ML18012A381

LER 96-010-02:on 960628,identified Surveillance Testing Deficiencies That Caused Past Entries Into TS 3.0.3.Caused by Personnel Error.Surveillance Test Procedures OST-1008 & OST-1092 revised.W/960927 Ltr

ML18012A381
Person / Time
Site:	Harris
Issue date:	09/27/1996
From:	Donahue J, Johnny Eads CAROLINA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
HNP-96-159, LER-96-010-02, LER-96-10-2, NUDOCS 9610040148
Download: ML18012A381 (10)

Text

UA.'l'P'(90RY REGULATORY INFORMATION DZSTRZBUTZ N SYSTEM (RIDE)

ACCESSION NBR:9610040148 DOC.DATE: 96/09/27 NOTARIZED: NO DOCKET FACIL:50-400 Shearon Harris Nuclear Power Plant, Unit 1, Carolina 05000400 AUTH. NAME AUTHOR AFFILIATION EADSFJ. Carolina Power S Light Co.

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

RECIP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER 96-010-02:on 960628,CVI Radiation Monitors provided C indication in Main Control Room. Caused by inadequate surveillance test procedures. Surveillance test procedure OST-1008 & OST-1092 have been revised.W/960927 ltr.

DISTRIBUTION CODE: ZE22T COPIES RECEIVED:LTR I ENCL TITLE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc.

3 SIZE:

NOTES:Application for permit renewal filed. 05000400 G RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD2-1 PD 1 1 LE,N 1 1 INTERNAL: ACRS 1 1 OD4QPD/RA 1 1 AEOD/SPD/RRAB 1 1 FILE CXN 1 1 NRR/DE/ECGB 1 1 NRR/DE/EELB 1 1 NRR/DE/EMEB 1 1 NRR/DRCH/HHFB 1 1 NRR/DRCH/HICB 1 1 NRR/DRCH/HOLB 1 1 NRR/DRCH/HQMB 1 1 NRR/DRPM/PECB 1 1 NRR/DSSA/SPLB 1 1 NRR/DSSA/SRXB 1 1 D RES/DSIR/EIB 1 1 RGN2 FILE 01 1 1 EXTERNAL: L ST LOBBY WARD 1 1 LITCO BRYCEFJ H 1 1 NOAC MURPHY,G.A 1 1 NOAC POOREFW. 1 1 NRC PDR 1 1 NUDOCS FULL TXT 1 1 U

E N

NOTE TO ALL "RIDS" RECIPZENTS:

PLEASE HELP US TO REDUCE WASTE. TO HAVE YOUR NAME OR ORGANIZATION REMOVED FROM DISTRIBUTION LISTS OR REDUCE THE NUMBER OF COPIES RECEIVED BY YOU OR YOUR ORGANIZATION, CONTACT THE DOCUMENT CONTROL DESK (DCD) ON EXTENSION 415-2083 FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 24 ENCL 24

Carolina Power & Light Cotnpany Harris Nuclear Plant PO Box 165 New Hill NC 27562 SEP 87 1996 U.S. Nuclear Regulatory Commission Serial: HNP-96-159 ATTN: NRC Document Control Desk 10CFR50.73 Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT UNIT 1 DOCKET NO. 50-400 LICENSE NO. NPF-63 LICENSEE EVENT REPORT 96-010-02

Dear Sir or Madam:

In accordance with Title 10 to the Code of Federal Regulations, the enclosed revision to Licensee Event Report 96-010 is submitted. This revision provides the safety significance discussion of the previously reported deficiencies that caused Technical Specification 3.0.3 entries during past testing. An updated status of corrective actions is also provided.

Sincerely, J. W. Donahue Director of Site Operations Harris Plant JHE/jhe Enclosure c: Mr. J. B. Brady (NRC Sr. Resident Inspector - HNP)

Mr. S. D. Ebneter (NRC Regional Administrator - RII)

Mr. N. B. Le (NRC - Project Manager/NRR) 9610040i48 960927 PDR ADQCK 05000400 S PDR State Road 1134 New Hill NC

l NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 FLS5) EXPIRES 04/30/96 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS MANDATORY INFORMATION CO(LECT)ON REOUEST: 5M HRS. REPORTED LESSONS LEARNED ARE LICENSEE EVENT REPORT (LER) DICORPORATED INTO THE UCENSING PROCESS ANO FED BACK TO O(DUSTRY.

FORWARD COMMENTS REGARDDIG BURDEN ESTIMATE TO THE INfORMATIDN AND RECORDS MANAGEMENT BRANCH IT@ F33L US. NUCLEAR REGU(ATORY COMMISSION, (See reverse for required nu(nber of WASHDIGTON, OC 20555000l, ANO TO THE PAPERWORK REDUCTION PR(LIECT (3150.

digits/characters for each block) 0(0(L Off)CE OF MANAGEMENT AND BUDGET, WASHINGTON, OC 20503.

FACILITYNAME (1) DOCKET NUMBER (3) PAGE (3)

Harris Nuclear Plant Unit-1 50-400 1 OF 6 TITLE (4)

Surveillance testing deficiencies that caused past entries into TS 3.0.3.

EVENT DATE (5) LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (6)

FACILITYNAME DOCKET NUMBER SEBUENTIAL REVISION MONTH OAY YEAR MONTH DAY YEAR NUMBER NUMBER 05000 FACIUTY NAME DOCKET NUMBER 06 28 96 96 010 02 09 27 96 05000 OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR 5: (Chock one or more) (11)

MODE {9) 20.2201 (b) 20.2203(a)(2) (v) X 50.73(a) I2)(i) 50.73(a)(2)(viii) 20.2203(a) {1) 20.2203(a)(3)(i) 50.73(al(2)(ii) 50.73(a) (2)(x)

POWER 100%

LEVEL (10) 20.2203(a) (2)(i) 20.2203(a)(3)(ii) 50,73(a) (2) (iii) 73.71 20.2203(a) (2) {ii) 20.2203(a) (4) 50.73(a)(2)(iv) OTHER 20.2203(a)(2) (iii) 50.36(c) (1) 50.73(a)(2)(v) specrly rn Abstract below or in NRC Form 366A 20.2203(a) (2) liv) 50.36(c) (2) 50.73(a) (2) (vii)

LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMBER ((ncrcrde Area Code)

Johnny Eads Project Engineer - Licensing {919) 362-2646 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

REPORTABLE REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER CAUSE SYSTEM COMPONENT MANUFACTURER TO NPROS TO NPROS re;; rc (Ic

@Sgc:'.jtkc cb. 'Tcc,4t)b SUPPLEMENTAL REPORT EXPECTED (14) EXPECTED MONTH DAY YEAR YES SUBMISSION

{It yes, complete EXPECTED SUBMISSION DATE). X NO DATE (15)

ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single. spaced typewritten lines) (16)

On June 14, 1996 with the plant operating in Mode-1 at 100% power, Operations personnel identified a deficiency in the quarterly Residual Heat Removal {RHR) System surveillance test procedures {OST-1008 Br OST-1092). Section 7.1 verifies that the RHR pump discharge check valves properly back seat during system operation. Prior to performing this section of OST-1008, Operators realized that during the check valve back seat test, a system alignment is established that cross connects the operable RHR train, with the inoperable train being tested, and in this condition the operable train would be incapable of providing the minimum required low head safety injection flow to the Reactor Coolant System. While in this cross-connected test alignment, the plant is actually in Technical Specification 3.0.3, which is reportable per 10CFR50.73. On June 28, 1996 while investigating LER 96-010 Revision 0, one additional surveillance testing deficiency was identified. With the plant operating in Mode-1 at 100% power, Operations personnel identified a deficiency in Maintenance Surveillance Test MST-(0417, "Containment Ventilation Isolation Area Radiation Monitors Relay Actuation Test." The surveillance test as written caused both trains of Containment Vacuum Relief System to become inoperable requiring entry into Technical Specification {TS) 3.0.3. This deficiency was first identified by Operations personnel on December 11, 1995. However, the significance and reportability of this deficiency was not recognized.

These conditions were caused by inadequate surveillance test procedures resulting from personnel errors during revisions to OST-1008 and OST-1092 in October 1992 and during original procedure development of MST-(0417 in December 1987.

Subsequent technical and safety reviews also failed to identify that the test procedures resulted in a Technical Specification 3.0.3 entry.

Immediate corrective actions included not performing the scheduled tests as written and placing these procedures on administrative hold. Procedure revisions were then completed for OST-1008( OST-1092 and MST-(0417 to allow surveillance testing without entry into TS 3.0.3. Additional actions included reviewing this event with appropriate personnel . The reportability requirements of TS 3.0.3 entry and restrictions related to voluntary TS 3.0.3 entry were a(so reinforced with licensed operators and included in Licensed Operator training programs. In addition, sampling of additional procedures to identify any similar deficiencies was also completed. No additional TS 3.0.3 entry problems were identified.

KRC FORM 366A US. NUCLEAR REGULATORT COMMISSION I6-96)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACIL)TT KAME II) OOCXET LBI NUMBER 16) PAGE Cn SEOUENTIAL REV610N TEAR NUMBER NUMBER Shaaron Harris Nucfaar Plant ~

Unit 0'1 50400 2 OF 6 96 - 010 - 02 TEXT Pl kooko spooorr koqvroC ooo okrCk)Cpoor onto or HRC Fokko 366lU i) 7)

EVENT DESCRIPTION:

On June 14, 1996 with the plant operating in Mode-1 at 100% power, Operations personnel in the main control room identified a deficiency in the quarterly interval Residual Heat Removal (RHR, EIIS Code - BP) Pump surveillance tests (OST-1008 and OST-1092). Section 7.1 of these tests are performed to satisfy the RHR pump In-Service-Testing Program requirements and verify that the RHR pump discharge check valves (1RH-70 and 1RH-34) properly back seat during system operation. Prior to performing this section of OST-1008, Operators realized that during the check valve back seat test, a system alignment is established that cross connects the operable RHR train, with the inoperable train being tested. Specifically, the A-train check valve (1RH-34) back seat test is performed by using the discharge pressure of the opposite loop RHR pump. The B-train RHR pump is started and is aligned to recirculate to the Refueling Water Storage Tank (RWST), with the opposite hot leg cross-over valve (1SI-326) shut. In this alignment, the B-train of RHR is completely isolated from the A-train and is fully operable. To verify that "A" RHR Pump check valve (1RH-34) is on it's backseat, 1SI-326 is then opened and a pump discharge flow measurement is taken. A flow rate increase of less than 50 gpm indicates that 1RH-34 is properly back seated. After observing and recording pump flow, 1SI-326 is shut. (reference page 6 for flow diagram)

During the time period when 1SI-326 is open, the two RHR trains are cross-connected. In this condition, the path of least resistance would be the recirculation line to the RWST. This would create the potential for a significant reduction in low head safety injection flow to the Reactor Coolant System in the event of an accident that required Safety Injection. Based on this, both trains of RHR are rendered inoperable and the plant is in Technical Specification 3.0.3, which is reportable per 10CFR50.73.

Investigation revealed that this back seat testing process was incorporated into OST-1008 and OST-1092 in October 1992 as a new testing methodology. Prior to this, the recirculation flow path to the RWST was secured prior to opening the cross-over valve, thus eliminating the potential for a reduction in low head safety injection flow during testing. Since October 1992, the A-train has been tested 18 times by performing OST-1008 and OST-1092 has been performed to test the B-train 15 times.

On June 28, 1996 while investigating LER 96-010 Revision 0, one additional surveillance testing deficiency was identified. With the plant operating in Mode-1 at 100% power, Operations personnel identified a deficiency in Maintenance Surveillance Test MST-I0417, "Containment Ventilation Isolation Area Radiation Monitors Relay Actuation Test." The surveillance test as written caused both trains of Containment Vacuum Relief System (EIIS Code - BF) to become inoperable requiring entry into Technical Specification (TS) 3.0.3.

The deficiency within MST-I0417 has existed since it was originally developed in December 1987. MST-I0417 provides instructions for performing a Relay Actuation Logic Test for Containment Ventilation Isolation System (EIIS Code - JM) actuation on a Two-of-Four High Radiation test signal from the Containment Ventilation Isolation Signal Area Radiation Monitors. This procedure satisfies part of the Monthly Surveillance Requirements of TS 4.3.2.1 (Table 4.3-2, Items 3.c.2 and 3.c.4.a) and TS 4.3.3.1 (Table 4.3-3, Item l.a). The surveillance as originally written generates a Containment Ventilation Isolation Signal which blocks the automatic Containment Vacuum Relief function which causes both trains of Containment Vacuum Relief to become inoperable. This MST deficiency has resulted in the Containment Vacuum Relief system being inoperable for approximately 45 minutes during each monthly performance of this surveillance test since December 1987.

Li NRC EORM 366A US. NUCLEAR REGULATORZ COMMISSION H.96)

LICENSEE EVENT REPORT (LERj TEXT CONTINUATION EACILITZ NAME ll) OOCKEf LER NUMOER (6) PAGE )3)

SEOUENHAL REVISION TEAR NUMOER kUMBER Shearon Harris Nuciear Plant ~

Unit )i)1 50400 3 OF 6 96 - 010 - 02 TEXT rsl mort sptstis rtqoimL ott tdittmotl rooms or Fir)C Form 3IRSU I)T)

EVENT DESCRIPTION Cont'd:

The four Containment Ventilation Isolation Radiation Monitors provide indication in the Main Control Room of the activity inside the Containment. These monitors provide a high radiation alarm when radiation levels reach preset limits. The receipt of these alarms, 2 of 4 logic, initiates a Containment Ventilation Isolation, and alerts Operations personnel of abnormal radiation inside Containment. In the event of a Containment Ventilation Isolation Signal, the Containment Vacuum Relief butterfly valve and damper of each train will receive a close signal and be prevented from opening until this signal is reset. The Containment Vacuum Relief valves are designed to prevent the differential pressure between Containment and the outside atmosphere from exceeding the design value as a result of inadvertent actuation of the Containment Spray system.

The MST-I0417 deficiency was first identified by Operations personnel on December 11, 1995. However, Operations personnel at that time did not recognize the reportability of short duration entry into TS 3.0.3 caused by surveillance testing. As a result, the Condition Report generated in December 1995 was improperly classified as a procedure improvement item and not as an adverse condition. As a result of the misclassification, the adverse condition did not receive timely corrective actions.

CAUSE:

These conditions were caused by inadequate surveillance test procedures. The RHR surveillance test deficiency resulted from personnel error during revisions to OST-1008 and OST-1092 in October 1992. These revisions implemented a change in the RHR pump discharge check valve back seat testing methodology without fully assessing the impact on the RHR system. The Containment Ventilation Isolation surveillance test deficiency resulted from a failure to fully assess the impact on Containment Vacuum Relief System operability during procedure development in December 1987. In both cases, subsequent technical and safety reviews also failed to identify that the test procedures resulted in a Technical Specification 3.0.3 entry.

The failure to recognize the significance and reportability of TS 3.0.3 entry on December 11, 1995 was the result of a personnel error.

SAFETY SIGNIFICANCE:

The safety significance discussion for both the RHR system surveillance test deficiency and the Containment Vacuum Relief system surveillance test deficiency are provided below:

RHR During performance of OST-1008 and OST-1092, the recirculation valve (1SI-331) from the RHR system back to the RWST was partially opened (9 1/2 turns per procedure step 7.1.14.b) to allow the OST required pump flow (3663 gpm 100 psid) to pass. This flow is the TS minimum flow for the RHR pumps. This is also the flow required for the testing of the injection path check valves (OST-1088). Since OST-1008 and OST-1092 (recirculation path) use the same flow requirement as OST-1088 (injection path), it is concluded that the line resistance of the recirculation path is approximately the same as the line resistance of the injection path. Since the line resistance of the recirculation path is approximately the same as the line resistance of the injection path, then the fiow through these two paths during an accident (with this test alignment) would be expected to split in half. Approximately one half of a single pump's flow would be expected to be delivered through the injection path. This delivered flow would not have met the minimum injection flow requirement for the RHR system. Also, it is expected that this one pump would exceed its maximum flow (run out) which could damage this pump.

NRC FORM 366A US. NUCLEAR REGULATORT COMMISSION FL.9SI LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (II OOCKET LER NUMBER I6) PAGE Iii SEOUENTML REVISION TEAR NUMBER NUMBER Shearon Harris Nuclear Plant - Unit Pl 50400 4 OF 6 96 - Ol0 - 02 TEXT Pl nrae guceir kkqvlkd, osr PCkfriaMI cod Pf lllICFnakk JSQ/ IITI SAFETY SIGNIFICANCE Cont'd:

However, with both trains of RHR available, using the same logic as above, it is concluded that one half of the flow from two pumps would be diverted through the recirculation path and one half delivered through the injection path.

With both pumps available, the delivered flow would be the equivalent of one pump's flow. Therefore, with both RHR trains available, the RHR system would be able to deliver the required fiow.

Additionally, an evaluation of the safety significance of this condition was made using probabilistic safety assessment (PSA). The test alignment was assumed to occur four times per year (quarterly testing) for a duration of one hour per test. It was further assumed that both RHR pumps are required to operate following a large break LOCA to provide sufficient safety injection (SI) flow to the reactor coolant system (RCS), and no credit was taken for proceduralized operator action to isolate the test flow path to the RWST for large break LOCA scenarios. For other conditions requiring operation of the RHR pumps, it was found that adequate time was available to permit restoration of the test alignment since precautions and limitations in OST-1008 and OST-1092 required an operator to be stationed at the recirculation valve to the RWST (1SI-331) and be in direct communication with the control room. The increase in annual core damage risk was determined to be 0.0065% of the nominal annual core damage frequency, which is about 6E-5 per year. This is a very small increase in core damage frequency; therefore, it is concluded that the safety significance of this condition with regard to core damage accidents is minimal.

Containment Vacuum Relief The consequences of the MST-I0417 surveillance deficiency are that both trains of the Containment Vacuum Relief system were inoperable for approximately 45 minutes during each monthly performance of MST-I0417 since December 1987. The Containment Vacuum Relief system is not relied upon to mitigate the consequences of any FSAR Chapter 15 accidents. The Containment Vacuum Relief system is designed to assure the structural integrity of the Containment against the differential pressure associated with the inadvertent operation of the Containment Spray system. If an inadvertent operation of the Containment Spray system had occurred during the performance of MST-I0417, operators would have been required to manually reset the Containment Ventilation Isolation signal from the Control Room to allow proper operation of the Containment Vacuum Relief system. A Control Room annunciator (ALB-028-5-1) is provided to alert the operators to a high vacuum in Containment. In response to the alarm, the Annunciator Panel Procedure (APP-ALB-028) requires the operator to verify that the vacuum relief valves are open if required. In addition, a Control Room annunciator (ALB-001-4-1) is provided to alert the operators to a containment spray pump start. Without appropriate operator intervention, the containment design limit of -2 psid could have been exceeded.

\

No instances of inadvertent containment spray system operation coincident with MST-I0417 performance have been identified at the Harris Plant.

These conditions are reportable per 10CFR50.73(a)(2)(i) and 10CFR50.73(a)(2)(vii).

A(4- )

NRC EORM 366A US. NUCLEAR REGUIATORT COMMISSION

)4.SS)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACIL)TT NAME (I) DOCKET LER NUMBER 16) PAGE I3)

SEOUENTIAL REVISION NUMBER NUMBER Shearon Harris Nuclear Plant Unit 0'1 50400 5 OF 6 96 - 010 - 02

'tEXT Pr moro sposo r's npr)rorL vso orrrprraool sop'os o/ FYRC Farm 3664) il))

PREVIOUS SIMILAR EVENTS:

There have been no previous Harris Plant LERs caused by deficient procedures that resulted in an inadvertent Technical Specification 3.0.3 entry.

CORRECTIVE ACTIONS COMPLETED:

1. Surveillance test procedure OST-1008 and OST-1092 have been revised to allow for required RHR system testing without cross-connecting the safety trains and entering Technical Specification 3.0.3. The revision to OST-1008 was approved on July 5, 1996 and the revision to OST-1092 was approved on June 28, 1996.

2. Surveillance test procedure MST-I0417 has been revised to prevent inoperability of both trains of Containment Vacuum Relief simultaneously. MST-I0417 was revised on July 26, 1996.

Operations personnel involved with the failure to recognize the significance and reportability of this condition on December 11, 1995 have been counselled. This counselling was completed by July 24, 1996.

This event has been reviewed with appropriate operations and maintenance personnel involved in developing and reviewing procedures. This review included insight on how this deficiency occurred and how it can be prevented in the future. This review was completed on August 1, 1996 for procedure group personnel or if unavailable by this date, the 'review was provided for personnel prior to writing and reviewing procedures.

5. The reportability requirements of TS 3.0.3 entry and restrictions related to voluntary TS 3.0.3 entry were reinforced with licensed operators. This subject was also included in Licensed Operator training programs on September 16, 1996.

An additional sample of procedures has been reviewed to identify any similar procedure deficiencies. This review was completed on September 10, 1996. No additional TS 3.0.3 entry problems were identified.

CORRECTIVE ACTIONS PLANNED:

No additional corrective actions planned.

EIIS Codes:

Residual Heat Removal System BP Containment Vacuum Relief System BF Containment Ventilation Isolation System JM A (4. )

IIRC FORM 366A US. NUCLEAR REGULATORY COMMISSION (4-95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACIUTT NAME 0) OOCXET LER NUMBER (6) PAGE (3)

SEOUENTIAL TEAR NUMBER NUMBER Shearon Harris Nuclear Plant - Unit ¹1 50400 6 OF '6 96 - 010 - 02 TEXT is/more specs/2 reqvied. vse ddcF(imd/coper ol //RC form 3agl (17)

HC/Sl PUMP RlrufIllC

~lfIDOL LAMINI SUCII ON wasf R SIORACL sc5.18 UIH lANK RCS LOOP I HCV I A 2 ISI- JJI HL, ISI JSS 151 Ils

~~RASI 28 POI Sa SSA e-Jls)

~PASS RCS LOOP 2 151- JJS I5 I J le If IRH le IRH el lf essa UIH le jf es2A Pl coda Pl eeU essa leds HA IA SA 151 522

'IRH 15 Jl 8 UIH Cw Cw RHR PuMP I ! rls

" edl IA SA ISI 5NS RCS lOOP I 5 C L.

IRH 2a 151 Jle 151 Jdd IRH J I AI eIH 58 POI Sacs

) IS I - Jl I RCS LOOP 2 US CL. )EA rf If If Pl 151 Jle ess ese8 RHR HK eel ess8 0 M

IRH SS IB 58 e Jls RCS L~ Jl~ ISI JII IXI PIH-ee C~L.

jI I L~J Rtal PUMP RI 58 151 (5) e Jel (Sl IRH ~ el lRH-121 CHC/Sl PUMP

! SuC I ION RCS LOOP I g VVI I g uN Na

~RCS PRI IRH ~5 t Iso M

p".-,~Z JS LRH INSIOI CONIAINULNI Oulelvf CONIAINMINI A(4- )