05000455/LER-2012-001-01, Regarding Manual Reactor Trip During Power Ascension Due to Steam Generator Level Approaching Turbine Trip Setpoint Caused by an Overly Complex Startup Procedure
| ML12097A354 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 04/06/2012 |
| From: | Tulon T Exelon Generation Co, Exelon Nuclear |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| Byron 2012-0041 LER 12-001-01 | |
| Download: ML12097A354 (4) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i) 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 |
| 4552012001R01 - NRC Website | |
text
April 6, 2012 L TR: Byron 2012 - 0041 File:
1.10.0101 wwwexeloncorpocom U. S. Nuclear Regulatory Commission A TIN: Document Control Desk Washington, DC 20555-0001 Byron Station, Unit 2 Facility Operating License No. NPF-66 NRC Docket No. STN 50-455 E
10 CFR 50.73 Subject: Licensee Event Report 2012-001-00, "Manual Reactor Trip During Power Ascension Due to Steam Generator Level Approaching Turbine Trip Setpoint Caused by an Overly Complex Startup Procedure" The enclosed Licensee Event Report (LER) is being submitted in accordance with 10 CFR 50.73, "Licensee event report system," paragraph (a)(2)(iv)(A). The LER involves a Unit 2 manual reactor trip and Auxiliary Feedwater system actuation.
There are no regulatory commitments contained in this letter. Should you have any questions concerning this submittal, please contact Mr. David Gudger, Regulatory Assurance Manager, at (815) 406-2800.
Tim J. Tulon Site f tee President Byrcfn Station
Enclosure:
LER Number 455-2012-001-00
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/3112013 (10-2010)
, the NRC may digits/characters for each block) not conduct or sponsor, and a person is not required to respond to, the information collection.
r' PAGE Byron Station, Unit 2 05000455 1 OF 3
- 4. TITLE Manual Reactor Trip During Power Ascension Due to Steam Generator Level Approaching Turbine Trip Setpoint Caused by an Overly Complex Startup Procedure
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED YEAR I SEQUENTIAL I REV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR NUMBER NO.
MONTH DAY YEAR N/A FACILITY NAME DOCKET NUMBER 02 06 2012 2012 - 001 -
01 04 06 2012 N/A
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check al/ that apply) o 20.2201 (b) o 20.2203(a)(3)(i) o 50.73(a)(2)(i)(C) o 50. 73(a)(2)(vii) 1 o 20.2201 (d) o 20.2203(a)(3)(ii) o 50.73(a)(2)(ii)(A) o 50.73(a)(2)(viii)(A) o 20.2203(a)(1) o 20.2203(a)(4) o 50.73(a)(2)(ii)(B) o 50.73(a)(2)(viii)(B) o 20.2203(a)(2)(i) o 50.36(c)(1 )(i)(A) o 50.73(a)(2)(iii) o 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL o 20.2203(a)(2)(ii) o 50.36(c)(1 )(ii)(A)
[8J 50.73(a)(2)(iv)(A) o 50.73(a)(2)(x) o 20.2203(a)(2)(iii) o 50.36(c)(2) o 50.73(a)(2)(v)(A) o 73.71(a)(4) 025 o 20.2203(a)(2)(iv) o 50.46(a)(3)(ii) o 50.73(a)(2)(v)(B) o 73.71(a)(5) o 20.2203(a)(2)(v) o 50.73(a)(2)(i)(A) o 50.73(a)(2)(v)(C) o OTHER o 20.2203(a)(2)(vi) o 50.73(a)(2)(i)(B) o 50.73(a)(2)(v)(D)
Specify in Abstract below or in NRC Form 366A
- 12. LICENSEE CONTACT FOR THIS LER FACILITY NAME iTELEPHONE NUMBER (Include Area Code)
David Gudger, Regulatory Assurance Manager (815) 406-2800 CAUSE SYSTEM COMPONENT MANU-REPORTABLE
CAUSE
SYSTEM COMPONENT MANU-REPORTABLE FACTURER TO EPIX FACTURER TO EPIX
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED MONTH DAY YEAR o YES (If yes, complete 15. EXPECTED SUBMISSION DATE)
SUBMISSION [8J NO DATE ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
At approximately 1700 hours0.0197 days <br />0.472 hours <br />0.00281 weeks <br />6.4685e-4 months <br />, on February 6, 2012, the Unit 2 Operating crew was in the process of conducting a unit startup in accordance with startup procedures. The reactor was at approximately 25 %
power. During the step that switched Feedwater (FW) flow from the upper nozzle of the 2C Steam Generator (SG) to its lower nozzle via the 2C FW Isolation Valve (FWIV), the Reactor Operator controlling the SG levels manually was also focused on maintaining subcooling in the tempering line and satisfying water hammer interlocks. Consequently, the RO was not closely monitoring the 2C SG level increase. Another assisting RO recognized the level increase was approaching the high level setpoint and informed the RO. Subsequently, he RO reduced FW flow; however, the level continued to increase and approach the high level setpoint of 80.8% (Le., P-14), which causes an automatic turbine trip and FW isolation. In anticipation of reaching this setpoint, the Unit 2 Senior RO directed a manual reactor trip at 1719 hours0.0199 days <br />0.478 hours <br />0.00284 weeks <br />6.540795e-4 months <br />. Almost simultaneous to the llanual reactor trip the 2C SG level reached the P-14 setpoint. The reactor trip signal and P-14 both generate
- In automatic turbine trip and FW isolation. The turbine tripped and FW isolated, as designed. In accordance
~i!hE reactor trip response procedures, the 2A and 2B Auxiliary Feedwater pumps were started. The cause of was ut:u::m II I lIt:d to IJI UI.;t:U crew pt:ifulli he startup procedure will be revised and personnel performance issues were remediated.
Description of Event
Event DatelTime: February 6, 2012/1719 hours Unit 2 was in Mode 1 - Power Operation
- 6. LER NUMBER
- 3. PAGE YEAR I SEQUENTIAL I REV NUMBER NO.
2 OF 3
2012 001 01 The Unit 2 Reactor Coolant (RC) [AS] System was at normal operating temperature and pressure for 25%
power. No structures, systems, or components were inoperable at the start of this event that contributed to the event. The 2C Feedwater (FW) [SJ] Isolation Valve (FWIV) was degraded due to mechanical binding issues when opening. A warming blanket was installed on the valve to aid in stroking it open without binding.
At approximately 1700 hours0.0197 days <br />0.472 hours <br />0.00281 weeks <br />6.4685e-4 months <br />, on February 6, 2012, the Unit 2 Operating crew was in the process of conducting a unit startup in accordance with startup procedures. At approximately 25% reactor power, the Reactor Operator (RO) responsible for Steam Generator (SG) FW flow and level control and the RO responsible for opening the FWIVs were ready to perform the steps that transfer FW flow from the Steam Generators (SGs) [SS] upper nozzles to the lower nozzles through the four FWIVs. The Unit 2 Senior Reactor Operator (SRO) was overseeing the evolution. To open the FWIVs, three water hammer prevention interlocks ensuring adequate FW flow and temperature needed to be satisfied. The startup procedure has the FW Regulating Valves (FRVs) in manual control for this evolution. Additionally, the 2C FWIV was to be the first valve to open due to issues with it potentially binding.
At the start of the evolution, the SG RO was focused on maintaining subcooling in the 2C FW tempering line by opening the 2C FRV and providing more FW through the line. The SG RO was also concerned with satisfying the water hammer prevention flow and temperature interlocks. The FWIV RO began opening the 2C FWIV and the 2C SG level began to rise as expected. As the 2C FWIV was opening, the SG RO continued to focus on maintaining subcooling and satisfying the waterhammer prevention interlocks and not closely monitoring the 2C SG level increase. During this time, the Unit 2 SRO was focused on the proper performance of the 2C FWIV due to its degraded condition and not maintaining proper oversight of the overall evolution.
Another RO, assisting in the startup, noticed and communicated to the SG RO and Unit 2 SRO that the 2C SG level was above 70% and rapidly rising. The SG RO reduced FW flow to the 2C SG; however, the level continued to increase and approach the high level setpoint of 80.8% (Le., P-14), which causes an automatic turbine trip and FW isolation. In anticipation of reaching this setpoint, the Unit 2 SRO directed a manual reactor trip at 1719 hours0.0199 days <br />0.478 hours <br />0.00284 weeks <br />6.540795e-4 months <br />. Almost simultaneous to the manual reactor trip, the 2C SG level reached the P-14 setpoint The reactor trip signal and P-14 both generate an automatic turbine trip and FW isolation signal. The turbine tripped and FW isolated, as designed. The Operating crew responded to the reactor trip using appropriate emergency procedures. At 1734 hours0.0201 days <br />0.482 hours <br />0.00287 weeks <br />6.59787e-4 months <br />, in accordance with reactor trip response procedures, the 2A and 2S Auxiliary Feedwater (A F) [SA] pumps were manually started to provide water to the SGs. The AF pumps operated as expected and were secured at 1954 hours0.0226 days <br />0.543 hours <br />0.00323 weeks <br />7.43497e-4 months <br /> and 2037 hours0.0236 days <br />0.566 hours <br />0.00337 weeks <br />7.750785e-4 months <br /> for the 2A AF and 2S AF pumps, respectively. Unit 2 remained stable in Mode 3 pending post reactor trip event review.
An Emergency Notification System notification was made to the NRC in accordance with 10 CFR 50.72 (b)(2}(iv){S) and (b)(3)(iv)(A). In addition, an LER is required to be submitted in accordance with 10 CFR (10-2010)
U.S. NUCLEAR REGULATORY COMMISSION
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE 3
OF YEAR I SEQUENTIAL I REV NUMBER NO.
Byron Station, Unit 2 05000455 3
2012 001 01
~--------------------------~--------~--------------,----~-----------~
A. Cause of the Event
The root cause was determined to be an overly complex procedure requiring the FRVs to be in manual control during this point in the reactor startup process. Coordination of the opening of the FRVs and FWIVs, while the SG RO was reacting to other issues such as maintaining subcooling in the tempering lines and satisfying water hammer prevention interlocks, became overly distracting and led to the 2C FRV to be opened to far. A benchmark of the Braidwood Station startup procedure indicates having the FRVs in automatic control at this point in the startup is less complex and a more effective means to perform this evolution.
Contributing causes include:
Inadequate personnel performance issues by the licensed operators involved in the event.
Distraction created by the degraded 2C FWIV.
Less than adequate Operations crew teamwork performance during the evolution.
Lack of Unit 2 specific SG level control simulator modeling. The Unit 1 and Unit 2 SGs are different models and have different level setpoints and level control. The site simulator is modeled after the Unit 1 configuration. Consequently, Operators do not receive simulator experience controlling the Unit 2 SG level.
B. Safety Significance
There were no safety consequences impacting plant or public safety as a result of this event. The reactor trip system, P-14, and the AF system functioned as designed. A risk analysis also indicates this event was not risk significant.
C. Corrective Actions
The Unit startup procedure will be revised to have the FRVs in automatic control during the opening of the FWIVs.
Other shutdown and startup procedures will be reviewed against Braidwood Station's corresponding procedures to ensure the sites are aligned to the best practices.
The priority of repair for the 2C FWIV will be reassessed.
Performance weaknesses with the operators involved have been remediated.
To correct the crew teamwork performance, enhanced pre-planned organizational responses are being developed and will be deployed when operating personnel performance issues warrant intervention.
The simulator's design will be reviewed to determine if the Unit 2 SG level control should be modeled.
E.
Previous Occurrences
LER 455 2010-001-00, "Reactor Protection and Auxiliary Feedwater System Actuation Signals from Low Steam Generator Level Due to Inadequate Surveillance Testing," dated April 19, 2010 Some personnel performance issues in this event were similar to this previous event. Beyond crew remediation for this event, management oversight and corrective actions to address crew team future high standards in individual and crew performance
~RC FORM 366A (10-2010)