05000266/LER-1999-005, :on 990514,steam Leak from Low Pressure FW Heater Was Noted.Caused by Rupture & Walls Thinnings Which Accelerated Corrosion.Repaired FW Heaters
| ML20195J761 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 06/11/1999 |
| From: | Krause C WISCONSIN ELECTRIC POWER CO. |
| To: | |
| Shared Package | |
| ML20195J753 | List: |
| References | |
| LER-99-005, LER-99-5, NUDOCS 9906210116 | |
| Download: ML20195J761 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(ii) 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(iv), System Actuation |
| 2661999005R00 - NRC Website | |
text
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPPSVED CY OMB NO. 3160-0104 (495)
- EXPlRES 04/30/98 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50 0 LICENSEE EVENT REPORT (LER)
NCORPO TED NTO THE ICENSING SS A FED BACK TO INDUSTRY. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE (See reverse for required number of INFORMATION AND RECORDS MANAGEMENT BRANCH digits / characters for each block)
(T-6 F33), U S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-000 t.
AND TO THE PAPERWORK REDUCTION PROJECT I FACILITY NAME (1)
DOCKET NUMBER (2)
PAGE (3)
Point Beach Nuclear Plant, Unit 1 05000266 1 of 5 I TITLE (4)
Steam Leak From Low Pressure Feedwater Heater l
EVENT DATE (6) l LER NUMBER (6)
REPORT DATE (7)
OTHER FACILITIES INVOLVED (8) l MONTHIYEAR SEQUENTIAL REVISION FACILn Y NAME DOCKET NUMBER DAY YEAR NUMBER NUMBE MONTH DAY YEAR 05000 l
1 l 05 l
FACILITY NAME DOCKET NUMBER l
14 1999l1999 -
005 -
00 06 11 1999 05000 j
OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR 6: (Check one or rnore) (11) l MODE (9)
N 20.2201(b) 2c.2203(a)(2)(v) 50.73(a)(2)(i) 50 73(a)(2)(vm)
POWER 20.2203(a)(1) 20.2203(a)(3)(i) 50.73(a)(2)(ii) 50 73(a)(2)(x)
LEVEL (10) 100 20.2203(a)(2)(i) 20 2203(a)(3)(n) 50.73(a)(2)(m) 73 71
)
20.2203(a)(2)(li) 20.2203(a)(4)
X 50.73(a)(2)(iv)
OTHER 20.2203(a)(2)(m) 50.36(c)(1) 50 73(a)(2)(v)
Spech m Abstract beh 20.2203(a)(2)(iv) 50.36(c)(2) 50.73(a)(2)(vu) or m NRC Form 366A LICENSEE CONTACT FOR THl3 LER (12) lNAME TELEPHONE NUMBER (include Area Code)
Charles Wm. Krause, Senior Regulatory Compliance Engineer (920) 755-6809 CCMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT {13)
CAUSE
SYSlEM COMPONENT MANUFACTURER REPORTABLE
CAUSE
SYSTEM COMPONENT MANUFACTURER REPORTABL TO EPlX B
E TO EPIX,
SUPPLEMENTAL REPORT EXPECTED (14)
EXPECTED MONTH DAY YEAR I
YES SUBMISSION (if yes, complete EXPECTED SUBMISSION DATE)
X NO DATE (16)
ABSTRACT (Limit to 1400 spaces. I e., approximately 15 single-spaced typewnttem lines) (16) on May 14, 1999, with the Unit 1 operating at full power, Point Beach Nuclear Plant (PBNP) experienced a steam leak from the rupture of the shell side of the #4B feedwater heater.
In response to this incident, the Unit 1 reactor was manually tripped and the steam to the 04B feedwater heater was secured.
In response to a decreasing pressurizer level, a manual safety injection signal was initiated.
The pressurizer level was restored with no need for actual ECCS injection.
Plant systems operated as anticipated in response to the incident. There were no personnel injuries or significant incidental equipment damage.
An inspection of the Unit 1 #4 feedwater heaters revealed shell wall thinning in the vicinity of the inlet nozzle deflector plates.
The apparent cause of the rupture and wall thinning was a combinat. ion of steam impingement and flow ac:elerated corrosion.
Inspection of the Unit 2 heaters revealed no comparable wall thinning.
Following repair of the feedwater heaters, Unit 1 was returned to operation after nine days.
9906210116 990611 PDR ADOCK 05000266 s
PM NRC FORM Soo (4-95)
NRC FoRQ 364A U.S. NUCLEAR REGULQToRY CoMMIS$loN 2
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LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
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YEAR SEQUENTIAL REVISlo Point Beach Nuclear Plant, Unit 1 05000266 NUMBER N
2 OF s NUMBER 1999 - 005 00 TEXT (11more space as reqwred, use additional copres of NRC Form 366A) (17)
Event Description
On May 14, 1999, at approximately 0711 (all times are CDT) with the Unit 1 operating at full power, Point Beach Nuclear Plant (PBNP) experienced a steam leak from the rupture of the shell side of #4B feedwater heater (HX-20B).
The steam leak was detected by a loud noise and steam throughout the Unit 1 turbine hall.
A manual reactor trip was initiated at 0713 after receiving a report that the steam was originating from the area around the Unit 1 feedwater heaters.
Emergency operating procedures for a reactor trip (EOP-0) and a secondary coolant leak (AOP-2A) were entered.
At 0716 a manual safety injection signal was initiated due to low level (less than 12% and decreasing) in the pressurizer.
At 0719 the pressurizer level was observed to be recovering with pressurizer pressure at 1841 psig. Safety injection was terminated at 0735 with no actual injection into the reactor coolant system (RCS) since the lowest RCS pressure did not drop below the 1550 psig shutoff head of the safety injection pumps.
At 0729 the main steam stop valves were shut and steam was removed fre' the turbine hall.
The leaking feedwater heater was isolated by shutting a manual isolation valve.
Plant components and systems required to function as a result of the reactor trip and manual sdfety injection operated as designed and no problems were encountered while responding to the steam leak event.
There were no personnel injuries as a result of the steam leak and damage to adjacent equipment was minima}.
In accord ance with 10 CFR 50.72 (b) (2) (ii), a four hour ENS notification was made to the NRC at 0813 as a result of the manual reactor protection system and engineered safety feature actuation.
Following an inspection and repairs to the Unit 1 fourth stage feedwater heaters (discussed further under Cause and Corrective Actions), Unit 1 was returned to service on May 22, 1999.
Component and System Description:
The Condensate and Feedwater System at PBNP consists of two condensate pumps, two feedwater pumps, four stages of low pressure and one stage of high pressure feedwater heaters, and associated valves, piping, and instrumentation arranged in two parallel trains.
The function of the feedwater heaters is to preheat the condensate in successive stages prior to returning the water to the steam generators and thereby increase the plant efficiency.
Steam from five extraction points in the turbine casings is piped to the shells of the two parallel strings of feedwater heaters.
The first point extraction originates from the 4'h stage of the high pressure turbine and 6
supplies steam to the #5 high pressure feedwater heaters.
The extraction point for the
- 4 feedwater heaters (HX-20 A and B) originates in the high pressure turbine exhaust nozzles.
The third, fourth and fifth extraction points originate in the low pressure turbine casing and supply the other three stages of low pressure feedwater heaters.
The fourth stage feedwater heaters, HX-20 A and B, receive drainage from the #5 high pressure feedwater heaters along with the extraction steam.
HX-20 A and B are horizontal half-size shell and "U" tube type heat exchangers. Each heater consists of a hemispherical channelhead welded to the tubesheet, which is welded to the shell. The shell is fabricated of carbon steel with ends closed by the tubesheet and a disned head.
The channelhead contains the inlet and outlet nozzles and is partitioned to direct the condensate flow through the stainless steel "U" tubes in 2 passes.
The extraction steam and drains from the #5 feedwater heaters enter the #4 feedwater NAC FORM 366A( 95)
a e
NRC FC.G; 344A U.S. NUCLEAR REGULATORY CosMISSloN (445)
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YEAR SEQUENTIAL REVISIO Point Beach Nuclear plant, Unit 1 05000266 NUMBER N
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~TExi pr more space os required, use additional coptes of NRc Fo,m 366A) (17) heaters through the top of the shell via separate nozzles, pass around stainless steel impingement plates which protect the tubes from erosion, and are directed among the tubes by transverse baffles which are positioned to provide maximum steam contact with the tube surfaces. As heat is transferred to the condensate flowing through the tubes, the steam is condensed and the drainage flows through the drain outlet to the heater drain tank.
Before reaching HX-20 A or B, the extr ction steam passes through preseparator tanks (T-94 A or B) where a portion of the moisture is removed from the extraction steam flow.
The shell design pressure and temperature of HX-20 A and B is 175 psig and 400*F respectively.
At full power operation, the actual steam pressure is approximately 125 psig and the temperature is about 350'F.
Cause
The source of the steam leak described in this event report was a rupture of the shell of HX-20B, the #4B feedwater heater.
The rupture was approximately 27 inches long and k inch wide.
The rupture occurred at the 11 o' clock position on the shell below and in line with the point at which the piping from the extraction steam enters the shell.
The nominal wall thickness of the shell is 0.5 inches.
The wall thickness at the point of rupture was 0.05 inches.
As of the date of this report, the root cause of the thinning has not been conclusively established; however, it appears that the apparent cause of the shell plate thinning was a combination of steam impingement and flow accelerated corrosion. A post event examination of the shell plate revealed that the thinning is in-line with the internal baffle plates which prevent direct impingement of the steam on the heater tubes.
Examinations of the shell wall on the opposite side of HX-20B and the comparable locations on HX-20A revealed similar wall thinning down to approximately 0.1 inches.
On May 15 power was reduced on Point Beach Unit 2 to allow ultrasonic NDE inspection of th that unit's 4 stage feedwater heaters.
We observed minimal shell wall thinning compared to that experienced on Unit 1.
After a brief shutdown to recover the feedwater heaters, Unit 2 was returned to full power.
Corrective Actions
Repairs have been completed on both Unit 1 #4 feedwater heaters.
The repairs consisted of mapping the thinned sections of the heat exchanger shell, cutting out those sections and welding in 5/9 inch thick replacement steel plates to restore the heat exchanger shells to a nominal 0.5 inches wall thickness. While the plate sections were removed, the tube bundle skid bars which facilitate removal of the heat exchanger tube bundle were modified to remove the possible steam flow dams which may have contributed to the erosion damage.
The plate section containing the rupture location has been sent to a metallurgica examination laboratory for an assessment of the failure mechanism.
e A investigation team was establisheci to perform a root cause evaluation of this equipment failure.
The final results of this evaluation were not available at the time of this LER submittal.
New informa tion or corrective actions which are-k as
NRC FORA 364A O.S. NUCLEAR REGULATORY CoMMISSloN (4-05)
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YEAR SEQUENTIAL REVISIO Point Beach Nuclear Plant, Unit 1 05000266 NUMBER N
4 OF s NUMBER 1999 - 005 00 TEXT (It more space os required, use additionalcopies of NRc Form 366A) (1T}
identified by this evaluation and which are significantly different than the preliminary information presented in this report will be provided in a supplement to this report.
- The extraction steam piping to the feedwater heaters, including the nozzles to the heater shells, have been included in the PBNP Flow Accelerated Corrosion (FAC)
Monitoring Program.
In response to this event, ths #4 feedwater heater shells will be added to the licensees FAC Monitoring Program.
- As discussed previously, the Unit 2 #4 feedwater heater shells were inspected for similar wall thinning with no significant thinning identified.
l Safety Assessa at:
The failure of the feedwater heater shell and the resulting steam leak had the potential to cause serious injury to the plant staff and significant equipment damage.
Fortunately, although the feedwater heater is located adjacent to a high traffic area, there were no personnel in the vicinity of the feedwater heater at the time of the rupture and no one was injured.
The damage to adjacent equipment was minimal and did not challenge the capability of the staff to contain and recover from the incident.,.
The response of the plant operating staff to this equipment failure was immediate and correct.
The reactor was manually tripped and the steam leak promptly isolated.
The response of the plant equipment to this transient, including the responses to the manual reactor trip and the manual actuation of safety injection, was appropriate.
'!Ae reactor coolant system was stabilized following the reactor trip and maintained in a hot standby condition during the duration of the feedwater heaters inspections and repairs.
The recovery operations and repairs went smoothly and the unit was returned to service after nine days.
There was no impact on the haalth and safety of the public as a result of this event.
Systc~. and Component Identifiers:
The Energy Industry Identification System component function identifier for each component / system referred to in this report are as follows:
Component / System Identifier t
heater, Feedwater HX Baffle BAF Turbine TRB Heater Drain Tank TK Safety Injection System BQ Condensate System SG Feedwater System SJ Main Turbine System TA LP Heater Drains and Vents System SM Engineered Safety Features Actuation System JE NRC f ORM 366A (4 96)
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YEAR SF.QUENTIAL REVISIO Point Beach Nuclear Plant, Unit 1 05000266 NUMBER N
5 OF 5 NUMBER 1999 - 005 00 TEXT (It more space sa required, use additionalcopies of NRC Form 366N (17)
Similar Occurrences:
A review of recent LERs (past two years) identified no other events reportable due to a j
manual reactor trip or reports resulting from significant steam leaks.
The following events involved inadvertent actuation of engineered safety features:
LER NUMBER Title 1
266/98-024-00 Inadvertent Emergency Diesel Generator Start 266/98-014-00 ESF Actuation, Automatic Start of Service Water Pump 266/98-006-00 Unanticipated Partial Service Water System Isolation 265/98-001-00 Failure of High Voltage Station Auxiliary Transformer 266/97-034-00 Unplanned Loss of Voltage on Train B Safeguards Buses i