Text

License Amendment Request to Remove or Modify Outdated License Information
	ML25058A477
Person / Time
Site:	Vogtle
Issue date:	02/27/2025
From:	Coleman J; Southern Nuclear Operating Co
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
References
	NL-24-0228
	Download: ML25058A477 (1)
	v • d • e

A Southern Nuclear February 27, 2025 Docket Nos.52-025 52-026 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Regulatory Affairs Southern Nuclear Operating Company Vogtle Electric Generating Plant Units 3 and 4 3535 Colonnade Parkway Birmingham, AL 35243 205.992.5000 NL-24-0228 10 CFR 50.90 License Amendment Request to Remove or Modify Outdated License Information Pursuant to 10 CFR 52.98(c) and in accordance with 10 CFR 50.90, Southern Nuclear Operating Company (SNC) requests an amendment to the combined license (COL) for Vogtle Electric Generating Plant (VEGP) Unit 3 (License Number NPF-91) and Unit 4 (License Number NPF-92). The requested amendment proposes to remove completed COL conditions, to modify other COL conditions, and to remove some unit-specific Technical Specification (TS) and COL language that is no longer applicable.

The Enclosure to this letter provides the description, technical evaluation, regulatory evaluation (including the Significant Hazards Consideration Determination) and environmental considerations for the proposed changes. to the Enclosure provides a markup depicting the requested changes to the VEGP Unit 3 and Unit 4 licensing basis documents, e.g., the Combined Licenses and the Technical Specifications. Attachment 2 provides the clean revised Technical Specification pages pending inclusion of the appropriate amendment number. Attachment 3 provides a markup of the existing TS Bases language to show an associated change for information only.

SNC requests NRC staff review and approval of this LAR no later than 12 months from acceptance. SNC expects to implement the proposed amendment within 30 days of approval of the LAR.

This letter contains no regulatory commitments. This letter has been reviewed and determined not to contain security-related information.

In accordance with 10 CFR 50.91, SNC is notifying the State of Georgia by transmitting a copy of this letter and its enclosures to the designated State Official.

If you have any questions, please contact Ryan Joyce at (205) 992-6468.

U. S. Nuclear Regulatory Commission NL-24-0228 Page 2 I declare under penalty of perjury that the foregoing is true and correct. Executed on the 27th of February 2025.

Respectfully submitted, Jamie M. Coleman Director, Regulatory Affairs Southern Nuclear Operating Company

Enclosure:

Evaluation of Proposed Changes cc:

NRC Regional Administrator, Region II NRR Project Manager - Vogtle 3&4 Senior Resident Inspector - Vogtle 3&4 Director, Environmental Protection Division - State of Georgia Document Services RTYPE: VND.LI.L00

ENCLOSURE to NL-24-0228 Evaluation of Proposed Changes

1.

SUMMARY

DESCRIPTION

2.

DETAILED DESCRIPTION 2.1 System Design and Operation 2.2 Current Requirements 2.3 Description of Proposed Change

3.

REASON FOR PROPOSED CHANGE and TECHNICAL EVALUATION

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. Licensing Basis Document Markups

2. Revised Technical Specification Pages

3. Associated Technical Specification Bases Change (For information only)

Enclosure to NL-24-0228 Evaluation of Proposed Changes

1.

SUMMARY

DESCRIPTION The proposed change would revise the Combined Licenses (COLs) for Vogtle Electric Generating Plant (VEGP) Unit 3 and Unit 4 by removing the completed license conditions, modifying other license conditions, and to remove some unit-specific Technical Specification (TS) and COL language that is no longer applicable in facility operating licenses NPF-91 and NPF-92, respectively.

2. DETAILED DESCRIPTION 2.1 System Design and Operation The COLs include numerous license conditions and TS notes which are outdated or obsolete and can be removed or modified to simplify the license. Many of the license conditions require specific activities to be completed by specific milestones such as the 10 CFR 52.103(g) finding, initial fuel loading, initial criticality, and commercial operation. The TS notes limit applicability of some requirements to after initial criticality of the respective unit.

2.2 Current Requirements The facility combined licenses NPF-91 and NPF-92 include Section 2.0 which identifies that "The license is subject to, and SNC shall comply with, the conditions specified and incorporated below." These conditions address:

(1)

Changes during Construction; (2)

Pre-operational Testing; (3)

Nuclear Fuel Loading and Pre-critical Testing; (4)

Initial Criticality and Low-Power Testing; (5)

Power Ascension Testing; (6)

Maximum Power Level; (7)

Reporting Requirements; (8)

Incorporation; (9)

Technical Specifications; (10) Operational Program Implementation; (11) Operational Program Implementation Schedule; (12) Site-and Unit-specific Conditions; and (13) Departures from Plant-specific DCD Tier 2* Information.

The facility combined licenses also include:

Section 2.H which addresses inspections, tests, analyses, and acceptance criteria (ITAAC) closure maintenance; Section 2.1 which addresses maintaining the guidance and strategies developed in accordance with 10 CFR 50.54(hh)(2); and Section 2.J which identifies the expiration of the COL as 40 years from the 10 CFR 52.103(g) finding.

E-2

Enclosure to NL-24-0228 Evaluation of Proposed Changes LAR-23-002 (approved as Unit 3 Amendment No. 189) added language for Unit 3 TS only, LAR-23-004 (approved as Unit 3 Amendment No. 190) added language for Unit 3 TS only, and LAR-23-005 (approved as Unit 4 Amendment No. 189) added language for Unit 4 TS and COL only, to delay OPERABILITY requirements prior to initial criticality of the respective unit.

2.3 Description of Proposed Change The proposed change would modify the COLs to replace license condition Sections 2.0.(1), 2.0.(4), 2.0.(5), 2.0.(7), 2.0.(9), 2.0.(10), 2.0.(11), 2.0.(12) and Sections 2.H and 2.1 with a statement of "Removed by Amendment No.###-"

where ##-# is the appropriate issued amendment number.

The proposed change would modify the COL for VEGP Unit 3 to replace license condition Sections 2.0.(2) and 2.0.(3) with a statement of "Removed by Amendment Nos. 192 and ##-#"where###- is the appropriate issued amendment number.

The proposed change would modify the COL for VEGP Unit 4 to replace license condition Sections 2.0.(2) and 2.0.(3) with a statement of "Removed by Amendment Nos. 194 and ##-#"where###- is the appropriate issued amendment number.

The proposed change would modify the COLs to remove the initial phrase of "Upon submission of the notification required by Section 2.D.(5)(d) of this license,"

from license condition Section 2.0.(6).

The proposed change would modify the COLs to remove the portions of Section 2.0.(13) that address matters rendered moot after the first full power operation pursuant to 10 CFR Part 52, Appendix D, Section VIII.B.6.c.

The proposed change would modify the COL for VEGP Unit 3 to modify Section 2.J to identify August 3, 2062, as the date on which the license would expire in lieu of the existing descriptive phrase "the date 40 years from the date that the Commission finds that the acceptance criteria in the combined license are met in accordance with 10 CFR 52.103(g)."

The proposed change would modify the COL for VEGP Unit 4 to modify Section 2.J to identify July 28, 2063, as the date on which the license would expire in lieu of the existing descriptive phrase "the date 40 years from the date that the Commission finds that the acceptance criteria in the combined license are met in accordance with 10 CFR 52.103(g)."

The proposed change to the COL Appendix A would remove language added, by LAR-23-002 (approved as Unit 3 Amendment No. 189) for Unit 3 TS only, by LAR-23-004 (approved as Unit 3 Amendment No. 190) for Unit 3 TS only, and by LAR-23-005 (approved as Unit 4 Amendment No. 189) for Unit 4 COL only, to delay OPERABILITY requirements prior to initial criticality. Unit 3 and Unit 4 have both achieved initial criticality and as such the language is no longer appropriate or useful. This change involves removal of obsolete notes from Unit 3 TS 3.4.11, E-3

Enclosure to NL-24-0228 Evaluation of Proposed Changes TS 3.4.12, and TS 3.4.13 for Unit 3 Amendment No. 189 and from Unit 3 TS 3.3.8, TS 3.3.9, TS 3.3.10, TS 3.3.16, TS 3.3.19, TS 3.3.20, TS 3.5.7 and TS 3.5.8 for Unit 3 Amendment No. 190, and removal of obsolete text from Unit 4 Limiting Condition for Operation (LCO) 3.0.7 and COL condition 2.D.(9) (including parts (a),

(b), and (c)) for Unit 4 Amendment No. 189.

Markups showing these changes are provided in Attachment 1.

3. REASON FOR PROPOSED CHANGE and TECHNICAL EVALUATION Many of the license conditions included in Section 2.D of the COLs have been implemented as required prior to their applicable milestone, i.e., the license conditions have been met with no further requirements to be met, and are thus, no longer requirements to be imposed on the unit. Others are no longer applicable or are modified for other specific reasons. Each of these and Sections 2.H, 2.1, and 2.J, are individually addressed below.

Condition 2.D.(1 ), Changes during Construction, provides a process under which SNC could proceed with proposed changes (at risk) during construction while an amendment request was under consideration. The process identified by this condition is no longer necessary or applicable since construction has been satisfactorily completed, fuel has been loaded, testing has been satisfactorily completed, and commercial operation has begun for Unit 3 and for Unit 4 [ADAMS Accession Nos.

ML23212A964 and ML24119A001, respectively]. The condition is thus moot as there will not be any further changes occurring during construction, and the condition can be removed as an administrative change.

Condition 2.D.(2), Item c) required that a) SNC perform certain tests prior to actual operation of the plant, b) review and evaluate the results, and confirm that the test results were satisfactory, and c) notify the NRC of successful completion. SNC notified the NRC of completion of this condition for both Unit 3 and Unit 4

[ML20048A055]. Thus, this condition is no longer necessary or applicable since the pre-operational testing has been satisfactorily completed, fuel has been loaded, startup testing has been satisfactorily completed, and commercial operation has begun. The condition is thus moot and the condition can be removed as an administrative change.

Condition 2.D.(2), Item d) required notification of the NRC of successful completion of all the ITAAC was addressed in recent License Amendment Request (LAR)23-001 for Unit 3 and LAR 23-008 for Unit 4. SNC notified the NRC that all lTAAC have been completed [ML22210A090 and ML23201A266, respectively], and NRC review acknowledged the completion in the 10 CFR 50.103(g) letter for Unit 3

[ML20290A284] and for Unit 4 [ML22348A094], recognizing that the Acceptance Criteria identified in the COL had been satisfied. The condition is thus moot and the condition can be removed as an administrative change.

Condition 2.D.(3), Nuclear Fuel Loading and Pre-critical Testing, provides for fuel load and pre-critical testing once Condition 2.D.(2) is complete; requires completion of the pre-critical tests, review and evaluation of the results, confirmation that the test results E-4

Enclosure to NL-24-0228 Evaluation of Proposed Changes were satisfactory, and notification of the NRC of successful completion. Completion of this condition is documented for Unit 3 and for Unit 4 by SNC letters [ML23061A121 and ML24039A055, respectively]. Thus, this condition is no longer necessary or applicable since the pre-critical testing has been satisfactorily completed, fuel has been loaded, startup testing has been satisfactorily completed, and commercial operation has begun. The condition is thus moot and the condition can be removed as an administrative change.

Condition 2.D.(4 ), Initial Criticality and Low-Power Testing, provides for facility operation at less than or equal to 5% thermal power and initial criticality and low-power testing once Condition 2.0(3) is complete; requires completion of the initial criticality and low-power tests, review and evaluation of the results, confirmation that the test results were satisfactory, and notification of the NRC of successful completion.

Completion of this condition is documented for Unit 3 and for Unit 4 in SNC letters

[ML23066A269 and ML24046A034, respectively]. Thus, this condition is no longer necessary or applicable since fuel has been loaded, the low-power testing has been satisfactorily completed, power ascension testing has been satisfactorily completed, and commercial operation has begun. The condition is thus moot and the condition can be removed as an administrative change.

Condition 2.D.(5), Power Ascension Testing, provides for facility operation at less than or equal to 100% thermal power for power ascension testing once Condition D( 4) is complete; requires completion of the power ascension tests, review and evaluation of the results, confirmation that the test results were satisfactory, and notification of the NRC of successful completion. Completion of this condition is documented for Unit 3 and for Unit 4 in SNC letters [ML23209A830 and ML24117A351, respectively]. Thus, this condition is no longer necessary or applicable since the power ascension testing has been satisfactorily completed and commercial operation has begun. The condition is thus moot and the condition can be removed as an administrative change.

Condition 2.D.(6), Maximum Power Level, identifies the steady state reactor core power level above which the facility is not to exceed. The initial phrase of this condition provides for facility operation at or below the 100% thermal power for any purpose "Upon submission of the notification required by Section 2.D.(S)(d) of this license." As noted above, the Condition 2.D.(5), item d), notification was completed and as noted in the paragraph addressing Condition 2.D.(1 ), the NRC acknowledged this required notification, recognizing that the testing had been completed. Since this notification has been provided, the initial phrase is moot and can be removed as an administrative change. The remainder of the condition provides a maximum power level for the unit and is retained. Any violations of this retained portion of the condition would be reported per Condition 2.D.(7), Reporting Requirements.

Condition 2.D.(7), Reporting Requirements, requires reporting of a) changes to the initial test program made without prior NRC approval, and b) violations of the Conditions in Section 2.D.(3), Section 2.D.(4), Section 2.D.(5), or Section 2.D.(6). As noted in the paragraphs above, the initial test program has been completed. Thus, there will be no further changes and no further reporting required. The change to the reporting portion of the License Condition change is consistent with the notice published in the Federal Register (FR) on November 4, 2005, 70 FR 67202 ("Notice of Availability of Model Application Concerning Elimination of Typical License Condition E-5

Enclosure to NL-24-0228 Evaluation of Proposed Changes Requiring Reporting of Violations of Section 2.C of Operating License Using the Consolidated Line Item Improvement Process [CLIIP]," dated November 4, 2005) as part of the consolidated line-item improvement process. SNC has reviewed the model safety evaluation (SE), the no significant hazards consideration (NSHC) determination. and the environmental assessment (EA) published in 70 FR 51098

("Notice of Opportunity To Comment on Model Safety Evaluation on Elimination of Typical License Condition Requiring Reporting of Violations of Section 2.C of Operating License Using the Consolidated Line Item Improvement Process," dated August 29, 2005), as part of the CLIIP Notice of Opportunity to Comment and concluded that the justifications presented in the SE, the NSHC determination and the EA prepared by the NRC staff are applicable to VEGP Units 3 and 4. Thus, this Condition can be removed as an administrative change.

Condition 2.D.(8), Incorporation, identifies Appendices that are incorporated into the license as well as the latest approved Amendment of the license. No changes are proposed for this condition other than to reflect the Amendment number for approval of this request.

Condition 2.D.(9), Technical Specifications, identifies that the Technical Specifications included in Appendix A become effective at the 10 CFR 53.103(g) finding. As noted above, the NRC acknowledged the completion in the 10 CFR 50.103(g) letter for Unit 3 [ML20290A284] and for Unit 4 [ML22348A094], making the finding that initiated the effectiveness of the Technical Specifications. Since that finding has been made, and the unit is operating under the requirements of the Technical Specifications, there is no further need to identify when the Technical Specifications become effective.

Thus, this Condition is moot and can be removed as an administrative change.

Condition 2.D.(10), Operational Program Implementation, identifies programs or portions of programs to be implement prior to specific milestones. As Commercial Operation has been achieved for each unit, the identified programs or portions of programs have been implemented, each at, or prior to, the required specific milestone

[ML23187A500 and ML24092A368]. Thus, this Condition is moot and can be removed as an administrative change.

Condition 2.D.(11 ), Operational Program Implementation Schedule, requires submittal of a schedule for implementation of operational programs, with periodic updates, until each of the identified programs is fully implemented. As Commercial Operation has been achieved for each unit, the identified programs have been fully implemented, each at, or prior to, the required specific milestone [ML23187A500 and ML24092A368]. Thus, this Condition is moot and can be removed as an administrative change.

Condition 2.D.(12), Site-and Unit-specific Conditions, requires seven site or unit specific conditions to be completed. Each of these conditions has been completed as noted below. Thus, this Condition is moot and can be removed as an administrative change.

Condition 2.D.(12)(a) addresses improvement of the soils directly above the blue bluff marl and under or adjacent to the Seismic Category I structures to eliminate any liquefaction potential. The soils directly above the blue bluff marl were E-6

Enclosure to NL-24-0228 Evaluation of Proposed Changes removed and replaced with structural backfill prior to construction of the seismic Category I structures. Thus, this Condition is moot and can be removed as an administrative change.

Condition 2.D.(12)(b) addresses completion of selected as-designed pipe rupture hazards analyses. SNC notified the NRC of the completion and availability of the as-designed pipe rupture hazards analysis reports as required by this license condition. These notifications included:

ND-13-2559 [ML13352A163] - Report Availability for Auxiliary Building Rooms

ND-14-0547 [ML14115A157] - Report Availability for Containment Rooms

ND-15-0573 [ML15086A076] - Report Availability for the Containment Building

ND-17-1177 [ML17178A457] - Report Availability for Auxiliary Building Room

ND-17-1191 [ML17187A179] - Report Availability for Auxiliary Building Rooms

ND-17-1459 [ML17230A362] - Report Availability for Auxiliary Building Rooms

ND-17-2116 [ML17353A113] - Report Availability for Turbine Building/ AB

ND-18-0271 [ML18059B013] - Report Availability for Auxiliary Building Since the required reports are complete and the required notifications have been completed, this Condition is moot and can be removed as an administrative change.

Condition 2.D.(12)(c) addresses completion of analysis of selected as-designed piping segments. SNC notified the NRC of the completion and availability of the analysis of the as designed individual piping segments as required by the license condition. These notifications included:

ND-13-2558 [ML13352A165] -Analysis Availability for PXS, RCS, RNS & SFS

ND-14-0337 [ML14062A048] -Analysis Availability for CCS

ND-14-1009 [ML14211A660] -Analysis Availability for CVS, SFS, VBS & VWS

ND-15-0681 [ML15121A902] -Analysis Availability for PXS, RCS, VBS & VWS

ND-16-1250 [ML16211A440] -Analysis Availability for PXS

ND-16-1613 [ML16243A189]-Analysis Availability for RCS, SGS & RNS

ND-16-1967 [ML16274A228] -Analysis Availability for PXS, RCS & CVS

ND-16-0815 [ML16319A043] -Analysis Availability for PXS & SGS

ND-17-0356 [ML 17067 AS00] -Analysis Availability for CCS, CVS & RCS

ND-17-1160 [ML17178A058] -Analysis Availability for PCS & SGS

ND-17-1394 [ML17215B053] -Analysis Availability for SGS

ND-17-1989 [ML17338A959] -Analysis Availability for SGS & RCS Since the required reports are complete and the required notifications have been completed, this Condition is moot and can be removed as an administrative change.

Condition 2.D.(12)(d) addresses (i) submittal of a fully developed set of plant-specific emergency action levels (EALs) that have been discussed and agreed upon with State and local officials and (ii) submittal of an assessment of emergency response staffing performed in accordance with NEI 10-05, Rev. 0.

These submittals were provided by SNC [ML19284C118 and ML16146A724, respectively]. Since the required submittals have been made, this Condition is moot and can be removed as an administrative change.

E-7

Enclosure to NL-24-0228 Evaluation of Proposed Changes Condition 2.D.(12)(e) precludes modification of portions of the Special Nuclear Material (SNM) Physical Protection Program prior to implementing the requirements of 10 CFR 73.55. As documented by SNC [ML23187A500], the requirements of 10 CFR 73.55 have been implemented through the Physical Security Program, the Safeguards Contingency Program, the Security Training and Qualification Program, and the Cyber Security Program. Thus, this Condition is moot and can be removed as an administrative change.

Condition 2.D.(12)(f) addresses submittal, and periodic update, of a schedule for implementation of several license conditions, until each license condition has been fully implemented, including:

1.

The construction and inspection procedures for steel concrete composite construction activities for seismic Category I nuclear island modules;

2.

The spent fuel rack Metamic Coupon monitoring program (before initial fuel load);

3.

Implementation of the flow accelerated corrosion (FAC) program including construction phase activities (before initial fuel load);

4.

A turbine maintenance and inspection program (before initial fuel load);

5.

The availability of documented instrumentation uncertainties to calculate a power calorimetric uncertainty (before initial fuel load);

6.

The availability of administrative controls to implement maintenance and contingency activities related to the power calorimetric uncertainty instrumentation (before initial fuel load);

7.

The site-specific severe accident management guidelines (before startup testing);

8.

The operational and programmatic elements of the mitigative strategies for responding to circumstances associated with loss of large areas of the plant due to explosions or fire (before initial fuel load); and

9.

The ITP procedures identified in U FSAR Section 14.2.3:

a. administrative manual (before the first preoperational test)

b. preoperational testing (before scheduled performance)

c. startup testing (before initial fuel load)

As documented internally and in SNC letters for Unit 3 [ML23187A500] and for Unit 4 [ML24092A368], the identified license conditions have been fully implemented and no further schedule submittals are necessary. Thus, this Condition is moot and can be removed as an administrative change.

Condition 2.D.(12)(g) requires SNC to complete the following prior to fuel load:

1.

Update the seismic interaction analysis in UFSAR Section 3.7.5.3 to reflect as-built information, which must be based on as-procured data, as well as the as-constructed condition;

2.

Reconcile the seismic analyses described in Section 3.7.2 of the UFSAR, to account for detailed design changes, including, but not limited to, those due to as-procured or as-built changes in component mass, center of E-8

Enclosure to NL-24-0228 Evaluation of Proposed Changes gravity, and support configuration based on as-procured equipment information;

3.

Calculate the instrumentation uncertainties of the actual plant operating instrumentation to confirm that either the design limit departure from nucleate boiling ratio (DNBR) values remain valid or that the safety analysis minimum DNBR bounds the new design limit DNBR values plus DNBR penalties;

4.

Update the pressure temperature (P-T) limits using the pressure temperature limits report (PTLR) methodologies approved in the UFSAR, using the plant-specific material properties or confirm that the reactor vessel material properties meet the specifications of and use the Westinghouse generic PTLR curves;

5.

Verify that plant-specific belt line material properties are consistent with the properties given in UFSAR Section 5.3.3.1 and Tables 5.3-1 and 5.3-3. The verification must include a pressurized thermal shock material data and the projected neutron fluence for the plant design objective. Submit this PTS evaluation report to the Director of NRO, or the Director's designee, in writing, at least 18 months before initial fuel load;

6.

Review differences between the as-built plant and the design used as the basis for the AP1000 seismic margin analysis. SNC shall perform a verification walkdown to identify differences between the as-built plant and the design. SNC shall evaluate any differences and must modify the seismic margin analysis as necessary to account for the plant-specific design and any design changes or departures from the certified design.

SNC shall compare the as-built structures, systems, and components (SSC) high confidence, low probability of failures (HCLPFs) with those assumed in the AP1000 seismic margin evaluation, before initial fuel load.

SNC shall evaluate deviations from the HCLPF values or assumptions in the seismic margin evaluation due to the as-built configuration and final analysis to determine if vulnerabilities have been introduced;

7.

Review differences between the as-built plant and the design used as the basis for the AP1000 probabilistic risk assessment (PRA) and UFSAR Table 19.59-18. SNC shall evaluate the plant-specific PRA-based insight differences and shall modify the plant-specific PRA model as necessary to account for the plant-specific design and any design changes or departure from the PRA certified in Rev. 19 of the AP1000 DCD;

8.

Review differences between the as-built plant and the design used as the basis for the AP1000 internal fire and internal flood analysis. SNC shall evaluate the plant-specific internal fire and internal flood analyses and shall modify the analyses as necessary to account for the plant-specific design and any design changes or departures from the design certified in Rev. 19 of the AP1000 DCD; and E-9

Enclosure to NL-24-0228 Evaluation of Proposed Changes

9.

Perform a thermal lag assessment of the equipment listed in UFSAR Tables 190-8 and 190-9 to provide additional assurance that this equipment can perform its severe accident functions during environmental conditions resulting from hydrogen burns associated with severe accidents.

SNC shall perform this assessment for equipment used for severe accident mitigation that has not been tested at severe accident conditions. SNC shall assess the ability of the equipment to perform during accident hydrogen burns using the environment enveloping method or the test based thermal analysis method described in Electric Power Research Institute (EPRI) NP-4354, "Large Scale Hydrogen Burn Equipment Experiments."

10.

Implement a surveillance program for explosively actuated valves (squib valves) that includes specific provisions for preservice testing and operational surveillance, in addition to the requirements specified in the edition of the ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code) as incorporated by reference in 10 CFR 50.55a.

This license condition noted to expire upon (1) incorporation of the above surveillance provisions for explosively actuated valves into the facility's inservice testing program, or (2) incorporation of inservice testing requirements for explosively actuated valves in new reactors (i.e., plants receiving a construction permit, or combined license for construction and operation, after January 1, 2000) to be specified in a future edition of the ASME OM Code as incorporated by reference in 10 CFR 50.55a, including any conditions imposed by the NRC, into the facility's inservice testing program.

Each of the identified license conditions have been fully implemented prior to the initial fuel loading, and no further actions are necessary. Thus, this Condition is moot and can be removed as an administrative change.

Condition 2.0.(13), Departures from Plant-specific DCD Tier 2* Information, provides an exemption from certain Tier 2* requirements. This condition was included via Amendment Nos. 142/141 to exempt SNC from the requirements in the regulations (in 10 CFR Part 52, Appendix D, Section VIII) for prior NRC approval of any departure from Tier 2* information or any departure from Tier 2 information that involves a change to or departure from Tier 2* information, provided that the criteria in the proposed license condition are met. However, portions of the new Condition 2.0.(13) are (or should have been identified as) moot following first full power operation of the unit pursuant to 10 CFR Part 52, Appendix D, Section VIII.B.6.c. The portions of the Condition related to the matters identified in Section VIII.B.6.c are removed from Condition 2.0.(13).

Item 1 of Condition 2.0.(13) addresses matters that "Involve a deviation from a code or standard credited in the plant-specific DCD for establishing the criteria for the design or construction of a structure, system, or component (SSC) important to safety." As noted in the Safety Evaluation (SE) for LAR 17-037 (ML18207A262),

this item encompasses matters from Paragraphs VIII.B.6.c, topics 3, 4, 13, and 16, each of which are identified as topics that revert to Tier 2 after the facility achieves first full power operation. Since both Unit 3 and Unit 4 have achieved full power E-10

Enclosure to NL-24-0228 Evaluation of Proposed Changes operation, these matters are no longer Tier 2* information and the Condition can be omitted.

Item 2 of Condition 2.D.(13) addresses matters that "Result in a change to a design process described in the plant-specific DCD that is material to implementation of an industry standard or endorsed regulatory guidance." As noted in the Safety Evaluation (SE) for LAR 17-037 (ML18207A262), this item encompasses matters from Paragraphs VIII.B.6.c, topics 9, 13, and 15, each of which are identified as topics that revert to Tier 2 after the facility achieves first full power operation. Since both Unit 3 and Unit 4 have achieved full power operation, these matters are no longer Tier 2* information and the Condition can be omitted.

Item 3(i) of Condition 2.D.(13) addresses matters that "Result in a change to the fuel criteria evaluation process, the fuel principal design requirements, or the nuclear design of the fuel or the reactivity control system that is material to a fuel or reactivity control system design function, or the evaluation process in WCAP-12488, "Westinghouse Fuel Criteria Evaluation Process"." As noted in the Safety Evaluation (SE) for LAR 17-037 (ML18207A262), this item encompasses matters from Paragraphs VIII.B.6.b, topics 2 and 3. No changes are proposed for these matters. However, this item also encompasses the matter identified in Paragraph VIII.B.6.c, topic 7, which is identified as a topic that reverts to Tier 2 after the facility achieves first full power operation. Since both Unit 3 and Unit 4 have achieved full power operation, this topic 7 portion of the matters is no longer Tier 2* information and the portion of the Condition "or the nuclear design of the fuel or the reactivity control system that is material to a fuel or reactivity control system design function," can be omitted.

Item 6 of Condition 2.D.(13) addresses matters that "Result in a change to the Passive Residual Heat Removal Heat Exchanger natural circulation test (first plant test), the Core Makeup Tank Heated Recirculation Tests (first three plants test), or the Automatic Depressurization System Slowdown Test (first three plants test) that is material to the test objectives or test performance criteria." As noted in the Safety Evaluation (SE) for LAR 17-037 (ML18207A262), this item encompasses matters from Paragraphs VIII.B.6.c, topics 10 and 11, both of which are identified as topics that revert to Tier 2 after the facility achieves first full power operation.

Since both Unit 3 and Unit 4 have achieved full power operation, these matters are no longer Tier 2* information and the Condition can be omitted.

Item 7 of Condition 2.D.(13) addresses matters that "Involve structural materials or analytical or design methods, including design codes and analytical assumptions, that deviate from those credited in the plant-specific DCD for critical sections." As noted in the Safety Evaluation (SE) for LAR 17-037 (ML18207A262), this item encompasses matters from Paragraphs VIII.B.6.c, topic 3, which is identified as a topic that reverts to Tier 2 after the facility achieves first full power operation.

Since both Unit 3 and Unit 4 have achieved full power operation, these matters are no longer Tier 2* information and the Condition can be omitted.

Item 8 of Condition 2.D.(13) addresses matters that "Result in a change to the design of the steel faceplates, internal trusses, tie bars, or headed studs of the steel-concrete (SC) module walls in the Nuclear Island or the Shield Building, E-11

Enclosure to NL-24-0228 Evaluation of Proposed Changes including SC-to-reinforced concrete (RC) connections." As noted in the Safety Evaluation (SE) for LAR 17-037 (ML18207A262), this item encompasses matters from Paragraphs VIII.B.6.c, topic 3, which is identified as a topic that reverts to Tier 2 after the facility achieves first full power operation. Since both Unit 3 and Unit 4 have achieved full power operation, these matters are no longer Tier 2*

information and the Condition can be omitted.

Item 9 of Condition 2.0.(13) addresses matters that "Result in an increase in the demand to capacity (D/C) ratio of a critical section of the structure. SNC shall determine the DIC ratio under this condition for each critical section structural member including, but not limited to, wall segments, wall sections, concrete panels, slabs, or basemat sections, affected by a departure by.... " As noted in the Safety Evaluation (SE) for LAR 17-037 (ML18207A262), this item encompasses matters from Paragraphs VIII.B.6.c, topic 3, which is identified as a topic that reverts to Tier 2 after the facility achieves first full power operation. Since both Unit 3 and Unit 4 have achieved full power operation, these matters are no longer Tier 2* information and the Condition can be omitted.

Condition 2.H addresses ITAAC closure maintenance until the Commission makes the finding under 10 CFR 52.103(g). NRC review acknowledged the completion of the ITAAC in the 10 CFR 50.103(g) letter for Unit 3 [ML20290A284] and for Unit 4

[ML22348A094], recognizing that the Acceptance Criteria identified in the COL had been satisfied. As such, there is no longer a requirement for ITAAC maintenance, and this condition can be removed as an administrative change.

Condition 2.1 requires SNC to maintain the guidance and strategies for mitigation of beyond-design-basis events developed in accordance with 10 CFR 50.54(hh)(2). This license condition was included because § 50.54(hh)(2) did not require maintenance of the guidance and strategies. On August 9, 2019, in 84 FR 39718, the 10 CFR 50.54(hh) regulations were revised and generally replaced with 10 CFR 50.155.

Paragraph 50.155(b) included a requirement to maintain the strategies and guidelines.

With the change to the regulations, the license condition is now a duplication of regulations and is no longer necessary. Therefore, this condition can be removed as an administrative change.

Condition 2.J identifies the expiration of the COL as 40 years from the 10 CFR 52.103(g) finding. As noted above, the NRC review acknowledged the completion of the IT AAC and issued the 10 CFR 50.103(g) letter for Unit 3 [ML20290A284] and for Unit 4 [ML22348A094]. Since this finding has been made, the date is known and a specific date of August 2, 2062, for Unit 3, and of July 28, 2063, for Unit 4, can be included as an administrative change.

VEGP Unit 3 and Unit 4 have both achieved initial criticality and as such the language limiting applicability of some TS to after initial criticality is no longer appropriate or useful. This change involves removal of obsolete notes from Unit 3 TS 3.4.11, TS 3.4.12, and TS 3.4.13 for Unit 3 Amendment No. 189 and from Unit 3 TS 3.3.8, TS 3.3.9, TS 3.3.10, TS 3.3.16, TS 3.3.19, TS 3.3.20, TS 3.5.7 and TS 3.5.8 for Unit 3 Amendment No. 190, and removal of obsolete text from Unit 4 LCO 3.0.7 and COL condition 2.0.(9) (including parts (a), (b), and (c)) for Unit 4 Amendment No. 189.

E-12

Enclosure to NL-24-0228 Evaluation of Proposed Changes 4

REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria 10 CFR 52.97 provides the requirements that must be met for the Nuclear Regulatory Commission to issue a combined license (COL). Such a COL was issued for Vogtle Electric Generating Plant (VEGP) Units 3 and 4. As identified in 10 CFR 52.97(c), the NRC may include terms and conditions as the NRC deems necessary and appropriate. Several such conditions were incorporated into the VEGP Units 3 and 4 COLs.

10 CFR 50.36 contains requirements for what is to be included in a licensee's TS.

The proposed note removal changes are consistent with implementation of 10 CFR 50.36.

4.2 Precedent The NRC approved a request for Shearon Harris Nuclear Power Plant that included a change to the Renewed Facility Operating License to delete License Condition 2.G (consistent with VEGP Units 3 and 4 Condition 2.D.(7) by letter dated March 10, 2022 [ML22020A007]. This change was requested as part of the consolidated line item improvement process and consistent with the model safety evaluation published in the Federal Register 70 FR 67202.

The request is also consistent with past NRC practice of approval of amendments to remove obsolete license conditions and TS notes. Two such precedent amendments are:

Vogtle Electric Generating Plant (VEGP) Units 1 and 2 Amendments Nos. 107 and 85, respectively (ML012390169)

Joseph M. Farley Nuclear Plant, Units 1 and 2 Amendment Nos. 152 and 144, respectively (ML003780854) 4.3 Significant Hazards Consideration Southern Nuclear Operating Company (SNC) is requesting an amendment to Combined License (COL) Nos. NPF-91 and NPF-92 for Vogtle Electric Generating Plant (VEGP) Units 3 and 4, respectively. The license amendment request (LAR) proposes to remove completed license conditions, to modify other license conditions, and to remove some unit-specific Technical Specification (TS) and COL language that is no longer applicable in the combined licenses. An evaluation to determine whether or not a significant hazards consideration is involved with the proposed amendment was completed by focusing on the three standards set forth in 10 CFR 50.92(c), "Issuance of amendment," as discussed below.

E-13

Enclosure to NL-24-0228 Evaluation of Proposed Changes

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

Response: No The proposed changes do not affect accident evaluations since there are no changes to the plant, no changes to analysis of the plant, and no changes to testing of the plant. The proposed changes do not adversely affect the operation of any structures, systems, or components (SSCs) associated with an accident initiator or initiating sequence of events. The proposed changes continue to maintain the initial conditions and operating limits assumed during normal operation, assumed by the accident analysis, and assumed in anticipated operational occurrences. Therefore, the proposed changes do not result in any increase in probability of an analyzed accident occurring.

The proposed changes do not involve a change to any mitigation sequence or the predicted radiological releases due to postulated accident conditions.

Thus, the consequences of the accidents previously evaluated are not adversely affected.

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

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

Response: No The proposed changes have been found to continue to provide the required functional capability of the safety systems for previously evaluated accidents and anticipated operational occurrences. The proposed changes do not change the function of the related systems, and thus, the changes do not introduce a new failure mode, malfunction or sequence of events that could adversely affect safety or safety-related equipment.

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 amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed changes continue to provide the required functional capability of the safety systems for previously evaluated accidents and anticipated operational occurrences. The proposed changes do not change the function of the related systems nor significantly affect the margins provided by the E-14

Enclosure to NL-24-0228 Evaluation of Proposed Changes systems. No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the requested changes.

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

Based on the above, it is concluded 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 Based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public. Therefore, it is concluded that the requested 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.

5 ENVIRONMENTAL CONSIDERATION A review has determined that the proposed changes require an amendment to the COL. A review of the anticipated operational effects of the requested amendment has determined that the requested amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9), in that the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b),

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

6 REFERENCES None E-15

ATTACHMENT 1 to NL-24-0228 Licensing Basis Document Markups License Amendment Request:

Remove or Modify Outdated License Information (This enclosure consists of 60 pages, including this cover page)

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 COMBINED LICENSE No change on this page.

Provided for context.

VOGTLE ELECTRIC GENERATING PLANT UNIT 3 SOUTHERN NUCLEAR OPERA TING COMPANY, INC.

GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MEAG POWER SPVM, LLC MEAG POWER SPVJ, LLC MEAG POWER SPVP, LLC CITY OF DAL TON, GEORGIA Docket No.52-025 License No. NPF-91

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for a combined license (COL) for Vogtle Electric Generating Plant (VEGP) Unit 3 filed by Southern Nuclear Operating Company, Inc. (SNC) acting on behalf of Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 1 and the City of Dalton, Georgia, an incorporated municipality in the state of Georgia acting by and through its Board of Water, Light and Sinking Fund Commissioners (City of Dalton), herein referred to as "the VEGP owners," which incorporates by reference Appendix D to 10 CFR Part 52 and Early Site Permit No. ESP-004, complies with the applicable standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission regulations set forth in 10 CFR Chapter I, and all required notifications to other agencies or bodies have been duly made; B.

There is reasonable assurance that the facility will be constructed and will operate in conformity with the application, as amended, the provisions of the Act, and the Commission regulations set forth in 10 CFR Chapter I, except as exempted from compliance in Sections 2.F and 2.G below; C.

There is reasonable assurance (i) that the activities authorized by this COL can be conducted without endangering the health and safety of the public and (ii) that such activities will be conducted in compliance with the Commission regulations set forth in 10 CFR Chapter I, except as exempted from compliance in Sections 2.F and 2.G below; On June 24, 2015, Municipal Electric Authority of Georgia transferred its ownership interest to its wholly owned subsidiaries:

MEAG Power SPVM, LLC; MEAG Power SPVJ, LLC; and MEAG Power SPVP, LLC as described in the SNC letter dated December 2, 2013 and in the Commission's Safety Evaluation available in the Agencywide Document Access and Management System (ADAMS) under Accession No. ML14072A340.

1 Amendment No. 36

No change on this page.

Provided for context.

D.

SNC2 is technically qualified to engage in the activities authorized by this license in accordance with the Commission regulations set forth in 10 CFR Chapter I.

SNC and the VEGP owners together are financially qualified to engage in the activities authorized by this COL in accordance with the Commission regulations set forth in 1 O CFR Chapter I; E.

SNC and the VEGP owners have satisfied the applicable provisions of 10 CFR Part 140, "Financial Protection Requirements and Indemnity Agreements;"

F.

The issuance of this license will not be inimical to the common defense and security or to the health and safety of the public; G.

After weighing the environmental, economic, technical, and other benefits of the facility against environmental and other costs and considering reasonable available alternatives, the issuance of this license subject to the conditions for protection of the environment set forth herein is in accordance with Subpart A of 10 CFR Part 51 and all applicable requirements have been satisfied; and H.

The receipt, possession, and use of source, byproduct, and special nuclear material as authorized by this license will be in accordance with the applicable regulations in 10 CFR Parts 30, 40, and 70.

2.

On the basis of the foregoing findings regarding this facility, COL No. NPF-91 is hereby issued to SNC, Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and the City of Dalton, Georgia (the licensees) to read as follows:

A.

This license applies to the VEGP Unit 3, a light-water nuclear reactor and associated equipment (the facility), owned by the VEGP Owners. The facility would be located adjacent to existing VEGP Units 1 and 2 on a 3, 169-acre coastal plain bluff on the southwest side of the Savannah River in eastern Burke County, GA, approximately 15 miles east-northeast of Waynesboro, GA, and 26 miles southeast of Augusta, GA, and is described in the licensees' updated final safety analysis report (UFSAR), as supplemented and amended.

B.

Subject to the conditions and requirements incorporated herein, the Commission hereby licenses:

(1)

SNC pursuant to Sections 103 and 185b. of the Act and 10 CFR Part 52, to construct, possess, use, and operate the facility at the designated location in accordance with the procedures and limitations set forth in this license; (2)

The VEGP owners pursuant to the Act and 10 CFR Part 52, to possess but not operate the facility at the designated location in Burke County, GA, in accordance with the procedures and limitations set forth in this license; 2 SNC is authorized by the VEGP owners to exercise responsibility and control over the physical construction, operation, and maintenance of the facility.

2 Amendment No. 125

No change on this page.

Provided for context.

(3)

(a)

SNC pursuant to the Act and 10 CFR Part 70, to receive and possess at any time, special nuclear material as reactor fuel, in accordance with the limitations for storage and in amounts necessary for reactor operation, described in the UFSAR, as supplemented and amended; (b)

SNC pursuant to the Act and 10 CFR Part 70, to use special nuclear material as reactor fuel, after a Commission finding under 10 CFR 52.103(g) has been made, in accordance with the limitations for storage and in amounts necessary for reactor operation, described in the UFSAR, as supplemented and amended; (4)

(a)

SNC pursuant to the Act and 10 CFR Parts 30 and 70, to receive, possess, and use, at any time before a Commission finding under 10 CFR 52.103(g), such byproduct and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts, as necessary; (b)

SNC pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use, after a Commission finding under 10 CFR 52.103(g), any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as necessary; (5)

(a)

SNC pursuant to the Act and 10 CFR Parts 30 and 70, to receive, possess, and use, before a Commission finding under 10 CFR 52.103(g),

in amounts not exceeding those specified in 10 CFR 30.72, any byproduct or special nuclear material that is (1) in unsealed form; (2) on foils or plated surfaces, or (3) sealed in glass, for sample analysis or instrument calibration or other activity associated with radioactive apparatus or components; (b)

SNC pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use, after a Commission finding under 10 CFR 52.103(g), in amounts as necessary, any byproduct, source, or special nuclear material without restriction as to chemical or physical form, for sample analysis or instrument calibration or other activity associated with radioactive apparatus or components but not uranium hexafluoride; and (6)

SNC pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C.

The license is subject to, and the licensees shall comply with, all applicable provisions of the Act and the rules, regulations, and orders of the Commission, including the conditions set forth in 10 CFR Chapter I, now or hereafter in effect.

3 Amendment No. 125

D.

The license is subject to, and SNC shall comply with, the conditions specified and incorporated below:

(1)

(2)

Changes during Construction ~

- Removed by Amendment No. ###

£~JC n=iay reeiuest use of a prelin=iinary amendment reeiuest (Pl\\R) proeess, for lieense amendments, at any time before a Commission finding under 1 O CP:R a2.103(g). To use the Pl\\R proeess, S~JC shall submit a written reeiuest to the Offiee of ~Jew Roasters (~JRO) in aooordanoe with COL I£G 02a, "Changes during Construotion under Part a2."

Before ~JRO's issuanoe of a written Pl\\R notifioation, £~JC shall submit the lieense amendn=ient reeiuest (LAR). Therea~er, ~JRO will issue a written PAR notifieation, setting forth whether S~JC may proeeed in aeeordanee with the Pl',R, LAR, and COL ISG 026. If S~JC eleets to proeeed and the Ll',R is subseeiuently denied, S~JC shall return the faeility to its eurrent lioensing basis.

Pre-operational Testing ~<:-----11 ~ Removed by Amendment Nos. 1921

_and###

f9j S~JC shall perform the design speeifie pre operational tests identified below:

£~JC shall rm1iew and 0 1.<<aluate the results of the tests identified in Seetion 2.D.(2)(a) of this lieense and eonfirm that these test results are within the range of aeeeptable values predieted or otherwise oonfirm that the tested systen=is perform their speoified funotions in aooordanoe with UP:£/\\R £eotion 14.2.Q, Amendment No. in each location including LC 2.D.(8) and at the bottom of each changed page to be inserted when determined.

4 Amendment No. 474

(3)

(4)

£~JG shall notify' the DiFeotoF of ~JRO, OF the DirnotoF's designee, in writing, upon sueeessful eoR=tpletion of the design speeifie pre opemtional tests identified in £eotion 2.D.(2)(a) of this lioense; and (ReR10Yed by l',R1endR=tent ~Jo. 192)

Nuclear Fuel Loading and Pre-critical Testing

- Removed by Amendment Nos. 192 and ###

f91 fsj fel:1 Until the subR1ission of the notifioation Fequirnd by

£eotion 2.D.(2)(o) of this lioense, £~JG shall not load fuel into the reaetor vessel; (ReR10Yed by l',R=tendR=tent ~Jo. 192)

£~JG shall pOFfoFR=t the pFe GFitioal tests identified in UP:£,1\\R

£eotion 14.2.10.1;

£~JG shall Fe1o1iew and e1o1aluate the Fesults of the tests identified in Seetion 2.D.(a)(e) of this lieense and eonfirR1 that these test Fesults arn within the mnge of aooeptable 1o1alues prndioted OF otherwise oonfiFR=i that the tested systeR=ts pOFfoFR=t theiF speoified funetions in aeeordanee with UFSl',R Seetion 14.2.1 O; and

£~JG shall notify' the DiFeotoF of ~JRO, OF the DirnotoF's designee, in WFiting, upon suooessful GOR=tpletion of the pFe GFitioal tests identified in Seetion 2.D.(a)(e) of this lieense.

_ ~

.... _- R-em_ o_v_e_d_b_y_A_m_e-nd_m_ e_n_t -.

Initial Criticality and Low-Power Testing /? I

.... N_o_._#_#_# ________ __.

~

f91 Upon subR=tission of the notifieation required by Seetion 2.D.(a)(e) of this lioense, £~JG is authoFiced to opeFate the faoility at rnaotoF steady state ooFe poweF le1o1els not to mmeed a pernent theFR=tal poweF in aoooFdanoe with the oonditions speoified hernin;

£~JG shall peFfoFR=t the initial GFitioality and low poweF tests identified in UP:£,1\\R £eotions 14.2.10.2 and 14.2.1 a.a, rnspeoti1o1ely; 5

Amendment No. ~

(5)

(6) f61

£~JG shall re1,ciew and 0 1,caluate the results of the tests identified in Seetion 2.D.(4)(b) of this lieense and eonfirRl that these test results are within the range of aeeeptable values predieted or otherwise oonfirR1 that the tested systeR1s perforRl their speoified funotions in aooordanoe with UP:£,1\\R eeotions 14.2.10.2 and 14.2.1 a.a; and fe-1

£~JG shall notify the Direotor of ~JRO, or the Direotor's designee, in writing, upon sueeessful eoR1pletion of initial eritieality and low power tests identified in eeotion 2.D.(4)(b) of this lioense, ineluding the design speeifie tests identified therein.

Power Ascension Testing ~(----11-Removed by Amendment No. tt##

fat Upon subR1ission ofthe notifioation required by eeotion 2.D.(4)(d) of this lioense, £NG is authoriced to operate the faoility at roaster steady state eore power levels not to mmeed 100 pereent therR1al power in aeeordanee with the eonditions speeified herein, but only for the purpose of perforRling power asoension testing; f91

£~JG shall perforRl the power asoension tests identified in UP:£,1\\R Seetion 14.2.10.4; f61

£~JG shall re1,ciew and 0 1,caluate the results of the tests identified in eeotion 2.D.(a)(b) of this lioense and oonfirRl that these test results are within the range of aeeeptable values predieted or otherwise eonfirRl that the tested systeR1s perforRl their speeified funotions in aooordanoewith UP:£,1\\R eeotion 14.2.10.4; and fe-1 S~JG shall notify the Direetor of ~JRO, or the Direetor's designee, in writing, upon suooessful 00R1pletion of power asoension tests identified in Seetion 2.D.(e)(b) of this lieense, ineluding the design speoifio tests identified therein.

Maximum Power Level Upon subR1ission of the notifioation required by eeotion 2.D.(a)(d) of this lieense, SNC is authorized to operate the facility at steady state reactor core power levels not to exceed 3400 MW thermal (100-percent thermal power), as described in the UFSAR, in accordance with the conditions specified herein.

6 Amendment No. 4eQ.

(7)

(8)

(9)

Reporting Requirements ~

_ Removed by Amendment No. ###

(a)

Within ao days of a ehange to the initial test progran=i deseribed in UF"S/\\R Seetion 14, Initial Test Progran=i, n=iade in aeeordanee with 1 0 GP:R a0.aQ or in aooordanoe with 1 0 GP:R Part a2, /\\ppendi>< D, Seotion VIII, "Prooesses for Changes and Departures," SNG shall report the ehange to the Direetor of ~JRO, or the Direetor's designee, in aeeordanee with 1 0 GF"R e0.e9(d).

fat-S~JG shall report any violation of a requiren=ient in Seetion 2.D.(a),

Seetion 2.D.(4 ), Seetion 2.D.(e), and Seetion 2.D.(6) of this lieense within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Initial notifioation shall be n=iade to the NRG Operations Genter in aeeordanee with 1 0 GF"R e0.72, with written follow up in aeeordanee with 1 0 GF"R e0.?a.

Incorporation The Technical Specifications and Environmental Protection Plan in Appendices A and B, respectively of this license, as revised through Amendment No. 4Qg., are hereby incorporated into this license.

Technical Specifications ""-<:---- Removed by Amendment No. ### I The teehnieal speei:f.ieations in l',ppendi>< /\\ to this lieense beeon=ie effeoti1w<<e upon a Gon=in=iission finding that the aooeptanoe oriteria in this lioense (ITAAG) are R=1et in aooordanoe with 1 0 GP:R a2.1 oa(g).

(10)

Operational Program Implementation...,_<--- Removed by Amendment No. ###

S~JG shall iR=lplen=ient the progran=is or portions of prograR=1s identified below, on or before the date S~JG aohie1w<<es the following n=iilestones:

fat E:nvironn=iental Quali:f.ieation Progran=i in=iplen=iented before initial fuel load; Reaetor >Jessel Material Surveillanee PrograR=I in=ipleR=1ented before initial eritieality; Preserviee Testing PrograR=I in=iplen=iented before initial fuel load; Gontainn=ient Leal~age Rate Testing Progran=i in=ipleR=1ented before initial fuel load; Fire Proteetion Progran=i The fire proteetion n=ieasures in aeeordanee with Regulatory Guide (RC) 1.189 for designated storage building areas (inoluding adjaoent fire areas that oould affeot the storage area) in=iplen=iented before initial reoeipt 7

Amendment No. 498

ffi1 fij fB of b,iproduot or speoial nuolear n=iaterials that are not fuel (mmluding mmn=ipt quantities as desoribed in 10 CP:R ao.18);

~

The fire proteetion n=ieasures in aeeordanee with R:G 1.189 for areas oontaining new fuel (inoluding adjaoent areas where a fire eould affeet the new fuel) implemented before reoeipt of fuel onsite; 3:-

/\\II fire proteotion program features in=iple n=iented before initial fuel load; Standard R:adiologieal E:ffluent Controls implen=iented before initial fuel load; Offsite Dose Caloulation Manual in=iplen=iented before initial fuel

~

R:adiologieal E:nvironmental Monitoring Progran=i in=iplen=iented before initial fuel load; Prooess Control Progran=i implen=iented before initial fuel load; Radiation Proteotion Progran=i (RPP) (inoluding the /\\LARI\\

prineiple) or applieable portions as identified in UrSI\\R: Seetion 12.a thereof:

4:-

RPP features applioable to reoeipt of b,i produot, souroe, or speeial nuelear n=iaterials (mmluding mcen=ipt quantities as desoribed in 1 0 CP:R a0.18) implemented before initial reeeipt of sueh n=iaterials;

~

R:PP features (ineluding the P,LAR:A prineiple) applieable to new fuel in=iplen=iented before reoeipt of initial fuel on site; 3:-

,1\\11 other RPP features (inoluding the Abl\\RA prinoiple) mcoept for those applioable to oontrol rad ioaoti1t1e waste shipn=ient in=iplen=iented before initial fuel load; 4:-

RPP features (ineluding the /\\LARA prineiple) applieable to radioaoti1t1e waste shipment implen=iented before first shipment of rad ioaetive waste; R:eaetor Operator Training Progran=i in=iplemented 18 n=ionths before the soheduled date of initial fuel load; Motor Operated Val1t1e Testing Progran=i in=iplen=iented before initial fuel load; 8

Amendment No. ~

IAitial Test Prograffl (ITP)

PreoperatioAal Test Prograffl ifflpleR1eAted before the first preoperatioAal test; Ad

£tart1,1p Test Prograffl ifflpleffleAted before iAitial f1,1el load* '

£peoial ~J1,1olear Material CoAtrol Ad l\\0001,1AtiAg Prograffl iR1pleffleAted before iAitial reeeipt of speeial A1,1elear ffl terial; Ad f01 Speeial ~J1,1elear Material Physieal ProteetioA Prograffl iR1pleffleAted before iAitial reoeipt of speoial A1,1olear ffl terial oA s+te-: *

- Removed by Amendment (11)

Operational Program Implementation Schedule k:- No. ###

~Jo later thaA 12 fflOAths after iss1,1 Aee of the COL, S~JC shall s1,1bfflit to the Direetor of NRG, or the Direetor's desigAee, a sehed1,1le for ifflpleffleAtatioA of the operatioAal prograR1s listed iA UP:£,1\\R Table 1 a. 4 201, iAel1,1diAg the assoeiated estifflated date for iAitial loadiAg of f1,1el. The sehed1,1le shall be 1,1pdated every 6 fflOAths 1,1Atil 12 R10Aths before sohed1,1led f1,1el loadiAg, Ad m,ery fflOAth thereafter l,lAtil all the operatioAal progrff'IS listed iA UP:£,1\\R Table 1 d.4 201 hm/0 beeA f1,1lly ifflpleffleAted.

(12)

Site-and Unit-specific Conditions ~L---11 -Removed by Amendment No. ###

W

£~JC shall either refflo1,ce Ad replaoe, or shall ifflpro1,ce, the soils direotly abo1,ce the bl1,1e bl1,1ff ffl rl for soils 1,1Ader or adj GeAt to Seisfflie Category I str1,1et1,1res, to eliffliAate Ry liq1,1efaetioA poteAtial.

f91 Before 60fflR1eAOiAg iAstallatioA of iAdivid1,1al pipiAg segR1eAtS Ad 0OAAeeted 0OfflpOAeAtS iA their fiAal loeatiOAS, S~JC shall 0Offlplete the as desigAed pipe r1,1pt1,1re hazmrds Aalysis for oofflp rtffleAts (rOOfflS) 0OAtiAiAg those segffleAtS iA aeeordaAee with the eriteria Ol,ltliAed iA the UFSI\\R SeetiOAS d.e.1.a.2 Ad d.e.2.e, Ad shall iAforffl the Direotor of ~JRO, or the Direotor's desigAee, iA writiAg, 1,1poA the oofflpletioA of this Aalysis Ad the a1,cailability of the as desigAed pipe r1,1pt1,1re ha2:ards Aalysis reports.

t61 Before GOfflffleAGiAg iAstallatioA of iAdi1,cid1,1al pipiAg segffleAts, ideAtified iA UFSI\\R SeetioA a.9.8.7, Ad eoAAeeted eofflpoAeAts iA their fiAal loeatioAs iA the faeility, g~JC shall eofflplete the Aalysis of the as desigAed iAdi1,cid1,1al pi pi Ag segffleAtS Ad shall iAforffl the Direotor of ~JRO, or the Direotor's 9

Amendment No. 4e4

designee, in writing, upon U:ie eon:ipletion of tl:iese an l}'ses and tl:ie a1o1ailabilit}' of tl:ie design reports for tl:ie seleoted piping pael(ages.

fel-1

~Jo later tl:ian 180 d J'S before initial fue l load, S~JG sl:iall subn:iit to tl:ie Direotor of NRG, or tl:ie Direotor's designee, in writing, a full}' de1o1eloped set of plant speeifie en:iergene}' aetion levels (E:l',Ls) for VE:GP Unit a in aooordanoe witl:i tl:ie oriteria defined in l\\n:iendn:ient ~Jo. 77. Tl:ie eALs sl:iall Rave been diseussed and agreed upon witR State and loeal offieials.

~Jo later tl:ian 180 d J'S before initial fuel load, S~JG sl:iall subn:iit to tl:ie Direotor of ~JRO, or tl:ie Direotor's designee, in writing, an assessn:ient of en:iergeno}'

response staffing perforn:ied in aooordanoe witl:i Nel 1 0 0a, "Assessn:ient of On £Rift en:iergene}' Response Organii!:ation Staffing and Capabilities,"

Revision 0.

fe1 S~JG st:iall not revise or n:iodif}' tt:ie provisions of Seetions e.a, e.4, e.e, e.9, and e.1 0 of tl:ie Speeial Nuelear Material (S~JM) PR}'Sieal Proteetion Progran:i until tl:ie requiren:ients of 1 0 GP:R 73.aa are in:iplen:iented.

fB

~Jo later tl:ian 12 n:iontl:is after issuanoe of tl:ie COL, S~JG sl:iall subn:iit to tl:ie Direotor of ~JRO, or tl:ie Direotor's designee, a sel:iedule for in:iplen:ientation of tl:ie following lieense eonditions.

Tl:ie sel:iedule sl:iall be updated every 6 n:iontl:is until 12 n:iontl:is before sel:ieduled fuel loading, and every n:iontl:i tl:iereafter until eael:i lieense eondition l:ias been full}' in:iplen:iented. Tl:ie sel:iedule sl:iall identif}' tl:ie oon:ipletion of or in:iplen:ientation of tl:ie following:

+.

Tl:ie eonstruetion and inspeetion proeedures for steel eonerete eon:iposite (SC) oonstruotion aoti1o1ities for seisn:iio Category I nuolear island n:iodules (ineluding sl:iield building SC n:iodules) deseribed in UFSl',R Seotion a.8.4.8;

~ Tl:ie spent fuel rael( Metan:iie Coupon n:ionitoring progran:i (before initial fuel load);

3e-ln:iplen:ientation of tl:ie flow aooelerated oorrosion (FAG) progran:i inoluding oonstruotion pl:iase aoti1o1ities (before initial fuel load);

4.-

A turbine n:iaintenanee and inspeetion progran:i, wl:iieR n:iust be eonsistent witR tl:ie n:iaintenanee and inspeetion progran:i plan aetivities and inspeetion inteFYals identified in UP:Sl\\R Seotion 10.2.3.6 (before initial fuel load);

e,-

Tl:ie a1o1ailabilit}' of dooun:iented instrun:ientation uneertainties to ealeulate a power ealorin:ietrie uneertaint}'

(before initial fuel load);

10 Amendment No. ~

ffij The availability of adlTlinistFative eontrnls to ilTlplen=ient IT'l intenanoe and oontingenoy aoti1w'ities Felated to the poweF oaloFin=ietFio 1,moertainty instn,11T1entation (beforn initial fl,1OI load);

The site speoifio se1t'OFe aooident n=ianagen=ient gc1idelines (befoFe startc1p testing);

The opOFational and prngmlTln=iatie elen=ients of the 1T1itigati1w'e stmtegies foF FOsponding to oiFoc1n=istanoes assoeiated with loss of l Fge aFeas of the plant dc1O to m(plosions OF fiFe de1t'eloped in aoooFdanoe with 1 O GF"R eO.e4(hh)(2) (beforn initial fc1el load); and Q.,.

The ITP pFOOOdclFOS identified in U F"SI\\R Seetion 1 4.2.a:

a. adn=iinistFatiiw'O ITl nc1al (befOFO the fiFst prnopemtional test)

b. prnopOFational testing (beforn sehedc1led per:fOFITl noe)

o. startc1p testing (befoFO initial fc1el load)

BefOFO initial fc1el load, SNG shall:

4,-

Update the seislTlie intemetion analysis in UF"SI\\R Seotion a.7.6.a to FOfleot as bc1ilt infoFITltion, whioh 1T1c1st be based on as pFoec1Fed data, as well as the as oonstFc1oted oondition;

~

Reoonoile the seisn=iio analyses desoFibed in Seotion a.7.2 of the UP:8.'\\R, to aoooc1nt foF detailed design ohanges, inelc1ding, bc1t not lilTlited to, those dc1O to as pFOOblFOd OF as bc1ilt ehanges in eo1T1ponent IT'l ss, eenteF of gFavity, and sc1pport oonfigc1Fation based on as pFooc1Fed eqc1ip1T1ent infoFITl tion; 4--:

Galec1late the instFc1n=ientation c1neertainties of the aetc1al plant opOFating instFbllTIOntation to oonfiFITI that eitheF the design lilTlit dep FtclFO fFOn=I nc1oleate boiling Ftio (D~JBR) valc1es FOIT'l in valid OF that the safety analysis n=iini1T1c11T1 D~JBR boc1nds the new design lilTlit D~JBR valc1es plc1s o~mR penalties; Update the pFOSSblFO ten=ip0Fatc1FO (P T) lilTlits c1sing the pFOSSblFO telTlpOFtblFO lin=iits Feport (PTLR) ITIOthodologies appFoved in the UF"SAR, c1sing the plant speeifie ITltOFial pFoperties OF oonfiFITI that the FO GtOF 1w'OSSOI ITl tOFial pFoperties n=ieet the speeifieations of and c1se the Westinghoc1se genOFiG PTLR GblPw'OS; eo-VOFify that plant speoifio belt line n=iatOFial pFoperties am eonsistent with the pFoperties given in UF"SI\\R Seetion e.d.d.1 and Tables e.d 1 and e.d d. The VOFifieation ITlblSt inelc1de a pFOSSblFi:z:ed thOFITl I shoal(

11 Amendment No. 4e4

(PTS) e>o<aluation based on as pFoeuFed FeaetoF >o<essel FAateFial data and the pmjeoted neutFon fluenoe foF the plant design objeeti>o<e. SubFAit this PTS e>o<aluation Feport to the 0iFeotoF of MRO, OF the 0iFeotoF's designee, in wFiting, at least 18 FAonths befoFe initial fuel load; Re>o<iew differnnees between the as built plant and the design used as the basis foF the l\\P1000 seisFAio FAargin anal}'sis. S~JC shall peFfoFFA a >o<eFifieation wall(down to identif}' differnnoes bet:ween the as built plant and the design. S~JC shall m,1aluate an}' diffeFenoes and FAust FAodif}' the seisFAio FAargin anal}'sis as neoessary to aooount foF the plant speoifio design and an}' design ehanges OF departuFes fFOFA the eertified design. S~JC shall GOFApaFO the as built stFUffiUFes, S}'SteFAS, and eoFAponents (SSC) high eonfidenee, low pmbabilit}' of failurns (l=ICLPP:s) with those assuFAed in the l\\P1000 seisFAie FAargin e>o<aluation, befoFe initial fuel load. SMC shall m.<<aluate de1o1iations fFOR=I the l=ICLPP: 1o1alues OF assuFAptions in the seisFAie R=laFgin e>o<aluation due to the as built oonfiguFation and final anal}'sis to deteFR=line if

>o<ulnembilities ha>o<e been intFodueed; 7-:

Re>o<iew differnnees between the as built plant and the design used as the basis foF the l\\P1000 pmbabilistio Fisl(

assessFAent (PRI\\) and UFSI\\R Table 19.69 18. SMC shall e1o1aluate the plant speoifio PRA based insight diffeFOnees and shall FAOdif}' the plant speeifie PRA FAOdel as neoessary to aooount foF the plant speoifio design and an}' design ohanges OF departuFe fFOFA the PRA oertified in Rev. 19 of the AP1 000 0C0 ;

g.,.

Re1o1iew differnnoes bet:ween the as built plant and the design used as the basis foF the AP1000 intemal fiFe and intemal flood anal}'sis. SMC shall e>o<aluate the plant speoifio intemal fiFe and intemal flood anal}'ses and shall FAodif}' the anal}'ses as neeessary to aeeount foF the plant speoifio design and an}' design ohanges OF departuFes fFOFA the design eertified in Re>o<. 19 of the l\\P1000 0C0; and Q.,.

PetlOFFA a theFR=lal lag assessFAent of the equipR=lent listed in UFSAR Tables 190 8 and 190 9 to pFOYide additional assuFanoe that this equipFAent oan perlOFR=I its se1o'eFe aeeident funetions duFing en>o<iFonFAental eonditions Fesulting fFOR=I h}'dFogen bums assooiated with se1o1eFe aeeidents. S~JC shall per;oFFA this assessFAent foF equipFAent used foF se1o1eFe aooident FAitigation that has not been tested at seveFe aeeident eonditions. S~JC shall assess the abilit}' of the equipFAent to peFfoFFA 12 Amendment No. ~

during accident hydrogen burns using the en1.iironment en1,eloping R=1ethod or the test based therR=1al analysis R=1ethod described in electric Power Research Institute (ePRI) ~JP 4354, "Large Scale Hydrogen Burn equipment experiR=lents."

4Q.,.

IR=1pleR=1ent a suri.ieillance prograR=I for explosi1.iely actuated 1.ialves (squib 1.ial1.ies) that includes the following pro1.iisions in addition to the requireR=1ents specified in the edition of the /\\SMe Gode for Operation and M-aintenanoe of N{;}o/ear P.ma,,er P-1-ants (OM Gode) as incorporated by reference in 10 ci;R 50.55a.

a-:-

Preservice Testing l\\II mcplosi1,ely actuated 1,aI1,es shall be preservice tested by 1,erifying the operational readiness of the actuation logic and associated electrical circuits for each mcplosi1, ely actuated 1, aI 1,e with its pyrotechnic charge reR=101,ed from the 1, aI1, e. This must include confirR=lation that sufficient electrical paraR=1eters

( 1.ioltage, current, resistance) are a1.iailable at the mcplosi1,ely actuated 1,aI1,e froR=I each circuit that is relied upon to actuate the 1.ial1.ie. In addition, a saR=1ple of at least 20% of the pyrotechnic charges in all mcplosi1, ely actuated 1,aI1,es shall be tested in the 1,aI1,e or a qualified test fo<ture to confirR=I the capability of each saR=1pled pyrotechnic charge to pro1,ide the necessary motii,e force to operate the 1,aI1, e to perforR=I its intended function without daR=1age to the 1.ial1.ie body or connected piping.

The saR=1pling R=1ust select at least one explosi1.iely actuated val1.io froR=I each redundant safety train.

Gorrectii.ie action shall be taken to resol1.ie any deficiencies identified in the operational readiness of tho actuation logic or associated electrical circuits, or tho capability of a pyrotechnic charge. If a charge fails to fire or its capability is not confirR=lod, all charges with tho saR=10 batch nuR=1bor shall be roR=101.iod, discarded, and replaced with charges from a different batch nuR=1ber that has deR=1onstrated successful 20% SaR=lpling of the charges.

Operational Surveillance explosi1.ioly actuated 1.ial1.ios shall be subject to tho following surveillance activities after coR=1mencing plant operation:

13

fh H+-e 14 l\\t least once e1Jery 2 years, each m~plosi1;ely actuated 1Jal 1Je shall undergo 1; isual m~ternal mmR=lination and reR=1ote internal e:xaR=lination (including e1Jaluation and reR=101Jal of fluids or contaR=linants that R=1ay interfere with operation of the 1Jal1Je) to 1Jerify the operational readiness of the 1Jal1Je and its actuator. This e:xaR=lination shall also 1;erify the appropriate position of the internal actuating R=1echanisR=1 and proper operation of reR=1ote position indicators.

Gorrecti1;e action shall be tal(;en to resol1Je any deficiencies identified during the e:xaR=lination with post R=laintenance testing conducted that satisfies the preservice testing reeiuireR=1ents.

l\\t least once e1Jery 1 O years, each m~plosi1;ely actuated 1Jal 1Je shall be disasseR=1bled for internal mmR=lination of the 1Jal1Je and actuator to 1Jerify the operational readiness of the 1Jal1Je asseR=1bly and the integrity of indi1Jidual coR=1ponents and to reR=101Je any foreign R=1aterial, fluid, or corrosion. The e:xaR=lination schedule shall pro1; ide for both of the two 1Jal1Je designs used for m~plosi1;ely actuated 1Jal1Jes at the facility to be included aR=1ong the e:xplosi1Jely actuated 1Jal1;es to be disasseR=1bled and mmR=lined e1;ery 2 years. Gorrecti1;e action shall be taken to resol1Je any deficiencies identified during the e:xaR=lination with post R=1aintonanco testing conducted that satisfies the preservice testing reeiuireR=1ents.

i;:or o:xplosi1Joly actuated 1Jal1Jos soloctod for test saR=1pling e1Jery 2 years in accordance 1,4.iith tho ASl\\4e 01\\4 Godo, tho operational readiness of tho actuation logic and associated electrical circuits shall be verified for each saR=1pled e:xplosi1Jely actuated 1Jal1Je following roR=101Jal of its charge. This R=1ust include confirR=lation that sufficient electrical paraR=1eters (voltage, current, resistance) are a1Jailablo for each val1Jo actuation circuit. GorroctiiJo action shall be taken to resol1Je any deficiencies identified in the actuation logic or associated electrical circuits.

+v-:-

For mcplosi>v<ely aetuated >v<al>v<es seleeted for test saR=1pling m.iery 2 years in aooordanoe with the ASME: OM Gode, the saR=tpling R=tust seleot at least one 0><plosi1o1ely aetuated >v<al>v<e froR=t eaeh redundant safety train. eaoh saR=tpled pyroteohnio oharge shall be tested in the valve or a qualified test fi Eture to oonfirR=i the oapability of the eharge to provide the neeessary R=totive foroe to operate the 1o1al1o10 to perforR=I its intended funetion without daR=1age to the 1o1al1o10 body or oonneoted piping. Gorreoti1o10 aetion shall be talrnn to resolve any defioienoies identified in the oapability of a pyroteehnie eharge in aeeordanee with the preservioe testing requireR=tents.

This lioense oondition shall mcpire upon (1) inoorporation of the abo1o10 surveillanoe pro1o1isions for 0><plosi1o1ely aotuated 1o1al1o1es into the faoility's inservioe testing prograR=t, or (2) ineorporation of inseFYiee testing requireR=tents for 0><plosi1o1ely aotuated 1o1al1o1es in new roasters (i.e., plants reeei>v<ing a eonstruetion perR=tit, or eoR=tbined lieense for oonstruotion and operation, after January 1, 2000) to be speeified in a future edition of the ASME: OM Gode as inoorporated by referenoe in 1 0 GP:R a0.aaa, inoluding any eonditions iR=iposed by the NRG, into the faeility's inseFYiee testing prog rR=I.

(13)

Departures from Plant-specific DCD Tier 2* Information (a)

SNC is exempt from the requirements of 10 CFR Part 52, Appendix D, Paragraphs VIII.B.6 and VIII.B.5.a for prior NRC approval of departures from Tier 2* information and departures from Tier 2 information involving a change to or departure from Tier 2* information; except for departures that:

1.

2.

3.

ln1o1ol1o10 a de1o1iation froR=t a oode or standard oredited in the plant speoifio DGD for establishing the oriteria for the design or eonstruetion of a strueture, systeR=t, or eoR=tponent (SSC) iR=iportant to safety,, __ ___, - Removed by Amendment No. ###

Result in a ohange to a design prooess desoribed in the plant speeifie DGD that is R=taterial to iR=ipleR=tentation of an (i)

R It.

h

- Removed by Amendment No. ###

esu 1n a c ange o.__ _____________

process, the fuel principal design requirements, 0f the nuolear design of the fuel or the reaotiio1ity oontrol systeR=I that is R=taterial to a fuel or reaeti>v<ity eontrol 15 Amendment No. ~

4.

5.

6.

7.

8.

9.

system design funetion, or the evaluation process in WCAP-12488, "Westinghouse Fuel Criteria Evaluation Process," or (ii)

Result in any change to the maximum fuel rod average burn-up limits; or the small break LOCA analysis information in UFSAR Subsections 15.6.5.4B.2.2 or 15.6.5.4B.2.3, Adversely affect the containment debris limits or debris screen design criteria, Change the Reactor Coolant Pump (RCP) type from a canned motor to a different type of RCP,

~ _ J -Removed by Amendment No. ### l suit In a e..

8

.tal l=leat e><Ghanger natural oiroulation test (first plant test), the Gore Mal<.eup Tani<. l=leated R:eeireulation Tests (first three plants test), or the l\\utomatio Depressuri;zation System Blowdown Test (first three plants test) that is material to the test objeoti1;es or test performanoe oriteria, ln'Wol1;0 struotul ~. ~:,~~.~:_d o~~.~~~~~~.e~~o~s~ ;,1;;:,~.Jds, ineluding design eodes and analytieal assumptions, that de1;iate from those oredited in the plant speoifio DGD for

=::l ~:~.~;:~:.~:R

~;::d.::~~~~~l=.1 internal trusses, tie bars, or headed studs of the steel oonorete (SC) module walls in the Nuolear Island or the Shield Building, ineluding SC to reinforeed eonerete (RC) 2'.j.~::~:v!:: :~:::~~.:~~~o;D: L.

of a eritieal seetion of the strueture. S~JG shall determine the me ratio under this oondition for cash oritioal seotion struetural member ineluding, but not limited to, wall segments, wall seotions, oonorete panels, slabs, or basemat seetions, affeeted by a departure by:

fij Using the Tier 2* information in the UF"SAR Scotian 3.8 or l\\ppendi>< 31=1 table that direotly states the me ratio or states the area of steel pro1;ided and the area of steel required for the affeeted struetural member, or Pro1;iding the same total area of steel aoross the entire eritieal seetion using any eombination of rebar si;zes and spaoing allowed by the design basis eodes used in the UF"SAR as the total area of steel speoified in UP:SAR: Scotian 3.8 and l\\ppendi>< 31=1 tables marl<.ed Tier 2*;

15a Amendment No. 142

No change on this page.

Provided for context.

(b)

For a departure from Tier 2* information that does not require prior NRC approval under the exemption in License Condition 2.D.(13)(a), SNC may take the departure provided that SNC complies with the requirements for Tier 2 departures in 10 CFR Part 52, Appendix D, Paragraph VIII.B.5, as modified by the exemption in License Condition 2.D.(13)(a). For each departure authorized by this License Condition:

1.

The departure or change to Tier 2* information shall remain Tier 2* information in the plant-specific DCD.

2.

SNC shall prepare and maintain a written evaluation that provides the bases for its determinations regarding the criteria in License Condition 2.D.(13)(a). In the report that 10 CFR Part 52, Appendix D, Section X.B.1 requires SNC to submit, SNC shall include a brief description of each departure and a summary of the evaluation of the departure.

E.

The licensees shall have and maintain financial protection of such type and in such amounts as the Commission shall require in accordance with Section 170 of the Atomic Energy Act of 1954, as amended, to cover public liability claims.

F.

Exemptions (1)

The following exemption from any part of the referenced design certification rule meets the requirements of 10 CFR 52.7 and Section VIII.A.4, VIII.B.4, or VIII.C.4 of Appendix D to 10 CFR Part 52, is authorized by law, will not present an undue risk to the public health or safety, and is consistent with the common defense and security. Special circumstances are present in that the application of the regulation in this particular circumstance is not necessary to achieve the underlying purpose of the rule (10 CFR 50.12(a)(2)(ii)) as described in the application and the staff SER dated August 5, 2011.

(a)

The licensees are exempt from the requirement of 10 CFR Part 52, Appendix D, Section IV.A.2.a to include a plant-specific DCD containing the same type of information and using the same organization and numbering as the generic DCD for the AP1000 15b Amendment No. 142

No change on this page.

Provided for context.

certified design. This exemption is specific to the organization and numbering scheme in the FSAR and is related to departure numberVEGP DEP 1-1.

(2)

The following exemptions from regulations were granted in the rulemaking for the design certification rule that is referenced in the application. In accordance with 10 CFR Part 52, Appendix D, Section V, Applicable Regulations, Subsection B, and pursuant to 10 CFR 52.63(a)(5), the licensees are exempt from portions of the following regulations:

(a)

Paragraph (f)(2)(iv) of 10 CFR 50.34-Plant Safety Parameter Display Console; (b)

Paragraph (c)(1) of 10 CFR 50.62-Auxiliary (or emergency) feedwater system; and (c)

Appendix A to 10 CFR Part 50, GDC 17-Second offsite power supply circuit.

(3)

For the reasons set forth below, the following specific exemptions, which are outside the scope of the design certification rule referenced in the application, are granted:

(a)

The licensees are exempt from the requirements of 10 CFR 70.22(b), 10 CFR 70.32(c), 10 CFR 74.31, 10 CFR 74.41, and 10 CFR 74.51 because the licensees meet the requirements of 10 CFR 70.17 and 74.7 as follows: The exemption is authorized by law, will not present an undue risk to the public health or safety, and is consistent with the common defense and security. Additionally, special circumstances are present in that the application of the regulations in this particular circumstance is not necessary to achieve the underlying purpose of the rule (1 O CFR 50.12(a)(2)(ii)) as described in the COL Application and the staff SER dated August 5, 2011.

(b)

The licensees are exempt from the requirements of 10 CFR 52.93(a)(1) as it relates to the exemption granted in Section 2.F.(1 )(a) of this license because the exemption meets the requirements of 10 CFR 52.7, because the exemption is authorized by law, will not present an undue risk to the public health or safety, and is consistent with the common defense and security. Additionally, special circumstances are present in that the application of the regulation in this particular circumstance is not necessary to achieve the underlying purpose of the rule (10 CFR 50.12(a)(2)(ii)) as described in the staff SER dated August 5, 2011.

16 Amendment No. 125

G.

Variances H.

I.

Having applied the technically relevant criteria applicable to the application for the Early Site Permit No. ESP-004, to the variances requested in the application, as described in NUREG-1923, the staff SER dated July 2009, the following variances from the early site permit (ESP) are granted:

(1)

A variance (VEGP VAR 1.6-1) from Section 1.6 of the VEGP ESP site safety analysis report (SSAR) as it references Revision 15 of the AP1000 DCD instead of Revision 19 of the AP1000 DCD, which is incorporated by reference in the FSAR; (2)

The variance (VEGP VAR 1.6-2) from Section 3.8.5, Foundations, of the VEGP ESP SSAR, which references Revision 15 of the AP1000 DCD, to reference Revision 19 of the AP1000 DCD, which is incorporated by reference in the FSAR; (3)

The variance (VEGP VAR 1.6-3) from Chapter 15, Accident Analysis, of the VEGP ESP SSAR which references Revision 15 of the AP1000 DCD, to reference Revision 19 of the AP1000 DCD, which is incorporated by reference in the FSAR; (4)

The variance (VEGP VAR 1.2-1) from the site layout information in Figures 1-4, 1-5, 13.3-2, and Part 5 Figure ii, of the VEGP ESP SSAR, which is superseded by the corresponding information in FSAR Section 1.1, Figure 1.1-202; (5)

The variance (VEGP VAR 2.2-1) from the information related to onsite chemical hazards in Section 2.2.3.2.3 and Table 2.2-6 of the VEGP ESP SSAR, which is superseded by the corresponding information contained in FSAR Sections 2.2 and 6.4; and (6)

The variance (VEGP VAR 2.3-1) from the information related to design-basis temperature characteristics in Section 2.3.1.5 and Table 1-1 of the VEGP ESP SSAR, which is superseded by the corresponding information contained in FSAR Section 2.3.1.5 and Table 2.0-201, which conforms to AP1000 DCD Revision 19.

~.~r+:i2kJ~1~~u~r:~~~:~~~SN~~:a~~ara~raph (cH1) ef 10 ci;:R 52.QQ until the Gemmissien makes the finEling unEler 1 Q ci;:R 52.10~(1),

SNG shall netify the NRG, in a timely manner, ef new infermatien that materially alters the bases fer Eletermining that either inspectiens, tests, er analyses were perfermeEI as reeiuireEI, er that acceptance criteria are met. The netificatien must centain sufficient infermatien te Elemenstrate that, netwithstanEling the nev, infermatien, the prescribes inspectiens, tests, er analyses ha11e been perfermeEI as reeiuireEI, anEI the prescribes acceptance criteria are met.

~

g"idaaoe aad slralegies de~ leped iA aGGer<laaoe wilh Removed by Amendment No. ###

17

!Amendment No. ###

J.

This license is effective as of February 10, 2012, and shall expire at midnight on Uie da-t:e 40 yeaFS tFom the da-t:e tl=la-f: tl=le Commission finds tl=la-t: the aeeeptanee cFitOFia in the combined license aFe met in accoFdance with 10 GFR 52.1 O@(g).

-I A-u-g-us_t_3_, 2_0_6_2 __ -~)

FOR THE NUCLEAR REGULATORY COMMISSION Appendices:

Michael R. Johnson, Director Office of New Reactors Appendix A - Technical Specifications Appendix B - Environmental Protection Plan Appendix C-Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) 18

!Amendment No. ###

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 COMBINED LICENSE No change on this page.

Provided for context.

VOGTLE ELECTRIC GENERATING PLANT UNIT 4 SOUTHERN NUCLEAR OPERATING COMPANY, INC.

GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MEAG POWER SPVM, LLC MEAG POWER SPVJ, LLC MEAG POWER SPVP, LLC CITY OF DAL TON, GEORGIA Docket No.52-026 License No. NPF-92

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for a combined license (COL) for Vogtle Electric Generating Plant (VEGP) Unit 4 filed by Southern Nuclear Operating Company, Inc. (SNC) acting on behalf of Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 1 and the City of Dalton, Georgia, an incorporated municipality in the state of Georgia acting by and through its Board of Water, Light and Sinking Fund Commissioners (City of Dalton), herein referred to as "the VEGP owners," which incorporates by reference Appendix D to 10 CFR Part 52 and Early Site Permit No. ESP-004, complies with the applicable standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission regulations set forth in 10 CFR Chapter I, and all required notifications to other agencies or bodies have been duly made; B.

There is reasonable assurance that the facility will be constructed and will operate in conformity with the application, as amended, the provisions of the Act, and the Commission regulations set forth in 10 CFR Chapter I, except as exempted from compliance in Sections 2.F and 2.G below; C.

There is reasonable assurance (i) that the activities authorized by this COL can be conducted without endangering the health and safety of the public and (ii) that such activities will be conducted in compliance with the Commission regulations set forth in 10 CFR Chapter I, except as exempted from compliance in Sections 2.F and 2.G below; 1 On June 24, 2015, Municipal Electric Authority of Georgia transferred its ownership interest to its wholly owned subsidiaries:

MEAG Power SPVM, LLC; MEAG Power SPVJ, LLC; and MEAG Power SPVP, LLC as described in the SNC letter dated December 2, 2013 and in the Commission's Safety Evaluation available in the Agencywide Document Access and Management System (ADAMS) under Accession No. ML14072A340.

1 Amendment No. 36

No change on this page.

Provided for context.

D.

SNC2 is technically qualified to engage in the activities authorized by this license in accordance with the Commission regulations set forth in 10 CFR Chapter I.

SNC and the VEGP owners together are financially qualified to engage in the activities authorized by this COL in accordance with the Commission regulations set forth in 1 O CFR Chapter I; E.

SNC and the VEGP owners have satisfied the applicable provisions of 10 CFR Part 140, "Financial Protection Requirements and Indemnity Agreements;"

F.

The issuance of this license will not be inimical to the common defense and security or to the health and safety of the public; G.

After weighing the environmental, economic, technical, and other benefits of the facility against environmental and other costs and considering reasonable available alternatives, the issuance of this license subject to the conditions for protection of the environment set forth herein is in accordance with Subpart A of 10 CFR Part 51 and all applicable requirements have been satisfied; and H.

The receipt, possession, and use of source, byproduct, and special nuclear material as authorized by this license will be in accordance with the applicable regulations in 10 CFR Parts 30, 40, and 70.

2.

On the basis of the foregoing findings regarding this facility, COL No. NPF-92 is hereby issued to SNC, Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and the City of Dalton, Georgia (the licensees) to read as follows:

A.

This license applies to the VEGP Unit 4, a light-water nuclear reactor and associated equipment (the facility), owned by the VEGP Owners. The facility would be located adjacent to existing VEGP Units 1 and 2 on a 3, 169-acre coastal plain bluff on the southwest side of the Savannah River in eastern Burke County, GA, approximately 15 miles east-northeast of Waynesboro, GA, and 26 miles southeast of Augusta, GA, and is described in the licensees' updated final safety analysis report (UFSAR), as supplemented and amended.

B.

Subject to the conditions and requirements incorporated herein, the Commission hereby licenses:

(1)

SNC pursuant to Sections 103 and 185b. of the Act and 10 CFR Part 52, to construct, possess, use, and operate the facility at the designated location in accordance with the procedures and limitations set forth in this license; (2)

The VEGP owners pursuant to the Act and 10 CFR Part 52, to possess but not operate the facility at the designated location in Burke County, GA, in accordance with the procedures and limitations set forth in this license; 2 SNC is authorized by the VEGP owners to exercise responsibility and control over the physical construction, operation, and maintenance of the facility.

2 Amendment No. 124

No change on this page.

Provided for context.

(3)

(a)

SNC pursuant to the Act and 10 CFR Part 70, to receive and possess at any time, special nuclear material as reactor fuel, in accordance with the limitations for storage and in amounts necessary for reactor operation, described in the UFSAR, as supplemented and amended; (b)

SNC pursuant to the Act and 10 CFR Part 70, to use special nuclear material as reactor fuel, after a Commission finding under 10 CFR 52.103(g) has been made, in accordance with the limitations for storage and in amounts necessary for reactor operation, described in the UFSAR, as supplemented and amended; (4)

(a)

SNC pursuant to the Act and 10 CFR Parts 30 and 70, to receive, possess, and use, at any time before a Commission finding under 10 CFR 52.103(g), such byproduct and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts, as necessary; (b)

SNC pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use, after a Commission finding under 10 CFR 52.103(g), any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as necessary; (5)

(a)

SNC pursuant to the Act and 10 CFR Parts 30 and 70, to receive, possess, and use, before a Commission finding under 10 CFR 52.103(g),

in amounts not exceeding those specified in 10 CFR 30.72, any byproduct or special nuclear material that is (1) in unsealed form; (2) on foils or plated surfaces, or (3) sealed in glass, for sample analysis or instrument calibration or other activity associated with radioactive apparatus or components; (b)

SNC pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use, after a Commission finding under 10 CFR 52.103(g), in amounts as necessary, any byproduct, source, or special nuclear material without restriction as to chemical or physical form, for sample analysis or instrument calibration or other activity associated with radioactive apparatus or components but not uranium hexafluoride; and (6)

SNC pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C.

The license is subject to, and the licensees shall comply with, all applicable provisions of the Act and the rules, regulations, and orders of the Commission, including the conditions set forth in 10 CFR Chapter I, now or hereafter in effect.

3 Amendment No. 124

D.

The license is subject to, and SNC shall comply with, the conditions specified and incorporated below:

(1)

(2)

Changes during Construction ~

- Removed by Amendment No. ###

g~JC n=iay reei1,1est 1,1se of a prelin=iinary aR1endn=ient reei1,1est (P,t\\R) proeess, for lieense an=iendR1ents, at any tin=ie before a CoR1n=iission finding 1,1nder 1 O CP:R a2.10d(g). To 1,1se the P,t\\R proeess, S~JC shall s1,1bn=iit a written reei1,1est to the Offiee of New Roasters (~JRO) in aooordanoe with COL IgG 02a, "Changes d1,1ring Constr1,1otion 1,1nder Part a2."

Before ~JRO's iss1,1anoe of a written P,t\\R notifioation, g~JC shall s1,1bn=iit the lieense aR1endn=ient reei1,1est (LAR). Thereafter, ~JRO will iss1,10 a written PAR notifieation, setting forth whether S~JC n=iay proeeed in aeeordanee with the Pl',R, LAR, and COL ISG 026. If S~JC eleets to proeeed and the Ll',R is s1,1bseei1,1ently denied, S~JC shall ret1,1rn the faeility to its e1,1rrent lioensing basis.

Pre-operational Testing --=~-----11 -Removed by Amendment Nos. 194 1

_and###

t91 S~JC shall perforn=i the design speeifie pre operational tests identified below:

S~JC shall re>v'iew and eval1,1ate the res1,1lts of the tests identified in geetion 2.D.(2)(a) of this lieense and eonfirn=i that these test res1,1lts are within the range of aeeeptable val1,1es predieted or otherwise oonfirR1 that the tested systen=is perforn=i their speoified f1,1netions in aeeordanee with UFSl',R Seetion 14.2.9, Amendment No. in each location including LC 2.D.(8) and at the bottom of each changed page to be inserted when determined.

4 Amendment No. 4eQ

(3)

(4)

S~JG shall notify the Direetor of ~JRO, or the Direetor's designee, in writing, upon sueeessful eompletion of the design speeifie pre operational tests identified in Seetion 2.D.(2)(a) of this lieense; and (Removed by Amendment ~Jo. 194)

Nuclear Fuel Loading and Pre-critical Testing

- Removed by Amendment Nos. 194 and ###

f9j t61 fe1 Until the submission of the notifieation required by Seetion 2.D.(2)(e) of this lieense, S~JG shall not load fuel into the reaetor vessel; (Removed by Amendment ~Jo. 194)

S~JG shall perform the pre eritieal tests identified in UP:SAR Seetion 14.2.10.1; S~JG shall review and evaluate the results of the tests identified in Seetion 2.D.(a)(e) of this lieense and eonfirm that these test results are within the range of aeeeptable values predieted or otherwise eonfirm that the tested systems perform their speeified funetions in aeeordanee with UP:SAR Seetion 14.2.10; and S~JG shall notify the Direetor of ~JRO, or the Direetor's designee, pre eritieal tests Initial Criticality and Low-Power Testing

- Removed by Amendment No.###

f9j Upon submission of the notifieation required by Seetion 2.D.(a)(e) of this lieense, S~JG is authoriced to operate the faeility at reaetor steady state eore power levels not to mmeed 5 pereent thermal power in aeeordanee with the eonditions speeified herein; S~JG shall perform the initial eritieality and low power tests identified in UP:SAR Seetions 14.2.10.2 and 14.2.10.a, respeetively; 5

Amendment No. 4-94

(5)

(6)

S~JG sl:iall Feview and evaluate tl:ie Fesults of tl:ie tests identified in Seetion 2.D.(4)(b) of tl:iis lieense and eonfiFFA tl:iat tl:iese test rnsults Fe witl:iin tl:ie mnge of aooeptable 1.ialues prndioted OF otl:ierwise oonfiFFA tl:iat tl:ie tested systeFAs peFfoFFA tl:ieiF speoified funetions in aeeoFdanee witR UFS/\\R Seetions 14.2.10.2 and 14.2.10.d; and S~JG sl:iall notify tl:ie DiFeotoF of ~JRO, OF tl:ie DiFeotoF's designee, in WFiting, upon suooessful GOFApletion of initial GFitioality and low poweF tests identified in Seetion 2.D.(4)(b) of tl:iis lieense, ineluding tl:ie design speeifie tests identified tl:iernin.

Power Ascension Testing ~<----11-Removed by Amendment No. ###

,{at Upon subR1ission of tl:ie notifioation Fequirnd by Seotion 2.D.(4)(d) of tl:iis lieense, SNG is autl:ioFii!:ed to opeFate tl:ie faeility at FeaetoF steady state eoFe poweF levels not to mmeed 100 pernent tl:ieFFA I poweF in aoooFdanoe witR tl:ie oonditions speoified ReFein, but only foF tl:ie puFpose of peFfoFFAing poweF asoension testing; f91 SNG sl:iall peFfoFFA tl:ie poweF asoension tests identified in UP:Sl\\R Seotion 14.2.10.4; f6j S~JG sl:iall Feview and evaluate tl:ie Fesults of tl:ie tests identified in Seotion 2.D.(a)(b) of tl:iis lioense and oonfiFFA tl:iat tl:iese test Fesults am witl:iin tl:ie mnge of aeeeptable values pFedieted OF otl:ierwise oonfiFFA tl:iat tl:ie tested systeFAs peFfoFR1 tl:ieiF speoified funetions in aeeoFdanee witR UFSI\\R Seetion 14.2.10.4; and fel-1 S~JG sl:iall notify tl:ie DiFeotoF of ~JRO, OF tl:ie DiFeotoF's designee, in wFiting, upon suooessful ooFApletion of poweF asoension tests identified in Seetion 2.D.(e)(b) of tl:iis lieense, ineluding tl:ie design speeifie tests identified tl:ieFein.

Maximum Power Level Upon subFAission of tl:ie notifioation rnquiFed by Seotion 2.D.(a)(d) of tl:iis lieense, SNC is authorized to operate the facility at steady state reactor core power levels not to exceed 3400 MW thermal (100-percent thermal power), as described in the UFSAR, in accordance with the conditions specified herein.

6 Amendment No. 4e7

(7)

(8)

(9)

Reporting Requirements ~,----11-Removed by Amendment No. ###

fat Within 30 days of a ehange to the initial test program deseribed in UrSAR Seetion 14, Initial Test Program, made in aeeordanee with 1 O GrR 50.59 or in aeeordanee with 1 O GrR Part 52, Appendi>< D, Seetion VIII, "Proeesses for Changes and Departures," S~JG shall report the ehange to the Direetor of ~JRO, or the Direetor's designee, in aeeordanee with 1 O GrR 50.59(d).

f91 S~JG shall report any violation of a requirement in Seetion 2.0.(3),

Seetion 2.0.(4), Seetion 2.0.(5), and Seetion 2.0.(6) of this lieense within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Initial notifieation shall be made to the NRG Operations Genter in aeeordanee v.iith 1 O GrR 50.72, with written follow up in aeeordanee with 1 O GrR 50.73.

Incorporation The Technical Specifications and Environmental Protection Plan in Appendices A and B, respectively, of this license, as revised through Amendment No. 494, are hereby incorporated into this license.

Technical Specifications ~<--..... I -Removed by Amendment No. ###

The teehnieal speeifieations in Append be A to this lieense beeome effeetive upon a Commission finding that the aeeeptanee eriteria in this lieense (ITMG) are met in aeeordanee with 1 O GrR 52.103(g) with the following mceeptions:

(a) Prior to initial eritieality of the reaetor eore while operating in plant operational Mode 5 (Gold Shutdown) or Mode 6 (Refueling) the following TS are temporarily mceluded from besoming effeetive:

TS 3.3.8, "E:ngineered Safety Feature Aetuation System (E:SFAS)

Instrumentation," Table 3.3.8 1

,Q, runetion 14, RCS Wide Range Pressure Low e

Funetion 15, Gore Malceup Tani< (GMT) Level Low 3 e

Funetion 16, GMT Level Lo*v\\' 6

,Q, runetion 18, IRWST Lower ~Jarrow Range Level Low 3 TS 3.3.9, "engineered Safety P:eature Aetuation System (eSrAS)

Manual Initiation," Table 3.3.9 1

,Q, runetion 1, Safeguards Aetuation Manual Initiation

,Q, runetion 6, ADS Stages 1, 2 & 3 Aetuation Manual Initiation e

Funetion 7, ADS Stage 4 Aetuation Manual Initiation e

Funetion 8, Passive Containment Cooling Aetuation Manual Initiation e

Funetion 9, Passive Residual Heat Removal Heat E:><ehanger Aetuation - Manual Initiation 7

Amendment No. 494

e runotion 12, In Containn=ient Reiueling Water £torage Tani(

(IRWST) lnjeetion Line \\lalve l'cetuation Manual Initiation e

runotion 1 d, IRW£T Containn=ient Reoiroulation \\£al1t10 l\\otuation Manual Initiation TS d.d.10, "E:ngineered Saiety Feature l'cetuation Systen=i (E:SFl'cS)

Roaster Coolant £ysten=i (RC£) l=lot Leg Le1t1el lnstrun=ientation" T£ d.d.14, "engineered £aiety reature l\\otuation £ysten=i (e£r,I\\£)

In oontainn=ient Reiueling Water £torage Tani( (IRW£T) and £pent Fuel Pool Level lnstrun=ientation," Table d.d.14 1 e

Funetion 1, Spent Fuel Pool Level Low 2 T£ d.d.1 Q, "Di1t1erse Aotuation £ysten=i (D,1\\£) Manual Controls,"

Table d.d.1 Q 1 e

Funetion 2, Passive Residual l=leat Ren=ioval l=leat E:)cel:langer (PRl=IR l=IX) oontrol and In Containn=ient Reiueling Water £torage Tani( (IRW£T) gutter oontrol 1t1al1t1es e

Fu notion 4, 1'1utoR1atie Depressuri2:ation Systen=i (l'cDS) stage 1

'11al>o<<es e

runotion a,,1\\0£ stage 2 1t1al1t1es e

runotion e,,1\\0£ stage d 1t1al1t1es e

Funetion 7, l'cDS stage 4 valves e

runotion 8, IRW£T injeotion squib 1t1al1t1es e

runotion Q, ContainR1ent reoiroulation 1t1al1t1es e

Fu notion 10, Passive eontainR1ent eooling drain valves e

Funetion 11, Seleeted eontainn=ient isolation valves T£ d.d.20, ",l\\uton=iatio Depressuri2:ation £ysten=i (,1\\0£) and In eontainn=ient Reiueling Water Storage Tani( (IRWST) lnjeetion Bloel(ing Deviee," Table d.d.20 1 e

runotion 2,,1\\0£ and IRW£T lnjeotion Blool( £witol:les ior Manual UnbloGl(ing TS d. 4.12, "l'cuton=iatie Depressuri2:ation Systen=i (l'cDS)

Sl:lutdown, RCS lntaef' T£ d. 4.1 d, ",l\\uton=iatio Depressuri2:ation £ysten=i (,1\\0£)

£1:lutdown, RC£ Open" TS d.e.e, "Passive Residual l=leat Ren=ioval l=leat E:)cel:langer (PRl=IR l=IX)

Sl:lutdown, Reaetor Coolant Systen=i (RCS) lntaef' T£ d.a.7, "In oontainn=ient Reiueling Water £torage Tani( (IRW£T)

£1:lutdown, MODe a" TS d.e.8, "In eontainn=ient Reiueling Water Storage Tani( (IRWST)

Sl:lutdown, MODE: e" T£ d.e.7, "Containn=ient Penetrations" T£ d.7.1 d, "£pent ruel Pool Cooling £ysteR1 (£r£) Containn=ient Isolation Valves"

?a Amendment No. 48-9

(b) Prior to initial sritisality of the roaster sore while operating in plant operational Mode 4 (£afe £hc1tdown) when any sold leg temperatc1re is

< 270°P: the following T£ are temporarily mmlc1ded from besoming effesth,ie:

TS a.a.8, "E:ngineered Safety Featc1re l',etc1ation Systen=i (E:Srl'.S) lnstrc1mentation," Table a.a.8 1 e

P:c1nstion 14, RC£ Wide Range Pressc1re Low e

P:c1nstion 1 a, Gore Mal'iec1p Tank (GMT) Le1;el Low a e

Fc1netion 16, GMT Le>v<el Low 6 e

Fc1netion 18, IRWST Lower ~Jarrow Range Le>v<el Low a e

P:c1nstion 1 Q, Roaster Coolant Pc1mp Bearing Water Temperatc1re l=ligh 2 TS a.a.9, "E:ngineered Safety Featc1re l',etc1ation Systen=i (E:Srl'.S)

Manc1al Initiation," Table a.a.9 1 e

P:c1nstion a, Gontainn=ient Isolation Manc1al Initiation e

Fc1netion 6, l',DS Stages 1, 2 & a l',etc1ation Manc1al Initiation e

Fc1netion 7, l',DS Stage 4 l',etc1ation Manc1al Initiation e

P:c1nstion 8, Passi1; 0 Containment Cooling l\\stc1ation Manc1al Initiation e

Fc1netion 12, In Containment Refc1eling Water Storage Tani'<

(IRW£T) lnjestion Line Val1;0 l\\stc1ation Manc1al Initiation e

Fc1netion 1 a, IRWST Containment Reeirec1lation Val>v<e Aetc1ation Manc1al Initiation T£ a.a.1 a, "engineered £afety P:eatc1re l\\stc1ation £ystem (e£P:,I\\£)

Main Control Roon=i Isolation, hr £c1pply Initiation, and eleotrisal Load De energi:zation," Table a.a.1 a 1 e

Fc1netion 1, Main Control Room Air Sc1pply Iodine or Partiec1late Radiation l=ligh 2 T£ a.3.1 Q, "Di1;erse Astc1ation £ystem (D,1\\£) Manc1al Controls,"

Table a.a.19 1 e

Fc1netion 4, l',c1ton=iatie Depressc1ri:zation System (l',DS) stage 1 1o'al>t,os e

Fc1netion e, l',DS stage 2 >v<al>v<es e

Fc1netion 6, l',DS stage a >v<al>v<es e

P:c1nstion 7,,1\\0£ stage 4 1Jal1JOS e

P:c1nstion 8, IRW£T injestion sqc1ib 1Ja l1JOS e

Fc1netion 9, Gontainn=ient reeirec1lation >v<al>v<es e

Fc1netion 10, Passi>v<e eontainn=ient eooling drain >v<al>v<es e

P:c1nstion 11, £elested sontainment isolation 1;al1;es 7b Amendment No. 48-9

T£ 3.3.20, "l\\c1toR=1atio DepFessc1Fication £ysteR=1 (l\\D£) and In eontainR=lent Refc1eling WateF StoFage Tani( (IRWST) IRjeetion Bloel(ing Deviee," Table 3.3.20 1 e

rc1notion 2, l\\D£ and IRW£T lnjeotion Blool( £witohes foF Manc1al Unbloel(ing TS 3. 4.11, "l',c1toR=1atie DepFessc1Fication SysteR=I (l',DS)

Ope Fating" T£ 3.a.1, "l\\GGblR=lbllatoFs" T£ 3.a.e, "In oontainR=lent Refc1eling WateF £torage Tani( (IRW£T)

OpOFating" TS 3.e.1, "GontainR=lent" T£ 3.e.2, "GontainR=lent l\\iF Lool(s" T£ 3.e.3, "GontainR=lent Isolation Val1t'es" T£ 3.e.e, "Passi1w'e GontainR=lent Cooling £ysteR=1 (PG£)"

TS 3.e.8, "pl=l l',djblStR=lent" T£ 3.7.4, "£eoondaFy £peoifio l\\oti1w'ity" T£ 3.7.10, "£teaR=1 GeneratoF (£G) Isolation Val1t'es" only feF PORV and PORV blool( 1t'al1t'es (£G blowdown isolation 1t'al1t'e not mmlc1ded)

(e) Prior to initial eritieality of the reaetor eore while operating in plant opeFational Mode 4 (£afe £hc1tdown) with all fec1F sold leg teR=1peratc1Fes

> 270°r the following T£ aFe teR=lpOraFily mmlc1ded fFOR=I beGOR=ling effeGti>t'e:

TS 3.3.8, "E:ngineered Safety F"eatc1re l',etc1ation SysteR=I (E:SF"l',S) lnstFc1R=1entation," Table 3.3.8 1 e

rc1notion 3, GontainR=lent Radioaoti>t'ity l=ligh e

rc1notion 18, IRW£T LoweF ~JaFFow Range Le1t'el Low 3 e

F"c1netion 19, Reaetor Coolant Pc1R=1p Bearing Water TeR=1peratc1re l=ligh 2 T£ 3.3.Q, "engineeFed £afety reatc1Fe l\\otc1ation £ysteR=1 (e£r,I\\£)

Manc1al Initiation," Table 3.3.9 1 e

F"c1netion 3, GontainR=lent Isolation Manc1al Initiation e

rc1notion e,,I\\D£ £tages 1, 2 & 3,l\\otc1ation Manc1al Initiation e

rc1notion 7,,I\\D£ £tage 4,l\\otc1ation Manc1al Initiation e

F"c1netion 8, Passive GontainR=lent Cooling l',etc1ation Manc1al Initiation e

rc1notion 13, IRW£T GontainR=lent ReoiFoc1lation Val1t'e,l\\otc1ation Manc1al Initiation e

F"c1netion 14, SG Power Operated Relief 'Jalve and Bloel( Valve Isolation Manc1al Initiation 7c Amendment No. 48-9

T£ 3.3.13, "engineOFed £a~ety P:eatc1re l\\otc1ation £ystem (e£P:,I\\£)

Main Control Roon=i Isolation, l',ir Sc1pply Initiation, and E:leetrieal Load De energi:zation," Table 3.3.13 1 e

P:c1notion 1, Main Control Room Air £c1pply Iodine or Partioc1late Radiation High 2 TS 3.3.19, "Diverse Aetc1ation System (01',S) Manc1al Controls,"

Table 3.3.1 Q 1 e

P:c1notion 4,,l\\c1ton=iatio Depressc1ri:zation £ystem (,1\\0£) stage 1

'11al>o<<es e

Fc1netion e, l',DS stage 2 valves e

P:c1notion e,,1\\0£ stage 3 *;al>o*es e

P:c1notion 7,,1\\0£ stage 4 *;al*;es e

Fc1netion 9, Containn=ient reeirec1lation valves e

P:c1notion 10, Passi*;e oontainn=ient oooling drain *;al*;es e

P:c1notion 11, £eleoted oontainment isolation *;al*;es TS 3. 4.11, "l',c1tomatie Depressc1ri:zation System (l',DS)

Operating" TS 3.6.1, "Aeec1mc1lators" T£ 3.a.e, "In oontainment Re~c1eling Water £torage Tanl'i (IRW£T)

Operating" only for oontainn=ient reoiroc1lation ~low paths (iRjeotion

~low paths not mcelc1ded)

TS 3.e.1, "Containment" T£ 3.e.2, "Containn=ient,l\\ir Lool'is" T£ 3.e.3, "Containment Isolation \\Jal*;es" TS 3.e.e, "Passive Containment Cooling System (PCS)"

TS 3.e.8, "pl=l P,djc1stn=ient" TS 3.7.4, "Seeondary Speei~ie l',etivity" T£ 3.7.10, "£team Generator (£G) Isolation \\Jal*;es" only for POR\\J and POR\\J blool'i *;al*;es (£G blowdown isolation *;al*;e not mcolc1ded) 7d Amendment No. ~

(10)

Operational Program Implementation ~<---ii -Removed by Amendment No. ### I S~JG st:iall i1T1ple1T1ent tt:ie progra1T1s or portions of prograFRs identified below, on or before tt:ie date S~JG aet:iieves tt:ie following IT'lilestones:

E:nvironFRental Qualifieation Prog rIT'I iFRplelTlented before initial fuel load; Reaetor >Jessel Material Surveillanee Prog rIT'I iFRplelTlented before initial oritioality; PreS0Pw'iG0 Testing PrograFR ilTlplelTlented before initial fuel load* '

GontainlTlent Leal<.age Rate Testing P rogrIT'I i1T1ple1T1ented before initial fuel load; P:ire Proteotion P rogrIT'I Tt:ie fire proteetion FReasures in aeeordanee witt:i Regulatory Guide (RG) 1.189 for designated storage building areas (ineluding adjaeent fire areas tt:iat eould affeot tt:ie storage area) i1T1ple1T1ented before initial reoeipt

?e Amendment No. 48-9

ffi1 t+1 fB of b;'produot or speoial nuolear materials that are not fuel (mmluding mmmpt quantities as deseribed in 10 CF"R ao.18);

~

The fire proteetion measures in aeeordanee with RG 1.189 for areas oontaining new fuel (inoluding adjaoent areas where a fire oould affeot the new fuel) implemented before reoeipt of fuel onsite; 3e-

/\\II fire proteotion program features implemented before initial fuel load; Standard Radiologieal effluent Controls implemented before initial fuel load; Offsite Dose Caloulation Manual implemented before initial fuel

~

Radiologieal E:nvironmental Monitoring Program implemented before initial fuel load; Proeess Control Program implemented before initial fuel load; Radiation Proteetion Program (RPP) (ineluding the l',LARA prineiple) or applieable portions as identified in UP:SA R Seetion 12.a thereof:

4-,.

RPP features applioable to reoeipt of b;' produot, souroe, or speeial nuelear materials (mceluding mcempt quantities as desoribed in 1 0 C P:R a0.18) implemented before initial reeeipt of sueh materials;

~

RPP features (ineluding the ALARA prineiple) applieable to new fuel implemented before reoeipt of initial fuel on site; 3e-

/\\II other RPP features (inoluding the ALARA prinoiple) mceept for those applieable to eontrol radioaetive waste shipment implemented before initial fuel load; 4,.

RPP features (inoluding the /\\LARA prinoiple) applioable to rad ioaetive waste shipment implemented before first shipment of radioaotiio10 waste; Roaster Operator Training Program implemented 18 months before the soheduled date of initial fuel load; Motor Operated Val1o10 Testing Program implemented before initial fuel load; 8

Amendment No. ~

fffi1 Initial Test ProgrR=I (ITP)

Preoperational Test ProgrR=I iR=lplen=iented before the first preoperational test; and Startup Test Progran=i in=iplen=iented before initial fuel load* '

~peeial ~Juelear Material Control and l'ceeounting Progran=i 1R=1plen=iented before initial reoeipt of speoial nuolear n=iaterial* and fe1

~peeial ~Juelear Material Ph;csieal Proteetion PrograR=I 1R=1pleR=1ented before initial reeeipt of speeial nuelear n=iaterial on s+te-: *

- Removed by Amendment (11)

Operational Program Implementation Schedule No. ###

~Jo la~er than 12 R=1onths after issuanoe of the COL, 8~JC shall subn=iit to

~he D1reetor of NRO, or the Direetor's designee, a sehedule for 1n=ipleR=1entation of the operational progran=is listed in UF8PcR Table 1 d.4 201, inoluding the assooiated estin=iated date for initial loading of fuel. The sohedule shall be updated m,iery e n=ionths until 12 n=ionths before_seheduled fuel loading, and every n=ionth thereafter until all the

?perat1onal prograR=1s listed in UP:8,1\\R Table 1 d.4 201 hm,ie been full;<<

1n=ipleR=1ented.

(12)

Site-and Unit-specific Conditions ~~---11 -Removed by Amendment No. ###

fa-1 8_~JC shall either ren=iove and replaee, or shall in=iprove, the soils d1r?otl~ abo1,e the blue bluff n=iarl for soils under or adjaoent to 8e1sn=i10 Category I struotures, to elin=iinate any liquefaotion potential.

f91 Before oon=1R=1enoing installation of indi'w'idual piping segR=1ents and oonneoted _oon=ipo~ents in their final looations, 8~JC shall oon=iplete the as des1gn?d_ pipe rupture ha:z:ards analysis for eon=ipartn=ients (ro~n=is) ?ontaInIng those segn=ients in aooordanoe with the oriteria

?utllned In t~e UP:8,1\\R 8eotions d.e.1.d.2 and d.e.2.a, and shall 1nforn=i the D1reet~r of ~JRO, or the Direetor's designee, in writing, upo_n the e~n=iplet1on of this analysis and the availability of the as designed pipe rupture ha:z:ards analysis reports.

t61

~efo~~ oo~n=ienoing installation of indi'w'idual piping segn=ients,

!dent1~1e~ In UF8~1R 8_eetion d.9.8.7, and eonneeted eon=iponents In the1~ final looat1ons In the faoilit;<<, 8~JC shall oon=iplete the

~nal;cs1s of t~e as designed individual piping segn=ients and shall 1nforn=i the D1reotor of ~JRO, or the Direotor's 9

Amendment No. ~

designee, in writing, upon U:ie eon:ipletion of tl:iese an l}'ses and tl:ie a1o1ailabilit}' of tl:ie design reports for tl:ie seleoted piping pael(ages.

(d) ~Jo later tl:ian 180 d J'S before initial fue l load, S~JC sl:iall submit to tl:ie Direetor of MRO, or tl:ie Direetor's designee, in writing, a full}' developed set of plant speoifio en:iergeno}' aotion le1o1els (e,t\\Ls) for Ve GP Unit 4 in aeeordanee witl:i tl:ie eriteria defined in Amendment ~Jo. 76. Tl:ie eALs sl:iall Rave been diseussed and agreed upon witR State and loeal offieials.

~Jo later tl:ian 180 d J'S before initial fuel load, S~JC sl:iall submit to tl:ie Direotor of ~JRO, or tl:ie Direetor's designee, in writing, an assessment of emergene}'

response staffing performed in aeeordanee witR MEI 10 05, "Assessn:ient of On Sl:iif.t En:iergene}' Response Organii!:ation Staffing and Capabilities,"

Re1o1ision 0.

fet S~JC st:iall not re1o1ise or n:iodif}' tt:ie pro1o1isions of Seotions a.d, a.4, a.6, a.9, and a.1 O of tl:ie Speoial Muolear Material (S~JM) PR}'Sioal Proteotion Progran:i until tl:ie requiren:ients of 1 0 CrR 7d.55 are implen:iented.

~ ~Jo later tl:ian 12 montl:is after issuanee of tl:ie COL, S~JC sl:iall submit to tl:ie Direotor of ~JRO, or tl:ie Direotor's designee, a sol:iedule for in:iplementation of tl:ie following lioense oonditions.

TAO SGR0dule SR II be updated 01o'0Fy 6 montRS until 12 n:iontRS before sol:ieduled fuel loading, and 01o1ery n:iontl:i tl:iereaf.ter until eael:i lieense eondition l:ias been full}' implemented. Tl:ie sel:iedule sl:iall identif}' tl:ie eompletion of or implementation of tl:ie following:

1. Tl:ie eonstruetion and inspeetion proeedures for steel eonerete eon:iposite (SC) eonstruetion aetivities for seisn:iie Category I nuelear island modules (inoluding sl:iield building SC modules) desoribed in UrS,t\\R Seetion d.8.4.8;

2. Tl:ie spent fuel raol( Metamio Coupon n:ionitoring program (before initial fuel load);

d. ln:iplementation of tl:ie flow aeeelerated eorrosion (Fl',C) program ineluding eonstruetion pl:iase aetivities (before initial fue l load);

4. A turbine maintenanoe and inspeotion program, wl:iioR must be oonsistent witR tl:ie maintenanoe and inspeotion program plan aoti1o1ities and inspeotion intervals identified in UrS,t\\R Seotion 10.2.d.6 (before initial fue l load);

5. Tl:ie availabilit}' of doeun:iented instrumentation uneertainties to ealeulate a power ealorimetrie uneertaint}'

(before initial fuel load);

6. Tl:ie a1o1ailabilit}' of adn:iinistrati1o10 oontrols to implen:ient maintenanee and eontingene}' aetivities related to tl:ie 10 Amendment No. ~

power oalorimetrio 1,moertainty instrc1mentation (before initial fc1el load);

7:

The site speeifie severe aeeident 1T1anage1T1ent gc1idelines (before startc1p testing);

g,.

The operational and progra1T1matio elements of the mitigati1,e strategies for respond ing to oiroc1mstanoes assooiated with loss of large areas of the plant dc1e to m~plosions or fire de1,eloped in aooordanoe with 1 O ci;;R a0.a4 (hh)(2) (before initial fc1el load); and The ITP prooedc1res identified in U i;;S/\\R Seotion 14.2.d:

a. administrative manc1al (before the first preoperational test)

b. preoperational testing (before sohedc1led performanee)

e. startc1p testing (before initial fc1el load) ffi1 Before initial fc1el load, S~JG shall:

4-:

Update the seis1T1io interaotion analysis in ui;;SAR Seotion d.7.e.d to refleet as bc1ilt information, whieh mc1st be based on as prooc1red data, as well as the as oonstrc1oted eondition; Reeoneile the seismie analyses deseribed in Seetion d.7.2 of the UFS/\\R, to aeeoc1nt for detailed design ehanges, inelc1ding, bc1t not lilTlited to, those dc1e to as prooc1red or as bc1ilt ohanges in oomponent mass, oenter of gra1, ity, and sc1pport oonfigc1ration based on as proec1red eqc1ipment infor1T1ation; Galoc1late the instrc1mentation c1noertainties of the aotc1al plant operating instrc11T1entation to oonfirlTl that either the design lilTlit departc1re from nc1eleate boiling ratio (D~JBR) valc1es re1T1ain valid or that the safety analysis minimc11T1 D~U~R boc1nds the new design limit D~JBR 1,alc1es plc1s D~JBR penalties; Update the pressc1re temperatc1re (P T) lilTlits c1sing the pressc1re te1T1peratc1re limits report (PTLR) methodologies approved in the UFSAR, c1sing the plant speeifie 1T1aterial properties or eonfirm that the reaetor vessel material properties meet the speoifioations of and c1se the Westinghoc1se generio PTLR oc1rves;

'Jerify that plant speeifie belt line material properties are oonsistent with the properties gi1,en in UFSAR Seotion e.d.d.1 and Tables e.d 1 and e.d d. The verifieation 1T1c1st inolc1de a pressc1riced thermal shool<. (PTS) e1,alc1ation based on as proec1red reaetor vessel material data and the prejeoted nec1tron flc1enoe for the 11 Amendment No. ~

plant design objeotii,ce_ SubA1it this PTS e1,caluation Feport to the DiFeotoF of NR:O, OF the DiFeotoF's designee, in wFiting, at least 18 A1onths befoFO initial fuel load; e-:-

Review differnnees between the as built plant and the design used as the basis foF the l\\P1000 seisR=liG AlaFgin analysis. S~JG shall perfoFR=I a veFifieation wall(down to identify diffeFenoes between the as built plant and the design. S~JG shall evaluate any diffeFenees and R=1ust A1odify the seisR=liG R=laFgin analysis as neoessary to aeeount foF the plant speeifie design and any design ohanges OF departuFes fFOR=I the oertified design. SNG shall GOR=lpaFe the as built stFUGtUFOS, systeAlS, and 00A1ponents (SSC) high oonfidenoe, low pFobability of failuFes (l=IGLPP:s) with those assuR=1ed in the l\\P1000 seisAliG AlaFgin e1,caluation, befoFO initial fuel load. SNG shall e1,caluate de1,ciations fFOR=I the l=IGLPP: 1,calues OF assuA1ptions in the seisR=liG R=laFgin e1,caluation due to the as built oonfiguFation and final analysis to deteFR=line if 1,culneFabilities ha1,ce been intFoduoed; Review differnnees between the as built plant and the design used as the basis foF the l\\P1000 pFObabilistio Fisl(

assessR=1ent (PR:A) and UP:SAR: Table 19.59 18. SNG shall e1,caluate the plant speoifio PR:A based insight differnnees and shall R=1odify the plant speeifie PR:A R=1odel as neoessary to aooount foF the plant speoifio design and any design ehanges OF departuFO fFOR=I the PR:A eertified in R:e1,c_ 1 Q of the AP1000 DGD; R:e1,ciew differnnoes between the as built plant and the design used as the basis foF the AP1000 internal fiFe and internal flood analysis. SNG shall e1,caluate the plant speeifie internal fire and internal flood analyses and shall R=1odify the analyses as neoessary to aooount foF the plant speeifie design and any design ehanges or departurns fFOR=I the design oertified in R:e1,c_ 1 Q of the l',P1000 DGD; and PerforAl a therR=lal lag assessR=1ent of the equipR=lent listed in UP:S,I\\R: Tables 1 QD g and 1 QD Q to pF01,cide additional assuranee that this equipR=lent ean perforR=I its severe aooident funotions duFing en1,ciFonR=1ental oonditions resulting froR=I hydrogen burns assoeiated with severe aooidents. S~JG shall perfoFR=I this assessA1ent foF equipAlent used for severe aeeident R=litigation that has not been tested at se1,ceFe aooident oonditions. S~JG shall assess the ability of the equipR=lent to perfoFR=I duFing aooident hydFOgen burns using the en1w'iFOnR=1ent en1,celoping A1ethod or the test based therA1al analysis 12 Amendment No. ~

method described in electric Power Research Institute (ePRI) ~JP 4354, "Large £ca le 1=4ydrogen Burn equipment e <periments. "

Implement a suri.ieillance program for explosi1.iely actuated 1;al1;es (squib 1;al1;es) that includes the following pro1;isions in addition to the requirements specified in the edition of the AS Me Goee for Operation ane,IJ,1-aintenanoe of N(;}ofear

,oov1,1er Plants (OM Code) as incorporated by reference in 10 ci;;R 50.55a.

a-:-

Preser;ice Testing

/\\II mcplosii;ely actuated 1;al1;es shall be preser;ice tested by 1;erifying the operational readiness of the actuation logic and associated electrical circuits for each mcplosii;ely actuated 1;al1;e with its pyrotechnic charge remoi;ed from the 1;al1Je. This must include confirmation that sufficient electrical parameters

('Joltage, current, resistance) are a1;ailable at the mcplosii;ely actuated 1;al1;e from each circuit that is relied upon to actuate the 1;al1Je. In addition, a sample of at least 20% of the pyrotechnic charges in all mcplosii;ely actuated 1;al1;es shall be tested in the 1;al1;e or a qualified test fi <ture to confirm the capability of each sampled pyrotechnic charge to proi;ide the necessary moti>;e force to operate the 1;al1;e to perform its intended function without damage to the 1;al1;e body or connected piping.

The sampling must select at least one 0><plosi1;ely actuated 1;al1;e from each redundant safety train.

Correcti1;e action shall be tal(;en to resol1;e any deficiencies identified in the operational readiness of the actuation logic or associated electrical circuits, or the capability of a pyrotechnic charge. If a charge fails to fire or its capability is not confirmed, all charges with the same batch number shall be remoi;ed, discarded, and replaced with charges from a different batch number that has demonstrated successful 20% sampling of the charges.

Operational Surveillance e

.ie action shall be taken to reso1,,e an,i deficiencies identified in the actuation logic or associated electrical circuits. P:or m~plosi1t1el,i actuated 1t1al1t1es selected for test sampling e1.ier,i 2,iears in accordance with the /\\£1\\4e 01\\4 Gode, the sampling 1T1ust seleet at least one mcplosively aotuated 1o<<al1o<<e fFOITI eaoh rndundant safety train. E:aeh sa!Tlpled pyFOteehnie ehar:ge shall be tested in the 1o<<al1o<<e OF a qualified test factuFO to eonfirlTI the eapability of the ohaFge to pF01o<<ide the neoessary 1T1oti1o<<e foree to operate the valve to perforR1 its intended funotion without daR1age to the 1o<<al1o<<e body OF oonneoted piping. GoFFeoti1o<<e aotion shall be tal(;en to FOSOl1o<<e any defioienoies identified in the oapability of a pyFOteehnie ehar:ge in aeeoFdanee with the pFeseFViGe testing rnquiFOITlents. This lioense oondition shall mcpiFe upon (1) inooFpoFation of the above su FYeillanee provisions for mcplosively aetuated 1o<<al1o<<es into the faoility's inseFViGe testing pFOgFalTI, OF (2) ineorporation of inseFYiee testing require!Tlents for 0>cplosi1o<<ely aotuated 1o<<al1o<<es in new FeaotoFs (i.e., plants reeeiving a eonstruetion per1T1it, or eo1T1bined lieense for oonstFuotion and opeFation, afteF January 1, 2000) to be speeified in a future edition of the ASME: OM Gode as inooFporated by Fefernnoe in 1 0 GP:R a0.aaa, inoluding any eonditions ilTlposed by the NRG, into the faeility's inseFYiee testing p Fog FR1. (13) Departures from Plant-specific DCD Tier 2* Information (a) SNC is exempt from the requirements of 10 CFR Part 52, Appendix D, Paragraphs VIII.B.6 and VIII.B.5.a for prior NRC approval of departures from Tier 2* information and departures from Tier 2 information involving a change to or departure from Tier 2* information; exc<-,1.LJ.....u..1.L..1.LwJ.L1.u..r..i.1.1..w.w..1J...u..u-----------.

1.

2.

3.

plant speeifie DGD for establishing the eriteria for the design OF oonstFUGtion of a stFUGtuFe, systelTI, OF GOITlponent (SSC) ~ - Removed by Amendment No. ### ~ a design proeess deseribed in the plant speoifio DGD that is 1T1at0Fial to i1T1ple1T1entation of an industry standard or endorsed regulatory guidanee, (i) Result in a change to the fuel criteria evaluation process, the fuel principal design requirements, ef the nuoleaF design of the fuel OF the Feaoti1o<<ity oontFOI systelTI that is 1T1aterial to a fuel or reaetivity eontrol 15 Amendment No. 441- system design funetion, or the evaluation process in WCAP-12488, "Westinghouse Fuel Criteria Evaluation Process," or (ii) Result in any change to the maximum fuel rod average burn-up limits; or the small break LOCA analysis information in UFSAR Subsections 15.6.5.4B.2.2 or 15.6.5.4B.2.3,

4.

Adversely affect the containment debris limits or debris screen design criteria,

5.

Change the Reactor Coolant Pump (RCP) type from a canned motor to - Removed by Amendment No. ###

6.

7.

8.

9.

l=leat e><Ghanger natural oiroulation test (first plant test), the Gore Mal<.eup Tani<. l=leated R:eeireulation Tests (first three plants test), or the l\\uton=iatio Depressuri;zation Systen=i Blowdown Test (first three plants test) that is n=iaterial to the ~

~=!d~r:~:e~!.~a~: ~.las, ineluding design eodes and analytieal assun=iptions, that de1o1iate fron=i those oredited in the plant speoifio DGD for

~ R d b A d t N ~ -to : : :~: ign ~f t: :~a; :~epl: ~es, internal trusses, tie bars, or headed studs of the steel oonorete (SC) n=iodule walls in the Nuolear Island or the Shield Building, ineluding SC to reinforeed eonerete (RC) ~ -Removed by Amendment No. ### 1 ~ ase in the demand to oapaoity (mG) rati of a eritieal seetion of the strueture. S~JG shall determine the me ratio under this oondition for cash oritioal seotion struetural men=iber ineluding, but not limited to, wall segn=ients, wall seotions, oonorete panels, slabs, or basen=iat seetions, a#eeted by a departure by: fij Using the Tier 2* information in the UF"SAR Scotian 3.8 or l\\ppendi>< 31=1 table that direotly states the me ratio or states the area of steel pro1o1ided and the area of steel required for the a#eeted struetural n=ien=iber, or Pro1o1iding the san=ie total area of steel aoross the entire eritieal seetion using any eon=ibination of rebar si;zes and spaoing allowed by the design basis eodes used in the UF"SAR as the total area of steel speoified in UP:SAR: Scotian 3.8 and l\\ppendi>< 31=1 tables n=iarl<.ed Tier 2*; 15a Amendment No. 141 No change on this page. Provided for context. (b) For a departure from Tier 2* information that does not require prior NRC approval under the exemption in License Condition 2.D.(13)(a), SNC may take the departure provided that SNC complies with the requirements for Tier 2 departures in 10 CFR Part 52, Appendix D, Paragraph VIII.B.5, as modified by the exemption in License Condition 2.D.(13)(a). For each departure authorized by this License Condition:

1.

The departure or change to Tier 2* information shall remain Tier 2* information in the plant-specific DCD.

2.

SNC shall prepare and maintain a written evaluation that provides the bases for its determinations regarding the criteria in License Condition 2.D.(13)(a). In the report that 10 CFR Part 52, Appendix D, Section X.B.1 requires SNC to submit, SNC shall include a brief description of each departure and a summary of the evaluation of the departure. E. The licensees shall have and maintain financial protection of such type and in such amounts as the Commission shall require in accordance with Section 170 of the Atomic Energy Act of 1954, as amended, to cover public liability claims. F. Exemptions (1) The following exemption from any part of the referenced design certification rule meets the requirements of 10 CFR 52.7 and Section VIII.A.4, VIII.B.4, or VIII.C.4 of Appendix D to 10 CFR Part 52, is authorized by law, will not present an undue risk to the public health or safety, and is consistent with the common defense and security. Special circumstances are present in that the application of the regulation in this particular circumstance is not necessary to achieve the underlying purpose of the rule (10 CFR 50.12(a)(2)(ii)) as described in the application and the staff SER dated August 5, 2011. (a) The licensees are exempt from the requirement of 10 CFR Part 52, Appendix D, Section IV.A.2.a to include a plant-specific DCD containing the same type of information and using the same organization and numbering as the generic DCD for the AP1000 certified design. This exemption is specific to the organization and numbering scheme in the FSAR and is related to departure numberVEGP DEP 1-1. 15b Amendment No. 141 No change on this page. Provided for context. (2) The following exemptions from regulations were granted in the rulemaking for the design certification rule that is referenced in the application. In accordance with 10 CFR Part 52, Appendix D, Section V, Applicable Regulations, Subsection B, and pursuant to 10 CFR 52.63(a)(5), the licensees are exempt from portions of the following regulations: (a) Paragraph (f)(2)(iv) of 10 CFR 50.34-Plant Safety Parameter Display Console; (b) Paragraph (c)(1) of 10 CFR 50.62-Auxiliary (or emergency) feedwater system; and (c) Appendix A to 10 CFR Part 50, GDC 17-Second offsite power supply circuit. (3) For the reasons set forth below, the following specific exemptions, which are outside the scope of the design certification rule referenced in the application, are granted: (a) The licensees are exempt from the requirements of 10 CFR 70.22(b), 10 CFR 70.32(c), 10 CFR 74.31, 10 CFR 74.41, and 10 CFR 74.51 because the licensees meet the requirements of 10 CFR 70.17 and 74.7 as follows: The exemption is authorized by law, will not present an undue risk to the public health or safety, and is consistent with the common defense and security. Additionally, special circumstances are present in that the application of the regulations in this particular circumstance is not necessary to achieve the underlying purpose of the rule (10 CFR 50.12(a)(2)(ii)) as described in the COL Application and the staff SER dated August 5, 2011. (b) The licensees are exempt from the requirements of 10 CFR 52.93(a)(1) as it relates to the exemption granted in G. Variances Section 2.F.(1 )(a) of this license because the exemption meets the requirements of 10 CFR 52.7, because the exemption is authorized by law, will not present an undue risk to the public health or safety, and is consistent with the common defense and security. Additionally, special circumstances are present in that the application of the regulation in this particular circumstance is not necessary to achieve the underlying purpose of the rule (10 CFR 50.12(a)(2)(ii)) as described in the staff SER dated August 5, 2011. Having applied the technically relevant criteria applicable to the application for the Early Site Permit No. ESP-004, to the variances requested in the application, as described in NUREG-1923, the staff SER dated July 2009, the following variances from the early site permit (ESP) are granted: 16 Amendment No. 124 H. I. (1) A variance (VEGP VAR 1.6-1) from Section 1.6 of the VEGP ESP site safety analysis report (SSAR) as it references Revision 15 of the AP1000 DCD instead of Revision 19 of the AP1000 DCD, which is incorporated by reference in the FSAR; (2) The variance (VEGP VAR 1.6-2) from Section 3.8.5, Foundations, of the VEGP ESP SSAR, which references Revision 15 of the AP1000 DCD, to reference Revision 19 of the AP1000 DCD, which is incorporated by reference in the FSAR; (3) The variance (VEGP VAR 1.6-3) from Chapter 15, Accident Analysis, of the VEGP ESP SSAR which references Revision 15 of the AP1000 DCD, to reference Revision 19 of the AP1000 DCD, which is incorporated by reference in the FSAR; (4) The variance (VEGP VAR 1.2-1) from the site layout information in Figures 1-4, 1-5, 13.3-2, and Part 5 Figure ii, of the VEGP ESP SSAR, which is superseded by the corresponding information in FSAR Section 1.1, Figure 1.1-202; (5) The variance (VEGP VAR 2.2-1) from the information related to onsite chemical hazards in Section 2.2.3.2.3 and Table 2.2-6 of the VEGP ESP SSAR, which is superseded by the corresponding information contained in FSAR Sections 2.2 and 6.4; and (6) The variance (VEGP VAR 2.3-1) from the information related to design-basis temperature characteristics in Section 2.3.1.5 and Table 1-1 of the VEGP ESP SSAR, which is superseded by the corresponding information contained in FSAR Section 2.3.1.5 and Table 2.0-201, which conforms to AP1000 DCD, Revision 19. IR db A d t N

emove y men men

o.

~ llowing g~JG's IT/\\/\\G olos1,1re notifioations 1,1nder paragraph (oH1) of 10 Gi;:R 52.QQ 1,1ntil tho GoR1R1ission Rlakos tho finding 1,1ndor 1 Q Gi;:R a2.10~(g), £NG shall notify the NRG, in a tiR=lel;i Rlanner, of new inforR1ation that R1ateriall;' alters the bases for deterRlining that either inspeotions, tests, or anal;ises were porforR1od as roei1,1irod, or that aoooptanoo oritoria are R1ot Tho notifioation R=11,1st oontain s1,1ffioiont inforR1ation to doR1onstrato that, notwithstanding tho nov, inforR=1ation, the presoribed inspeotions, tests, or anal;ises hm.ie been perforR1ed as reei 1,1ired, and the presoribed aooeptanoe oriteria are R1et £NG shall R=1aintain the g1,1idanoe and strategies de1,celoped in aooordanoe with 10 GFR aQ.a4(hh)(2). ~ Removed by Amendment No. #1#1 17 !Amendment No. ### J. This license is effective as of February 10, 2012, and shall expire at midnight on the date 40 yeaFS from the date that the Commission finds that the aooeptanoe eriteria in the eombined lioense are met in aooordanoe with 10 CFR 62.103(g). .... I J-u-ly-2-8,-2-06_3 __ --) Vv\\4cau1 fl~--- FOR THE NUCLEAR REGULATORY COMMISSION Appendices: Michael R. Johnson, irector Office of New Reactors Appendix A-Technical Specifications Appendix B - Environmental Protection Plan Appendix C - Inspections, Tests, Analyses, and Acceptance Criteria {ITAAC) 18 !Amendment No. ### 3.0 LCO Applicability LCO 3.0.5 LCO 3.0.6 LCO 3.0.7 VEGP Units 3 and 4 Technical Specifications LCO Applicability 3.0 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to LCO 3.0.2 for the system returned to service under administrative control to perform the test required to demonstrate OPERABILITY. When a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support system LCO ACTIONS are required to be entered. This is an exception to LCO 3.0.2 for the supported system. In this event, additional evaluations and limitations may be required in accordance with Specification 5.5.7, "Safety Function Determination Program (SFDP)." If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LCO 3.0.2. Test Exception LCOs 3.1.8 and 3.1.10 allow specified Technical Specification (TS) requirements to be changed to permit performance of special tests and operations. Unless otherwise specified, all other TS requirements remain unchanged. Compliance with Test Exception LCOs is optional. When a Test Exception LCO is desired to be met but is not met, the ACTIONS of the Test Exception LCO shall be met. When a Test Exception LCO is not desired to be met, entry into a MODE or other specified condition in the Applicability shall be made in accordance with the other applicable Specifications. l',dditionally, for Unit 4 only, ColTlbined Lieense Condition 2.0(9) pro>v<ides te1T1porary mmlc1sions for speoified T£ reqc1ire1T1ents prior to beoor=i=iing per1T1anently effeoti>.ie at initial oritioality of the roaster sore. ColTlplianoe with T£ reqc1ire1T1ents that are O)(Glc1ded fron=i bOGOITling eff:eoti1.ie while operating in MODES 4, e, and 6 in aeeordanee with the COL Condition is optional. 3.0 - 2 Amendment No. 117 (Unit 3) Amendment No. ~ (Unit 4) Technical Specifications ESFAS Instrumentation 3.3.8 Table 3.3.8-1 (page 1 of 2) Engineered Safeguards Actuation System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED FUNCTION CONDITIONS CHANNELS

1.

Containment Pressure

a. -Low 1,2,3,4,S'*l,5(*l 4

b. -Low 2 1,2,3,4,S'*l,6'*l 4

2.

Containment Pressure - High 2 1,2,3,4 4

3.

Containment Radioactivity - High 1,2,3,4(b) 4

4.

Containment Radioactivity - High 2 1,2,3 4

5.

Pressurizer Pressure - Low 3 1,2,3(c)(I) 4

6.

Pressurizer Water Level - Low 1,2 4

7.

Pressurizer Water Level - Low 2 1,2,3,4(b) 4 4(d),5(e) 4

8.

Pressurizer Water Level - High 1,2,3 4

9.

Pressurizer Water Level - High 2 1,2,3,4(I) 4

10.

Pressurizer Water Level - High 3 1,2,3,4(I) 4

11.

RCS Cold Leg Temperature (Tco1d) - Low 2 1,2,3(c)(I) 4 per loop

12.

Reactor Coolant Average Temperature (Tavg) 1,2 4 -Low

13.

Reactor Coolant Average Temperature (Tavg) 1,2 4 -Low 2

14.

RCS Wide Range Pressure - Low 1,2,3,4 4 5/Al 4 5(g)/Al 4 (a) Without an open containment air flow path <!: 6 inches in diameter. (b) With the RCS not being cooled by the Normal Residual Heat Removal System (RNS). (c) Above the P-11 (Pressurizer Pressure) interlock. (d) With the RCS being cooled by the RNS. (e) With RCS not VENTED and CMT actuation on Pressurizer Water Level - Low 2 not blocked. (f) With all four cold leg temperatures > 275°F. (g) With upper internals in place. CONDITIONS p p H E D F J Q E D D H K L (I) Below the P-11 (Pressurizer Pressure) interlock and RCS boron concentration is less than that necessary to meet the SOM requirements at an RCS temperature of 200°F. (R) FBF 61Ril a SRI',', RBI reei1,1ired le be GPeP!ABbe 13rier le iRilial srilisalily. VEGP Units 3 and 4 3.3.8 - 7 Amendment No. 4W (Unit 3) Amendment No. 147 (Unit 4) Technical Specifications ESFAS Instrumentation 3.3.8 Table 3.3.8-1 (page 2 of 2) Engineered Safeguards Actuation System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED FUNCTION CONDITIONS CHANNELS

15.

Core Makeup Tank (CMT) Level - Low 3 1,2,3,4(b) 4 per tank 4(d), 5(h) 4 per OPERABLE tank

16.

CMT Level - Low 6 1,2,3,4(b) 4 per tank 4(d), 5(h)(Aj 4 per OPERABLE tank

17.

Source Range Neutron Flux Doubling 2(il,3(i),40l 4 50l 4

18.

IRWST Lower Narrow Range Level - Low 3 1,2,3,4(b) 4 4(d), 5/Al 4 5(g)(Aj 4

19.

Reactor Coolant Pump Bearing Water 1,2,3,4 4 per RCP Temperature - High 2

20.

SG Narrow Range Water Level - Low 2 1,2,3,4(b) 4 per SG

21.

SG Wide Range Water Level - Low 2 1,2,3,4(b) 4 per SG

22.

SG Narrow Range Water Level High 1,2,3,4 4 per SG

23.

SG Narrow Range Water Level - High 3 1,2 4 per SG 3,4 4 per SG

24.

Steam Line Pressure - Low 2 1,2,3(c)(l)(m) 4 per steam line

25.

Steam Line Pressure - Negative Rate - High 3(k) 4 per steam line (b) With the RCS not being cooled by the Normal Residual Heat Removal System (RNS). (c) Above the P-11 (Pressurizer Pressure) interlock. (d) With the RCS being cooled by the RNS. (g) With upper internals in place. (h) With RCS not VENTED. CONDITIONS F J F J F M N 0 F F D G (i) With unborated water source flow paths not isolated except when critical or except during intentional approach to criticality. U) With unborated water source flow paths not isolated. (k) Below the P-11 (Pressurizer Pressure) interlock when Steam Line Pressure - Low 2 is blocked. (I) Below the P-11 (Pressurizer Pressure) interlock and RCS boron concentration is less than that necessary to meet the SOM requirements at an RCS temperature of 200°F. (m) Below the P-11 (Pressurizer Pressure) interlock when Steam Line Pressure - Low 2 is not blocked. (R) Fer URit 3 eRly, Rat reei1,1ired te be OJ2eRABLe 13rier te iRitial eritieality. VEGP Units 3 and 4 3.3.8 - 8 Amendment No. 4W (Unit 3) Amendment No. 147 (Unit 4) Technical Specifications Table 3.3.9-1 (page 1 of 2) ESFAS Manual Initiation 3.3.9 Engineered Safeguards Actuation System Instrumentation FUNCTION

1.

Safeguards Actuation - Manual Initiation

2.

Core Makeup Tank (CMT) Actuation - Manual Initiation

3.

Containment Isolation - Manual Initiation

4.

Steam Line Isolation - Manual Initiation

5.

Feedwater Isolation - Manual Initiation

6.

ADS Stages 1, 2 & 3 Actuation - Manual Initiation

7.

ADS Stage 4 Actuation - Manual Initiation

8.

Passive Containment Cooling Actuation - Manual Initiation

9.

Passive Residual Heat Removal Heat Exchanger Actuation - Manual Initiation

10.

Chemical and Volume Control System Makeup Isolation - Manual Initiation

11.

Normal Residual Heat Removal System Isolation - Manual Initiation APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS 1,2,3,4 5 1,2,3,4(a) 4(bl, 5(d) 1,2,3,4 1,2,3,4 1,2,3,4 1,2,3,4 5(d) 1,2,3,4 5(i) 5(e)fi) 1,2,3,4 5(1) 5(1) 1,2,3,4 5(c) 1,2,3,4(h) 1,2,3 (a) With the RCS not being cooled by the Normal Residual Heat Removal System (RNS). (b) With the RCS being cooled by the RNS. (c) With the RCS pressure boundary intact. (d) With RCS not VENTED. (e) With upper internals in place. (f) With decay heat> 7.0 MW!. (h) With all four cold leg temperatures> 275°F. (i) Fer URil d ORiy, ROI reEt~ireel le ee OPERABLE prier le iRilial eriliealily. VEGP Units 3 and 4 3.3.9 - 5 REQUIRED CHANNELS CONDITIONS 2 switches E 2 switches J 2 switches D 2 switches G 2 switches E 2 switches F 2 switches F 2 switch sets E 2 switch sets H 2 switch sets E 2 switch sets H 2 switch sets 2 switches E 2 switches J 2 switches K 2 Switches E 2 switches G 2 switches F 2 switch sets F Amendment No. 4W (Unit 3) Amendment No. 132 (Unit 4) Technical Specifications Table 3.3.9-1 (page 2 of 2) ESFAS Manual Initiation 3.3.9 Engineered Safeguards Actuation System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED FUNCTION CONDITIONS

12.

In-Containment Refueling Water Storage Tank 1,2,3,4(a) (IRWST) Injection Line Valve Actuation - 4(b),5fil Manual Initiation 6fil

13.

IRWST Containment Recirculation Valve 1,2,3,4(a) Actuation - Manual Initiation 4(b),5fil 5fil

14.

SG Power Operated Relief Valve and Block 1,2,3,4(a) Valve Isolation - Manual Initiation

15.

Containment Vacuum Relief Valve Actuation - 1,2,3,4,5'9l,6(g) Manual Initiation (a) With the RCS not being cooled by the Normal Residual Heat Removal System (RNS). (b) With the RCS being cooled by the RNS. (g) Without an open containment air flow path ;:: 6 inches in diameter. (i) FBF 61Ril a SRI',', RBI F0E!blireel le BO GPe~ABbe 13rier le iRilial srilisalily. VEGP Units 3 and 4 3.3.9 - 6 REQUIRED CHANNELS CONDITIONS 2 switch sets D 2 switch sets J 2 switch sets K 2 switch sets D 2 switch sets J 2 switch sets K 2 switches D 2 switches L Amendment No. 4W (Unit 3) Amendment No. 13 (Unit 4) Technical Specifications Table 3.3.10-1 (page 1 of 1) ESFAS RCS Hot Leg Level Instrumentation 3.3.10 Engineered Safeguards Actuation System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED FUNCTION CONDITIONS CHANNELS CONDITIONS

1.

Hot Leg Level - Low 4 5 (alfel,6 (b)fel 1 per loop C

2.

Hot Leg Level - Low 2 5 (c)fel 1 per loop D 5 (d)fel 1 per loop E (a) With CMT actuation on Pressurizer Water Level - Low 2 blocked. (b) With upper internals in place and with CMT actuation on Pressurizer Water Level - Low 2 blocked. (c) Below the P-12 (Pressurizer Level) interlock. (d) With the water level < 23 feet above the top of the reactor vessel flange. (e) Fer URil d ORiy, ROI reei~ireel le se OPERABLE prier le iRilial eriliealily. VEGP Units 3 and 4 3.3.10 - 5 Amendment No. 4W (Unit 3) Amendment No. 117 (Unit 4) Technical Specifications ES FAS Actuation Logic -Shutdown 3.3.16 3.3 INSTRUMENTATION 3.3.16 Engineered Safety Feature Actuation System (ESFAS) Actuation Logic - Shutdown LCO 3.3.16 Four divisions with one subsystem for each of the following Functions shall be OPERABLE: APPLICABILITY: ACTIONS

a. ESF Coincidence Logic; and

b. ESF Actuation.

- NOTES - 4-:- Only the divisions necessary to support Main Control Room Isolation, Air Supply Initiation, and Electrical Load De-energization are required to be OPERABLE during movement of irradiated fuel assemblies when not in MODE 1, 2, 3, 4, 5, or 6.

2. P:or Unit d onl;i, e8P: aotc1ation P:c1notion for ADS stage 4 flow paths, In GontainlTlent Refc1eling Water Storage +anl( ifljeetion and reeirec1lation flow paths, and GV8 letdown isolation valves, not reqc1ired to be OPeRABLe prior to initial oritioalit;i.

MODES 5 and 6, During movement of irradiated fuel assemblies. - NOTE - Separate condition entry is allowed for each Function. CONDITION A. One or more Functions A.1 within one required division inoperable. VEGP Units 3 and 4 REQUIRED ACTION Restore required division to OPERABLE status. COMPLETION TIME 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 3.3.16 - 1 Amendment No. 4W (Unit 3) Amendment No. 107 (Unit 4) Technical Specifications DAS Manual Controls 3.3.19 Table 3.3.19-1 (page 1 of 1) DAS Manual Controls

1.

2.

FUNCTION Reactor trip manual controls Passive Residual Heat Removal Heat Exchanger (PRHR HX) control and In-Containment Refueling Water Storage Tank (IRWST) gutter control valves

3.

Core Makeup Tank (CMT) isolation valves

4.

Automatic Depressurization System (ADS) stage 1 valves

5.

ADS stage 2 valves

6.

ADS stage 3 valves

7.

ADS stage 4 valves

8.

IRWST injection squib valves

9.

Containment recirculation valves

10.

Passive containment cooling drain valves

11.

Selected containment isolation valves APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS 1,2 1,2,3,4,5(al 1,2,3,4,5(al 1,2,3,4,5(al 1,2,3,4,5(a) 1,2,3,4,5(a) 1,2,3,4,5fat,6(c)fat 1,2,3,4,5fat,6fat 1,2,3,4,5fat,6fat 1,2,3,4,5(b),6(b) 1,2,3,4,5,6 (a) With Reactor Coolant System (RCS) pressure boundary intact. (b) With the reactor decay heat> 7.0 MWt. (c) With upper internals in place. (d) P:or Unit d only, not reqc1ired to be OPER/\\BU: prior to initial oritioality. REQUIRED CONTROLS 2 switches 2 switches 2 switches 2 switches 2 switches 2 switches 2 switches 2 switches 2 switches 2 switches 2 switches VEGP Units 3 and 4 3.3.19 - 3 Amendment No. 4W (Unit 3) Amendment No. 137 (Unit 4) Technical Specifications ADS and IRWST Injection Blocking Device 3.3.20 Table 3.3.20-1 (page 1 of 1) ADS and IRWST Injection Blocking Device FUNCTION

1.

Core Makeup Tank Level for Automatic Unblocking(a)

2.

ADS and IRWST Injection Block Switches for Manual Unblocking APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS 1,2,3,4(b) 1,2,3,4(b) REQUIRED CHANNELS PER DIVISION 2 SURVEILLANCE REQUIREMENTS SR 3.3.20.2 SR 3.3.20.3 SR 3.3.20.5 SR 3.3.20.3 SR 3.3.20.4 SR 3.3.20.1 SR 3.3.20.3 SR 3.3.20.4 (a) Not required to be OPERABLE with associated divisional ADS and IRWST Injection Block switch in the "unblock" position. (b) With the Reactor Coolant System (RCS) not being cooled by the Normal Residual Heat Removal System (RNS). (c) With the RCS being cooled by the RNS. fElt Fer URit 3 eRly, Rat re~l:lireel te se OJ2E;RABLE; ~rier te iRitial oritioality. VEGP Units 3 and 4 3.3.20 - 3 Amendment No. 4W (Unit 3) Amendment No. 166 (Unit 4) Technical Specifications 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.11 Automatic Depressurization System (ADS) - Operating LCO 3.4.11 Ten ADS flow paths shall be OPERABLE. NOTE: ADS - Operating 3.4.11 P:or Unit d only, in MODe 4, l\\Qg stage 4 ~ow paths are not reqc1ired to be OPE:RI\\BLE: prior to initial eritieality. APPLICABILITY: MODES 1, 2, 3, and 4. ACTIONS CONDITION A. One flow path in ADS stage 1, 2, or 3 inoperable. B. One flow path in ADS stage 4 inoperable. C. One flow path in ADS stage 1 inoperable and one flow path in ADS stage 2 or 3 inoperable. OR Two flow paths in ADS stage 1 inoperable. VEGP Units 3 and 4 A.1 B.1 C.1 REQUIRED ACTION COMPLETION TIME Restore flow path to 7 days OPERABLE status. Restore flow path to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months OPERABLE status. Restore one flow path to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months OPERABLE status. 3.4.11-1 Amendment No. ~ (Unit 3) Amendment No. 117 (Unit 4) Technical Specifications ADS - Shutdown, RCS Intact 3.4.12 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.12 Automatic Depressurization System (ADS) - Shutdown, RCS Intact LCO 3.4.12 APPLICABILITY: ACTIONS CONDITION A. With reactor subcritical for < 28 hrs:

1.

Five flow paths in ADS stage 1, 2, and 3 shall be OPERABLE; and

2.

Four flow paths in ADS stage 4 shall be OPERABLE B. With reactor subcritical for~ 28 hrs:

1.

Three flow paths in ADS stage 1, 2, and 3, with a minimum of two flow paths in ADS stage 2 or 3, shall be OPERABLE; and

2.

Three flow paths in ADS stage 4 shall be OPERABLE. NOTE P:or Unit d only, /\\Qg stage 4 ilow paths are not required to be OPeR/\\BU: prior to initial oritioality. MODE 5 with RCS pressure boundary intact and pressurizer level ~ 20%. REQUIRED ACTION COMPLETION TIME A. One required flow path A.1 Restore required flow path 7 days in ADS stage 1, 2, or 3 inoperable. B. One required flow path B.1 in ADS stage 4 inoperable. VEGP Units 3 and 4 to OPERABLE status. Restore required flow path 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to OPERABLE status. 3.4.12 - 1 Amendment No. ~ (Unit 3) Amendment No. 117 (Unit 4) Technical Specifications ADS - Shutdown, RCS Open 3.4.13 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.13 Automatic Depressurization System (ADS) - Shutdown, RCS Open LCO 3.4.13 APPLICABILITY: VEGP Units 3 and 4 A. With reactor subcritical for < 28 hrs:

1.

Five flow paths in ADS stage 1, 2, and 3 shall be open; and

2.

Four flow paths in ADS stage 4 shall be OPERABLE. B. With reactor subcritical for~ 28 hrs:

1.

Three flow paths in ADS stage 1, 2, and 3, with a minimum of two flow paths in ADS stage 2 or 3, shall be open; and

2.

Three flow paths in ADS stage 4 shall be OPERABLE. - NOTES - 4-:- In MODE 5, required flow paths in ADS stage 1, 2, and 3 may be closed provided they meet OPERABILITY requirements of LCO 3.4.12, ADS - Shutdown, RCS Intact, for the following:

a.

To facilitate RCS vacuum fill operations until a pressurizer level of~ 20% is established; or

b.

To facilitate LCO compliance during transitions between LCO 3.4.12 and LCO 3.4.13. ~ P:or Unit d only, l\\0£ stage 4 ilow paths are not reeic1ired to be OPE:RABLE: prior to initial eritieality. MODE 5 with pressurizer level < 20%, MODE 5 with RCS pressure boundary open, MODE 6 with upper internals in place. 3.4.13 - 1 Amendment No. ~ (Unit 3) Amendment No. 117 (Unit 4) Technical Specifications IRWST - Shutdown, MODES 3.5.7 3.5 PASSIVE CORE COOLING SYSTEM (PXS) 3.5.7 In-containment Refueling Water Storage Tank (IRWST) - Shutdown, MODE 5 LCO 3.5.7 The IRWST, with one injection flow path and one containment recirculation flow path, shall be OPERABLE. NOTE P:or Unit d only, not reeic1ired to be OPeRABU: prior to initial oritioality. APPLICABILITY: MODE 5. ACTIONS CONDITION A. Required motor operated containment recirculation isolation valve not fully open. B. Required IRWST injection flow path with noncondensible gas volume in one squib valve outlet line pipe stub not within limit. C. Required IRWST injection flow path with noncondensible gas volume in both squib valve outlet line pipe stubs not within limit. VEGP Units 3 and 4 A.1 B.1 C.1 REQUIRED ACTION COMPLETION TIME Open required motor 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months operated containment recirculation isolation valve. Restore noncondensible 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months gas volume in squib valve outlet line pipe stub to within limit. Restore noncondensible 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months gas volume in one squib valve outlet line pipe stub to within limit. 3.5.7 - 1 Amendment No. 4W (Unit 3) Amendment No. 13 (Unit 4) Technical Specifications IRWST - Shutdown, MODE6 3.5.8 3.5 PASSIVE CORE COOLING SYSTEM (PXS) 3.5.8 In-containment Refueling Water Storage Tank (IRWST) - Shutdown, MODE 6 LCO 3.5.8 The IRWST, with one injection flow path and one containment recirculation flow path, shall be OPERABLE. NOTE For Unit a only, not required to be OPE:RABLE: prior to initial eritieality. APPLICABILITY: MODE 6. ACTIONS CONDITION A. Required motor operated containment recirculation isolation valve not fully open. B. Required IRWST injection flow path with noncondensible gas volume in one squib valve outlet line pipe stub not within limit. C. Required IRWST injection flow path with noncondensible gas volume in both squib valve outlet line pipe stubs not within limit. VEGP Units 3 and 4 A.1 B.1 C.1 REQUIRED ACTION COMPLETION TIME Open required motor 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months operated containment recirculation isolation valve. Restore noncondensible 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months gas volume in squib valve outlet line pipe stub to within limit. Restore noncondensible 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months gas volume in one squib valve outlet line pipe stub to within limit. 3.5.8 - 1 Amendment No. 4W (Unit 3) Amendment No. 80 (Unit 4) ATTACHMENT 2 to NL-24-0228 Revised Technical Specifiation Pages License Amendment Request: Remove or Modify Outdated License Information (This enclosure consists of 15 pages, including this cover page) 3.0 LCO Applicability LCO 3.0.5 LCO 3.0.6 LCO 3.0.7 VEGP Units 3 and 4 Technical Specifications LCO Applicability 3.0 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to LCO 3.0.2 for the system returned to service under administrative control to perform the test required to demonstrate OPERABILITY. When a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support system LCO ACTIONS are required to be entered. This is an exception to LCO 3.0.2 for the supported system. In this event, additional evaluations and limitations may be required in accordance with Specification 5.5.7, "Safety Function Determination Program (SFDP)." If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LCO 3.0.2. Test Exception LCOs 3.1.8 and 3.1.10 allow specified Technical Specification (TS) requirements to be changed to permit performance of special tests and operations. Unless otherwise specified, all other TS requirements remain unchanged. Compliance with Test Exception LCOs is optional. When a Test Exception LCO is desired to be met but is not met, the ACTIONS of the Test Exception LCO shall be met. When a Test Exception LCO is not desired to be met, entry into a MODE or other specified condition in the Applicability shall be made in accordance with the other applicable Specifications. 3.0 - 2 Amendment No. 117 (Unit 3) Amendment No. '##-# (Unit 4) Technical Specifications ESFAS Instrumentation 3.3.8 Table 3.3.8-1 (page 1 of 2) Engineered Safeguards Actuation System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED FUNCTION CONDITIONS CHANNELS

1.

Containment Pressure

a. -Low 1,2,3,4,S'*l,5(*l 4

b. -Low 2 1,2,3,4,S'*l,6'*l 4

2.

Containment Pressure - High 2 1,2,3,4 4

3.

Containment Radioactivity - High 1,2,3,4(b) 4

4.

Containment Radioactivity - High 2 1,2,3 4

5.

Pressurizer Pressure - Low 3 1,2,3(c)(I) 4

6.

Pressurizer Water Level - Low 1,2 4

7.

Pressurizer Water Level - Low 2 1,2,3,4(b) 4 4(d),5(e) 4

8.

Pressurizer Water Level - High 1,2,3 4

9.

Pressurizer Water Level - High 2 1,2,3,4(I) 4

10.

Pressurizer Water Level - High 3 1,2,3,4(I) 4

11.

RCS Cold Leg Temperature (Tco1d) - Low 2 1,2,3(c)(I) 4 per loop

12.

Reactor Coolant Average Temperature (Tavg) 1,2 4 -Low

13.

Reactor Coolant Average Temperature (Tavg) 1,2 4 -Low 2

14.

RCS Wide Range Pressure - Low 1,2,3,4 4 5 4 5(g) 4 (a) Without an open containment air flow path <!: 6 inches in diameter. (b) With the RCS not being cooled by the Normal Residual Heat Removal System (RNS). (c) Above the P-11 (Pressurizer Pressure) interlock. (d) With the RCS being cooled by the RNS. (e) With RCS not VENTED and CMT actuation on Pressurizer Water Level - Low 2 not blocked. (f) With all four cold leg temperatures> 275°F. (g) With upper internals in place. CONDITIONS p p H E D F J Q E D D H K L (I) Below the P-11 (Pressurizer Pressure) interlock and RCS boron concentration is less than that necessary to meet the SOM requirements at an RCS temperature of 200°F. VEGP Units 3 and 4 3.3.8 - 7 Amendment No. "##-# (Unit 3) Amendment No. 147 (Unit 4) Technical Specifications ESFAS Instrumentation 3.3.8 Table 3.3.8-1 (page 2 of 2) Engineered Safeguards Actuation System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED FUNCTION CONDITIONS CHANNELS

15.

Core Makeup Tank (CMT) Level - Low 3 1,2,3,4(b) 4 per tank 4(d), 5(h) 4 per OPERABLE tank

16.

CMT Level - Low 6 1,2,3,4(b) 4 per tank 4(d), 5(h) 4 per OPERABLE tank

17.

Source Range Neutron Flux Doubling 2(il,3(i),40l 4 50l 4

18.

IRWST Lower Narrow Range Level - Low 3 1,2,3,4(b) 4 4(dl_5 4 5 (9) 4

19.

Reactor Coolant Pump Bearing Water 1,2,3,4 4 per RCP Temperature - High 2

20.

SG Narrow Range Water Level - Low 2 1,2,3,4(b) 4 per SG

21.

SG Wide Range Water Level - Low 2 1,2,3,4(b) 4 per SG

22.

SG Narrow Range Water Level High 1,2,3,4 4 per SG

23.

SG Narrow Range Water Level - High 3 1,2 4 per SG 3,4 4 per SG

24.

Steam Line Pressure - Low 2 1,2,3(c)(l)(m) 4 per steam line

25.

Steam Line Pressure - Negative Rate - High 3(k) 4 per steam line (b) With the RCS not being cooled by the Normal Residual Heat Removal System (RNS). (c) Above the P-11 (Pressurizer Pressure) interlock. (d) With the RCS being cooled by the RNS. (g) With upper internals in place. (h) With RCS not VENTED. CONDITIONS F J F J F M N 0 F F D G (i) With unborated water source flow paths not isolated except when critical or except during intentional approach to criticality. U) With unborated water source flow paths not isolated. (k) Below the P-11 (Pressurizer Pressure) interlock when Steam Line Pressure - Low 2 is blocked. (I) Below the P-11 (Pressurizer Pressure) interlock and RCS boron concentration is less than that necessary to meet the SOM requirements at an RCS temperature of 200°F. (m) Below the P-11 (Pressurizer Pressure) interlock when Steam Line Pressure - Low 2 is not blocked. VEGP Units 3 and 4 3.3.8 - 8 Amendment No. "##-# (Unit 3) Amendment No. 147 (Unit 4) Technical Specifications Table 3.3.9-1 (page 1 of 2) ESFAS Manual Initiation 3.3.9 Engineered Safeguards Actuation System Instrumentation FUNCTION

1.

Safeguards Actuation - Manual Initiation

2.

Core Makeup Tank (CMT) Actuation - Manual Initiation

3.

Containment Isolation - Manual Initiation

4.

Steam Line Isolation - Manual Initiation

5.

Feedwater Isolation - Manual Initiation

6.

ADS Stages 1, 2 & 3 Actuation - Manual Initiation

7.

ADS Stage 4 Actuation - Manual Initiation

8.

Passive Containment Cooling Actuation - Manual Initiation

9.

Passive Residual Heat Removal Heat Exchanger Actuation - Manual Initiation

10.

Chemical and Volume Control System Makeup Isolation - Manual Initiation

11.

Normal Residual Heat Removal System Isolation - Manual Initiation APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS 1,2,3,4 5 1,2,3,4(a) 4(bl, 5(d) 1,2,3,4 1,2,3,4 1,2,3,4 1,2,3,4 5(d) 1,2,3,4 5 5(e) 1,2,3,4 5(1) 5(1) 1,2,3,4 5(c) 1,2,3,4(h) 1,2,3 (a) With the RCS not being cooled by the Normal Residual Heat Removal System (RNS). (b) With the RCS being cooled by the RNS. (c) With the RCS pressure boundary intact. (d) With RCS not VENTED. (e) With upper internals in place. (f) With decay heat> 7.0 MW!. (h) With all four cold leg temperatures> 275°F. VEGP Units 3 and 4 3.3.9 - 5 REQUIRED CHANNELS CONDITIONS 2 switches E 2 switches J 2 switches D 2 switches G 2 switches E 2 switches F 2 switches F 2 switch sets E 2 switch sets H 2 switch sets E 2 switch sets H 2 switch sets 2 switches E 2 switches J 2 switches K 2 Switches E 2 switches G 2 switches F 2 switch sets F Amendment No. "##-# (Unit 3) Amendment No. 132 (Unit 4) Technical Specifications Table 3.3.9-1 (page 2 of 2) ESFAS Manual Initiation 3.3.9 Engineered Safeguards Actuation System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED FUNCTION CONDITIONS

12.

In-Containment Refueling Water Storage Tank 1,2,3,4(a) (IRWST) Injection Line Valve Actuation - 4(b),5 Manual Initiation 6

13.

IRWST Containment Recirculation Valve 1,2,3,4(a) Actuation - Manual Initiation 4(b),5 6

14.

SG Power Operated Relief Valve and Block 1,2,3,4(a) Valve Isolation - Manual Initiation

15.

Containment Vacuum Relief Valve Actuation - 1,2,3,4,5'9l,6(g) Manual Initiation (a) With the RCS not being cooled by the Normal Residual Heat Removal System (RNS). (b) With the RCS being cooled by the RNS. (g) Without an open containment air flow path ;:: 6 inches in diameter. VEGP Units 3 and 4 3.3.9 - 6 REQUIRED CHANNELS CONDITIONS 2 switch sets D 2 switch sets J 2 switch sets K 2 switch sets D 2 switch sets J 2 switch sets K 2 switches D 2 switches L Amendment No. "##-# (Unit 3) Amendment No. 13 (Unit 4) Technical Specifications Table 3.3.10-1 (page 1 of 1) ESFAS RCS Hot Leg Level Instrumentation 3.3.10 Engineered Safeguards Actuation System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED FUNCTION CONDITIONS CHANNELS CONDITIONS

1.

Hot Leg Level - Low 4 5 (al, 5 (b) 1 per loop C

2.

Hot Leg Level - Low 2 5 (c) 1 per loop D 5 (d) 1 per loop E (a) With CMT actuation on Pressurizer Water Level - Low 2 blocked. (b) With upper internals in place and with CMT actuation on Pressurizer Water Level - Low 2 blocked. (c) Below the P-12 (Pressurizer Level) interlock. (d) With the water level < 23 feet above the top of the reactor vessel flange. VEGP Units 3 and 4 3.3.10 - 5 Amendment No. "##-# (Unit 3) Amendment No. 117 (Unit 4) Technical Specifications ES FAS Actuation Logic -Shutdown 3.3.16 3.3 INSTRUMENTATION 3.3.16 Engineered Safety Feature Actuation System (ESFAS) Actuation Logic - Shutdown LCO 3.3.16 Four divisions with one subsystem for each of the following Functions shall be OPERABLE: APPLICABILITY: ACTIONS

a. ESF Coincidence Logic; and

b. ESF Actuation.

- NOTE - Only the divisions necessary to support Main Control Room Isolation, Air Supply Initiation, and Electrical Load De-energization are required to be OPERABLE during movement of irradiated fuel assemblies when not in MODE 1, 2, 3, 4, 5, or 6. MODES 5 and 6, During movement of irradiated fuel assemblies. - NOTE - Separate condition entry is allowed for each Function. CONDITION A. One or more Functions A.1 within one required division inoperable. VEGP Units 3 and 4 REQUIRED ACTION Restore required division to OPERABLE status. COMPLETION TIME 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 3.3.16 - 1 Amendment No. '##-# (Unit 3) Amendment No. 107 (Unit 4) Technical Specifications DAS Manual Controls 3.3.19 Table 3.3.19-1 (page 1 of 1) DAS Manual Controls

1.

2.

FUNCTION Reactor trip manual controls Passive Residual Heat Removal Heat Exchanger (PRHR HX) control and In-Containment Refueling Water Storage Tank (IRWST) gutter control valves

3.

Core Makeup Tank (CMT) isolation valves

4.

Automatic Depressurization System (ADS) stage 1 valves

5.

ADS stage 2 valves

6.

ADS stage 3 valves

7.

ADS stage 4 valves

8.

IRWST injection squib valves

9.

Containment recirculation valves

10.

Passive containment cooling drain valves

11.

Selected containment isolation valves APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS 1,2 1,2,3,4,5(al 1,2,3,4,5(al 1,2,3,4,5(al 1,2,3,4,5(al 1,2,3,4,5(al 1,2,3,4,5,6(c) 1,2,3,4,5,6 1,2,3,4,5,6 1,2,3,4,5(b),6(b) 1,2,3,4,5,6 (a) With Reactor Coolant System (RCS) pressure boundary intact. (b) With the reactor decay heat> 7.0 MWt. (c) With upper internals in place. REQUIRED CONTROLS 2 switches 2 switches 2 switches 2 switches 2 switches 2 switches 2 switches 2 switches 2 switches 2 switches 2 switches VEGP Units 3 and 4 3.3.19 - 3 Amendment No. '##-# (Unit 3) Amendment No. 137 (Unit 4) Technical Specifications Table 3.3.20-1 (page 1 of 1) ADS and IRWST Injection Blocking Device 3.3.20 ADS and IRWST Injection Blocking Device FUNCTION

1.

Core Makeup Tank Level for Automatic Unblocking(a)

2.

ADS and IRWST Injection Block Switches for Manual Unblocking APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS 1,2,3,4(b) 1,2,3,4(b) REQUIRED CHANNELS PER DIVISION 2 SURVEILLANCE REQUIREMENTS SR 3.3.20.2 SR 3.3.20.3 SR 3.3.20.5 SR 3.3.20.3 SR 3.3.20.4 SR 3.3.20.1 SR 3.3.20.3 SR 3.3.20.4 (a) Not required to be OPERABLE with associated divisional ADS and IRWST Injection Block switch in the "unblock" position. (b) With the Reactor Coolant System (RCS) not being cooled by the Normal Residual Heat Removal System (RNS). (c) With the RCS being cooled by the RNS. VEGP Units 3 and 4 3.3.20 - 3 Amendment No. "##-# (Unit 3) Amendment No. 166 (Unit 4) Technical Specifications 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.11 Automatic Depressurization System (ADS) - Operating LCO 3.4.11 Ten ADS flow paths shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. ACTIONS CONDITION REQUIRED ACTION A. One flow path in ADS A.1 Restore flow path to stage 1, 2, or 3 OPERABLE status. inoperable. B. One flow path in ADS B.1 Restore flow path to stage 4 inoperable. OPERABLE status. ADS - Operating 3.4.11 COMPLETION TIME 7 days 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months C. One flow path in ADS C.1 Restore one flow path to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months stage 1 inoperable and one flow path in ADS stage 2 or 3 inoperable. OR Two flow paths in ADS stage 1 inoperable. VEGP Units 3 and 4 OPERABLE status. 3.4.11-1 Amendment No. '##-# (Unit 3) Amendment No. 117 (Unit 4) Technical Specifications ADS - Shutdown, RCS Intact 3.4.12 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.12 Automatic Depressurization System (ADS) - Shutdown, RCS Intact LCO 3.4.12 APPLICABILITY: ACTIONS CONDITION A. With reactor subcritical for < 28 hrs:

1.

Five flow paths in ADS stage 1, 2, and 3 shall be OPERABLE; and

2.

Four flow paths in ADS stage 4 shall be OPERABLE B. With reactor subcritical for~ 28 hrs:

1.

Three flow paths in ADS stage 1, 2, and 3, with a minimum of two flow paths in ADS stage 2 or 3, shall be OPERABLE; and

2.

Three flow paths in ADS stage 4 shall be OPERABLE. MODE 5 with RCS pressure boundary intact and pressurizer level ~ 20%. REQUIRED ACTION COMPLETION TIME A. One required flow path A.1 Restore required flow path 7 days in ADS stage 1, 2, or 3 inoperable. B. One required flow path B.1 in ADS stage 4 inoperable. VEGP Units 3 and 4 to OPERABLE status. Restore required flow path 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to OPERABLE status. 3.4.12 - 1 Amendment No. '##-# (Unit 3) Amendment No. 117 (Unit 4) Technical Specifications ADS - Shutdown, RCS Open 3.4.13 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.13 Automatic Depressurization System (ADS) - Shutdown, RCS Open LCO 3.4.13 APPLICABILITY: VEGP Units 3 and 4 A. With reactor subcritical for < 28 hrs:

1.

Five flow paths in ADS stage 1, 2, and 3 shall be open; and

2.

Four flow paths in ADS stage 4 shall be OPERABLE. B. With reactor subcritical for~ 28 hrs:

1.

Three flow paths in ADS stage 1, 2, and 3, with a minimum of two flow paths in ADS stage 2 or 3, shall be open; and

2.

Three flow paths in ADS stage 4 shall be OPERABLE. - NOTE - In MODE 5, required flow paths in ADS stage 1, 2, and 3 may be closed provided they meet OPERABILITY requirements of LCO 3.4.12, ADS - Shutdown, RCS Intact, for the following:

a.

To facilitate RCS vacuum fill operations until a pressurizer level of~ 20% is established; or

b.

To facilitate LCO compliance during transitions between LCO 3.4.12 and LCO 3.4.13. MODE 5 with pressurizer level< 20%, MODE 5 with RCS pressure boundary open, MODE 6 with upper internals in place. 3.4.13 - 1 Amendment No. '##-# (Unit 3) Amendment No. 117 (Unit 4) Technical Specifications IRWST - Shutdown, MODES 3.5.7 3.5 PASSIVE CORE COOLING SYSTEM (PXS) 3.5.7 In-containment Refueling Water Storage Tank (IRWST) - Shutdown, MODE 5 LCO 3.5.7 The IRWST, with one injection flow path and one containment recirculation flow path, shall be OPERABLE. APPLICABILITY: MODE 5. ACTIONS CONDITION A. Required motor operated containment recirculation isolation valve not fully open. B. Required IRWST injection flow path with noncondensible gas volume in one squib valve outlet line pipe stub not within limit. C. Required IRWST injection flow path with noncondensible gas volume in both squib valve outlet line pipe stubs not within limit. VEGP Units 3 and 4 A.1 B.1 C.1 REQUIRED ACTION COMPLETION TIME Open required motor 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months operated containment recirculation isolation valve. Restore noncondensible 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months gas volume in squib valve outlet line pipe stub to within limit. Restore noncondensible 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months gas volume in one squib valve outlet line pipe stub to within limit. 3.5.7 - 1 Amendment No. '##-# (Unit 3) Amendment No. 13 (Unit 4) Technical Specifications IRWST - Shutdown, MODE6 3.5.8 3.5 PASSIVE CORE COOLING SYSTEM (PXS) 3.5.8 In-containment Refueling Water Storage Tank (IRWST) - Shutdown, MODE 6 LCO 3.5.8 The IRWST, with one injection flow path and one containment recirculation flow path, shall be OPERABLE. APPLICABILITY: MODE 6. ACTIONS CONDITION A. Required motor operated containment recirculation isolation valve not fully open. B. Required IRWST injection flow path with noncondensible gas volume in one squib valve outlet line pipe stub not within limit. C. Required IRWST injection flow path with noncondensible gas volume in both squib valve outlet line pipe stubs not within limit. VEGP Units 3 and 4 A.1 B.1 C.1 REQUIRED ACTION COMPLETION TIME Open required motor 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months operated containment recirculation isolation valve. Restore noncondensible 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months gas volume in squib valve outlet line pipe stub to within limit. Restore noncondensible 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months gas volume in one squib valve outlet line pipe stub to within limit. 3.5.8 - 1 Amendment No. '##-# (Unit 3) Amendment No. 80 (Unit 4) ATTACHMENT 3 to NL-24-0228 Associated Technical Specification Bases Change {For information only) License Amendment Request: Remove or Modify Outdated License Information (This enclosure consists of 2 pages, including this cover page) to NL-24-0228 Associated Technical Specification Bases Change (for information only) In association with the changes to Unit 4 TS LCO 3.0.7, associated with Test Exceptions, the Bases for LCO 3.0.7 are modified as shown below (Bases page B 3.0 - 9). This change consistent with removal of pre-initial criticality exceptions. LCO 3.0.7 Additionally, for Unit 4 only, Combined License (COL) Condition 2.D(Q) provides temporary exclusions for specified TS requirements commencing 1Nith the 1 O CFR 52.103(g) finding prior to initial criticality. The TS exclusions listed in the COL Condition are only portions of the TS requirements Applicable in MODeS 4, 5, and 6. All TS requirements become permanently effective at initial criticality of the reactor core.

v t e Letter Sequence Request
EPID:L-2025-LLA-0039, (Open)
Initiation Acceptance Administration Questions Answers Results Approval Other:
MONTHYEAR2025-03-06 [Table View]

NL-24-0228, License Amendment Request to Remove or Modify Outdated License Information

Text

Enclosure:

SUMMARY

SUMMARY

Navigation menu

NL-24-0228, License Amendment Request to Remove or Modify Outdated License Information

Text

Enclosure:

SUMMARY

SUMMARY

Navigation menu

Search