ML18011A636

LER 94-003-01:on 940718,improperly Analyzed Single Failure in Esws Occurred That Could Potentially Result in Charging/ Safety Injection Pumps Not Performing Safety Function.Caused by Inadequate Reviews.Review conducted.W/941110 Ltr

ML18011A636
Person / Time
Site:	Harris
Issue date:	11/10/1994
From:	Donahue J, Verrilli M CAROLINA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
HO-941057, LER-94-003-01, LER-94-3-1, NUDOCS 9411160233
Download: ML18011A636 (6)

Text

P&IC3R.I I"V 1 (ACCELERATED RIDS PROCESSli REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)ACCESSION NBR:9411160233 DOC.DATE: 94/11/10 NOTARIZED:

NO FACIL:50-400 Shearon Harris Nuclear Power Plant, Unit 1, Carolina AUTH.NAME AUTHOR AFFILIATION VERRILLI,M.

Carolina Power 6 Light Co.DONAHUE,J.W.

Carolina Power 6 Light Co.RECIP.NAME RECIPIENT AFFILIATION DOCKET N 05000400

SUBJECT:

LER 94-003-01:on 940718,improperly analyzed single failure in ESWS occurred that could potentially result zn charging/safety injection pumps not performing safety function.Caused by inadecpxate reviews.Review conducted.W/941110 ltr.DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR

!ENCL (SIZE: TITLE: 50.73/50.9 Licensee Event Report (LER), Tncident Rpt, etc.NOTES:Application for permit renewal filed.05000400 RECIPIENT ID CODE/NAME PD2-1 PD INTERNAL: ACRS AEOD/SPD/RRAB NRR/DE/EELB NRR/DORS/OEAB NRR/DRCH/HICB NRR/DRSS/PRPB NRR/DSSA/SRXB RES/DSIR/EIB EXTERNAL: L ST LOBBY WARD NOAC MURPHY,G.A NRC PDR COPIES RECIPIENT LTTR ENCL ID CODE/NAME 1 1 LE,N 1 OD/ROA'B/DSP 1 FIM C~EN ERA 02 1 NRR/DE/EMEB 1 NRR/DRCH/HHFB 1 NRR/DRCH/HOLB 2 NRR/DSSA/SPLB 1 NRR/PMAS/IRCB-E 1 RGN2 FILE Ol 1 1 LITCO BRYCE,J H 1 1 NOAC POORE,W.1 1 NUDOCS FULL TXT COPIES LTTR ENCL 1 1 2 2 1 1 1 1 1 1 1 1 1 1 1 1 1 1 2 2 1 1 1 1 NOTE TO ALL"RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE iVASTE!CONTACTTHE DOCL'ifENTCONTROL DESK, ROOM P I-37 (EXT.504-2083)TO EL Lif I NATE YOl R NA if E F ROil DISTRIBUTION LISTS I'OR DOC!.'iif ENTS YOU DON" I'EED!FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 27 ENCL 27 Carolina Power&Light Company Harris Nuclear Plant PO Box 165 New Hill NC 27562 g0V i O 198't~Letter Number: 'HO-941057 U.S.Nuclear Regulatory Commission ATTN: NRC Document Control'Desk Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT UNIT 1 DOCKET NO.50-400 LICENSE NO.NPF-63 LICENSEE EVENT REPORT 94-003-01 Gentlemen:

In accordance with Title 10 to the Code of Federal Regulations, the enclosed Licensee Event Report is submitted.

The original report fulfilled the requirement for a written submittal within thirty (30)days of a reportable occurrence and was in accordance with the format set forth in NUREG-1022, September 1983.This revision is being submitted to provide supplemental information following completion of additional engineering evaluation related to the identified condition.

Sincerely, J.W.D nahue General Manager Harris Plant MV Enclosure cc: Mr.S.D.Ebneter (NRC-RII)Mr.N.B.Le (NRC-PM/NRR)Mr.S.A.Elrod (NRC-HNP)Mr.W.R.Robinson 94iii602M 941110 PDR ADOCK 05000400 S PDR State Road 1134 New Hill NC NRC FORM 366 (5-92)U.S.NUCLEAR REGULATORY COHMISSION LICENSEE EVENT REPORT (LER)(See reverse for required number of digits/characters for each block)APPROVED BY OHB NO.3150.0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MHBB 7714), U.S.NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150.0104), OFFICE OF MANAGEHENT AND BUDGET WASHINGTON DC 20503.FACILITY NAME (1)Shearon Harris Nuclear Plant-Unit

¹I DXKET NUMBER (2)05000/400 PAGE (3)1OF4 TITLE (4)Improperly analyzed single failure in the Emergency Service Water System that could potentially result in the Charging/Safety Injection Punps not performing their safety function.EVENT DATE (5 LER NUMBER 6)REPORT DATE 7 OTHER FACILITIES INVOLVED B MONTH DAY YEAR 7 18 94 YEAR SEQUENTIAL NUMBER 94--003 REVISION NUMBER MONTH DAY 11 10 YEAR FACILITY HAME FACILITY NAME DOCKET NUMBER 05000 DOCKET NUMBER 05000 OPERATING NODE (9)POMER LEVEL (10)100X THIS REPORT IS SUBMITTED PURSUANT 20'02(b)'0.405(a)(1)(i) 20.405(a)(1)(ii) 20.405(a)(1)(iii) 20.405(a)(1)(iv) 20.405(a)(1)(v) 20.405(c)50.36(c)(1) 50.36(c)(2) 50'3(a)(2)(i) 50.73(a)(2)(II) 50.73(a)(2)(iii) 50.73(a)(2)(iv) 50.73(a)(2)(v) 50.73(a)(2)(vii) 50.73(a)(2)(viii)(A) 50.73(a)(2)(viii)(B) 50.73(a)(2)(x) 73.71(b)73.71(c)OTHER (Specify in Abstract belox and in Text, HRC Form 366A)TO THE RE UIREMENTS OF 10 CFR 5: (Check one or mor e)(11)LICENSEE CONTACT FOR THIS LER (12)Michael Verrilli Sr.Specialist

-Licensing TELEPHONE NUMBER (Include Area Code)(919)362-2303 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE TO HPRDS CAUSE SYSTEM COMPOHENT MANUFACTURER REPORTABLE TO HPRDS SUPPLEMENTAL REPORT EXPECTED 14 YES (If yes, complete EXPECTED SUBMISSION DATE).X NO EXPECTED SUBMISSION DATE (15)MONTH DAY YEAR ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typexritten tines)(16)On July 18, 1994, a potential single failure was identified in one train of the Emergency Service Water System (ESW)that, without reasonably expected operator action, might prevent the Charging/Safety Injection Pumps (CSIP)from performing their safety function due to inadequate oil cooling.This postulated scenario would occur due to elevated temperature cooling water being supplied to the CSIP Oil Coolers.Specifically, if the"A" train auxiliary reservoir return valve (1SW-270)failed to open, the"A" train of ESW would be dead-headed at a higher pressure than the"B" train.This difference in pressure, coupled with a pressure gradient within the"A" train due to the action of the Service Water Booster Pump, would result in cooling water back-flowing through the Emergency Diesel Generator (EDG)jacket water coolers before entering the CSIP oil coolers.With the EDG running and loaded, the back flow would be heated above the oil cooler maximum design temperature.

This condition was caused by not adequately identifying and considering the effects of an active valve failure on the cross-connection at the CSIP's during initial ESW design and operational valve line-up development.

Subsequent system design assessments and reviews also did not identify the abnormal flow path.This was due to the scope of these reviews, which focused on passive type failures such as line breaks in the common piping between headers and their effects on opposite train operability, not the detailed flow and temperature analysis that eventually identified the cited condition.

Additional engineering analysis concluded that, upon normally expected identification and mitigating actions by operators in the main control room, the event would have been terminated prior to any adverse effects on the CSIP's.Immediate corrective actions for this event were to close the train cross connects at the CSIP lube oil coolers, thus eliminating the possibility of the described scenario.Additional corrective actions include further engineering review and evaluation in the Emergency and Normal Service Water Systems as well as other systems that may have similar cross connection problems.

NRC FORH 366A (5-92)U.S.NUCLEAR REGULATORY COHHISSION LICENSEE EVENT REPORT (LER)BY OHB NO.3150.0104 EXPIRES 5/31/95 EST IHATED BURDEN PER RESPONSE TO COHPLY WITH THIS INFORHATION COLLECTION REQUEST: 50.0 HRS.FORWARD COHHENTS REGARDING BURDEN ESTIHATE TO THE INFORHATION AND RECORDS HANAGEHENT BRANCH (MNBB 7714), U.S.NUCLEAR REGULATORY COHMISSION, WASHINGTON, DC 20555-0001 AND TO THE PAPERWORK REDUCTION PROJECT (3140-0104), OFFICE OF MANAGEMENT AND BUDGET WASHINGTON DC 20503.FACILITY NAME (1)Shearon Harris Nuclear Plant-Unit¹1 DOCKET NUHBER (2)05000/400 LER NUMBER (6)YEAR SEQUENTIAL REVISION PAGE (3)2OF4 TEXT (If more space fs required, use additional copies of NRC Form 366A)(17)94 003 01 EVENT DESCRIPTION:

On July 18, 1994, a potential single failure was identified in one train of the Emergency Service Water System (ESW)that, without reasonably expected operator action, might prevent the Charging/Safety Injection Pumps (CSIP)from'performing their safety function due to inadequate oil cooling.This postulated scenario would occur due to elevated temperature cooling water being supplied to the CSIP Oil Coolers, which would eventually lead to oil breakdown and subsequent loss of lubrication.

This condition was identified when seat leakage through a shut Normal Service Water isolation valve caused unexpected pressurization of one of the ESW return headers.Because both trains of ESW were cross-connected at the CSIP Oil Coolers at both the supply and return headers, plant personnel became concerned about the effect of backpressure on flow adequacy through the coolers.Failure modes leading to pressurization were reviewed and the effects of the pressurization evaluated.

As a result, an unanalyzed single active failure effect was identified.

Specifically, if the"A" train auxiliary reservoir return valve (1SW-270)failed to open, the"A" train of ESW would be dead-headed at a higher pressure than the"B" train.This difference in pressure, coupled with a pressure gradient within the"A" train due to the action of the Service Water Booster Pump, would result in cooling water back-flowing through the Emergency Diesel Generator (EDG)jacket water coolers before entering the CSIP oil coolers.With the EDG running and loaded, the back flow would be heated above the oil cooler maximum design temperature.

The most conservative minimum time for this heated water to arrive at the CSIP inlet was just over nine minutes.Adequate flow would exist during this time interval and the CSIP's would continue to operate normally prior to the arrival of the hot water.The scenario involving the postulated failure of 1SW-270 was reported to the Nuclear Regulatory Commission on July 18, 1994 as a four hour non-emergency notification per the immediate notification requirements of 10CFR50.72 and was followed by a written submittal on August 17, 1994.Additional engineering analysis was performed by Carolina Power&Light and vendor engineering personnel following submittal of the original LER.A similar scenario was also run on the plant simulator by six operating crews.It was concluded that upon normally expected identification and mitigating actions by operators in the main control room, the event would be terminated prior to any adverse effects on the CSIP's.Detailed aspects of the engineering analysis are provide in the"Safety Significance" section of this report.No previous reports have been submitted related to the failure to properly perform analysis of pertinent safety system component functions.

CAUSE:This condition was caused by not adequately identifying and considering the effects of an active valve failure on the cross-connection at the CSIP's during initial ESW design and operational valve line-up development.

Subsequent system design assessments and reviews also did not identify the abnormal flow path.This was due to the scope of these reviews, which focused on passive type failures such as line breaks in the common piping between headers and their effects on opposite train operability, not the detailed flow and'temperature analysis that eventually identified the cited condition.

NRC FORM 3 A (5-92)

NRC FORM 366A (5-92)U.S.NUCLEAR REGULATORY COHHISSION LICENSEE EVENT REPORT (LER)APPROVED BY OHB NO.3150.0104 EXPIRES 5/31/95 ESTIHATED BURDEN PER RESPONSE TO COHPLY MITH THIS INFORHATION COLLECTIOH REQUEST: 50.0 HRS.FORMARD COHHENTS REGARDING BURDEN ESTIHATE TO THE INFORHATION AND RECORDS HANAGEHENT BRANCH (HNBB 7714);U.S.NUCLEAR REGULATORY COHHISSION, MASHINGTON, DC 20555-000'I AND TO THE PAPERMORK REDUCTION PROJECT (3150-0104), OFFICE OF HANAGEHENT AND BUDGET MASHINGTON DC 20503.FACILITY NAHE (1)Shearon Harris Nuclear Plant-Unit¹1 DOCKET NUHBER (2)YEAR 05000/400 94 LER NUHBER (6)SEQUENTIAL 003 REVISION 01 PAGE (3)3 OF 4 TEXT (If more space is required, use additional copies of NRC Form 366A)(17)SAFETY SIGNIFICANCE:

To address the significance, consequences and impact on CSIP safety function capability, an Engineering Evaluation (PCR-7379) was performed.

This evaluation represents a conservative assessment of the impact resulting from the postulated failure of a single plant component (failure of 1SW-270 to open), coincident with a loss of off-site power.This analysis determined through system flow analysis using the KYPIPE code, that under the worst case failure scenario, an adequate cooling water flow rate would be delivered to the CSIP oil coolers.However, it also identified that this cooling water would be at elevated temperatures.

This elevated temperature would result from the ESW back flow through the"A" EDG jacket water cooler.The back flow potential was caused by the difference in pressure between the"A" train and"B" train, coupled with a pressure gradient within the"A" train caused by the action of the"A" ESW Booster Pump.(A similar condition would not develop in the"B" train due to differences in piping configuration.)

The analysis determined the maximum cooling water temperature exiting the EDG jacket water cooler to be approximately 274 degrees F and the most conservative transit duration for this high temperature water to reach the CSIP lube oil coolers to be just over nine minutes.The most limiting component affected by this high temperature cooling water and resulting high temperature CSIP lube oil is the motor-to-pump speed increaser gear box.As reported by the manufacturer, continued operability could not be guaranteed at these elevated temperatures.

To prevent adverse effects and possible failure of the CSIP's, operators in the main control room would have to identify the abnormal ESW system alignment and resulting degraded flow condition within the approximate nine to ten minute interval, then terminate the event by either opening 1SW-270, securing the"A" train ESW Pump, securing the running EDG, or restoring the Normal Service Water alignment to the"A" train ESW header.Based on procedure guidance and demonstrated operator performance on the plant simulator, the abnormal ESW"A" train flow patli would be readily apparent from indicated ESW flow rates and pressures on the main control board.Verification of a running ESW pump is directed by Emergency Operating Procedures at this point, which would prompt operators to verify expected flow rates.There will also be several main control board annunciators in alarm that would alert the operators to this condition.

Additionally, an independent verification of Engineered Safety Feature Actuation is performed by the Shift Technical Advisor, which would clearly indicate which components had not gone to their required position.Based on this combination of control board indication, alarm annunciation, procedural guidance and the STA's verification, there is a very high level of confidence that the control room staff would appropriately identify the condition and take corrective actions.This confidence was validated by plant simulator training, where six HNP operating crews were faced with a similar scenario and each properly took measures to terminate the event in less than four minutes.CORRECTIVE ACTIONS: 1.Immediate corrective actions were to close the system cross connect valves at the CSIP lube oil coolers.2.PCR-7379 was generated to evaluate the impact and consequences of this condition.

3.The Service Water System Single Failure Review was updated to include the described potential scenario.NRC FORH 366A (5-92)

NRC FORM 366A (5-92)U.S.NUCLEAR REGULATORY COHHISSION LXCENSEE EVENT REPORT (LER)PROVE 0104 OHB NO.3150.EXP IRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORHATION COLLECTION REQUEST: 50.0 HRS.FORNARD COHMENTS REGARDING BURDEN ESTIHATE TO THE INFORHATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714)i U S NUCLEAR REGULATORY COMMISSIONS MASHINGTON, DC 20555-0001 AND TO THE PAPERIIORK REDUCTION PROJECT (3110-0104), OFFICE OF HANAGEHENT AND BUDGET lrASHINGTON DC 20503.FACILITY NAME (1)Shearon Harris Nuclear Plant-Unit¹1 DOCKET NUMBER (2)05000/400 YEAR LER NUHBER (6)SEQUENTIAL 003 REVISION 01 PAGE (3)4OF4 TEXT (If more space fs required, use additional copies of NRC Form 366A)(17)CORRECTIVE ACTIONS: (cont.)4.A general review was performed on other Decay Heat Removal Systems for cross connections that could have adverse system operational effects.This included the High and Low Head Safety Injection, Component Cooling Water and Auxiliary Feedwater systems.No similar vulnerabilities were found.5.Additional engineering review will be performed by conducting a Self Service Water Operational Performance Inspection, which includes a reconstitution of actions taken as a result of Generic Letter¹89-13.6.Single Failure Training and guidance will be provided for appropriate Nuclear Engineering Personnel.

EIIS INFORMATION:

N/A NRC FORH 366A (5-92)