ML061370622
| ML061370622 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 05/11/2006 |
| From: | Balduzzi M Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 2.06.034 LER 06-01-000 | |
| Download: ML061370622 (8) | |
Text
s'Entergy Entergy Nuclear Operations, Inc.
Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth, MA 02360 Michael A. Balduzzi Site Vice President May 11, 2006 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555
SUBJECT:
Entergy Nuclear Operations, Inc.
Pilgrim Nuclear Power Station Docket No: 50-293 License No. DPR-35 Licensee Event Report 2006-001-00 LETTER NUMBER: 2.06.034
Dear Sir or Madam:
The enclosed Licensee Event Report (LER) 2006-001-00, "Manual Scram due to High Offgas Recombiner Temperature Resulting from Inadequate Preventive Maintenance of Recombiner Preheater Pressure Control Valve Controller," is submitted in accordance with 10 CFR 50.73.
This letter contains no commitments.
Please do not hesitate to contact Bryan Ford, (508) 830-8403, if there are any questions regarding this subject.
Sincerely, M.A. Balduzzi DWE/dm Enclosure cc: Mr. James Shea, Project Manager Mr. Samuel J. Collins Plant Licensing Branch 1-1 Regional Administrator, Region 1 Division of Operator Reactor Licensing U.S. Nuclear Regulator Commission Office of Nuclear Reactor Regulation 475 Allendale Road U.S. Nuclear Regulatory Commission King of Prussia, PA 19406 One White Flint North O-8C2 11555 Rockville Pike Rockville, MD 20852 INPO Records Senior Resident Inspector 700 Galleria Parkway Pilgrim Nuclear Power Station Atlanta, GA 30399-5957
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- 1. NRC Form 366 U.S. NUCLEAR REGULATORY COMMISSION Estimated burden per response to comply with this mandatory information collection request: 50 hrs. Reported lessons learned are Incorporated Into the licensing process and fed back to industry. Forward comments regarding burden LICENSEE EVENT REPORT (LER) I estimate to the Records Management Branch (T-6 F33), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and to the Paperwork Reduction Project (3150-0104), Office of Management and Budget, Washington. DC 20503.
FACILITY NAME (1) lDOCKET NUMBER (2) l PAGE(3)
PILGRIM NUCLEAR POWER STATION 05000-293 1of 7 TITLE (4)
Manual Scram due to High Offgas Recombiner Temperature resulting from Inadequate Preventive Maintenance of Recombiner Preheater Pressure Control Valve Controller EVENT DATE (5)l LER NUMBER (6) REPORT DAE(7 OTHER FACILITIES INVOLVED (8)
SEQUENTIAL NUMBER l REVISION FACILrIY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER MONTH DAY YEAR N/A 05000 03 13 2006 I
_ l 2006 001 00 05 11 2006 l FACILITY NAME N/A DOCKET NUMBER 05000 OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR: (Check one or more) (11)
MODE (9) N 20.2201(b) 22.2203(a)(3)(i) 50.73(a)(2)(§)(C) l 50.73(a)(2)(vii)
POWER 22.2202(d) _ 20.2203(a)(3)(ii) 50.73(a)(2)(il)(A) 50.73(a)(2)(viis)(A)
LEVEL (10) 43 20.2203(a)(1) 20.2203(a)(4) 50.73(a)(2)(ii)(B) 50.73(a)(2)(vlii)(B) 20.2203(a)(2)(i) _ 0.36(3)(1)(l)(A) 50.73(a)(2)(i11) l 50.73(a)(2)(ix)(A) 20.2203(a)(2)(i) _ 50.36(3)(1)(i)(A) X 50.73(a)(2)(iv)(A) I 50.73(a)(2)(x) 20.2203(a)(2)(lii 1 50.36(c)(2) 50.73(a)(2)(v)(A) I 73.71(a)(4) 20.2203(a)(2)(iv) 50.46(a)(3)(1) 50.73(a)(2)(v)(B) 73.71(a)(5) 20.2203(a)(2)(v) _ 50.73(a)(2)(i)(A) _ 50.73(a)(2)(v)(C) OTHER 20.2203(a)(2)(0) _ 50.73(a)(2)(i)(B) 1 50.73(a)(2)(v)(D) Specify in Abstract below
____m _ _ __ _ _I _ _Lor In NRC Form 366A LICENSEE CONTACT FOR THIS LER (12)
NAME TELEPHONE NUMBER (Include Area Code)
Bryan Ford, Licensing Manager 1 (508) 830-8403 (Umit to 1400 spaces, i.e., approxmately 15 single-spaced typewritten lines) (18)
ABSTRACT On March 13, 2006, at 1808 hours0.0209 days <br />0.502 hours <br />0.00299 weeks <br />6.87944e-4 months <br />, an unplanned scram was manually initiated while at 43 percent reactor power. This procedurally required action was taken because of high temperature in the Augmented Offgas (AOG) system recombiner. There were no complications encountered during the response to the scram.
The root cause of the event was the failure of the controller of a pressure control valve (PCV) that supplies steam dilution flow PCV and steam heating to the recombiner's preheater. The failure was due to inadequate preventive maintenance of the controller. Corrective action taken included replacement of the pressure control valve controller and repair of a hand operated valve that enables steam to bypass the pressure control valve.
Corrective action planned includes developing preventive maintenance for the AOG pressure controllers, training, and evaluating changes to procedures The event occurred during power operation. The reactor vessel pressure was about 967 psig with the reactor vessel water temperature at the saturation temperature for that pressure. The event posed no threat to public health and safety.
NRC FORM 366
NRC Form 366A U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1) DOCKET NUMBER 2 LER NUMBER 6 PAGE (3)
. l SEQUENTIAL l REVISION YEAR NUMBER I NUMBER PILGRIM NUCLEAR POWER STATION 05000-293 2006 001 00 2 of 7 TEXT (If more space Is required, use additional copies of NRC Form 366A) (17)
BACKGROUND The main condenser gas removal system functions to remove air, non-condensible gases, and water vapor from the main condenser. The non-condensible gases and water vapor are discharged from the after-condenser for subsequent processing in the Offgas system that includes a hold-up line.
The Augmented Offgas (AOG) system provides additional treatment. The Offsite Dose Calculation Manual (ODCM) requires the AOG system be in service at greater than 50 percent reactor power with allowances to remove the system from service. When the AOG system is in service, the offgas upstream of the Offgas system hold-up line is directed to the AOG system. The AOG system includes steam jet compressors, pre-heaters, catalytic recombiners, after-condensers and water separators, hydrogen analyzers, cooler condensers, moisture separators, pre-filters, charcoal absorbers, and radiation monitors. After the offgases are processed in the AOG system, non-condensible gases from the AOG system are directed to the Offgas system hold-up line.
The principal mechanical components of the AOG system are connected in a series-parallel arrangement that consists of two trains. The system's two in-parallel recombiners are located downstream of the system's two in-parallel preheaters and upstream of the system's two in-parallel offgas condensers. A recombiner functions to recombine oxygen and hydrogen formed through the radiolytic decomposition of water exposed to radiation in the reactor vessel. A preheater functions to preheat diluted offgas before entry into the recombiner.
The diluted offgas piping is equipped with a valve for pressure relief and instrumentation for sensing and indicating temperature and pressure. Steam for heating the diluted offgas is piped to the preheater. The piping for heating is equipped with instrumentation including those for pressure sensing, pressure indication, pressure alarm, and control of the steam pressure supplied to the preheater. The offgas heated in the preheater is piped to the respective recombiner. A recombiner is equipped with a catalyst and instrumentation. The instrumentation functions to sense the pressure and temperature of the offgas from the preheater, the temperature of the recombiner vessel, and temperature of the offgas exiting the recombiner.
As planned, Pilgrim Station was removed from electrical power generation service on March 11, 2006, for the planned reconnection of the Unit Auxiliary Transformer (UAT). While the turbine-generator was removed from electrical power generation service, the reactor was maintained in a critical condition, at about 14 percent power, with steam directed to the main condenser via the turbine steam bypass valves.
After reconnection of the UAT, and as reactor power was gradually increased, Pilgrim Station returned to electrical power generation service at 1832 hours0.0212 days <br />0.509 hours <br />0.00303 weeks <br />6.97076e-4 months <br /> on March 11 t. Routine plant operations continued for the planned return to full reactor power operation, including the following activities.
At 0438 hours0.00507 days <br />0.122 hours <br />7.242063e-4 weeks <br />1.66659e-4 months <br /> on March 2l the AOG system was put into service. As part of this activity, the hand operated valve (1-HO-154) that enables the bypass of steam around the AOG pressure valve PV-9239 would not open.
NRC FORM 366A
~
NRC~~~~~~ ~ ~ ~ ~ Fom36 .. NCER EUAOYCMISO NRC Form 366A U.S. NUCLEAR REGULATORY COMMISSION FACILITY NAME (1)
LICENSEE EVENT REPOR,T (LER)
TEXT CONTINUATION DOCKET NUMBER (2 l
LER NUMBER (6)
SEQUENTIAL REVISION PAG 1
YEAR NUMBER NUMBER PILGRIM NUCLEAR POWER STATION 05000-293 2006 001 00 3c TEXT (If more space Is required, use additional copies of NRC Form 366A) (17)
At about 1610 hours0.0186 days <br />0.447 hours <br />0.00266 weeks <br />6.12605e-4 months <br /> on March 13th, a reactor operator noted the AOG preheater steam heating pressure was about 280 psi, with the pressure valve PV-9239 in manual control. Downstream of PV-9239, the AOG dilution steam was being supplied through a hand operated valve that enables steam bypass around pressure valve PV-9238. Reactor power was about 82 percent, and reactor power was being increased gradually, at about 15 MW thermal per hour. The operations crew anticipated that as reactor power increased the inlet pressure to PV-9239 would increase.
At 1745 hours0.0202 days <br />0.485 hours <br />0.00289 weeks <br />6.639725e-4 months <br /> on March 13 , at about 83 percent reactor power, a Control Room alarm occurred indicating low steam dilution flow supplied to the AOG system. As part of a pre-evolution briefing, two non-licensed operators were directed to adjust the steam pressure at PV-9239. The briefing included the possibility of bypassing PV-9239 in accordance with the AOG system operating procedure.
By 1753 hours0.0203 days <br />0.487 hours <br />0.0029 weeks <br />6.670165e-4 months <br />, communications at PV-9239 were established and initial attempts were made to throttle open PV-9239. One of the operators reported the valve did not open when an increased pressure demand signal to the controller of PV-9239 was attempted twice. On the third attempt the controller of PV-9239 indicated the valve was closing. Control Room personnel observed the steam pressure from PV-9239 steadily decreasing, and the operators were directed to open hand operated valve 1-HO-1 54 to enable steam bypass around PV-9239. The bypass valve could not be opened. At 1757 hours0.0203 days <br />0.488 hours <br />0.00291 weeks <br />6.685385e-4 months <br />, the vapor valves, closed automatically due to low steam dilution flow. The low dilution trip was bypassed and the vapor valves were reopened. At 1803 hours0.0209 days <br />0.501 hours <br />0.00298 weeks <br />6.860415e-4 months <br />, a Control Room alarm occurred indicating an AOG recombiner high temperature condition of greater than about 875 degrees F and the procedure for abnormal recombiner temperature was entered.
At 1805 hours0.0209 days <br />0.501 hours <br />0.00298 weeks <br />6.868025e-4 months <br />, the licensed shift control room supervisor directed reactor power be decreased in accordance with the power reduction procedure and reactor power was being decreased by the adjustment of reactor recirculation flow and insertion of selected control rods. Meanwhile, additional operator attempts were made to adjust the steam pressure from PV-9239. The pressure controller of PV-9239 was placed in the automatic mode and then back to the manual mode several times in the attempt to open PV-9239.
This action ultimately resulted in the return of the pressure controller demand and position signals and the controller began to control PV-9239. AOG steam dilution flow was restored and the pressure from PV-9239 was observed to increase and steady at about 350 psig. Despite having restored normal AOG recombiner parameters, the recombiner vessel bottom temperature increased to a reported value of 1074 degrees F. After verifying the recombiner temperatures were >1000 degrees F and in accordance with approved procedure, the shift senior licensee operator directed a manual scram of the reactor.
EVENT DESCRIPTION On March 13, 2006, at about 1808 hours0.0209 days <br />0.502 hours <br />0.00299 weeks <br />6.87944e-4 months <br />, a manual actuation of the reactor protection system (RPS) and reactor scram occurred while at about 43 percent reactor power. The RPS actuation was the expected response to intentionally depressing the manual scram pushbuttons of RPS Channels 'A' and 'B', and was taken in accordance with procedure because the AOG recombiner temperature was greater than 1000 degrees F.
NRC FORM 366A
NRC Form 366A U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE 3 SEQUENTIAL l REVISION YEAR NUMBER I NUMBER PILGRIM NUCLEAR POWER STATION 05000-293 2001 00 4 of 7 TEXT (If more space Is required, use additional copies of NRC Form 366A) (17)
The reactor vessel water level decreased to about -21 inches (narrow range level) due to the decrease in the void fraction in the reactor vessel water. Plant systems responded to the event as designed including an automatic actuation of the RPS, Primary Containment Isolation Control System (PCIS) and Reactor Building Isolation Control System (RBIS) due to the decrease in reactor water level.
PCIS actuation resulted in the automatic closing of the Primary Containment system (PCS) Group 2 and Group 6/Reactor Water Cleanup (RWCU) system isolation valves that were open. The normally closed PCS Group 3/residual heat removal system valves remained closed.
The RBIS actuation resulted in the automatic start of the Standby Gas Treatment system (SGTS) trains 'A' and 'B', and automatic closing of the Reactor Building/Secondary Containment system (SCS) trains 'A' and 'B' supply and exhaust ventilation dampers.
At about 1856 hours0.0215 days <br />0.516 hours <br />0.00307 weeks <br />7.06208e-4 months <br />, the RPS was reset and the RBIS was reset at about 1859 hours0.0215 days <br />0.516 hours <br />0.00307 weeks <br />7.073495e-4 months <br /> and subsequently, the SGTS was returned to normal standby service, and the reactor building ventilation system was returned to service.
CAUSE The direct cause of the scram signal that initiated the scram was intentionally depressing the manual scram pushbuttons of RPS Channels 'A' and 'B' while at about 43 percent reactor power. This action was taken in accordance with procedure because the AOG system recombiner temperature was greater than 1000 degrees Fahrenheit.
The root cause of the event was failure of PIC-9239 due to inadequate preventive maintenance. Relay K-1 exhibited high contact resistance. This relay functions to determine which signal, manual or automatic, is sent to the operational amplifier of the pressure indicating controller PIC-9239. PIC-9239 functions to control pressure valve PV-9239 that was supplying steam to the AOG preheater and steam dilution flow PCV at the time of the event. The controller was manufactured by the Foxboro Corporation, model 62-HB-4E-OH.
A factor contributing to the cause of the event was no recognition of a degraded material condition or vulnerability of the AOG system to a single component failure (e.g. PIC-9239 or PV-9239).
An additional factor contributing to the event was inadequate communication. A corrective action was not initiated, i.e. a corrective action program document or maintenance document, as a result of not being able to open a hand operated valve (1-HO-1 54) on or about January 13, 2006. The hand operated valve enables a steam bypass pathway around pressure valve PV-9239. At that time the operator who attempted to open 1-HO-1 54 mistakenly believed the valve had been successfully opened later in the day on January 13, 2006 and did not initiate a corrective action document (i.e. condition report). This deficient condition was again experienced on March l2 th when the AOG system was being put into service. A condition report was initiated prior to the scram, but not until early March 13. Upon the failure of PIC-9239 on March 13t, valve 1-HO-1 54 was unable to be opened and the recombiner temperature consequently exceeded 1000 degrees F.
NRC FORM 366A
NRC Form 366A U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1) DOCKET NUMBER 2 LER NUMBER (6 PAGE 3 SEQUENTIAL REVISION YEAR NUMBER NUMBER PILGRIM NUCLEAR POWER STATION 05000-293 2006 001 00 5 of 7 TEXT (If more space Is required, use additional copies of NRC Form 366A) (17)
CORRECTIVE ACTION Corrective actions taken included the following:
- The 300# pressure indicating controller installed at the time of the event for PV-9239 (Steam heating to Recombiner Preheater) was removed from service and bench tested. Bench testing revealed unacceptable contact resistance in the K-1 relay. A replacement pressure indicating controller (different model number) was obtained from stores, calibrated, bench tested and installed in the field. After installation of the replacement pressure indicating controller, PV-9239 was manually operated and observed to operate as designed.
. The 40# pressure indicating controller installed at the time of the event for PV-9238 (Dilution Steam to Vapor Valves) was removed from service and bench tested. Bench testing revealed unacceptable contact resistance in the K-1 relay. A replacement pressure indicating controller was obtained from stores, calibrated, bench tested, and installed in the field. After installation of the replacement controller, PV-9238 was manually operated and was observed to operate as designed.
- Hand operated valve 1-HO-1 54 was repaired.
- Night orders were issued to reinforce management expectations for timeliness of initiating corrective action documents for deficient conditions.
Corrective actions planned include the following:
- Develop and implement preventive maintenance for PIC-9239 and PIC-9238. Capture lessons learned from this event in the PM.
- Develop a forced outage work package for replacement of the K-1 relay installed in PIC-9238 and PIC-9239.
- Complete single point vulnerability review, review Scram Frequency Reduction Committee (SRFC) items, and perform a design comparison review for the AOG system.
- Develop a case study from this event and provide training to operations, maintenance, and technical programs personnel.
- Evaluate procedure revisions to allow steam jet air ejector, 300# and 40# regulators with their respective bypass valves throttled to auctioneer steam supplies following the single point vulnerability review.
- Evaluate the AOG system operating procedure to allow bypassing AOG at rated power.
- Perform applicability review for other controllers during development of PMs.
These planned corrective actions may be supplemented or modified in accordance with the corrective action process.
NRC FORM 366A
NRC Form 366A U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1) DOCKET NUMBER 2 LER NUMBER (6) PAGE 3 SEQUENTIAL REVISION YEAR NUMBER NUMBER PILGRIM NUCLEAR POWER STATION 05000-293 2006 001 00 6 of 7 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
SAFETY CONSEQUENCES The Offgas system and AOG system function to process non-condensable offgas resulting from the nuclear fuel fission process. After processing, gases from the Offgas/AOG system are directed to the main stack for release to the environment. The systems each contain separate radiation monitors that monitor the activity of the gases to be released into the atmosphere, and automatically initiate isolation signals to the main stack isolation valve if predetermined setpoints are achieved. The offgases being processed at the time of the event did not exceed the setpoints that would have required the isolation of the main stack isolation valve.
An engineering evaluation was conducted of event parameters related to the AOG system recombiner.
The evaluation determined there was no damage to the recombiner vessel.
The decrease in the reactor vessel water level was the expected response to the scram and accompanying shrink in the water. The consequent RPS, PCIS and RBIS actuations were the expected designed responses to a low reactor vessel water level condition, i.e. less than about +12 inches (narrow range).
During the event, the lowest reactor vessel water level that occurred was approximately -21 inches (narrow range). This level is significantly greater than the core standby cooling systems set point (approximately -
46 inches), and the level corresponding to the top of the active fuel zone (approximately -127 inches).
All safety systems performed as designed. This event did not result in a challenge to fuel limits or release of radioactive material above the expected normal operating level. The event posed no threat to public health and safety.
REPORTABILITY This report was submitted in accordance with 10 CFR 50.73(a)(2)(iv) because the actuation of the RPS and consequent actuations of the PCIS and RBIS, although a designed response to intentionally depressing the manual scram pushbuttons of RPS Channels 'A' and 'B' while at about 43 percent reactor power, were not planned.
SIMILARITY TO PREVIOUS EVENTS A review was conducted of Pilgrim Station LERs submitted since 2004. The review focused on similarities involving an automatic or manual scram(s) and degrading main condenser vacuum conditions. The review identified no similar events.
NRC FORM 366A
NRC Form 366A U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) 1 I SEQUENTIAL l REVISION YEAR I NUMBER I NUMBER PILGRIM NUCLEAR POWER STATION 05000-293 2006 001 00 7 of 7 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
ENERGY INDUSTRY IDENTIFICATION SYSTEM (EIIS) CODES The EIIS codes for this report are as follows:
COMPONENTS CODES Control, indicating, pressure (PIC-9239) PIC Recombiner RCB Relay (K-1) RLY Valve (PV-9239, 1-HO-154) V SYSTEMS CODES Condenser system SC Condenser vacuum system (MCGRS) SH Containment isolation control system (PCIS/RBIS) JM Engineered safety features actuation system (RPS, PCIS, RBIS) JE Offgas system (Offgas, AOG) WF Plant protection system (RPS) JC Reactor building (SCS) NG Reactor recirculation system AD Reactor water cleanup system (RWCU) CE Standby gas treatment system (SGTS) BH NRC FORM 366A