ML16343A246
| ML16343A246 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 04/11/2017 |
| From: | Andrew Hon Plant Licensing Branch II |
| To: | William Gideon Duke Energy Progress |
| Wentzel M, NRR/DORL/LPL2-2, 415-6459 | |
| References | |
| CAC MF7602, CAC MF7603 | |
| Download: ML16343A246 (21) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 April 11, 2017 Mr. William R. Gideon, Vice President Brunswick Steam Electric Plant Duke Energy Progress, LLC 8470 River Rd., SE, M/C BNP001 Southport, NC 28461
SUBJECT:
BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENTS REGARDING REACTOR PROTECTION SYSTEM ELECTRICAL PROTECTION ASSEMBLY ELECTRIC POWER MONITORING SURVEILLANCE REQUIREMENTS 3.3.8.2.2 AND 3.3.8.2.3 (CAC NOS. MF7602 AND MF7603)
Dear Mr. Gideon:
The Commission has issued the enclosed Amendment Nos. 273 and 301 to Renewed Facility Operating License Nos. DPR-71 and DPR-62, respectively, for Brunswick Steam Electric Plant, Units 1 and 2. The amendments are in response to your application dated April 13, 2016, as supplemented by letter dated March 1, 2017. The amendments change the Technical Specifications by revising the Reactor Protection System electric power monitoring assembly Allowable Values for overvoltage and undervoltage contained within Surveillance Requirements 3.3.8.2.2 and 3.3.8.2.3.
A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register Notice.
Sincere~,
/J~:\
An~~n, Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-325 and 50-324
Enclosures:
- 1. Amendment No. 273 to DPR-71
- 2. Amendment No. 301 to DPR-62
- 3. Safety Evaluation cc w/enclosures: Distribution via Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY PROGRESS. LLC DOCKET NO. 50-325 BRUNSWICK STEAM ELECTRIC PLANT. UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 273 Renewed License No. DPR-71
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment filed by Duke Energy Progress, LLC, dated April 13, 2016, as supplemented by letter dated March 1, 2017, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
Enclosure 1
- 2. Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment; and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-71 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 273, are hereby incorporated in the license. Duke Energy Progress, LLC shall operate the facility in accordance with the Technical Specifications.
- 3. This license amendment is effective as of the date of its issuance and shall be implemented within 120 days.
FOR THE NUCLEAR REGULATORY COMMISSION
-~-21~
Benjamin G. Beasley, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Operating License and Technical Specifications Date of Issuance: Apr i 1 11 , 2 o1 7
ATTACHMENT TO LICENSE AMENDMENT NO. 273 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 1 RENEWED FACILITY OPERATING LICENSE NO. DPR-71 DOCKET NO. 50-325 Replace Page 6 of Renewed Facility Operating License No. DPR-71 with the attached Page 6.
Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Remove Pages Insert Pages 3.3-73 3.3-73 3.3-74 3.3-74
(c) Transition License Conditions
- 1. Before achieving full compliance with 10 CFR 50.48(c), as specified by 2. below, risk-informed changes to the licensee's fire protection program may not be made without prior NRC review and approval unless the change has been demonstrated to have no more than a minimal risk impact, as described in 2. above.
- 2. The licensee shall implement the modifications to its facility, as described in Table S-1, "Plant Modifications Committed," of Duke letter BSEP 14-0122, dated November 20, 2014, to complete the transition to full compliance with 10 CFR 50.48(c) by the startup of the second refueling outage for each unit after issuance of the safety evaluation. The licensee shall maintain appropriate compensatory measures in place until completion of these modifications.
- 3. The licensee shall complete all implementation items, except item 9, listed in LAR Attachment S, Table S-2, "Implementation Items," of Duke letter BSEP 14-0122, dated November 20, 2014, within 180 days after NRC approval unless the 1801h day falls within an outage window; then, in that case, completion of the implementation items, except item 9, shall occur no later than 60 days after startup from that particular outage. The licensee shall complete implementation of LAR Attachment S, Table S-2, Item 9, within 180 days after the startup of the second refueling outage for each unit after issuance of the safety evaluation.
C. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1) Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2923 megawatts thermal.
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 273, are hereby incorporated in the license. Duke Energy Progress, LLC shall operate the facility in accordance with the Technical Specifications.
For Surveillance Requirements (SRs) that are new in Amendment 203 to Renewed Facility Operating License DPR-71, the first performance is due at the end of the first surveillance interval that begins at implementation of Amendment 203. For SRs that existed prior to Amendment 203, including SRs with modified acceptance criteria and SRs whose frequency of Renewed License No. DPR-71 Amendment No. 273
RPS Electric Power Monitoring 3.3.8.2 ACTIONS (continued)
COMPLETION CONDITION REQUIRED ACTION TIME Immediately D. Required Action and D.1 Initiate action to fully insert associated Completion Time all insertable control rods of Condition A or B not met in core cells containing in MODE 3, 4, or 5 with any one or more fuel control rod withdrawn from a assemblies.
core cell containing one or more fuel assemblies.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.8.2.1 --------------------------------NOTE-------------------------------
Only required to be performed prior to entering MODE 2 from MODE 3 or 4, when in MODE 4 for 184 days
~ 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Perform CHANNEL FUNCTIONAL TEST.
24 months SR 3.3.8.2.2 Perform CHANNEL CALIBRATION for each RPS motor generator set electric power monitoring assembly. The Allowable Values shall be:
- a. Overvoltage .:::; 127 V.
- b. Undervoltage ~ 107 V.
- c. Underfrequency ~ 57.2 Hz.
(continued)
Brunswick Unit 1 3.3-73 Amendment No. 273
RPS Electric Power Monitoring 3.3.8.2 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY 24 months SR 3.3.8.2.3 Perform CHANNEL CALIBRATION for each RPS alternate power supply electric power monitoring assembly. The Allowable Values shall be:
- a. Overvoltage ::=:; 127 V.
- b. Undervoltage;:::: 107 V.
- c. Underfrequency;:::: 57.2 Hz.
24 months SR 3.3.8.2.4 Perform a system functional test.
Brunswick Unit 1 3.3-74 Amendment No. 273
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY PROGRESS. LLC DOCKET NO. 50-324 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 301 Renewed License No. DPR-62
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment filed by Duke Energy Progress, LLC, dated April 13, 2016, as supplemented by letter dated March 1, 2017, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
Enclosure 2
- 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment; and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-62 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 301, are hereby incorporated in the license. Duke Energy Progress, LLC shall operate the facility in accordance with the Technical Specifications.
- 3. This license amendment is effective as of the date of its issuance and shall be implemented within 120 days.
FOR THE NUCLEAR REGULATORY COMMISSION Benjamin G. Beasley, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Operating License and Technical Specifications Date of Issuance: April 11, 2017
ATTACHMENT TO LICENSE AMENDMENT NO. 301 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 2 FACILITY OPERATING LICENSE NO. DPR-62 DOCKET NO. 50-324 Replace Page 6 of Renewed Facility Operating License No. DPR-62 with the attached Page 6.
Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Remove Pages Insert Pages 3.3-73 3.3-73 3.3-74 3.3-74
(c) Transition License Conditions
- 1. Before achieving full compliance with 10 CFR 50.48(c), as specified by 2. below, risk-informed changes to the licensee's fire protection program may not be made without prior NRC review and approval unless the change has been demonstrated to have no more than a minimal risk impact, as described in 2. above.
- 2. The licensee shall implement the modifications to its facility, as described in Table S-1, "Plant Modifications Committed," of Duke letter BSEP 14-0122, dated November 20, 2014, to complete the transition to full compliance with 10 CFR 50.48(c) by the startup of the second refueling outage for each unit after issuance of the safety evaluation. The licensee shall maintain appropriate compensatory measures in place until completion of these modifications.
- 3. The licensee shall complete all implementation items, except Item 9, listed in LAR Attachment S, Table S-2, "Implementation Items," of Duke letter BSEP 14-0122, dated November 20, 2014, within 180 days after NRC approval unless the 1801h day falls within an outage window; then, in that case, completion of the implementation items, except item 9, shall occur no later than 60 days after startup from that particular outage. The licensee shall complete implementation of LAR Attachment S, Table S-2, Item 9, within 180 days after the startup of the second refueling outage for each unit after issuance of the safety evaluation.
C. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1) Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2923 megawatts (thermal).
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 301, are hereby incorporated in the license. Duke Energy Progress, LLC shall operate the facility in accordance with the Technical Specifications.
For Surveillance Requirements (SRs) that are new in Amendment 233 to Renewed Facility Operating License DPR-62, the first performance is due at the end of the first surveillance interval that begins at implementation of Amendment 233. For SRs that existed prior to Amendment 233, Renewed License No. DPR-62 Amendment No. 301
RPS Electric Power Monitoring 3.3.8.2 ACTIONS (continued)
COMPLETION CONDITION REQUIRED ACTION TIME Immediately D. Required Action and D.1 Initiate action to fully insert associated Completion Time all insertable control rods of Condition A or B not met in core cells containing in MODE 3, 4, or 5 with any one or more fuel control rod withdrawn from a assemblies.
core cell containing one or more fuel assemblies.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3. 8.2.1 -------------------------------NOTE--------------------------------
Only required to be performed prior to entering MODE 2 from MODE 3 or 4, when in MODE 4 for ~ 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
184 days Perform CHANNEL FUNCTIONAL TEST.
SR 3.3.8.2.2 Perform CHANNEL CALIBRATION for each RPS motor 24 months generator set electric power monitoring assembly. The Allowable Values shall be:
- a. Overvoltage::; 127 V.
- b. Undervoltage ~ 107 V.
- c. Underfrequency ~ 57.2 Hz.
(continued)
Brunswick Unit 2 3.3-73 Amendment No. 301
RPS Electric Power Monitoring 3.3.8.2 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.3.8.2.3 Perform CHANNEL CALIBRATION for each RPS alternate 24 months power supply electric power monitoring assembly. The Allowable Values shall be:
- a. Overvoltage::; 127 V.
- b. Undervoltage ~ 107 V.
- c. Underfrequency ~ 57.2 Hz.
SR 3.3.8.2.4 Perform a system functional test. 24 months Brunswick Unit 2 3.3-74 Amendment No. 301
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 273 AND 301 TO RENEWED FACILITY OPERATING LICENSES NOS. DPR-71 AND DPR-62 DUKE ENERGY PROGRESS, LLC BRUNSWICK STEAM ELECTRIC PLANT. UNITS 1 AND 2 DOCKET NOS. 50-325 AND 50-324
1.0 INTRODUCTION
By letter dated April 13, 2016 (Agencywide Documents and Management System (ADAMS)
Accession No. ML16111B203), as supplemented by letter dated March 1, 2017 (ADAMS Accession No. ML17087A263), Duke Energy Progress, LLC (Duke Energy or the licensee) requested amendments to Renewed Facility Operating License Nos. DPR-71 and DPR-62 for the Brunswick Steam Electric Plant (BSEP), Units 1 and 2, respectively. The proposed amendments would revise the Technical Specifications (TSs) for the reactor protection system (RPS) electrical protection assembly (EPA) electric power monitoring surveillance requirements (SRs) 3.3.8.2.2 and 3.3.8.2.3. The proposed change amends the allowable values (AVs) of SRs contained in TS 3.3.8.2, "RPS Electric Power Monitoring." Specifically, the TS change proposes to revise the RPS EPA AVs for overvoltage and undervoltage contained within SRs 3.3.8.2.2 and 3.3.8.2.3. The proposed change in SR 3.3.8.2.2 revises item "a. Overvoltage" from a value of s [less than or equal to] 129 V [Volts] to a value of s 127 V and item "b. Undervoltage" from a value of~ [greater than or equal to] 105 V to a value of ~ 107 V.
The proposed change in SR 3.3.8.2.3 revises item "a. Overvoltage" from a value of s 132 V to a value of s 127 V and item "b. Undervoltage" from a value of ~ 108 V to a value of~ 107 V.
Duke Energy will update TS Bases B 3.3.8.2 for BSEP, Units 1 and 2 to reflect these TS changes. The proposed SR changes result in no physical change to the plant configuration or method of operation.
The supplement dated March 1, 2017, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination, as published in the Federal Registeron June 7, 2016 (81 FR 36613).
2.0 REGULATORY EVALUATION
The regulatory requirements and guidance that the U.S. Nuclear Regulatory Commission (NRC) staff considered in its review of the applications are as follows:
- 1. Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix A, "General Design Criteria for Nuclear Power Plants," General Design Criterion (GDC)-13, Enclosure 3
"Instrumentation and control," requires that instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, 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. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.
BSEP Unit 1 and Unit 2 were not licensed to the 10 CFR Part 50, Appendix A, GDC.
The BSEP equivalent of GDC-13 is described in the BSEP Updated Final Safety Analysis Report, Section 3.1.2.2.4 as: "Instrumentation and suitable means of control are provided to monitor the operation of the reactor core [reactor coolant pressure boundary], and the containment and its associated systems over the anticipated normal operating and accident conditions, and to maintain these systems within the prescribed operating ranges."
- 2. Section 50.90 of 10 CFR provides direction to licensees seeking to revise their license to file an application for amendment with the NRG. The TSs constitute Appendix A to the Operating License for each facility. This would require a license amendment to revise any portion of the TSs, such as requested here.
- 3. Section 50.36 of 10 CFR, "Technical specifications," states, "Each applicant for a license authorizing operation of a production or utilization facility shall include in his application proposed technical specifications in accordance with the requirements of this section." As it pertains to the requested change, 10 CFR 50.36(c)(2)(ii) sets forth four criteria to be used in determining whether a limiting condition for operation (LCO) is required to be included in the TS: Criterion 1. Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. Criterion 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Criterion 4. A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety. Additionally, 10 CFR 50.36(c)(3) require the TS to contain surveillance requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.
3.0 TECHNICAL EVALUATION
3.1 Background 3.1.1 The Brunswick Steam Electric Plant and Reactor Protection System
BSEP is made up of two General Electric Boiling-Water Reactors (BWRs) (BWR/4 version with Mark I containments) located near the town of Southport, North Carolina. Units 1 and 2 share a common control room.
Each unit has two RPS trip systems (systems A and B). The power for each RPS is supplied, by a separate bus, via its own high inertia alternating current motor-generator (MG) set. The high inertia is provided by a flywheel and is sufficient to maintain voltage and frequency within 5 percent of rated values for at least 1 second following a total loss of power to the drive motor.
Alternate power is available to either RPS bus from standby electrical power. An alternate power switch prevents simultaneously feeding both buses from the same source and prevents placing an MG set in parallel with the alternate supply.
3.1.2 The Electrical Protection Assemblies In RPS electronic circuits, like the ones at BSEP, power is provided by either a primary or secondary power source such that if the primary is lost, there is an automatic switchover to the secondary power source. An undetected degradation of the voltage (i.e., the voltage is present, but outside the design limits) could potentially damage and impact the safety function of RPS components. Electric power monitoring is used to prevent occurrence of an undetected single failure that would allow the RPS power supply output voltage to remain outside the voltage rating of the connected Class 1E loads. This includes devices called electrical protection assemblies (EPAs). The EPAs meet the criteria of the BSEP plant-specific equivalent of GDC-13, "Instrumentation and control," because they are instrumentation that provides monitoring of the voltage parameter of the normal and alternate electric power supplies tor the RPS electronic circuits over their anticipated ranges tor normal operation, tor anticipated operational occurrences, and for accident conditions and take protective action as appropriate to help assure adequate safety. The EPAs provide protection to the RPS components downstream from the power supply buses by acting to disconnect the RPS from a power supply under specified conditions that could damage the RPS, such as overvoltage and undervoltage.
Thus, appropriate controls of the EPA circuits are provided to maintain these variables and systems within prescribed operating ranges.
In the event of an undervoltage condition tor an extended period of time, the RPS scram solenoids can chatter and potentially lose their safety control function capability. Further, in the event of an overvoltage condition, the RPS logic relays and scram solenoids may experience a voltage higher than their design voltage. If the overvoltage condition persists for an extended period of time, the equipment could undergo a degraded condition threatening the RPS safety function.
Two EPAs are in series between each MG set and its respective bus (labeled EPA 1 and 2 for RPS Bus A and EPA 3 and 4 tor RPS Bus B). In addition, there is a set of two EPAs connected to the single alternate power source, which is available to feed either of the RPS buses (labeled EPA 5 and 6). Duke Energy states that each EPA is seismically and environmentally qualified Class 1E.
3.2. The Electrical Protection Assembly Trip Safety Function 3.2.1 Allowable Value for Trip The electric power monitoring instrument channels of the RPS EPAs provide overvoltage and undervoltage protection at all times for the loads connected to the RPS 120 V power buses.
The EPA circuits are set to trip and disconnect from the power source when the voltage parameter value is greater than a preset value (overvoltage setpoint) or less than a preset value (undervoltage setpoint). The AV settings account for instrument drift uncertainty, test and measurement equipment uncertainty, and calibration tolerance.
Duke Energy states in the license amendment request (LAR), "The final nominal trip setpoints of the EPAs for overvoltage and undervoltage conditions were determined from measurements and calculations made in the field after final wiring and cabling were installed."
3.2.2 Historical Perspective on Electrical Protection Assembly Setpoints Duke Energy stated that a downstream component of the RPS, ASCO' (ASCO) Scram Solenoid Pilot Valves (SSPVs), are approaching the end of their qualified life. To address this issue, an Engineering Change was initiated to evaluate the need to rebuild or replace the existing ASCO SSPVs.
Duke Energy determined the best solution was to replace the SSPVs with Alloy Valves and Control, Inc. (AVCO) solenoid valves during the Unit 1 outage in 2012 and the Unit 2 outage in 2013. During the review of the Engineering Change, issues were identified with the current RPS EPA setpoints that could potentially allow the SSPV coils to operate above their design maximum voltage rating and below their minimum design voltage rating based on worst-case conditions. Based on the field measurements and calculations, the AVCO SSPVs became the most limiting voltage requirement for determining the settings for the RPS EPAs. As a result of this change, Duke Energy determined the TSs AVs should be revised based on the revised calculation.
The current AVs for the instrument settings for EPA 1 through EPA 6 are not the same. The AVs for the normal power supply electrical power monitoring assembly (i.e., EPAs 1 through 4) are based on the RPS MG sets providing ~ 57 Hertz (Hz) and 117 V +/- 10%. The A Vs for the instrument settings of the alternate power supply electrical power monitoring assembly (i.e.,
EPAs 5 and 6) are based on the alternate power supply providing ~ 57 Hz and 120 V +/- 10%.
Duke Energy indicated that they believed this difference in voltage and AV settings created the perception of different setpoints being required when powered from the two different power sources.
Duke Energy stated that when the EPAs were first installed, there was no voltage regulator for the alternate power supply. A voltage regulator has since been installed on the alternate power supply with specified settings to regulate the voltage at 117 V. This will regulate the normal and alternate power sources at the same voltage. The A Vs for both the normal and alternate power supply will be the same.
3.2.3 Proposed New EPA Allowable Values Section 50.36(c)(2)(ii) of 10 CFR requires, in part, that a technical specification LCO of a nuclear reactor must be established for each item meeting one of the four criteria. Criterion 3 includes "a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier." The EPAs provide protection to the RPS components by acting to disconnect the RPS from the power supply under specified conditions that could damage the RPS equipment and threaten achieving the safety function. Therefore, RPS electric power monitoring by the EPAs satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).
The SRs are used to help assure the EPAs are capable of accomplishing their safety function by performing periodic channel calibration. SR 3.3.8.2.2 and SR 3.3.8.2.3 associated with TS 3.3.8.2 contain the AVs for the EPA instruments. SR 3.3.8.2.2 performs the channel calibration for each EPA for the MG set, EPAs 1 through 4. SR 3.3.8.2.3 performs the channel calibration for each EPA for the alternate power supply, EPAs 5 and 6.
The licensee's request states in part that issues were identified during the review of the proposed change from ASCO to AVCO SSPVs that could subject the SSPVs to voltage conditions above or below their design limits. On January 31, 2017 (ADAMS Accession No. ML17031A292), the NRC staff requested the licensee to explain the reason, type, or severity of the issues identified and to provide a copy of the revised calculation for NRC staff review in order for the staff to verify the settings were conservative with respect to the solenoid voltage ratings. By letter dated March 1, 2017 (ADAMS Accession No. ML17087A263), the licensee responded to the NRC staff request and provided Calculation 1C71-0016, Revision, 1, Reactor Protection System Power Monitor Overvoltage, Undervoltage, Underfrequency, and Time Delay Uncertainty and Setpoint Calculation, as an enclosure (ADAMS Accession No. ML17087A264) to the letter. The original ASCO SSPVs were rated at 115 V +/- 1O V (i.e., 105 V to 125 V). The original calculation was based on transient conditions and did not address a sustained high or low voltage condition to assess if the equipment was subject to voltages beyond their design limits. As noted previously, sustained overvoltage or undervoltage can damage the equipment.
In its response to the NRC staff request for additional information, the licensee explained that the original ASCO SSPVs could potentially be exposed to overvoltage and undervoltage conditions beyond their deign limits under worst-case conditions when a nominal line voltage drop was taken into consideration. The design maximum and minimum operating voltages for the replacement AVCO SSPVs are 132 V and 90 V, respectively. However, the qualified life for the replacement AVCO SSPVs are based on a maximum and minimum voltage of 125 V and 102 V, respectively. The EPAs are rated up to 134 V for overvoltage and 95 V for undervoltage.
Therefore, the revised overvoltage and undervoltage settings of the SSPVs are the limiting values for the new protective settings. The revised calculation methodology is consistent with the methodology described in International Society of Automation Standard 67.04, Setpoints for Nuclear Safety-Related Instrumentation, which is referenced in Regulatory Guide 1.105, Rev. 3.
The calculation accounts for minimum and maximum voltage drops between the output of the power supplies and the solenoid valves to ensure that the calculation is representative of field conditions. Based on the revised overvoltage setpoint of 125 V and the minimum expected voltage drop of 2 V, the associated AV is 127 V, which is within the design limits for the
replacement AVCO SSPVs. Based on the revised undervoltage setpoint of 102 V and the maximum expected voltage drop of 5 V, the associated AV is 107 V, which is within the design limits of the replacement AVCO SSPVs.
The NRC staff finds that the proposed Technical Specification A Vs described above are acceptable and support the safety function of the EPAs. The proposed new AV for overvoltage is revised in a conservative direction (i.e., closer to the desired control value of the power supply voltage, so the EPA will continue to provide the safety function). The actuation setpoints selected will not increase the chance of a spurious trip. Having the AV values the same for EPAs 1 through 6 simplifies the TS and is an improvement in human factors for the plant operators. The calculation submitted by the licensee indicates that the As-Found Tolerance band established for use with required periodic surveillances accounts for instrument setting tolerance, drift, and calibration instrument uncertainty. The method for establishing this allowance is consistent with NRC guidance provided in Regulatory Issue Summary 2006-17, NRG Staff Position on the Requirements of 10 CFR 50.36, "Technical Specifications,"
Regarding Limiting Safety System Settings During Periodic Testing and Calibration of Instrument Channels (ADAMS Accession No. ML051810077).
Regarding the installation of a voltage regulator on the alternate power supply, the original controlling value for EPAs 5 and 6 was ~ 57 Hz and 120 V +/- 10%. Duke Energy states that the current controlling value with margin for EPAs 5 and 6 is now~ 57 Hz and 117 V +/- 10%, similar to EPAs 1 through 4. The 10% margin is+/- 11.7 V. For overvoltage, 117 V+ 11.7 V = 128.7 V; or with round off a value of 129 V; or if truncated, a value of 128 V; thus the proposed new AV of 127 Vis conservative. In a like manner, for undervoltage, 117 V - 11.7 V = 105.3 V; or rounded up, a value of 106 V, or truncated, a value of 105 V; thus the proposed new AV of~ 107 Vis conservative.
There are no changes to the frequency settings. Per Calculation 1C71-0016, Revision, 1, the AVs and device setpoint remain unchanged at 57.2 Hz and 57.7 Hz, respectively. The Design Limit for the RPS EPA underfrequency function is 57.0 Hz. This value is based on the protective circuitry requirement of -5.0% of 60 Hz. This is conservative based on the minimum qualified frequency of 56.5 Hz, per Brunswick Design Report 296, Rev. 0, Wyle Laboratories Test Report 44400R96-1.
A summary of the previous and the revised new settings is presented in the table below:
Parameter Current EPA 1 Current EPA 5 Proposed new through 4 through 6 EPA 1 through (SR 3.3.8.2.2) (SR 3.3.8.2.3) 6 Controlling value of ~57 Hz and ~57 Hz and ~57 Hz and Voltage for Original 117 v +/- 10%. 120 v +/- 10%. 117 v +/- 10%.
AV and new AV Overvoltage AV $129 v $132 v $127 v Undervoltage AV ~105 v ~108 v ~107 v
Regarding compliance with applicable regulatory requirements, the NRC staff finds that:
- The BSEP equivalent to GDC-13 is met because the requested change provides appropriate controls to maintain the RPS EPAs within their prescribed operating ranges.
- 10 CFR 50.36 requirements are met because TS 3.3.8.2 contains an LCO for RPS electric power monitoring by the EPAs (10 CFR 50.36(c)(2)(ii)) and the changes to the EPA AVs will continue to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met (10 CFR 50.36(c)(3)).
Based on the above, the NRC staff finds the licensee's modification of TS 3.3.8.2 of overvoltage and undervoltage AVs contained in SRs 3.3.8.2.2 and 3.3.8.2.3 acceptable.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the State of North Carolina official was notified of the proposed issuance of the amendment on April 3, 2017. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes surveillance requirements. The NRC 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, and that 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 (81 FR 36613, June 7, 2016).
Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 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 Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that 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 Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor: Eugene Eagle Date: April 11, 2017.
W. Gideon
SUBJECT:
BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENTS REGARDING REACTOR PROTECTION SYSTEM ELECTRICAL PROTECTION ASSEMBLY ELECTRIC POWER MONITORING SURVEILLANCE REQUIREMENTS 3.3.8.2.2 AND 3.3.8.2.3 (CAC NOS. MF7602 AND MF7603)
DATED APRIL 11, 2017 DISTRIBUTION:
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