ML20056D399

LER 93-029-00:on 930708,partially Withdrawn Control Rod 26-31 Scrammed.Caused by RPS Scram Signal from Two of Four Reactor Pressure Scram Switches.Scram Reset & Pressure Testing completed.W/930806 Ltr

ML20056D399
Person / Time
Site:	Cooper
Issue date:	08/06/1993
From:	Gardner R, Myers J NEBRASKA PUBLIC POWER DISTRICT
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
CNSS933202, LER-93-029, LER-93-29, NUDOCS 9308130125
Download: ML20056D399 (4)

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5' COOPER NUCLI AR ST AllON 9? PO Ef 0 x 98, BROWNVf LLE., NE BR ASKA 683?J x -

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CNSS933202 August 6, 1993 U.S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555

Dear Sir:

Cooper Nuclear Station Licensee Event Report 93-029, Revision 0, is forwarded as an attachment to this letter.

Sincerely, r

i-R. L. Cardner Plant Manager RLG/j u Attachment cc: J. L. Milboan G. R. Horn J. M. Meacham R. E. Wilbur V. L. Wolstenholm D. A. Whitman INPO Records Center NRC Resident Inspector R.J. Singer CNS Training CNS Quality Assurance 130036 9308130125 930806 PDR ADOCK 05000298 fYS 1 -

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LICENSEE EVENT REPORT (LER)

Fl4tLITY hAME H) DUCEET NUMBER (2) P'6E 43' Cooper Nuclear Station o 151010 l 0 l 2 l 9 l 8 1 lOFl0 l3 TITLE tai Spurious RPS Actuation During Performance Of The ASME Class 1-N System 1.eakage Test EVENT DATE (5) LER NUMSER 46) REPORT DATE (71 OTHE R F ACILITIEE INVOLVED (S)

MONTH DAY YEAR YEAR ley h*'  :

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l 0l 7 0l 8 9 3 9l 3 0l2l9 0l0 0l 8 0l6 9l 3 oistogogog OPE uT M THIS REPORT f5 SUBMITTED PUR5 DANT TO THE REQUIREMENTS OF 10 CFR {: (Chece one or mom af the fehowg1011 "N 8 N 20.e2t : 20. mm y so.73c.H2H 73 71te)

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GAME TELEPwONE NUMBER ARE A CODE.

John R. Myers 4;Ol2 8;2 3 5g3 1 1 8 1 1l1 COMPLETE ONf LINE FOR EACH COMPONENT F AILURE DESCRfBED IN THIS REPORT (13P CAUSE SYSTEM CDMPONfNT

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$USMi&5 ION s VES (19 ven. commere EXPECTED SUBnUS90N DA TE) NO l l l u.sTR ACT omir , re m r e.. .-me., , u .,- r,-r- ae nm On July 8, 1993, at 11:42 pm, a Reattor Protection System (RPS) scram signal was generated from two of the four Reactor Pressure-Scram switches, resulting in the scramming of partially withdrawn control Rod 26-31. The Reactor Coolant System was pressurized to perform the ASME Class 1-N System Leakage Test in accordance with ASME Code requirements prior to startup from the 1993 Refueling Outage. In concert with this test, other activities were being performed, including replacement of instrument line excess flow check valves and scram time testing. The common sensing line for the two Reactor Pressure-Scram switches was apparently vibrated during activities associated with replacing an excess flow check valve, resulting in a minor pressure perturbation at the switches. With the Reactor at approximately 1030 psig, only a small margin existed between the Reactor Coolant System pressure and the switches' setpoint.

The scram was reset, and pressure testing completed without further difficulties. Procedures related to ASME Class IN system leakage testing will be revised to require better coordination of activities in the vicinity of the High Pressure-Scram switches and sensing lines which could result in an RPS actuation.

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NRC Faste 306 W. 3)

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U.S. NUCLEAR REGULATORY COMMISSION UCENSEE EVENT REPORT ILER) TEXT CONTINUATION AreRovEo oms No. 3 so-oio.

EXPIRES: 8/3U88 f ACILITV NAME 4:ll DOCF.ET NUMBER (2) LER NUMBER (6) PAGE (3) .

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0l0 0l2 OF 0 l3 TEXT <a-, a-. . w. - .e-w une m m4v nn A. Event Description On July 8,1993, at 11:42 pm, a Reactor Protection System (RPS) scram signal was generated from NBI-PS-55C and D, two of the Reactor Pressure-Scram switches. As a result, Control Rod 26-31, which was partially withdrawn for performing scram timing, automatically scrammed.

The Reactor Coolant System was pressurized to approximately 1030 psig to:

perform the ASME Ciass 1-N System Leakage Test, in accordance with ASME Code requirements, prior to startup from the 1993 Refueling' Outage. In.

concert with the leakage testing, other activities, including replacement of instrument line excess flow check valves and scram time testing, were being performed. In response to the scram, correct plant response was verified and an investigation of the cause was begun,.since the immediate indications did not identify a specific cause for the actuation. Further investigation found that at the time of the actuation replacement of an excess flow check valve was in progress on an instrument line emanating i' rom the penetration which contains the common sensing line for these Reactor Pressure-Scram switches.

B. Plant Status The plant was in cold shutdown, in the 1993 Refueling Outage. The Reactor was pressurized to npproximately 1030 psig, undergoing the ASME.

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Class 1-N System Leakage. test, with a Reactor coolant temperature of approximately 190 degrees Fahrenheit.

C. Basis for Report Actuation of the Reactor Protection System resulting in the scram of a single, partially withdrawn, control rod, report.able in accordance with the criteria prescribed by 10CFR50. 73(a)(2)(iv) .

D. Cause The common sensing line for the two Reactor Pressure-Scram switches was apparently vibrated while completing the activities associated with replacing the excess flow check valve, resulting in a minor pressure perturbation at the switches. With the Reactor at approximately 1030 psig, only a small margin existed between the-Reactor Coolant System pressure and the switches' setpoint. Under these conditions, a minor pressure perturbation would be sufficient to actuate the switches.

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The procedures for performing the pressure testing, excess flow check valve testing and maintenance, and control rod scram timing did not contain guidance to ensure that the activities being conducted would not result in a scram should minor pressure perturbations occur.

E. Safety Sirnificance Reactor Coolant System pressure was being maintained at approximately 1030 psig, below the nominal 1045 psig setpoint of the Reactor Pressure-Scram switches. Under these conditions, the spurious pressure perturbation did .not represent a challenge to the integrity of the Reactor Coolant System.

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F. Safety Implications 6

t The high pressure scram is intended'to prevent exceeding the Technical Specification safety limit of 1337 psig for Reactor Coolant System overpressurization during operational transient conditions. Under transient conditions, a scram at 1045 psig will ensure that the safety limit is not challenged. The. pressure testing which was'being performed is conducted only while the Reactor is shutdown, and thus the pressure '

rise associated with an operational transient condition is not-applicable.

I G. Corrective Action i

The scram was reset, and a reduced pressure maintained while investigating the scram. Subsequently, pressure testing was completed i without further difficulties. Procedures related to ASME Class IN system leakage testing will be revised to require better coordination of activities in the vicinity of the High Pressure-Scram switches and

• sensing lines which could result in an RPS actuation.

H. Similar Events .

LER 87-01 discusses Reactor scrams which occurred during performance of ,

similar testing in 1987. Procedures were revised, however, the >

revisions did not place appropriate restrictions upon the operational '

conditions required to prevent RPS actuations.

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