ML17335A553

LER 99-023-00:on 990907,inadequate TS Surveillance Testing of ESW Pump ESF Response Time Noted.Caused by Inadequate Understanding of Plant Design Basis.Surveillance Tests Will Be Revised & Implemented

ML17335A553
Person / Time
Site:	Cook
Issue date:	10/07/1999
From:	Depuydt M INDIANA MICHIGAN POWER CO.
To:
Shared Package
ML17335A552	List: ML17335A551 ML17335A553
References
LER-99-023, NUDOCS 9910130194
Download: ML17335A553 (7)

Text

NRC FORM 366 U.S. NUCLEAR REGULAT COMMISSION APPROV MB NO. 3150-0104 EXPIRES 06/30/2001 ie-1999)

Estimated burden per response to comply with this mandatory information LICENSEE EVENT REPORT (LER) cogection request: 50 hrs. Reported lessons learned are incorporated into the licensing process and fed back to industry. Forward comments regarding burden estimate to the Records Management Branch (T4 F33), U.S. Nuclear Regutatory (See reverse for required number of Commission. Washington, DC 205554&1, and to the PapenNork Reduction digits/characters, for each block) Project (3t504104), Oirice of Management and Budget. Washington. DC 20503.

lf an information collection does not display a currently valrd OMB control number. the NRC may not conduct or sponsor, and a person is not required to respond to. the information collection.

FACILITYNAME (1) DOCKET NUMBER I2) PAGE I3)

Cook Nuclear Plant Unit 1 05000-315 1OF4 TITLE t4)

Inadequate Technical Specification Surveillance Testing of Essential Service Water Pump Engineered Safety Feature Response Time EVENT DATE {5) LER NUMBER {6) REPORT DATE (7) OTHER FACILITIES INVOLVED (8)

MONTH DAY YEAR YEAR SEQUENTIAL REVISION MONTH DAY YEAR FACILITY NAME DOCKET NUMBER NUMBER NUMBER 1999 FACIUIYNAME DOCKET NUMBER OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR 5: (Check one or moro) {11)

MODE (9) 20.2201 (b) 20.2203(a) (2)(v) 50.73(a)(2)(i) 50.73(a) (2) {viii)

POWER 20.2203(a) (1) 20.2203(a)(3)(I) 50.73(a)(2)(ii) 50.73(a) (2)(x)

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) Specify in Abstract below or In NRC Form 366A 20.2203(a) {2)(iv) 50.36(c)(2) 50.73(a) (2)(vii)

LICENSEE CONTACT FOR THIS LER {12)

NAME TELEPHONE NUMBER IInrSude Area Code)

Mary Beth Depuydt, Regulatory Compliance (616) 465-5901 X1 589 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13) cAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE To EPIX To EPIX SUPPLEMENTAL REPORT EXPECTED 14 EXPECTED MONTH DAY YEAR YEs NO

{Ifyes, complete EXPECTED SUBMISSION DATE).

ABSTRACT (Limit to 1400 spaces, i.o., approximately 15 single-spaced typewritten linos) (16)

On June 24, 1999, during a review of the findings of Expanded System Readiness Review (ESRR) team by Performance Assurance (PA), it was discovered that no testing program could be identified which verifies the capability of the Essential Service Water (ESW) pumps to meet the Engineered Safety Feature (ESF) response time speciTied in the Technical Specifications (TS) or the Updated Safety Analysis Report. Subsequent investigation confirmed that in-place TS surveillance testing measured the ESF response time for the ESW pumps as the elapsed time from actuation of the channel sensor until pump breaker closure, but did not include the time until a specified pump discharge pressure is reached or until the ESW pump discharge valve is open; as required by the definition of Engineered Safety Feature Response Time.

Since existing surveillance testing did not satisfy the TS definition of ESF response time, the identified condition constitutes a missed surveillance test. This is an operation or condition prohibited by TS and was determined to be reportable pursuant to the requirements of 10CFR50.73(a)(2)(i)(B) on September 7, 1999. The apparent cause of this event was the inadequate understanding of the plant design basis. Surveillance tests will be revised and implemented to include the time to achieve prescribed pump discharge pressure/flow and/or discharge valve position as part of the overall ESF response time testing for the ESW system prior to restart of each respective unit. The ESW ESF response times in UFSAR Table 7.2-7 will be evaluated and revised, if necessary, prior to restart of each respective unit. ESW system performance records and surveillance test results provide reasonable assurance that the system has remained capable of performing its interlded functions. Therefore, there were minimal safety implications to the health and safety of the public as a result of this event.

'I)9i0i30i94 99i007 PDR ADGCI{l 050003i5 S PDR

NRC FORM 366A U.S. NUCLEAR RE TORY COMMISSION l+,16.1998)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME I1) DocKET I2) LER NUMBER (6) PAGE I3)

Cook Nuclear Plant Unit 1 05000-315 "EAR SEQUENTIAL NUMBER REVISION 2 QF 4 NUMBER 1999 023 00 TEXT llfmore spaceis required, use additional copies of NRC Form 366A J )17)

CONDITIONS PRIOR TO EVENT Unit 1 was defueled Unit 2 was defueled DESCRIPTION OF THE EVENT On June 24, 1999, during a review of the findings of Expanded System Readiness Review (ESRR) team by Performance Assurance (PA), it was documented that no testing program could be identified which verifies the capability of the Essential Service Water (ESW) pumps to meet the Engineered Safety Feature (ESF) response time specified in the Technical Specifications or the Updated Safety Analysis Report. Subsequent investigation of this condition by Engineering, completed September 1, 1999, confirmed that the acceptance criteria for in-place Technical Specification surveillance testing defined the ESF response time for the ESW pumps as the elapsed time from actuation of the channel sensor until pump breaker closure. Testing did not include the time until a specified pump discharge pressure is reached or the ESW pump discharge valve is open, as required by the definition of Engineered Safety Feature Response Time. The Technical Specification (TS) and UFSAR definition of Engineered Safety Feature Response Time. is that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor to until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Since existing surveillance testing did not satisfy the TS definition of ESF response time, the identified condition constitutes a missed surveillance test.

CAUSE OF THE EVENT The apparent cause of this event was inadequate understanding of the design basis of the plant. During the development of the ESW ESF response times, the design basis requirements for ESW availability during an accident were inadequately understood. This resulted in surveillance procedures for ESW which did not satisfy the UFSAR and Technical Specification definition of ESF response time.

ANALYSIS OF THE EVENT The Technical Specification (TS) and UFSAR definition of Engineered'Safety Feature Response Time is that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor to until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Since existing surveillance testing did not satisfy the TS definition of ESF response time, the identified condition constitutes a missed surveillance test. This is an operation or condition prohibited by TS and was determined to be reportable pursuant to the requirements of 10CFR50.73(a)(2)(i)(B) on September 7, 1999.

Response times for Engineered Safety Features are provided in the UFSAR, Section 7.2, Table 7.2-7. The ESF Response Time Basis Procedure specifies the strategy used at Cook Nuclear Plant to demonstrate the operability of various Engineered Safety Features, systems and sub-systems. This procedure defines "Device Response Time as the time from Safeguards Master Relay closing until the component reaches its ESF position. Additionally, "ESF Response Time" is defined as the time interval from when the monitored parameter exceeds its ESF actuation se'tpoint at the channel sensor until the ESF equipment is capable of performing it's safety function.

Technical Specifications Surveillance Requirements for ESF response time in Section 4.3.2.1.3 and Table 3.3-3 specify that each Engineered Safety. Feature Actuation Signal (ESFAS) function will be demonstrated to be within limits at least once per 18 moriths.

Review of Emergency Diesel Generator Load Sequencing and ESF Testing revealed the ESF response time for the ESW NRC FORM 366A I6-1998)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION i6-1999)

LICENSEE EVENT REPORT tLER)

TEXT CONTINUATION FACIUTY NAME I1) DOCKET {2) LER NUMBER I6) PAGE I3)

Cook Nuclear Plant Unit 1 05000-315 YEAR SEQUENTIAL NUMBER REYIsI0N 3 OF 4 NUMBER 1999 023 00 TEXT (Ifmore spaceis required, use additional copies of NRC Form 366AJ {17) pumps is measured from the initiating sensor channel to the pump breaker closure. Testing does not include the time for the pump to reach the required discharge pressure or for the ESW pump discharge valve to open.

In early 1975, operational problems identified with the ESW system, including severe water hammer at pump start-up, lead to testing being performed under various operational transients. This testing did not result in a significant water hammer, however, a previous test and operating experience showed that the water hammer did not occur when'an idle pump was started with a throttled discharge valve even though its header had not been pressurized for as long as twelve hours. Determination was made that the water hammers were induced upon the start of an idle ESW pump with a fully open discharge valve even when the header had been depressurized for no more than a few minutes. This determination lead to modification of the design of the ESW pump discharge valves, such that the valves remain closed when the ESW pump is idle, and are interlocked to open on ESW pump start at breaker closure.

Response times for sensor actuation to ESW pump breaker closure and ESW pump discharge MOV stroke times are measured under the Surveillance Test Program. However, these times are not combined to provide an overall ESF response time which meets the TS definition and which is compared to an acceptance criteria. The ESF response time test procedure was reviewed to verify that ESF pumps other than ESW are tested from pump start to required system pressure/flow. Each was verified to include requirements to measure the overall response time from sensor actuation until an acceptable discharge pressure or flow prescribed by acceptance criteria.

Although the ESW ESF response times are included in UFSAR Table 7.2-7, ESW response times are not explicitly included in the UFSAR Chapter 14.0 accident analysis assumptions. ESW is not immediately required to support the containment spray system (CTS) and Emergency Diesel Generator during a design basis Loss of Coolant Accident (LOCA). ESW system performance records and surveillance test results provide reasonable assurance that the system has remained capable of performing its intended functions. Based upon the above information, there were minimal safety implications to the health and safety of the public as a result of this event.

CORRECTIVE ACTIONS Surveillance tests will be revised and implemented to include the time to achieve prescribed pump discharge pressure/flow and/or discharge valve position as part of the overall ESF response time testing for the ESW system prior to restart of each respective unit.

The ESW ESF response times in UFSAR Table 7.2-7 will be evaluated and revised, if necessary, prior to restart of each respective unit.

As discussed in letter AEP:NRC:1260GH, "Donald C. Cook Nuclear Power Plant, Units 1 and 2, Enforcement Actions98-150, 98-151,98-152 and 98-156, Reply To Notice Of Violation Dated October 13, 1998," dated March 19, 1999, a surveillance program owner and manager position has been established, reporting to the Work Control Director. A Leadership Plan has been developed which includes the creation of a detailed surveillance data base to align surveillance requirements to specific implementing procedures and a comprehensive adequacy review of surveillance testing procedures.

As previously discussed in LER 315/99-021-00 and as part of Restart Action Plan 0 0001 for the Programmatic Breakdown in Surveillance Testing, the adequacy of the TS surveillance program will be evaluated. This evaluation includes verification that TS surveillance requirements for all modes of plant operation are incorporated into TS surveillance test procedures. Also, as part of the Restart effort, System and programmatic assessments in the Expanded System Readiness Reviews and Licensing Basis Reviews are reestablishing and documenting the plant's Design and Licensing Basis.

NRC FORM 366A i6-1999)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION I6-1998)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME I1) DOCKET I2) LER NUMBER I6) PAGE I3)

Cook Nuclear'lant Unit 1 05000-315 YEAR SEQUENTIAL NUMBER REVISION 4 OF 4 NUMBER 1999 023 00 TEXT (lfmore spaco is required, uso additional copies of NRC Form 386A/ I 17)

SIMILAR EVENTS 315/99-010-00 315/99-015-00 315/99-016-00 315/99-021-00 NRC FORM 366A I6.1998)

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