ML17264B127

LER 97-005-00:on 971031,undetected Unblocking of SI Actuation Signal Occurred at Low Pressure Condition,Due to Faulty Bistable Which Resulted in Inadvertent SI Actuation Signal.Sias,Ci & CVI Signals Were Reset

ML17264B127
Person / Time
Site:	Ginna
Issue date:	12/01/1997
From:	Martin J ROCHESTER GAS & ELECTRIC CORP.
To:
Shared Package
ML17264B126	List: ML17264B125 ML17264B127
References
LER-97-005, LER-97-5, NUDOCS 9712050136
Download: ML17264B127 (9)

Text

NRC FORM 366 U.S. CLEAR REGULATORY COMMISSIO I'PROVED BY OMB NO. 3150-0104 (4-95) EXPIRES 04/30/96 ESTIMATED BURDEN PER RESPONSE To COMPLY WITH THIS MANDATORY INFORMATION COLLECTION REQUEST: 50.0 HRS.

LICENSEE EVENT REPORT (LER) REPORTED LESSONS LEARNED ARE INCORPORATED INTO THE LICENSING PROCESS AND FED BACK To INDUSTRY. FORWARD COMMENTS REGARDING BURDEN ESTIMATE To. THE (See reverse for required number of INFORMATION AND RECORDS MANAGEMENT BRANCH IT-6 F33),

digits/characters for each block) U.s. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND To THE PAPERWORK REDUCTION PROJECT FACILITYNAME (1) DOCKEf NUMBER(2I PAGE (3I R.E. Ginna Nuclear Power Plant 05000244 1 OF 7 TITLE (4I Undetected Unblocking of Safety Injection Actuation Signal While at Low Pressure Condition, Due to Faulty Bistable, Resulted in Inadvertent Safety Injection Actuation Signal EVENT DATE (5) LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (6)

SEQUENTIAL FACILITY NAME DOCKET NUMBER REVISION MONTH DAY YEAR NUMBER MONTH DAY YEAR NUMBER FACIL(ry NAME DOCKET kUMSER 10 31 97 97 005 00 12 01 97 OPERATING THIS REPORT IS SUBMITTED PUR SUANT TO THE REQUIREMENTS OF 10 CFR E: (Check one or more) (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) 000 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) X 50.73(a)(2)(iv) OTHER 20.2203(a) (2)(nr) 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 NUM8ER (Incrude Area Code)

John T. St. Martin - Technical Assistant (716) 771-3641 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DES CRIBED IN'THIS REPORT (13)

SYSTEM COMPONENT REPORTABLE REPORTABLE CAUSE MANUFACTURER To NPRDS CAUSE SYSTEM COMPONENT MANUFACTURER TO NPRDS JE JS F180 NO SUPPLEMENTAL REPORT EXPECTED (14) MONTH DAY YEAR EXPECTED YES SUBMISSION (If yes, complete EXPECTED SUBMISSIONDATE). X NO DATE (15)

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

On October 31, 1997, at approximately 1640 EST, the plant was in Mode 6 with the reactor coolant system being maintained at a temperature of 80 degrees F, and the reactor cavity filled to greater than 23 feet. The plant was shut down for refueling. An inadvertent automatic Safety Injection Actuation occurred.

Immediate corrective action was to monitor the automatic start of engineered safety features components and secure unneeded equipment.

The underlying cause of the inadvertent automatic Safety Injection Actuation was an undetected faulty bistable circuit board. During the performance of a periodic test, this faulty bistable caused the Safety Injection Actuation signaI to be inadvertently unblocked with pressurizer pressure less than the setpoint for Safety Injection Actuation.

Corrective action to preclude repetition is outlined in Section V.B.

97i2050i36 97i20i PDR ADQCK 05000244 S PDR NRC FORM 366 (4-95)

NRC FORM 366A U.s. NUCLEAR REGULATORY COMMISSION (4-95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1) DOCKET LER NUMBER (6) PAGE (3)

YFAR SEQUENTIAL REVISION NUMBER NUMBER 2 OF 7 R.E. Ginna Nuclear Power Plant 05000244 97 005 00 TEXT Iifmore speceis required, use edditionel copies of NRC Form 366AI (17)

PRE-EVENT PLANT CONDITIONS:

On October 31, 1997, the plant was in Mode 6 with the reactor coolant system (RCS) being maintained at a temperature of approximately 80 degrees F, and the reactor cavity filled to greater than 23 feet. Both residual heat removal (RHR) pumps were operating, and fuel movement was on hold. Two service water (SW) pumps were operating. The plant was shut down for refueling. Performance Monitoring technicians were conducting periodic test procedure PT-32.1 (Plant Safeguard Logic Test A or B Train). Testing of the "B" train was started at approximately 1606 EST.

As part of the conduct of procedure PT-32.1, simulated signals are inserted to provide normal relay states for miscellaneous signals. Prior to inserting these signals, individuals from Performance Monitoring and Instrument and Control (ILC) verified that indications were appropriate on the Main Control Board, in safeguards racks in the Relay Room, and in protective racks in the Control Room. IRC technicians then performed these simulations in accordance with procedure PT-32.1. IRC connected non-powered transmitter simulators to three pressurizer (PRZR) pressure channel test injection jacks (for channels P-429, P-430 and P-431). Test injection switches for P-429 and P-430 were taken to the test position and simulator output was checked for the desired output.

DESCRIPTION OF EVENT:

A. DATES AND APPROXIMATE TIMES OF MAJOR OCCURRENCES'ctober 31, 1997, 1606 EST: Performance Monitoring starts procedure PT-32.1 for "B" train.

October 31, 1997, 1640 EST: Event date and time.

October 31, 1997, 1640 EST: Discovery date and time.

October 31, 1997, 1643 EST: Safety Injection Actuation and Containment Isolation signals are reset.

October 31, 1997, 1649 EST: Containment Ventilation Isolation signal is reset.

October 31, 1997, 1700 EST: Pre-event refueling shutdown conditions are restored.

B. EVENT:

On October 31, 1997, the plant was in Mode 6 with the RCS being maintained at a temperature of approximately 80 degrees F, and the reactor cavity filled to greater than 23 feet. The plant was shut down for refueling.

The Safety Injection (Sl) Actuation Signal (SIAS) is, provided with a block signal which prevents a SIAS from occurring for Low PRZR Pressure. SIAS was blocked, which is the normal configuration when PRZR pressure is intentionally reduced below the SIAS setpoint during plant shutdown conditions.

NRC FORM 366A (4.95)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION I4-95)

LICENSEE EVENT REPORT (LER)

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

YEAR SEQUENTIAL REVISION NUMBER NUMBER 3 OF 7 R.E. Ginna Nuclear Power Plant 05000244 97 005 00 TEXT (Ifmore space is required, use additional copies of NRC Form 366A/ (17)

At approximately 1640 EST, I&C technicians took the third channel (P-431) test injection switch to the test position. Immediately after the switch was in the test position, simulator output was observed to be approximately 30 milliamperes, which simulates pressure above the unblock setpoint of 1992 PSIG. This resulted in unblocking SIAS from Low PRZR Pressure. This also resulted in Sl actuation on Low PRZR Pressure.

The Control Room operators immediately responded to the inadvertent Sl actuation. They responded to Main Control Board annunciator D-19 (Pressurizer Lo Press Sl 1750 PSIG). They confirmed that automatic Sl actuation had occurred and verified that all operable engineered safety features (ESF) components functioned properly. No immediate actions were required for the RCS, since the reactor cavity level remained stable. The Control Room operators referred to emergency operating procedures E-0 (Reactor Trip or Safety Injection) and ES-1.1 (Sl Termination) for guidance in securing and restarting equipment.

Both RHR pumps continued to operate, and the two selected SW pumps started. The "A" emergency diesel generator (EDG) started and did not energize any busses, since all safeguards busses remained energized from off-site power. All other operable ESF components were also observed to function properly, with the exception of valve position for air-operated valve AOV-5392 (instrument air to containment isolation valve). The valve position indicated both open and closed. However, subsequent investigation confirmed that the valve did, in fact, travel to the fully closed position.

Due to plant conditions, many ESF components were not operable at the start of this event. The "B" EDG was inoperable for periodic manufacturer inspection and overhaul. All three SI pumps were rendered inoperable. Therefore, the "B" EDG did not start and no Sl pumps started. No injection flow from the Sl pumps to the RCS occurred.

I&C removed the safeguards train DC supply fuses to prevent a re-occurrence of Sl actuation. I&C then commenced trouble-shooting for the cause of the Sl actuation.

Subsequent root cause analysis revealed that, prior to this activity, there had been a fault in bistable circuit board PC-430 E/F. This undetected fault resulted in the bistable de-energizing. A de-energized output results in both a 1/3 Unblock Sl signal and a 1/3 Low PRZR Pressure Sl signal.

The SIAS for Low PRZR Pressure was automatically unblocked as soon as a second channel (the P-431 simulated signal) was simulated above the unblock setpoint, since a first channel (failed channel P-430) was already inserting an unblock signal. Simulated PRZR pressure (as measured by channel P-429) was below the setpoint for automatic SIAS (thus inserting a 1/3 Low PRZR Pressure Sl signal). The faulted channel P-430 was also inserting a 1/3 Low PRZR Pressure Sl signal. This 2/3 Sl signal was blocked until the P-431 simulator output was inserted. At that instant, the 2/3 signal was unblocked, resulting in automatic SI actuation from 2/3 Logic for PRZR pressure less than 1750 PSIG. Automatic Sl actuation caused automatic actuation of CNMT Isolation (Cl) and CNMT Ventilation Isolation (CVI).

NRC FORM 366A (4.95)

I NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION I4.95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME I1) DOCKET LER NUMBER IB) PAGE I3)

YEAR SEQUENTIAL REVISION NUMBER NUMBER 4 OF 7 R.E. Ginna Nuclear Power Plant 05000244 97 005 00 TEXT llfmore spaceis required, use additional copies of IVRC Form 366A/ (17)

INOPERABLE STRUCTURES, COMPONENTS, OR SYSTEMS THAT CONTRIBUTED TO THE EVENT:

Trouble-shooting by l&C identified that bistable PC-430 E/F had a blown fuse, which caused its outputs to be de-energized. No output from this bistable (de-energized state) provides a 1/3 Unblock Sl and 1/3 Low PRZR Pressure Sl signal.

D. OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED:

None E. METHOD OF DISCOVERY:

This event was immediately apparent due to numerous Main Control Board Annunciator alarms in the Control Room.

F. OPERATOR ACTION:

The Control Room operators responded to the Annunciator alarms. They diagnosed that automatic Sl actuation had occurred, and verified that all operable ESF components functioned properly.

They observed the dual position indication for AOV-5392 and confirmed that Instrument Air to CNMT had been properly isolated.

The Control Room operators reset the SIAS and Cl and CVI signals. Unneeded equipment was secured. All Cl valves and CVI components (which changed position due to the SI actuation) were returned to their positions prior to the event. Pre-event conditions were restored.

Subsequently, the Control Room operators notified higher supervision and the NRC. The Shift Supervisor notified the NRC per 10CFR50.72 (b) (2) (ii), non-emergency four hour notification, at approximately 1946 EST on October 31, 1997.

G. SAFETY SYSTEM RESPONSES:

All safety systems and components that were operable responded as designed, except that the valve position indication for AOV-5392 indicated both open and closed. All other operable ESF components were observed to function properly after the automatic Sl actuation. This included autostart of two SW pumps and the "A" EDG.

NRC FORM 366A I4.95)

NRC I'ORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4.95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1) DOCKET LER NUMBER (6) PAGE (3)

YEAR SEQUENTIAL REVISION NUMBER NUMBER 5 OF 7 R.E. Ginna Nuclear Power Plant 05000244 97 005 00 TEXT Iffmore spaceis required, use eddidonal copies of NRC Form 386A/ (17)

III. CAUSE OF EVENT:

A. IMMEDIATECAUSE:

The immediate cause of the inadvertent automatic Sl actuation was from 2/3 logic from Low PRZR Pressure with SIAS unblocked.

B. INTERMEDIATE CAUSE:

The intermediate cause of automatic SIAS being unblocked was simulating the signal for'PRZR channel P-431 Sl Unblock to be above the unblock setpoint with a concurrent undetected fault in channel P-430.

ROOT CAUSE:

The underlying cause was faulty bistable circuit board PC-430 E/F on channel P-430, which functions to provide 1/3 Unblock Sl signals and 1/3 Low PRZR Pressure Sl signals. In both cases, the bistable de'-energizes to provide this function. The bistable's supply fuse had blown and caused both signal outputs to go to zero. This caused the safety system to see a 1/3 signal for unblocking and for initiating Sl. When channel P-431 was simulated, the output was greater than the unblock setpoint, so SIAS was unblocked from both P-430 (faulted) and P-431 (simulated).

This event is NUREG-1022 Cause Code (B), "Design, Manufacturing, Construction/Installation".

IV. ANALYSIS OF EVENT:

This event is reportable in accordance with 10 CFR 50.73, Licensee Event Report System, item (a) (2) (iv),

which requires a report of, "Any event or condition that resulted in a manual or automatic actuation of any engineered safety feature (ESF)". The inadvertent automatic SI actuation is an automatic actuation of an ESF.

An assessment was performed considering both the safety consequences and implications of this event with the following results and conclusions:

There were no operational or safety consequences or implications attributed to the inadvertent automatic Sl actuation because:

~ The plant was in Mode 6 (refueling shutdown mode) with high head Sl pumps rendered inoperable. The RHR system was aligned for decay heat removal, and steam generator (SG) nozzle dams were in place.

NRC FORM 366A (4-95)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4-95)

LICENSEE EVENT REPORT (LER)

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

YEAR SEQUENTIAL REVISION NUMBER NUMBER 6 OF 7 R.E. Ginna Nuclear Power Plant 05000244 97 005 00 TEXT (Ifmore spaceis required, use additional copies of NRC Form 366A/ (17)

~ Plant conditions precluded any over-pressure condition. The head was removed from the reactor and the reactor cavity was filled to greater than 23 feet. All Sl pumps were inoperable and the Refueling Water Storage Tank was not aligned to the suction of the RHR pumps. There was no addition of water inventory to the RCS. RCS temperature continued to be maintained stable via the RHR system. Therefore, reactivity was not affected by RCS temperature changes.

~ The SW system and Component Cooling Water system remained in service during this event, providing adequate cooling to the RCS and Spent Fuel Pool.

~ The RHR cooling capability remained in service and RCS integrity was maintained.

Therefore, heat removal from the reactor was assured.

Based on the above, it can be concluded that the public's health and safety was assured at all times.

V. CORRECTIVE ACTION:

A. ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL STATUS:

IRC decreased the simulated output for P-431 to below the Sl unblock setpoint, thus terminating the SI Unblock logic, and allowing the Control Room operators to again block SIAS.

The SIAS, Cl, and CVI signals were reset. Unneeded equipment was secured.

All Cl valves and CVI components (which changed position due to the Sl actuation) were returned to their positions prior to the event. Pre-event conditions were restored.

B. ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE:

A method to visually indicate Sl Unblock status from PRZR pressure will be evaluated.

The faulty bistable circuit board was replaced with a like for like replacement, and calibrated and tested satisfactorily.

CPI-TRIP TEST-5.20 (Reactor Protection System Bistable Trip Test/Calibration for Channel 2 (White) Bistable Alarms) was completed for all Channel 2, Rack 1 bistables to check for proper operation and to verify there were no other bistables with similar faults. No other faulty bistables were identified.

NRC FORM 366A I4 95)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4.95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1) DOCKET LER NUMBER (6) PAGE (3)

YEAR SEQUENTIAL REVISION NUMBER NUMBER 7 OF 7 R.E. Ginna Nuclear Power Plant 05000244 97 005 00 TEXT llfmore space is required, use additional copies of NRC Form 366Al (17)

VI. ADDITIONALINFORMATION:

A. FAILED COMPONENTS:

None PREVIOUS LERs ON SIMILAR EVENTS' similar LER event historical search was conducted with the following results: no documentation of similar LER events with the same root cause at Ginna Station could be identified. However, the following LERs were similar events with different root causes:

~ LER 84-006

~ LER 85-004

~ LER 89-003

~ LER 95-003 C. SPECIAL COMMENTS'one NRC FORM 366A (4 95)