ML15043A270

Issuance of Amendment Diesel Generator Initiation - Degraded Voltage Time Delay Setting Change

ML15043A270
Person / Time
Site:	Nine Mile Point
Issue date:	03/12/2015
From:	Bhalchandra Vaidya Plant Licensing Branch 1
To:	Orphanos P Exelon Generation Co
Vaidya B
References
TAC MF1022
Download: ML15043A270 (25)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 12, 2015 Mr. Peter M. Orphanos Site Vice President - Nine Mile Point Nuclear Station Exelon Generation Company, LLC 348 Lake Road Oswego, New York 13126

SUBJECT:

NINE MILE POINT NUCLEAR STATION, UNIT NO. 1 - ISSUANCE OF AMENDMENT RE: DIESEL GENERATOR INITIATION - DEGRADED VOLTAGE TIME DELAY SETTING CHANGE (TAC NO. MF1022)

Dear Mr. Orphanos:

The Commission has issued the enclosed Amendment No. 217 to Renewed Facility Operating License No. DPR-63 for the Nine Mile Point Nuclear Station, Unit No. 1 (NMP1 ). The amendment consists of changes to the Technical Specifications (TSs) in response to your application transmitted by letter dated March 8, 2013, as supplemented by letters dated May 16, 2013, July 8, July 16, August 29, 2014, and January 22, 2015.

Further, as a part of its application for the license transfer and conforming amendment of the Renewed Facility Operating License for Nine Mile Point Units 1 and 2, in the letter dated March 28, 2014, (ADAMS Accession No. ML14087A274), Exelon Generation Company, LLC has stated that:

Prior to the license transfers, GENG [Constellation Energy Nuclear Group, LLC]

made docketed submittals to the NRC [U.S. Nuclear Regulatory Commission]

that requested specific licensing actions, such as license amendment requests, relief requests, exemption requests, etc. Furthermore, in the application for the license transfers, Exelon stated that upon transfer of the licenses, Exelon would assume all current regulatory commitments made for these units. Accordingly, Exelon hereby adopts and endorses those docketed requests currently before the NRC for review and approval. Exelon requests that the NRC continue to process those pending actions on the schedules previously requested by GENG.

The amendment to the NMP1 Renewed Facility Operating License DPR-63 modified TS Table 3.6.2i, "Diesel Generator Initiation," by revising the existing 4.16kV Power Board (PB) 102/103 Emergency Bus Undervoltage (Degraded Voltage) Operating Time value and updating the Set Point heading title. In addition, subsequent to the issuance of the proposed amendment by the U.S. Nuclear Regulatory Commission, the NMP1 Updated Final Safety Analysis Report (UFSAR) Table XV-9, "Significant Input Parameters to the Loss-Of-Coolant Accident (LOCA)

Analysis," should be revised based on the issued amendment, to add a note regarding maximum allowable delay time from initiating signal to rated pump speed settings, to address the scenario of degraded grid voltage coincident with a LOCA.

P. M. Orphanos The TS and UFSAR revisions are being made to resolve the green non-cited violation (NCV) associated with the vital bus degraded voltage protection time delay documented in NRC Inspection Report (IR) 05000220/201101, "Nine Mile Point Nuclear Station - NRC Unresolved Item Follow-up Inspection Report," dated January 23, 2012, specifically, NCV05000220/20 11011-01, "Vital Bus Degraded Voltage Time Delay Not Maintained within LOCA Analysis Assumptions."

A copy of the related Safety Evaluation is enclosed. A Notice of Issuance will be included in the Commission's next regular biweekly Federal Register notice.

Sincerely, Bhalchandra Vaidya, Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-220

Enclosures:

1. Amendment No. 217 to DPR-63

2. Safety Evaluation cc w/encls: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY. LLC NINE MILE POINT NUCLEAR STATION. LLC DOCKET NO. 50-220 NINE MILE POINT NUCLEAR STATION. UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 217 Renewed License No. DPR-63

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

A The application for amendment by Exelon Generation Company, LLC (the licensee) dated March 8, 2013, as supplemented by letter dated May 16, 2013, July 8, July16, August 29, 2014, and January 22, 2015, 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.

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-63 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, which is attached hereto, as revised through Amendment No. 217, is hereby incorporated into this license.

Exelon Generation Company, 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 60 days.

FOR THE NUCLEAR REGULATORY COMMISSION Benjamin Beasley, Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the License and Technical Specifications Date of Issuance: March 12, 2015

ATTACHMENT TO LICENSE AMENDMENT NO. 217 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-63 DOCKET NO. 50-220 Replace the following page of the Renewed Facility Operating License with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove Page Insert Page Page 3 Page 3 Replace the following page of Appendix A, Technical Specifications, with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove Pages Insert Pages 238 238

(2) Exelon Generation pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (3) Exelon Generation pursuant to the Act and 10 CFR Parts 30, 40, and 70 to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Exelon Generation pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument and equipment calibration or associated with radioactive apparatus or components.

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

C. This renewed operating 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; Section 50.54 and 50.59 of Part 50; and Section 70.32 of Part 70. This renewed license 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 also 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 1850 megawatts (thermal).

(2) Technical Specifications The Technical Specifications contained in Appendix A, which is attached hereto, as revised through Amendment No. 217 is hereby incorporated into this license. Exelon Generation shall operate the facility in accordance with the Technical Specifications.

(3) Deleted Renewed License No. DPR-63 Amendment No. 191 through 210, 211, 213, 214, 215, 216, 217 Correction Letter Dated August 7, 2012

TABLE 3.6.2i (cont'd)

DIESEL GENERATOR INITIATION Limiting Condition for Operation Parameter Set Point Loss of Power Relay Dropout Operating Time

a. 4.16kV PB 102/103 Emergency Bus ~3200 volts 0 volts ::;3.2 seconds(a)

Undervolt (Loss of Voltage)

b. 4.16kV PB 102/103 Emergency Bus ~3705 volts >3.4 seconds(b)

Undervoltage (Degraded Voltage)  ::;24 seconds (c)

(a) The operating time indicated in the table is the time required for the relay to operate its contacts when the voltage is suddenly decreased from operating voltage level values to the voltage level listed in the table above.

(b) The operating time indicated in the table is the minimum time required to clear voltage transients due to load sequencing to avoid spurious separation from offsite power.

(c) The operating time indicated in the table is the maximum time allowable to preclude load damage or trip device actuation at voltages below the degraded voltage setpoint of 3705 volts.

AMENDMENT NO. 142. 148.217

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 217 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-63 EXELON GENERATION COMPANY, LLC NINE MILE POINT NUCLEAR STATION, LLC NINE MILE POINT NUCLEAR STATION UNIT NO. 1 (NMP1)

DOCKET NO. 50-220

1.0 INTRODUCTION

By letter dated March 8, 2013 (Reference 1), as supplemented by letter dated May 16, 2013 (Reference 2), July 8 (Reference 3), July 16 (Reference 4), August 29, 2014 (Reference 5), and January 22, 2015 (Reference 6), Exelon Generation Company, LLC (the licensee) submitted a request for changes to the Nine Mile Point Nuclear Station Unit No. 1, (NMP1) Technical Specifications (TSs).

The supplements dated May 16, 2013, July 8, July 16, August 29, 2014, and January 22, 2015, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRG) staff's initial proposed no significant hazards consideration determination noticed in the Federal Register on June 11, 2013, (78 FR 35062) (Reference 7).

Further, as a part of its application for the license transfer and conforming amendment of the Renewed Facility Operating License for Nine Mile Point Units 1 and 2, in the letter dated March 28, 2014, (Reference 8) Exelon Generation Company, LLC has stated that:

Prior to the license transfers, GENG [Constellation Energy Nuclear Group, LLC]

made docketed submittals to the NRC that requested specific licensing actions, such as license amendment requests [LARs], relief requests, exemption requests, etc. Furthermore, in the application for the license transfers, Exelon stated that upon transfer of the licenses, Exelon would assume all current regulatory commitments made for these units. Accordingly, Exelon hereby adopts and endorses those docketed requests currently before the NRG for review and approval. Exelon requests that the NRG continue to process those pending actions on the schedules previously requested by GENG.

The proposed amendment to the NMP1 Renewed Facility Operating License DPR-63 would modify Technical Specification (TS) Table 3.6.2i, "Diesel Generator Initiation," by revising the existing 4.16kV Power Board (PB) 102/103 Emergency Bus Undervoltage (Degraded Voltage)

Operating Time value and updating the Set Point heading title. In addition, subsequent to the issuance of the proposed amendment by U.S. Nuclear Regulatory Commission, the NMP1 Updated Final Safety Analysis Report (UFSAR) Table XV-9 (Reference 9), "Significant Input Parameters to the Loss-Of-Coolant Accident (LOCA) Analysis," should be revised based on the issued amendment, to add a note regarding maximum allowable delay time from initiating signal to rated pump speed settings, to address the scenario of degraded grid voltage coincident with a LOCA. The TS and UFSAR revisions are being made to resolve the Green non-cited violation (NCV) associated with the vital bus degraded voltage protection time delay documented in NRC Inspection Report (IR) 05000220/201101, "Nine Mile Point Nuclear Station - NRC Unresolved Item Follow-up Inspection Report," dated January 23, 2012 (Reference 10), specifically, NCV05000220/2011011-01, "Vital Bus Degraded Voltage Time Delay Not Maintained within LOCA Analysis Assumptions." The TS Basis was not changed.

The proposed amendment includes the following TS revisions:

• TS Table 3.6.2i, Operating Time Setting: Replace the,"< 60 seconds," upper time limit for the 4.16kV PB 102/103 Emergency Bus Undervoltage (Degraded Voltage) with, "~ 24 seconds."

• TS Table 3.6.2i, table heading for Set Point: Remove the parenthetical statement categorizing the type of relay.

The licensee (in Reference 1) proposed to add two notes in the NMP1 UFSAR Table XV-9.

Note (1) reads:

Note (1 ): This value is added to the maximum degraded voltage time delay in TS Table 3.6.2i for a degraded grid voltage coincident with a LOCA (section XV-C.2.2.5).

Note (2) was proposed to be added in response to NRC staff's RAI (Reference 5):

Note (2): 2 seconds are added to the maximum time delay for Core Spray start on Reactor Low-Low Level for LOCA coincident with LOOP [Loss of Offsite Power].

2.0 REGULATORY EVALUATION

The following explains the use of general design criteria (GDC) for NMP1. The construction permit for NMP1 was issued by the Atomic Energy Commission (AEC) on April 12, 1965, and the operating license was issued on December 26, 1974. The plant design criteria for NMP1 are listed in the Updated Final Safety Analysis Report (UFSAR)Section I, "Principal Design Criteria," (Reference 9). The AEC published the final rule that added Appendix A to Title 10 of the Code of Federal Regulations (1 O CFR) Part 50, "General Design Criteria for Nuclear Power Plants," in the Federal Register (36 FR 3255) on February 20, 1971 (Reference 11 ), with the rule effective on May 21, 1971. In accordance with an NRC staff requirements memorandum

from S. J. Chilk to J. M. Taylor, "SECY-92-223 - Resolution of Deviations Identified During the Systematic Evaluation Program," dated September 18, 1992 (Reference 12), the Commission decided not to apply the final GDC to plants with construction permits issued prior to May 21, 1971, which includes NMP1.

The NMP1 was not licensed to the 10 CFR 50, Appendix A GDC, while NMP2 was licensed to the GDC. The NMP1 Updated Final Safety Analysis Report (UFSAR) provides an assessment against the GDC in Table 1-1. This UFSAR table refers to the NMP1 Technical Supplement to Petition for Conversion from Provisional Operating License to Full-Term Operating License, July 1972, for the details of the assessment against the GDC current at that time. The NRC staff safety evaluation for Amendment No. 1, dated January 9, 1975, determined that the plant-specific requirements for NMP1 are sufficiently similar to the Appendix A GDC as related to the proposed change. [Reference 13: The NRC staff safety evaluation for Amendment No. 1, dated December 27, 1974, published on January 9, 1975, in the Federal Register, Volume 40, No. 6, Page 1760]

Therefore, the NRC staff reviews the amendment requests for the NMP1 license using the 1O CFR 50 Appendix A GDC (Reference 14) unless there are specific criteria identified in the UFSAR.

The regulatory requirements and guidance documents the NRC staff considered in its review of the proposed amendment included the following:

(a) Regulatory Requirements

• Title 10, Section 50.36, of the Code of Federal Regulations (1 O CFR) (Reference 15) requires that the facility's TS will include a section addressing limiting conditions for operation (LCO). 1O CFR Part 50.36, also states that each license authorizing operation of a production or utilization facility of a type described in§ 50.21 or§ 50.22 will include technical specifications. The technical specifications incorporated in a license will be designed to include those significant design features, operating procedures and operating limitations which are considered important in providing reasonable assurance that the facility will be constructed and operated without undue hazard to public health and safety.

• Title 10, Section 50.36(c)(2)(i), of the Code of Federal Regulations states, in part:

Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility.

When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met. ..

• Title 10, Section 50.36(c)(2)(ii), of the Code of Federal Regulations states, in part:

A technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the following criteria:

(B) 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.

(C) 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.

• Title 10, Section 50.36(c)(3), of the Code of Federal Regulations states the following:

Surveillance requirements are 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 satety limits, and that the limiting conditions for operation will be met.

• 1O CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," (Reference 16) establishes standards for the calculation of emergency core cooling system (ECCS) performance and acceptance criteria for that calculated performance.

• 1O CFR 50, Appendix K, "ECCS Evaluation Models," (Reference 17) establishes required and acceptable features of evaluation models for heat removal by the ECCS after the blowdown phase of a LOCA.

• General Design Criterion (GDC) 17, "Electric power systems," of Appendix A," General Design Criteria for Nuclear Power Plants," to Title10, Part 50, of the Code of Federal Regulations (1 O CFR) (Reference 14) requires, in part, that nuclear power plants have onsite and offsite electric power systems to permit the functioning of structures, systems,

and components that are important safety. The onsite system is required to have sufficient independence, redundancy, and testability to perform its safety function, assuming a single failure. The offsite power system is required to be supplied by two physically independent circuits that are designed and located so as to minimize, to the extent practical, the likelihood of their simultaneous failure under operating and postulated accident and environmental conditions. In addition, this criterion requires provisions to minimize the probability of losing electric power from the remaining electric power supplies as a result of loss of power from the unit, the offsite transmission network, or the onsite power supplies.

(b) Regulatory Guidance Regulatory Guide (RG) 1.105, "Setpoints for Safety-Related Instrumentation,"

Revision 3, issued December 1999 (ADAMS Accession No. ML993560062)

(Reference 18), describes a method that the NRG staff considers acceptable for complying with the agency's regulations for ensuring that setpoints for safety-related instrumentation are initially within and remain within the TS limits. The RG 1.105 endorses Part I of Instrument Society of America (ISA) Standard 67.04-1994, "Setpoints for Nuclear Safety-Related Instrumentation," subject to NRG staff clarifications. The NRG staff used this guide to establish the adequacy of the licensee's setpoint calculation methodologies and the related plant surveillance procedures.

3.0 TECHNICAL EVALUATION

3.1 Electrical Engineering Technical Evaluation:

NRG Staff Evaluation The NRG staff has reviewed the licensee's regulatory and technical analyses in support of its proposed license amendment, which is described in the Enclosure of the LAR and the supplemental responses to the NRG staff's request for additional information.

The undervoltage protection for NMP1 calculation is titled "4.16kV PB 102/103," which is designed to ensure that sufficient voltage is available to the loads connected to PB 102/103.

Two levels of undervoltage protection are provided; loss of voltage and degraded voltage. The loss of voltage relay setpoints specified in TS Table 3.6.2i are not affected by this change.

For degraded voltage, NMP1 calculation titled "4.16KVAC-PB102/103SETPT/27" determined that the time delay for the Degraded Voltage Relay (DVR) should be set at 21 plus or minus 3 seconds. The basis for the maximum allowable relay time delay setpoint is to preclude motor insulation degradation or actuation of protective devices. The most limiting time duration was determined to be 200 seconds for the limiting electrical components and breakers on down-stream of PB 16B and 17B. The current TS limit approved in NMP1 License Amendment 148 (ADAMS Accession No. ML011070061) (Reference 19) selected 60 seconds as the maximum time the degraded voltage condition could be sustained and preclude damage to loads or trip device actuation. The licensee stated that changing the limit of the maximum time the degraded voltage condition could be sustained to !:>24 seconds is conservative in that the in-plant settings for the degraded voltage relay operating time are currently set at 21 +/- 3 seconds. The

licensee also stated that changing the TS limit from <60 seconds to ~24 seconds is bounded by the current calculations and analysis for delays up to 200 seconds.

The proposed amendment would modify the existing 4.16kV PB 102/103 Emergency Bus Undervoltage (Degraded Voltage) Operating Time value only and the related dropout setpoint and current value of <::3705 volts of the DVR remains unchanged.

Component Design Basis Inspections (CDBI) have identified inadequate voltage setpoints at several nuclear plants and the NRC issued Regulatory Issue Summary (RIS) 2011-12 "Adequacy of Electric Distribution System Voltages" (Reference 20) to inform the industry about the findings at some plants. The NRC staff asked the licensee to provide additional information on the voltage setpoint in order for the NRC staff to complete the review of the proposed amendment request.

The additional information requested by the NRC staff included the following:

1. Validation of the voltage setpoint of 3705V to satisfy the criterion delineated in the RIS as related to the starting and running voltage requirements for the safety related loads at all busses.

2. Provision of excerpts from calculation(s) that establish the limiting voltage at various safety buses for equipment operability with the 102/103 Power Boards at 3705V.

In a letter dated May 16, 2013, (Reference 2) the licensee provided supplemental information related to the NMP1 License Amendment Request. The supplement included a summary clarifying that the current NMP1 DVR setpoint complies with the criterion in RIS 2011-12 (Reference 20). The supplement also included excerpts from a preliminary degraded voltage study conducted to confirm the degraded voltage setpoint and safety related bus voltage values.

The excerpts included the scope, design inputs and assumptions, acceptance criteria, and conclusions of the study. The licensee stated that the results of a preliminary degraded voltage study validated that the voltage requirements (starting and running) for all safety related equipment were preserved by the DVR dropout voltage setpoint. The conclusion of the study showed that adequate starting and running voltages are provided to all safety related equipment during bounding accident conditions with the OVA-monitored buses at the DVR dropout setting of 3705V.

Subsequently, in a letter dated July 8, 2014, (Reference 3) the licensee confirmed that the results of the preliminary study for establishing the degraded voltage set points had been finalized and accepted as a formal calculation. The licensee also stated that there were no changes to the summary clarifying that the current NMP1 DVR setpoint complies with the criterion in RIS 2011-12, the inputs, and the assumptions utilized in the study. The licensee concluded that the final study refined the preliminary study results with improved margin obtained. However, the licensee clarified that the final study results changed the conclusions provided in the May 16, 2013, submittal.

The NRC staff reviewed the July 8, 2014, submittal, and during several teleconferences requested the licensee to identify the specific changes made to the study conclusions and

provide a markup of the final degraded voltage study results to the NRC staff showing those changes.

In the letter dated August 29, 2014, (Reference 5) the licensee submitted a second supplement to provide a summary of the changes in the final degraded voltage study. This supplement included a revised copy of the preliminary study conclusions reflecting the new information from the final study.

The NRC staff's review of the calculation changes was limited to the summary of loads that were impacted at the degraded voltage relay setting and general assumptions. Based on the understanding that the licensee has performed the calculations in accordance with the criterion delineated in RIS 2011-12, the NRC staff concluded that:

(1) the 3705V DVR dropout setpoint is supported by the formal degraded voltage study.

(2) The licensee's supplemental submittal addresses the adequacy of the DVR design in ensuring that safety-related systems are supplied with adequate voltages.

Task Interface Agreement (TIA 2011-003) "Nine Mile Point Nuclear Station Unit 1 Licensing Basis for Degraded Grid Relay Time Delays," dated June 29, 2011, (ADAMS Accession No.

ML11171A702) (Reference 21) provided NRC staff position related to DVR time delay requirements. The TIA clarified that the allowable time delay, including any margins, shall not exceed the maximum time delay that is assumed in the UFSAR accident analysis. The specific requirement is that the degraded voltage protection time delay should be set such that the ECCS is able to inject water in to the core within the maximum allowable time assumed in the UFSAR accident analysis; even if a sustained degraded grid voltage condition is present.

The licensee has stated that the 24 second DVR time delay is set such that the ECCS is able to inject water in to the core within the maximum allowable time assumed in the UFSAR accident analyses with a sustained degraded voltage of ~3705 V at the 4160 V safety busses. Based on the review of the licensee's excerpts of analyses and clarifications in support of the maximum time and dropout setpoint for the DVRs, the NRC staff finds that the proposed change in maximum allowable delay time for DVR to be acceptable. The NRC staff therefore finds the proposed change to TS Table 3.6.2i, Operating Time Setting acceptable.

The LAR identified that the table heading for "Set Point" in TS Table 3.6.2i includes a categorization of the relay type in a parenthetical statement as, "Inverse Time Undervoltage Relays." The NRC staff agrees that this type of information in the table heading is not necessary. The parameters of interest are critical attributes such as operating limits of the time delay and the voltage setpoints. The NRC staff concludes that removal of the relay type from the TS Table 3.6.2i heading is acceptable.

On October 10, 2014, the NRC completed a CDBI at the NMPNS. Inspection Report No. 05000220/2014007 (Reference 22), dated November 20, 2014, identified three findings of very low risk significance related to DVRs. By letter dated December 10, 2014, (Reference 23),

the NRC staff asked the licensee to provide additional information related to the CDBI findings identified below:

1. The inspection team identified a failure to adequately evaluate the transient voltages to the Class 1E accident initiated motors and motor operated valves (MOVs) on the-safety related buses and motor control centers (MCC's). Specifically, the calculations incorrectly used 115 kilovolts (kV) grid voltage instead of incorporating the 3.5% grid voltage sag into calculation NIMO-ELMSAC01. Consequently, the licensee did not verify and assure adequate voltages would be available to Unit 1 Class 1E accident initiated motors, MOVs, and control circuits powered from the 4160 V, 600 V, and 120 V distribution systems during a design-basis LOCA with subsequent unit trip and resulting sag of the 115 kV grid.

2. The inspection team noted that the NMP1 electrical design calculations had not evaluated tor the following conditions:

a. Connected Class 1 E loads would not be degraded or rendered inoperable for a design basis LOCA and a sustained degraded voltage condition between the degraded voltage dropout setting (3705 V) and the loss of voltage setting (3200 V) tor the degraded voltage time delay of 21 +/- 3 seconds and subsequent reconnection to the emergency diesel generator.

b. Safety-related equipment that is operating or safety-related loads that are required to start (motors, MOVs, etc.) had not been evaluated to ensure that their protective devices would not actuate during a sustained degraded grid condition coincident with a design basis LOCA.

For the conditions identified above, the required equipment may not be available after transfer of safety busses to the onsite power sources.

For CDBI finding 1 above, the NRG staff requested a summary of the evaluations provided during the CDBI to verify that the critical Class 1E loads that operate during the first 24 seconds of a LOCA would not be damaged or become unavailable for a design basis LOCA coincident with a sustained degraded voltage condition. The NRG staff asked tor input and assumptions used tor each preliminary evaluation, including load tap changer performance during the 24 second period.

For CDBI finding 2 above, the NRG staff requested a summary of evaluations performed to evaluate conditions 2.a and 2.b described above along with each input and assumptions used for each preliminary evaluation.

For CDBI findings 1 and 2 above, the NRG staff requested details on the corrective actions planned and taken, including the review of the extent-of-condition for components required during a LOCA.

In a letter dated January 22, 2015, (Reference 6) the licensee provided supplemental information related to staff's questions on CDBI findings. The supplement included a summary of the preliminary evaluations provided during the CDBI, for both findings, to verify that the critical Class 1E loads that operate during the first 24 seconds of a LOCA would not be damaged or become unavailable for a design basis LOCA coincident with a sustained degraded

voltage condition. The licensee also provided corrective actions planned and taken, including the review of the extent-of-condition for components required during a LOCA.

For CDBI finding 1 above, the licensee stated that during the recent 2014 CDBI inspection, it was noted that the formal calculation used 115kV as a design input but failed to include the 3.5 percent grid voltage sag into the calculation. The licensee provided a summary of the reevaluation of plant voltage profile using the 3.5 percent grid voltage sag and concluded that the resultant change incorporated in the calculation did not impact the previous conclusions provided in the previously submitted Second Supplement to Nine Mile Point Nuclear Station LAR for Diesel Generator Initiation - Degraded Voltage Time Delay Setting Change, dated August 29, 2014 (Reference 5). The licensee has stated that the conclusion remains the same, in that all safety related equipment will operate during degraded voltage conditions at the degraded voltage setting of 3705V during the first 21 +/- 3 seconds. The acceptance criteria noted in the Second Supplement for this equipment was still achieved after incorporating the 3.5 percent grid voltage sag. The NRG staff reviewed the summary of evaluations supporting this conclusion submitted as "Equipment Technical Evaluations" and found them acceptable.

For CDBI finding 2 above, the licensee stated that Conditions 2.a and 2.b above were evaluated using a preliminary scenario utilizing the formal calculation, NIMO-ELMS-AC01 Revision 1, "Performance of the Electrical Auxiliary System." The licensee explained that a preliminary model was used to validate that the safety-related equipment required to operate during the first 24 seconds under sustained degraded voltage condition concurrent with a unit trip and LOCA would not be degraded, rendered inoperable and/or would not actuate the associated protective devices. Also, the licensee stated that the preliminary scenario simulated the 4160 V power boards at 3200V which is the Loss of Voltage Relay (LVR) setpoint, and that this provides the minimum voltage at the terminals for the loads before separating from the power grid. The licensee stated that Class 1E motor protective functions were evaluated based on the consideration that the Class 1E motors do not stall within the DVR time delay of 21 +/- 3 seconds. Also, the licensee stated that the criterion of motor stall was selected because it results in the protective device being subjected to a locked rotor current (LRA) condition for a time period that may be long enough to challenge protective devices. The licensee concluded that all connected Class 1E loads would not be degraded or rendered inoperable for a design-basis LOCA and a sustained degraded voltage condition between the degraded voltage dropout setting {3705V) and the loss of voltage setting (3200V) for the degraded voltage time delay of 21 +/- 3 seconds. The equipment will be available for reconnection to the emergency diesel generators to mitigate the consequences of an accident. The evaluations supporting this conclusion are included in Enclosure 1, Equipment Technical Evaluations of the licensee's submittal. The NRG staff finds this portion of the response acceptable.

The NRG staff notes that regarding the review of the extent-of-condition for components required during a LOCA, the licensee stated that the incorporation of the preliminary findings into the necessary calculations and design documents is being tracked as part of the station's corrective action program. The actions taken include the issuance of a modification to increase the circuit breaker size for select Core Spray MOVs. Increasing the breaker size will provide additional design margin for MOV operation during degraded voltage conditions for these valves. The licensee has evaluated the performance of the plant equipment considering the grid voltage sag of 3.5%. Formal calculations will be updated to reflect the changes in the voltage profile.

Among the actions planned, the licensee identified the following:

1. The updated grid voltage analysis will be used as the basis for calculation updates to ensure grid voltage sag is documented and addressed for degraded voltage scenarios.

2. Update the formal calculation to include case models that address safety related equipment and protective device operation between the degraded voltage relay setpoint of 3705V and loss of voltage relay setpoint of 3200V.

3. Develop and issue calculation(s) for MOV contactors and include evaluation of control circuit protective fusing.

The licensee also explained that an extent of condition evaluation was performed to identify the Class 1E safety related components required to perform their safety function within the first 24 seconds of degraded voltage conditions. The components identified for this extent of condition will be addressed in formal engineering evaluations. The licensee has included the actions to upgrade the existing calculations to include the additional analysis for grid voltage sag conditions and MOV protective devices and plans to upgrade select breakers to increase design margin, as part of the corrective action program.

Conclusions - Electrical Engineering Evaluation The NRC staff has reviewed the licensee's proposed TS changes and supporting documentation. Based on the evaluation discussed above, the NRC staff has determined that the proposed amendment to the TS Table 3.6.2i is consistent with the requirements in GDC 17 and supplemental clarification in TIA 2011-003. The licensee has evaluated the CDBI findings and concluded that the DVR time delay and dropout setpoint is adequate for protecting the safety-related equipment needed to ensure cooling of the core and the maintenance of containment integrity and other vital functions even in the event of postulated accidents. The staff also evaluated the responses to 2014 CDBI findings related to DVR setpoints. Based on the summary of evaluations performed in response to the findings, the staff has concluded that the results of existing calculations remain valid. The removal of details on the relay type from the heading of TS Table 3.6.2i is editorial in nature and does not impact the functional requirements of the relay. Therefore, the NRC staff finds the proposed changes acceptable and consistent with the NRC Regulations for compliance with GDC 17 requirements.

3.2 Reactor Systems Technical Evaluation:

In the Technical Evaluation" section of its March 8, 2013, submittal [Reference 1], the licensee stated that the undervoltage protection for NMP1 calculation is titled "4.16kV PB 102/103,"

which is designed to ensure that sufficient voltage is available to the loads connected to PB 102/103. Two levels of undervoltage protection are provided; loss of voltage and degraded voltage. The loss of voltage relay setpoints specified in TS Table 3.6.2i are not affected by this change. For degraded voltage, NMP1 calculation "4.16KVAC-PB102/103SETPT/27" determined that the time delay for the degraded voltage relay should be set at 21 plus or minus 3 seconds. The currently approved TS limit is 60 seconds as the maximum time the degraded

voltage condition could be sustained and preclude damage to loads or trip device actuation.

The proposed change limits to~ 24 seconds in that the in-plant settings for the degraded voltage relay operating time are currently set at 21 +/- 3 seconds.

The licensee further stated that a 24 second time delay for diesel generator initiation under degraded voltage conditions results in a maximum time delay of 59 seconds from initiating signal to core spray pump at rated speed for the special scenario of degraded voltage conditions coincident with a LOCA. The value of 59 (= 21+3+10+25) seconds was determined by adding the nominal 21 second time delay for diesel generator initiation, plus the maximum 3 second uncertainty, to the current 1O second diesel generator start time and the 25 second time for the core spray pumps to attain rated speed. An evaluation performed by General Electric Hitachi Nuclear Energy (GEH) using approved methodologies and 1O CFR 50 Appendix K assumptions to assess the sustained degraded voltage condition coincident with a LOCA showed that the analysis results remained below the 1O CFR 50.46 acceptance criteria of 2200 F for Peak Cladding Temperature (PCT) and 17% for Maximum Local Oxidation (MLO). The GEH analysis demonstrated that the ECCS will perform its safety function with a time delay of 60 seconds from event initiation to core spray pump at rated speed, resulting in insignificant differences in the PCT and MLO for both GE11 and GNF2 fuel types in use at NMP1.

In response to the NRG staff questions [Reference 4] concerning the analyses for the special scenario of degraded voltage conditions coincident with a LOCA, and the most limiting LOCA analysis of record for the current licensing basis (CLB) for NMP1, the licensee submitted supplemental information which included the following:

A 60 second time delay from initiation to core spray pump at rated speed was used in the LOCA analysis coincident with a degraded voltage condition. The resulting PCT was 2113 F for GNF2 fuel and 2141 F for GE11 fuel. The licensee updated the analysis of record to incorporate the degraded voltage condition scenario. No additional changes are expected to be required as a result of this amendment. The degraded voltage LOCA scenario was analyzed in response to the NRG NCV in 2012 that was cited in the LAR.

The most limiting LOCA analysis of record for the CLB for NMP1 is the Large Recirculation Line Break with a Loss of Offsite Power (LOOP) using Appendix K assumptions and bottom peaked axial power shape at rated plant conditions. The PCT for this event was 2144 F for GNF2 fuel and 2142 F for GE11 fuel. Therefore, the LOOP-LOCA design basis analysis remains the CLB, setting the PCT for both GNF2 and GE11 fuel. The degraded voltage LOCA results satisfy the 1O CFR 50.46 acceptance criteria and are bounded by the CLB scenario.

In addition, the NRG staff requested the licensee to explain an apparent discrepancy between the current and the proposed Table XV-9 of NMP1 UFSAR. In the current table, the maximum allowable delay time from initiating signal to pump at rated speed (sec) is listed as 35 sec for GE11 and 37 sec for GNF2 fuel; whereas, in the proposed table, it is listed as 35 sec for both the fuel types. The licensee stated (in Reference 4) that for the NMP1 GNF2 and GE11 CLB LOCA analysis coincident with a LOOP, the core spray delay time used in the analyses for both fuel types is 37 seconds (35 seconds+ 2 seconds for low-low level indication). As the 2 seconds is a conservative assumption, not part of the actual timing of the event, the NMP1 UFSAR will show 35 seconds in Table XV-9. In order to further clarify the additional

conservatism applied to the LOOP/LOCA, the licensee proposed that a second note will be added to the table stating [Reference 5]:

"(2) 2 seconds are added to the maximum time delay for Core Spray start on Reactor Low-Low Level for LOCA coincident with LOOP."

The NRC staff reviewed the licensee's submittal [Reference 1], supplemental information provided in response to the NRC staff's questions [References 4 and 5], and related documentation (e.g., TS, UFSAR). The NRC staff determined that the NRC Inspection Report (NCV05000220/20 11011-01, "Vital Bus Degraded Voltage Time Delay Not Maintained within LOCA Analysis Assumptions") identified conflicting values in TS Table 3.6.2.i and UFSAR Table XV-9. The UFSAR LOCA significant input parameter of 35 seconds for maximum allowable delay time from initiating signal to pump at rated speed documented in UFSAR Table XV-9 was determined by the NRC to be non-conservative when compared to the degraded voltage relay setpoint of< 60 seconds specified in TS Table 3.6.2.i. The degraded voltage protection time delay should be set such that the ECCS is able to inject water in to the core within the maximum allowable time assumed in the UFSAR accident analyses, even if a sustained degraded grid voltage condition is present.

The NRC staff concluded that the proposed revision of the maximum delay time for the degraded voltage relays in TS Table 3.6.2i to 24 seconds results in a maximum ECCS injection time of 59 seconds (24 seconds added to 35 seconds according to Note 1, Table XV-9, UFSAR) from event initiation. The ECCS analysis demonstrated that a maximum delayed injection time of 59 secs from event initiation is acceptable because the results of the analysis using a conservative delay time of up to 60 secs employing NRC approved methodologies comply with the requirements of 1O CFR 50.46, as discussed above. The NRC staff further concludes that the change proposed in this amendment request will align the TS limits for degraded voltage documented in TS Table 3.6.2i with the current plant configuration settings, and make the TS Table 3.6.2.i consistent with the UFSAR Table XV-9. The NRC staff, therefore, concludes that the proposed amendment of the NMP1 TS continues to meet the 10 CFR 50.36, 10 CFR 50.46 and 10 CFR 50, Appendix K regulatory requirements, and is acceptable.

Conclusions - Reactor Systems Technical Evaluation:

Based on the discussion above, the NRC staff determined that the applicable regulatory requirements will continue to be met, adequate defense-in-depth will be maintained, and sufficient safety margins will be maintained. The NRC staff, therefore, finds that the proposed LAR is acceptable.

3.3 Instrumentation and Controls Technical Evaluation:

The current 4.16kV PB 102/103 Emergency Bus Undervoltage (degraded voltage) relay time delay is allowed to be set between 3.4 seconds and 60 seconds. However, the associated relays are actually set to 21 +/- 3 seconds. The 60 second upper limit of this setting was determined to be unacceptable. It is inconsistent with the UFSAR LOCA maximum allowable delay time from initiating signal to pump at rated speed documented in UFSAR Table XV-9 when considering degraded grid voltage coincident with a LOCA. The maximum allowable time for the core spray pump to reach its rated speed in the LOCA analysis is 35 seconds from the

initiating signal. This reduces the allowable time delay for detection of 4.16 KV bus undervoltage and initiation of diesel start. To account for this reduction, the upper specified Operating Time in Table 3.6.2i for the 4.16kV PB 102/103 Emergency Bus Undervoltage (Degraded Voltage) is being changed from <60 seconds to</= 24 seconds.

The setpoint heading title is also being changed from "Set Point (Inverse Time Undervoltage Relays)" to "Set Point." Justification for this title change was provided by the licensee stating the parenthetical information in the table heading was not necessary. Information on the type of relay used to accomplish the required function in not relevant to the limiting conditions for operation presented in the table.

The NRC staff confirmed the setpoint limits used in the plants surveillance tests provide adequate margin between relay actuation and the revised analytical limit value of 24 seconds.

No changes to the plant relay settings are being made because the current plant settings for the delay time are between 18 and 24 seconds which meets the UFSAR criteria for detection of 4.16 KV bus undervoltage and initiation of diesel start. This LAR is therefore correcting the Technical Specification Table 3.6.2i maximum value to less than or equal to 24 seconds which is consistent with the current instrument setting and this current instrument setting remains consistent with the UFSAR LOCA analysis assumptions. The NRC inspectors have also verified the NMPI Degraded Voltage time delay relays to be set at a nominal 21 seconds.

Conclusions - Instrumentation and Controls Technical Evaluation:

Because no changes to the degraded voltage relay setpoints or time delay settings are being made by this amendment and because current settings remain within the time delay limits defined in the revised technical specifications, the NRC staff determined the proposed changes to Technical Specification Table 3.6.2i are acceptable and existing instrumentation will remain capable of meeting the maximum time delay for degraded voltage assumed in the FSAR accident analysis. The NRC staff finds the proposed Technical Specification change to be compliant with 10 CFR 50.36(c)(2) and 10 CFR 50.36(c)(3).

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the New York State official was notified of the proposed issuance of 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 1O CFR Part 20, and 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 in the Federal Register on June 11, 2013, (78 FR 35062)

(Reference 7).

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 NRC staff 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.

7.0 REFERENCES

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March 8, 2013, ADAMS Accession No. ML13073A103.

2. Costanzo, C. R., Vice President-Nine Mile Point, letter to Document Control Desk, U.S. Nuclear Regulatory Commission (NRC), Nine Mile Point Nuclear Station, Unit No. 1, Docket No. 50-220, "Supplement to Nine Mile Point Nuclear Station License Amendment Request for Diesel Generator Initiation - Degraded Voltage Time Delay Setting Change," May 16, 2013, ADAMS Accession No. ML13144A068.

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ML14251A233.

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January 22, 2015, ADAMS Accession No. ML15026A132.

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dated January 23, 2012, ADAMS Accession No. ML12023A119.

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ML11171A702.

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Inspection Report No. 05000220/2014007,"0ctober 10, 2014, ADAMS Accession No. ML14325A019.

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December 10, 2014, ADAMS Accession No. ML14342A097.

24. Federal regulation U.S. Code of Federal Regulations, "Environmental Protection R3gulations For Domestic Licensing and Related Regulatory Functions - Cirterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review," Title 10 of the Code of Federal Regulations (1 O CFR) Part 51, Section 22.

Principal Contributors: M. M. Razzaque, NRR/DSS/SRXB R. Stattel, NRR/DE/EICB T. Martinez, NRR/DE/EEEB Date: March 12, 2015

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