05000366/LER-2009-003, Regarding Main Generator Runback Due to High Stator Water Cooling Water Temperature Results in Reactor Scram
| ML092230148 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 08/10/2009 |
| From: | Madison D Southern Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-09-1222 LER 09-003-00 | |
| Download: ML092230148 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 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)(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 |
| 3662009003R00 - NRC Website | |
text
Dennis R. Madison Southern Nuclear Vice President - Hatch Operating Company, Inc.
Plant Edwin I. Hatch 11028 Hatch Parkway North Baxley, Georgia 31513 Tel 912537.5859 Fax 912.366.2077 SOUTHERN.\\
COMPANY August 10, 2009 Docket No.:
50-366 1\\1L-09-1222 U. S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant Licensee Event Report Main Generator Runback Due to High Stator Water Cooling Water Temperature Results in Reactor Scram Ladies and Gentlemen:
In accordance with the requirements of 10 CFR 50.73(a)(2)(iv)(A), Southern Nuclear Operating Company is submitting the enclosed Licensee Event Report (LER) concerning a reactor scram resulting from high stator water cooling temperature.
This letter contains no NRC commitments. If you have any questions, please advise.
Sincerely,
~~Y77~':
D. R. Madison Vice President - Hatch DRM/MJK/
Enclosure: LER 2-2009-003 cc:
Southern Nuclear Operating Company Mr. J. T. Gasser, Executive Vice President Ms. P. M. Marino, Vice President - Engineering RTYPE: CHA02.004 U. S. Nuclear Regulatory Commission Mr. L. A. Reyes, Regional Administrator Mr. R. E. Martin, NRR Project Manager - Hatch Mr. J. A. Hickey, Senior Resident Inspector - Hatch
06 NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION 9-2007)
LICENSEE EVENT REPORT (LER)
- 1. FACILITY NAME Edwin I. Hatch Nuclear Plant Unit 2
~. TITLE Main Generator Runback Due to High Stator Water Cooling Water Temperature Results in Reactor Scram
- 6. LER NUMBER SEOUENTIAL
- 5. EVENT DATE REV YEAR YEAR MONTH DAY NUMBER NO.
0 20 2009 2009 - 003
- 7. REPORT DATE MONTH DAY YEAR 10 2009 08 MANU REPORTABLE SYSTEM COMPONENT
CAUSE
CAUSE FACTURER TO EPIX y
B TJ TCV F130 B
- 14. SUPPLEMENTAL REPORT EXPECTED o YES (If yes, complete 15. EXPECTED SUBMISSION DATE) 181 NO APPROVED BY OMB: NO. 3150-0104 EXPIRES: 08/31/2010
, the NRC may not conduct or sponsor, and a person is not reqUired to respond to, the information collection.
- 2. DOCKET NUMBER
- 13. PAGE 05000366 1 OF 4
- 8. OTHER FACILITIES INVOLVED FACILITY NAME DOCKET NUMBER 05000 DOCKET NUMBER FACILITY NAME 05000
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply)
~. OPERATING MODE o 20.2201 (b) o 20.2203(a)(3)(i) o 50.73(a)(2)(i)(C) o 50.73(a)(2)(vii) o 20.2201 (d) o 20.2203(a)(3)(ii) o 50.73(a)(2)(ii)(A) o 50.73(a)(2)(viii)(A) 1 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) o 20.2203(a)(2)(ii) o 50.36(c)(1 )(ii)(A) 181 50.73(a)(2)(iv)(A) o 50.73(a)(2)(x)
- 10. POWER LEVEL o 20.2203(a)(2)(iii) o 50.36(c)(2) o 50.73(a)(2)(v)(A) o 73.71 (a)(4) o 20.2203(a)(2)(iv) o 50.46(a)(3)(ii) o 50.73(a)(2)(v)(B) o 73.71(a)(5) 100 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 I~ELEPHONE NUMBER (Include Area Code)
Edwin I. Hatch / Steve Tipps, Principal Licensing Engineer 912-537-5880 MANU REPORTABLE FACTURER TO EPIX y
F130 MONTH DAY YEAR SYSTEM COMPONENT TJ 23
- 15. EXPECTED SUBMISSION DATE ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
On June 20,2009 at 14:17 EDT, Unit 2 was in mode 1 with an approximate reactor power of 2804 CMWTh. At this time protective circuitry initiated a main generator runback due to a high Stator Water Cooling (SWC) outlet temperature. The electro-hydraulic control (EHC) Mark VI cores processor began reducing turbine load by closing the turbine control valves (TCV) and subsequently opening the turbine bypass valves (TBV). Operations personnel began reducing reactor load in response to the main generator runback by reducing recirculation flow. The reactor power was reduced to approximately 66 percent reactor thermal power and could not be reduced any lower without the insertion of control rods. Due to turbine load reducing faster than the manual actions taken to reduce reactor load, reactor pressure began to increase. Pressure had reached the setpoint to actuate the Reactor Protection System, and a reactor scram was initiated due to high reactor pressure.
This event was caused by the improper set-up of the SWC temperature control instrument loop during 2R20 outage.
Corrective actions consist of: set-up of SWC valve 2N43F100 was corrected and the control system was calibrated, review of other work during the refueling outage was performed to ensure a similar condition does not exist on other equipment, and review of the Unit 1 SWC system was performed to confirm proper operation.
PRINTED ON RECYCLED PAPER NRC FORM 366 (9-2007)
(If more space is required, use additional copies of NRC Form 366A)
PLANT AND SYSTEM IDENTIFICATION
General Electric - Boiling Water Reactor Energy Industry Identification System codes appear in the text as (EllS Code XX).
DESCRIPTION OF EVENT
On June 20,2009 at 14:17 EDT, Unit 2 was in mode 1 with an approximate reactor power of 2804 CMWTh. At this time the control room received a 'GEN INLET TEMP HIGH' alarm followed within minutes by a 'GEN PROTECTION CIRCUIT ENERGIZED' alarm. This protective circuitry initiated a main generator (EllS Code TB) runback due to a high Stator Water Cooling (SWC, EllS Code TJ) outlet temperature. The electro-hydraulic control, Mark VI system (EHC, EllS Code TG) cores processor began reducing turbine load by closing the turbine control valves (TCV, EllS Code TA) and subsequently opening the turbine bypass valves (TBV, EllS Code TA). Operations personnel began reducing reactor load in response to the main generator runback by reducing recirculation flow (EllS Code AD). The reactor power was reduced to approximately 66 percent reactor thermal power and could not be reduced any lower without the insertion of control rods (EllS Code AA). Due to turbine load reducing faster than the manual actions taken to reduce reactor load, reactor pressure began to increase. Reactor pressure increased to a high of 1074 psjg and the Reactor protection System (RPS) (EllS Code IG), initiated a reactor scram due to high reactor pressure.
Reactor water level decreased resulting in a primary containment valve Group 2 (EllS Code JM) valve isolation per design. Safety Relief Valves (EllS Code SB) did not actuate nor were they required to based on the maximum reactor pressure reached.
CAUSE OF EVENT
The direct cause of this event was improper set-up of a valve controller.
During the investigation of the 'GEN INLET TEMP HIGH' alarm, it was found that the generator SWC heat exchangers were on full bypass when they should have been providing flow through the heat exchanger given the current system operating temperature. When the valve was manually stroked by maintenance, it was noticed that the piping temperatures down stream of the valve immediately dropped. From review of the trend data taken by rounds it was determined that the valve was not maintaining the SWC temperature by design. Therefore, the cause of this event was failure of valve 2N43F1 00 (SWC Temperature Controller) to properly control the SWC temperature.
The SWC temperature instrument loop consists of (2N43F100) a Fisher model 667-Y AOV three way valve with a model 3582 pneumatic valve positioner. A Fisher 4160B temperature controller feeds the instrument air signal to the above positioner and controls the SWC temperature by its established set points. The function of this instrument loop is to control the inlet SWC temperature within a given operating range. The valve controls this temperature by opening a bypass around the SWC heat exchangers when SWC inlet temperatures beQin to drop or vice-a-versa when the temperature increases. This function PRINTED ON RECYCLED PAPER(9-2007)
LICENSEE EVENT REPORT (LER) u.s. NUCLEAR REGULATORY COMMISSION CONTINUATION SHEET
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE Edwin 1. Hatch Nuclear Plant Unit 2 05000366 I
SEQUENTIAL
/REVISION NUMBER NUMBER YEAR 3
OF 4
2009 003 o
was lost when maintenance was performed on the instrument loop during the 2R20 refuel outage. The positioner and the controller were not set so that the valve would function per design. Both the positioner for valve 2N43F100 and its temperature controller 2N43R310, proportional band were set up as "Direct" acting. In this configuration, the valve responded opposite of its design. Therefore; due to seasonal increases in river temperature, the temperature control set points were reached and the SWC flow was automatically placed on full bypass. Removal of system heat exchangers from the system caused a rapid heat build up in the stator cooling water since the generator was online. When the system went on bypass, the generator stator bar temperatures increased to a point where the generator runback set point was reached.
REPORTABILITY ANALYSIS AND SAFETY ASSESSMENT This event is reportable per 10 CFR 50.73(a)(2)(iv)(A) because unplanned actuations of a safety feature system listed in 10 CFR 50.73 occurred. In this instance, a reactor protection system (RPS) actuation resulting in a reactor scram.
An increase in reactor vessel pressure during reactor operation compresses the steam voids and results in a positive reactivity insertion. This causes the neutron flux and thermal power transferred to the reactor coolant to increase which could challenge the integrity of the fuel cladding and the reactor coolant pressure boundary. Therefore, the reactor is shut down automatically on high reactor vessel steam dome pressure to limit the neutron flux and thermal power increase. The automatic reactor shutdown on high pressure, along with the SRVs, limits the peak reactor vessel pressure to less than the American Society of Mechanical EngineersSection III Code limits.
In this event, protective circuitry initiated a main generator runback due to a high SWC outlet temperature. EHC, Mark VI, system began reducing turbine load by closing the turbine control valves (TCV) and subsequently opening the turbine bypass valves. Operations personnel began reducing reactor load in response to the main generator runback by reducing recirculation flow. The reactor power was reduced to approximately 66 percent reactor thermal power and could not be reduced any lower without the insertion of control rods. Due to turbine load reducing faster than the manual actions taken to reduce reactor load, reactor pressure began to increase. Pressure increased to a high of 1074 psig and the RPS initiated a reactor scram due to high reactor pressure. Reactor water level decreased resulting in primary containment Group 2 valve isolation per design. No SRVs opened nor were any required to open to limit or reduce reactor vessel pressure. No Emergency Core Cooling Systems actuated nor were any reqUired to actuate to recover or maintain water level during or following this event. All automatic functions operated per design in response to the pressure increase and the automatic reactor shutdown.
Based on this analysis, it is concluded that this event had no adverse impact on nuclear safety.
PRINTED ON RECYCLED PAPER (9*2007)
LICENSEE EVENT REPORT (LER) u.s. NUCLEAR REGULATORY COMMISSION CONTINUATION SHEET
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE Edwin I. Hatch Nuclear Plant Unit 2 05000366 YEAR I SEQUENTIAL IREVISION NUMBER NUMBER 4
OF 4
2009 003 0
CORRECTIVE ACTIONS
The set-up of SWC valve 2N43F100 was corrected and the control system was calibrated.
An instrument loop check on the valve was performed and it was confirmed that the valve will operate as designed on increase and decrease in temperature.
A review of work performed during the recent refuel outage was performed to determine if there was a potential for a similar configuration to exist on another valve. It was determined that no other vulnerabilities of this nature exist.
Plant personnel confirmed that the Unit 1 SWC instrument loop is set up so it will function as designed.
ADDITIONAL INFORMATION
Other Systems Affected: None
Failed Components Information
Master Parts List Number: 2N43F100 Manufacturer: Fisher Controls (F130)
Model Number: 3582 Type: Valve Positioner
Failed Components Information
Master Parts List Number: 2N43R310 Manufacturer: Fisher Controls (F130)
Model Number: 4160B Type: Temperature Controller EllS System Code: TJ Reportable to EPIX: Yes Root Cause Code: B EllS Component Code: TCV EllS System Code: TJ Reportable to EPIX: Yes Root Cause Code: B EllS Component Code: 23 Commitment Information:
This report does not create any new permanent licensing commitments.
Previous Similar Events
There are no similar events within the past two years in which a valve control system was set-up improperly that resulted in a reactor scram.
PRINTED ON RECYCLED PAPER