ML18038A478
| ML18038A478 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 05/16/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML18038A477 | List: |
| References | |
| CON-IIT07-439-91, CON-IIT7-439-91 GL-83-28, NUDOCS 8906060108 | |
| Download: ML18038A478 (41) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EYALUATION REPORT GENERIC LETTER 83-28 ITEM 4.5.2 REACTOR TRIP SYSTEM RELIABILITY NINE MILE POINT 2 DOCKET NO. 50-410 1.0 ItITRODUCTION AND
SUMMARY
On February 25, 1983, both of the scram circuit breakers at Unit I of the Salem Nuclear Power Plant failed to open upon an automatic reactor trip signal from the reactor protection system.
This incident was terminated manually by the operator about 30 seconds after the initiation of the automatic trip signal.
The failure of the circuit breakers was determined to be related to the stickinq of the undervoltage trip attachment.
Prior to this incident, on February 22, 1983, at Unit 1 of the Salem Nuclear Power Plant, an automatic trip signal was generated based on steam generator low-low level during plant start-up.
In this
- case, the reactor was tripped manually hy the operator almost coincidentally with the automatic trip.
Following these incidents, on February 28, 1983, the NRC Executive Director For Operations (EDO), directed tIIe staff to investigate and report on the generic implications of these occurrences at Unit 1 of the Salem Nuclear Power Plant.
The results of the staff's inquiry into the generic implications of the Salem unit incidents are reported in NUREG-1000, "Generic Implications of the ATIIS Events at the Salem Nuclear Power Plant."
As a result of this investigation, the>Comission (NRC) requested (by Generic Letter 83-28 dated July 8, 1983
) all licensee of operating reactors, applicants for an operating
- license, and holders of construction permits to respond to generic issues raised by the analyses of these two AT)IS events.
'This report is based on our contractor's evaluation of the response submitted by Niagara Mohawk Power Corporation, the licensee for Nine Mile Point Unit 2, for Item 4.5.2 of Generic Letter 83-28 (Ref. 4).
The actual documents reviewed as part of this evaluation are listed in the references at the end of the report.
Item 4.5.2 requires licensees with plants not currently designed to permit on-line testing to justify not making provisions for such testing.
Alternatives to on-line testing proposed by the licensees will be considered if the objectives of high reliability can be met in another way.
This review will:
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~ ~
2
. I.
Confirm that the licensee has identified those portions of the Reactor Trip System (RTS) that are not on-line testable.
If the entire RTS is verified to be on-line testable, with those exceptions addressed
- above, no further review is required.
2.
Evaluate modifications proposed by the licensee to permit on-line testing against the existing criteria for the design of the protection systems for the plant being modified.
3.
Evaluate proposed alternatives to on-line testing of the RTS where the impracticality of the modifications necessary to permit on-line testing exists.
2.0 EVALUATION Niagara t'tohawk Power Corporation, the licensee for Nine Nile Point 2, responded to the Generic Letter on April 10, 1984, December 20, 1985, and April 15, 1986.
The licensee's responses affirm that Nine Nile Point 2
is designed to permit on-line functional testing of the Reactor Protection System, with the exception of the backup scram valves.
The licensee's responses state that Nine l1ile Point 2 will per form functional testing of the backup scram valves during refueling outages.
3;0 CONCLUS ION Inasmuch as the Reactor Protection System includes those components necessary to trip the reactor, we find that the licensee's stated position on Item 4.5.2 of the Generic Letter, including their justification for not performing periodic on-line testing of the backup scram valves, meets the requirements and is, therefore, acceptable.
4.0 REFERENCES
1.
NRC Letter, D. G. Eisenhut to all licensee of Operating Reactors, Applicants for Operating License, and Holders of Construction
- Permits, "Required Actions Based on Generic Implications of Sa'1em AT>IS Events (Generic Letter 83-28)," July 8, 1983.
2.
Niagara mohawk Power Corporation letter to NRC, T. E.
Lempges to Director of Nuclear Reactor Regulation, November 8, 1983.
3.
Niagara Mohawk Power Corporation letter to NRC, T. E.
Lempges to Director of Nuclear Reactor Regulation, July 31, 1984.
h
4.
EGG-NTA-7470, "Technical Evaluation Report Reactor Trip System Reliability Conformance to Item 4.5.2 of Generic Letter 83-28 Arnold, Enrico Fermi-2, Hope Creek, LaSalle Count-1 and -2, Hillstone-l, Monticello, Nine bile Point-1, Nine Mile Point-2, Oyster Creek," F. G. Farmer, Idaho National Engineering Laboratory, August 1987.
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RB CLOSE LOOP COOLING WATER FLOW TO DRYWELL UNIT COOLERS 2CNM-38A LSK-4-1.3K 2CNM-68A LSK-6-3 2CCP-135 LSK-9-1D 2CNM-45A LSK-4-1.1J 2CNM-104 LSK-4-1.3K 2CNM-138 LSK-4-1.3K 2DER-123 LSK-32-9D FO FO CONDENSATE BOOSTER PUMP P2A SUCTION FLOW'OSS OF CONTROL OF FV38A REACTOR FD PUMP 2FWS-P1A SUCTION FLOW/LOSS OF CONTROL OF 2FWR-FV2A RB CLOSE LOOP COOLING WATER FLOW FROM DW UNIT COOLERS CONDENSER CND1A VACUUM CONDENSER BOOSTER PUMPS DISCH HDR PR START-UP LVL CONTROL VALVE FLOW DW EQUIPMENT DRAIN PUMPS P3A&B FLOW LP14 UPS NO: 2VBB-UPS1A PNL NO: 2VBS-PNLA101 DIST PNL NO: 2CEC-PNL825 BKR NO: 14 PAGE 3 OF 3 LOOP 2FWS-9A LSK-6-1.1D PLANT IMPACT: LOSS OF POWER TO 2CEC-PNL825 O "v + + A ~o ~o 'V 6 "g 4+ DESCRIPTION/OTHER 2FWS-P1A PUMP DISCH PRESS 2HDL-4A LSK-4-2.1B 2HDL-35A LSK-4-2.1C 2HDL-4lA LSK-4-2.1B 2MSS-238 PID-1G FC FO 4TH POINT HEATER 2CNM-E4A WATER LEVEL/LOSS OF CONTROL OF LV4A 4TH POINT HEATER 2CNM-E4A DRAIN PUMP P1A FLOW/LOSS OF CONTROL OF FV35A 4TH POINT HEATER DRAIN PUMP P1A DISCHARGE PRESSURE MAIN STEAM PRESSURE INDICATION LP14 r ~%4~ ' LOOP UPS NO: 2VBB-UPS1A PLANT PNL NO: 2VBS-PNLA101 BKR NO: 5 PNL NO: 2VBS-PNLB101 BKR NO: 4 0 DIST PNL NO: 2CEC-PNL826 +~+ ~+8 IMPACT: LOSS OF POWER TO 2CEC-PNL826 ~O/+ + A /~O e ~o o~ ~ o~ 'V 6 M 4+ DESCRIPTION/OTHER 2CCP-126 LSK 9-L 2CCP-105 LSK-9-1L 2CCP-107 LSK-9-1M 2CCP-74B LSK-9-1D 2CCP-24D LSK-9-1D 2CCP-24B LSK-9-1D 2CCP-30D LSK-9-1D X X X X X X X X FC RB CLOSE LOOP COOLING WATER FLOW FROM REACTOR WATER CLEANUP HT EXCH. 2WCS-E3/LOSS OF CONTROL OF FV126LCW RB CLOSE LOOP COOLING WATER BOOSTER PUMPS 2CCP-P3Ai B &C DISCHARGE HDR FLOW RB CLOSE LOOP COOLING WATER BOOSTER PUMPS DISCHARGE HEADER PRESSURE RB CLOS E LOOP COOLING WATER FROM DRYWELL 2DRS-UC3B TEMP RB CLOSE LOOP COOLING WATER FROM DRYWELL 2DRS-UC2D-TEMP RB CLOSE LOOP COOLING WATER FROM DRYWELL 2DRS-UC2B-TEMP RB CLOSE LOOP COOLING WATER FROM DRYWELL 2DRS-UC1D-TEMP LP5 1' PAGE 2 OF 3 2CCP-30B LSK 9-1D 2CNM-70B LSK-4-1.3K 2SFC-2B LSK-34-2B 2CCP-51B LSK-9-1D 2HDL-41B LSK-4-2.1B UPS NO: 2VBB-UPS1A PLANT PNL NO: 2VBS-PNLA101 BKR NO: 5 PNL NO: 2VBS-PNLB101 BKR Noo 4 0 DIST PNL NO: 2CEC-PNL826 g LOOP ~+ ~+ c+ IMPACT: LOSS OF POWER TO 2CEC-PNL826 ~o 'V 6 ~ 4+ DESCRIPTION/OTHER RB CLOSE LOOP COOLING WATER FLOW FROM DRYWELL 2DRS-UC1B-TEMP REACTOR FEEDWATER PUMP P1B SUCT. PRESS SPENT FUEL POOL WATER LEVEL RB CLOSE LOOP COOLING WATER FLOW FROM 2RCS*P1B COOLERS 4TH POINT HEATER NO. 2CNM-E4B DRAIN PUMP DISCHARGE PRESSURE 2CNM-68B LSK-6-3 2HDL-4B LSK 4-2. 1B X FO FC REACTOR FEED PUMP 2FWS-P1B SUCTION FLOW/LOSS OF CONTROL OF 2FWR-FV2B 4TH PT HTR 2CNM-E4B WATER LEVEL/LOSS OF CONTROL OF LV4B LP5 PAGE 3 OF 3 UPS NO: 2VBB-UPS 1A PLANT PNL'O: 2VBS-PNLA101 BKR NO 5 PNL NO: 2VBS-PNLB101 BKR NO 4 0 DIST PNL NO: 2CEC-PNL826 ~C e LOOP ~+ z+ ~+ ~+ 4 IMPACT: LOSS OF POWER TO 2CEC-PNL826 0 + ~~~ 0+ ~~ y~ e~ ~C ~ o 'V 6 M 4 DESCRIPTION/OTHER 2CNM-45B LSK-4-1.1j 2FWS-9B LSK-6-1.1D 2CNM-38B LSK-4-1.3K 2HDL-35B LSK-4-2.1C 2FWS-40B LSK-6-1.1E 2CCP-108 LSK-9-1B 2CCP-125 LSK-9-1A X X FO FO FO CONDENSER CND1B VACUUM REACTOR FEEDWATER PUMP P1B DISCHARGE PRESSURE CONDENSATE BOOSTER PUMP P2B FLOW/LOSS OF CONTROL OF FV38B 4TH POINT HEATER 2CNM-E4B DRAIN PUMP P1B FLOW/LOSS OF CONTROL OF FV35B REACTOR INLET PRESSURE CLOSED LOOP COOLING SYSTEM HEAT EXCHANGE DISCHARGE TEMP/LOSS OF CONTROL OF TV108 RB CL LP COOLING HEADER PRESSURE LP5 'I f qE E UPS NO: 2VBB-UPS1A PNL NO: 2VBS-PNLA102 DIST PNL NO: 2CEC-PNL886 BKR NO 6 PAGE 1 OF 3 LOOP E 2AAS-113 2ARC-20B 2ASS-116 2ASS-122 2ASS-127 2CMS-72E,F 2CMS-168 2CND-267 2CND-281 2CPS-102 2CPS-103 PLANT IMPACT: LOSS OF POWER TO 2CEC-PNL886 0 "v + + A ~o 6 M 4 DESCRIPTION/OTHER X BRTHG AIR HDR PRESS AIR REMOVAL PUMP-1B SEAL WTR TEMP HI CLN STM REBLR STM PRESS AIR EJECTOR 2A STM PRESS OFFGAS STM PRESS CONTAINMENT DEW POINT SUPPR CHAMBER PRESS NORM LOW CONDUCTIVITY WASTE TK13 LEVEL NEUTRALIZING TK12 LEVEL ENABLES 2CPS-FN1 (DISCH TO STOP PERMISSIVES DRYWELL) ENABLES 2CPS-FN1 (DISCH TO STOP PERMISSIVES SUPPR PL) 2CPS-125 2CPS-126 FC PRIMARY CONTAINMENT PRESS N2 INLET FLOW/LOSS OF CONTROL OF FV125 PRIMARY CONTAINMENT INLET NITROGEN FLOW LP6 UPS NO: 2VBB-UPS1A PNL NO 2VBS-PNLA102 DIST PNL NO: 2CEC-PNL886 BKR NO: 6 PAGE 2 OF '3 LOOP PLANT 'IMPACT: LOSS OF POWER TO 2CEC-PNL886 0 "v + O + 4 ~O G 'V G 'V 4 DESCRIPTION/OTHER 2CPS-127 2GSN-138 2HRS-7C 2HRS-8C 2HRS-9C 2IAS-101 2IAS-178 2IAS-183 2MWS-114 2MWS-116 2SSR-124 2SSR-125 2SST-106 2SST-113 X X X X X X X X X X X X X X X X X X X X X X X X X X X PRIMARY CONTAINMENT INLET NITROGEN PRESS PRIMARY CONTAINMENT N2 PURGE TEMP LOW 2TMS-T2C LP TURB INL FR 2MSS-E1A 2TMS-T2C HRS FROM 2MSS-E1A 2TMS-T2C HRS FROM 2MSS-E1B INSTR AIR HDR PRESS ADS CPRSR RCVR TANK
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X X
X X
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X X
X X
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PAGE 1 OF 3
UPS NO 2VBB-UPS 1B PLANT PNL NO:
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X X
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FC CNDS STOR TK1B LEVEL CNDS XFR HDR FLOW HIGH/INHIBITAUTO START OF 2CNS-P1A OR 2CNS-P1B COND HOT WELL LEVEL 4TH PT HTR EXTR ST PRESS 3RD PT HTR EXTR ST PRESS 2ND PT HTR SHELL PRESS ElB STEAM SUPPLY PRESS E1B REHEATING STEAM TEMP E1B REHEATING STEAM FLOW TURB 4TH STG EXTR ST PRESS 5TH PT HTR 2CNM-ESB EXTR ST.
PRESS MS/R REGULATED STEAM PRESS/LOSS OF CONTROL OF REHEAT CONTROL VALVES PV28B&PV29B LP4B
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X X
X X
X FO FC FO FAI 2TMS-T2B HRS FROM 2MSS-E1B AUX. STM FLOW TO AIR REMOVAL DRAIN RECEIVER HIGH LEVEL/LOSS OF CONTROL OF LV68B NORM WTR LVL DR CONTROL/LOSS OF CNTRL OF LVX65B, LVY65B AND LVZ65B MOISTURE SEP.
DRAIN RECEIVER TK4B WTR LEVEL MOIST SEP DR RCVR TK4B WTR LEVEL/LOSS OF CONTROL OF LV78B SCAVENGING STEAM LINE, B PRESSURE/LOSS OF CONTROL OF MOV 86B 1ST PT HTR 2CNM-E1B SHELL PRESS 6TH PT HTR 2FWS-E6B EXTR ST.
PRESS LP4B
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UPS NO:
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X FC REHEATER 1B SHELL PRESSURE/LOSS OF CONTROL OF PV28Ai
- PY28B, PV29A PY29B HIGH PRESSURE TURBINE EXH.
COLD REHEATER STEAM TO 2MSS-E1B SHELL PRESSURE.
HP TURB EXH COLD REHEATER STEAM TO 2MSS-E1B SHELL PRESSURE.
HP TURB EXH COLD REHEATER STEAM TO 2MSS-E1B SHELL PRESSURE.
HP TURB EXH.
COLD REHEATER STEAM TO 2MSS-E1B SHELL PRESS'EHEATER B COLD REHEAT PRESS.
CONTAINMENT DEW POINT LP4B
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UPS NO:
2VBB-UPS1A PNL NO 2VBS-PNLA102 DIST PNL NO:2CEC-PNL884 BKR NO:
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LOOP IMPACT:
LOSS OF POWER TO 2CEC-PNL884 O
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DESCRIPTION/OTHER 2ASS-6A 2MSS-103 2WTS-243 2DRS-68A 2FOF-101 2DSR-78A 2MSS-104 2CNS-8A X
X X
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X X
X X
X X
X X
X FO FAI FO AUX. STEAM FLOW TO AIR REMOVAL SYSTEM 2TMS-T1 FIRST STAGE PRESSURE/GROUP 1
DRAIN VALVES NEUTRALIZER TANK LEVEL HIGH/INHIBIT THE START OF REGENERATION REHEATER DRAIN RECEIVER TANK 6A WATER LEVELS LOSS OF CONTROL OF LV-68A FUEL OIL STORAGE TANK LEVEL FIRE COMPUTER ALARM SCAVENGING STEAM LINE A PRESSURE LOSS OF CONTROL OF MOV-86A 2TMS-TI 1ST STAGE PRESSURE/LOSS OF CONTROL OF AOV85AiBiCiDi 2CNM-AOV101 AND 2CNM-AOV109 CNDS.
STOR TK1A LEVEL LP4
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UPS NO:
2VBB-UPS1A PNL NO: 2VBS-PNLA102 DIST PNL NO:2CEC-PNL884 BKR NO:
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LOOP 2MMS-101 2MSS-24A 2DSR-103 PLANT IMPACT: 'OSS OF POWER TO 2CEC-PNL884 O
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DESCRIPTION/OTHER 2TMS-Tl MAIN STEAM INLET HDR.
PRESSURE LOSS OF 2ASS-STV112
& 2ASS-AOV145 ElA REHEATING STEAM TEMP SCAVENGING STEAM
'EADER PRESSURE 2MSS-22A 2CRS-1A 2CRS-2A 2CRS-3A 2CRS-16A 2CRS-19A FC MS/R REGULATED STEAM PRESSURE LOSS OF CONTROL OF 2MSS-
'V28A AND 29A HIGH PRESSURE TURBINE EXHAUST COLD REHEATER STEAM TO 2MSS-E1A SHELL PRESSURE.
H.P.
TURB. EXH.
COLD REHEATER STEAM TO 2MSS-E1A SHELL PRESSURE.
H.P.
TURBINE EXH.
COLD REHEATER STEAM TO 2MSS-E1A SHELL PRESSURE.
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COLD REHEATER STEAM TO 2MSS-E1A SHELL PRESSURE REHEATER A COLD REHEAT PRESSURE LP4
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2VBB-UPS1A PNL NO 2VBS-PNLA102 DIST PNL NO:2CEC-PNL884 BKR NO 4
PAGE 3
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LOOP 2ESS-1A 2ESS-4A 2ESS-7A 2ESS-12A 2ESS-17A 2ESS-25A 2ESS-31A 2HRS-7A 2HRS-8A 2HRS-9A 2MSS-23A PLANT IMPACT:
LOSS OF POWER TO 2CEC-PNL884 O
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2CNM-E4A EXTR.
STEAM PRESSURE 5TH POINT HTR 2CNM-E5A EXTR STEAM PRESSURE 2TMS-T2A LP TURBINE INL PRESS.FR 2MSS-E1A 2TMS-T2A HRS FROM 2MSS-E1B 2TMS-T2A HRS FROM 2MSS-E1B E1A REHEATING STEAM FLOW LP4
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2VBB-UPS1A PNL NO:
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.. ~I AIIQ 5I.~ TITSI TECHNICAL EVALUATION REPORT:
A REVIEW OF REACTOR TRIP SYSTEM AVAILABILITYANALYSES FOR GENERIC LETTER 83-28, ITEM 4.5. 3, RESOLUTION 5 Avfv01ISI David P.
Mackowiak John A. Schroeder
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Box 1625 Idaho Fall s, ID 83415 I STQIISQ<<sssG 010Asss IATIQIIssAIII AIIOIsAICI<<0 A001ISS
~ svfsvw CA Cssfvs Instrumentation and Control Systems Branch Division of Engineering and System Technology Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, OC 20555
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ZQT~ <<<<ss The Idaho National Engineering Laboratory (INEL) conducted a technical review of the commercial nuclear reactor licensees'esponses to the requirements. of the Nuclear Regulatory Commission's (NRC's) Generic Letter 83-28 (GL 83-28),
Item 4.5.3.
The results of this review, if all plants are shown to be covered by an adequate analysis, will provide the NRC staff with.a basis to close out this issue with no further review.
The licensees, as the four vendors'wners'roups, submitted analyses to the NRC either directly in response to GL 83-28, Item 4.5.3, or to provide a basis for requesting changes to the Technical Specifications (TSs) that would extend the Reactor Protection System (RPS) surveillance test intervals (STIs).
To conduct the review, the INEL defined three criteria to determine the adequacy, the plant applicability, and the acceptability of the results.
The INEL examined the Owners Groups'eports to determine if the analyses and results met the established criteria.
Fort St. Vrain's responses to Item 4.5.3 were also reviewed.
The INEL review results show that all licensees of currently opera-ting commercial nuclear reactors have adequately demonstrated that their current on-line RPS test intervals meet the requirements of GL 83-28, Item 4.5.3.
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