05000260/LER-1997-001-05, :on 970424,reactor Scram Occurred During Surveillance Testing.Caused by Personnel Error.Affected Sys Were Restored to pre-event Conditions,Administered Personnel Corrective Actions & Briefed Maintenance
| ML20140A544 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 05/23/1997 |
| From: | Deroche M TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML20140A534 | List:
|
| References | |
| LER-97-001-05, LER-97-1-5, NUDOCS 9706040267 | |
| Download: ML20140A544 (8) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2) |
| 2601997001R05 - NRC Website | |
text
APPROVED BY GMB NO. 315o-o104 U.S. NUCLEAR REGULATORY COMMISSION EXPIRES 04/3 oleo NRc FORM 366 (4 55)
ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS 50.0 MANDATORY INFORMATION COLLECTION AEQl8MT:
REPORTED LESSONS LEARNED ARE INCORPORATED HRS.
LICENSEE EVENT REPORT (LER)
INTO THE LICENSING PROCESS AND FE0 SACK TO ESTE'TO"YHE
'N NFO AT O AND R (See reverse for required number of
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PAGEC4 DOCKET NUMt;ER (2)
FACluf V NAME (1) 05000260 1OF8 Browns Ferry Nuclear Plant (BFN) Unit 2 YtTLE141 R2 actor Scram as a Result of Personnel Error During Surveillance Testing l
OTHER FACILITIES INVOLVED (8)
REPORr DATE (7)
LER NUMBER (6)
DOCl41 NUMbt.H EVENT DATE (5)
FACluT f NAME SE E
N MONTH DAY YEAR MONTH DAY YEAR
" EAR NU NU R
FACIUlY NAME DOCKET NUMBER 001 00 05 23 97 NA 04 24 97 97 (Check one c,r rnare) (11)
S MP RT IS SUBMITTED PURSUANT TO T23HE REQulREMENTS OF lo CFR ft:
So.73(aH2)d) bo.73(aH2)(vm) l OPERATING N
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LICENSEE CONTACT FoR THIS LER (12)
T ELEPMONE NUMBER (metude Area Cocal NAME (205) 729 - 4889 Mark DeRoche, Industry Affairs Specialist COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13 R OT
CAUSE
SYSTEM COMPONENT MANUFACTURER NPRDS
CAUSE
SYSTEM COMPONENT MANUFACTURER ONPRD MONTH DAY YEAR SUPPLEMENT AL REPORT EXPECTED (14)
EXPECTED SUBMISSION X
No D ATE (15)
YEs (If yes, complete EXPECTED SuBM:sSION DATE).
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ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten knes) bi ti eon April 24,1997, at 1814 Central Daylight Time (CDT), Unit 2 scrammed fr tripped when a high reactor water level trip signal was generated during caused by personnel error when a voit-ohm meter being used in the test was inadvertently connected ac terminals of a companion Channel A relay instead of the intended Channel C relay. When Channel C per the test instruction with the meter connected to Channel A, the two signal associated with the transient. The closure of the MSlVs in this event was an unexpected response t turbine trip transient. The most probable cause of the high steam flow signa decreased to -45 inches as a result of the loss of feedwater and the reactor pressure increase from the MS closure. Water level was restored and maintained by manualinitiation of the Reactor Core Isolation Cooli and automatic initiation of the High Pressure Coolant injection System. The MSRVs functioned to control reac pressure. TVA is reporting this event in accordance with 10 CFR 50.73 (a)(2)(ivi, as any event or co resulted in manual or automatic actuation of any engineered safety feature in addition, TVA willisp R dasy change to increase the response time for the main steam line high flow isolatio function.
I 9706040267 970523 PDR ADOCK 05000260 S
PDR NRC FORM 366 (4-9 0
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NRC FORM 3GGA U.S. NUCLEAR REGULATORY Cor#4SSloN (445) l LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME ~(1)
DOCKET LER NUMBER (6)
PAGE (3)
YEAR LEQUCNTIAL REVlGION j
NUMBER NUMBER i
- '~
2 of 8 Browns Ferry Unit 2 05000260 97 --
001 00
' TEXT {lf more space is required, use naditEal copies of NRc Form 366K) (1/)
I.
PLANT CONDITIONS
Units 2 and 3 were at approximately 100 percent power (3293 megawatts thermal). Unit 1 was shutdown and defueled.
II.
DESCRIPTION OF EVENT
A.
Event on April 24, 1997, at 1814 hours0.021 days <br />0.504 hours <br />0.003 weeks <br />6.90227e-4 months <br /> Central Daylight Time (CDT),
Unit 2 received engineered safety feature actuations (ESF) [JE]
and a reactor scram from full power due to a turbine trip caused by a reactor high water level signal.
At 1814 CDT, Instrument Maintenance personnel [ utility, nonlicensed) were performing surveillance instruction 2-SI-4.2.B-ATU(C), Core and Containment Cooling Systems Analog Trip Unit Functional Te.st.
As part of the test, Instrument Maintenance personnel were to connect the volt-ohm meter across contacts associated with relay 2-62-3-208C.
However, they placed the test leads acreas contacts associated with relay 2-62-3-208A.
Subremmatly, the craftsman inserted a trip signal to 2-LS-3-208C and the logic for a high reactor water level trip was completed.
The main turbine [TA) and all three reactor feed pumps [SJ) tripped as a result of the high water level trip signal. The reactor automatically scrammed as a result of the turbine trip.
The main steam isolation valves (MSIV) [ISV) closed due to a high main steam line flow signal, PCIS Group 1.
This signal was caused by the instrument response to the pressure wave, which resulted from the turbine stop valve closure and process noise from safety relief valve operation.
At 1815 CDT, the unit operator manually initiated Reactor Core Isolation Cooling (RCIC) [BN] to maintain reactor water level.
Water level centinued to decrease and at 1817 CDT, when level reached -45 irches, the High Pressure Coolant Injection System (HPCI) [BJ) automatically ir.itiated and injected into the vessel.
In addition to the above actuations, the scram resulted in the actuation or isolation of the following Primary Containment Isolation [JE] [PCIS) systems / components.
PCIS group 2, shutdown cooling mode of Residual Heat Removal [BO) system; Drywell floor drain isolation valve; Drywell equipment drain sump isolation valve [WP).
PCIS group 3, Reactor Water Cleanup [CE).
Il
NRC FORM 3GGA U.S. NUCLEAR REGULATORY cCf44S310N 1
(445)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION 1
FACILITY NAME (1)
DOCKET 11 LER NUMBER (6)
PAGE (3)
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YEAR GEQUENTIAL REVIGION NUMBER NUMBER Browns Ferry Unit 2 05000260 3 of 8 97 --
001 00
~ TEXT (if more space is required, une addibonal copies oTRRC Form 366Af(11)
PCIS group 6, Primary Containment Purge and Ventilation l
[JM); Unit 2 Reactor Zone Ventilation [VB); Refuel Zone Ventilation [VA); Standby Gas Treatment (SGT) [BH] system; Control Room Emergency Ventilation (CREV) [VI).
PCIS group 8, Transverse Incore Probe [IG).
The reactor scram was reset by 1824 CDT.
The affected systems were returned to pre-event status by 1940 CDT.
All safety systems responded as expected during the reactor scram, except for the MSIV closure. The MSIV closure is further discussed in Section II.G.
This event is reportable in accordance with 10 CFR S0.73 (a) (2 ) (iv), as any event or condition that resulted in manual or automatic actuation of any engineered safety fea' cure including the reactor protection system.
B.
Inoperable Structures, components, or systems that contributed to l
the Event:
None.
C.
Dates and Approximate Times of Major occurrences
1 April 24, 1997 at 1814 CDT The Unit 2 Reactor received a j
full scram due to a turbine trip caused by a resctor high water level signal. An MSIV closure also occurred due to a high steam line flow signal.
April 24, 1997 at 1835 CDT After verifying that no steam line break had occurred, the operating crew re-opened the MSIVc and re-established the normal heat sink.
April 24, 1997 at 1914 CDT TVA made a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> nonemergency notification to NRC in accordance with 10 CFR 50.72 (b) (1) (iv) and l
a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> nonemergency notification to NRC in accordance with 10 CFR 50.72 (b) (2) (ii).
April 24, 1997 at 1920 CDT The PCIS actuations were reset.
SGT and CREV systems are returned to standby readiness.
NRc FORM 386A (445)
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l NRC PORM 386A U.S. NUCLEAR REGULATORY coNWESSION (445)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION I
FACILITY NAME (1)
DOCKET LER NUMBER (6)
PAGE (3)
YEAR SEQUENTIAL REVislON NUMBER NUMBER Browns Ferry Unit 2 05000260 4 of 8 3
97 --
001 00 TEXT (if more space is required, use additional copies of NRc Form 366A) (14)
D.
Other systems or secondary Functions Affected
l None.
E.
Method of Discovery
The Unit 2 Operator received alarms associated with the full reactor scram and main steam isolation valve closure.
7 F.
Operator Actions
Operator actions taken during this event were as expected. Main l
steam relief valves (MSRV) [RV) were used to control reactor pressure. RCIC and HPCI were used to increase reactor water l
level. After verifying that no main steam line break existed, I
the control room operators opened the MSIVs and re-established l
the normal heat sink.
G.
Safety System Responses:
The safety systems listed in section IIA of this report responded to the reactor scram as designed, with the exception of the MSIVs.
I An unexpected PCIS [JE] group 1 isolation occurred on high steam i
flow approximately 500 milliseconds after the turbine trip was j
initiated. The high flow signal occurred on three of four PCIS i
channels. This signal was of short duration and subsequently the logic relays dropped out for approximately 20 milliseconds. This action is not expected in a turbine trip event and has not been previously experienced at Browns Ferry.
When a turbine trip occurs, a pressure wave originates at the turbine stop valves and is transmitted back toward the reactor vessel. The magnitude of the pressure exceeds reactor pressure vessel dome pressure because the large volume of the vessel dissipates the pressure wave.
Following the reactor scram, MSRVs (RV) 1-31 and 1-34 (both on main steam line C) opened approximately 300 milliseconds after the turbine trip. This was attributed to the passage of the pressure wave through main steam line C.
One complete wave cycle is approximately 600 milliseconds as observed on the Integrated Computer System (ICS)
(ID).
Opening of MSRVs concurrent with the initie.1 pressure wave cycle would have the effect of increasing the amplitude of the wave.
It was also determined from ICS data that the flow indicator for main steam line C had process noise resulting from MSRV operation on that line. TVA believes that the high flow signal was caused by the additive combination of process noise on steam line C flow element and the effects of the passing pressure wave.
NRc PoRM 366A (445)
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i NRC PORM 366A U.S. NUCLEAR REGULATORY COfmBSSloN
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(4-95) j LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET LER NUMBER (6)
PAGE (3)
YEAR GEQUENTIAL REVislON NUMBER NUMBER Browns Ferry Unit 2 05000260 5 of 8 97 001 00 TEXT (if more space es reqwred, use additional copies of NRc Form 306A) (1/)
III.
CAUSE OF THE EVENT
I i
A human performance evaluation was conducted and the following causes l
were identified.
l l
A.
Immediate Cause:
l The immediate cause of the main turbine trip was a high reactor l
water level signal. This was followed by a reactor scram.
B.
Root Cause:
]
The root cause of the event was personnel error in that the craftsmen did not perform self-checking continuously. They properly located the relay to be tested but then broke eye l
contact with the component to physically access the test jacks.
While connecting the test leads, the craftsmen focused on connecting the leads to the correct terminal and did not re-l verify that they were on the correct relay.
Subsequently, the j
test leads were incorrectly placed on relay 2-62-3-208A instead of relay 2-62-3-208C, as required by the surveillance instruction.
1 C.
Contributing Factors:
Labels identifying relays 2-62-3-208A and -208C are clearly visible from a standing position. However, they are not visible t
when connecting test equipment to any of the two lower rows of terminals on the relay base.
IV.
ANALYSIS OF THE EVENT
This transient was initiated from an unexpected high reactor water level trip signal generated during the performance of a surveillance instruction. The required safety systems performed as needed to properly control the event.
One unexpected equipment response did occur during the transient.
Main steam isolation valves are not expected to close after a turbine trip. The main steam isolation valves closed upon receipt of a main steam line high flow signal.
i d
NCC FORM 366A (4-95) l 1
l
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NRC FORM 3GSA U.S. NUCLEAR REGULATORY CofMSSION (4-Ob)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET LER NUMBER (6)
PAGE (3)
YEAR GEQUENTIAL Revision NUMEER NUMBER Browns Ferry Unit 2 05000260 6 of 8 97 --
001 00 TEXT (if more space is required, use additaonal copies of RRC7orm 366A) (17)
Closure of MSIVs during a turbine trip transient initiated by the effects of the trip is bounded by existing analyses for turbine trip without bypass, feedwater controller failure, and MSIV closure with flux scram. The transient pressure and flux effects of a turbine trip occur in a much shorter time frame than those of MSIV closure since MSIVs have a closing time of three seconds and both turbine valves and MSIVs interrupt steam flow in the same path. Therefore, the analyzed transients individually produce more limiting results. This event did not affect the health and safety of plant personnel or the public.
V.
CORRECTIVE ACTIONS
A.
Immediate Corrective Actions
The affected systems were restored to pre-event conditions.
B.
Corrective Actions to Prevent Recurrence:
TVA will administer personnel corrective actions in accordance l
with TVA policy to those involved in the event.
Appropriate maintenance personnel have been briefed on management's expectations for the performance of instruction steps requiring second party verification.
l TVA management has instructed instrument maintenance personnel that all components in steps requiring verification must be identified such that if visual contact with the component is subsequently lost, the tag will enable the craftsman to easily locate the correct component.
TVA will issue a design change to increase the response time of 2
main steam line flow instruments.
TVA will place supplemental labels under the sub ect relays to facilitate placement of test leads when required l
1 TVA does not consider these actions P.egulatory Commitments. The TVA corrective action program will track completion of the corrective actions.
PCC FORM 366A (4-95)
NRC FORM 3GGA U.S. NUCLEAR REGULATORY COf411SSIOIC (4-95)
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LICENSEE EVENT REPORT (LER) j TEXT CONTINUATION FACILITY NAME (1)
DOCKET LER NUMBER (6)
PAGE (3)
YEAR LEQUENTiAL REVIsloN NUMBER NUMBER Browns Ferry Unit 2 05000260
~
7 of B 97 --
001 00
~ TEX f (if more space is required, use additional copese of NRc Form 366A) (17) l TVA has taken several actions to address the human performance aspect of this and previous events:
TVA has included specific human performance lessons learned in pre-job briefings.
In order to focus attention on critical activities, TVA has
+
modified the Scheduled Surveillance sections of the Browns Ferry " Plan of the Day" to indicate which Surveillance Instructions could potentially cause a half-scram, an Engineered Safety Feature, or a turbine trip.
TVA has increased management observation of the performance of Surveillance Instructions.
TVA has focused on improving pre-job briefings and making better use of pre-job briefings.
To foster a deeper sense of accountability for the maintenance shops and crews, TVA has emphasized atountability for personnel actions at the general foreman, foreman, and shop t
manager level.
VI.
ADDITIONAL INFORMATION
A.
Failed Components:
None B.
Previous LERs on similar Events:
The folbwing L3Rs describe similar events, however, the correct ive actions implemented for these events could not prevent the event under consideration.
LER 7.96/96004: Unplanned Manual Start of Emergency Diesel Generator During a Scheduled Redundant Start Test: During a scheduled performance of the Diesel Generator 3C Redundant Start Test, EDG 3D wa.s manually started from the Unit 3 Main Control Room. When the operator was requested to start EDG 3C, the individual instead started EDG 3D.
The root cause of the event was personnel error due to inattention to detail.
LER 260/97002: Unit 3 HPCI System Unexpected Isolation: While performing a surveillance instruction for the functional testing of Unit 3 HPCI steam supply low pressure switches, a volt-ohm meter was inadvertently placed across a wrong pressure switch.
The cause of this event was personnel error, as a result of mis-positioning a volt-ohm meter lead. This was the result of a lack of self-checking and second party verification.
r 1
NRC FORM 3GGA U.S. NUCLEAR REGULATORY CONMSSION (4-95)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET IIR NUMBER (6)
PAGE (3)
YEAR SEQUENTIAL.
REVISION NUMBER NUMBER Browns Ferry Unit 2 05000260 8 of 8 97 --
001 00
~TEft (if more space is required, use additional copies of NRc Form 366A) (17)
LER 296/96002: Unit 3 Scrammed Following Loss of Reactor Feed Pump 3C: A low reactor water level scram occurred on Unit 3 as a result of the loss of Reactor Feed Pump 3C while aligning RFP 3C's oil purification system. The loss of the reactor feed pump was caused by personnel error. An Assistant Unit Operator improperly aligned oil valves resulting in draining the RFP oil tank.
LER 260/95004: Reactor Scram Resulting From Personnel Error During a Surveillance Test: Unit 2 reactor scrammed during the performance of the 2-SI-4.2.B-ATU(C), Core and Containment System Analog Trip Unit Functional Test.
The root cause of the event was personnel error.
I&C personnel prematurely repositioned the ATWS mode switch from the ' Test' to the ' Normal' position prior to resetting the ATWS/ARI logic which caused a low scram pilot air header pressure and reactor scram.
VII.
Col 44ITMENTS TVA will administer personnel corrective actions in accordance with TVA policy to those involved by June 15, 1997.
Energy Industry Identification System (EIIS) system and component codes are identified in the text with brackets (e.g.,
[XX])._ _ _ _ _ - - _ _