05000354/LER-2005-002, Regarding Through-Wall Leak on B Reactor Recirculation System Decontamination Port
| ML051540027 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 05/26/2005 |
| From: | Massaro M Public Service Enterprise Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LR-N05-0261 LER 05-002-00 | |
| Download: ML051540027 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(x) |
| 3542005002R00 - NRC Website | |
text
PSEG Nudear LC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236
§ PSEG NuclearLLC LR-N05-0261 MAY 2 6 2005 U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 LER 35412005-002-00 HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NO. NPF-57 DOCKET NO. 50-354 This Licensee Event Report entitled, 'Through-Wall Leak on 'B' Reactor Recirculation System Decontamination Port," is being submitted pursuant to the requirement of I OCFR50.73(a)(2)(ii)(A).
Sincerely, Michael assaro Plant Manager - Hope Creek Attachment BJT C
Distribution LER File 3.7 95-2168 REV. 7/99
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSIO APPROVED BY OMB: NO. 3150-0104 EXPIRES: 06130/2007 6-2004)
, the NRC may digts/harctes or achblok)not conducto sponsor, and a person Is not required to respond to, the digits/characters for each block)
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- 3. PAGE Hope Creek Generating Station 05000354 1 OF 4 4 TITLE Through-Wall Leak on 'B' Reactor Recirculation System Decontamination Port
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILMES INVOLVED FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER NOEV MONTH DAY YEAR FACILrTY NAME DOCKET NUMBER 03 27 2005 2005 - 002 -
00 05 26 2005 9.OPERATING MODE
- 11.
THIS REPO{T IS SUBMrrED PURSUANT TO THE REQUIREMENTS OF10 CFR§: (Check al that epply) 0 20.2201(b) 0 20.2203(a)(3Xi) 0 50.73(aX2)(i)(C) 0 50.73(a)(2Xv A) 2 0 20.2201(d) 0 20.2203(a)(3XXi) 0D 50.73(aX2)(iiXA) 0 50.73(a)(2XvIiIXA) o 20.2203(aX1) 0 20.2203(a)(4) 0 50.73(aX2)(iiXB) 0 50.73(a)(2)(viii)(B) 0 20.2203(aX2)(1) 0 50.36(c)(1XIXA) 0 50.73(aX2)(iii) 0 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL 0 20.2203(aX2)(ii) 0 50.36(cXl XiiXA) 0 50.73(aX2)(ivXA) 0 50.73(a)(2)(x)
O 20.2203(aX2)(iii) 0 50.36(c)(2) 0 50.73(aX2)(v)(A) 0 73.71(a)(4) o 20.2203(aX2)(iv) 0 50.46(aX3)(ii) 0 50.73(aX2)(v)(1) 0 73.71(a)(5) 6 0 20.2203(aX2)(v) 0 50.73(aX2XI)(A) 0 50.73(aX2)(vXC) 0 OTHER 0 20.2203(aX2)(vi) 0 50.73(aX2XI)(B) 0 50.73(aX2)(v)(D)
Specify In Abstract below nr in NRC. Fnwm.WJA
- 12. LICENSEE CONTACT FOR THIS LER FACILiTY NAME TELEPHONE NUMBER (incude Aea Code)
Brian Thomas, Licensing Engineer 856-339-2022CAUSE SYSTEM COMPONENT MANU-REPORTABLE
CAUSE
SYSTEM COMPONENT MANU-REPORTABLE FACTURER TO EPIX FACTURER TO EPIX X
=
Y
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED MONTH DAY YEAR SUBMISSION o YES (If yes, complete 15. EXPECTED SUBMISSION DATE) 0 NO DATE ABSTRACT (Umit to 1400 spaces, I.e., approximately 15 single-spaced typewritten lines)
On March 27, 2005, reactor power was reduced to approximately 6% power (Operational Condition 2) to perform a primary containment entry to identify the source of increasing unidentified reactor coolant system (RCS) leakage.
Beginning on February 8, 2005, unidentified RCS leakage had been trending upward and was at approximately 0.73 gpm on March 27, 2005. Hope Creek personnel identified a steam leak coming from the insulated decontamination port that is connected to the 'B' reactor recirculation system between the suction of the pump and the suction isolation valve. At 2329 hours0.027 days <br />0.647 hours <br />0.00385 weeks <br />8.861845e-4 months <br /> on March 27, 2005, the reactor mode switch was taken to shutdown to perform a controlled shutdown and cooldown of the plant to precisely identify and repair the leakage source. At 1035 hours0.012 days <br />0.288 hours <br />0.00171 weeks <br />3.938175e-4 months <br /> on March 28, 2005, after removal of the insulation from the decontamination port, the steam leakage was determined to be coming from a crack located on the 4-inch nominal pipe size (NPS) portion of the decontamination port.
The through-wall crack in the decontamination connection was determined to be fatigue initiation and propagation as caused by the following: presence of an original subsurface weld defect and presence of a decontamination port geometry that experienced damaging high cycle vibration during certain recirculation system operating conditions.
Corrective actions consisted of the modification of the A and B reactor recirculation decontamination ports and an extent of condition review to determine if other connections to the reactor recirculation system were vulnerable to the decontamination port failure.
This event is being reported in accordance with 1 OCFR50.73(a)(2)(ii)(A), 'any event or condition that resulted in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded."
NRC FORM 366 (6-201)U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
FACILITY NAME (1)
DOCKET (2)
LER NUMBER (6)
PAGE (3)
SEQUENTIAL I REVISION YEAR NUMBER NUMBER Hope Creek Generating Station 05000354 2 OF 4 l
2005 002 00 TEXT (If more space Is required, use additional copies of NRC Form 366A) (17)
PLANT AND SYSTEM IDENTIFICATION
General Electric - Boiling Water Reactor (BWR/4)
Reactor Recirculation System {AD)*
- Energy Industry Identification System {EIIS} codes and component function identifier codes appear as {SS/CCC}
IDENTIFICATION OF OCCURRENCE Event Date: March 27, 2005 Discovery Date: March 27, 2005 CONDITIONS PRIOR TO OCCURRENCE A voluntary controlled plant power reduction was performed to determine the source of increasing reactor coolant system (RCS) {AB) unidentified leakage. On March 27, 2005, while in Operational Condition 2 with reactor power at approximately 6%, a primary containment entry identified a steam leak from the insulated decontamination port that is sealed by a bolted flange within the isolable boundary of the 'B' recirculation system. Hope Creek then proceeded to cold shutdown to precisely identify and repair the leakage source. With the plant in Operational Condition 3 (HOT SHUTDOWN), a through-wall leak at the welded junction of the 4-inch decontamination port and the 'B' reactor recirculation system suction piping was identified. Unidentified reactor coolant system leakage was at approximately 0.73 gpm prior to the plant shutdown, which was well below the Technical Specification limit of 5 gpm. There was no equipment out of service that impacted this event.
DESCRIPTION OF OCCURRENCE On March 27, 2005, reactor power was reduced to approximately 6% power (Operational Condition 2) to perform a primary containment entry to Identify the source of Increasing unidentified reactor coolant system (RCS) {AB) leakage.
Beginning on February 8, 2005, unidentified RCS leakage had been trending upward and was at approximately 0.73 gpm on March 27, 2005. Hope Creek personnel identified a steam leak coming from the insulated decontamination port that is connected to the 'B' reactor recirculation system {AD} between the suction of the pump and the suction isolation valve. At 2329 hours0.027 days <br />0.647 hours <br />0.00385 weeks <br />8.861845e-4 months <br /> on March 27, 2005, the reactor mode switch was taken to shutdown to perform a controlled shutdown and cooldown of the plant to precisely identify and repair the leakage source.
At 1035 hours0.012 days <br />0.288 hours <br />0.00171 weeks <br />3.938175e-4 months <br /> on March 28, 2005 while in Operational Condition 3 after removal of the insulation from the decontamination port, the steam leakage was determined to be coming from a crack located on the 4-inch NPS portion of the decontamination port. The decontamination port is 4-inch diameter SA-376 Type 304 stainless steel Schedule 80 ASME Code Class 1 piping.
Hope Creek entered cold shutdown at 1452 hours0.0168 days <br />0.403 hours <br />0.0024 weeks <br />5.52486e-4 months <br /> on March 28, 2005. The RCS unidentified leakage was well below the Technical Specification limit of 5 gpm.
This event is being reported in accordance with 1 OCFR50.73(a)(2)(ii)(A), 'any event or condition that resulted in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded."U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
FACILITY NAME (1)
DOCKET (2)
LER NUMBER (6)
PAGE (3)
SEQUENTLAL REVISION YEAR NUMBER INUMBER Hope Creek Generating Station 05000354 3OF4 l_
II 2005 002 00 CAUSE OF OCCURRENCE The cause of the through-wall crack in the decontamination connection was fatigue initiation and propagation as a result of the following:
original subsurface weld defect, and presence of a decontamination port geometry that experienced damaging high cycle vibration during certain recirculation system operating conditions.
The fatigue crack was initiated at a subsurface (near the outside diameter (OD) surface) weld defect in combination with the presence of the decontamination spool geometry that experienced damaging high cycle vibration during certain recirculation system operating conditions.
The metallurgical examinations identified fatigue initiation and propagation. The as-found 'B' decontamination port contained an approximately 4-inch long circumferential crack. A significant portion of the fracture surface was covered with a heavy oxide scale suggesting that this portion existed for an extended period. A minor portion of the fracture surface nearest the inside surface was much less oxidized suggesting a more recent fracture that was responsible for the through-wall leak.
Analysis revealed that the natural frequencies of the 'B' decontamination port were coincident with the 5X vane passing frequencies of the reactor recirculation pump at normal operating speeds. The natural frequencies of the decontamination port ranged from 122 to 135 Hz. The 5X vane passing frequency of the reactor recirculation pump at normal operating speeds is 120 to 125 Hz.
An analysis of pump operating data indicated that on February 8, 2005, the pump experienced a transient 5X vibration acceleration at a speed of 1501 rpm that was high in amplitude and short in duration, apparently due to the convergence of the 5X vane passing frequency with acoustic resonance of the recirculation piping system. An analysis of the drywell floor leakage demonstrated an increasing trend In unidentified RCS leakage following this event.
PREVIOUS OCCURRENCES
A review of LERs at Salem and Hope Creek generating stations for the previous two years was performed to identify prior similar occurrences. LER 354/04-010-00 identified an October 10, 2004, pipe failure of the A moisture separator drain line that led to a manual reactor scram. Although the October 10, 2004, event identifies a pipe failure, the mechanism of this pipe failure was different from the current event. The October 10, 2004 pipe rupture was the result of fatigue failure due to 25 days of high dynamic loading caused by the flow through a failed open valve. The corrective actions associated with the October 10, 2004 event would not have prevented this current event from occurring.
I I
ILU.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
FACILITY NAME (1)
DOCKET (2)
LER NUMBER (6)
PAGE (3)
SEQUENTIAL REVISION lYEAR NUMBER INUMBER Hope Creek Generating Station 05000354 l 4OF4 2005 002 00
SAFETY CONSEQUENCES AND IMPLICATIONS
An analysis of pump operating data indicated that on February 8, 2005, the 'B' recirculation pump experienced a transient 5X vibration acceleration at a speed of 1501 rpm that was high in amplitude and short in duration. An analysis of the drywell floor leakage demonstrated an increasing trend in unidentified RCS leakage following this event. After the February 8, 2005, transient 5X vibration acceleration, an administrative limit for the 'B' recirculation pump speed was put in place to prevent entering this region.
As the result of monitoring the increasing trend in RCS unidentified leakage, Hope Creek performed a controlled power reduction of the plant on March 27, 2005, to perform a primary containment entry and identified the source of the leakage as the insulated 'B' decontamination port. Hope Creek then proceeded to cold shutdown to precisely identify and repair the leakage source. After removal of the insulation from the decontamination port, with the plant in Operational Condition 3 (hot shutdown), a through-wall leak at the welded junction of the 4-inch decontamination port and the 'B' reactor recirculation system suction piping was identified. The 'B' decontamination port is located between the suction of the 'B' recirculation pump and the suction isolation valve. Therefore, in the unlikely event that the crack would have propagated, the decontamination port could have been isolated by the closure of the 'B' recirculation loop suction and discharge isolation valves. RCS unidentified leakage was approximately 0.73 gpm, which was well below the Technical Specification limit of 5 gpm. The crack on the decontamination port was repaired prior to restart of the plant.
Based on the above, there was no impact to the health and safety of the public.
A review of this event determined that a Safety System Functional Failure (SSFF) has not occurred as defined in Nuclear Energy Institute (NEI) 99-02. Hope Creek was brought to a controlled safe cold shutdown condition following the identification of the 'B' decontamination port through-wall leak.
CORRECTIVE ACTION
- 1. The decontamination ports for both the A and B recirculation loops were modified. The 4-inch decontamination port was shorted to increase the natural frequencies so they are not coincidental with the normal recirculation pump vane passing frequencies. These modifications were completed prior to the plant startup.
- 2. A review of connections to the reactor recirculation loops was performed to determine if other similar connections would be susceptible to the failure experienced with the decontamination port. This review consisted of finite element analysis, vibration analysls, modal analysis, isometric review and previous ISI inspections. A total of twenty-three (23) welds were inspected during the plant shutdown using a combination of nondestructive examination (NDE) techniques. These NDE inspections were satisfactory.
COMMITMENTS
This LER contains no commitments.