05000331/LER-2006-005, Regarding Reactor Scram During Main Turbine Testing
| ML070110194 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 01/05/2007 |
| From: | Vanmiddlesworth G Florida Power & Light Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NG-07-0004 LER 06-005-00 | |
| Download: ML070110194 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 3312006005R00 - NRC Website | |
text
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Duane Arnold Energy Center FPL Energy Duane Arnold, LLC 3277 DAEC Road Palo, Iowa 52324 January 5, 2007 NG-07-0004 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555-0001 Duane Arnold Energy Center Docket 50-331 License No. DPR-49 Licensee Event Report #2006-005-00 Please find attached the subject Licensee Event Report (LER) submitted in accordance with 10 CFR 50.73. This letter contains no new NRC commitments.
Gary Van Middlesworth Site Vice President, Duane Arnold Energy Center FPL Energy Duane Arnold, LLC cc:
Administrator, Region Ill, USNRC Project Manager, DAEC, USNRC Resident Inspector, DAEC, USNRC
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 06/30/2007 (6-2004)
, the NRC (See reverse for required number of may not conduct or sponsor, and a person is not required to respond to, the diqits/characters for each block) information collection.
- 3. PAGE Duane Arnold Energy Center 05000 331 I
OF 5
- 4. TITLE Reactor Scram During Main Turbine Testing
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED
'OR YE SEQUENTIAL REV MONTH A
AR FACILITY NAME DOCKET NUMBER NUMBER NO.
05000 FACILITY NAME DOCKET NUMBER 11 06 2006 2006 5
0 01 05 2007 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) 1 J 20.2201(b)
[]
20.2203(a)(3)(i)
[] 50.73(a)(2)(i)(C)
El 50.73(a)(2)(vii)
El 20.2201(d)
E] 20.2203(a)(3)(ii)
[] 50.73(a)(2)(ii)(A)
[] 50.73(a)(2)(viii)(A)
E] 20.2203(a)(1)
[] 20.2203(a)(4)
El 50.73(a)(2)(ii)(B)
El 50.73(a)(2)(viii)(B)
[1] 20.2203(a)(2)(i)
E] 50.36(c)(1)(i)(A)
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50.73(a)(2)(iii)
[] 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL Q] 20.2203(a)(2)(ii)
[] 50.36(c)(1)(ii)(A)
[] 50.73(a)(2)(iv)(A)
[] 50.73(a)(2)(x)
[] 20.2203(a)(2)(iii)
[] 50.36(c)(2)
E] 50.73(a)(2)(v)(A)
El 73.71(a)(4) 100%
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E] 50.46(a)(3)(ii)
[] 50.73(a)(2)(v)(B)
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E] 50.73(a)(2)(i)(A)
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El OTHER 0
20.2203(a)(2)(vi)
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El 50.73(a)(2)(v)(D)
Specify in Abstract below or in NRC Form 366A
- 12. LICENSEE CONTACT FOR THIS LER
- 14. SUPPLEMENTAL REPORT EXPECTED Q
YES (If yes, complete 15. EXPECTED SUBMISSION DATE)
()
NO ABSTRACT (Umit to 1400 spaces, Le., approximately 15 single-spaced typewritten lines)
On November 6, 2006, at 0110, while operating at 100% power, during the performance of Main Turbine Operational Tests, Surveillance Test Procedure (STP) NS93001, Section 7.3, "Overspeed Trip Device and Mechanical Trip Valve Test (Test A)," a reactor scram occurred. Investigation into this event has determined that the scram was initiated by Reactor Protection System (RPS) actuation due to at least three Turbine Stop Valves less than 90 percent open.
The cause of this Turbine Stop Valve closure was a noise spike, of some level, in combination with normally open relays being closed (possibly stuck). Corrective actions included the replacement of a suspect cable, that was determined to be the most likely cause of the noise spike, the burnishing of relay contacts, and the replacement of a relay board.
There were no actual safety consequences and no effect on public health and safety as a result of this event.
NRC FORM 366 (6-2004)
PRINTED ON RECYCLED PAPER NRC FORM 366 (6-2004)
PRINTED ON RECYCLED PAPERU.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME I1)
DOCKET NUMBER (2)
LER NUMBER (6)
PAGE (3)
YEARI SEQUENTIAL I REVISION Duane Arnold Energy Center 05000331 NUMBER NUMBER 2006 005 00 2of5 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
1. Description of Event
On November 6, 2006, at 0110, while operating at 100% power, during the performance of Main Turbine Operational Tests, Surveillance Test Procedure (STP) NS93001, Section 7.3, "Overspeed Trip Device and Mechanical Trip Valve Test (Test A)," a reactor scram occurred. The testing is performed with the mechanical trip lockout in place to prevent an actual turbine trip. The scram occurred just prior to the portion of the test that resets the Mechanical Trip Valve (MTV), the front standard trip finger, and the Lockout Valve. The scram was initiated by Reactor Protection System (RPS) actuation due to least three Turbine Stop Valves less than 90 percent open.
An investigation into this event determined that an erroneous speed error signal (signal spikes) was induced into the speed control circuitry, most likely when the limit switch for the MTV actuated during the test. The investigation determined that these signal spikes on the speed control circuit may have been transmitted from the front standard due to multiple degraded cables. Combined Intermediate Valve Fast Closure was the first signal received by the First Hit Panel on the Electro Hydraulic Control system due to a speed error signal versus valve position of greater than 5 percent. Turbine Stop Valve #2 (SV2) closure, normally defeated when the main generator is synched to the grid, was allowed to occur in response to the perceived overspeed condition due to the normally open contacts of a mercury-wetted relay being closed (possibly stuck). The remaining three Turbine Stop Valves also closed due to a slaved relation to SV2 which should be defeated except during startup, turbine trip, shell and chest warming.
The slave relation was allowed to occur due to the normally open contacts of another mercury-wetted relay being closed (possibly stuck).
The erroneous speed error signal had not cleared prior to the Turbine Trip because the Intercept Valves had remained closed. Thus, the closure of the Intercept Valves and Control Valves would have occurred in the same timeframe in response to the erroneous speed error signal, even if the Stop Valves had not closed (i.e., mercury wetted relay contacts were open per design), and closure of the normal steam flow path would cause reactor pressure to increase rapidly to the scram setpoint. Additionally, during this event, reactor pressure increased to the scram setpoint approximately three seconds following the scram and RPT from Stop Valve position (i.e., reactor power had been significantly reduced). Therefore, the speed control unit noise sensitivity is cited as the single root cause since the erroneous speed error signal alone would have resulted in a scram. The noise generation and closed relay contacts were determined to be contributing factors.
There were no structures, systems, or components inoperable at the start of the event.
I1. Assessment of Safety Consequences:
Scram #06-001 was an automatic RPS trip initiated by TSV position - 90% open. From the On-Shift Analysis, the key plant responses related to nuclear significance are:
" "A" and "B" Recirculation Pumps tripped [RPT from TSV position]
Maximum reactor pressure was 1075 psig No safety-relief valves opened Minimum reactor water level was 164"U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET NUMBER (2)
LER NUMBER (6)
PAGE (3)
RI SEQUENTIAL REVISION Duane Arnold Energy Center 05000331 Y
NUMBER NUMBER 2006 005 00 3of5 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
" Maximum reactor water level was 208"
" PCIS Group 2 isolation [reactor water level below 1701 PCIS Group 3 isolation [reactor water level below 170"]
PCIS Group 4 isolation [reactor water level below 1701 All safety-related equipment functioned as expected The Turbine Trip with Bypass event (an Anticipated Operational Occurrence) described in Updated Final Safety Analysis Report (UFSAR) 15.1.2.2.1 results in safety-relief valves opening in response to the reactor pressure increase. As described above, the plant response (no safety-relief valves opening) for this scram was much milder and clearly bounded by the Turbine Trip with Bypass event described in UFSAR 15.1.2.2.1. Therefore, in terms of nuclear safety, this scram was non-significant.
Ill. Cause of Event:
One root cause (RC) and three contributing factors (CFs) were identified for this event:
RCI - An inadequately shielded cable allowed a noise spike of some level to interrupt normal Speed Control Unit operation, causing a false overspeed signal.
CF1 - Noise was generated, likely from manipulation of a component in the Front Standard, which coupled to the Speed Control Unit circuitry.
CF2 - Relay K4B17 normally open contacts were closed (possibly stuck), allowing speed error to be connected to SV2 amplifier. This contact should be open at full power.
CF3 - K1 B12 relay normally open contacts were closed (possibly stuck), allowing the slave relation between SV2 and the other SVs to exist while synchronized to the grid such that when SV2 closed, all main stop valves closed.
IV. Corrective Actions
The following corrective actions have resulted from this event:
Corrective Actions to Restore CWO-A74355 (complete) - New mercury-wetted relay boards were installed during troubleshooting and left in place.
CWO-A74357 (complete) - The Main Turbine Mechanical Trip Interlock Switch contacts were bumished.
- CWO-A79431 (complete) - The Mechanical Trip Valve Switch cable was meggered and replaced.U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET NUMBER (2)
LER NUMBER (6)
PAGE (3)
Duane Arnold Energy Center 05000331 YEAR SEQE I
NUMRER 2006 005 00 4of5 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
Interim Corrective Actions (mitigation)
" PWR-35701 (complete) - STP NS930001 was quarantined until Refuel Outage (RFO) 20.
- CA044709 (complete) - Revised STP NS930001 per PWR-36036 to un-quarantine the STP to permit performance of Sections 7.1,7.2, 7.6, 7.7, 7.8, 7.9 and 7.12. Sections 7.3, 7.4, 7.5, 7.10 and 7.11 remain quarantined and are not to be performed until concerns have been resolved per CAPs 44120 and 45313. STP NS930001 was approved and issued on 12/14/2006 as Revision 19.
Corrective Actions to Prevent Recurrence (CATPRs)
CA044782 (due 03/01/2007) - As part of the Heater Bay Cable Replacement Project:
Replace all Front Standard wiring between TJB1 and field devices. Specify optimum cable design for heat and oil environment. This action will be performed under CWO-A74574.
Replace SSPU cables (T00002HA, T00002JA, and T00002KA) from TJB1 to IC049. Cable is original, shows signs of insulation cracking at TJB1, and appears from troubleshooting data to be noise sensitive.
Other Corrective Actions
" CA044783 (due 03/23/2007) - Investigate/test components that may be sources of significant noise and implement a small modification to install noise suppression as warranted to reduce the generation of noise.
CA044784 (due 03/2312007) - Employ voltage checks at startup following RFO20 for suspect relays KL4B17 and K1B12 to ensure proper contact state.
0 CA044785 (due 02/02/2007) - Institute a Pre Planned Task (PPT) for each refueling outage to remove all EHC mercury-wetted relay boards, mechanically agitate them, test them in a test fixture, and return them to service. The PPT is to be first performed during RFO 20.
0 CA044786 (due 02/02/2007) - Develop a test to monitor the front standard and speed control unit, and re-perform STP NS930001, Section 7.3, just prior to RFO20, without tripping the unit.
" CA044787 (due 10/03/2007) - Revise PPT-05960 to inspect the wiring harness in the turbine front standard to include detailed instructions and adjust the frequency from 1 YR to 1 RD to allow a more intrusive inspection. Revised PPT to be implemented beginning with RFO21 due to replacements being performed under CA044782.U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET NUMBER (2)
LER NUMBER (6)
PAGE (3)
Duane Arnold Energy Center 05000331 YR SEQ AL R
N 2006 005 00 5of5 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
V. Additional Information
Previous Similar Occurrences:
A review of LERs at the DAEC over the last 3 years identified no LERs with similar events with similar
causes
EIIS System and Component Codes:
TG Main Turbine Control Fluid System
Reporting Requirements
This report is being submitted under 10 CFR 50.73(a)(2)(iv)(A).