ML18239A198

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Safety Evaluation - Vogtle Electric Generating Plant Units 3 and 4 (LAR 17-031)
ML18239A198
Person / Time
Site: Vogtle  Southern Nuclear icon.png
Issue date: 09/27/2018
From: Paul Kallan
NRC/NRO/DLSE/LB4
To:
City of Dalton, GA, Georgia Power Co, MEAG Power, Oglethorpe Power Corp, Southern Nuclear Operating Co
KALLAN P/415-2809
Shared Package
ML18239A192 List:
References
EPID L-2017-LLA-0097, LAR-17-031
Download: ML18239A198 (11)
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Text

SAFETY EVALUATION BY THE OFFICE OF NEW REACTORS RELATED TO AMENDMENT NOS. 144 AND 143 TO THE COMBINED LICENSE NOS. NPF-91 AND NPF-92, RESPECTIVELY 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 DALTON, GEORGIA VOGTLE ELECTRIC GENERATING PLANT UNITS 3 AND 4 DOCKET NOS.52-025 AND 52-026

1.0 INTRODUCTION

By letter dated September 8, 2017 (Agency Documents Access and Management System (ADAMS) Accession No. ML17251A458), the Southern Nuclear Operating Company (SNC) requested that the U.S. Nuclear Regulatory Commission (NRC) amend Vogtle Electric Generating Plant (VEGP) Units 3 and 4, Combined License (COL) Numbers NPF-91 and NPF-92, respectively.

License Amendment Request (LAR)17-031 would revise the Updated Final Safety Analysis Report (UFSAR) in the form of departures from the incorporated plant-specific Design Control Document (DCD) Tier 2 information, with corresponding changes to the associated COL Appendix A, Technical Specifications (TS). Specifically, the requested amendment proposes changes to (1) the design of the Protection and Safety Monitoring (PMS) system and associated changes to Chapter 15 transient and accident analyses, (2) changes to TS for the moderator temperature coefficient (MTC), and (3) additional changes to TS for power distributions, and the On-Line Power Distribution Monitoring System (OPDMS). The proposed changes to the PMS system and the crediting of trips in the Chapter 15 transient and accident analyses address issues caused by increased uncertainties in the ex-core nuclear instrumentation during mechanical shim (MSHIM) operations. The proposed changes to the TS for MTC modify the surveillance of MTC to address surveillance issues at beginning of life (BOL) and end of life (EOL). The proposed changes to TS for the power distribution and OPDMS update these TS to accurately reflect system capabilities.

2.0 REGULATORY EVALUATION

The LAR proposed changes to COL Appendix A and UFSAR Tier 2 information where the amendment would revise the licensing basis information for the design of the PMS automatic reactor trips and the crediting of PMS automatic reactor trips necessary to prevent exceeding fuel design limits including the power range high neutron flux (high setpoint), the power range high positive flux rate trip, the overpower T trip, and the overtemperature T trip. Also, changes to the TS for maintaining MTC and maintaining power distributions within the required absolute power generation limits.

The staff considered the following regulatory requirements in reviewing the LAR that included the proposed UFSAR changes.

Title 10 of the Code of Federal Regulations (10 CFR) Part 52, Appendix D,Section VIII.B.5.a allows an applicant or licensee who references this appendix to depart from Tier 2 information, without prior NRC approval, unless the proposed departure involves a change to or departure from Tier 1 information, Tier 2* information, or the TS, or requires a license amendment under paragraphs B.5.b or B.5.c of the section.

In 10 CFR 50.36, "Technical specifications," the NRC establishes regulatory requirements related to the content of TS by imposing limits, operating conditions, and other requirements upon reactor facility operation for the public health and safety. Per the Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors (July 22, 1993; 58 FR 39132), the TS are derived from the analyses and evaluations in the safety analysis report.

Pursuant to 10 CFR 50.36, TS are required to include items in the following five specific categories related to station operation: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements; (4) design features; and (5) administrative controls. This regulation is applicable since the licensees proposed change concerns the TS LCOs.

10 CFR 50.55a(h)(3), Protection and Safety Systems, requires compliance with Institute of Electrical and Electronics Engineers (IEEE) Standard (Std.) 603-1991, IEEE Standard Criteria for Safety Systems for Nuclear Power Generating Stations, and the correction sheet dated January 30, 1995. Clause 6.1 of IEEE Std. 603-1991, Automatic Control, requires, in part, that means shall be provided to automatically initiate and control all protective actions. Clause 6.8, Setpoints, requires, in part, that the device setpoint shall be determined using a documented methodology.

10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, Appendix A, General Design Criteria for Nuclear Power Plants, General Design Criterion (GDC) 10, Reactor design, requires that the reactor core and associated coolant, control, and protection systems be designed with appropriate margin to assure that specified acceptable fuel design limits (SADFLs) are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences (AOOs).

GDC 13, "Instrumentation and control," requires that instrumentation be provided to monitor variables and systems over their anticipated ranges for normal operation, for AOOs, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems.

GDC 15, Reactor coolant system design, requires that the reactor coolant system and associated auxiliary, control, and protection systems be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including AOOs.

GDC 20, Protection system functions, requires that the protection system be designed to (1) initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that SAFDLs are not exceeded as a result of AOOs and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.

GDC 25, Protection system requirements for reactivity control malfunctions, requires that the protection system be designed to assure that SAFDLs are not exceeded for any single malfunction of the reactivity control systems, such as accidental withdrawal (not ejection or dropout) of control rods.

3.0 TECHNICAL EVALUATION

OF THE REQUESTED CHANGES

3.1 TECHNICAL EVALUATION

OF PROPOSED CHANGES TO ADDRESS INCREASED UNCERTAINTIES IN POWER RANGE DETECTOR Sections 2.1 through 2.3 of Enclosure 1 to LAR-17-031 describe proposed changes to address the impacts of MSHIM operation. MSHIM operation increases the uncertainty in excore detector measurements due to the effect of rod shadowing (i.e., control rod movement during operation affects the flux observed by the excore detectors). This increased uncertainty affects the excore detector measurements of axial flux difference (AFD) and reactor power. To address these uncertainties, the licensee proposes (1) to change the AFD measurement from a core average AFD (CAAFD) to a weighted peripheral AFD (WPAFD), and (2) to eliminate dependence on the neutron flux high setting in the Chapter 15 transient and accident analyses. The licensee also proposes changes to TS to reflect these design changes.

3.1.1 AXIAL FLUX DIFFERENCE Section 2.1 of Enclosure 1 to LAR-17-031 describes a proposed change to the AFD measurement. The licensee describes the use of WPAFD rather than CAAFD because essentially all of the neutron flux measured by the excore detectors comes from a very limited number of fuel assemblies on the periphery of the reactor core. Staff agrees with this statement because it is consistent with previously observed detector weighting functions (Reference 4).

Additionally, APP-GLR-GW-137, Bases of Digital Overpower and Overtemperature Delta-T Reactor Trips, Revision 1 (ADAMS Accession No. ML110620129) describes that the Overpower T and Overtemperature T trips provide protection against peak fuel centerline temperature (FCT) and departure from nuclear boiling (DNBR) safety limits, respectively. The Overpower T and Overtemperature T trips account for the impact of the axial flux shape on FCT and DNBR by having penalty functions that reduce the trip setpoint based on the axial flux shape, as measured by the AFD. Accordingly, the change from CAAFD to WPAFD affects the penalty functions used in the Overpower T and Overtemperature T.

During an audit, staff observed (1) that the Overpower T AFD and Overtemperature T AFD penalty functions are regenerated for the WPAFD in 3D Final Acceptance Criteria (FAC) analysis and (2) that the WPAFD values are used directly in the development of the Overpower T AFD and Overtemperature T AFD penalty functions (ADAMS Accession No. ML18262A098). Because the 3D FAC analysis uses the WPAFD values when generating the Overpower T AFD and Overtemperature T AFD penalty functions, staff finds that the updated penalty functions continue to provide adequate protection against the FCT and DNBR safety limits. Therefore, staff finds the change from CAAFD to WPAFD continues to provide adequate margin to assure that SAFDLs are not exceeded during AOOs, and the change from CAAFD to WPAFD satisfies GDC 10 and 20.

3.1.2 USE OF OVERPOWER T AND POWER RANGE POSITIVE FLUX RATE TRIP FOR REACTOR PROTECTION DURING MSHIM OPERATION Section 2.2 of Enclosure 1 to LAR-17-031 describes that uncertainties in the power range detector indicated reactor power level cause the error allowances for the power range high neutron flux (high setpoint) trip to be exceeded, and that the current power range high neutron flux (high setpoint) trip does not provide adequate reactor protection such that the acceptance limit of 118 percent rated thermal power (RTP) is not exceeded. To address this issue, the licensee proposes to credit a combination of the power range high positive flux rate, Overpower T, and Overtemperature T trips to replace the power range high neutron flux (high setpoint) trip in the transient and accident analyses presented in Chapter 15 of the UFSAR. Accordingly, the licensee proposed UFSAR markups in the following sections to clarify that (1) the high power range neutron flux (high setpoint) provides backup to the power range high positive flux rate, Overpower T, and Overtemperature T trips and (2) that the 118 percent maximum power range neutron flux trip setpoint is no longer credited in the Chapter 15 analyses:

  • UFSAR, Tier 2, Subsection 7.2.1.1.2, Nuclear Overpower Trips
  • UFSAR, Tier 2, Subsection 7.2.1.1.3, Core Heat Removal Trips
  • UFSAR, Tier 2, Subsection 15.0.6, Protection and Safety Monitoring System Setpoints and Time Delays to Trip Assumed in Accident Analyses
  • UFSAR, Tier 2, Subsection 15.0.7, Instrumentation Drift and Calorimetric Errors, Power Range Neutron Flux
  • UFSAR, Tier 2, Subsection 15.0.11.2, LOFTRAN Computer Code
  • UFSAR, Tier 2, Subsection 15.0.16, References
  • UFSAR, Tier 2, Table 15.0-4a, Protection and Safety Monitoring System Setpoint and Time Delay Assumed in Accident Analyses
  • UFSAR, Tier 2, Table 15.0-5, Determination of Maximum Power Range Neutron Flux Channel Trip Setpoint, Based on Nominal Setpoint and Inherent Typical Instrumentation Uncertainties
  • UFSAR, Tier 2, Table 15.0-6, Plant Systems and Equipment Available for Transient and Accident Conditions
  • UFSAR, Tier 2, Section 15.1.2.2.2, Results (Feedwater System Malfunctions that Result in an Increase in Feedwater Flow)
  • UFSAR, Tier 2, Section 15.1.3.1, Identification of Causes and Accident Description (Excessive Increase in Secondary Steam Flow)
  • UFSAR, Tier 2, Section 15.1.4.1, Identification of Causes and Accident Description (Inadvertent Opening of a Steam Generator Relief or Safety Valve)
  • UFSAR, Tier 2, Section 15.4.2, Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power The licensee identified, in Section 2.2 of Enclosure 1 to LAR 17-031, that the Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power (URWP) event in UFSAR Section 15.4.2 is the only event in UFSAR Chapter 15 that currently credits the high power range neutron flux (high setpoint) trip. Accordingly, the licensee reanalyzed this event and provided markups to UFSAR, Tier 2, Section 15.4.2. The staff reviewed the current UFSAR, Tier 2, Chapter 15 (ADAMS Accession No. ML17172A218) to determine which events credit the high power range neutron flux (high setpoint) trip. Based on this review, staff agrees with the licensee that the URWP event is the only event that credits the high power range neutron flux (high setpoint) trip as the means of primary protection.

The staff evaluated the proposed markups to UFSAR, Tier 2, Section 15.4.2 which reflect reanalysis of the URWP event. Based on these markups, staff finds (1) the markups clearly reflect that the high power range neutron flux (high setpoint) trip is not credited, (2) the markups clearly reflect that the high positive flux rate trip, Overpower T trip, and Overtemperature T trip provide the necessary protection for this event, (3) there are no changes to the evaluation model outside of the addition of the power range high positive flux rate trip, (4) the results of the reanalysis show that the DNBR continues to remain above the safety limit for all cases, and (5) the results of the reanalysis show that the maximum reactor coolant system pressure continues to remain below 110 percent of the design pressure for all cases. In addition, during the audit, staff observed that the revised calculations supporting UFSAR Sections 15.4.2, and 15.4.8 (1) do not credit the high power range neutron flux (high setpoint) trip and (2) demonstrate that the combination of the power range high positive flux rate, Overpower T, and Overtemperature T trips provide adequate protection against the acceptance criteria for the URWP events and the Spectrum of Rod Cluster Control Assembly Ejection Accidents (Reference 6). Based on the markups provided in UFSAR, Tier 2, Chapters 7 and 15, and the results of the reanalysis of the URWP event staff finds that GDC 10, 15, 20, and 25 continue to be satisfied and the proposed markups are acceptable.

In addition to the UFSAR markups discussed above, the licensee proposes corresponding COL Appendix A changes:

  • Limiting Condition for Operation (LCO) 3.2.1, FQ ( ), as approximated by FC Q ( ) and W

FQ ( ), shall be within the limits specified in the Core Operating Limits Report (COLR)

  • LCO 3.7.1, Six Main Steam Safety Valves (MSSVs) shall be operable The proposed changes to LCO 3.2.1 removes the required actions to reduce the power range neutron flux (high) trip setpoint. Staff finds these changes acceptable because (1) the safety analyses show that the power range neutron flux (high setpoint) trip is not required to provide adequate protection, (2) the power range high positive flux rate trip setpoints are available to provide protection and are not required to be changed, and (3) the required actions to reduce the Overpower T trip setpoint are retained.

The proposed change to LCO 3.2.2 changes Required Action A.1.2.2 from reduce Power Range Neutron Flux - High to reduce Overpower T trip setpoint. Section 2.2 of Enclosure 1 to LAR 17-031 states that the overlap between the responses of the power range high positive flux rate trip and Overpower T trip with reduced setpoint are adequate to prevent exceeding fuel design limits for design basis events initiated within the applicability of LCO 3.2.2. Staff agrees with this statement because Required Action A.1.2.1 (which reduces the thermal power to less than 50 percent RTP) provides significant margin to fuel design limits such that there is reasonable assurance that the combination of the high positive flux rate trip and Overpower T trip (at a trip setpoint of less than or equal to 55 percent RTP) provides adequate protection of the fuel design limits.

The proposed change to LCO 3.7.1 changes Required Action A.2 from reduce Power Range Neutron Flux - High to reduce Overpower T trip setpoint. Section 2.2 of Enclosure 1 to LAR 17-031 states that the Overpower T trip provides margin to the 118 percent RTP acceptance limit as long as the setpoints are reduced to compensate for the number of operable MSSVs available. During the audit, staff observed that (1) analyses were performed to confirm that secondary side pressures remain below the analysis limit of 110 percent design pressure for cases analyzed with the digital Overpower T setpoint reduced to the Maximum Allowable Power specified in TS Table 3.7.1-1, (2) the analyses were performed with LOFTRAN (an NRC approved analysis code), (3) the analyses assume the Overpower T setpoint is set conservatively, (4) the calculation includes justification for the modeling of the MSSVs in LOFTRAN, and (5) the results of the analyses show that the peak secondary side pressure remains below 110 percent design pressure for all cases (Reference [6]). Based upon conservative analysis confirming that the Overpower T trip provides adequate secondary side protection, staff finds the proposed change to LCO 3.7.1 acceptable.

3.1.3 SURVEILLANCE TOLERANCE LIMITS Sections 2.1 and 2.2 of Enclosure 1 to LAR 17-031 provide the following proposed changes to TS 3.3.1:

  • SR 3.3.1.3 to change the tolerance on the T power calculation measurement
  • SR 3.3.1.4 to change the tolerance on the AFD measurement The proposed changes to SR 3.3.1.2 (1) clarify that the surveillance is applicable when reactor power is greater than or equal to 15 percent RTP and (2) increases the tolerance between the NI power measurement and calorimetric heat balance from 1 percent RTP to 5 percent RTP.

During the audit, staff observed that (1) an analysis was conducted to evaluate the instrumentation uncertainty for the excore Power Range Neutron Flux-High, Power Range Neutron Flux-Low, and High Positive Flux Rate reactor trip setpoints, (2) the proposed change to SR 3.3.1.2 is incorporated into the setpoint evaluations, and (3) the analysis showed margin between the setpoints and the safety analysis limits (Reference [6]). Because (1) the Power Range Neutron Flux-High reactor trip is not credited in the safety analysis, and therefore has no safety analysis limit, and (2) an analysis that incorporates the change to SR 3.3.1.2 shows that the Power Range Neutron Flux-Low and High Positive Rate trips continue to provide margin to the safety analysis limits, staff finds the proposed change to SR 3.3.1.2 acceptable.

The proposed changes to SRs 3.3.1.3 and 3.3.1.4 adjust the tolerances for the T power and AFD measurements, respectively. During the audit that occured July 26 - August 17, 2018, (ADAMS Accession No. ML18262A098), staff observed that (1) an analysis was conducted to evaluate the instrumentation uncertainty for the Overpower T and Overtemperature T reactor trip setpoints, (2) the proposed changes to SRs 3.3.1.3 and 3.3.1.4 are incorporated into the setpoint evaluations, and (3) the analysis showed margin between the setpoints and the safety analysis limits. Because an analysis that incorporates the changes to SRs 3.3.1.3 and 3.3.1.4 shows that the Overpower T and Overtemperature T reactor trip setpoints continue to provide margin to the safety analysis limits, staff finds the proposed changes to SRs 3.3.1.3 and 3.3.1.4 acceptable.

Additionally, staffs review confirmed that there is no proposed change to the setpoint calculation methodology for the reactor trips which was previously reviewed and approved by staff for the certified AP1000 design. Because the proposed changes are implemented using an approved setpoint calculation methodology and the results of these calculations show adequate margin to safety analysis limits, staff finds that the proposed changes continue to meet 10 CFR 50.55a(h)(3) and GDC 13.

3.2 TECHNICAL EVALUATION

OF PROPOSED CHANGES TO TECHNICAL SPECIFICATION FOR MTC Section 2.4 of Enclosure 1 to LAR 17-031 describes the proposed changes to TS 3.1.3, Moderator Temperature Coefficient with corresponding changes to TS 5.6.3, Core Operating Limits Report. The changes to TS 3.1.3 include modifications to the LCO 3.1.3, Required Action A.1 and the addition of a Note modifying SR 3.1.3.2; these changes impact MTC SRs at BOL and EOL, respectively.

The modification to LCO 3.1.3, Required Action A.1 changes Establish administrative withdrawal limits for control banks to maintain MTC within limit, to Restore MTC within limit.

Staff finds this change acceptable because it produces the same result (i.e., MTC is restored to be within the upper limit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or the plant is brought to a subcritical condition in accordance with LCO 3.1.3, Required Action B.1).

The addition of a Note to SR 3.1.3.2 provides a condition under which MTC surveillance near EOL would not be required. The licensee is proposing the addition of this Note to eliminate the potential disruption to normal plant operations that is associated with performing the MTC surveillance near EOL. To support the addition of this Note, the licensee references approved topical report WCAP-13749-P-A, Safety Evaluation Supporting the Condition Exemption of the Most Negative EOL Moderator Temperature Coefficient Measurement, dated March 31, 1997.

The approved topical report methodology allows replacement of the requirement to measure the MTC near EOL with a revised prediction of MTC at EOL.

The added Note to SR 3.3.1.2 states Not required to be performed provided applicable criteria in the COLR are satisfied. In addition, the following references were added to TS 5.6.3 as methodology for TS 3.1.3 :

WCAP-16045-P-A, Qualification of the Two-Dimensional Transport Code PARAGON, August 2004 (Westinghouse Proprietary) and WCAP-16045-NP-A, (Non- Proprietary),

WCAP-16045-P-A, Addendum 1-A, Qualification of the NEXUS Nuclear Data Methodology, August 2007 (Westinghouse Proprietary) and WCAP-16045-NP-A, Addendum 1-A, (Non-Proprietary), and WCAP-13749-P-A, dated March 1997 (Westinghouse Proprietary).

The staff safety evaluation for WCAP-13749-P-A has concluded that the methodology described therein is acceptable as a reference for proposed changes to relevant TS provided the following conditions: (1) only PHOENIX/ANC calculation methods are used for the individual plant analyses relevant to determinations for the EOL MTC plant methodology and (2) the predictive correction is reexamined if changes in core fuel designs or continued MTC calculation/measurement data show significant effect on the predictive correction.

The licensee identified, in Section 2.4 of Enclosure 1 to LAR 17-031, that (1) as described in UFSAR Subsection 4.3.3.2, the AP1000 plant core design calculations are performed using the PARAGON lattice code with NEXUS methodology, (2) Section 4.0, Conditions and Limitations of the NRC's safety evaluation for WCAP-16045-P-A, the NRC stated The PARAGON code can be used as a replacement for the PHOENIX-P lattice code, whenever the PHOENIX-P code is used in NRC approved methodologies, (3) Section 5.0, Conclusion, of the NRC's safety evaluation for WCAP-16045-P-A, Addendum 1-A states, The NRC staff has reviewed the TR submitted by Westinghouse and determined that the NEXUS/ANC code system is adequate to replace the PARAGON/ANC code system wherever the latter is used in NRC approved methodologies. The staff confirmed the licensee references to the safety evaluation for WCAP-16045-NP-A (ADAMS Accession No. ML042250322). Based on the use of an approved methodology to replace PARAGON, staff finds that Condition 1 for WCAP-13749 is satisfied.

Section 2.4 of Enclosure 1 to LAR 17-031, states that the licensee will confirm, prior to the use of the conditional exemption technique, that core design changes and MTC calculation and measurement data are confirmed to not show a significant effect on the predictive correction.

The licensee further states that if a significant effect is found, the use of the predictive correction will be re-examined. The WCAP-13749-P-A methodology also requires certain core performance criteria, such as startup physics tests and cycle reactivity measurements, to be met to allow exemption of the EOL MTC measurement. This requirement is captured in the licensee's proposed changes to SR 3.1.3.2. Because the licensee has (1) stated that the effects of core design changes will be verified relative to the use of the predictive correction and (2) the TS requires adherence to the core acceptance criteria of WCAP-13749-P-A, staff determined that Condition 2 for WCAP-13749-P-A is satisfied.

Because the licensee satisfied both conditions in WCAP-13749-P-A, staff finds the methodology applicable. Staff verified that the corresponding changes to TS 5.6.3 incorporate approved and applicable methodologies. Based on the use of approved methodologies, staff finds the proposed change to SR 3.1.3.2 and corresponding changes to TS 5.6.3 acceptable.

Accordingly, staff finds that the proposed changes continue to meet 10 CFR 50.36.

3.3 TECHNICAL EVALUATION

OF PROPOSED CHANGES TO TECHNICAL SPECIFICATION FOR POWER DISTRIBUTIONS AND OPDMS 3.3.1 POWER DISTRIBUTIONS Sections 2.5 and 2.6 of Enclosure 1 to LAR 17-031 describe the proposed changes to COL Appendix A, LCOs 3.2.1 and 3.2.2, respectively. The licensee proposes changes to address issues with (1) the applicability of LCOs 3.2.1 and 3.2.2 extending below 20 percent RTP where there is no mechanism to meet the SR, and (2) Notes in SRs 3.2.1.3, 3.2.1.4, and 3.2.2.2 inappropriately allowing for 31 days of operation with the OPDMS not functional before verification is required.

The licensee identified that the surveillance requirements associated with LCOs 3.2.1 and 3.2.2 utilize the fixed incore detectors which cannot adequately measure the power distribution at low power levels below approximately 20 percent RTP. Accordingly, the licensee proposed to change (1) the applicability of LCOs 3.2.1 and 3.2.2 to limit the applicability to powers greater than and equal to 25 percent RTP, and (2) change LCO 3.2.1 Required Action C.1 and LCO 3.2.2 Required Action B.1 to reduce thermal power to 25 percent RTP. Staff finds the proposed change acceptable because large thermal margin exists at the reduced powers as reflected in the frequency for SRs 3.2.1.1, 3.2.1.2, and 3.2.2.1 (i.e., these surveillances are required prior to thermal power exceeding 75 percent RTP).

The licensee identified that Notes modifying SRs 3.2.1.3, 3.2.1.4, and 3.2.2.2 inappropriately allow for 31 days of operation with the OPDMS not functional before verification is required.

Accordingly, the licensee proposed to change the Notes modifying SRs 3.2.1.3, 3.2.1.4, and 3.2.2.2 to require verification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after OPDMS not monitoring parameters. Staff finds the proposed changes acceptable because they result in more stringent surveillance requirements.

3.3.2 OPDMS Section 2.7 of Enclosure 1 to LAR 17-031 describes the proposed changes to COL Appendix A, LCO 3.2.5. The licensee proposes one editorial change and another change to increase the applicability of LCO 3.2.5. The proposed editorial change replaces Power Density with Heat Rate, in LCO 3.2.5. Staff finds the proposed change acceptable because it has no technical impact on LCO 3.2.5.

The licensee proposed to increase the range of applicability of LCO 3.2.5. Section 2.7 of to LAR 17-031 states that this change is proposed because the original assumption that the rapid power reduction system would be blocked if the OPDMS is not functioning is being removed, resulting in the possibility that fuel design limits can be exceeded following rapid power reduction system actuation at power levels between 25 percent RTP and 50 percent RTP. Accordingly, the licensee proposed to change the applicability of LCO 3.2.5 from MODE 1 with THERMAL POWER > 50 % RTP with OPDMS monitoring parameters a, b, and c, to MODE 1 with THERMAL POWER 25% RTP and with OPDMS monitoring parameters a, b, and c to ensure that the monitored parameters remain within the values assumed in the safety analyses at reduced powers. Additionally, the licensee proposed a corresponding change to LCO 3.2.5 Required Action B.1 to require a reduction in power to less than or equal to 25 percent RTP. Staff finds the proposed changes acceptable because they result in an increase in the range of applicability over which the OPDMS is monitoring parameters.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations in 10 CFR 50.91(b)(2), on August 17, 2018, the Georgia State official was consulted regarding the amendment. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20, Standards for Protection Against Radiation. The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite. Also, there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (Federal Register, 82 FR 49234, dated October 24, 2017).

Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Under 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

The staff has concluded, based on the considerations discussed in Section 3.0 that there is reasonable assurance that: (1) the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that 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. Therefore, the staff finds the changes proposed in this license amendment acceptable.

7.0 REFERENCES

1. Request for License Amendment (LAR 17-031): Effect of Rod Shadowing on Excore Power Range Detectors, Changes to Nuclear Overpower Reactor Trips, and Changes to Monitoring of Moderator Temperature Coefficient and Core Power Distribution, September 8, 2017 (ADAMS Accession No. ML17251A458).
2. Combined License NPF-91 for Vogtle Generating Plant Unit 3, Southern Nuclear Operating Company, (ADAMS Accession No. ML14100A106).
3. Combined License NPF-92 for Vogtle Electric Generating Plant Unit 4, Southern Nuclear Operating Company, (ADAMS Accession No. ML14100A135).
4. G. A. Ahn, N. Z. Cho and J. E. Kuh, Generation of Spatial Weighting Functions for Ex-Core Detectors by Adjoint Transport Calculation, Nuclear Technology, pp. 114-121, July 1993.
5. APP-GW-GLR-137, Bases of Digital Overpower and Overtemperature Delta-T Reactor Trips, Rev. 1," February 24, 2011 (ADAMS Accession No. ML110620129).
6. Report of Regulatory Audit for License Amendment Realted to Effect of Rod Shadowing on Excore Power Range Detectors, Changes to Nuclear Overpower Reactor Trips, and Changes to Monitoring of Moderator Temperature Coefficient and Core Power Distribution, September 20, 20180 (ADAMS Accession No. ML18262A098).
7. Vogtle Electric Generating Plant, Units 3 and 4 Updated Final Safety Analysis Report, Revision 6 and Tier 1, Revision 5, June 13, 2011 (ADAMS Accession No. ML17172A218).
8. WCAP-13749-P-A, Safety Evaluation Supporting the Conditional Exemption of the Most Negative EOL Moderator Temperature Coefficient Measurement, March 31, 1997.
9. WCAP-16045-NP-A, Qualification of Two-Dimensional Transport Code PARAGON, August 31, 2004 (ADAMS Accession No. ML042250322).
10. AP1000 Design Control Document, Revision 19, June 13, 2011 (ADAMS Accession No. ML11171A500).
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