05000339/LER-2002-001
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded |
| 3392002001R00 - NRC Website | |
DOCKET FACILITY NAME (1) LER NUMBER (6) NORTH ANNA POWER STATION 05000 - 339 2002 � --001 -- � 00 2 � of 11 1.0 DESCRIPTION OF THE EVENT On September 8, 2002, North Anna Unit 2 was shutdown for a scheduled refueling outage. During the outage, a qualified, barehead visual inspection of the reactor vessel head (EIIS-RPV) and penetrations (EIIS-PEN) for evidence of leakage was performed as required by NRC Bulletin 2001-01, Circumferential Cracking of Reactor Pressure Vessel Head Penetration Nozzles. On September 14, 2002, with Unit 2 in Mode 6, the qualified, barehead visual inspection on penetrations 21 and 31 identified through-wall leakage based on the presence of boric acid deposited on the reactor head at these penetrations.
Additional inspections identified four more penetrations that were suspected of leaking and several penetrations that were masked with boric acid residue due to a conoseal leak that made their status indeterminate.
Based on the apparent through-wall leak, a non-emergency 8-hour notification was made to the NRC, at 2214 hours0.0256 days <br />0.615 hours <br />0.00366 weeks <br />8.42427e-4 months <br />, on September 14, 2002, in accordance with 10CFR50.72(b)(3)(ii)(A), any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded. This event is reportable per 10 CFR 50.73(a)(2)(ii)(A), for any event or condition that resulted in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.
In addition, Technical Specification (TS) 3.4.6.2 prohibits reactor coolant system (RCS) (EIIS-AB) pressure boundary leakage in Modes 1 through 4. Although the apparent leakage was identified in Mode 6, it is reasonable to assume that the leakage occurred during Modes 1 through 4. Therefore, this event is also reportable in accordance with 10CFR50.73(a)(2)(i)(B) for a condition prohibited by TS.
In addition to the qualified, visual barehead inspection that was performed, volumetric inspections were performed as required by NRC Bulletin 2002-02, Reactor Pressure Vessel Head And Vessel Head Penetration Nozzle Inspection Program. Nondestructive examination techniques performed included eddy current testing (ET), ultrasonic testing (UT), and dye penetrant testing (PT).
A summary of nondestructive examination (NDE) inspections performed on the reactor vessel head penetration tubes and J-groove welds during the North Anna 2 refueling outage is provided below.
A remote ET examination of the J-groove weld surface was performed on 59 penetrations.
The remote ET examination identified crack-like indications on 57 of the 59 penetrations inspected. The remaining six (6) penetrations were manually inspected by PT examination of the J-groove weld surface. These included four (4) thermocouple penetrations (numbers N2-51, 53, 55 and 57) that were inaccessible for remote inspection LER NUMBER (6)
DOCKET
05000 - 339 FACILITY NAME (1)
NORTH ANNA POWER STATION
3 � of 11 and two (2) penetrations (numbers N2-62 and 63) that were previously repaired by weld overlay of the J-groove weld. Thermocouple penetration N2-51 was also repaired by weld overlay of the J-groove weld during the Fall 2001 outage. This penetration had boric acid identified during the barehead inspection and was suspected to be leaking. Each of the six penetrations inspected by manual PT examination of the J-groove weld surface had evidence of rejectable indications.
Based on the large number of penetrations with J-groove weld surface indications and/or boric acid deposits found during the barehead visual inspection (from leakage, suspected leakage, or masking from a conoseal leak), a conservative decision was made to remove thermal sleeves and fully inspect each suspect penetration tube in the area of interest using an open housing scanner. A best effort inspection using the blade probe technique was not considered to be adequate for these penetrations. Therefore, inspection using an ET blade probe was only performed on a total of six (6) penetrations (numbers N2-1, 17, 22, 23, 24 and 25). Of these six penetrations, numbers N2-1, 22, 23, 24 and 25 are part- length control rod penetrations that were initially planned to be inspected with the open housing scanner since the part-length drive shaft and control mechanism was previously removed. It was discovered during the inspection process that a part-length thermal sleeve remained inside the penetration housing that prevented performing an open housing inspection. As a result, an inspection using an ET blade probe was performed on the part-length penetrations.
Thermal sleeves were removed on a total of 31 penetrations to facilitate open housing inspection using a combination of ET and UT inspection techniques. Since the 4 thermocouple penetrations do not have thermal sleeves, the open housing scanner was used to inspect a total of 35 penetrations.
A detailed summary of NDE inspection results is provided below.
J-Groove Weld and Penetration Tube OD Surface Eddy Current Examinations Grooveman ET examinations were conducted on fifty-nine CRDM penetration J-groove welds and on the outside diameter (OD) surfaces of forty-seven penetration tubes. These examinations were performed to identify the presence of primary water stress corrosion cracking on the outside diameter surfaces of the penetrations and on the surface of the J- groove welds attaching the penetrations to the reactor vessel head. The following table provides a summary of all Grooveman ET reactor vessel head penetration nondestructive inspections performed at North Anna Unit 2 during the refueling outage.
U.S. NUCLEAR REGULATORY COMMISSION FACILITY NAME (1)
NORTH ANNA POWER STATION
DOCKET
05000 - 339 LER NUMBER 6) PAGE (3) 4 � of 11 (I YEAR � I � SEQUENTIAL � REVISION Eddy Current � J-Groove Weld � Eddy Current Penetration # � J-Groove Weld � Coverage � Penetration Tube OD 1 Reportable (RI) 360° � No Detectable Indications (NDI) 2 Reportable (RI) 360° � No Detectable Indications (NDI) 3 Reportable (RI) 360° � No Detectable Indications (NDI) 4 Reportable (RI) 360° � No Detectable Indications (NDI) 5 Reportable (RI) 360° � No Detectable Indications (NDI) 6 Reportable (RI) 360° � No Detectable Indications (NDI) 7 Recordable (NRI) 360° � No Detectable Indications (NDI) 8 Reportable (RI) 360° � Recordable (NRI) 9 Reportable (RI) 360° � No Detectable Indications (NDI) 10 Recordable (NRI) 360° � Not Inspected 11 Recordable (NRI) 360° � Recordable (NRI) 12 Recordable (NRI) 360° � No Detectable Indications (NDI) 13 Reportable (RI) 360° � No Detectable Indications (NDI) 15 Reportable (RI) 360° � No Detectable Indications (NDI) 17 No Detectable Indications (NDI) 360° � No Detectable Indications (NDI) 19 Reportable (RI) 360° � No Detectable Indications (NDI) 21 Reportable (RI) 360° � No Detectable Indications (NDI) 22 Reportable (RI) 360° � No Detectable Indications (NDI) 23 Recordable (NRI) 360° � No Detectable Indications (NDI) 24 No Detectable Indications (NDI) 360° � Not Inspected 25 Reportable (RI) 360° � No Detectable Indications (NDI) 26 Recordable (NRI) 360° � No Detectable Indications (NDI) 27 Reportable (RI) 360° � No Detectable Indications (NDI) 28 Recordable (NRI) 360° � No Detectable Indications (NDI) 29 Reportable (RI) 360° � No Detectable Indications (NDI) 30 Recordable (NRI) 360° � No Detectable Indications (NDI) 31 Reportable (RI) 360° � No Detectable Indications (NDI) 32 Recordable (NRI) 360° � No Detectable Indications (NDI) 33 Recordable (NRI) 360° � Not Inspected 34 Recordable (NRI) 360° � No Detectable Indications (NDI) 35 Reportable (RI) 360° � Not Inspected 36 Reportable (RI) 360° � No Detectable Indications (NDI) 37 Recordable (NRI) 360° � No Detectable Indications (NDI) 38 Reportable (RI) 360° � No Detectable Indications (NDI) 39 Reportable (RI) 360° � No Detectable Indications (NDI) 40 Reportable (RI) 360° � No Detectable Indications (NDI) 41 Reportable (RI) 360° � Not Inspected 42 Reportable (RI) 360° � Not Inspected 43 Recordable (NRI) 360° � No Detectable Indications (NDI) 44 Reportable (RI) 360° � No Detectable Indications (NDI) FACILITY NAME (1)
NORTH ANNA POWER STATION
DOCKET
05000 - 339 LER NUMBER (6) PAGE (3) 5 of II Eddy Current Penetration # J-Groove Weld J-Groove Weld Coverage Eddy Current Penetration Tube OD 45 Reportable (RI) 360° No Detectable Indications (NDI) 46 Reportable (RI) 360° No Detectable Indications (NDI) 47 Recordable (NRI) 360° No Detectable Indications (NDI) 48 Reportable (RI) 360° No Detectable Indications (NDI) 49 Reportable (RI) 360° No Detectable Indications (NDI) 50 Reportable (RI) 360° No Detectable Indications (NDI) 51 PT Not Inspected 52 Reportable (RI) 360° No Detectable Indications (NDI) 53 PT Not Inspected 54 Reportable (RI) 360° No Detectable Indications (NDI) 55 PT Not Inspected 56 Reportable (RI) 360° No Detectable Indications (NDI) 57 PT Not Inspected 58 Reportable (RI) 360° No Detectable Indications (NDI) 59 Reportable (RI) 360° No Detectable Indications (NDI) 60 Reportable (RI) 360° No Detectable Indications (NDI) 61 Reportable (RI) 360° No Detectable Indications (NDI) 62 PT Not Inspected 63 PT Not Inspected 64 Recordable (NRI) 360° No Detectable Indications (NDI) 65 Reportable (RI) 360° No Detectable Indications (NDI) 66 Reportable (RI) 360° No Detectable Indications (NDI) 67 Reportable (RI) 360° No Detectable Indications (NDI) 68 Reportable (RI) 360° No Detectable Indications (NDI) 69 Reportable (RI) 360° No Detectable Indications (NDI) Forty-two J-groove welds were found to contain reportable indications (RI), fifteen contained recordable (non-reportable) indications (NRI) and two contained no detectable indications (NDI). Two penetration tubes contained recordable indications (NRI) and forty- five contained no detectable indications.
RI — Reportable Indication (> 9 mm in length) NRI — Recordable Indication (> 6 mm and NDI — No Detectable Indication ( PT — Penetrant Test All fifteen of the NRIs were considered crack-like in nature. Seven of the NRIs were actually conservatively classified as NRIs.
Open Housing Scanner Ultrasonic and Eddy Current Examinations An open housing scanner was used on 35 reactor vessel head penetrations to: 1) use time-of-flight diffraction (TOFD) to detect axial and circumferential reflectors on the penetration tube OD surfaces, 2) use a straight beam UT at 2.25 MHz, to interrogate the J-grove weld zone, and at 5.0 MHz to identify possible leak paths in the shrink fit region between the head penetrations and the reactor vessel head, and 3) perform eddy current examinations capable of detecting axial and circumferential degradation on the penetration tube ID surfaces. The following table provides a summary of the open housing scanner nondestructive inspections performed at North Anna Unit 2 during the refueling outage.
Penetration # Tube (Volumetric) J-Groove Weld Zone Shrinkfit Region Tube ID Surface Axial TOFD � Circ TOFD Channel 1 � Channel 2 2.25 Mhz 0° 5.0 Mhz 0° ET Results 10 CRDM NDD WII LOF LOF NDD 12 CRDM PTI-ID NDD NDD NDD 3 Axial 15 CRDM PTI/BBP/NDD NDD NDD NDD NDD 19 CRDM PTI/IPA/NDD NDD NDD NDD 1 Axial 21 CRDM NDD NDD PLP PLP NDD 31 CRDM NDD PTI/I PA/NDD PLP PLP NDD 35 CRDM PTI PTI WVI WVI 1 Axial 38 CRDM NDD WII/IPA/NDD NDD NDD NDD 40 CRDM NDD WII/IPA/NDD NDD NDD 2 Axial 41 CRDM PTI-ID PTI LOF LOF 6 Axial 43 CRDM NDD WII/IPA/NDD NDD NDD NDD 44 CRDM LCG PTI/W I I/NDD LOF LOF 3 Axial 46 CRDM NDD WII/BBP/NDD NDD NDD NDD 47 CRDM NDD WII/IPA/NDD NDD LOF NDD 48 CRDM LCG WII/IPA/NDD NDD LOF NDD 49 CRDM NDD NDD NDD NDD NDD 50 CRDM PTI/IPA/NDD PTI/I PA/NDD NDD NDD 6 Axial 51 T/C PTI-ID WII/IPA/NDD NDD PLP 9 Axial 52 CRDM PTI/IPA/NDD PTI/IPA/NDD LOF LOF 3 Axial 53 T/C PTI-ID NDD LOF NDD 3 Axial 54 CRDM NDD PTI NDD NDD 1Axial 55 T/C PTI-ID WII/IPA/NDD NDD NDD 3 Axial 56 CRDM NDD WII/IPA/NDD NDD NDD NDD DOCKET FACILITY NAME (1) (I � YEAR I SEQUENTIAL REVISION NORTH ANNA POWER STATION 05000 - 339 2002 � --001 -- � 00 7 � of 11 Tube (Volumetric) J-Groove Weld Zone Shrinkfit Region Tube ID Surface Axial TOFD Circ TOFD Penetration # Channel 1 Channel 2 2.25 Mhz 0° 5.0 Mhz 0° ET Results 57 T/C NDD NDD NDD NDD NDD 58 CRDM WII/IPA/NDD WPI/IPA/NDD NDD NDD NDD 59 CRDM PTI PTI NDD NDD 1 Axial 60 CRDM WII/IPA/NDD WII/IPA/NDD NDD NDD NDD 61 CRDM PTI-ID WII/IPA/NDD NDD NDD 5 Axial 62 CRDM PTI/IPA/NDD WII/IPA/NDD NDD NDD 19 Axial 63 CRDM PTI-ID NDD PLP PLP 2 Axial 64 CRDM NDD WII/IPA/NDD NDD NDD 14 Axial 65 CRDM PTI-ID W I I/PTI NDD NDD 9 Axial 66 CRDM PTI-ID PTI/IPA/NDD NDD NDD 9 Axial 67 CRDM PTI-ID PTI/W I I LOF NDD 1 Axial 68 CRDM NDD NDD LOF LOF NDD Legend:
NDD — No Detectable Defect IPA — Indication Profile Analysis Resolution of Indication WII — Weld Interface Indication WVI — Weld Volume Indication PTI — Parent Tube Indication LCS — Loss of Coupling-Scanner LCG — Loss of Coupling-Geometry LIF — Loss of Interference Fit LOF — Lack of fusion at the tube to weld interface BBP — B and B Prime Analysis Resolution VOL — Volumetric Indication PLP — Possible Leak Path Gapscanner Penetration Tube ID Surface Eddy Current Examinations Gapscanner eddy current examinations were conducted on the ID surfaces of six reactor vessel head penetration tubes. These examinations are capable of identifying the presence of primary water stress corrosion cracking (PWSCC) on the inside diameter surfaces of the penetration tubes. The following table provides a summary of all Gapscanner ET examinations performed at North Anna Unit 2 during the refueling outage.
Penetration # � ET Results 1 P/L � NDD 17 CRDM � NDD 22 P/L � 2 Axial 23 P/L � NDD 24 P/L � 2 Axial 25 P/L � 5 Axial Of the six penetrations inspected with the Gapscanner ET system, three showed axial indications on the ID surfaces and three showed no detectable degradation.
Discussion of Results Results from the Grooveman eddy current examinations of fifty-nine J-groove welds and forty-seven penetration tube OD surfaces showed forty-two J-groove welds with reportable indications (RI), fifteen with recordable indications (NRI) and two with no detectable indications (NDI). Indication orientations were found to be axial and circumferential with respect to the welding direction and ranged in length from 0.12" to about 7.0". In some cases, the longer flaws that are reported are actually a series of small flaws with very short distances in between.
Eddy current results from tube inside diameter surface examinations with the Westinghouse 7010 Open Housing Scanner examinations showed twenty of thirty-five penetration tubes had axial indications. These indications are believed to be less than 0.12" (3.0 mm) deep based on the PCS24 axial TOFD results. More accurate sizing would require additional time-of-flight interrogation with probes with smaller PCS spacings.
Therefore, refined depth sizing was not performed.
Time-of-flight ultrasonic examinations with the Westinghouse 7010 Open Housing Scanner showed a number of penetration tubes with indications. Indication profile analyses and "B and B-Prime" resolution analyses were performed in order to assess the significance of these reflectors. Four penetrations (#21, #31, #51, and #63) also showed evidence of a leak path in the shrink fit area between the vessel head and the tube.
Penetrations #51 and 63 were identified as leaking in Fall 2001. Repairs were determined to have been improperly applied because the weld overlay repair did not extend out far enough to cover the previous NDE indications.
The six penetration welds inspected by PT had rejectable (greater than 1/16" linear) indications.
Penetration tubes with results indicative of degradation are summarized below.
FACILITY NAME (1)
NORTH ANNA POWER STATION
DOCKET
05000 - 339 LER NUMBER (6) PAGE (3)
NUMBER
2002 --001 -- 00 Penetration Characteristics Length Depth #15 OD Circumferential 7.5 degrees to 12 degrees 0.226" #21 Potential Leak Path 220 degrees N/A* #35 OD Axial 0.80" 0.223" #41 OD Circumferential 357 degrees to 43 degrees 0.097" #46 OD Circumferential 4 degrees to 20 degrees 0.072" #51 Potential Leak Path 210 degrees to 260 degrees N/A* #54 OD Circumferential 119 degrees to 198 degrees 0.226" #54 OD Circumferential 344 degrees to 16 degrees 0.156" #59 OD Circumferential 347 degrees to 63 degrees 0.149" #59 OD Circumferential 156 degrees to 206 degrees 0.149" #63 Potential Leak Path 320 degrees to 0 degrees N/A* #65 OD Circumferential 330 degrees to 42 degrees 0.152" #65 OD Circumferential 160 degrees to 190 degrees 0.078" #67 OD Circumferential 343 degrees to 27 degrees 0.094" * Indications were not indicative of tube degradation.
2.0 SIGNIFICANT SAFETY CONSEQUENCES AND IMPLICATIONS Reactor Coolant System leakage, including unidentified leakage, is quantified every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> per TS 3.4.6.2. Prior to the Unit 2 shutdown to perform the reactor vessel head inspection, unidentified leakage was measured at 0.1321 gpm and containment sump inleakage was measured at 0.22 gpm.
No discernable head wastage was identified during the initial head inspection. The head was subsequently cleaned and re-inspected and the head was determined to be in good condition with no evidence of head wastage identified.
An evaluation was performed and determined that weld metal cracking by itself does not result in an immediate safety concern since cracking that is contained entirely within the weld metal, even if 360° around the nozzle, will not lead to nozzle ejection. The portion of the weld that is attached to the outside surface of the nozzle will not be able to pass through the tight annular fit. Additionally, the outward distortion in the penetration from weld shrinkage would further prevent the nozzle from passing through the tight annular fit i.e., the nozzle maintains its circumference, but becomes oval in shape at the weld, thus resisting ejection through the round penetration). Through-weld cracking to the annulus has the same consequence as a leaking nozzle and is evaluated as RCS pressure boundary leakage.
WCAP-14552, "Structural Evaluation of Reactor Vessel Upper Head Penetrations to Support Continued Operation: North Anna and Surry Units," was prepared to support determination of appropriate corrective actions. The WCAP documents that as much as 83.9% of the weld may be unfused, and the allowable stress limits can still be met. Even a complete lack of fusion in the zone between the weld and the head would not result in rod ejection because the weld to the tube would prevent it. Therefore, catastrophic failure of a penetration is unlikely. The health and safety of the public were not affected at any time during this event.
Also, a through-wall circumferential crack growth curve was developed using the same bases as those contained in WCAP-14552 except that the latest (i.e., more conservative) crack growth model recommended by the EPRI Material Reliability Program and stress intensities as developed by an NRC contractor were used. The curve shows that a circumferential, through-wall crack, if initiated early in the life of the plant, would not grow to 270 degrees, the criteria for determining instability of the tube, until the unit has operated for 37.5 years. The maximum circumferential flaw size found in North Anna Unit 2 was contained within a central angle of 79 degrees and was part through-wall.
However, it is estimated that this flaw would grow to through-wall in 1.8 years and would require another 24.2 years to reach the 270 degree stability limit. Therefore, the tube would remain stable for approximately an additional 26 operating years.
3.0 CAUSE The apparent cause of the reportable NDE indications was hot-short cracking, which occurred during original fabrication of the reactor vessel head. The hot-short cracking was due to the presence of low melting point contaminants in the weld metal. The reactor vessel head may not have been sufficiently cleaned prior to welding which would indicate a lapse in process control. Once cracking progressed through the J-groove welds, primary water stress corrosion cracking affected the penetration tubes.
A boat sample was obtained from penetration 62 during the Fall 2001 outage. Analysis of this boat sample confirmed that the indications were most likely associated with original fabrication.
A boat sample from the previous weld overlay repair of penetration 51 was removed and sent to Westinghouse for laboratory analysis. Analysis of the boat sample indicated that the weld overlay repair did not extend out far enough to cover previous NDE indications.
4.0 IMMEDIATE CORRECTIVE ACTION(S) Based on the qualified, visual barehead inspection results, additional NDE examinations were initiated to characterize the nature of the indications.
5.0 ADDITIONAL CORRECTIVE ACTIONS An evaluation using methodology obtained from WCAP-14552, (with supplemental stress intensity and crack growth rate curves) was performed of the NDE indications to DOCKET FACILITY NAME (1) LER NUMBER (6) NORTH ANNA POWER STATION 05000 - 339 2002 � --001 -- � 00 11 � of 11 determine the additional service life allowable before repair. The results indicate that 1.5 years of service could be achieved prior to the indication growing to 75% through-wall, 1.8 years before becoming completely through-wall, and 26.0 years remain before the flaw would become unstable.
A boat sample from the previous weld overlay repair of penetration 51 was removed and sent to Westinghouse for laboratory analysis. Analysis of the boat sample indicated that the weld overlay repair did not extend out far enough to cover previous NDE indications.
Additional information related to the structural integrity of the reactor pressure vessel head penetration nozzles, including the extent of the leakage and indications, and the inspections and repairs undertaken to satisfy regulatory requirements was provided in the written response to NRC Bulletin 2002-02.
A root cause evaluation of the event is being performed.
6.0 ACTIONS TO PREVENT RECURRENCE A management decision was made to replace the North Anna Unit 2 reactor head constructed of materials that are known to be more resistant to cracking rather than perform multiple penetration repairs.
7.0 SIMILAR EVENTS North Anna 2 reactor pressure vessel head penetrations identified three penetrations as being rejectable due to the existence of boric acid on the reactor pressure vessel barehead surface.
8.0 MANUFACTURER/MODEL NUMBER Rotterdam Dockyard Company/Serial Number 30662 9.0 ADDITIONAL INFORMATION None
LI