ML18100B061

LER 94-007-00:on 940407,automatic Reactor Trip Occurred, Immediately Followed by ECCS Si.Caused by Personnel Error. Operating Procedures Revised & Simulator Training Conducted W/All Operating shifts.W/940506 Ltr

ML18100B061
Person / Time
Site:	Salem
Issue date:	05/06/1994
From:	Hagan J, Pastva M Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-94-007, LER-94-7, NUDOCS 9405130189
Download: ML18100B061 (11)

Text

Public Service Electric and Gas Company P.O. Box 236

• Hancocks Bridge, New Jersey 08038 Salem Generating Station May 6, 1994 U. s. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555

Dear Sir:

SALEM GENERATING STATION LICENSE NO. DPR-70 DOCKET NO. 50-272 UNIT NO. 1 LICENSEE EVENT REPORT 94-007-00 This Licensee Event Report is being submitted pursuant to the requirements of Code of Federal Regulation 10CFR50.73(a) (2) (iv).

In addition, this report fulfills the requirement for a Special Report pursuant to Technical Specification 3.5.2. Issuance of this report is required within thirty (30) days of event discovery.

Sincerely yours, J. J. H an General Manager -

Salem Operations MJPJ:pc Distribution

~ ') (!* t~ 1 :-,

9405130189 940506 PDR ADOCK 05000272 s PDR The power is in your hands.

95-2169 REV 7-92

I NRC FORM 366 .S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 (5-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD LICENSEE EVENT REPORT (LER) COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF (See reverse for required number of digits/characters for each block) MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2) PAGE (3)

Salem Generatin~ Station - Unit 1 05000 272 1 OF09 TITLE (4) Reactor Trip From 25% Power/Two Safety Injections, Manually Initiated Main Steam Isolation- And Discretionary Declaration Of ALERT.

EVENT DATE (5) LER NUMBER (6 REPORT NUMBER (7) OTHER FACILITIES INVOLVED lB FACILITY NAME DOCKET NUMBER SEQUENTIAL REVISION MONTH DAY YEAR YEAR MONTH DAY YEAR NUMBER NUMBER 05000 FACILITY NAME DOCKET NUMBER

-- -- 05000 04 07 94 94 007 00 05 06 94 OPERATING THIS REPORT IS SUBMITIED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check one or more (11 MODE (9) 1 20.402(b) 20.405(c) x 50.73(a)(2)(iv) 73.71(b)

POWER 20.405(a) (1) (i) 50.36(c)(1) 50.73(a)(2)(v) 73.71 (c)

LEVEL (10) 073 20.405(a)(1)(ii) 50.36(c)(2) 50.73(a) (2) (vii) x OTHER (Specify in Abstract 20.405(a) (1) (iii) x 50. 73 (a)(2)(i) 50.73(a) (2) (viii) (A)

I-below and in Text, NRC 20.405(a)(1)(iv) 50.73(a)(2)(ii) 50.73(a) (2) (viii) (B) Form 366A) 20.405(a)(1 )(v) 50.73(a)(2)(iii) 50.73(a)(2)(x) Special Rep LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMBER (Include Area Code)

M. J. Pastva. Jr. - LER.Coordinator (609) 339-5165.

COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

REPORTABLE REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER CAUSE SYSTEM COMPONENT MANUFACTURER TO NPRDS TO NPRDS SUPPLEMENTAL REPORT EXPECTED (14 EXPECTED MONTH DAY YEAR I YES (If yes, complete EXPECTED SUBMISSION DATE) x NO SUBMISSION DATE (15)

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

At 1050 hours0.0122 days <br />0.292 hours <br />0.00174 weeks <br />3.99525e-4 months <br /> on 4/7/94, an automatic Reactor trip occurred, was immediately followed by an Emergency Core Cooling System {ECCS) Safety Injection {SI) and, at 1100 hours0.0127 days <br />0.306 hours <br />0.00182 weeks <br />4.1855e-4 months <br /> an Unusual Event was declared. At 1105 hours0.0128 days <br />0.307 hours <br />0.00183 weeks <br />4.204525e-4 months <br />,* the SI signal was reset and ECCS flow reduction began.

Reactor Coolant System temperature increased, Pressurizer level increased to >100%, steam generator pressure increased and main steam safety valves lifted, and at 1128 hours0.0131 days <br />0.313 hours <br />0.00187 weeks <br />4.29204e-4 months <br />, a second automatic SI occurred. At 1316 hours0.0152 days <br />0.366 hours <br />0.00218 weeks <br />5.00738e-4 months <br />, a precautionary ALERT was declared. HOT SHUTDOWN was achieved at 0106 hours0.00123 days <br />0.0294 hours <br />1.752645e-4 weeks <br />4.0333e-5 months <br /> on 4/8/94, and at 1124 hours0.013 days <br />0.312 hours <br />0.00186 weeks <br />4.27682e-4 months <br /> (same day), COLD SHUTDOWN was achieved. The trip resulted from assigning inappropriate priority of actions and improperly monitoring reactor power while withdrawing rods. The first SI resulted from inadequate control Of primary loop temperature, concurrent with a false high steam flow signal. The second SI resulted from low Pressurizer pressure due to lifting a steam generator safety valve. Involved personnel have completed remedial training and evaluation. Operating procedures have been revised, as appropriate. Component testing, repairs, and modifications have been made, as required.

NRG FORM 366 (5-92)

• REQUIRED NUMBER OF DIGITS/CHARACTERS FOR EACH BLOCK

• BLOCK NUMBER OF TITLE NUMBER DIGITS/CHARACTERS 1 UP TO 46 FACILITY NAME 8 TOTAL 2 DOCKET NUMBER 3 IN ADDITION TO 05000 3 VARIES PAGE NUMBER 4 UP TO 76 TITLE 6 TOTAL 5 EVENT DATE 2 PER BLOCK 7 TOTAL 2 FOR YEAR 6 LER NUMBER 3 FOR SEQUENTIAL NUMBER 2 FOR REVISION NUMBER 6 TOTAL 7 REPORT DATE 2 PER BLOCK UP TO 18 -- FACILITY NAME 8 OTHER FACILITIES INVOLVED 8 TOTAL- DOCKET NUMBER 3 IN ADDITION TO 05000 9 1 OPERATING MODE 10 3 POWER LEVEL 1

11 REQUIREMENTS OF 1O CFR CHECK BOX THAT APPLIES UP TO 50 FOR NAME 12 LICENSEE CONTACT 14 FOR TELEPHONE CAUSE VARIES 2 FOR SYSTEM 13 4 FOR COMPONENT EACH COMPONENT FAILURE 4 FOR MANUFACTURER NPRDS VARIES 1

14 SUPPLEMENTAL REPORT EXPECTED CHECK BOX THAT APPLIES 6 TOTAL 15 EXPECTED SUBMISSION DATE 2 PER BLOCK

• Salem Generating Station DOCKET NUMBER

• LICENSEE EVENT REPORT (LER) TEXT CONTINUATION LER NUMBER PAGE Unit 1 5000272 94-007-00 2 of 9 PLANT AND SYSTEM IDENTIFICATION:

Westinghouse - Pressurized Water.Reactor Energy Industry Identification System (EIIS) codes are identified in the text as {xx}

IDENTIFICATION OF OCCURRENCE:

Reactor Trip From 25% Power/Two Safety Injections, Manually Initiated Main Steam Isolation, And Discretionary Declaration Of ALERT Event Date: 4/7/94 Report Date: 5/6/94 This report was initiated by Incident Report No.94-102.

CONDITIONS PRIOR TO OCCURRENCE:

Mode 1 Reactor Power 73% - Unit Load 800 MWe T at 562 degrees Fahrenheit (F). Control Rods in manual control w~~fi Bank D rods at 195 steps.

The Unit was at reduced power due to seasonal problems with excessive Delaware River marsh grass/debris affecting the Circulating Water (CW)

{UA} intake structure. The amount of grass/debris loading in the river in was excess of four times the seasonal average recorded over a 17 year period.

Operational challenges were being encountered maintaining the CW circulators {UA} and traveling screens in service. Between 1016 and 1043 hours0.0121 days <br />0.29 hours <br />0.00172 weeks <br />3.968615e-4 months <br /> on April 7, 1994, a load reduction was in progress to take the Main Turbine {TA} off-line following "emergency" tripping of 13A and 13B CW traveling screens and subsequent trips of llA, llB, and 12A circulators. Reactor power had been reduced to 7% with Unit load at 80 MWe. llA and 12B circulators were in service prior to the trip.

In response to decreasing Tave' at approximately 1049 hours0.0121 days <br />0.291 hours <br />0.00173 weeks <br />3.991445e-4 months <br /> (same day) control rods were being manually withdrawn to increase Reactor Coolant System (RCS) {AB} temperature.

DESCRIPTION OF OCCURRENCE:

During rod withdrawal to restore Reactor Coolant System (RCS) temperature, Reactor power increased to 25% and, at 1050 hours0.0122 days <br />0.292 hours <br />0.00174 weeks <br />3.99525e-4 months <br />, on April 7, 1994, an automatic Reactor Protection System (RPS) {JC} trip occurred. This was immediately followed by an Emergency Core Cooling System {BQ} Safety Injection (SI), (Train A) and, at 1100 hours0.0127 days <br />0.306 hours <br />0.00182 weeks <br />4.1855e-4 months <br />, an Unusual Event (UE) was declared. Following the reactor trip/safety

• Salem Generating Station DOCKET NUMBER

• LICENSEE EVENT REPORT (LER) TEXT CONTINUATION LER NUMBER PAGE Unit 1 5000272 94-007-00 3 of 9 DESCRIPTION OF OCCURRENCE: (cont'd) injection, the Main Steam isolation valves were closed due to the primary plant temperature decrease below 547 degrees F. The RCS temperature started to increase at-this time.

At 1105 hours0.0128 days <br />0.307 hours <br />0.00183 weeks <br />4.204525e-4 months <br />, the SI signal was reset on Train A. The ECCS pumps were secured and normal charging was placed in service. Pressurizer level increased to greater than 100% indication (solid condition) and pressure increased due to the SI charging flow and increasing RCS temperature. At 2335 pounds per square inch gauge (psig), the Pressurizer power operated*relief valves (PORVs) {AB} cycled automatically. Steam Generator (SG) pressure also increased and two safety valves on 11 SG loop lifted causing RCS temperature and pressure to drop rapidly. At 1128 hours0.0131 days <br />0.313 hours <br />0.00187 weeks <br />4.29204e-4 months <br />, a second SI automatically occurred on Train B. After the second SI was reset at 1143 hours0.0132 days <br />0.318 hours <br />0.00189 weeks <br />4.349115e-4 months <br />, the Pressurizer Relief Tank (PRT) {SB} rupture disc operated due to discharge from the PORVs. At 1316 hours0.0152 days <br />0.366 hours <br />0.00218 weeks <br />5.00738e-4 months <br />, an ALERT was declared, in accordance with .Event Classification Guide 17B, as a precautionary step to mobilize engineering resources for assistance, if needed.

Required notifications were made in accordance with 10CFR50.72 and the Salem Emergency Plan.

NRC discretionary enforcement was obtained, to provide an additional 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> beyond the six hours to HOT SHUTDOWN, required by Technical Specification (TS) 3.0.3, due to the blocking of the automatic SI signals. The Pressurizer bubble was reestablished at approximately 1500 hours0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.7075e-4 months <br />. At 0106 hours0.00123 days <br />0.0294 hours <br />1.752645e-4 weeks <br />4.0333e-5 months <br /> on April 8, 1994, cooldown to HOT SHUTDOWN was achieved and at 1124 hours0.013 days <br />0.312 hours <br />0.00186 weeks <br />4.27682e-4 months <br /> (same day), COLD SHUTDOWN was achieved.

ANALYSIS OF OCCURRENCE:

On the morning of April 7, 1994, Salem Unit 1 encountered problems maintaining Main Condenser vacuum due to the ongoing seasonal river grass/debris influx affecting CW circulator availability. A Unit load reduction was in progress to take the Main Turbine off-line.

Reactor power was reduced to 7% with Unit load at 80 MWe. Reduction of power to less than 10% automatically reinstated low power trip setpoints. Due to the power reduction, Tave was 553 degrees F.

Two manual borations were performed and control rods were manually inserted to return Tave to program. During this time, the Senior Reactor Operator (SRO) directed the primary Nuclear Control Operator (NCO) to transfer the power supply to the Group Buses from the station Auxiliary Power Transformer to the 11 and 12 Station Power Transformers. During this evolution, Tave decreased to 530 degrees F. .

Control rods were then withdrawn to increase Tave and Reactor power

• Salem Generating Station DOCKET NUMBER

• LICENSEE EVENT REPORT (LER) TEXT CONTINUATION LER NUMBER PAGE Unit 1 5000272 94-007-00 4 of 9 ANALYSIS OF OCCURRENCE: (cont'd) increased to 25%. Power Range channels 1N42 and 1N44 initiated an automatic Reactor trip and trip of the Main Turbine. An SI occurred immediately thereafter, when the steam line high steam flow bistables actuated on a short duration pressure pulse, concurrent with Tave below 543 degrees F. SI Train A logic partially actuated and SI Train B logic did not actuate due to the short duration of the high steam flow signal.

The high steam flow signal was due to a pressure pulse in the main steam lines caused by closure of the turbine stop valves. Emergency Operating Procedures (EOPs) were entered and components were positioned in response to the SI signal. The SI Train A was reset with the automatic actuation in the "blocked" condition. The Train B automatic logic remained armed. After the Main Steam isolation valves were closed, Tave increased due to decay heat and Reactor Coolant Pump {AB} operation. Pressurizer pressure increased, due to increasing TQV and SI charging flow and the Pressurizer power operated relier valves, lPRl and 1PR2, automatically cycled at 2335 psig. SG pressures also increased in response to increasing Taye*

The secondary NCO did not open the Main steam atmospheric relier valves (MSlOs) {SB} in response to the increasing SG pressures. Two safety valves {SB} on 11 SG loop lifted causing Tave and primary pressure to drop rapidly. Operators were in the process of initiating a manual SI to respond to the plant condition, however, a second SI, from the Train B logic automatically occurred. The Pressurizer Relief Tank (PRT) rupture disc operated due to the PORVs relieving to the PRT. The SI was terminated, the Pressurizer bubble was reestablished and COLD SHUTDOWN was achieved.

Personnel Performance For approximately six weeks prior to the event, the Salem operating shift crews were challenged by the marsh grass/debris affecting the CW intake structure. This has resulted in extended periods of load reductions and numerous transients regarding maintaining operation of the CW circulators.

The Reactor trip is attributed to personnel error, including inadequate command and control. This occurred when the operating crew took inappropriate action, which resulted in an automatic RPS actuation on the Nuclear Instrumentation System

{IG} power range low setpoint. The control rod withdrawal to correct Tave was not correctly implemented and resulted in reactor power increasing at a faster rate than anticipated by the NCO. The Nuclear Shift Supervisor (NSS) did not maintain adequate oversight of changing plant conditions and inappropriately prioritized the actions of the operating crew.

• LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station DOCKET NUMBER LER NUMBER PAGE Unit 1 5000272 94-007-00 5 of 9 ANALYSIS OF OCCURRENCE: (cont'd)

Personnel Performance (cont'd)

He directed the primary NCO to transfer the power supply to the Group Buses from the station Auxiliary Power Transformer to the 11 and 12 station Power Transformers. As a result, the NCO's focus was divided between a number of monitoring activities.

The NSS recognized the low TQve condition and withdrew control rods a few steps, but realizing this was counter to management expectations and training he discontinued this action. After the electrical bus transfer was completed, the NSS directed the NCO to restore Tave*

Following the reduction of Reactor power to 7% and transfer of the Group Buses, the primary Nuclear control-Operator (NCO) recognized that Tave was below the program value. Because of his focused attention on restoring Tav~' the NCO did not properly monitor reactor power while withdrawing rods.

The MSlOs were set in automatic control, but did not respond to the increasing pressure. The operating crew did not adequately communicate RCS temperature and no trending of the Tave value was performed by the NCOs. The required action of the secondary NCO, to take manual control of the valves and open them to prevent lifting of the SG safety relief valves, was not done in a timely.manner.

Equipment Performance At the time of the event, rod control for the Unit was in manual for troubleshooting of suspected problems with automatic rod control. Subsequent troubleshooting, which included testing of the Rod Speed circuitry, showed the Rod Control System was fully functional.

Due to "shadowing" by rod position and Tave being off program low, the Nuclear Instrument System (NIS) Intermediate Range (IR)

Rod Stop at 20% did not actuate to prevent the increase in power to above 25%. It was concluded that the system, functioned, as designed. (The NIS is not an Engineered Safety Feature and credit for it is not taken in the plant accident analysis.)

The first SI occurred due to TQve below program coincident with an erroneous high steam line flow signal. Due to the short duration of the high steam line pressure pulse, the SI signal was only generated by the Train A Solid state Protection System (SSPS) {JC}. Train B SSPS did not respond to the SI signal due to acceptable differences in the actuation time of the SSPS.

• LICENSEE EVENT REPORT (LER) TEXT CONTINUATION PAGE Salem Generating Station DOCKET NUMBER LER NUMBER Unit 1 5000272 94-007-00 6 of 9 ANALYSIS OF OCCURRENCE: (cont'd)

Equipment Performance (cont'd)

The high steam line flow signal occurred when the turbine stop valves closed following the Reactor trip signal. This generated a pressure pulse of sufficient magnitude and duration to actuate the steam line high steam flow bistables. Post event testing verified both channels of high steam flow were functioning within overall time response required by TS and showed no indication of degradation.

Following the first SI, main steam isolation valves (MSIVs) {SB}

13 and 14 MS167 closed, while MSIVs 11 and 12MS167 did not automatically close. The 11 and 12MS167 did not close due to differences in the response of the actuation circuitry to the short duration pulse of the SI signal.

The closure of the Main Turbine stop valves caused a pressure pulse of sufficient magnitude and duration to initiate a high steam flow signal. Due to the short duration of this signal, the SI cleared before some plant equipment could latch and operate to allow completion of all component actions. Although Train "B" did not respond due to the short duration of the pulse, it operated within design specifications and no equipment failures were noted.

Several main steam safety valves operated, per design, during the event, due to the increase in secondary loop pressure.

Operation of the PRT rupture disc occurred per design.

During the cycling of PORVs 1PR1 and 1PR2, the valves performed as designed.

Response of the MS10s to open in automatic is a previously identified condition. The valves have a delay in opening due to the valve controller being below its setpoint for an extended period of time. The design of the valve controller allows the controller output to saturate low when the process is below the control setpoint. This necessitates manual action by the control operator. Following this event, individual problems involving a binding servo drive in the 11MS10 controls, a logic transfer circuit board in the 13MS10 controls, and a missing gear tooth and a misaligned drive shaft in the 14MS10 controls were also identified.

• LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station DOCKET NUMBER LER NUMBER PAGE Unit 1 5000272 94-007-00 7 of 9 ANALYSIS OF OCCURRENCE: {cont'd)

Equipment Performance (cont'd)

The following SI components did not respond to the first SI signal:

Train A 11 and 12MS167, main steam isolation valves for 11 and 12 SGs, did not close.

11, 12, 13, and 14BF13, SG feedwater motor-operated inlet isolation valves did not close.

11 and 12 SG feed pumps did not trip.

Train B SSPS Train B did not respond to the high steam flow SI.

Subsequent testing and analysis indicates the pressure pulse from closure of the main turbine stop valves was not of sufficient duration to initiate the complete train logic. Therefore, it is concluded the above-listed equipment responded, as designed.

The second SI of this event constituted the 21st accumulated SI actuation cycle to date.

APPARENT CAUSE OF OCCURRENCE:

This event is attributed to "Personnel Error", as classified in Appendix B of NUREG-1022. The Reactor trip and initial SI occurred when the NSS failed to maintain adequate command and control, communications, and assigned inappropriate priority of actions in response to the changing plant conditions. The NCO added positive reactivity change at a rate which caused power to increase too quickly, resulting in the reactor trip. The response of the operating crew to the changing conditions of the event was affected by some equipment problems and procedural guidance.

PREVIOUS OCCURRENCES:

Prior events involving excessive CW intake grass/debris have been reported in LERs 272/83-033/0lT, 272/93-011-00, and 311/89-013-00.

A prior event involving greater than 100% level (solid condition) in

• LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station DOCKET NUMBER LER NUMBER PAGE Unit 1 5000272 94-007-00 8 of 9 PREVIOUS OCCURRENCES: (cont'd) the Pressurizer was reported in LER 311/89-005-00.

SAFETY SIGNIFICANCE:

This event did not affect the health and safety of the public. This event is reportable pursuant to 10CFR50.73(a) (2) (iv), due to the RPS and SI actuations and 10CFR50.73(a) {2) (i) (B), due to entry into TS 3.0.3. In addition, this report fulfills the requirement for a Special Report within 90 days of an SI, as required by TS 3.5.2.,

ACTION: b.

The combination of all personnel actions and equipment performance contributed to the plant response. An analysis of that response was performed which addressed the safety significance of all contributing factors. The plant response was reviewed against Condition II safety criteria from Chapter 15 of the Salem Updated Final Safety Analysis Report. This review, which included the safety limits on peak primary and secondary system pressure, and minimum Departure from Nucleate Boiling Ratio, showed these limits were not exceeded. In addition, similar consideration was given to plant component fatigue, fuel integrity, and the effects of lower than normal Tave* This showed all component fatigue analytical conclusions remain valid, no fuel failures have resulted from the event, and the effects of the lower than normal Tave were insignificant with respect to plant safety.

CORRECTIVE ACTION:

The PRT rupture disc has been replaced.

The CW traveling screens were repaired and returned to service.

Operating procedures have been revised, as appropriate.

Simulator training on this event has been conducted with all operating shifts.

The MSlOs controls have been tested and repaired, as required.

Modifications have been made to the MSlOs to improve performance.

Changes to the plant design have been implemented to dampen/filter the erroneous high main steam flow signal generated by closure of the Main Turbine stop valves.

The involved licensed personnel were removed from Licensed Operator duties. Remedial training and evaluation will be performed for these

• LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station DOCKET NUMBER LER NUMBER PAGE Unit 1 5000272 94-007-00 9 of 9 CORRECTIVE ACTION: (cont'd) personnel, prior to their resuming licensed duties.

The PORVs have been inspected and greater than expected wear was noted on several components. Internal parts will be replaced, as required, prior to return to power.

The Salem Emergency Operating Procedures will be reviewed and revised, as required.

General Manager -

Salem Operations MJPJ:pc SORC Mtg.94-039