05000280/LER-1997-008-01, :on 971011,invalid Actuation of ESF Occurred. Caused by Personnel Errors.Main CR Bottled Air Sys Isolated & Containment Hydrogen Analyzer Heat Tracing Actuation Signal Reset
| ML18153A179 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 11/07/1997 |
| From: | Christian D VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | |
| Shared Package | |
| ML18153A178 | List:
|
| References | |
| LER-97-008-01, LER-97-8-1, NUDOCS 9711120275 | |
| Download: ML18153A179 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability |
| 2801997008R01 - NRC Website | |
text
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB NO. 3150-0104 (4-95)
EXPIRES 4/30/98 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 INFORMATION AND RECORDS MANAGEMENT BRANCH (T-6 F33),
(See reverse for required number of digits/characters for each block)
U.S. NUCLEAR REGULATORY COMMISSION. WASHINGTON, DC 20555-0001. AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104 ),
OFFICE OF MANAGEMENT AND
- BUDGET, WASHINGTON, DC 20503.
FACILITY NAME (1)
DOCKET NUMBER (2)
PAGE(3)
SURRY POWER STATION, Unit 1 05000 - 280 1 OF5 TITLE (4)
Invalid Actuation of Engineered Safety Features Due to Personnel Errors EVENT DATE (5)
LER NUMBER (6)
REPORT DATE (7)
OTHER FACILITIES INVOLVED 8)
MONTH YEAR YEAR SEQUENTIAL REVISION FACILITY NAME DOCUMENT NUMBER DAY MONTH DAY YEAR Surry Power Station, Unit 2 05000-281 NUMBER NUMBER 10 11 97 97
-- 008 --
00 11 07 97 FACILITY NAME DOCUMENT NUMBER 05000-OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check one or more) (11)
MODE (9)
N 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) 100,,
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)(iii) 50.36(c)(1) 50.73(a)(2)(v)
Specify in Abstract below 20.2203(a)(2)(iv) 50.36(c)(2) 50.73(a)(2)(vii) or in NRG Form 366A LICENSEE CONTACT FOR THIS LER (12)
NAME I (;;;~N;;;:;~~~de Area Code)
D. A. Christian, Station Manager COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (131
CAUSE
SYSTEM COMPONENT MANUFACTURER REPORTABLE
~
CAUSE
SYSTEM COMPONENT MANUFACTURER REPORTABLE TONPRDS TONPRDS SUPPLEMENTAL REPORT EXPECTED (14)
EXPECTED MONTH DAY YEAR I YES XI NO SUBMISSION (If yes, complete EXPECTED SUBMISSION DATE).
DATE ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16)
On October 11, 1997, with Unit 2 at Refueling Shutdown, an inadvertent actuation of the Unit 2 "B,, train of the High-Consequence Limiting Safeguards (Hi-CLS) system occurred at 14:22 during the replacement of electrical relay 2-CLS-RLY-1812. As designed, the Hi-CLS signal initiated a Unit 2 "B" train Safety Injection (SI) system actuation.
Plant systems and components responded to the Hi-CLS and SI signals, as designed. After verifying that the Hi-CLS signal was not valid, the affected systems were restored to their pre-event configuration.
A Root Cause Evaluation concluded that this event was caused by personnel errors.
To prevent recurrence, training will be conducted to discuss the lessons learned. The event did not affect Unit 2 reactor core cooling, or result in any radiological releases or personnel injuries.
Although the engineered safety feature actuations were invalid, the CLS and SI systems were not completely removed from service.
Therefore, the NRC was notified pursuant to 10 CFR 50. 72 (b)(2)(ii) on October 11, 1997, at 16:39. This report is being submitted pursuant to 10 CFR 50. 73(a)(2)(iv).
9711120275 971107 PDR ADOCK 05000280 s
PDR NRG FORM 366 (4-95)
- rr==============t NR_C FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4-95)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
Surry Power Station Unit 1 DOCKET 05000 - 280 LER NUMBER (6)
YEAR I SEQUENTIAL I REVISION NUMBER NUMBER 97
--008 --
00 PAGE (3) 20F 5 TEXT (If more space is required, use additional copies of NRC Form 366A) (17) 1.0 DESCRIPTION OF THE EVENT On October 11, 1997, with Unit 2 at Refueling Shutdown (RSD), an inadvertent actuation of the Unit 2 "B" train of the High-Consequence Limiting Safeguards (Hi-CLS) system [EIIS-JE] occurred at 14:22 during the replacement of electrical relay 2-CLS-RL Y-1812.
As designed, the Hi-CLS signal initiated a Unit 2 "B" train Safety Injection (SI) system actuation.
Plant systems and components responded to the Hi-CLS and SI signals as designed.
After verifying that the Hi-CLS signal was spurious and that the containment was at atmospheric pressure, control room operators reset the SI signal.
The actuations resulting from tlie SI signal and the operator response actions are described below.
SI system valve, 2-SI-MOV-2867D, opened and was immediately closed by a control room operator. The brief opening of the valve allowed gravity driven flow from the refueling water storage tank [EIIS-BP,TK] to the reactor vessel [EIIS-RPV], resulting in a vessel level increase of 0.2 feet.
Emergency Diesel Generator (EOG) No. 3 started. Although not required, the EOG [EIIS-EK,DG] was conservatively declared inoperable. The requirements of Technical Specification (TS) 3.16. B.1.a were verified to be satisfied and the EOG was returned to an operable status at 16:35.
One bank of the Main Control Room Bottled Air system [EIIS-VI] actuated.
The system was isolated and the affected bank was returned to service at 16:08.
Auxiliary Building Ventilation system fans, 1-VS-F-58A/B, started and the system realigned, as expected. The air flow through the fans was lower than normal since two Unit 2 charging pumps [EIIS-CB, P] and the associated ventilation system dampers [EIIS.:.VF,DMP] were out of service. While control room operators attempted to remedy the low air flow condition, the fans tripped. The low air flow trip signal was immediately cleared and the ventilation system was restored to its normal configuration.
The containment hydrogen analyzer heat tracing [EIIS-FE] received an actuation signal, which initiated a time-delayed start of the system. The signal was manually reset before the heat tracing began to operate.
Auxiliary Feedwater system valves, 2-FW-MOV-251 B/D/F, opened and were immediately closed.
- rr============j NR.C FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4-95)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET LER NUMBER (6)
YEAR I SEQUENTIAL l REVISION NUMBER NUMBER Surry Power Station Unit 1 05000 - 280 97
--008 --
00 PAGE (3) 30F 5 TEXT (If more space is requirea. use additional copies of NRC Form 366A) (17) 1.0 DESCRIPTION OF THE EVENT (Continued)
Containment integrity valves closed as follows:
Component Cooling system valve, 2-CC-TV-209B, closed. While the closure of 2-CC-TV-209B did not affect the operating train of the residual heat removal (RHR) system, the valve was promptly reopened by a control room operator to ensure the standby train of the RHR system [EIIS-BP] remained available.
Aerated Drains system
- valve, 2-DA-TV-200B, Charging system
- valve, 2-CH-TV-2204B, and Gaseous Vents system valve, 2-VG-TV-2098, closed. The valves were immediately reopened.
Although the engineered safety feature (ESF) actuations described above were invalid, the CLS and SI systems were not completely removed from service. Therefore, the NRC was notified pursuant to 10 CFR 50.72 (b)(2)(ii) on October 11, 1997, at 16:39. This report is being submitted pursuant to 10 CFR 50. 73(a)(2)(iv).
2.0
SIGNIFICANT SAFETY CONSEQUENCES AND IMPLICATIONS
The Hi-CLS actuation resulted from an invalid signal that was inadvertently generated during a maintenance activity.
The unit conditions (i.e., RSD) did not warrant the initiation of the Hi-CLS signal, nor was the Hi-CLS system required to be operable. The other ESF actuations described above occurred as a result of the Hi-CLS signal and were also invalid.
After verifying that the Hi-CLS signal was not valid, the affected systems were restored to the pre-event configuration. The event did not affect Unit 2 reactor core cooling or result in any radiological releases or personnel injuries. Therefore, this event resulted in no safety consequences or significant implications and the health and safety of the public were not affected at any time.
3.0
CAUSE
A Category 2 Root Cause Evaluation (RCE) was performed to determine the cause of this event and to recommend corrective actions. The RCE concluded that this event was caused by 1) a "slip" (i e., an unintentional error of execution of a correctly intended action) by utility maintenance personnel during the replacement of electrical relay 2-CLS-RLY-1B12 [EIIS-JE,RLY] and, 2) an error by utility personnel in that the assumed limiting conditions in the Safety Evaluation for the relay replacement were not properly incorporated into the appropriate implementing document. These causes are discussed in more detail below. !
- (4-95)
U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET LER NUMBER (6)
YEAR I SEQUENTIAL I REVISION NUMBER NUMBER Surry Power Station Unit 1 05000 - 280 97
--008 --
00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17) 3.0 CAUSE (Continued)
PAGE (3) 40F 5 While attempting to connect an electrical lead to the appropriate contact during the replacement of 2-CLS-RL Y-1B12, an electrician inadvertently touched the lead to an adjacent contact, thereby creating a short circuit.
This "slip" was caused by the confined space and location of 2-CLS-RLY-1B12 within the electrical cabinet, which made the relay replacement difficult to accomplish.
The short circuit caused a CLS system fuse [EIIS-JE,FU] to fail, which allowed the Hi-CLS signal to be generated.
The potential for a Hi-CLS signal had been recognized and addressed in tr-s Safety Evaluation for the relay replacement.
The Safety Evaluation also recognized that a Hi-CLS signal would initiate an SI system actuation, but noted that the Unit 2 "B" train of the SI system would be defeated by a procedurally controlled temporary modification (TM), which was associated with an unrelated maintenance activity. The use of the TM to prevent a potential SI system actuation was valid.
However, the Safety Evaluation preparer failed to utilize an adequate tracking mechanism, as required by Station Administrative Procedure VPAP-3001, "Safety Evaluations," to ensure that the TM would be in place during the relay replacement. This inadequacy was also not detected by the personnel who reviewed and approved the Safety Evaluation. As a result, the personnel involved with the replacement of 2-CLS-RL Y-1 B 12 were not aware that the TM had been removed when the relay replacement began. The absence of the TM allowed the SI system to actuate in response to the Hi-CLS signal.
4.0 IMMEDIATE CORRECTIVE ACTION(S)
Control room operators verified that the Hi-CLS signal was spurious and that the containment was at atmospheric pressure. The SI signal was subsequently reset and valves 2-SI-MOV-28670 [EIIS-BP,20],
2-FW-MOV-251 B/0/F [EIIS-BA,20],
2-CC-TV-209B [EIIS-CC,ISV],
2-DA-TV-200B [EIIS-WD,ISV],
2-CH-TV-2204B [EIIS-CB, ISV], and 2-VG-TV-209B [EIIS-WE, ISV] were repositioned to their pre-event configuration.
The Main Control Room Bottled Air system was isolated and the containment hydrogen analyzer heat tracing actuation signal was reset.
The low air flow trip signal for 1-VS-F-58A/B [EIIS-VF, FAN] was immediately cleared and the ventilation system was restored to its normal configuration.
EOG No. 3 was declared inoperable and two independent offsite power sources were verified to be operable in accordance with TS 3.16. B.1.a.
A Deviation Report was submitted to document the deviating condition and to initiate
corrective actions
,rr==============1
" (4-95)
U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET LER NUMBER (6)
YEAR I SEQUENTIAL I REVISION NUMBER NUMBER Surry Power Station Unit 1 05000 - 280 97
--008 --
00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17) 5.0
ADDITIONAL CORRECTIVE ACTIONS
PAGE (3)
SOF 5 The affected bank of the Main Control Room Bottled Air system was returned to service at 16:08.
The failed CLS fuse and 2-CLS-RLY-1 B12 were replaced. The Hi-CLS system was reset at 16:12.
EOG No. 3 was secured and returned to an operable status at 16:35.
6.0 ACTIONS TO PREVENT RECURRENCE The lessons learned from this event will be presented to Safety Evaluation preparers/reviewers.
This event will also be discussed in Technical Staff Continuing training.
7.0
SIMILAR EVENTS
LER 50-280/50-281 /95-009-00, "Personnel Error Results in Loss of 4160V Transfer Bus and Start of Emergency Diesel Generators" LER 50-281 /95-006-00, "Unit 2 Auto Reactor Trip Due to Main Transformer Protective Differential Relay Actuation" LER 50-280/~3-001-00, "Reactor Trip and Safety Injection Due to Spurious High Consequence Limiting Safeguards Signal Caused by Malfunctioning Relay" 8.0
ADDITIONAL INFORMATION
Although EDG No. 3 had been conservatively declared inoperable, it would have responded automatically to a Unit 1 demand signal after the Unit 2 SI signal had been reset, except for a brief period during the EOG shutdown evolution.
The status of the Unit 2 charging pumps and the associated ventilation system dampers would not have affected the ability of the Auxiliary Ventilation system to respond to a Unit 1 SI signal. During this event, the Auxiliary Ventilation system would have automatically responded to a Unit 1 demand signal, except for the momentary reset of the low air flow trip signal for 1-VS-F-58A/B.
Unit 1 was operating at 100% power during this event and was otherwise not affected.