ML20043G201

LER 90-005-00:on 900510,steam Generator lo-lo Level Reactor Trip Occurred During Cold Shutdown.Caused by Inadequate Warning Sign Posting.Nonconformance Rept Written to Document Event & Recommend evaluation.W/900608 Ltr

ML20043G201
Person / Time
Site:	Point Beach
Issue date:	06/08/1990
From:	Fay C WISCONSIN ELECTRIC POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
CON-NRC-90-084, CON-NRC-90-84 LER-90-005, LER-90-5, VPNPD-90-274, NUDOCS 9006200035
Download: ML20043G201 (7)

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. POWER COMPANY 231 W Michigart PO Bcm 2046. MJwoukee,WI 53201 (414)221-2345 VPNPD 274 ' c CFR 50.73 NRC 084 June 8, 1990

-U. S. NUCLEAR REGULATORY COMMISSION Document Control Desk

Mail Station P1-137

Washington, D.C. 20555 Gentlemen

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DOCKET 50-266

-LICENSEE EVENT REPORT 90-005-00

- STEAM GENERATOR LOW-LOW LEVEL REACTOR TRIP DURING COLD SHUTDOWN POINT BEACH NUCLEAR PLANT, UNIT 1 Enclosed is Licensee' Event Report 90-005-00 for Point Beach Nuclear? Plant, Unit 1. This report details the occurrence of a

' reactor trip' signal while preparing for primary system testing

during-La refueling outage. This report is provided in accordance 5l- with 10.CFR 50.73(a)(2)(iv), "Anyl event or condition that

' /; .resulted in-the manual or automatic actuation of any engineered I safety-system including the reactor protection system."

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/- If any further information is required, please contact us.

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-Very truly yours,

.C..W. ay-

=Vice President Nuclear Power y Enclosure Copies to NRC Regional Administrator, Region III NRC Resident. Inspector

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On May 10, 1990 at 1608, the PBNP Unit 1 Reactor Protection Sytem (RPS) generated a reactor trip signal upon receipt of a 2/3 trip logic for Low-Low Steam Generator Water Level. When the event occurred, Unit 1 was nearing the end of a refueling outage and was in a Cold Shutdown (CSD) condition. The reactor trip breakers were open at the time the trip signal was generated.

PBNP personnel were making preparations for cold rod drop testing.

Dummy test signals had been inserted into 4 of the 6 steam generator narrow range level channels in order to simulate steam generator levels above the low-low level trip setpoint. The dummy test signals were being supplied by a test cart which was powered by a lighting receptacle through an extension cord. A technician working in the area disconnected the extension cord from the cart. With the loss of the test signals, the RPS received low-low steam generator level signals, and generated the trip function.

Because the reactor was in a CSD condition with reactor trip breakers open, at.d the reactor protection system functioned as designed, there were no safety implications. A modification to the test equipment is in progress which will make disconnection errors in this application highly unlikely.

NRC Form 354 (649)

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Taxi,,--. . w. . ecw me.nm BACKGROUND AND EVENT DESCRIPTION on May 10, 1990, Unit 1 was nearing the end of a scheduled refueling outage. Unit 1 was in a cold shutdown (CSD) condition and Unit 2 was at 100% power. Both the A and B Steam Generators for Unit 1 were drained to the 24 inch level which is off-scale low on the narrow range level instrumente ("A"SG - LT-461B, 462A, 463C; "B"SG - LT-471B, 472A, 473C). Steam generator (SG) level is a parameter which is utilized in the PBNP Reactor Protection System (RPS) in order to trip the reactor under two adverse operational conditions: 1) Steam Flow-Feed Flow Mismatch coincident with Low SG 1evel, and 2) Low-Low SG level.

The signals from the narrow range level transmitters cannot, by design, be isolated from the RPS at low power or shutdown conditions.

Therefore, all SG level transmitters were sending an off-scale low analog signal to the RPS. The RPS was, as a result, in a " tripped" condition since the 2 of 3 logic for low-low SG level (setpoint equals 15% of narrow range level) conditions were satisfied. In the shutdown condition, this is of no consequence as the reactor trip breakers are open. However, this " tripped" condition prevents closure of the reactor trip breakers, and thus, prevents energization of the control rod drive mechanisms (CRDMs).

PBNP Technical Services and Operations personnel were making preparations to conduct Primary System Test procedure RESP 3.1 which includes, among other items, cold control rod drop testing under full-flow conditions. This test requires latching and movement of control rods, which in turn requires clearing of any reactor trip conditions and subsequent closure of the reactor trip breakers.

This is accomplished through a step in the " Initial Conditions" section of the procedure which directs Instrumentation and Control (I&C) personnel to " clear any active reactor trips using analog simulators or other means as necessary." I&C personnel accomplished tbc step by inserting " dummy" signals into the test jacks of four (tuo for the "A" SG and two for the "B" SG) of the six SG narrow range level instruments. These signals simulate a mid-range level above the low-low SG level trip setpoint of 15%. In this condition, the RPS sensed 2 of 3 channels from each SG in a " permissable" condi-tion, thus clearing the trip. The signals were generated by four analog simulators mounted on a portable test cart. The test cart was powered through an extension cord by a non-safeguards 120V AC receptacle. A temporary sign was hung on the front of the test cart warning against deenergization.

NRC Form 366A (6496

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At 1608, prior to beginning the test procedure, and with the reactor trip breakers still in the open position, an I&C technician in need of an extension cord unplugged the cord which was supplying power to the test cart. The analog simulators were deenergized and the RPS again sensed the necessary conditions to meet the required 2 of 3 trip logic for low-low SG level. As a result a new reactor trip signal was generated. Upon hearing an audible alarm initiated coincident with his unplugging the extension cord, the technician plugged the cord back in and reported his error to the control room operating personnel. The cause of the trip signal generation was determined, and the reportability of the event was subsequently evaluated.

In accordance with 10CFR50.72, the NRC Operations Center was notified of the RPS actuation at 1715.

PLANT SYSTEM RESPONSES The Unit 1 reactor protection system responded appropriately to a low-low SG 1evel condition. Upon deenergization of the test cart, the RPS received off-scale low signals from each of the three level transmitters associated with each steam generator (one channel off-scale low due to actual level being out of range low, and two channels off-scale low due to the simulated, " dummy" signals becoming deenergized). The 2 of 3 trip logic for either SG was satisfied, and the reactor trip was activated. However, because the reactor trip breakers were already open, there was no effect on physical plant status other than receipt of the reactor trip alarm.

An additional Engineered Safety Feature (ESF) associated with low SG

. level is activation of the Auxiliary Feedwater System. This activa-tion also occurs upon receipt of 2 of 3 narrow range SG level detectors at the low-low level setpoint of 15%. However, in the shutdown condition, this activation is defeated by placing the Main Feed Pump control switches-in " pullout". This is the normal condition for those control switches when shutdown, and was their condition at the time of the event. Therefore, auxiliary feedwater was not actuated.

Unit 2 operation was unaffected by the Unit 1 reactor trip signal.

NRC Form 366A (6496

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010 014 OF O l6 YuT tu -. am. = w. em unc rom. .uwenno SYSTEM DESCRIPTIONS The RPS provides protective action for the reactor by interrupting power to the rod drive power cabinet and DC hold bus through the tripping of either reactor trip breaker A (RTA served by Train A instrumentation and logic) or reactor trip breaker B (RTB served by Train B instrumentation and logic). The opening of either breaker interrupts dedicated 260v 3-phase power from the rod-drive MG sets to the Rod Control Cluster Assembly (RCCA) stationary and moveable gripper coils, causing the control rods to drop.

Two conditions related to steam generator level can cause a reactor trip, 1) Steam Flow-Feed Flow mismatch with coincident SG low level, and 2) Low-Low SG level. The first protective action uses a low SG 1evel setpoint of 20%, while the second utilizes the low-low SG level setpoint of 15%. Both protective measures use the narrow range SG level instruments.

The low-low SG 1evel bistables combine in a 2 of 3 logic for each steam generator to provide a reactor trip signal when either SG level measures less than or equal to 15% of narrow range level. This feature protects against a loss of heat sink. The 2 of 3 low-low logic also provides signals which start the auxiliary feedwater pumps.

CAUSE The cause of this event can be traced to two factors: 1) human error and 2) inadequate warning sign posting.

The I&C technician disconnected the extension cord at the test cart without checking the front of the cart to see if it was in use. Had i

he done so, the posted warning sign would have alerted him to the

! cart's being in use. This was a cognitive error on the part of the

~ technician.

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The second error concerns how the test cart and its power supply were posted. A posting of the extension cord at the test cart and the l receptacle, warning against deenergization, could have alerted the l technician not to disconnect the cord.

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010 015 F 0l6 1sxT cu,,.m am a o. .sewac r 3aw,>.m REPORTABILITY 2his item is reported in accordance with 10CFR50.73(a)(2)(iv) which states, "The licensee shall report . . ,, any event or condition that resulted in a manual or automatic actuation of any engineered safety feature (ESF), including the reactor protection system (RPS)."

CORRECTIVE ACTION A Nonconformance Report (NCR) was written to document the event and to recommend evaluation of preventive measures. The I&C group is addressing the situation from a design standpoint and intends to discontinue the use of the test cart which requires an external power source for this application. A different device will be used which is less prone to accidental deenergization. I&C is developing a small module which will plug into the test jacks for the instrument channel to be simulated. It will be powered by the instrument loop itself, and will require no external (to the racks) interface. The device will be conspicuously placarded to warn against unauthorized removal.

I&C personnel have built a prototype for this device and are in the process of testing it. It is anticipated testing will be completed and the device will be utilized during the fall 1990 Unit 2 refueling outage.

SAFETY ASSESSMENT Because the Reactor Trip Breakers were open at the time of the event, and no control rods were latched, there was no effect on core reactivity or on the status of the plant. Similarly, because the plant was shutdown, Auxiliary Feedwater Initiation did not occur due to the Main Feed Pump controllers being in the pullout position.

Had Primary System Test procedure RESP 3.1 been in progress,.with rods latched, the effect of this event on plant conditions would have been minimal with no safety implications. Initial conditions for RESP 3.1 specify two steps regarding core reactivity: 1) Reactor is shutdown by at least 5,000 pcm with all RCCA's fully inserted and boron concentration is greater than 1800 ppm; and 2) The RCS boron concentration required to assure 1% shutdown at 530 F with all rods out has been verified to be less than 1800 ppm. These checks, along

-with the conservative requirement to not have more than eight rods fully withdrawn at one time, assure control of reactivity below criticality during testing. The consequences of the event occurring

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Ol 0 Ol6 nxTrn = ==., x - ac % w . inn during rod testing would be the removal of reactivity by the dropping of any withdrawn control rods. While the test would be disrupted, there would be no safety implications.

I&C personnel are required to simulate SG level signals during two other tests conducted during refueling outages, testing of AMSAC (ATWS Mitigating System Actuation Circuitry) and Operations Refueling Test (ORT) 443. The consequences of the event occurring during these tests would be similar, and in no way less conservative, than the event occurrence during RESP 3.1. Therefore, there are no safety implications in this regard.

GENERIC IMPLICATIONS There are no known generic implications from this event.

NRC Perm 366A (6-89)

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