW573 7 )

modifier for Axial Flux Difference (AFD), for all AFD f2(AFD) = 0% RTP INSERT 5 for new COLR Figure 7

FIGURE 7 REACTOR CORE SAFETY LIMITS

v t e Letter Sequence Request
EPID:L-2023-LLA-0053, TSTF-339 (Approved, Closed)
Initiation Request, Request, Request Acceptance... Administration Questions 12 July 2023 5 September 2023 Answers 11 August 2023 Results Approval Other:
MONTHYEAR2023-04-11 [Table View]

NL-23-0152, License Amendment Request: Technical Specification Revision to Adopt TSTF-339-A, Relocate Technical Specification Parameters to the COLR Consistent with WCAP-14483

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NL-23-0152, License Amendment Request: Technical Specification Revision to Adopt TSTF-339-A, Relocate Technical Specification Parameters to the COLR Consistent with WCAP-14483

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Enclosure:

SUMMARY

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