ML17296A133
| ML17296A133 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 02/02/2018 |
| From: | Andrew Hon Plant Licensing Branch II |
| To: | James Shea Tennessee Valley Authority |
| Hon A DORL/LPL2-2 301-415-8480 | |
| References | |
| CAC MF9398, CAC MF9399, EPID L-2017-LLA-0188 | |
| Download: ML17296A133 (18) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 February 2, 2018 Mr. Joseph W. Shea Vice President, Nuclear Licensing Tennessee Valley Authority 1101 Market Street, LP 4A Chattanooga, TN 37402-2801
SUBJECT:
SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENTS RE: TECHNICAL SPECIFICATION MODIFYING SR 3.8.1.17 OF TS 3.8.1 "AC SOURCES - OPERATING (CAC NOS. MF9398 AND MF9399; EPID L-2017-LLA-0188)
Dear Mr. Shea:
The Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment Nos. 340 and 333 to Renewed Facility Operating License Nos. DPR-77 and DPR-79, respectively, for Sequoyah Nuclear Plant, Units 1 and 2, in response to the Tennessee Valley Authority application dated March 13, 2017, as supplemented by a "CNL-17-008, License Amendment Request to Modify Technical Specification (TS) Surveillance Requirement 3.2.4, QPTR, and TS 3.3.1, [[Equipment trip" contains a listed "[" character as part of the property label and has therefore been classified as invalid. System (RTS) Instrumentation, Condition D (SQN-TS-17-02 and WBN-TS-17-014)|letter dated August 7, 2017]].
The amendments delete a Note associated with Technical Specification Surveillance Requirement 3.8.1.17 to allow the performance of the Surveillance Requirement in Modes 1 through 4 when the associated load is out of service for maintenance or testing.
A copy of our related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
Docket Nos. 50-327 and 50-328
Enclosures:
- 1. Amendment No. 340 to DPR-77
- 2. Amendment No. 333 to DPR-79
- 3. Safety Evaluation cc: Listserv Sincerely, Andrew Hon, Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-327 SEQUOYAH NUCLEAR PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 340 Renewed License No. DPR-77
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Tennessee Valley Authority (the licensee),
dated March 13, 2017, as supplemented by a "CNL-17-008, License Amendment Request to Modify Technical Specification (TS) Surveillance Requirement 3.2.4, QPTR, and TS 3.3.1, [[Equipment trip" contains a listed "[" character as part of the property label and has therefore been classified as invalid. System (RTS) Instrumentation, Condition D (SQN-TS-17-02 and WBN-TS-17-014)|letter dated August 7, 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, as amended; 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 1 O 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-77 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.340 are hereby incorporated into this renewed license. The licensee 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 no later than 60 days from the date of its issuance.
Attachment:
Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: February 2, 2018 FOR THE NUCLEAR REGULATORY COMMISSION fta.i~~O/
Undine Shoop, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
ATTACHMENT TO LICENSE AMENDMENT NO. 340 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-77 SEQUOYAH NUCLEAR PLANT, UNIT 1 DOCKET NO. 50-327 Replace page 3 of Renewed Facility Operating License No. DPR-77 with the attached revised page 3. The revised page is identified by amendment number and contains a marginal line indicating the area of change.
Replace the following page of the Appendix A Technical Specifications with the attached revised page. The revised page is identified by amendment number and contains marginal line indicating the areas of change.
REMOVE 3.8.1-14 INSERT 3.8.1-14 (3)
Pursuant to the Act and 1 O 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)
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 calibration or associated with radioactive apparatus or components; and (5)
Pursuant to the Act and 1 O CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the Sequoyah and Watts Bar Unit 1 Nuclear Plants.
C.
This renewed license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 1 O CFR Chapter I and 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)
(2)
(3)
Maximum Power Level The Tennessee Valley Authority is authorized to operate the facility at reactor core power levels not in excess of 3455 megawatts thermal.
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.340 are hereby incorporated into this renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.
Initial Test Program The Tennessee Valley Authority shall conduct the post-fuel-loading initial test program (set forth in Section 14 of Tennessee Valley Authority's Final Safety Analysis Report, as amended), without making any major modifications of this program unless modifications have been identified and have received prior NRC approval. Major modifications are defined as:
- a.
Elimination of any test identified in Section 14 of TVA's Final Safety Analysis Report as amended as being essential;
- b.
Modification of test objectives, methods, or acceptance criteria for any test identified in Section 14 of TVA's Final Safety Analysis Report as amended as being essential; Amendment No. 340 Renewed License No. DPR-77
SURVEILLANCE REQUIREMENTS continued SR 3.8.1.17 SURVEILLANCE Verify setpoint for each load sequence timer is within +/- 5% of design.
SEQUOYAH - UNIT 1 3.8.1-14 AC Sources - Operating 3.8.1 FREQUENCY In accordance with the Surveillance Frequency Control Program Amendment~. 340 I
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-328 SEQUOYAH NUCLEAR PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 333 Renewed License No. DPR-79
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Tennessee Valley Authority (the licensee),
dated March 13, 2017, as supplemented by a "CNL-17-008, License Amendment Request to Modify Technical Specification (TS) Surveillance Requirement 3.2.4, QPTR, and TS 3.3.1, [[Equipment trip" contains a listed "[" character as part of the property label and has therefore been classified as invalid. System (RTS) Instrumentation, Condition D (SQN-TS-17-02 and WBN-TS-17-014)|letter dated August 7, 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, as amended; 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 1 O 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-79 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 333, are hereby incorporated into the license. The licensee 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 no later than 60 days from the date of its issuance.
Attachment:
Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: February 2, 2018 FOR THE NUCLEAR REGULATORY COMMISSION 1/f)J~~
Undine Shoop, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
ATTACHMENT TO LICENSE AMENDMENT NO. 333 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-79 SEQUOYAH NUCLEAR PLANT, UNIT 2 DOCKET NO. 50-328 Replace page 3 of Renewed Facility Operating License No. DPR-79 with the attached revised page 3. The revised page is identified by amendment number and contains a marginal line indicating the area of change.
Replace the following page of the Appendix A Technical Specifications with the attached revised page. The revised page is identified by amendment number and contains marginal line indicating the areas of change.
REMOVE 3.8.1-14 INSERT 3.8.1-14 (3)
Pursuant to the Act and 1 O 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)
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 calibration or associated with radioactive apparatus or components; and (5)
Pursuant to the Act and 1 O CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the Sequoyah and Watts Bar Unit 1 Nuclear Plants.
C.
This renewed license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and 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 Tennessee Valley Authority is authorized to operate the facility at reactor core power levels not in excess of 3455 megawatts thermal.
(2)
Technical Specifications (3)
The Technical Specifications contained in Appendices A and B, as revised through Amendment No.333 are hereby incorporated into this renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.
Initial Test Program The Tennessee Valley Authority shall conduct the post-fuel-loading initial test program (set forth in Section 14 of Tennessee Valley Authority's Final Safety Analysis Report, as amended), without making any major modifications of this program unless modifications have been identified and have received prior NRG approval. Major modifications are defined as:
- a.
Elimination of any test identified in Section 14 of TVA's Final Safety Analysis Report as amended as being essential;
- b.
Modification of test objectives, methods or acceptance criteria for any test identified in Section 14 of TVA's Final Safety Analysis Report as amended as being essential; Amendment No. 333 Renewed License No. DPR-79
SURVEILLANCE REQUIREMENTS continued SR 3.8.1.17 SURVEILLANCE Verify setpoint for each load sequence timer is within +/- 5% of design.
SEQUOYAH-UNIT 2 3.8.1-14 AC Sources - Operating 3.8.1 FREQUENCY In accordance with the Surveillance Frequency Control Program Amendment~. 333 I
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 340 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-77, AND AMENDMENT NO. 333 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-79 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328
1.0 INTRODUCTION
By letter dated March 13, 2017 (Agency wide Document Access Management System (ADAMS)
Accession No. ML17073A018), as supplemented by a "CNL-17-008, License Amendment Request to Modify Technical Specification (TS) Surveillance Requirement 3.2.4, QPTR, and TS 3.3.1, [[Equipment trip" contains a listed "[" character as part of the property label and has therefore been classified as invalid. System (RTS) Instrumentation, Condition D (SQN-TS-17-02 and WBN-TS-17-014)|letter dated August 7, 2017]] (ADAMS Accession No. ML17219A517) Tennessee Valley Authority (TVA), the licensee for Sequoyah Nuclear Plant (SQN), Units 1 and 2, requested U.S. Nuclear Regulatory Commission (NRC) staff approval of a license amendment that would delete a Note associated with Technical Specification (TS) Surveillance Requirement (SR) 3.8.1.17 to allow the performance of the SR in Modes 1 through 4 when the associated load is out of service for maintenance or testing.
The supplemental letter, dated August 7, 2017, provided additional information that clarified the application, did not expand the scope of the application as originally noticed and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register on July 5, 2017 (82 FR 31102).
The NRC staff reviewed and found the amendment request to be acceptable because it is consistent with the guidance provided in NUREG-1431, "Standard Technical Specifications (STS), Westinghouse Plants," Revision 4. Details of the NRC staff's evaluation are summarized below.
2.0 REGULATORY EVALUATION
=
System Description===
The Class 1 E Alternating Current (AC) Electrical Power Distribution System AC sources consist of the offsite power sources (preferred power sources), and the onsite standby power sources (Train A and Train B emergency diesel generators (DGs)). The design of the AC electrical power system provides independence and redundancy to ensure an available source of power to the Engineered Safety Feature (ESF) systems.
The onsite Class 1 E AC Electrical Power Distribution System is divided into two redundant and independent load groups with two 6.9 Kilovolt (kV) Shutdown Boards in each load group. Each 6.9 kV Shutdown Board has a connection to a preferred offsite power source and a dedicated DG. The 6.9 kV Shutdown Boards in a load group (i.e., 1 A-A and 2A-A, or 1 B-B and 2B-B) are normally powered by the same offsite power circuit. Two DGs associated with one load group can provide all safety related functions to mitigate a loss-of-coolant accident in one unit and safely shut down the other unit. The Train A and Train B ESF systems each provide for the minimum safety functions necessary to shut down the plant and maintain it in a safe shutdown condition.
Applicable Regulatory Requirements The staff applied the below listed regulatory requirements in its review of this License Amendment Request (LAR).
Under Title 1 O of the Code of Federal Regulations (10 CFR), Section 50.90, whenever a holder of a license wishes to amend the license, including TSs in the license, an application for amendment must be filed, fully describing the changes desired. Under 1 O CFR 50.92(a),
determinations on whether to grant an applied-for license amendment are to be guided by the considerations that govern the issuance of initial licenses or construction permits to the extent applicable and appropriate. Both the common standards for licenses and construction permits in 1 O CFR 50.40(a), and those specifically for issuance of operating licenses in 1 O CFR 50.57(a)(3), provide that there must be reasonable assurance that the activities to be authorized will not endanger the health and safety of the public.
The SON was designed to meet the intent of the Proposed General Design Criteria [GDC] for Nuclear Power Plant Construction Permits published in July 1967 (32 FR 10213). The SON construction permit was issued in May 1970. However, in its Updated Final Safety Analysis Report (UFSAR), Section 3.1.2, the licensee stated that it meets the intent of the NRC GDC published as Appendix A to 1 O CFR Part 50 in July 1971, to the extent described in Section 3.1.2 of UFSAR.
In UFSAR Section 3.1.2, the licensee describes how the plant meets the 1 O CFR Part 50, Appendix A, Criterion 17, "Electric power systems." Criterion 17 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 to safety. The safety function for each system (assuming the other system is not functioning) must be to provide sufficient capacity and capability to assure that (1) specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded as a result of anticipated operational occurrences and (2) the core is cooled and containment integrity and other vital functions are maintained in the event of postulated accidents.
The regulation, 1 O CFR 50.36(b), requires that each Part 50 operating license include TSs derived from the facility's safety analysis.
Section 50.36(c)(2) "Limiting conditions for operation," states, in part, that limiting conditions for operation (LCOs) are the lowest functional capability performance levels of equipment required for safe operation of the facility. When an LCO is not met, the licensee must shut down the reactor or follow any remedial action permitted by the TSs until the condition can be met.
Section 50.36(c)(3), "Surveillance requirements," requires that TSs includes SRs, which 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 safety limits, and that the LCOs will be met.
Applicable Regulatory Guidance In its review of this LAR, the NRC staff considered the following guidance.
NUREG-1431, "Standard Technical Specifications (STS), Westinghouse Plants" Revision 4.
The STS SR 3.8.1.18 contains the recommended surveillance to be performed on a DG load sequencer (and corresponds to the SR 3.8.1.17 in the SON TS). A Reviewer Note in the basis for STS SR 3.8.1.18 states that the MODE restrictions for performing this DG load sequencer SR may be deleted subject to satisfying the following criteria:
(a) Performance of the SR will not render any safety system or component inoperable, (b) Performance of the SR will not cause perturbations to any of the electrical distribution systems that could result in a challenge to steady state operation or to plant safety systems, and (c) Performance of the SR, or failure of the SR, will not cause, or result in an AOO
[anticipated operational occurrence] with attendant challenge to plant safety systems.
3.0 TECHNICAL EVALUATION
3.1 Proposed TS Changes
Currently, in SR 3.8.1.17 of each unit, the required surveillance states as follows:
NOTE------------------------------------------------------------------
For DGs 1 A-A and 1 B-B [for Unit 1] [2A-A and 2B-B for Unit 2], this Surveillance shall not normally be performed in MODE 1, 2, 3, or 4. However, this Surveillance may be performed to reestablish OPERABILITY provided an assessment determines the safety of the plant is maintained or enhanced. Credit may be taken for unplanned events that satisfy this SR.
Verify setpoint for each load sequence timer is within +/- 5% of design.
In the LAR, the licensee proposed to delete the above NOTE, so that SR 3.8.1.17 can also be performed in Modes 1 through 4 (without any mode restriction). The licensee requested this change to allow testing, currently required to be performed during plant outages (Mode 5 or 6), to be moved to any Mode of operation or periods with lighter, less complex maintenance schedules and, therefore, make more efficient use of existing LCO allowed outage times.
3.2 Evaluation of the proposed change In the LAR, the licensee stated that it meets the three criteria stated in NUREG-1431 for removal of mode restrictions as follows:
( a) Performance of the SR will not render any safety system or component inoperable.
This criterion is met by doing the work on systems coupled with existing out-of-service conditions.
(b) Performance of the SR will not cause perturbations to any of the electrical distribution systems that could result in a challenge to steady state operation or to plant safety systems.
SQN utilizes individual discrete time delay relays, in lieu of a single load sequencer, to sequentially load the EDGs [DGs] following a loss of offsite power.
Calibration of the time delay relays, associated with the load sequencing, only affects downstream equipment, which will already be out-of-service for testing or maintenance activities. Therefore, there would be no challenge to steady state operation or to plant safety systems.
(c) Performance of the SR does not cause or result in an anticipated operational occurrence with attendant challenge to plant safety systems.
The work will be performed with existing out-of-service conditions so it would not cause or result in an anticipated operational occurrence.
In the LAR, the licensee stated that SQN uses discrete timers that sequentially apply the individual loads to the DG to respond to an accident or loss of offsite power condition (i.e., the discrete timers function in the role of a load sequencer). In its request for additional information (RAI) response dated August 7, 2017, the licensee stated that each shutdown board has an associated "logic panel" located adjacent to the shutdown board. Relays implementing the load sequence scheme are distributed through the switchgear control compartments and the adjacent logic panel. In the event of an accident signal without loss of offsite power (LOOP), the discrete load sequencer timer circuit is not activated. Discrete blackout relays associated with each sequenced load are configured in parallel start circuits for each load. Blackout relays are latching control relays that change state when loss of voltage is detected. The blackout relays enable either an instantaneous start signal if offsite power is available or the discrete load sequencer timer for the associated load when connected to the onsite AC. Based on clarification provided by the licensee, the staff understands that the discrete load timers come into action in case of a LOOP event or a LOOP concurrent with an accident event.
In the LAR, the licensee also stated:
Calibration of the time delay relays, associated with automatic load sequence timers, may defeat certain functions in the downstream equipment energized by the timer, which may require entry into LCO Actions for the associated function.
However, the time required to perform the calibration is typically a small fraction of the LCO Completion Time. Maintenance and testing of the associated functions typically require periodic entry into these same LCO Actions. Coupling the calibrations with existing maintenance or testing does not increase the unavailability of the equipment.
In its RAI response dated August 7, 2017, the licensee further stated that during the performance of the surveillance, testing of a relay only affects the specific safety system component associated with that relay. The relay testing does not affect the associated diesel generator since the testing is performed with the end device out-of-service and tagged under a hold order, per plant procedures. The licensee further stated that performance of relay testing for SR 3.8.1.17 typically requires 2 to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.
The NRC staff finds that justifications provided by the licensee to perform SR 3.8.1.17 during any Mode are reasonable and meet the criteria stated in NUREG-1431, because:
Criterion (a), performance of the SR at any given time will not render any safety system or component inoperable. Because this SR is performed when the downstream affected safety system is already taken out of service and the plant enters the associated LCO, performing this SR does not result in additional safety system or component inoperability. In addition, this SR is performed in a very short period of time (2 to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> versus 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 7 days LCO completion time for the downstream associated equipment). Furthermore, because of the TS Section 1.1 Definition of Operable-Operability, the TS require entering an LCO for the affected safety system or component when its attendant system, such as the sequencing relays associated with a load sequencer, is taken out-of-service the plant for testing or maintenance.
Criteria (b) and (c), the SR would not cause any perturbations to any upstream electrical distribution system or result in an anticipated operational occurrence. Because the SR is performed only on one individual relay at a time, it only affects limited downstream associated equipment (that is already taken out of service). Furthermore, during the SR performance, the relay is isolated from its upstream electrical bus and thus will not cause perturbations to the other plant equipment powered by the electrical bus.
The NRC staff also finds the remainder of the note "However, this Surveillance may be performed to reestablish OPERABILITY provided an assessment determines the safety of the plant is maintained or enhanced" becomes unnecessary after the removal of the Mode restrictions, because this change allows the SR to be performed during power operations. In addition, "Credit may be taken for unplanned events that satisfy this SR" is also not necessary for SQN's design, because actuations of the relay from unplanned events might not satisfy the specific elements of the test procedure for the SR. The staff finds that the deletion of other language in the Note is conservative, and therefore acceptable.
3.3 Summary The staff finds that the proposed change to remove the Mode restrictions on TS SR 3.8.1.17 for SON Units 1 and 2 is consistent with the guidance provided in NUREG-1431. In addition, the change would not impact the licensee's current compliance with the requirements of GDC 17 because the reliability of onsite and offsite electrical power system is not affected. It would also not impact the licensee's current compliance with 1 O CFR 50.36(c)(3) because the necessary qualification of systems and components is still maintained.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Tennessee State official was notified of the proposed issuance of the amendments on November 13, 2017. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes SRs.
The NRC staff has determined that the amendments involve 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 amendments involve no significant hazards consideration, and there has been no public comment on such finding published in the Federal Register on July 5, 2017 (82 FR 31102). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 1 O CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
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 the 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 amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor: V. Goel Date: February 2, 2018
- via memorandum ML172768314
+via e-mail OFFICE NRR/DORL/LPL2-2/PM NRR/DORL/LPL2-2/LA NRR/DE/Eeob/BC*
NRR/DSS/STSB/BC+
NAME AHon BClayton JQuichocho VCusumano DATE 11/09/17 11 /09/17 10/19/17 11 /28/17 OFFICE OGC /NLO NRR/DORL/LPL2-2/BC NRR/DORL/LPL2-2/PM NAME MYoung UShoop (RSchaff for)
AHon DATE 01/31/18 02/02/18 02/02/18