Text

Application to Revise Technical Specifications to Adopt TSTF-576, Revise Safety/Relief Valve Requirements
	ML25339A156
Person / Time
Site:	Hatch
Issue date:	12/05/2025
From:	Coleman J; Southern Nuclear Operating Co
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
References
	NL-25-0255
	Download: ML25339A156 (0)
	v • d • e

Regulatory Affairs 3535 Colonnade Parkway Birmingham, AL 35243 205 992 5000 December 5, 2025 Docket Nos.: 50-321 NL-25-0255 50-366 10 CFR 50.90 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant - Units 1 & 2 Application to Revise Technical Specifications to Adopt TSTF-576, Revise Safety/Relief Valve Requirements Ladies and Gentlemen:

Pursuant to 10 CFR 50.90, Southern Nuclear Operating Company (SNC) is submitting a request for an amendment to the Technical Specifications (TS) for Edwin I. Hatch Nuclear Plant (HNP)

Units 1 and 2.

Southern Nuclear Operating Company (SNC) requests adoption of TSTF-576, Revise Safety/Relief Valve Requirements. The proposed change revises the Safety/Relief Valve (S/RV) TS to align the overpressure protection requirements with the safety limits and the regulations.

The enclosure provides a description and assessment of the proposed changes. Attachment 1 provides the existing TS pages marked to show the proposed changes. Attachment 2 provides revised (clean) TS pages. Attachment 3 provides the existing TS Bases pages marked to show revised text associated with the proposed TS changes and is provided for information only. provides an example revised Core Operating Limits Report, for information, that illustrates the addition of the S/RV limits.

SNC requests that the amendment be reviewed under the Consolidated Line Item Improvement Process (CLIIP). Approval of the proposed amendment is requested within six months of acceptance. Once approved, the amendment shall be implemented within 90 days.

There are no regulatory commitments made in this submittal.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated Georgia Official.

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

U. S. Nuclear Regulatory Commission NL-25-0255 Page 2 I declare under penalty of perjury that the foregoing is true and correct. Executed on the 5th day of December 2025.

Respectfully submitted, Jamie M. Coleman Director, Regulatory Affairs Southern Nuclear Operating Company JMC/agq/cbg

Enclosure:

Description and Assessment Attachments: 1.

Proposed Technical Specification Changes (Mark-up)

2.

Revised Technical Specification Pages (Clean-typed)

3.

Proposed Technical Specification Bases Changes (Mark-up) - For Information Only

4.

Example Updated Core Operating Limits Report - For Information Only cc:

Regional Administrator, Region ll NRR Project Manager - Hatch 1 & 2 Senior Resident Inspector - Hatch 1 & 2 Director, Environmental Protection Division - State of Georgia RType: CHA02.004 Digitally signed by JAMIEMCO Date: 2025.12.05 14:05:58

-06'00'

Edwin I. Hatch Nuclear Plant - Units 1 & 2 Application to Revise Technical Specifications to Adopt TSTF-576, Revise Safety/Relief Valve Requirements Enclosure to NL-25-0255 Description and Assessment

Enclosure to NL-25-0255 Description and Assessment E-1

1.0 DESCRIPTION

Southern Nuclear Operating Company (SNC) requests adoption of TSTF-576, Revise Safety/Relief Valve Requirements. The proposed change revises the Safety/Relief Valve (S/RV) Technical Specifications (TS) to align the overpressure protection requirements with the safety limits and the regulations.

2.0 ASSESSMENT

2.1 Applicability of Safety Evaluation SNC has reviewed the safety evaluation for TSTF-576 provided to the Technical Specifications Task Force in a letter dated September 11, 2024 [ML24249A148]. This review included a review of the NRC staffs evaluation, as well as the information provided in TSTF-576. As described herein, SNC has concluded that the justifications presented in TSTF-576 and the safety evaluation prepared by the NRC staff are applicable to Edwin I. Hatch Nuclear Plant (HNP) Units 1 and 2 and justify this amendment for the incorporation of the changes to the HNP Units 1 and 2 TS.

The NRC-approved overpressure protection analysis methodology for HNP Units 1 and 2 is Global Nuclear Fuel Report NEDE-24011-P-A, General Electric Standard Application for Reactor Fuel (GESTAR II).

2.2 Optional Changes and Variations SNC is proposing the following variations from the TS changes described in TSTF-576 or the applicable parts of the NRC staffs safety evaluation:

TSTF-576 revises the Standard Technical Specifications (STS) table of contents (TOC) to change STS 3.4.3, Safety/Relief Valves (S/RVs) to Overpressure Protection System (OPS). The TOC of the HNP TS is licensee-controlled and is not included in Attachments 1 and 2.

TSTF-576 revises Condition A, B, and C of STS 3.4.3, Safety/Relief Valves (S/RVs).

The HNP TS contain requirements for safety/relief valves (S/RVs) that differ from the STS on which TSTF-576 was based but are encompassed in the TSTF-576 justification.

Condition C that is revised in STS 3.4.3 is analogous to Condition A in the HNP TS, and HNP TS 3.4.3 does not contain any Conditions corresponding to Condition A or B of STS 3.4.3.

TSTF-576 revises Surveillance Requirement (SR) 3.4.3.1 by deleting a Note that is not contained in the HNP TS SR 3.4.3.1. The Note provides flexibility, and removing the Note from the STS is conservative. By not containing this note, the HNP TS SR 3.4.3.1 currently reflects this conservatism and is encompassed in the TSTF-576 justification.

The remaining changes to SR 3.4.3.1 are applicable to the HNP TS.

TSTF-576 revises SR 3.4.3.2 by replacing the surveillance, in its entirety, with new language that directs operators to the Core Operating Limits Report (COLR) for as-found lift pressure limits. SR 3.4.3.2 does not currently exist in the HNP TS. However, SR 3.4.3.2, as-replaced and justified in TSTF-576, does apply to the HNP adoption of TSTF-576 and is included in this request.

Enclosure to NL-25-0255 Description and Assessment E-2

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration Analysis Southern Nuclear Operating Company (SNC) requests adoption of TSTF-576, Revise Safety/Relief Valve Requirements. The proposed change revises the Safety/Relief Valve (S/RV) Technical Specifications (TS) to align the overpressure protection requirements with the safety limits and the regulations. The Limiting Condition for Operation (LCO) is revised to replace requirements on each credited S/RV with a requirement that the Overpressure Protection System (OPS) be operable. The Surveillance Requirements (SRs) are revised to move the as-found S/RV lift pressure limits to the licensee-controlled Core Operating Limits Report. The TS Actions are revised to be consistent with the changes to the LCO and SRs.

Administrative changes are made to the TS for clarity and consistency. The Core Operating Limits Report specification is revised to reference the Overpressure Protection System specification.

SNC has evaluated if a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:

1.

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

Response: No The proposed change revises the S/RV TS to align the overpressure protection requirements with the safety limits and the regulations. The overpressure protection system must accommodate the most severe pressurization transient. Evaluations have determined that the most severe transient is the closure of all main steam isolation valves (MSIVs), followed by reactor scram on high neutron flux (i.e., failure of the direct scram associated with MSIV position). The proposed change does not affect the MSIVs and would have no effect on the probability of the MSIVs closing or generation of a reactor scram signal on MSIV position. Therefore, the probability of the event is unaffected. The consequences of the accident are based on the peak reactor pressure vessel pressure. Both the current and proposed TS ensure the overpressure Safety Limit is not exceeded. The accident analyses consider the aggregate operation of the credited S/RVs, not the performance of individual valves. The proposed change moves the S/RV as-found lift pressure limits to the licensee-controlled Core Operating Limits Report which uses NRC-approved methodologies. Altering the control process for these values has no effect on the accident evaluations. As a result, the consequences of the accident are not changed.

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

Enclosure to NL-25-0255 Description and Assessment E-3 The proposed change revises the S/RV TS to align the overpressure protection requirements with the safety limits and the regulations. The proposed change does not alter the design function or operation of the S/RVs. The proposed change does not create any new credible failure mechanisms, malfunctions, or accident initiators not already considered in the design and licensing basis.

Therefore, the proposed change does 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 change revises the S/RV TS to align the overpressure protection requirements with the safety limits and the regulations. The proposed change ensures that the S/RVs can protect Safety Limit 2.1.2. Although the as-found S/RV lift pressure limits are moved to the licensee-controlled Core Operating Limits Report, the safety margin provided by the S/RVs, which ensures the Safety Limit is protected, is not changed. The conservatisms in the evaluation and the analysis are described in the NRC-approved methods for each licensee, which are not altered by the proposed change. The proposed change does not alter a design basis limit or a safety limit, and, therefore, does not reduce the margin of safety.

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

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

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

4.0 ENVIRONMENTAL EVALUATION A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or 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.

Edwin I. Hatch Nuclear Plant - Units 1 & 2 Application to Revise Technical Specifications to Adopt TSTF-576, Revise Safety/Relief Valve Requirements to NL-25-0255 Proposed Technical Specification Changes (Mark-up)

OPSS/RVs 3.4.3 HATCH UNIT 1 3.4-5 Amendment No. 266 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 Overpressure Protection System (OPS)Safety/Relief Valves (S/RVs)

LCO 3.4.3 The OPS safety function of 10 of 11 S/RVs shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

OPS inoperableTwo or more S/RVs inoperable.

A.1 Be in MODE 3.

AND A.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 36 hours

OPSS/RVs 3.4.3 HATCH UNIT 1 3.4-6 Amendment No. 323 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 Verify the as-left safety functionOPS lift setpoints pressures of the required safety/relief valves (S/RVs) are within +/- 1% of the nominal setpointas follows:

Nominal Number of Setpoint OPS S/RVs (psig) 1011 1160 34.8 Following testing, lift settings shall be within 1%.

In accordance with the INSERVICE TESTING PROGRAM SR 3.4.3.2 Verify the as-found OPS lift pressures of the required S/RVs are within the limits specified in the COLR.

In accordance with the INSERVICE TESTING PROGRAM

Reporting Requirements 5.6 HATCH UNIT 1 5.0-21 Amendment No. 299 5.6 Reporting Requirements 5.6.2 Annual Radiological Environmental Operating Report (continued) format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.

5.6.3 Radioactive Effluent Release Report

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

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

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

5.6.4 Deleted.

5.6.5 CORE OPERATING LIMITS REPORT (COLR)

a.

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

1)

The Average Planar Linear Heat Generation Rate for Specification 3.2.1.

2)

The Minimum Critical Power Ratio (MCPR) for Specification 3.2.2 and the MCPR99.9% value used to calculate the Specification 3.2.2 MCPR.

3)

The Linear Heat Generation Rate for Specification 3.2.3.

4)

The as-found Overpressure Protection System Lift Pressures for Specification 3.4.3.

(continued)

OPSS/RVs 3.4.3 HATCH UNIT 2 3.4-5 Amendment No. 210 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 Overpressure Protection System (OPS)Safety/Relief Valves (S/RVs)

LCO 3.4.3 The OPSsafety function of 10 of 11 S/RVs shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

OPS inoperableTwo or more S/RVs inoperable.

A.1 Be in MODE 3.

AND A.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 36 hours

OPSS/RVs 3.4.3 HATCH UNIT 2 3.4-6 Amendment No. 268 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 Verify the as-left safety functionOPS lift setpoints pressures of the required safety/relief valves (S/RVs) are within 1% of the nominal setpointas follows:

Nominal Number of Setpoint OPS S/RVs (psig) 1011 1160 34.8 Following testing, lift settings shall be within +/- 1%.

In accordance with the INSERVICE TESTING PROGRAM SR 3.4.3.2 Verify the as-found OPS lift pressures of the required S/RVs are within the limits specified in the COLR.

In accordance with the INSERVICE TESTING PROGRAM

Reporting Requirements 5.6 HATCH UNIT 2 5.0-21 Amendment No. 244 5.6 Reporting Requirements 5.6.2 Annual Radiological Environmental Operating Report (continued) table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.

5.6.3 Radioactive Effluent Release Report

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

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

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

5.6.4 Deleted.

5.6.5 CORE OPERATING LIMITS REPORT (COLR)

a.

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

1)

The Average Planar Linear Heat Generation Rate for Specification 3.2.1.

2)

The Minimum Critical Power Ratio (MCPR) for Specification 3.2.2 and the MCPR99.9% value used to calculate the Specification 3.2.2 MCPR.

3)

The Linear Heat Generation Rate for Specification 3.2.3.

4)

The as-found Overpressure Protection System Lift Pressures for Specification 3.4.3.

(continued)

Edwin I. Hatch Nuclear Plant - Units 1 & 2 Application to Revise Technical Specifications to Adopt TSTF-576, Revise Safety/Relief Valve Requirements to NL-25-0255 Revised Technical Specification Pages (Clean-typed)

OPS 3.4.3 HATCH UNIT 1 3.4-5 Amendment No. 266 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 Overpressure Protection System (OPS)

LCO 3.4.3 The OPS shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

OPS inoperable.

A.1 Be in MODE 3.

AND A.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 36 hours

OPS 3.4.3 HATCH UNIT 1 3.4-6 Amendment No. 323 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 Verify the as-left OPS lift pressures of the required safety/relief valves (S/RVs) are within +/- 1% of the nominal setpoint:

Nominal Number of Setpoint OPS S/RVs (psig) 10 1160 In accordance with the INSERVICE TESTING PROGRAM SR 3.4.3.2 Verify the as-found OPS lift pressures of the required S/RVs are within the limits specified in the COLR.

In accordance with the INSERVICE TESTING PROGRAM

Reporting Requirements 5.6 HATCH UNIT 1 5.0-21 Amendment No. 299 5.6 Reporting Requirements 5.6.2 Annual Radiological Environmental Operating Report (continued) format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.

5.6.3 Radioactive Effluent Release Report

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

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

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

5.6.4 Deleted.

5.6.5 CORE OPERATING LIMITS REPORT (COLR)

a.

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

1)

The Average Planar Linear Heat Generation Rate for Specification 3.2.1.

2)

The Minimum Critical Power Ratio (MCPR) for Specification 3.2.2 and the MCPR99.9% value used to calculate the Specification 3.2.2 MCPR.

3)

The Linear Heat Generation Rate for Specification 3.2.3.

4)

The as-found Overpressure Protection System Lift Pressures for Specification 3.4.3.

(continued)

OPS 3.4.3 HATCH UNIT 2 3.4-5 Amendment No. 210 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 Overpressure Protection System (OPS)

LCO 3.4.3 The OPS shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

OPS inoperable.

A.1 Be in MODE 3.

AND A.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 36 hours

OPS 3.4.3 HATCH UNIT 2 3.4-6 Amendment No. 268 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 Verify the as-left OPS lift pressures of the required safety/relief valves (S/RVs) are within 1% of the nominal setpoint:

Nominal Number of Setpoint OPS S/RVs (psig) 10 1160 In accordance with the INSERVICE TESTING PROGRAM SR 3.4.3.2 Verify the as-found OPS lift pressures of the required S/RVs are within the limits specified in the COLR.

In accordance with the INSERVICE TESTING PROGRAM

Reporting Requirements 5.6 HATCH UNIT 2 5.0-21 Amendment No. 244 5.6 Reporting Requirements 5.6.2 Annual Radiological Environmental Operating Report (continued) table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.

5.6.3 Radioactive Effluent Release Report

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

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

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

5.6.4 Deleted.

5.6.5 CORE OPERATING LIMITS REPORT (COLR)

a.

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

1)

The Average Planar Linear Heat Generation Rate for Specification 3.2.1.

2)

The Minimum Critical Power Ratio (MCPR) for Specification 3.2.2 and the MCPR99.9% value used to calculate the Specification 3.2.2 MCPR.

3)

The Linear Heat Generation Rate for Specification 3.2.3.

4)

The as-found Overpressure Protection System Lift Pressures for Specification 3.4.3.

(continued)

Edwin I. Hatch Nuclear Plant - Units 1 & 2 Application to Revise Technical Specifications to Adopt TSTF-576, Revise Safety/Relief Valve Requirements to NL-25-0255 Proposed Technical Specification Bases Changes (Mark-up) - For Information Only

RCS Pressure SL B 2.1.2 (continued)

HATCH UNIT 1 B 2.0-5 REVISION 70 B 2.0 SAFETY LIMITS (SLs)

B 2.1.2 Reactor Coolant System (RCS) Pressure SL BASES BACKGROUND The SL on reactor steam dome pressure protects the RCS against overpressurization. In the event of fuel cladding failure, fission products are released into the reactor coolant. The RCS then serves as the primary barrier in preventing the release of fission products into the atmosphere. Establishing an upper limit on reactor steam dome pressure ensures continued RCS integrity. Per 10 CFR 50, Appendix A, GDC 14, "Reactor Coolant Pressure Boundary," and GDC 15, "Reactor Coolant System Design" (Ref. 1), the reactor coolant pressure boundary (RCPB) shall be designed with sufficient margin to ensure that the design conditions are not exceeded during normal operation and anticipated operational occurrences (AOOs).

During normal operation and AOOs, RCS pressure is limited from exceeding the design pressure by more than 10%, in accordance with Section III of the ASME Code (Ref. 2). To ensure system integrity, all RCS components are hydrostatically tested at 125% of design pressure, in accordance with ASME Code requirements, prior to initial operation when there is no fuel in the core. Any further hydrostatic testing with fuel in the core may be done under LCO 3.10.1, "Inservice Leak and Hydrostatic Testing Operation."

Following inception of unit operation, RCS components shall be pressure tested in accordance with the requirements of ASME Code,Section XI (Ref. 3).

Overpressurization of the RCS could result in a breach of the RCPB, reducing the number of protective barriers designed to prevent radioactive doses from exceeding the limits specified in 10 CFR 50.67 (Ref. 4). If this occurred in conjunction with a fuel cladding failure, fission products could enter the containment atmosphere.

APPLICABLE The Overpressure Protection System RCS safety/relief valves and the SAFETY ANALYSES Reactor Protection System Reactor Vessel Steam Dome Pressure -

High Function have settings established to ensure that the RCS pressure SL will not be exceeded.

The RCS pressure SL has been selected such that it is at a pressure below which it can be shown that the integrity of the system is not endangered. The reactor pressure vessel is designed to Section III of the ASME Boiler and Pressure Vessel Code, 1965 Edition, including

Control Rod Scram Times B 3.1.4 (continued)

HATCH UNIT 1 B 3.1-20 REVISION 0 BASES APPLICABLE "AVERAGE PLANAR LINEAR HEAT GENERATION RATE SAFETY ANALYSES (APLHGR)"], which ensure that no fuel damage will occur if these (continued) limits are not exceeded. Above 800 psig, the scram function is designed to insert negative reactivity at a rate fast enough to prevent the actual MCPR from becoming less than the MCPR SL, during the analyzed limiting power transient. Below 800 psig, the scram function is assumed to perform during the control rod drop accident (Ref. 5) and, therefore, also provides protection against violating fuel damage limits during reactivity insertion accidents (see Bases for LCO 3.1.6, "Rod Pattern Control"). For the reactor vessel overpressure protection analysis, the scram function, along with the Overpressure Protection Systemsafety/relief valves, ensures that the peak vessel pressure is maintained within the applicable ASME Code limits.

Control rod scram times satisfy Criterion 3 of the NRC Policy Statement (Ref. 8).

LCO The scram times specified in Table 3.1.4-1 (in the accompanying LCO) are required to ensure that the scram reactivity assumed in the DBA and transient analysis is met (Ref. 6). To account for single failures and "slow" scramming control rods, the scram times specified in Table 3.1.4-1 are faster than those assumed in the design basis analysis. The scram times have a margin that allows up to approximately 7% of the control rods (e.g., 137 x 7% 10) to have scram times exceeding the specified limits (i.e., "slow" control rods) assuming a single stuck control rod (as allowed by LCO 3.1.3, "Control Rod OPERABILITY") and an additional control rod failing to scram per the single failure criterion. The scram times are specified as a function of reactor steam dome pressure to account for the pressure dependence of the scram times. The scram times are specified relative to measurements based on reed switch positions, which provide the control rod position indication. The reed switch closes ("pickup") when the index tube passes a specific location and then opens ("dropout") as the index tube travels upward. Verification of the specified scram times in Table 3.1.4-1 is accomplished through measurement of the "dropout" times. To ensure that local scram reactivity rates are maintained within acceptable limits, no more than two of the allowed "slow" control rods may occupy adjacent locations.

Table 3.1.4-1 is modified by two Notes, which state that control rods with scram times not within the limits of the Table are considered "slow" and that control rods with scram times > 7 seconds are considered inoperable as required by SR 3.1.3.4.

RPS Instrumentation B 3.3.1.1 (continued)

HATCH UNIT 1 B 3.3-9 REVISION 16 BASES APPLICABLE 2.c. Average Power Range Monitor Neutron Flux - High (continued)

SAFETY ANALYSES, LCO, and Flux - High Function is assumed to terminate the main steam isolation APPLICABILITY valve (MSIV) closure event and, along with the Overpressure Protection System safety/relief valves (S/RVs), limits the peak reactor pressure vessel (RPV) pressure to less than the ASME Code limits.

The control rod drop accident (CRDA) analysis (Ref. 7) takes credit for the Average Power Range Monitor Neutron Flux - High Function to terminate the CRDA.

The Allowable Value is based on the Analytical Limit assumed in the CRDA analyses.

The Average Power Range Monitor Neutron Flux - High Function is required to be OPERABLE in MODE 1 where the potential consequences of the analyzed transients could result in the SLs (e.g., MCPR and RCS pressure) being exceeded. Although the Average Power Range Monitor Neutron Flux - High Function is assumed in the CRDA analysis, which is applicable in MODE 2, the Average Power Range Monitor Neutron Flux - High (Setdown)

Function conservatively bounds the assumed trip and, together with the assumed IRM trips, provides adequate protection. Therefore, the Average Power Range Monitor Neutron Flux - High Function is not required in MODE 2.

2.d. Average Power Range Monitor - Inop This Function (Inop) provides assurance that the minimum number of APRM channels is OPERABLE.

For any APRM channel, any time: 1) its mode switch is in any position other than "Operate," 2) an APRM module is unplugged, or

3) the automatic self-test system detects a critical fault with the APRM channel, an Inop trip signal is sent to all four voter channels. Inop trips from two or more unbypassed APRM channels result in a trip output from all four voter channels to their associated trip system.

This Function was not specifically credited in the accident analysis, but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.

There is no Allowable Value for this Function.

This Function is required to be OPERABLE in the MODES where the APRM Functions are required.

RPS Instrumentation B 3.3.1.1 (continued)

HATCH UNIT 1 B 3.3-12 REVISION 16 BASES APPLICABLE

3. Reactor Vessel Steam Dome Pressure - High (continued)

SAFETY ANALYSES, LCO, and along with the Overpressure Protection System S/RVs, limits the peak APPLICABILITY RPV pressure to less than the ASME Section III Code limits.

High reactor pressure signals are initiated from four pressure transmitters that sense reactor pressure. The Reactor Vessel Steam Dome Pressure - High Allowable Value is chosen to provide a sufficient margin to the ASME Section III Code limits during the event.

Four channels of Reactor Vessel Steam Dome Pressure - High Function, with two channels in each trip system arranged in a one-out-of-two logic, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. The Function is required to be OPERABLE in MODES 1 and 2 when the RCS is pressurized and the potential for pressure increase exists.

4. Reactor Vessel Water Level - Low, Level 3 Low RPV water level indicates the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, a reactor scram is initiated at Level 3 to substantially reduce the heat generated in the fuel from fission. The Reactor Vessel Water Level - Low, Level 3 Function is assumed in the analysis of the recirculation line break (Ref. 3). The reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the Emergency Core Cooling Systems (ECCS),

ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.

Reactor Vessel Water Level - Low, Level 3 signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel.

Four channels of Reactor Vessel Water Level - Low, Level 3 Function, with two channels in each trip system arranged in a one-out-of-two logic, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal.

The Reactor Vessel Water Level - Low, Level 3 Allowable Value is selected to ensure that: (a) during normal operation the steam dryer skirt is not uncovered (this protects available recirculation pump net

RPS Instrumentation B 3.3.1.1 (continued)

HATCH UNIT 1 B 3.3-13 REVISION 91 BASES APPLICABLE

4. Reactor Vessel Water Level - Low, Level 3 (continued)

SAFETY ANALYSES, LCO, and positive suction head (NPSH) from significant carryunder) and, APPLICABILITY (b) for transients involving loss of all normal feedwater flow, initiation of the low pressure ECCS subsystems at Reactor Vessel Water - Low Low Low, Level 1 will not be required.

The Function is required in MODES 1 and 2 where considerable energy exists in the RCS resulting in the limiting transients and accidents. ECCS initiations at Reactor Vessel Water Level - Low Low, Level 2 and Low Low Low, Level 1 provide sufficient protection for level transients in all other MODES.

5. Main Steam Isolation Valve - Closure MSIV closure results in loss of the main turbine and the condenser as a heat sink for the nuclear steam supply system and indicates a need to shut down the reactor to reduce heat generation. Therefore, a reactor scram is initiated on a Main Steam Isolation Valve - Closure signal before the MSIVs are completely closed in anticipation of the complete loss of the normal heat sink and subsequent overpressurization transient. However, for the overpressurization protection analysis of Reference 4, the Average Power Range Monitor Neutron Flux - High Function, along with the Overpressure Protection SystemS/RVs, limits the peak RPV pressure to less than the ASME Code limits. That is, the direct scram on position switches for MSIV closure events is not assumed in the overpressurization analysis.

Additionally, MSIV closure is assumed in the transients analyzed in Reference 2 (e.g., low steam line pressure, manual closure of MSIVs, high steam line flow).

The reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the ECCS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. The reactor scram resulting from an MSIV closure due to an isolation on low steam line pressure also ensures THERMAL POWER is 24% RTP before reactor steam dome pressure decreases below 685 psig to conserve Reactor Core SL 2.1.1.1.

MSIV closure signals are initiated from position switches located on each of the eight MSIVs. Each MSIV has two position switches; one inputs to RPS trip system A while the other inputs to RPS trip system B. Thus, each RPS trip system receives an input from eight Main Steam Isolation Valve - Closure channels, each consisting of one position switch. The logic for the Main Steam Isolation Valve -

Closure Function is arranged such that either the inboard or outboard valve on three or more of the main steam lines must close in order for a scram to occur. In addition, certain combinations of valves closed in two lines will result in a half-scram.

ATWS-RPT Instrumentation B 3.3.4.2 (continued)

HATCH UNIT 1 B 3.3-87 REVISION 22 BASES APPLICABLE b.

Reactor Steam Dome Pressure - High (continued)

SAFETY ANALYSES, LCO, and which could potentially result in fuel failure and APPLICABILITY overpressurization. The Reactor Steam Dome Pressure -

High Function initiates an RPT for transients that result in a pressure increase, counteracting the pressure increase by rapidly reducing core power generation. For the overpressurization event, the RPT aids in the termination of the ATWS event and, along with the Overpressure Protection Systemsafety/relief valves, limits the peak RPV pressure to less than the ASME Section III Code limits.

The Reactor Steam Dome Pressure - High signals are initiated from four pressure transmitters that monitor reactor steam dome pressure. Four channels of Reactor Steam Dome Pressure - High, with two channels in each trip system, are available and are required to be OPERABLE to ensure that no single instrument failure can preclude an ATWS-RPT from this Function on a valid signal. The Reactor Steam Dome Pressure - High Allowable Value is chosen to provide an adequate margin to the ASME Section III Code limits.

ACTIONS A Note has been provided to modify the ACTIONS related to ATWS-RPT instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable ATWS-RPT instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable ATWS-RPT instrumentation channel.

A.1 and A.2 With one or more channels inoperable, but with ATWS-RPT capability for each Function maintained (refer to Required Actions B.1 and C.1 Bases), the ATWS-RPT System is capable of performing the intended function. However, the reliability and redundancy of the ATWS-RPT instrumentation is reduced, such that a single failure in the remaining

LLS Instrumentation B 3.3.6.3 (continued)

HATCH UNIT 1 B 3.3-187 REVISION 124 BASES SURVEILLANCE SR 3.3.6.3.2, SR 3.3.6.3.3, and SR 3.3.6.3.4 (continued)

REQUIREMENTS function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.

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

A portion of the S/RV tailpipe pressure switch instrument channels are located inside the primary containment. The Note for SR 3.3.6.3.3, "Only required to be performed prior to entering MODE 2 during each scheduled outage > 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months when entry is made into primary containment," is based on the location of these instruments, ALARA considerations, and compatibility with the Completion Time of the associated Required Action (Required Action B.1).

Clarification for SR 3.3.6.3.3 Note:

a.

Outage duration is measured from the time the generator is removed from the grid to the time the generator is tied to the grid, i.e., breaker-to-breaker.

b.

Scheduled outage is defined as either a refueling outage or an outage for which at least a 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months period exists between discovery of an off-normal condition requiring shutdown and a corresponding change in power level to shut the unit down.

SR 3.3.6.3.5 CHANNEL CALIBRATION is a complete check of the instrument loop and sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations, consistent with the plant specific setpoint methodology.

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

SR 3.3.6.3.6 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required actuation logic for a specified channel.

The system functional testing performed in LCO 3.4.3, "Overpressure Protection System (OPS)Safety/Relief Valves (S/RVs)" and LCO 3.6.1.6, "Low-Low Set (LLS) Safety/Relief Valves (S/RVs)," for S/RVs overlaps this test to provide complete testing of the assumed safety

OPSS/RVs B 3.4.3 (continued)

HATCH UNIT 1 B 3.4-10 REVISION 79 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.3 Overpressure Protection System (OPS)Safety/Relief Valves (S/RVs)

BASES BACKGROUND The Overpressure Protection System (OPS) prevents overpressurization of the nuclear system by discharging reactor steam to the suppression pool. This action protects the reactor coolant pressure boundary (RCPB) from failure which could result in the release of fission products (Ref. 1).

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Ref. 2) requires the reactor pressure vessel be protected from overpressure during upset conditions by self-actuated safety valves. As part of the nuclear pressure relief system, tThe size and number of safety/relief valves (S/RVs) are selected such that peak pressure in the nuclear system will not exceed the ASME Code limits for the reactor coolant pressure boundary (RCPB).

The S/RVs are located on the main steam lines between the reactor vessel and the first isolation valve within the drywell. Each S/RV discharges steam through a discharge line to a point below the minimum water level in the suppression pool.

The S/RVs can actuate by either of two modes: the safety mode or the relief mode.

In the safety mode, the spring loaded pilot valve opens when steam pressure at the valve inlet expands the bellows to the point that the bellows force overcomes the force holding the pilot valve closed.

Opening the pilot valve allows steam to pass to the second stage operating piston which causes the second stage disc to open. This vents the chamber over the main valve disc to the downstream side of the valve, which causes a pressure differential to develop across the main valve piston and opens the main valve. The safety mode is credited for overpressure protection.This satisfies the Code requirement.

In the relief mode (or power actuated mode of operation), a pneumatic piston or cylinder and mechanical linkage assembly are used to open the valve by overcoming the spring force, even with the valve inlet pressure equal to 0 psig. The pneumatic operator is arranged so that its malfunction will not prevent the valve disk from lifting if steam inlet pressure reaches the spring lift set pressures. In the relief mode, valves may be opened manually or automatically at the selected preset pressure. S/RVs operating in relief mode are not credited for overpressure protection.

OPSS/RVs B 3.4.3 (continued)

HATCH UNIT 1 B 3.4-11 REVISION 13 BASES BACKGROUND (continued)

Some of the S/RVs operating in relief mode also provide the low-low set relief function, specified in LCO 3.6.1.6, Low-Low Set (LLS)

Valves, and the Automatic Depressurization System function, specified in LCO 3.5.1, ECCS - Operating. The instrumentation associated with the low-low set relief function is discussed in the Bases for LCO 3.3.6.3, Low-Low Set (LLS) Instrumentation, and instrumentation for the ADS function is discussed in LCO 3.3.5.1, Emergency Core Cooling System (ECCS) Instrumentation.

Each S/RV discharges steam through a discharge line to a point below the minimum water level in the suppression pool. The S/RVs that provide the relief mode are the low-low set (LLS) valves and the Automatic Depressurization System (ADS) valves. The LLS requirements are specified in LCO 3.6.1.6, "Low-Low Set (LLS)

Valves," and the ADS requirements are specified in LCO 3.5.1, "ECCS - Operating."

APPLICABLE The OPS overpressure protection system must accommodate the SAFETY ANALYSES most severe pressurization transient. Evaluations have determined that the most severe transient is the closure of all main steam isolation valves (MSIVs), followed by reactor scram on high neutron flux (i.e., failure of the direct scram associated with MSIV position)

(Ref. 1). The S/RV discharge piping is designed to accommodate forces resulting from the relief action including interactions with the suppression pool and is supported for reactions due to flow at maximum S/RV discharge capacity. For the purpose of tThe overpressure protection analyses, (Ref. 1) assume 10 ten of eleven11 S/RVs are assumed to operate in the safety mode of operation. The analysis results demonstrate that the OPSdesign S/RV capacity is capable of maintaining reactor pressure below the ASME Code limit of 110% of vessel design pressure (110% x 1250 psig = 1375 psig).

This LCO helps to ensure that the acceptance limit of 1375 psig is met during the dDesign bBasis eEvent.

From an overpressure standpoint, the design basis events are bounded by the MSIV closure with flux scram event described above.

Reference 42 discusses additional events that are expected to actuate the S/RVs.

The OPS satisfies S/RVs satisfy Criterion 3 of the NRC Policy Statement (Ref. 3).

OPSS/RVs B 3.4.3 (continued)

HATCH UNIT 1 B 3.4-12 REVISION 96 BASES (continued)

LCO The OPS is OPERABLE when it can ensure the ASME Code limit on peak reactor pressure, as stated in Safety Limit 2.1.2, will be protected using the safety modefunction of the10 of 11 S/RVs. are required to be OPERABLE to satisfy the assumptions of the safety analysis (Refs. 1, 2, and 4), although margins to the ASME Vessel Overpressure Limit are substantial. The requirements of this LCO are applicable only toThe OPERABILITY of the OPS is only dependent on the abilitycapability of the S/RVs to mechanically open to relieve excess pressure and maintain reactor pressure below Safety Limit 2.1.2, and may credit less than the full complement of installed S/RVs.when the lift setpoint is exceeded (safety function).

The S/RV setpoints are established to ensure that the ASME Code limit on peak reactor pressure is satisfied. The ASME Code specifications require the lowest safety valve to be setsetpoint to be at or below vessel design pressure (1250 psig) and the highest safety valve to be set so that the total accumulated pressure does not exceed 110% of the design pressure for overpressureization conditions. The transient evaluations in the FSARReference 4 are based on these setpoints, but also include the additional uncertainties of +/- 3% of the nominal setpoint drift to provide an added degree of conservatism.

An inoperable OPSOperation with fewer valves OPERABLE than specified, or with setpoints outside the ASME limits, could result in a more severe reactor response to a transient than predicted, possibly resulting in Safety Limit 2.1.2.the ASME Code limit on reactor pressure being exceeded.

APPLICABILITY In MODES 1, 2, and 3, the OPS10 of 11 S/RVs must be OPERABLE, since there may be considerable energy may be in the reactor core and the limiting design basis transients are assumed to occur in these MODES. The OPSS/RVs may be required to provide pressure relief to discharge energy from the core until such time that the Residual Heat Removal (RHR) System is capable of dissipating the core heat.

In MODE 4, decay heat is low enough for the RHR System to provide adequate cooling, and reactor pressure is low enough that the overpressure limit is unlikely to be approached by assumed operational transients or accidents. In MODE 5, the reactor vessel head is unbolted or removed and the reactor is at atmospheric pressure. The OPSS/RV function is not needed during these conditions.

OPSS/RVs B 3.4.3 (continued)

HATCH UNIT 1 B 3.4-13 REVISION 96 BASES (continued)

ACTIONS A.1 and A.2 With 1 SR/V inoperable, no action is required, because an analysis demonstrated that the remaining 10 SR/Vs are capable of providing the necessary overpressure protection. (See Ref. 5.)If the OPS is inoperable,With two or more S/RVs inoperable, a transient may result in the violation of the ASME Code limit on reactor pressure. The plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.3.1 REQUIREMENTS This Surveillance verifies that the S/RVs credited by the OPS have their as-left settings within 1% of the nominal opening pressure setpoint. The OPS may credit less than the full complement of installed S/RVs and the SR only applies to those S/RVs required to meet the LCO. The verification of the S/RV as-left settings is performed in accordance with the requires that the S/RVs will open at the pressures assumed in the safety analysis of Reference 5. The demonstration of the S/RV safety lift settings must be performed during shutdown, since this is a bench test, to be done in accordance with the INSERVICE TESTING PROGRAM. The nominal setpointlift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures. The S/RVs setpoint is +/- 3% for OPERABILITY; however, the valves are reset to +/- 1% during the Surveillance to allow for drift.

The Frequency of this SR is in accordance with the INSERVICE TESTING PROGRAM.

SR 3.4.3.2 This Surveillance verifies that the as-found lift pressures of the S/RVs credited by the OPS are consistent with the assumptions of the overpressure analysis. The OPS may credit less than the full complement of installed S/RVs and the SR only applies to those S/RVs required to meet the LCO. The measurement of the S/RV lift pressures must be performed in accordance with the INSERVICE TESTING PROGRAM. The OPS S/RV lift pressures are specified in the COLR.

OPSS/RVs B 3.4.3 HATCH UNIT 1 B 3.4-14 REVISION BASES (continued)

REFERENCES

1.

FSAR, Appendix M.

2.

ASME Boiler and Pressure Vessel Code,Section III.

32.

NRC No.93-102, Final Policy Statement on Technical Specification Improvements, July 23, 1993.Unit 2 FSAR, Chapter 15.

43.

FSAR, Section 15.NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

45.

NEDC-32041P, "Safety Review for Edwin I. Hatch Nuclear Power Plant Units 1 and 2 Updated Safety/Relief Valve Performance Requirements," April 1996.

56.

GEH Report NEDC-34126P, Rev. 0, Edwin I. Hatch Nuclear Power Plant Units 1 and 2 Safety/Relief Valve Setpoint Increase, March 2024.

Reactor Steam Dome Pressure B 3.4.10 (continued)

HATCH UNIT 1 B 3.4-53 REVISION 110 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.10 Reactor Steam Dome Pressure BASES BACKGROUND The reactor steam dome pressure is an assumed value in the determination of compliance with reactor pressure vessel overpressure protection criteria and is also an assumed initial condition of design basis accidents and transients.

APPLICABLE The reactor steam dome pressure of 1058 psig is an initial condition SAFETY ANALYSES of the vessel overpressure protection analysis of Reference 1. This analysis assumes an initial maximum reactor steam dome pressure and evaluates the response of the Overpressure Protection Systempressure relief system, primarily the safety/relief valves, during the limiting pressurization transient. The determination of compliance with the overpressure criteria is dependent on the initial reactor steam dome pressure; therefore, the limit on this pressure ensures that the assumptions of the overpressure protection analysis are conserved.

Reference 2 also assumes an initial reactor steam dome pressure for the analysis of design basis accidents and transients used to determine the limits for fuel cladding integrity [see Bases for LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)"] and 1%

cladding circumferential strain [see Bases for LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)"].

Reactor steam dome pressure satisfies the requirements of Criterion 2 of the NRC Policy Statement (Ref. 3).

LCO The specified reactor steam dome pressure limit of 1058 psig ensures the plant is operated within the assumptions of the overpressure protection analysis. Operation above the limit may result in a response more severe than analyzed.

APPLICABILITY In MODES 1 and 2, the reactor steam dome pressure is required to be less than or equal to the limit. In these MODES, the reactor may be generating significant steam and events which may challenge the overpressure limits are possible.

LLS Valves B 3.6.1.6 (continued)

HATCH UNIT 1 B 3.6-32 REVISION 79 B 3.6 CONTAINMENT SYSTEMS B 3.6.1.6 Low-Low Set (LLS) Valves BASES BACKGROUND The safety/relief valves (S/RVs) can actuate in either the safety mode as part of the Overpressure Protection System, the Automatic Depressurization System mode, or the LLS mode. In the LLS mode (or power actuated mode of operation), a pneumatic diaphragm and stem assembly overcome the spring force and open the second stage disc. As in the safety mode, opening the second stage disc allows a differential pressure to develop across the main valve piston and opens the main valve. The main valve can stay open with valve inlet steam pressure as low as 50 psig. Below this pressure, steam pressure may not be sufficient to hold the main valve open against the spring force of the main stage spring. The pneumatic operator is arranged so that its malfunction will not prevent the valve disk from lifting if steam inlet pressure exceeds the safety mode pressure setpoints.

Four of the S/RVs are equipped to provide the LLS function. The LLS logic causes the LLS valves to be opened at a lower pressure than the relief or safety mode pressure setpoints and stay open longer, so that reopening more than one S/RV is prevented on subsequent actuations.

Therefore, the LLS function prevents excessive short duration S/RV cycles with valve actuation at the relief setpoint (Ref. 1).

Each S/RV discharges steam through a discharge line and quencher to a location near the bottom of the suppression pool, which causes a load on the suppression pool wall. Actuation at lower reactor pressure results in a lower load.

APPLICABLE The LLS relief mode functions to ensure that the containment SAFETY ANALYSES design basis is met (Ref. 1). In other words, multiple simultaneous openings of S/RVs (following the initial opening), and the corresponding higher loads, are avoided. The safety analysis demonstrates that the LLS functions to avoid the induced thrust loads on the S/RV discharge line resulting from "subsequent actuations" of the S/RV during Design Basis Accidents (DBAs). Furthermore, the LLS function justifies the primary containment analysis assumption that simultaneous S/RV openings occur only on the initial actuation for DBAs. Even though four S/RVs are designated for the LLS function, all four LLS S/RVs do not operate in any DBA analysis. Thus, operation with three of four LLS S/RVs OPERABLE is acceptable (Ref. 4).

RCS Pressure SL B 2.1.2 (continued)

HATCH UNIT 2 B 2.0-5 REVISION 75 B 2.0 SAFETY LIMITS (SLs)

B 2.1.2 Reactor Coolant System (RCS) Pressure SL BASES BACKGROUND The SL on reactor steam dome pressure protects the RCS against overpressurization. In the event of fuel cladding failure, fission products are released into the reactor coolant. The RCS then serves as the primary barrier in preventing the release of fission products into the atmosphere. Establishing an upper limit on reactor steam dome pressure ensures continued RCS integrity. Per 10 CFR 50, Appendix A, GDC 14, "Reactor Coolant Pressure Boundary," and GDC 15, "Reactor Coolant System Design" (Ref. 1), the reactor coolant pressure boundary (RCPB) shall be designed with sufficient margin to ensure that the design conditions are not exceeded during normal operation and anticipated operational occurrences (AOOs).

During normal operation and AOOs, RCS pressure is limited from exceeding the design pressure by more than 10%, in accordance with Section III of the ASME Code (Ref. 2). To ensure system integrity, all RCS components are hydrostatically tested at 125% of design pressure, in accordance with ASME Code requirements, prior to initial operation when there is no fuel in the core. Any further hydrostatic testing with fuel in the core may be done under LCO 3.10.1, "Inservice Leak and Hydrostatic Testing Operation."

Following inception of unit operation, RCS components shall be pressure tested in accordance with the requirements of ASME Code,Section XI (Ref. 3).

Overpressurization of the RCS could result in a breach of the RCPB, reducing the number of protective barriers designed to prevent radioactive doses from exceeding the limits specified in 10 CFR 50.67 (Ref. 4). If this occurred in conjunction with a fuel cladding failure, fission products could enter the containment atmosphere.

APPLICABLE The Overpressure Protection System RCS safety/relief valves and the SAFETY ANALYSES Reactor Protection System Reactor Vessel Steam Dome Pressure -

High Function have settings established to ensure that the RCS pressure SL will not be exceeded.

The RCS pressure SL has been selected such that it is at a pressure below which it can be shown that the integrity of the system is not endangered. The reactor pressure vessel is designed to Section III of the ASME, Boiler and Pressure Vessel Code, 1968 Edition, including

Control Rod Scram Times B 3.1.4 (continued)

HATCH UNIT 2 B 3.1-20 REVISION 0 BASES APPLICABLE "AVERAGE PLANAR LINEAR HEAT GENERATION RATE SAFETY ANALYSES (APLHGR)"), which ensure that no fuel damage will occur if these (continued) limits are not exceeded. Above 800 psig, the scram function is designed to insert negative reactivity at a rate fast enough to prevent the actual MCPR from becoming less than the MCPR SL, during the analyzed limiting power transient. Below 800 psig, the scram function is assumed to perform during the control rod drop accident (Ref. 5) and, therefore, also provides protection against violating fuel damage limits during reactivity insertion accidents (see Bases for LCO 3.1.6, "Rod Pattern Control"). For the reactor vessel overpressure protection analysis, the scram function, along with the Overpressure Protection Systemsafety/relief valves, ensures that the peak vessel pressure is maintained within the applicable ASME Code limits.

Control rod scram times satisfy Criterion 3 of the NRC Policy Statement (Ref. 8).

LCO The scram times specified in Table 3.1.4-1 (in the accompanying LCO) are required to ensure that the scram reactivity assumed in the DBA and transient analysis is met (Ref. 6). To account for single failures and "slow" scramming control rods, the scram times specified in Table 3.1.4-1 are faster than those assumed in the design basis analysis. The scram times have a margin that allows up to approximately 7% of the control rods (e.g., 137 x 7% 10) to have scram times exceeding the specified limits (i.e., "slow" control rods) assuming a single stuck control rod (as allowed by LCO 3.1.3, "Control Rod OPERABILITY") and an additional control rod failing to scram per the single failure criterion. The scram times are specified as a function of reactor steam dome pressure to account for the pressure dependence of the scram times. The scram times are specified relative to measurements based on reed switch positions, which provide the control rod position indication. The reed switch closes ("pickup") when the index tube passes a specific location and then opens ("dropout") as the index tube travels upward. Verification of the specified scram times in Table 3.1.4-1 is accomplished through measurement of the "dropout" times. To ensure that local scram reactivity rates are maintained within acceptable limits, no more than two of the allowed "slow" control rods may occupy adjacent locations.

Table 3.1.4-1 is modified by two Notes, which state that control rods with scram times not within the limits of the Table are considered "slow" and that control rods with scram times > 7 seconds are considered inoperable as required by SR 3.1.3.4.

RPS Instrumentation B 3.3.1.1 (continued)

HATCH UNIT 2 B 3.3-9 REVISION 21 BASES APPLICABLE 2.c. Average Power Range Monitor Neutron Flux - High (continued)

SAFETY ANALYSES, LCO, and Flux - High Function is assumed to terminate the main steam isolation APPLICABILITY valve (MSIV) closure event and, along with the Overpressure Protection Systemsafety/relief valves (S/RVs), limits the peak reactor pressure vessel (RPV) pressure to less than the ASME Code limits.