ML20078R442
| ML20078R442 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 11/07/1983 |
| From: | Mills L TENNESSEE VALLEY AUTHORITY |
| To: | Adensam E Office of Nuclear Reactor Regulation |
| References | |
| GL-83-28, NUDOCS 8311150111 | |
| Download: ML20078R442 (39) | |
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TENNESSEE VALLEY AUTHORITY CHATTANOOGA. TENNESSEE 374ot 6
400 Chestnut Street Tower II November 7, 1983 Director of Nuclear Reactor Regulation Attention:
Ms. E. Adensam, Chief Licensing Branch No. 4 Division of Licensing U.S. Nuclear Regulatory Coinnission Washington, D.C.
20555
Dear Ms. Adensam:
In the Matter of
)
Docket No.
50-327 Tennessee Valley Authority
)
50-328 Enclosed is our response to D. G. Eisenhut's July 8, 1983 letter to "All Licensees of Operating Reactors,... regarding the required actions based on the generic implications of the Salem ATWS events. (Generic Letter 83-28). The enclosed information provides the status of TVA's conformance with the positions identified by Generic Letter 83-28. Accordingly, this information is being provided pursuant to 10 CFR 50.54(f) and the operating license. for our Sequoyah Nuclear Plant should not be modified, suspended, or revoked.
If you have any questions concerning this matter, please get in touch with J. E. Wills at FTS 858-2683.
Very truly yours, 8311150111 831107 PDR ADOCK 05000327 h
L. M. Mills, knager Nuclear Licensing d subscr ed before me Sworn t g-dayof me, 1983 this si s1/s Nota? Public My C ssion Expires >/
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Enclosure cc:
U.S. Nuclear Regulatory Commission (Enclosure) gg Region II Attn:
Mr. James P. O'Reilly Administrator f
f 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30303 1983-TVA SOTH ANNIVERSARY An Equal Opportunity Employer
4 ENCLOSURE RESPONSE TO D. G. EISENHUT'S JULY 8, 1983 LETTER TO "ALL LICENSEES..."
REQUIRED ACTIONS BASED ON GENERIC IMPLICATIONS OF SALEM ATWS EVENTS (GENERIC LETTER 83-28)
SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2
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l Scqu yah Nucisar Plant 1.1 - Pret-Trip Rrview (Program Dxcriptien and Proordure)
The Division of Nuclear Power (NUC PR) has delineated the requirements O
for the reactor scram and turbine trip reports through its division procedures. All operating plants are required to have procedures which as a minimum meet the guidelines setforth in the corporate procedures.- The comments contain under the following items as outlined by the Generic Letter describes the program, procedures, and methods used at Sequoyah Nuclear Plant (SQN) to perform a post trip
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review.
1.1.1 Criteria for Determining the Acceptability of Restart
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The plant trip report is completed by the Unit Operator (UO),
Assistant Shift Engineer ( ASE), or Shift Technical Advisor l
(STA), following any unscheduled reactor or turbine trip. This procedure provides a step-by-step guidance to assist the user in ensuring that all safety systems actuated and operated correctly. After completion of the report, which includes all pertinent charts such as the trip sequence of events, the report is reviewed by the shift engineer (SE) and STA to verify that'all systems operated correctly. Once the SE is satisfied I
that all systems operated as expected and the root cause of the trip has been determined and corrected, he has the authority to authorize unit restart. This is now documented in the General Operating Instructions. If the root cause cannot be determined or unexpected operations occurred during the trip, additional discussions or investigations by plant personnel and/or plant management will be initiated to resolve the problem and make recommendations to the SE.
1.1.2 The Responsibilities and Authorities of Personnel Who Will Perform the Review and Analysis of These Events The primary personnel involved in post-trip review are the SE,
.L ASE, UO, and STA. These personnel review the plant parameters and the chain of events that resulted in the trip; however, it is the SE who has the authority for declaring the unit safe for l
restart after the trip cause is known and a safety review performed. Available to the SE are plant maintenance and engineering sections to assist in any investigation into the trip that the SE may determine necessary.
1.1 3 The Necessary Qualifications and Training for the Responsible Personnel l
All personnel involved in the primary ircle of event assessmer.t are licensed operators by NRC except for the STA. The STA is a degreed engineer who has received plant specific training. All 1
of these personnel have been trained in a systematic safety assessment approach to reactor trips including simulator i
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. training. In addition, TVA is also in the process of formally training various management and engineering personnel at the e'
plant in system operations.
1.1.4 The Sources of Plant Information Necessary To Conduct the Review and Analysis.
There are many sources that the Operations personnel have available to use in a post-trip analysis. The present pro-cedure requires the attaching of the charts from steam gener-ator level, feedwater inlet flow and steam flow, nuclear instru-mentation, pressurizer pressure, pressurizer level, overpower /
over temperature and delta temperature, turbine reference pressure, reactor average temperature, turbine speed / governor valve.positon, sequence of events recording, and post-morten review program recording.
In addition, interviews with personnel who were directly, involved with the trip are assimulated with the observations from the UO to provide a narrative discussion of the event in the report.
This combined information allows the primary personnel to accurately reconstruct the event in sufficient detail for a better understanding and complete evaluation of the event.
1.1.5 The Methods and Criteria For Comparing Event Information with Known or Expected Plant Behavior The reactor trip report provides the necessary step-by-step checklist to verify that operations occurred as expected.
Plant behavior is compared to limiting values contained in the technical specifications and expected behavior as described in the final safety analysis report to ensure operation was as expected and within limits. Any deviation is noted in the report and evaluated.
1.1.6 The Criteria For Determining the Need For Independent Assessment of An Event of and Guidelines on the Preservation of Physical Evidence to Support Independent Analysis of the Event As noted in paragraph 1.1.2, the SE has the responsibility for making the decision as to whether the plant is safe to restart after a trip. Plant management is normally contacted in the ts event of an trip and is kept cognizant of the event assessment; Therefore, the SE does have the input from an independent group in making his decision, but it still remains the SE's responsibility to determine if it is safe to restart.
t As a post-event assessment, the Plant Operations Review Committee (PORC) reviews the trip event, using the trip report and attached information, to concur with the actions and recom-mendations described in the report. The PORC chairman signs and dates the report to denote it has been reviewed and then transmits it to the plant document control unit for permanent record retention.
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Conclusion We believe that TVA is in compliance with the guidelines outlined in the Generic Letter 83-28 in that systematic assessment procedures do exist and adequately establish a complete evaluation of the event. The primary personnel involved in the assessment are well-trained and and cognizant of the operation of the plant. In addition, the personnel effectively use the procedures which address post-trip review and assessment to ensure a safe restart of_the unit.
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Sequoyah Nuclear Plant 1.2 Post-Trip' Review - Data and Information Capability Position Licensees and applicants shall have or have planned a capability to record, recall and display data and information to permit diagnosing the causes of unscheduled reactor shutdowns before restart and for ascertaining the proper functioning of safety-related equipment.
Adequate data and information shall be provided to correctly diagnose the cause of unscheduled reactor shutdowns and the proper functioning of safety-related equipment during these events using systematic safety assessment procedures (Action 1.1).
The data and information shall be displayed in a form that permits ease of assimilation and analysis by persons trained in the use of systematic safety assessment procedures.
A report shall be prepared which describes and justifies the adequacy of equipment for diagnosing an unscheduled reactor shutdown.
Response
l '. 2.1 Capability for assessing sequence of events 1.
Description of equipment Plant computer - Westinghouse PRODAC P250 process computer.
2.
Parameters Monitored Parameters which result in s reactor trip are listed in attachment 2'.
The parameters which result in a partial reactor trip are not listed. However, both types of parameters are printed on the Sequence of Event record. Attachment 1 contains a listing of key symbols.
3 Time discrimination between events Ary one of the sequence of event inputs can initiate scanning of the inputs. If any one of the inputs has changed state, its new condition is recorded. This scanning is repeated every 16.667 MS for 1 minute or until the recording buffer is full.
4.
Format for displaying data and information See attached printout (attachment 3).
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5.
Capability for retent's.' of data and infonmation The recording buffer may store up to 25 events or hold data for 1 minute from the initiating event before transferring data to the printing buffer. The recording buffer is then free to record new data. The contents of the printing buffer is then made available for storage in a hardcopy form.
6.
Power source Power is from an inverter with station battery as backup power. It is considered non-Class 1E.
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Key Symbols Delta D
F Flow L
Level Q
Flux P
- Pressure
- T Temperature TB Turbine Caus Re - Causing Reactor Trip Set Point SP-IN - Inlet OUT - Outlet RCL - Reactor coolant loop FPT - Feed pump turbine XFMR - Transformer-SG Steam generator m ~
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Sequence of Events Parameters F0403D RCL LO F Above P-8 CAUS RE F0423D RCL LO F Above P-7 CAUS RE F0490D Stm Line HI F SI Block TR A LO406D Sta Gen A LO LO L CAUS RE-
-LO426D Sta Gen B LO LO L CAUS RE LO446D-Stm Gen C LO LO L CAUS RE LO466D-Sta Gen D LO LO L CAUS RE LO483D Pressurizer HI L and P7 CAUS RE N0005D PWR RNG CHAN HI Q CAUS RE N0010D PWR RNG CHAN LO Q CAUS RE N0020D INTERM RNG 1 HI Q Initiates RE N0021D INTERM RNG 2 HI Q Initiates RE N0024D INTERM RNG HI Q CAUS RE N0029D PWR RNG CHAN HI Q RATE CAUS RE
.N0036D Source RNG HI C CAUS RE P0407D Stm Line A HI DP SI CAUS RE PO427D Sta Line B HI DP SI CAUS RE PO447D Sta Line C HI DP SI CAUS RE PO467D Sta Line D HI DP SI CAUS RE PO483D Pressurizer HI P CAUS RE PO488D Pressurizer LO P and P7 CAUS RE P1003D' Containe HI P SI CAUS RE T0498D RCL OVERTEMP DT CAUS RE T0499D RCL OVERPWR DT CAUS RE V0324D RCP BUS UNDER FREQ and P7 CAUS RE YO335D UNIT ON LINE YO390D TB Stop Valves CL and P9 CAUS RE YO401D Stm Gen A LO L and FW F CAUS RE Y0421D Sta Gen B LO L and FW F CAUS RE YO441D Stm Gen C LO L and FW F CAUS RE YO461D Stm Gen D LO L and FW F CAUS RE YO480D Pressurizer LO PRESS SI CAUS RE YO721D SG A HI HI L CAUS TURB TR YO722D SG B HI HI L CAUS TURB TR YO723D SG C HI HI L CAUS TURB TR YO724D SG D HI HI L CAUS TURB TR YO920D SFTY INJ SET MANUAL 1 CAUS RE
.YO121D~
SFTY INJ SET MANUAL 2 CAUS R" Y2000D TB TRIP - COND VACUUM Y2001D TB TRIP - HYD FLUID PRESS Y2003D TB TRIP - OVERSPEED CAUS Y2004D TB TRIP - STATOR COOLANT Y2005D TB TRIP - BRG OIL LEVEL Y2007D TB TRIP - FPT S Y2008D TB TRIP - SOL ENERGIZED Y2009D TB TRIP - HYD FLUID LEVEL Y2010D TB TRIP - VIBRATION Y2011D-TB TRIP - EH CONTROL PWR Y2012D TB TRIP - THRUST BRG WEAR Y2013D TB TRIP - BRG OIL PRESS
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Y2400D FPT A TRIP - PMP TB BRG OIL PRESS Y2402D FPT A TRIP - THRUST BRG WEAR Y2403D FFT A TRIP - SUCTION VALVE Y2405D FPT A TRIP - INJECTION WATER Y2406D FPT A TRIP - COND VACUUM Y2407D SSPS TURB TRIP TRAIN A Y2410D FPT B TRIP - PMP TB BRG OIL PRESS Y2412D FPT B TRIP - THRUST BRG WEAR Y2413D FPT B TRIP - SUCTION VALVE CL
-Y2415D FPT B TRIP - INJECTION WATER Y2416D.
FPT B TRIP - COND VACUUM Y2417D SSPS TURB TRIP TRAIN B Y2801D GEN DIFF Y2802D GEN MAIN BRK Y2803D GEN NEG PHASE SEQ Y2804D GEN BACK'JP and MN XFMR FDR DIFF Y2805D GEN NEUTRAL OVERVOLT Y2806D
. GEN OVERCURRENT Y2807D GEN REVERSE POWER Y2808D MAIN XFMR DIFF - SUD PRESS Y2809D USS XFMR DIFF - PRESS - OC - NEUT OC Y2815D CRD M SET A Y2816D.
CRD E SET B Y2906D ENVIRON MON SYS i
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Sequence of Event Printout 1231 SEQUENCE OF-EVENTS RECORD. FIRST EVENT AT H12 M29 557
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F04650 STM GEN D LO FW F 2 PART RE TR C
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.F04650.STM GEN D LO FW F 2 PART RE NT TR C 128 F0465D STM GEN D LO FW.F 2 PART RE TR C 146 F04650 STM GEN D LO FW F 2 PART RE NT TR C 158 F04650 STM GEN D LO FW F 2 PART RE TR C 226 F0465D STM GEN D LO FW F-2 PART RE' NT TR C 252 LO461D STM GEN D LO L 2 PART RE TR C 2585 LO461D-STM GEN D LO L 2 PART RE NT TR C 2596 LO4600 STM GEN D LO L 1 PART RE TR C 2647 LO461D STM GEN D LO L 2 PART RE TR C 2834 LO461D STM, GEN D LO L 2 PART RE NT TR C 2385 LO461D STM GEN D.LO L 2 PART RE TR C 2931 LO461D STM GEN D LO L 2 PART RE NT TR C 2982 LO461D STM GEN D LO L 2 PART RE TR C 3011 LO461D STM GEN D LO L 2 PART RE-NT TR C 3211' L0461D STM GEN D LO L 2 PART RE TR C 3244 LO461D STM GEN D LO L 2 PART RE tat TR C 3266 L0461D STM GEN D LO L 2 PART RE TR C 3316 LO461D STM GEN D LO L 2 PART RE NT TR C 3399 LO461D STM GEN D LO L.2 PART RE TR C 3437 1232 END SEQUunCE OF. EVENTS RECORD 1232 SEQUEN.CE OF EVENTS RECORD. FIRST-EVENT AT H12 M31 510 LO461D STM GEN D LO L 2 PART RE NT TR-C 0
L0461D STM GEN D LO L 2 PART RE
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16 LO4610 STM GEN D LO L 2 PART RE NT TR C 144 LO461D STM GEN D LO L 2 PART RE TR C 200 LO461D STM GEN D LO L 2 PART RE NT TR C 2004 LO461D STM GEN D LO L 2 PART RE TR C 2045 LO461D STM GEN D LO L 2 PART RE NT TR C 2117 LO4610 STM GEN D LO L 2 PART RE TR C 2160 LO4$1D STM CEN D LO L 2 PART RE NT TR - C 2201 LO4010 STM GEN D LO L 2 PART RE TR C 2237 LO4460' STM GEN C LO LO L CAUS RE TR C 2362 Y0006D REAC MAIN TR BKR A TR C 2367 Y00070.REAC MAIN TR BKR B TR C 2367 P0397D TB HYD OIL LO P 2 PART RE TR C 2372 Y0390D TB STOP VALVES CL G P7 CAUS RE TR C 2372
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P0396D TB HYD OIL LO P 1 PART RE TR C 237.2 P0398D TB HYD OIL LO P 3 PART RE TR C 2372 LO446D STM GEN C LO LO L CAUS RE NT TR C 2375 I
N00270 PWR RNG CHAN 3 HI Q RATE PART RE TR C 2378
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N00290 PWR RNG CHAN HI Q RATE CAUS RE TR C 2379
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N00250 PWR RNG CHAN 1 HI Q RATE PART RE TR C 2379 N00260 PWR RNG CHAN 2 HI Q RATE PART RE TR C 2379 N0028n PWR RNG CHAN 4 HI Q RATE PART RE TR C 2380 Y29170 NUCL PWR 1 RE TR P9 PART PERM RESET C 2392 Y2919D NUCL P A 3 RE TR P9 PART PERM RESET C 2392 Y29203 NUCL PWR 4 RE TR P9 PART PERM RESET C 2392 Y29210 NUCLEAR POWER P9 PERMISSIVE SET C 2392 YO390D TB STOP VALVES CL & P7 CAUS RE NT TR C 2392 Y2918D NUCL PWR 2 RE TR P9 PART PERM RESET C 2393 y
LO461D STM GEN D LO L 2 PART RE NT TR C 2399 1233 END SEQUENCE OF EVENTS RECORD
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Sequoyah Nuclear Plant 1.2.2 Capability for assessing the time history of analog variables needed to determine the cause of unscheduled reactor shutdowns, and the functioning of safety-related equipment.
s 1.
Description of equipment Plant computer - Westinghouse PRODAC P250 process computer
.2.
P'arameters monitored, sampling rate, and basis for selecting
-parameters and sampling rate I-There are two sets of parameters monitored by the post-trip program. Group 1 points are scanned at 8 second intervals l
and are listed in attachment 4.
Group 2 points are scanned at 2 second intervals and are listed in attachment 5.
See L
attachment 1 for Key Symbols.
3 Duration of time history Group 1 points are monitored for 2 minutes before a trip and 3 minutes after a trip.
Group 2 points are monitored for 3 seconds before and after-a reactor trip.
4.
Format for displaying data including scale of time histories Trip time is displayed in hours, minutes, and seconds. The data collection time for-each group of parameters is indicated in minutes,' seconds, and tenths of a second.
Printed columns Time 1, Time 2,'etc., indicate the data I
collection time. See attachment 6.
5.
Capability for retention of data, information, and physical evidence Data is not stored on magnetic tape, a hardcopy only (printout) is available for storage.
6.
Power source Power is from an inverter with station battery as backup power. It is considered non-Class 1E.
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1.2 3 Other data and information provided to assess the cause of unscheduled reactor shutdowns.
The first cut annunciator system provides the operator with information which alerts him to the probable cause of reactor r trip. The operator also has strip chart recording of selected
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phrameters for use in determining causes of reactor trips.
1.2.4 Schedule for any pla'nned changes to existing data and information capability.
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There are not any planned changes to Sequoyah's system at this time.
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Group 1 Post-Trip Review F0403A STM GEN A FEED WTR IN 1 F F0404A STM GEN A FEED WR IN 2 P F0405A STM GEN A STM OUT 1 F F0406A STM GEN A STM OUT-2 P
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F0423A STM GEN B FEED W R IN 1 F F0424A STM GEN B FEED WTR IN 2 F F0425A STM GEN B STM OUT 1 F F0426A STM GEN B STM OUT 2 F F0443A STM GEN C FEED WTR IN 1 F F0444A STM GEN C FEED WTR IN 2 F
' F0443A STM GEN C STM OUT 1 F F0446A STM GEN C STM OUT 2 F F0463A-STM GEN D FEED WTR IN 1'F F0464A STM GEN D FEED WTR IN 2 F F0465A STM GEN D STM OUT 1 F i
F0466A STM GEN D STM OUT 2 F F2250A' FW PMP A DISCH F F2251A FW PMP B DISCH F LO400A STM GEN A NAR RNG 1 L LO401A STM GEN A NAR RNG 2 L
- LO402A STM GEN A NAR RNG 3 L LO403A STM GEN A WIDE RNG L L0420A STM GEN B NAR RNG 1.L
-LO421A STM GEN B NAR RNG 2 L
- LO422A STM GEN B NAR RNG 3 L LO423A STM GEN B WIDE RNG L L0440A STM GEN C NAR RNG 1 L
' LO441A~
STM GEN C NAR RNG 2 L' LO442A STM GEN C NAR RNG 3 L LO443A STM GEN C WIDE RNG L
- L0460A STM GEN D NAR RNG 1 L LO461A STM GEN D NAR RNG 2 L LO462A STM GEN D NAR RNG 3 L LO463A STM GEN D WIDE RNG L.
- LO480A Pressurizer 1 L LO481A Pressurizer 2 L LO482A
~ Pressurizer 3 L
- LO483A
- Pressurizer LVL CONTROL SP N0031A Source RNG DETECTOR 1 LOG Q N0032A Source RNG DETECTOR 2 LOG Q N0035A Intern RNG DETECTOR 1. LOG Q $U0019 <
N0036A Intern RNG DETECTOR 2 LOG Q $UOO20 <
N0041A PWR RNG CH1 $ QUAD 4< TOP DETECT Q N0042A PWR RNG CH1 % QUAD 4< BOT DETECT Q N0043A PWR RNG CH2 $ QUAD 2< TOP DETECT Q N0044A PWR RNG CH2 $ QUAD 2< BOT DETECT Q O
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~2-N0045A PWR RNG CH3 $ QUAD 1< TOP DETECT Q N0046A PWR RNG CH3 % QUAD 1 < BOT DETECT Q N0047A PWR RNG CH4 % QUAD 3 < TOP DETECT Q N0046A PWR RNG CH4 % QUAD 3 <B0" DETECT Q N0049A PWR RNG CH1 % QUAD 4 <Q N0050A PWR RhG QI? $ QUAD 2 <Q N0051A PWR RNG CH3 $ QUAD 1 <Q N0052A PWR RNG CH4 % QUAD 3 <Q P0398A TB FIRST STG 1 P P0399A TB FIR 3T STG 2 P P0400A STM GEN A STM OUT 1 P PO401A STM GEN A STM OUT 2 P PO402A STM GEN A STM OUT 3 P P0420A STM GEN B STM OUT 1 P PO421A STM GEN B STM OUT 2 P PO422A STM GEN B STM-OUT 3 P P0440A STM GEN C STM OUT 1 P P0441A STM GEN C STM OUT 2 P P0442A STM GEN C STM OUT 3 P P0460A STM GEN D STM OUT 1 P PO461A STM GEN D STM OUT 2 P PO462A STM GEN D STM OUT 3 P PO480A Pressurizer 1 P PO481A' Pressurizer 2 P PO482A Pressurizer 3 P PO483A Pressurizer 4 P P1000A Containment 1 P P1001A Containment 2 P P1002A Containment 3 P P1003A Containment 4 P P2224A Feedwater Pumps Suction Header P P2226A Condensate BSTR PMPS SUCT HDR P P2270A COND A HOTWELL P P2273A FEEDWATER HTRS 1 OUTLET HDR P QO340A UNIT GENERATOR GROSS MW T0403A RCLA 1 DT T0406A RCLA COLD T T0407A RCLA OVERPWR DT 1 SP 70410A RCLA OVERTEMP DT 1 SP T0420A RCLB 1 TAVG T0423A RCLB 1 DT T0426A RCLB COLD T T0427A RCLB OVERPWR DT 1 SP T0430A RCLB OVERTEMP DT 1 SP 70440A RCLC 1 TAVG
- T0443A RCLC 1 DT 70446A RCLD COLD T T0447A RCLD OVERPOWER DT 1 SP 70450A RCLD OVERTEMP DT 1 SP T0460A RCLD 1 TAVO T0463A RCLD 1 DT 70466A RCLD COLD T t
70467A RCLD OVERPWR 1 SP 70470A RCLD OVERTEMP DT 1 SP 70481A PRESSURIZER STM T T0496A' RC TREF T0497A RCL AUCT DT T0499A RCL AUCT TAVG These points are' scanned every 6 seconds and stored for 2 minutes before
.and 3 minutes after a trip.
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Group 2 Post-Trip Review N0049A PWR RNG CHANNEL 1 % QUAD 4<Q N0050A PWR RNO CHANNEL 2 % QUAD 2<Q N0051A PWR RNG CHANNEL 3 % QUAD 1< Q l
N0052A PWR RNG CHANNEL 4 % QUAD 3<Q P0398A TB FIRST STAGE 1 P P0399A TB FIRST STAGE 2 P Q0340A UNIT GENERATOR GROSS MW T0496A RC TREF These points are scanned every 2 seconds and stored for 8 seconds before and after a trip.
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5 Ssquoyth Nuclect Pitnt 2.1 Equipment Classification and Vendor Interface (Rea'otor Trip System Components)
Presently TVA's Division of Nuclear Power (NUC PR), identifies all components whose functioning is required to trip the reactor as safety-related.
These components which include the reactor protection system, the solid state protection system, and all other components whose function is defined as safety-related are now outlined in TVA's Operational Quality Assurance manual as critical systems, structures, or components (CSSC) which is a corporate document. Each individual plant has incorporated the applicable portions of this document into their procedures. In addition, TVA's corporate proceduras raquire all maintenance or modification activities to ba documented prior to performing the work. This documentation is then reviewed by the appropriate plant organizations to ensure that it is properly identi-fied as CSSC or non-CSSC and to ensure that the applicable procedures and quality requirements for the idenitified work will be adhered to.
Furthermore, NUC PR requires that all procurement documents be identi-fled as pertaining to CSSC or non-CSSC equipment. These procurement documents are reviewed by plant organizations or division central office organizations (depending on their point of origination) to ensure they are properly identified and contain the appropriate and required quality controls and specifications. Depending on the quality grouping, as outlined in division procedures that the procure-ment documents come under, many of them are also reviewed by other division central office organizations to further ensure that they meet all requirements.
In view of the present division and plant procedures pertaining to safety-related equipment identification, and information handling systems used to control safety-related activities, we believe TVA is in compliance with the NRC staffs position.
TVA's NUC prs vendor interface program is presently centered around the division's operating experience review (OER) program which was developed to ensure that vendor and other related information would be handled from a systematic approach to continually inform the plants and other cognizant organizations of revisions, modifications, or defic *.encies in plant equipment or procedures. The vendor interface program hinges around the original nuclear steam supply system (NSSS) l supplier who supplied all reactor trip system components. Any infor-mation supplied by the NSSS vendor to the division corporate office is I
acknow' edged upon receipt by the division and forwarded to the OER organization and entered in the system. This information is then forwarded to the cognizant organizations and plants for review, comments, recommendation, or incorporation into plant activities.
l The review, comments, or recommendations are documented and returned I
to the operating experience review group (OERG)'normally within l
30 days as presently required by the division procedures. Any recom-i sendations are forwarded to the plant for incorporation in plant l
activities or resolution. This information is tracked and documented l
by the OERG during the entire process until it has been incorporated or resolved. This documentation is then stored for the life of the plant for further reference.
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Conclusion As previously stated, we believe that TVA's programs properly identify the reactor trip system and related components as safety related. We also believe that TVA adequately controls activities such as maintenance, modi-fication, and procurement on reactor trip system components. In addition, we believe that TVA's operating experience review prograa has established a comprehensive vendor interface program and ensures that vendor activities are reviewed and incorporated as necessary for the reactor trip system. In conclusion, we believe TVA's program is in compliance with NRC position and recommendations as stated in Generic Letter 83-28.
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8equoyah Nudest Plant Lt 30tt1Mert CLA881FICAt10N AND Y2NDOR INTRRfACE (pe00 RAMS FOR ALL 8APITT -
RSLATBD 00Mp0NENf8)
M Licensees and appileants shall tubmit, for staff review, a description of their programs for estety-related equipment classification and vender interface as describsd belows i.61 Poe equipment elassifloation, 11oensees and appilaants shall describe their program for ensuring that all acepenents of safety-related systema neessaary for aeoompilshing required safety functions are identified as safety-related on doousents, procedures, and information ineluding asintenaries, work ceders, and replacement parts. This desoription shall include 2 2.1.1 The criteria for identifying ocepenents as safety.
related within oratens currently cleasified as safety =
relatea. This shall gL be interpretad to require changes in' safety classafication at the ayatens level.
Rasmonen During dealen and construction, the equipment clasaffieatens were identified in various design output dooumenta auch as drawin8s and construction project speelfloations. Thj a elassification was to be identified if the itema fell under the requirements of a quality assurance program, not necessarily it it was safety-related.
W C pH espanded this eeneept by establishing a Critical Structurea systema and odsponents (0880) list for each nuclear plant. The 0880 defines the scope of applicability of TVA's GA preseas on operating plants. All M1vities that could affect CSSC equipment are performed in arsordanee with QA program requirements. The prenant eriteria, which are a part of our
.0peestional Qusiity Amauranes Manual (00AM), are used for inclusion of itema on the C880 1146 are as follows.
1.
OtttfALCf.itdria A.
Those itema that'are necessary te ensure is g,gprityoftheremotorcoolantpressure 2.
The espability to shut down the reaeter and maintain it in a safe sendition 3.
The espability to prevent or alligste the consequences of an ineident which 60uld result in potential offaite espesures eenparable to those speelfied in 10 CFR part 100
B.
Those items which the CSSC Subcommitte consider should receive the same level of quality assurance coverage as those listed in the general criteria above.
II. Specific Guidelines for Inclusion of Items on the CSSC List Specific systems, structures, or components should be added to the CSSC list if they perform any of the following safety-related functions.
A.
Maintains core reactivity control under emergency conditions including those covered by anticipated transients without scram (scram mechanisms).
B.
Instruments and controls which are essential for cmcrgency reactor shutdown, containment isolation, reactor core cooling, and containment and reactor heat removal or are otherwise required for preventing significant release of radioactive material to the environment. Instrumentation and controls that perform an essential recondary function shall be considered safety-related if they are designed primarily to accomplish one of the above functions or where their failure would prevent accomplishing one of the above functions.
This includes those instruments and controls that are designed as safety-related and:
1.
Automatically keep the reactor operating within safe region by shutting down the reactor whenever the limits of the region are approached (reactor trip signal instrumentation)
.2.
Initiate actuation of one or more of the engineered safety features in order to prevent or mitigate damage to the core and water coolant system components and ensure containment integrity-(engineered safety features activation system instrumentation) 3 Provids protective interlocks to prevent an operator error which could lead to incidents or events representing limiting plant design cases (permissive and interlock circuits).
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Indicators and recorders and associated channels which are essential to:
Perform manual safety functio $ and to a.
perform postaccident monitoring following a reactor trip due to any condition up to and including the design limiting. fault 7
(containment pressure indicators).
b.
Maintain the plant in a hot shutdown condition or to proceed to a cold shutdown condition while meeting the limits of the plant's technical specification (system pressure monitor).
c.
Monitor conditions in the reactor core, reactor coolant systems, mainsteam and feedwater systems and containment (auxiliary fe'edwater flow monitor).
C.
Provides a barrier for containing reactor coolant within the reactor coolant pressure boundary (reactor coolant piping, valves, and fittings).
D.
Cools the reactor core under emergency conditions (residual core heat removal systems).
E.
Maintains fuel clad integrity (fuel clad, core power monitoring systems).
l F.
Provides power, control, logic, indication, and protection to systems or components to enable them to accomplish'their safety function (diesel generators, vital ao and de power).
G.
Supports or houses equipment that performs a safety function or protects that safety-related equipment from potential natural phenomena, equipment failure, and manmade hazards (seismic class I containment and structures, fire protection systems).
i H.
Maintains specifie'd environment (e.g., temperature, pressure humidity, radiation) as required in vital areas.to maintain equipment operability and personnel l
access (control room habitability systems).
I.
Supplies cooling water for the arpose of heat removal from the systems and r,omponents which provide a safety function (essentia! component cooling and i
l service water systems).
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Contains radioactive waste such that its failure could rasult in the release of radioactive waste to the offsite environments in violation of criteria A.3 (low-level radioactive waste discharge isolation valves).
K.
Controls fuel storage to prevent inadvertent criticality (fuel storage racks).
L.
Ensures adequate cooling for irradiated fuel in spent fuel storage (spent fuel cooling system).
M.
Minimizes the pro'0 ability of dropping objects on stored fuel (overhead crane).
N.
Maintains primary containment as required by the FSAR to meet General Design Criteria 54, 55, 56, and 57 (containment penetrations and associated isolation and boundary valves).
O.
Doors and hatches which serve one or more of the following functions for safety-related equipment and areas:
(1) pressure confinement, (2) leakage confinement, (3) missile protection, (4) pipe whip and jet impingement barrier, (5) equipment rupture flood protection, (6) natural flood protection, or (7) fire protection
.The items in parentheses are examples of items which would be considered as applicable to the listed guidelines and therefore eligible for inclusion on the CSSC list. These guidelines are continually reviewed and updated by the CSSC Review Committee to include changes in NRC requirements and plant design and safety criteria as they occur.
III. The CSSC list is supplemented by TVA EN DES identified Class lE equipment and requirements.
2.2.1.2 A description of the information handling system used to identify safety-related components (e.g., computerized equipment list) and the methods used for its development and validation.
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Response
The overall development and maintenance of7the CSSC list is the responsibility of the CSSC committee. The CSSC committee is a review group comprised of multidisciplined nuclear-experienced engineers and quality, assurance representatives. The various technical branches under the oversight of the CSSC committee developed the initial CSSC list and evaluate all changes which are reviewed and approved by the CSSC committee. The CSSC line is issued and controlled manually as part of the OQAM.
2.2.1 3 A description of the process by which station personnel use this information handling system to determine that an activity is safety-related and what procedures for maintenance, surveillance, parts replacement, and other activities defined in the introduction of 10 CFR 50, Appendix B, apply to safety-related j
components.
Response
Plant activities that could affect equipment on the CSSC list are prescribed by instructions appropriate to the circumstances.
These instructions are prepared, reviewed, and approved in accordance with section 6.0 of the plant's technical j
specifications and.the plant QA program.
2.2.1.4 A description of the management controls utilized to verify that the procedures for preparation, validation and routine utilization of the information handling system have been-followed.
4
Response
After licensing, the inplant Quality Engineering Section routinely and independently verifies that the plant instructions appropriately utilize the CSSC list and meet the plant's quality assurance requirements.
The Office of Quality Assurance performs audits of the central office activities and plant activities to verify that the QA
- requirements are met.
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i 2.2.1.5 A demonstration that appropriate design 3
verification and qualification testing is specified for procurement of safety-related components. The specifications shall include qualification testing for expected safety service conditions and provide support for the licensess' receipt of testing documentation to support the limits of life recommended by the supplier.
Response
Predefined specification for various components and materials have-been prepared by various technical branches for items such as ASME code valve parts, pump parts and materials, and class lE equipment. Also, when original specifications cannot be verified, the technical branches prepare specifications that are used in the procurement process..In addition, TVA prepares and utilizes substitution guides for standardized industry items such as bearings, V belts, capacitors and resistors. The quality assurance program requires that for items that have storage and shelf life, the vendor furnish such information.
The quality assurance program requires that the original design specification or the TVA originated specifications, supplemented by class lE requirements, are used in procurement of CSSC components. All CSSC procurements are reviewed independently by a quality j
'nce or quality engineering group.
ams are receipt inspected to ensure that the a.
.ed contract documentation and requirements are met.
2.2.1.6 Licensees and applicants need only to submit for staff review the equipment classification program for safety-related components. Although not required to be submitted for staff review, your equipment classification program should also include the broader class of structures, systems, and components important to safety required by GDC-1 (defined in 10 CFR Part 50, Appendix A, " General Design Criteria, Introduction").
Response--None Required 9
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7 2.2.2 For vendor interface,... safety-related equipment are provided.
Response
TVA is actively participating in the NUTAC associated with NRC Generic letter 83-28, Section'2.2.2. The results of NUTAC are expected to be available for approval during February 19811. Upon receipt of the NUTAC recommendations, TVA will evaluate and provide a plan for implementation.
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Sequoyah Nuclear Plant a
31 Post-Maintenance Testing (Reactor Trip System Components)
Action The following actions are applicable to post-maintenance testing.
1.
Licensees and applicants shall submit the results of their review of test and maintenance procedures and Technical Specifications to assure that post-maintenance operability testing of safety-related components in the reactor trip system is required to be conducted and that the testing demonstrates that the equipment is capable of performing its safety functions before being returned to service.
Response
Standard Practice Pr,ocedure SQM2 at Sequoyah establishes the method and responsibilities necessary to conduct maintenance through the use of a Maintenance Request (MR) form TVA 6436.
Maintenance on critical structures, systems, and components (CSSC)
(reactor trip system is CSSC) is required to be initiated by this form. These MRs are originated by the persen requesting maintenance. The MR planner (cognizant in area of requested maintenance) checks the MR and completes tht areas necessary for identifying maintenance including post-maintinance testing.
Technical Instructions (tis) 54 and 59 are used by the MR planner for identifying testing. These tis contain lists of process instruments and their Surveillance Instructions (sis) which are listed in technical specifications and/or required operable by technical specifications, and/or needed to obtain data Lo satisfy technical specification requirements. The sis either contain the steps to test the instrument or they schedule the performance of a functional test or calibration.
Before any maintenance or testing can be performed, Quality Assurance (QA) must review the form to assure that the format and content are in compliance with QA requirements. After post-maintenance testing is completed, the MR must be signed by the section completing the test and the SI covering the test must be g
listed. Plant Operations must also sign the MR acknowledging that testing is completed. Once the MR is completed, QA must again review it to assure that the format and content are in compliance with QA requirements. The completed SI is also reviewed for accuracy and completeness.
During review of material concerning ATWS and the procedures designed to test the reactor trip system, it was identified that the procedure used to test the manual reactor trip system could be enhanced by providing for completely independent testing of the shunt trip and undervoltage (uy) trip functions. This procedure has been changed to independently verify operation of the uv trip
1 coil, shunt trip coil, handswitch contacts, and all associated wiring.
Based upon our review, Sequoyah's program does require post-maintenance testing and the procedures for this testing require operability before the reactor trip system can be returned to service.
2.
Licensees and applicants shall submit the results of.their check of vendor and engineering recommenda.tions to ensure that any appropriate test guidance is included in the test and maintenance procedures or the technical specifications, where required.
Response
i The Area Plan concept of NUC PR addresses 3.1.2 through its Regulatory j
Compliance Program. The Regulatory Compliance Program includes the program element entitled Nuclear Experience Review which makes bresd-base use of industry experience information. The Reactor Engineeriag Branch is responsible for ensuring that this information is distrib-
'uted to the appropriate sections, maintaining files, references, and responses for each item. Nuclear central office and plant staff
. ithin the area of their expertise review this information and inform w
the responsbile section as to its applicability. However, the plant with assistance from central office staffs, has the responsibility for implementing any corrective actions or recommendations.
As a part of this program, Westinghouse Electric Corporation was requested to furnish all technical bulletins and data letters which NUC PR did not have in their files. Most'of these safety-related bulletins and data letters have since been reviewed for applicability to test and maintenance procedures. The remaining safety-related bulletins and data letters are being reviewed.
Our Nuclear Experience Review Program does ensure that vendor and engineering recommendations receive the appropriate distribution /
review and verifies that test and maintenance procedures contain 4
appropriate vendor and engineer,ing recommendations.
i 3
Licensees and applicants shall identify, if applicable, any post-maintenance test requirements in existing technical specifications which can be demonstrated to degrade rather than enhance safety.
Appropriate changes to these test requirements, with supporting justification, shall be submitted for staff approval.
(Note that action 4.5 discusses online system funogional testing.)
Response
At the present time, we cannot identify any post-maintenance testing requirement which degrades 'rather than enhances safety.
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t In March 1983 we received permission to change technical specifi-cations to extend the time intervals required to conduct analog channel functional tests of the reactor trip system from one to three months. This change was based upon actual operating experience and.
reflects the continuing effort to maintain the best possible testing program.
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e 32 Post-Maintenance Testing ( All Other Safety-Related Components)
Action 1.
Licensees and applicants shall submit a report documenting the extending of test and maintenance procedures and technical specifications review to ensure that post-maintenance operability testing of all safety-related equipment is required to be conducted and that the testing demonstrates that the equipment is capable of performing its safety functions before being returned to service.
Response
The Division of Nuclear Power Operaticnal Quality Assurance Manual (0QAM) ' requires that maintenance instructions shall contain measures to cover the following.
"Upon completion of maintenance on any item of the CSSC list and before release for aervice, appropriate testing shall be performed to verify operational acceptability. Functional tests or industrial standard testa may be used for this purpose.'
The OQAM also requires review of the maintanance request (MR) by the responsible section and the Field Quality Engineering (QE) Section before performance of maintenance on CJSC equipment. Standardized guidelines which include the following are provided for preparation / review of hrs.
1.
Specify appropriate post-maintenance testing and, where applicable, reference the proper plant instruction.
2.
Consider compliance with plant technical specifications.
Specifically:
a.
Will removal of-equipment from service for this maintenance violate any limiting conditions for operations?
b.
Are adequate post-maintenance tests (sis) specified to ensure the equipment's readiness for operation?
3 Provide for return of equipment to normal status as required.
The MR requires that the section responsbile for the performance of the post-maintenance test and also the operations section shall sign to concur that the post-maintenance test was performed and the equigaent is ready for return to service.
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-R-Based upon our review, the NUC PR progree does require post-maintenance testing to demonstrate operability before safety-related components are returned to service. These requirements are implemented at each plant through plant specific instructions.
Smauavah E mlaar Plant Standard Practice SQM2, " Maintenance Management System,"_ establishes the method and responsibilities for managing the initiation, planning, scheduling, execution, status tracking, and documentatun of maintenance work. This procedure identifies those responsible for completing the portion of the MR addressing post-maintenance testing, provides instructions on how to address this ites, and provides l
guidance for the QA review before the performance of the work.
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2.
Licensees and applicants shall submit the results of their check of vendor and engineering recommendations to ensure that any appropriate test guidance is included in the test and maintenance procedures or the technical specifications where required.
Response
TVA's philosophy has always been to utilize engineering judgment, operating experience, TVA policy, and industry experience in conjunction with vendor and engineering recommendations to ensure that any appropriate test guidance is included in the test and maintenance procedures or the technical specifications where required.
This is supplemented by a program dealing with the review of operating experience reports. This program establishes a system to ensure the review of operating experience reports to document their applicability i
to TVA plants, to provide required written responses, and to ensure proper disposition of all applicable items.
Also, in order to comply with IE Ilulletin 79-01B and NUREG-0588, class 1E electrical equipment is being reviewed for applicable maintenance instructions required to maintain the environmental qualification of the equipment. This activity will be completed in accordance with the NRC ruling on environmental qualification.
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In addition to the above.-periodic review of procedures and
' instructions is required by the OQAM to determine if changes are necessary or desirable. This review is conducted no less frequently than every two years by an individual knowledgeable in the area affected by the procedure / instruction.
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t SEQUOYAH NUCLEAR PLANT 1
4.1 Reactor Trip Systen Reliability (Vendor-Related Modifications) h All vendor-recommended reactor trip breaker modifications shall be reviewed to verify that either: (1) each modification has, in fact been implemented; or (2) a written evaluation of the technical reasons i
for not implementing a modification exists.
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Reanonne The urElervoltage trip attachment (WTA) at Sequoyah for the Westinghouse DB-50 breakers is considered post-1972 (modified WTAs) as stated in the Westinghouse letter on vendor-related modifications.
An MR has been issued for visual inspection of the DB-50 breakers to ensure that they have the post-1972 WTAs. We plan for the inspection to be performed during the next refueling outages of units 1 and 2.
An inspection of Power Stores' inventory to ensure that spare parts reflect the modification that was performed will be made prior to their use.
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1 4.2 Reactor Trip System Reliability (Preventive Maintenance and Surveillance Program for Reactor Trip Breakers)
Ant 1GA Licensees and applicants shall describe their preventive maintenance and surveillance program to ensure reliable reactor trip breaker operation. The program shall include the following.
1.
A planned program of periodic maintenance, including lubrication, housekeeping, and other items recommended by the equipment supplier._
2.
Trending of parameters affecting operation and measured during testing to forecast degradation of operability.
3 Life testing of the breakers (including the trip attachments) on an acceptable sample size.
4.
Periodic replacement of breakers or components consistent with demonstrated life cycles.
Reanonse 1.
Maintenance Ins,truction (MI) Procedure 10.9 and Technical Specification $.3 1 at Sequoyah govern the programs that are in d
place for the periodic maintenance of DB-50 reactor trip 1
breakers. A technical standard on reactor trip breakers has been drafted to include the' appropriate portions of the Westinghouse i
reactor trip switchgear maintenance program for DB-50 breakers.
i This program is a description of periodic maintenance, including lubrication, housekeeping, and other items recommended by
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Westinghouse. Sequoyah's MI 10 9 will be revised to reflect the
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recommendations in the technical standard on reactor trip breakers. We expect the technical standard to be issued by i
December 15, 1983, and incorporated into MI 10 9 within 90 days.
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A program for trending of parameters is recommended in the technical standard on reactor trip breakers. The technical standard states that the program should consist of the following.
a.
The compilatio? of all maintenance activity records into a history file.
b.
The use of the Nuclear Plant Reliability Data System for breaker failure data.
c.
An HR system. The HR system is described in section 31.1 under response, paragraphs 1-3 These suggestions will be used at Sequoyah to develop a program for trending of parameters to assess any possibility of performance degradation.
3 Life cyle testing of the shunt trip attachment and the undervoltage trip attachment of the reactor trip switchgear is being conducted by the Westinghouse Electric Corporation for the Westinghouse Owners Group. This program is aimed toward establishing the service life of these devices and substantiating periodic test requirements with proper maictenance. The test program is scheduled for completion in the second quarter of 1984. Once this information is available, the technical standard on reactor trip breakers will be revised as applicable to incorporate this information.
4.
A maintenance program for the periodic replacement of breakers or components consistent with demonstrated life cycles is addressed in the technical standard on reactor trip breakers. The maintenance program will be established after life cycle testing of the shunt trip attachment and the undervoltage trip attachment of the reactor trip switchgear information has been made available by Westinghouse for the Westinghouse Owners Group.
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It is TVA's opinion that the above programs and philosophy provide sufficient checks and balances to provide reasonable assurance that vendor ard engineering recossendations are incorporated as appropriate.
St'andard Practice SQM57 outlines the preventive maintonance program.
It states that,. 'The appropriate mechanical, electrical, or instrument sections shall identify the type and frequency of maintenance to be performed from vendors manuals, manufacturers' bulletins, technical
' specifications requirements, division procedures, standard practices, or other directives. This type and frequency of maintenance shall be refined and changed as operating and maintenance experience is gained with the equipment."
This philosophy has been used in establishing maintenance and test programs. The periodic review of instructions is implemented by Administrative Instruction, AI-4, " Plant Instructions - Document Control." Standard Practice SQA26, " Review, Reporting, and Feedback of Operating Experience Items," implements at.the plant level the TVA i
program for review of operating' experience reports within the nuclear industry.
3 Licensees and applicants shall identify, it. applicable, any post-maintenance test requirements in existing technical specifications which are perceived to degrade rather than enhance safety.
Appropriate changes to these test requirements, with supporting justi.fication, shall be submitted for staff approvs1.
Resnonse All Plants It is not TVA's philosophy to propose changes in existsing technical
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specifications which are perceived to degrade rather than enhance safety. ' When items are identified, they will be submitted along with j
the supporting justification.
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Sequoyah Nuclear Plant 4.3 Reactor trip system reliability (automatic actuation of shunt trip attachment for Westinghouse and Babcock & Wilcox (B&W) plants).
Action Westinghouse and B&W reactors shall be modified by providing automatic reactor trip system actuation of the breaker shunt trip attachments.
The ahunt trip attachment shall be considered safety-related (Class 1E).
Response
The generic design package submitted by the Westinghouse Owner's Group for incorporation of an automatic shunt trip feature is being pro-cessed as a design change request for Sequoyah. This generic design package includes design basis, functional requirements, conceptual design, and addresses comformance to safety criteria. The deaign also includes hard-wired and component installation provisions for online surveillance testing that independently verifies the operability of the automatic undervoltage and shunt trip functions. Specifically, implementation of this modification will include the generic design plus provide the necessary responses to the items in the safety evaluation report. Scheduling for this modification will be incorporated into the integrated schedule.
4.4 Reactor trip system reliability (improvements in maintenance and test procedures for B&W plants)
No response required for this item because this item does not apply to the Sequoyah Nuclear Plant.
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e Sequoyoh Nuclear Plant 4.5. Reactor trip system reliability'(system functional testing)
Action Online functional testing of the reactor trip system, including independent testing of the diverse trip features, shall be performed on all plants.
4.5.1 Action The diverse trip features to be tested include the breaker uv and shunt trip features on Westingbouse, B&W (see action 4.3 above) and General Electric Company (GE) plants; the circuitry used for power interruption with the silicon controlled rectifiers on B&W plants (see action 4.4 above); and the scram pilot valve and backup scram valves. (including all initiating circuitry) on GE plants.
Response
The specific design package for the automatic shunt trip modification will include provisions for online surveillance testing that will independently verify operation of the shunt and the uv trip function.
.The existing reactor trips are the automatic uv trip and. the manual uv trip or manual shunt trip. The automatic trip function is tested online on a monthly basis. Alternate trains are tested each month as required by technical specifications. The manual trip function is required to be tested at each startup if it has not been tested in the past seven days. The manual trip test includes independent verification of both the uv trip and shunt trip functions as well as all handswitch contacts and associated wiring. Since this procedure results in a reactor trip, it is not feasible to perform this test at power.
4.5.2 -
Action Plants not currently designed to permit periodic online testing shall justify not making modifications to permit such testing. Alternatives
-to online testing proposed by licensees will be considered where special circumstances exist and where the objective of high 4'
reliability can be met in another way.
Response
Design changes necessary to implement the automatic shunt trip function will include provisions for online surveillance testing which provide for independent verification of the diverse trip function.
Present procedures provide for this testing *during startup if not performed in the previous seven days, Additional testing provides for monthly online testing of the automatic uv trip function. Alternate trains are tested each month as required by technical specifications.
Our present testing accomplishes the objective of high reliability.
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Sequoyah Nuclear Plant 4.5.3 Action Existing intervals for online functional testing required by Technical Specifications shall be reviewed to determine that the intervals are consistent with achieving high raactor trip system availability when accounting for considerations such as:
1.
uncertainties in component failure rates 2.
uncertainty in c0mmon mode failure rates 3
reduced redundancy during testing 4.
operator errors during testing 5.
component " wear-out" caused by the testing Licensees currently not performing periodic online testing shall determine sppropriate test intervals as described above. Changes to existing required intervals for online testing as well as the inksevals to be determined by licensees currently not performing online testing shall be justified by information on the sensitivity of eeactor trip system availability to parameters such as the test intervals, component failure rates, and common failure rates.
Response
'TVA
- has recently reviewed technical specification requirements concerning _ surveillance intervals for functional testing of the reactor trip system and the engineered safety features actuation system. This information was submitted to NRC and in March 1983, permission was received to increase the ir.terval for conducting analog channsl functional testing from one to three months. This review dessnatrates our concern to maintain a high reactor trip system availability.
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