ML14356A012
| ML14356A012 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 12/29/2014 |
| From: | Lyon C Plant Licensing Branch IV |
| To: | Cortopassi L Omaha Public Power District |
| Lyon C | |
| References | |
| TAC MF5143 | |
| Download: ML14356A012 (29) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 29, 2014 Mr. Louis P. Cortopassi Site Vice President and Chief Nuclear Officer Omaha Public Power District Fort Calhoun Station 9610 Power Lane, Mail Stop FC-2-4 Blair, NE 68008
SUBJECT:
FORT CALHOUN STATION, UNIT NO. 1 -ISSUANCE OF AMENDMENT RE:
ONE-TIME EXTENSION OF A LIMITED NUMBER OF TECHNICAL SPECIFICATION SURVEILLANCE REQUIREMENTS (TAC NO. MF5143)
Dear Mr. Cortopassi:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 279 to Renewed Facility Operating License No. DPR-40 for the Fort Calhoun Station, Unit No. 1. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated November 7, as supplemented by letters dated November 21 and December 10 and 19, 2014.
The amendment revises a limited number of TS Surveillance Requirements by adding a note or footnote permitting a one-time extension from a refueling frequency (i.e., at least once per 18 months) to a maximum of 28 months.
A copy of the related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.
Docket No. 50-285
Enclosures:
- 1. Amendment No. 279 to DPR-40
- 2. Safety Evaluation cc w/encls: Distribution via Listserv Sincerely, Carl F. Lyon, Project Manager Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 OMAHA PUBLIC POWER DISTRICT DOCKET NO. 50-285 FORT CALHOUN STATION, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 279 Renewed License No. DPR-40
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Omaha Public Power District (the licensee), dated November 7, as supplemented by letters dated November 21 and December 10 and 19, 2014, 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 license 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, Renewed Facility Operating License No. DPR-40 is amended by changes as indicated in the attachment to this license amendment, and paragraph 3.B. of Renewed Facility Operating License No. DPR-40 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 279, are hereby incorporated in the license. Omaha Public Power District shall operate the facility in accordance with the Technical Specifications.
- 3.
The license amendment is effective as of its date of issuance and shall be implemented within 30 days from the date of issuance.
Attachment:
Changes to the Renewed Facility Operating License No. DPR-40 and Technical Specifications FOR THE NUCLEAR REGULATORY COMMISSION Eric R. Oesterle, Acting Chief Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: December 29, 2014
ATTACHMENT TO LICENSE AMENDMENT NO. 279 RENEWED FACILITY OPERATING LICENSE NO. DPR-40 DOCKET NO. 50-285 Replace the following pages of the Renewed Facility Operating License No. DPR-40 and the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
License Page REMOVE INSERT Technical Specifications REMOVE 3.1-Page 6 3.1 -Page 7 3.1-Page 10 3.1 -Page 13 3.1-Page 14 3.1-Page 23 3.7-Page 1 3.16-Page 16 INSERT 3.1-Page 6 3.1-Page 7 3.1-Page 10 3.1-Page 13 3.1-Page 14 3.1-Page 23 3.7-Page 1 3.16-Page 16 (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 when associated with radioactive apparatus or components; (5)
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 operation of the facility.
- 3.
This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter 1: 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; 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:
A.
Maximum Power Level Omaha Public Power District is authorized to operate the Fort Calhoun Station, Unit 1, at steady state reactor core power levels not in excess of 1500 megawatts thermal (rate power).
B.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 279 are hereby incorporated in the license. Omaha Public Power District shall operate the facility in accordance with the Technical Specifications.
C.
Security and Safeguards Contingency Plans The Omaha Public Power District shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled: "Fort Calhoun Station Security Plan, Training and Qualification Plan, Safeguards Contingency Plan," submitted by letter dated May 19, 2006.
OPPD shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The OPPD CSP was approved by License Amendment No. 266.
Renewed Operating License No. DPR-40 Amendment No. 279
TECHNICAL SPECIFICATIONS TABLE 3-1 (Continued)
MINIMUM FREQUENCIES FOR CHECKS, CALl BRA TIONS AND TESTING OF REACTOR PROTECTIVE SYSTEM Channel Description Surveillance Function Frequency Surveillance Method
- 6.
Steam Generator Level
- a.
Check s
- a.
CHANNEL CHECK
- b.
Test Q(1)
- b.
CHANNEL FUNCTIONAL TEST
- c.
Calibrate R<4>
- c.
CHANNEL CALIBRATION
- 7.
- a.
Check s
- a.
CHANNEL CHECK Pressure
- b.
Test Q(1)
- b.
CHANNEL FUNCTIONAL TEST
- c.
Calibrate R
- c.
CHANNEL CALIBRATION
- 8.
Containment Pressure
- a.
Test Q(1)
- a.
CHANNEL FUNCTIONAL TEST
- b.
Calibrate R
- b.
CHANNEL CALIBRATION
- 9.
Loss of Load
- a.
Test p
- a.
CHANNEL FUNCTIONAL TEST
- 10. Manual Trips
- a.
Test p
- a.
CHANNEL FUNCTIONAL TEST 11. Steam Generator
- a.
Check s
- a.
CHANNEL CHECK Differential Pressure
- b.
Test Q(1)
- b.
CHANNEL FUNCTIONAL TEST
- c.
Calibrate R
- c.
CHANNEL CALIBRATION 3.1-Page 6 Amendment No. 77.163. 182, 2-&7 279
TECHNICAL SPECIFICATIONS TABLE 3-1 (Continued)
MINIMUM FREQUENCIES FOR CHECKS, CALl BRA TIONS AND TESTING OF REACTOR PROTECTIVE SYSTEM Channel Description Surveillance Function Frequency Surveillance Method
- 12. Reactor Protection System Logic Units
- a.
Test Q(1)
- a.
CHANNEL FUNCTIONAL TEST
- 13. Axial Power Distribution NOTES:
- a.
Check:
- 1) Axial Shape Index Indication
- 2) Upper Trip Setpoint Indication
- 3) Lower Trip Setpoint Indication
- b.
Test
- c.
Calibrate s
- a.
- 1) CHANNEL CHECK
- 2) CHANNEL CHECK
- 3) CHANNEL CHECK Q(1)
- b.
CHANNEL FUNCTIONAL TEST R
- c.
CHANNEL CALIBRATION (1)
The quarterly tests will be done on only one of four channels at a time to prevent reactor trip.
(2)
Calibrate using built-in simulated signals. *
(3)
Not required unless the reactor is in the power operating condition and is therefore not required during plant startup and shutdown periods.
(4)
During PLANT OPERATING CYCLE 27 only, the maximum allowed surveillance test interval shall not exceed 28 months.
3.1 -Page 7 Amendment No. 77.122.163. 182, ~
279
TECHNICAL SPECIFICATIONS TABLE 3-2 (continued)
MINIMUM FREQUENCIES FOR CHECKS, CALl BRA TIONS AND TESTING OF ENGINEERED SAFETY FEATURES, INSTRUMENTATION AND CONTROLS Channel Description
- 6.
(continued)
- 7.
Manual Safety Injection Actuation
- 8.
Manual Containment Isolation Actuation
- 9.
Manual Containment Spray Actuation
- 10. Automatic Load Sequencers
- 11. Diesel Testing Surveillance Function
- b.
Test
- c.
Calibrate
- a.
Test
- a.
Check
- b.
Test
- a.
Test
- a.
Test See Technical Specification 3.7 Frequency Q
R R
R<9l R<9l R
Q 3.1-Page 10 Surveillance Method
- b.
- c.
- a.
- a.
- b.
- a.
- a.
CHANNEL FUNCTIONAL TEST Secondary and Electronic Calibration performed at refueling frequency.
Primary calibration performed with exposure to radioactive sources only when required by the secondary and electronic calibration.
CHANNEL FUNCTIONAL TEST Observe isolation valves closure.
CHANNEL FUNCTIONAL TEST CHANNEL FUNCTIONAL TEST CHANNEL FUNCTIONAL TEST Amendment No. 54.111.152.163.173.1 82, 2e7-279
TECHNICAL SPECIFICATIONS TABLE 3-2 (continued)
MINIMUM FREQUENCIES FOR CHECKS 1 CALIBRATIONS AND TESTING OF ENGINEERED SAFETY FEATURES 1 1NSTRUMENTATION AND CONTROLS Channel Description Surveillance Function
- 18. SIRW Tank Temperature
- a.
Check
- b.
Test
- 19. Manual Recirculation
- a.
Test Actuation
- 20. Recirculation Actuation
- a.
Test Logic
- b.
Test
- 21. 4.16 KV Emergency Bus
- a.
Check Low Voltage (Loss of Voltage and Degraded Voltage) Actuation Logic
- b.
Test
- c.
Calibrate
- 22. Manual Emergency Off-site
- a.
Test Power Low Trip Actuation Frequency 0(6)
R R<1o)
Q R(7l(1o) s Q
R R
3.1-Page 13 Surveillance Method
- a.
Verify that temperature is within limits.
- b.
Measure temperature of SIRW tank with standard laboratory instruments.
- a.
CHANNEL FUNCTIONAL TEST
- a.
CHANNEL FUNCTIONAL TEST
- b.
CHANNEL FUNCTIONAL TEST
- a.
Verify voltage readings are above alarm initiation on degraded voltage level - supervisory lights "on".
- b.
CHANNEL FUNCTIONAL TEST (Undervoltage relay)
- c.
CHANNEL CALIBRATION
- a.
CHANNEL FUNCTIONAL TEST Amendment No. 41.153.163.172.182.249,2 279
TECHNICAL SPECIFICATIONS TABLE 3-2 (continued)
MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS AND TESTING OF ENGINEERED SAFETY FEATURES, INSTRUMENTATION AND CONTROLS Channel Description Surveillance Function Frequency Surveillance Method
- a.
Check:
s
- a.
- 24. Manual Auxiliary Feedwater Actuation
- 25. Manual Steam Generator Slowdown Isolation
- 26. Automatic Steam Generator Slowdown Isolation
- 1) Steam Generator Water Level Low (Wide Range)
- 2) Steam Generator Pressure Low
- b.
Test:
- 1) Actuation Logic
- c.
Calibrate:
- a.
- a.
- a.
- 1) Steam Generator Water Level Low (Wide Range)<10>
- 2) Steam Generator Pressure Low
- 3) Steam Generator Differential Pressure High Test Test Test
- 1)
CHANNEL CHECK
- 2)
CHANNEL CHECK QR<7>
- b.
- 1)
CHANNEL FUNCTIONAL TEST R
- c.
- 1)
CHANNEL CALl BRA TION
- 2)
CHANNEL CALIBRATION
- 3)
CHANNEL CALIBRATION R
- a.
CHANNEL FUNCTIONAL TEST R
- a.
CHANNEL FUNCTIONAL TEST R
- a.
CHANNEL FUNCTIONAL TEST NOTES: (1) Not required unless pressurizer pressure is above 1700 psia.
(2) CRHS monitors are the containment atmosphere gaseous radiation monitor and the Auxiliary Building Exhaust Stack gaseous radiation monitor.
(3) Not required unless steam generator pressure is above 600 psia.
(4) QP -Quarterly during designated modes and prior to taking the reactor critical if not completed within the previous 92 days (not applicable to a fast trip recovery).
(5) Not required to be done on a SIT with inoperable level and/or pressure instrumentation.
(6) Not required when outside ambient air temperature is greater than 50°F and less than 1 05°F.
(7) Tests backup channels such as derived circuits and equipment that cannot be tested when the plant is at power.
(8) SGLS is required for containment spray pump actuation only. SGLS lockout relays are not actuated for this test.
(9) During PLANT OPERATING CYCLE 27 only, the maximum allowed surveillance test interval for valves PCV-1849A&B shall not exceed 28 months.
(10) During PLANT OPERATING CYCLE 27 only, the maximum allowed surveillance test interval shall not exceed 28 months.
3.1 -Page 14 Amendment No. 41.54.65.122.163.171.172.182.255.257, ~
279
TECHNICAL SPECIFICATIONS TABLE 3-3A (Continued)
MINIMUM FREQUENCY FOR CHECKS, CALl BRA TIONS AND FUNCTIONAL TESTING OF ALTERNATE SHUTDOWN PANELS (AI-185 AND Al-212)
AND EMERGENCY AUXILIARY FEEDWATER PANEL (AI-179) INSTRUMENTATION AND CONTROL CIRCUITS Surveillance Channel Description Function Frequency Surveillance Method
- 7.
- a. CHECK M
- a.
CHANNEL CHECK LEVEL, WIDE RANGE (AI-179)
- b. CALIBRATE R(1l
- b.
CHANNEL CALIBRATION
- 8.
- a. CHECK M
- a.
CHANNEL CHECK LEVEL, NARROW RANGE (AI-179)
- b. CALIBRATE R<1J
- b.
CHANNEL CALIBRATION
- 9.
- a. CHECK M
- a.
CHANNEL CHECK PRESSURE (AI-179)
- b. CALIBRATE R
- b.
CHANNEL CALIBRATION
- 10. PRESSURIZER PRESSURE
- a. CHECK M
- a.
CHANNEL CHECK (AI-179)
- b. CALIBRATE R
- b.
CHANNEL CALIBRATION
- 11. EAFW CONTROL
- a. TEST R
- a.
CHANNEL FUNCTIONAL TEST CIRCUITS (AI-179)
<1> During PLANT OPERATING CYCLE 27 only, the maximum allowed surveillance test interval shall not exceed 28 months.
3.1-Page 23 Amendment No. 125.182, 2-a+
279
TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE REQUIREMENTS 3.7 Emergency Power System Periodic Tests Applicability Applies to periodic testing and surveillance requirements of the emergency power system.
Objective To verify that the emergency power system will respond promptly and properly when required.
Specifications The following tests and surveillance shall be performed as stated:
(1)
Diesel generators:
- a.
Each diesel engine shall be started at least once per 31 days on a staggered basis.
The engine shall be run with all protective devices operable. The test shall verify that:
- i.
The diesel starts and accelerates to idle speed. Following a warm-up period as recommended by the manufacturer, the diesel generator will be accelerated to rated speed and voltage.
However, at least once per 184 days in these surveillance tests, the diesel generator shall demonstrate that it can be started and accelerated to rated speed and voltage in less than or equal to 10 seconds without a prior warm-up.
The signal initiated to start the diesel shall be varied from one test to another to verify all manual and auto start circuits. (1) ii. With the diesel running at rated speed and voltage, the generator shall be synchronized with the 4.16 KV bus and the diesel breaker manually closed from the electrical control board. The generator shall then be loaded to at least the continuous(2l KW rating and run for at least 60 minutes before being off-loaded and the diesel breaker tripped.
- b.
The auto-start initiating circuit for each diesel shall be tested prior to each plant startup if not done during the previous week.
- c.
Tests shall be conducted during each refueling outage to demonstrate the satisfactory overall automatic operation of each diesel system. This test shall be conducted by:(*)
(*) During PLANT OPERATING CYCLE 27 only, the maximum allowed surveillance test interval shall not exceed 28 months.
3.7-Page 1 Amendment No. 44,444,-MG 279
TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE REQUIREMENTS 3.16 Residual Heat Removal System Integrity Testing Applicability Applies to determination of the integrity of the residual heat removal (RHR) systems and associated components.
Objective To verify that the leakage from the residual heat removal system components is within acceptable limits.
Specifications (1) a. The portion of the shutdown cooling system that is outside the containment, and the piping between the containment spray pump suction and discharge isolation valves, shall be examined for leakage at a pressure no less than 300 psig. This shall be performed on a refueling frequency.
- b. Piping from valves HCV-383-3 and HCV-383-4 to the suction isolation valves of the low pressure safety injection pumps and containment spray pumps and to the high pressure safety injection pumps shall be examined for leakage at a pressure no less that 82 psig.
This shall be performed at the testing frequency specified in (1 )a. above. (1)
- c. The portion of the high pressure safety injection (HPSI) system that is located outside the containment and downstream of the HPSI pumps shall be examined for leakage when subjected to the discharge pressure of a HPSI pump operating in the minimum recirculation mode. This test shall be performed at the frequency specified in ( 1 )a. above.
The leakage contribution from this section shall be the observed leakage from this piping at the test pressure multiplied by the square root of the ratio 1500/P, where P is the test discharge pressure (in psig) of the operating HPSI pump.
- d. An internal leakage test shall be performed on a refueling frequency. The test shall measure and quantify the leakage to the safety injection refueling water tank (SIRwr) from applicable water leakage paths.
- e. Visual inspection of the system's components shall be performed at the frequency specified in (1 )a. above to uncover any significant external leakage to atmosphere (including leakage from valve stems, flanges, and pump seals). The leakage shall be measured by collection and weighing or by any other equivalent method.
(2) a. The sum of leakages from section (1 )a, (1 )b, (1 )c, and (1 )d above shall not exceed 3800 cc/hour.
- b. Repairs shall be made as required to maintain leakage within the acceptable limits.
(1) During PLANT OPERATING CYCLE 27 only, the maximum allowed surveillance test interval shall not exceed 28 months.
3.16-Page 1 Amendment No. 87,122,128,136,157,201,2-W 279
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 279 TO RENEWED FACILITY
1.0 INTRODUCTION
OPERATING LICENSE NO. DPR-40 OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION, UNIT NO. 1 DOCKET NO. 50-285 By application dated November 7, 2014, as supplemented by letters dated November 21 and December 10 and 19, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML14311A158, ML14343A425, ML14344B017, and ML14353A461, respectively), Omaha Public Power District (OPPD) requested changes to the Technical Specifications (Appendix A to Renewed Facility Operating License No. DPR-40) for the Fort Calhoun Station, Unit No. 1 (FCS).
The proposed amendment would revise a limited number of Technical Specification (TS)
Surveillance Requirements (SRs) by adding a note or footnote permitting a one-time extension from a refueling frequency (i.e., at least once per 18 months) to a maximum of 28 months.
These SRs include (1) manual containment isolation actuation, (2) manual recirculation actuation and recirculation actuation logic. (3) steam generator level calibration, (4) visual examination of the high-efficiency particulate air (HEPA) and charcoal filters in the containment recirculating air cooling and filtering system, (5) emergency diesel generators, and (6) residual heat removal system integrity. An extension is requested because these tests will expire before the next refueling outage begins on April 11, 2015.
In the licensee's letter dated December 10, 2014, OPPD withdrew the proposed change to TS 3.6(3)d regarding the visual examination of the HEPA and charcoal filters in the containment recirculating air cooling and filtering system because OPPD successfully completed this examination on December 6, 2014.
The supplemental letters dated November 21 and December 10 and 19, 2014, 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 (NRC) staffs original proposed no significant hazards consideration determination as published in the Federal Register on November 17, 2014 (79 FR 68487).
Specifically, the licensee proposed the following changes:
- 1.
TS 3.1, Table 3-2, Item 8.a and 8.b Add note "During PLANT OPERATING CYCLE 27 only, the maximum allowed surveillance test interval for valves PCV-1849A&B shall not exceed 28 months."
- 2.
TS 3.1, Table 3-2, Items 19.a and 20.b Add note "During PLANT OPERATING CYCLE 27 only, the maximum allowed surveillance test interval shall not exceed 28 months."
- 3.
TS 3.1, Table 3-1, Item 6.c; TS 3.1, Table 3-2, Item 23.c.1; TS 3.1, Table 3-3A, Item 7.b and 8.b Add footnote "During PLANT OPERATING CYCLE 27 only, the maximum allowed surveillance test interval shall not exceed 28 months."
- 4.
TS 3.6(3)d Add footnote "During PLANT OPERATING CYCLE 27 only, the maximum allowed surveillance test interval for the charcoal filters shall not exceed 28 months." In the licensee's letter dated December 10, 2014, OPPD withdrew this proposed change because OPPD successfully completed a visual examination on December 6, 2014, of the HEPA and charcoal filters in the containment recirculating air cooling and filtering system to ensure that leak paths do not exist.
- 5.
TS 3.7(1)c.ii(2)
Add footnote "During PLANT OPERATING CYCLE 27 only, the maximum allowed surveillance test interval shall not exceed 28 months." In the licensee's letter dated December 10, 2014, OPPD clarified the request to apply toTS 3.7(1)c instead of only TS 3.7(1)c.ii(2).
- 6.
TS 3.16(1)b Add footnote "During PLANT OPERATING CYCLE 27 only, the maximum allowed surveillance test interval shall not exceed 28 months."
PLANT OPERATING CYCLE is defined in the FCS TSs as, "The time period from a REFUELING SHUTDOWN to the next REFUELING SHUTDOWN."
2.0 REGULATORY EVALUATION
The regulatory requirements and guidance which the NRC staff considered in its review of the application are as follows:
Title 1 0 of the Code of Federal Regulations ( 1 0 CFR) Part 50 establishes the fundamental regulatory requirements with respect to the domestic licensing of nuclear production and utilization facilities. Specifically, Appendix A, "General Design Criteria [GDC] for Nuclear Power Plants," to 10 CFR Part 50 provides, in part, the necessary design, fabrication, construction, testing, and performance requirements for structures, systems, and components important to safety.
GDC 13, "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.
GDC 18, "Inspection and testing of electric power systems," requires that electric power systems that are important to safety must be designed to permit appropriate periodic inspection and testing of important areas and features to assess the continuity of the systems and the condition of their components.
GDC 20, "Protective system functions," requires the protection system be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.
GDC 21, "Protection system reliability and testability," requires that the protection system be designed for high functional reliability and in service testability, with redundancy and independence sufficient to preclude loss of the protection function from a single failure and preservation of minimum redundancy despite removal from service of any component or channel.
GDC 22, "Protection system independence," requires that the protection system be designed so that the effects of natural phenomena, and of normal operating, maintenance, testing, and postulated accident conditions on redundant channels do not result in loss of the protection function.
The regulation in 10 CFR 50.36, "Technical specifications," which states, in part, that "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." Specifically, 10 CFR 50.36(c)(2)(ii) sets forth four criteria to be used in determining whether a limiting condition for operation is required to be included in the TS. Also, 10 CFR 50.36(c)(3), "Surveillance requirements," states that SRs 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 limiting conditions for operation will be met.
10 CFR 50.65, "Requirements for monitoring the effectiveness of maintenance at nuclear power plants," requires licensees to monitor or control through preventive maintenance the performance or condition of structures, systems, or components in a manner sufficient to provide reasonable assurance that these structures, systems, and components are capable of fulfilling their intended functions. The objective of preventing failures of structures, systems, and components through maintenance must be appropriately balanced against the objective of minimizing unavailability of structures, systems, and components due to monitoring or preventive maintenance.
3.0 TECHNICAL EVALUATION
In its application, the licensee stated, in part, that An extension is necessary because even with the maximum allowable 25%
extension of the surveillance interval permitted by TS 3.0.1, these tests will expire before the next refueling outage (RFO) begins on April 11, 2015.
The NRC staff reviewed each of the proposed TS changes individually to ensure that they continue to comply with the regulations listed in Section 2.0 above. The following are evaluations for each of these TS changes in the order they were presented in the license amendment request.
3.1 Manual Containment Isolation Actuation (i.e.. TS 3.1. Table 3-2. Item 8.a and 8.b)
This proposed change would add a note to the TS which would allow the associated surveillance intervals to be extended from a refueling interval of 18 months to 28 months. The note includes the condition that this extended interval only applies to plant operating cycle 27.
Justification for allowing this interval extension was provided by the licensee. Components tested under the affected surveillance test (IC-ST-IA-0001) are the containment instrument air pressure switches and the containment instrument air isolation valves. IC-ST-IA-0001 was last performed on March 15, 2013, and will require an interval extension to allow uninterrupted plant operation during cycle 27.
Two other surveillance tests, OP-ST-ESF-0013 and OP-ST-ESF-0011, were cited as providing alternate means of partially satisfying the TS SRs. The first of these tests, OP-ST-ESF-0013, includes testing of the Containment Pressure High Signal for automatic engineered safeguards initiation which actuates the Steam Generator Isolation Signal. The NRC staff finds that it is acceptable to assign partial credit for this test because it initiates a containment pressure high signal. The test does not, however, exercise the pressure switches associated with the containment instrument air isolation on loss of containment instrument air pressure nor does this test exercise the containment isolation valves to ensure containment instrument air isolation.
This test was last performed on September 20, 2013. The second test, OP-ST-ESF-0011, is the Steam Generator Isolation Signal isolation test. This test ensures the containment instrument air isolation valves properly close to isolate instrument air to containment; however, this test does not verify the setpoints for the associated pressure switches. This test was last performed on June 22, 2013.
Based on the satisfactory completion of these related surveillance tests and on the demonstrated past performance of the affected components, the NRC staff concludes that there is reasonable assurance that these components will remain operable during the extended surveillance interval period and, therefore, that the requested extension is acceptable.
3.2 Manual Recirculation Actuation and Recirculation Actuation Logic (i.e..
TS 3. 1. Table 3-2. Items 19.a and 20. b. respectively)
This proposed change would add a note to the TS which would allow the associated surveillance intervals to be extended from a refueling interval of 18 months to 28 months. The note includes the condition that this extended interval only applies to plant operating cycle 27.
Justification for allowing this interval extension was provided by the licensee. Components tested under the affected surveillance test (OP-ST-ESF-0019) are the valves receiving a recirculation actuation signal (RAS) and two low-pressure safety injection (LPSI) pump breakers. OP-ST-ESF-0019 was last performed on February 20, 2013, and will require an interval extension to allow uninterrupted plant operation during cycle 27.
Two other surveillance tests, OP-ST-ESF-0011 and OP-ST-ESF-0002, were cited as providing an alternate means of partially satisfying the TS SRs. The first of these tests, OP-ST-ESF-0011, includes a channel A and B automatic and manual engineered safeguard actuation signal test which verifies that the same test switches are actuated and that the same lockout relays are tripped as those tested under surveillance OP-ST-ESF-0019. This test does not test the LPSI pump breaker trip Functions; however, the lockout relays that trip the breakers are tested. This surveillance was last performed on June 22, 2013. The second test, OP-ST-ESF-0002, verifies that the breakers for the LPSI pumps cycle open. This test was last performed in June 2013.
Based on the satisfactory completion of these related surveillance tests and on the demonstrated past performance of the affected components, the NRC staff concludes that there is reasonable assurance that these components will remain operable during the extended surveillance interval period and, therefore, that the requested extension is acceptable.
3.3 Steam Generator Level Calibration (i.e., TS 3.1. Table 3-1, Item 6.c; TS 3.1. Table 3-2. Item 23.c.1; TS 3.1. Table 3-3A Item 7.b and 8.b)
This proposed change would add a footnote to the TS which would allow the associated surveillance intervals to be extended from a refueling interval of 18 months to 28 months. The note includes the condition that this extended interval only applies to plant operating cycle 27.
Justification for allowing this interval extension was provided by the licensee. Components tested under the affected surveillance tests are as follows:
IC-ST-MS-0018-Steam Generator RC-2A Narrow Range Level instrument A/L-901. In its letter dated November 21, 2014, the licensee stated, in part, that "The extension proposed by LAR 14-10 applies only to testing of components located in containment."
IC-ST-MS-0005-Steam Generator RC-2A Wide Range Level instrument D/L-911. In its letter dated November 21, 2014, the licensee stated, in part, that "The extension proposed by LAR 14-10 applies only to testing of components located in containment."
IC-ST-MS-0034-Steam Generator RC-2A Narrow Range Level instrument L-903Y-1. This surveillance includes testing of a loop power supply and an indicator which are located outside of containment. The requested interval extension applies to these components of the level instrument and additional justification was provided by the licensee in its response to RAI No. 2 in the letter dated November 21, 2014. This justification included data from completed surveillances, graphical and tabular data analysis indicating no outliers and no adverse trends, and performance of monthly channel checks per OP-ST-ASP-0001, "Alternate Shutdown Capability Instrumentation Functional Check."
IC-ST-MS-0035-Steam Generator RC-28 Narrow Range Level instrument L-906Y-1. This surveillance includes testing of a loop power supply and an indicator which are located outside of containment. The requested interval extension applies to these components of the level instrument and additional justification was provided by the licensee in its response to RAI No. 2 in the letter dated November 21, 2014. This justification included data from completed surveillances, graphical and tabular data analysis indicating no outliers and no adverse trends, and performance of monthly channel checks per OP-ST-ASP-0001, "Alternate Shutdown Capability Instrumentation Functional Check."
Based on the demonstrated past performance of the loop components to which this surveillance interval extension applies, and on the capability for ongoing channel checks performed on these instruments to detect instrument malfunctions, the NRC staff concludes that there is reasonable assurance that these components will be maintained in an operable condition during the extended surveillance interval period and, therefore, that the requested extension is acceptable.
3.4 Emergency Diesel Generators (i.e.. TS 3.7(1 )c)
This proposed change would add a footnote to the TS which would allow the associated surveillance intervals to be extended from a refueling interval of 18 months to 28 months. The note includes the condition that this extended interval only applies to plant operating cycle 27.
Background
According to FCS Updated Safety Analysis Report (USAR), Chapter 8.4, "Electrical Systems-Emergency Power Sources," two emergency diesel generators (EDGs) are installed and designed to furnish reliable in-plant alternating current (AC) power adequate for safe shutdown and operation of engineered safeguards loads when no energy is available from the 345 kiloVolt (kV) or 161 kV offsite electrical power systems.
Each EDG is connected to one of the two 4.16 kV systems' safety buses 1A3 and 1A4. The engineered safeguards and other essential auxiliaries are divided between the 4.16 kV systems.
The division of loads is such that operation of either system alone provides minimum Engineered Safeguards required for the design basis accident conditions. Each EDG is rated at 4.16 kV, 0.8 power factor lagging, 60 Hertz for 2000-hour kilowatt (HR KW) for the engine (USAR Table 8.4-1). The capacity of each EDG is adequate to support the operation of required engineered safeguards loads under the most restrictive design-basis accident (DBA) from initiation through long-term post-accident cooling. There is no bus-tie breaker between 4.16 kV buses 1A3 and 1A4, and interlocks prevent the interconnection of the diesels auxiliaries at the 480 Volt (V) level. These provisions ensure the two systems supplying engineered safeguards buses and loads are operated independently.
Transfer of the 4.16 kV buses 1A3 and/or 1A4 to the respective EDGs is automatic upon loss-of-offsite power. All loads connected to buses 1A3 and 1A4 are shed except the LPSI pump (if manually started prior to the loss of offsite power) and 4160/480 V transformers supplying 480 V buses before the EDGs breakers can close. All motors connected directly to the 480 V buses are load shed before closure of the EDGs breakers. Upon breaker closure, the system is reloaded manually if no accident signal is present. In the event of a pressurizer pressure low signal (PPLS) or containment pressure high signal (CPHS), the resulting safety injection actuation signal (SIAS) initiates shedding of selected non-essential loads supplied from the 480 V motor control centers. Selected vital loads are sequentially and automatically reconnected by the automatic load sequencers. In the event of a PPLS and CPHS, all vital loads associated with the automatic load sequencers are sequentially reconnected.
TS 3.7(1) provides SRs for the EDG periodic tests.
Current TS SR 3.7(1)c states:
(1)
Diesel generators:
- c.
Tests shall be conducted during each refueling outage to demonstrate the satisfactory overall automatic operation of each diesel system. This test shall be conducted by:
Technical Evaluation
- i.
Initiation of a simulated auto-start signal to verify that the diesel starts.
ii.
Initiation of a simulated simultaneous loss of 4.16 KV supplies to bus 1A3 (1A4). Proper operation will be verified by observation of:
(1)
De-energization of bus 1A3 (1A4),
(2)
Load shedding from bus (both 4160 V and 480 V),
(3)
Energization of bus 1A3 (1A4),
(4)
Automatic sequence start of emergency load, and (5)
Operation of~ 5 minutes while its generator is loadedwith the emergency load.
iii.
Verification that emergency loads do not exceed the 2000-HR KW rating of the engine.<2l In the application dated November 7, 2014, the licensee proposed to add a footnote to TS SR 3. 7(1 )c.ii(2) permitting a one-time extension to the surveillance frequency from 18 months to a maximum of 28 months. The licensee stated that SR 3.7(1)c.ii(2) is performed in surveillance test OP-ST-ESF-0015, "480 Volt Load Shed." OP-ST-ESF-0015 ensures that the main turbine electrohydraulic control (EHC) pumps, the hydrogen purge blowers, and the waste disposal system respond properly to the initiation of an SIAS. The load shed is initiated by manual actuation of lockout relays of the SIAS circuitry. OP-ST-ESF-0015 was last performed satisfactorily on February 15, 2013, due to the extended shutdown of FCS for the refueling outage (RFO) that ended in December 2013. Using the maximum allowable 25 percent grace period of the surveillance interval permitted by TS SR 3.0.1, OP-ST-ESF-0015 must be performed again by January 2, 2015 (22.5 months). However, the next RFO is scheduled to begin in April 2015. Thus, the due date for the performance of OP-ST-ESF-0015 needs to be extended in order for the test to be performed during the 2015 RFO. The licensee also stated that the surveillance test OP-ST-ESF-0002, "Diesel Generator No. 1 and No. 2 Auto Operation," verifies actuation of the same load shed relays and therefore provides additional assurance for the surveillance frequency extension request.
On November 25, 2014, the NRC staff held a conference call with the licensee to clarify the request in the amendment. During the call, the licensee stated that OP-ST-ESF-0002 meets the requirements of SR 3. 7(1 )c.ii by simulating simultaneous loss of power to both 4.16 kV buses 1A3 and 1A4 and observing the five sequence of events in SR 3.7(1)c.ii.(1)-(5).
However, the OP-ST-ESF-0002 testing procedure includes a supplemental part (i.e.,
OP-ST-ESF-0015) that verifies loads are shed from the 480 V bus at the initiation of the SIAS.
The licensee also stated that OP-ST-ESF-0002 does not need an extension, only OP-ST-ESF-0015 does. The staff requested the licensee to clarify the need for an amendment if OP-ST-ESF-0002 meets the requirements of SR 3.7(1)c.ii.
On December 2, 2014, the licensee provided additional information regarding the surveillance tests (ADAMS Accession No. ML14337A089). The licensee clarified that OP-ST-ESF-0002 meets the requirements of SR 3.7(1)c.ii by simulating an accident (PPLS) signal coincident with a loss-of-offsite power and verifying that 480 V load shed occurs to the extent that it is necessary to support the operation of the sequencers with emergency loads supplied by offsite power and by the EDGs. The load shedding of the 480 V loads including the EHC pumps, hydrogen purge blowers, and waste disposal components is verified by observing that the 480 V load shed relays change state as a result of the simulated accident signal. The 4.16 kV load shed is verified by the observation that the EDG output breaker, which is interlocked with the 4.16 kV loads, closes. Successful operation of the EDGs during sequencing verifies that load shed is functioning properly. The licensee further clarified that OP-ST-ESF-0015 fulfills the requirement of TS 3.7(1)c.ii.(2) associated with the Auxiliary Building and Turbine Building portions of the test (i.e., EHC pumps, hydrogen purge blowers, and waste disposal components). This surveillance test verifies that these 480 V loads respond properly to the initiation of the SIAS and is performed by manual tripping of the SIAS lockout relays without having to start the respective EDG. This provides redundant testing of the 480 V load shed relays to satisfy the requirements of NRC Generic Letter (GL) 96-01, "Testing of Safety Related Logic Circuits," dated January 10, 1996 (ADAMS Accession No. ML031110002). GL 96-01 requires the licensee to ensure that all portions of the logic circuitry including interlocks, bypass circuits, relay contacts and other relevant electrical components performing a safety function in the logic circuits, associated with the EDG load shedding and sequencing and the engineered safety features (ESF) systems actuation logic are adequately covered in the surveillance procedures to fulfill the TS requirements. OP-ST-ESF-0015 provides testing for the load shed relays and the SIAS lockout relays within the ESF system for the portion of 3.7(1 )c.ii.(2) that ensures that certain 480 V loads respond properly to the SIAS.
During the conference call held with the licensee on December 4, 2014, the NRC staff informed the licensee that since OP-ST-ESF-0015 is necessary to fulfill the requirement of TS 3.7{1)c.ii.(2) as stated in the December 2, 2014, response, both OP-ST-ESF-0002 and OP-ST-ESF-0015 must be completed to satisfy the requirements of 3.7(1)c.ii. In addition, the EDG must be started as required by 3.7(1)c.i to complete the portion of 3.7(1)c.ii tested by OP-ST-ESF-0002. Therefore, the licensee's request should be a one-time 28-month extension for SR 3.7(1)c, rather than SR 3.7(1)c.ii.(2). NRC GL 91-04, "Changes in Technical Specification Intervals to Accommodate a 24-month Fuel Cycle," dated April2, 1991 (ADAMS Accession No. ML013100215), provides guidance for extending the time limit for completing relevant surveillances from the current 18 months to 24 months, for a maximum interval of 30 months, including the 25 percent grace period. During the call, the licensee clarified that OP-ST-ESF-0002 and OP-ST-ESF-0015 completion dates are out of sequence because some relays were replaced and tested by OP-ST-ESF-0002 during the extended spring RFO.
By letter dated December 10, 2014, the licensee supplemented the application with a request to extend the surveillance test interval for SR 3.7(1)c to 28 months to allow the OP-ST-ESF-0015 to be performed during the April 2015 RFO. The proposed change would add the footnote
"(*)During PLANT OPERATING CYCLE 27 only, the maximum allowed surveillance test interval shall not exceed 28 months" to SR 3.7(1 )c. The licensee provided the evaluation for the requested change in accordance with GL 91-04.
The licensee clarified in an e-mail dated December 12, 2014 (ADAMS Accession No. ML14352A013), that the main portion of 3.7(1)c.i. is covered in OP-ST-ESF-0002, and 3.7(1)c.iii. is covered in SE-ST-DG-0001, "Verification of Diesel Generator Loads."
OP-ST-ESF-0002 and SE-ST-DG-0001 were completed in June 2013 and October 2013, respectively. These two tests are within the required surveillance intervals and do not require an extension to comply with the TS.
Although the footnote was added to SR 3.7(1)c to include all the sub-parts of the SR, the NRC staff evaluated the request for the portion of SR 3.7(1 )c.ii. that is tested by OP-ST-ESF-0015, since it is the only surveillance test that requires an extension to complete the performance of 3.7(1)c during the April2015 RFO.
The GL 91-04 requires the licensee to: (1) evaluate the effects on safety due to the change to a 24-month fuel cycle; (2) confirm that historical maintenance and surveillance data support the conclusion that the effect of the proposed changes on safety is small; and (3) confirm that the performance of surveillances at the bounding surveillance interval limit provided to accommodate a 24-month fuel cycle would not invalidate any assumption in the plant licensing basis The licensee stated that OP-ST-ESF-0015 does not involve instrument setpoint or measurement, thus instrument drift is not a factor. The load shed relays (General Electric (GE)
CR120A and HFA) and the SIAS lockout relays (GE HEA) that are tested change state from normally de-energized to energized and vice versa during the test. The relay subcomponents (relay coils and relay contacts) are also tested during the test. The duration of relay actuation has no impact on the EDG loading/unloading. This ensures that the EDGs will not be overloaded when loads are sequenced onto the bus. In addition, the ESF system employs derived signal actuation which ensures that automatic actuation of one channel would result in actuation of the redundant channel. The actuation logic for each load shed circuit is designed to provide complete load shed as required by SR 3.7(1)c.ii.(2). Manual actuation of either SIAS lockout relays, if required, would provide complete load shed. The licensee confirmed in an e-mail dated December 18, 2014 (ADAMS Accession No. ML14352A424), that the same load shed relays (GE CR120A and GE HFA) were satisfactorily tested by OP-ST-ESF-0002 in the last extended RFO. Thus, the effect on safety is small as testing ensures that the EDGs will be available to mitigate a degraded voltage condition during a DBA.
The licensee also stated that a review of the results from the past four tests, completed in accordance with OP-ST-ESF-0015, found that the load shed and lockout relays performed as expected and provided immediate load shed. In addition, the NRC staff did a historical review of GE CR120A, HFA, and HEA relays which determined that these relays have performed reliably over the years. Thus, the one-time extension of the testing interval of these components from 18 months to 28 months does not adversely affect the safety of the 480 V load shed circuitry due to the proven reliability of these components as a result of years of use at FCS.
According to FCS USAR Section 7.3, "Engineered Safeguards Controls and Instrumentation,"
the principal operations that occur coincident with a SIAS include containment isolation, automatic start of the EDGs, and sequential starting of ESF equipment. Automatic start of the EDGs and sequencing of the ESF loads have been successfully tested by OP-ST-ESF-0002 and OP-ST-ESF-0015 over the past several years. Additionally, relays that have been determined to have reached their end of service life have been replaced so as to ensure their continued reliability. This ensures that the Engineered Safeguards Control System will perform its design basis function and automatically actuate ESF equipment and essential support systems. Thus, the proposed change does not invalidate assumptions or commitments contained in the plant licensing basis.
The NRC staff reviewed the evaluation provided by the licensee for the one-time extension of the testing interval to 28-months for the portion of SR 3.7(1)c.ii tested by OP-ST-ESF-0015 and finds that the request is consistent with the GL 91-04 guidance. Furthermore, the remainder of the tests required by SR 3.7(1)c were performed satisfactorily and are within the required test interval. Therefore, the staff concludes that the one-time extension of the testing interval for 3.7(1)c from 18 months to 28 months is acceptable.
Conclusion forTS 3.7(1)c Based on the above evaluation, the NRC staff concludes that the proposed changes to TS SR 3.7(1)c surveillance intervals from 18 months to 28 months is consistent with GL 91-04 and will not impact the licensee's compliance with the regulatory requirements of GDC 18, 10 CFR 50.36(c), and 10 CFR 50.65. Therefore, the staff concludes that this one-time change is acceptable.
3.5 Residual Heat Removal {RHR) System Integrity (i.e.. TS 3.16{1)b)
This proposed change would add a footnote to the TS which would allow the associated surveillance intervals to be extended from a refueling interval of 18 months to 28 months. The note includes the condition that this extended interval only applies to plant operating cycle 27.
Background
As described in USAR Section 9.3, "Auxiliary Systems-Shutdown Cooling System," the limiting leakage to atmosphere from the RHR system (3800 cubic centimeters per hour; cc/hr) is based upon a plant-specific leak rate analysis for RHR system components operating after a DBA.
The test pressures in sections 3.16(1 )a and 3.16(1 )b, and the pressure correction factors in sections 3.16(1)c give adequate margins over the highest pressures within the lines after a DBA. A RHR system leakage of 3800 cc/hr will limit offsite exposures due to leakage to insignificant levels relative to those calculated for direct leakage from the containment in the DBA.
As stated by the licensee in its application, TS 3.16(1 )b surveillance requires piping from the containment's sump isolation valves HCV-383-3 and HCV-383-4 to the suction isolation valves of the LPSI pumps, containment spray pumps, and the high-pressure safety injection pumps to be examined for leakage at a pressure not less than 82 pounds per square inch gauge (psig).
To satisfy this requirement, SE-ST-SI-3027 performs a hydrostatic pressure test on the RHR piping between safety injection and refueling water tank (SIRWT) Sl-5 isolation valves, LCV-383-1 and LCV-383-2, and containment sump isolation valves, HCV-383-3 and HCV-383-4.
TS 3.16(2)a surveillance consists of a series of leakage tests which measure and total observed leakage from the RHR headers. Per the TS, the total leakage shall not exceed 3800 cc/hr. To comply, in part, with TS 3.16(2)a, SE-ST-SI-3027 measures seat leakage across LCV-383-1 and LCV-383-2 as well as back leakage across RHR header check valves, Sl-139 and Sl-140.
The quantified leakage found in SE-ST-SI-3027 is combined with the leakages measured from SE-ST-SDC-3003, and QC-ST-SI-3008 to verify TS 3.16(2)a compliance.
The surveillance was last performed satisfactorily on April 26, 2013. The test was performed twice during the extended outage with the previous test being completed in May 2011. The next required due date including the 25 percent allowed extension is March 14, 2015. The surveillance performs a hydrostatic pressure test on the RHR headers "A" and "8" piping between SIRWT Sl-5 isolation valves, LCV-383-1 and LCV-383-2, and containment sump isolation valves, HCV-383-3 and HCV-383-4. The test also measures seat leakage across LCV-383-1 and LCV-383-2 as well as back leakage across RHR header check valves, Sl-139 and Sl-140. This testing ensures pressure boundary integrity of the RHR piping and quantifies leakage past LCV-383-1, LCV-383-2, Sl-139, and Sl-140 to ensure that the combined leakage from components complies with TS 3.16(2)a.
The test aligns safety injection equipment such that the system cannot perform its normal safety functions and can only be performed when the refueling cavity is filled with water from the SIRWT and the SIRWT itself is empty. Therefore, this test must be performed when the plant is in Refueling Shutdown (Mode 5).
Technical Evaluation The RHR headers have primarily been in standby since Cycle 26 operations and the associated valves and piping see little active wear or degradation. A search of condition reports and work orders for valves LCV-383-1, LCV-383-2, Sl-139, and Sl-140 found no relevant recent history.
Given the satisfactory recent performance of the subject valves and with the knowledge that the system has operated mostly in standby throughout operating Cycle 26, there is reasonable assurance that all tested components will continue to perform well until beyond the scheduled outage start date. Additional assurance is provided by the requirement that primary system leakage must be monitored and maintained constantly in accordance with TS 2.1.4, "Reactor Coolant System Leakage Limits."
Conclusion forTS 3.16(1)b Surveillance tests performed prior to startup demonstrate that the RHR piping is robust and that the pressure boundary integrity of the RHR piping will be maintained to ensure that the combined leakage complies with TS 3.16(2)a. In addition, primary system leakage is monitored continuously in accordance with TS 2.1.4. Therefore, the NRC staff concludes that the proposed one-time surveillance extension to SR 3.16(1)b is acceptable.
4.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION
DETERMINATION The Commission published notice of this proposed license amendment in the Federal Register on November 17, 2014 (79 FR 68487) for a 30-day comment period and a 60-day request for hearing period. According to 10 CFR 2.1 05(a)(4)(i), the Commission may issue a license amendment before the expiration of the 60-day request for hearing period if the Commission determines that the amendment involves no significant hazards consideration. In this instance, some of the SRs that are requested to be extended would expire before the expiration of the 60-day request for hearing period as well as before the next refueling outage. This would require a plant shutdown before the next refueling outage in order to perform the SRs. Since this would be an unnecessary plant transient, the NRC staff is making a final no significant hazards consideration determination in order to issue the amendment prior to the expiration of the 60-day request for hearing period.
The Commission's regulations at 10 CFR 50.92(c) state that the Commission may make a final determination that a proposed amendment involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not: ( 1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. As required by 10 CFR 50.91 (a), the licensee has provided its analysis of the issue of no significant hazards consideration which is presented below.
- 1.
Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The requested action is a one-time extension to the performance interval of certain TS surveillance requirements. The performance of the surveillances, or the failure to perform the surveillances, is not a precursor to an accident. Performing the surveillances or failing to perform the surveillances does not affect the probability of an accident.
Therefore, the proposed delay in performance of the surveillance requirements in this amendment request does not increase the probability of an accident previously evaluated.
A delay in performing the surveillances does not result in a system being unable to perform its required function. Additionally, the defense-in-depth of the system design provides additional confidence that the safety function is maintained. In the case of this one-time extension request, the relatively short period of additional time that the systems and components will be in service before the next performance of the surveillance will not affect the ability of those systems to operate as designed. Therefore, the systems required to mitigate accidents will remain capable of performing their required function.
No new failure modes have been introduced because of this action and the consequences remain consistent with previously evaluated accidents.
Therefore, the proposed delay in performance of the surveillance requirement in this amendment request does not involve a significant increase in the consequences of an accident.
- 2.
Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed amendment does not involve a physical alteration of any system, structure, or component (SSC), or a change in the way any sse is operated. The proposed amendment does not involve operation of any SSCs in a manner or configuration different from those previously recognized or evaluated. No new failure mechanisms will be introduced by the onetime surveillance extension being requested.
Therefore, the proposed change does not create the possibility of a new or different kind, of accident from any previously evaluated.
- 3.
Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No.
The proposed amendment is a one-time extension of the performance-interval of certain TS surveillance requirements. Extending the surveillance requirements does not involve a modification of any TS Limiting Conditions for Operation. Extending the surveillance frequency does not involve a change to any limit on accident consequences specified in the license or regulations. Extending the surveillance frequency does not involve a change to how accidents are mitigated or a significant increase in the consequences of an accident. Extending the surveillance frequency does not involve a change in a methodology used to evaluate the consequences of an accident. Extending the surveillance frequency does not involve a change in any operating procedure or process.
The systems and components involved in this request have exhibited reliable operation based on the results of the most recent performances of their 18-month surveillance requirements and the associated functional surveillances. Based on the limited additional period of time that the systems and components will be in service before the surveillance is next performed, as well as FCS operating experience provides reasonable assurance these surveillances will be successful when performed. Thus, it is reasonable to conclude that the margin of safety associated with the surveillance requirement will not be affected by the requested extension.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's no significant hazards consideration analysis.
Based on this review and on the NRC staff's safety evaluation of the underlying license amendment request, the NRC staff concludes that the three standards of 10 CFR 50.92(c) are satisfied. Accordingly, the NRC staff makes a final determination that no significant hazards consideration is involved.
5.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Nebraska State official was notified of the proposed issuance of the amendment. The State official had no comments.
6.0 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 made a final determination that the amendment involves no significant hazards consideration. 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.
7.0 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 Contributors: R. Stattel, NRR/DE/EICB A. Foli, NRR/DE/EEEB F. Lyon, NRR/DORLILPL4-1 Date: December 29, 2014
- memo dated
- email dated OFFICE NRR/DORL/LPL4-1/PM NRR/DORL/LPL4-1/LA NRR/DSS/STSB/BC NRR/DE/EICB/BC NAME FLyon JBurkhardt REIIiott MChernoff for**
JThorp*
DATE 12/22/14 12/22/14 12/23/14 12/10/14 OFFICE NRR/DE/EEEB/BC OGC NRR/DORLILPL4-1/BC(A)
NRR/DORLILPL4-1/PM NAME JZimmerman*
JWachutka** NLO EOesterle FLyon DATE 12/19/14 12/24/14 12/24/14 12/29/14