05000280/LER-1998-014, :on 981126,manual Reactor Trip in Response to Main Feedwater Regulating Valve Failure Occurred.Caused by Dislocation of Retaining Clip in Positioner.Control Room Operators Placed Unit in Safe,Shutdown Condition
| ML18152B578 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 12/16/1998 |
| From: | Grecheck E VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | |
| Shared Package | |
| ML18152B577 | List: |
| References | |
| LER-98-014, LER-98-14, NUDOCS 9812230237 | |
| Download: ML18152B578 (4) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii) 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability |
| 2801998014R00 - NRC Website | |
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NRC FORM 366 U.S. NUC[ AR REGULATORY COMMISSION APPROVEDB B NO. 3150-0104 EXPIRES 06/30/2001 (6-1998)
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
FACILITY NAME (1)
DOCKET NUMBER (2)
PAGE (3)
SURRY POWER STATION, Unit 1 OF 4 05000-280 1
TITLE (4)
Manual Reactor Trip in Response to Main Feedwater Regulating Valve Failure EVENT DATE (5)
LER NUMBER 6)
REPORT DATE (7)
OTHER FACILITIES INVOLVED (8)
FACILITY NAME DOCKET NUMBER I
SEQUENTIAL REVISION MONTH DAY YEAR YEAR NUMBER NUMBER MONTH DAY YEAR 05000 --
FACILITY NAME DOCKET NUMBER 11 26 1998 1998 -
014 -- 00 12 16 1998 ni;nnn --
THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check one or more) (11)
N 20.2201/b) 20.2203(a)(2)(v) 50.73/a)(2)(i)
- 50. 73(a)(2)(viii)
ER 20.2203(a)(1) 20.2203(a)(3)(i) 50.73(a)(2)(ii)
- 50. 73(a)(2)(x)
L 10 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 or i
20.2203(a)(2)(iv)
- 50.36(c)(2) 50.73(a)(2)(vii) in NRC Form 366A LICENSEE CONTACT FOR THIS LER (12)
NAME TELEPHONE NUMBER (Include Area Code)
E. S. Grecheck, Site Vice President (757) 365-2000 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)
REPORTABLE If~
REPORTABLE
CAUSE
SYSTEM COMPONENT MANUFACTURER TO EPIX
CAUSE
SYSTEM COMPONENT MANUFACTURER TO EPIX B
SJ V
Bailey Controls y
Division
- Ar SUPPLEMENTAL REPORT EXPECTED (14)
EXPECTED MONTH DAY YEAR IYES Ix SUBMISSION (If yes, complete EXPECTED SUBMISSION DATE).
NO DATE (15)
ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single~spaced typewritten lines) (16)
On November 26, 1998, at 21 :08, the Unit 1 steam generator (SG) "BIi main feedwater regulating valve, 1-FW-FCV-1488, failed closed. A control room annunciator alarmed indicating a difference between the SG "B" feedwater and steam flow parameters. In response, the control room operator placed 1-FW-FCV-1488 in the manual control mode and attempted to open the valve.
1-FW-FCV-1488 remained closed, however, and the "B" SG level continued to decrease. To avert an automatic reactor trip due to low SG level coincident with a steam/feedwater flow*mismatch, the control room operator initiated a manual reactor trip. The Unit 1 reactor tripped from 81 % power and was followed by an automatic turbine trip, as designed. Appropriate operator actions were taken in accordance with emergency operating procedures to* ensure the performance of system automatic actions and to respond to abnormal conditions. The event was caused by the dislocation of a retaining clip in the positioner pilot valve, which caused 1-FW-FCV-1488 to fail closed.
Approved Root Cause Evaluation recommendations, designed to-prevent the recurrence of a similar event, will be implemented. The NRG was notified pursuant to 10 CFR 50.72 (b)(2)(ii) on November 26, 1998 at 23:35. This report is being submitted pursuant to 1 O CFR 50.73 (a)(2)(iv).
9812230237 981216 PDR ADOCK 05000280
=-
S PDR NRC FORM 366 (6-1998)
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U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)
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SEQUENTIAL REVISION YEAR NUMBER NUMBER SURRY POWER STATION, Unit 1 05000 -- 280 1998 -
014 --
00 2 OF 4 TEXT (If more space is required, use additional copies of NRG Form 366A) (17) 1.0 DESCRIPTION OF THE EVENT On November 26, 1998, at 21:08, the Unit 1 steam generator (SG) [EIIS-AB,SG] "B" main feedwater regulating valve [EIIS-SJ,FCV], 1-FW-FCV-1488, failed closed. A control room annunciator [EIIS-IB] alarmed indicating a difference between the SG "B" feedwater and steam flow parameters. In response, the control room operator placed 1-FW-FCV-1488 in the manual control mode and attempted to open the valve. 1-FW-FCV-1488 remained closed, however, and the "B" SG level continued to decrease. To avert an automatic reactor trip due to low SG level coincident with a steam/feedwater flow mismatch, the control room operator initiated a manual reactor trip [EIIS-JC]. The Unit 1 reactor tripped from 81% power and was followed by an automatic turbine trip [EIIS-TA,TRB], as designed.
The auxiliary feedwater pumps [EIIS-EIIS-BA-P] started on low-low SG water level and provided flow to the SGs. The Anticipated Transient Without Scram Mitigation System Actuation Circuitry (AMSAC) [EIIS-JE] armed and initiated, as designed. The main steam dump valves [EIIS-SB,TCV] automatically opened to admit steam to the main condenser.
The RCS reached a minimum temperature of approximately 538°F and subsequently stabilized at 547°F. The reactivity shutdown margin was calculated following the RCS cooldown to ensure that Technical Specification and administrative shutdown margin limits were satisfied.
The following discrepancy was noted during the post-trip response:
Moisture separator reheater control valve, 1-MS-FCV-1040 [EIIS-FCV], failed to fully close.
The NRC was notified pursuant to 1 O CFR 50.72 (b)(2)(ii) on November 26, 1998 at 23:35. This report is being submitted pursuant to 10 CFR 50.73 (a)(2)(iv) as an event that resulted in the actuation of engineered safety features and the reactor protection system.
2.0
SIGNIFICANT SAFETY CONSEQUENCES AND IMPLICATIONS
This event resulted in no safety consequences or implications. Appropriate operator actions were taken in accordance with emergency operating procedures to ensure the performance of system automatic actions and to respond to abnormal conditions. The unit was quickly brought to a stable, no-load condition. Therefore, the health and safety of the public were not affected at any time during this event.
~-------=------~---=========----~-~----- --
-(6-1998)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION U.S. NUCLEAR REGULATORY COMMISSION FACILITY NAME (1)
DOCKET (2)
LEA NUMBER (6)
PAGE (3)
YEAR I SEQUENTIAL I REVISION NUMBER NUMBER SURRY POWER STATION, Unit 1 05000 -- 280 1998 -
014 --
00 3 OF 4 TEXT (If more space is required, use additional copies of NRC Form 366A) (17) 3.0
CAUSE
A Category 1 Root Cause Evaluation (RCE) was initiated on November 27, 1998, to determine the cause of this event and to recommend corrective actions. The RCE has preliminarily concluded that a retaining clip in the positioner pilot valve became dislocated, which caused 1-FW-FCV-1488 to fail closed. The RCE is continuing to investigate the cause of the retaining clip dislocation.
4.0 IMMEDIATE CORRECTIVE ACTION($)
Following the reactor trip, Control Room Operators acted promptly to place the unit in a safe, shutdown condition in accordance with emergency and other operating procedures.
The Shift Technical Advisor monitored the critical safety function status trees to ensure that plant parameters remained acceptable.
5.0
ADDITIONAL CORRECTIVE ACTIONS
1-MS-FCV-104D was examined and found to be held in a partially ~pen position as a result of the stem binding. The stem was cleaned and the valve was tested satisfactorily.
The pilot valve assembly for 1-FW-FCV-1488 and the retaining clip in the positioner were replaced. 1-FW-FCV-1488 was subsequently tested satisfactorily.
The RCE team is evaluating unit conditions and systems response contributing to the RCS cooldown following the reactortrip.
6.0 ACTIONS TO PREVENT RECURRENCE Approved RCE recommendations, designed to prevent the recurrence of a similar event, will be implemented.
7.0
SIMILAR EVENTS
LER 50-281/1990-003-00 Manual Reactor Trip Due to Failure of "A" Main Feedwater Regulating Valve LER 50-281/1991-011-00 High Steam Generator Level Due to Main Feedwater Regulating Valve Oscillations Results in ESF Actuations and Reactor Trip
,\\============ (6-1998)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION U.S. NUCLEAR REGULATORY COMMISSION ii--------- _FA_C_l:;:.;LITY.c....;_N....ccA....:.Mc:.:E:....,('"'"'1) ________
11 __
D_O_CK:....E_T_,_(2-'-)---ll---~L-ER_NU_M_B_E_R-;.(6-'-)----it--P-A_G_E_,_(3-'-) --ii YEAR I SEQUENTIAL I REVISION SURRY POWER STATION, Unit 1 05000 -- 280 TEXT (If more space is required, use additional copies of NRG Form 366A) (17) 8.0 MANUFACTURER/MODEL NUMBER Bailey Controls Division AV1 Series Positioner 9.0
ADDITIONAL INFORMATION
Unit 1 was returned to service on November 27, 1998.
NUMBER NUMBER 1998 -
014 --
00 4 OF 4 Unit 2 was operating at ~ 00% power and was not affected by this event.