05000289/LER-2001-002
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| Initial Reporting | |
| 2892001002R00 - NRC Website | |
BACKGROUND
Thermocouple (TIC) Nozzles A thermocouple (T/C) nozzle is a one inch diameter schedule 160 pipe machined to a controlled diametrical fit with the bore in the reactor pressure vessel (RPV) head IRCTI. The nozzle material ;s SB-167 (alloy 600). A total of e'ght thermocouple *[THC] nozzles were installed in the RPV head. These nozzles are located outbcard of the RPV head's Control Rod Drive IOW] Mechanisms (CRDMs).
The original TIC nozzles were intended to provide instrumentation access into the vessel in order to verify that the internal reactor vessel plenum veni valves were not leaking. This was later determined to be unnecessary and blind flanges were added to the T/C nozzles that established the Reactor Cooiant System (RCS) *[AB] pressure b;untary. Two of the T/C nozzles were subsequently modified tc support the Reactor Coolant Inventory Tracking Systein (RUTS) and Reactor High Point Vent System. The other six T/C nozzles serve nn current function other than RCS pressure boundary.
A typical TIC nozzle head penetration consists of an approximate 1 inch outside diameter (OD) by 0.218 inch nominal wall alloy 600 pipe tha',: is inserted vertically into the RF/ head. The TIC pipe is connected to the inside surface of the RPV head by a J-groove partial penetration weld. The thermocouple nozzles have an overall length of approximately 62 inches. About 8 inches of each T/C nozzle extends past the J-groove weld on the inside surface of the RPV head.
Control Rod Drive Mechanism (CRDM) Nozzles There are 69 Control Rod Drive Mechanism (CRDM) IAA] nozzles that penetrate the RPV head. The CRDM nozzles are approximately 5 feet !cng and are welded to the RPV head at various radial locations from the centerline of the RPV head. The nozzles are constructed from 4 inch OD alloy 600 material. The lower end of the nozzle extends about 6 inches below the inside of the RPV head.
T.ER NUMBER (6) .AGE (3) DOCKET (2) Each CRDM nozzle was machined to final dimensions to assure a match between the RPV head bore and the OD of each nozzle. The nozzles were tack welded and then permanently welded to the RPV head usIng 182 weld metal. The final weld surface was ground and PT inspected.
NRC Bulletin 2001-01 Based on past RPV head penetration leakage found at the Oconee and ANO plants, the NRC issued Bulletin 2001-01 to solicit input from each Pressurized Water Reactor (PWR) to determine the various Utility response and actions to address this issue. TMI Unit-1 response stated that all RPV head penetrations would be visually inspected, and any "suspect" nozzles would be subsequently examined by liquid dye penetrant (PT) and Ultrasonic (UT) inspections. Any RPV head penetration determined to be leaking would be repaired.
In accordance with approved plant procedures, the initial results of the visual inspection classified the "as-found" condition of the RPV head penetrations into three categories:
1. Acceptable. Those in the Acceptable category showed no evidence of leakage at the base of the nozzle and the outer RPV baad surface.
2. Masked. This is an interim category. Those in the Masked category had loose debris or obstructions around the nozzle that prevented an entire 360-degree inspection. The obstruction or loose debris was vacuumed (while videotaping the area) to allow for complete inspection. Based on the results of leaking RPV nozzles at other stations, the boric acid residue associated with leaking penetrations is characterized as tightly adhering to the nozzle/head interface area.
Vacuuming would not remove this type of boric acid residue. After vacuuming, the nozzle was classified as either Acceptable or Suspect. Any nozzle that remained "masked" in the area of interest (annular gap) would be determined to be Suspect and subject to subsequent U"'.- and PT inspections.
3. Suspect. Those in the Suspect category showed signs of toric acid residue at the nozzle base.
11DOCKET (2) ..ER NUMBER (6) Leakage would be confirmed by additional PT and UT inspections.
The suspect CRDM locations were evaluated with visible dye penetrant (PT) method at the surface of the J-groove weld, the OD of the CRDM nozzle protruding into the RPV, and at the end of the CRDM nozzle. All suspect CRDMs had the drives removed and a top down inspection was performed utilizing the EPKI demonstrated Framatome ANP ultrasonic inspectic," equipment. The ultrasonic inspection consisted of two complete scans cf each suspect nozzle. One axial scan was used to identi::If circumferential flaws, and one circumferential scan identified any axial flaws.
Since no additional PT or UT inspections were planned for the TIC nozzles, any masked condition placed the T/C nozzle into the Suspect cs,:agory. Videotapes of previous inspections were available to identify any prior boric acid residue conditions.
Additionally, if any circumferential cracking was found above the J-groove weld in CRDMs, then the UT inspection would be expanded to include those CRDM nozzles that are made accessible (i.e. had CRDM motor tubes removed as a result of visual inspections or to permit repairs). Repairs would be made to any leaking RPV nozzle and any flavrt found not acceptable. These repairs would use Framatome ANP repair techniques similar to the repairs completed at the Oconee and Crystal River Units.
EVENT DESCRIPTION
On October 11/12, 2001 with Three Mile Island Nuclear Station (TMI-1) Unit 1 in refueling outage, 1R14, a periodic, qualified visual inspection of the top surface of the RPV head revealed boric acid deposited on the head surface. The RPV head inspection was performed in response to Generic Letter 88-5, "Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components lri PWR Plants", and NRC Bulletin 2001-01, " Circumferential Cracking of Reactor Pressure Vessel Head Penetration Nozzles".
DOCKET (2) PAGE (3) LER NUMBER (6) Boric acid deposits were located at the base of all eight TIC nozzles. After reviewing tapes of the last TIC nozzle inspection, all thermocouple nozzles were deemed to be leaking (since no additional PT or UT inspections were planned for the T/C nozzles). At approximately 1230 on October 12, 2001, Engineering evaluation of the thermocouple nozzle visual inspection results confirmed that the boron deposits around the eight T/C nozzles indicate a RCS pressure boundary leak. The T/C nozzle leakage was reported [EN # 38383] as a non-emergency [6-hour] in accordance with 10 CFR 50.72(b)(3)(ii)(A).
The initial visual inspection of the CRDM nozzles only categorized two (2) to be "Suspect". These were CRDM numbers 35 and 37. However, forty-five (45) CRDM nozzles were categorized as "Masked". These locations were videotap,;d as the loose debris was va:-.:uumed to allow for complete inspection of the base of the CRDM nozzles. Subsequently an additional ten (10) CRDM nozzles were deemed to be "Suspect" from the population of forty-five (45) masked. These were CRDM numbers 11, 20, 29, 32, 41, 44, 48, 51, 64,and 65. This hrought the total number of CRDM nozzles requiring additional PT and UT inspections tc twelve (12).
After the RPV head was removed and placed on the storage stand, additicnai PT ar.d UT inspections were performed on the twelve (12) suspect CRDM nozzles. At approximately 1357 on October 22, 2001, Engineering evaluation of the visual inspection, Liquid Penetrant Test (PT) data and Ultrasonic Test (UT) data identified through-wall indicaticr► s on three (3) CRDM nozzles. This initial engineering evaluation concluded that the visual indications around CRDM numbers 35, 37, and 44 indicate a RCS pressure boundary leak. Since the condition resulted in leakage through the RCS pressure boundary, it was reported [EN # 38416] as a ncn-emergency [8-hour] report in accordance with 10 CFR 50.72(b)(3)(ii)(A). Two days later, a revised non-emergency 8-hour report stated that CRDM nozzles number 29, and 64 were also determined to be leaking.
Details PT Inspection Results The results of the PT inspection identified four CRDM locations with positive indications. None of the DOCKET (2) PAGE (3) LER NUMBER (6) indications detected were safety significant, and no indications were found which could contribute to loose parts. However, all CRDM locations with positive PT indications were repaired. The other eight nozzles exhibited no PT indication. The PT indications are as follows:
1. CRDM nozzle 35 had two (2) axiai indications in the weld and one circumferential indication approximately 23 degree in the weld l'3ward the RPV cladding.
2. CRDM nozzle 37 had one axial indication in the weld and one circumferential indication approximately 100 degree in the weld toward the RPV cladding.
3. CRDM nozzle 44 had four (4) axial indications, one in the weld and three at the end of the nozzle; and one circumferential indication approximately 23 degree in the weld toward the PPV cladding.
4. CRDM nozzle 64 had one circumferentia! indication approximately 60 degree in the weld toward the RPV cladding.
UT Inspection Results The results of the UT inspection identified seven (7) CRDM nozzles with axial flaws. No circumferential flaws were detected either above or below the .1-groove weld in the nozzle material.
The identified flaws were evaluated in accordance with the acceptance criteria contained in the September 24, 2001 draft letter from J. Strosnider (NRC NRR) to A. Marion (NE!). Three of the CRDM nozzles were determined to require repair based on the ultrasonic inspections. These CRDM locations were nozzles 44 (also determined to be leaking based on PT), 29, and 51. CRDM #29 was the only nozzle to show an OD flaw, the other flaws were all located on the ID. Based on fracture mechanics and crack growth it was determined that the flaw in CRDM nozzle 51 required repair. ID flaws in the other four (4) CRDM nozzles were analyzed to be acceptable. Note that CRDM #35 and #64 were repaired based on PT results. The other five nozzles had no flaws based on UT. The UT indications are as follows:
1. CRDM nozzle 11 had one ID axial indication. The flaw was 0.12 inch in depth and 0.36 inch long.
The flaw was located 1.91 inch below the J-groove weld. The flaw was analyzed as acceptable for at least an additional cycle of operation.
2. CRDM nozzle 29 had one OD axial indication. The flaw was 0.11 inch in depth and 0.91 inch long.
DOCKET (2) LER NUMBER (6) OF 12 The flaw was located 0.13 inch above the J-groove weld and extended to 0.34 inch above the face of the weld. It was determined that the flaw most likely entered the weld material and was too small to be seen as a PT indication. The flaw was determined to be unacceptable and repaired.
3. CRDM nozzle 35 had three (3) ID axial indications. Flaw #1 was 0.35 inch in depth and 0.44 inch long. The flaw was located 1.49 inch below the J-groove weld. Flaw #2 was 0.21 inch in depth and 0.61 inch long. The flaw was located 1.08 inch below the J-groo"e weld. Flaw #3 was 0.21 inch in depth and 0.52 inch long. The flaw was located 1.32 inch below the J-groove weld. The three flaws were closely spaced and evaluated as a combination, the combined flaw was analyzed as acceptable for an additional cycle of operation.
4. CRDM nozzle 44 had one ID axial indication. The flaw was 0.34 inch in depth and 1.53 inch long.
The flaw was located 0.33 inch beiow the J-groove weld. The flaw growth was analyzed and was deemed unacceptable and repaired.
5. CRDM nozzle 51 had five (5) ID axial indications. Flaw #1 was 0.35 inch in depth and 1.7 inch long. The flaw was located 0.97 inch below the J-circove weld. Flaw #2 ',vas 0.43 inch in depth and 2.06 inch long. The flaw was located 0.48 inch J-griove weld. Flaw #3 was 0.15 inch in depth and 0.47 inch long. The flaw was !ocated 1.99 inch below the J-groove weld. Flaw #4 was 0.17 inch in depth and 0.55 inch lor. The flaw was located 0.75 inch above the J-groove weld. Flaw #5 was 0.12 inch in depth and 0.33 inch long. The flaw was :ocated 1.1 inch above the J-groove weld. Flaws #1, #2, and #3 were closely spaced and evaluated as a combination, the combined flaw was analyzed as unacceptable and repaired.
6. CRDM nozzle 64 had one ID axial indication. The flaw was 0.24 inch in depth and 01.17 inch long.
The flaw was located 1.03 inch below the J-groove weld. The flaw was analyzed as acceptable for an additional cycle of operation (this nozzle was repaired based on PT results).
7. CRDM nozzle 65 had one !D axial indication. The flaw was 0.12 inch in depth and 0.4 inch long.
The flaw was located 0.89 inch below the J-groove weld. The flaw was analyzed as acceptatle for at least an additional cycle of operation.
DOCKET (2) PAGE (3) LER NUMBER (6)
CAUSAL FACTORS
The apparent root cause of the RPV head penetration nozzle crack was Primary Water Stress Corrosion Cracking (PWSCC). Previous industry experience with PWSCC in Alloy 600 components was evaluated as part of NRC Generic Letter 97-01, "Degradation of Control Rod Drive Mechanism Nozzle and Other Closure Head Penetrations." The findings from TMI Unit 1 are similar to the PWSCC cracking previously experienced r4t the Oconee Nuclear Stations, Crystal River, and Arkansas Nuclear. However, no circumferential cracking was found either above or below the J- groove weld on any of the twelve CRDM nozzles that were ultrasonically inspected.
TMI and Exelon will continue to participate with industry owner groups, NEI, and EPRI to better understand the PWSCC issues with alloy e00 component failures.
Discussion Alloy 600 is used extensively in nozzle applications in the reactor vessel. It is also used in the pressurizer, hot and cold leg piping as we!! as steam generator tubing in Babcock & Wilcox fabricated plants. It is generally recognized that these small-bore nozzles have experienced cracking, and the industry has evaluated the results of multiple failure analyses. The result from these analyses is that the failure mechanism is a form of stress corrosion cracking referred to as PWSCC.
PWSCC has been assumed to initiate at the inside surface of the nozzle that is adjacent to the partial penetration J-groove welds. This area has been shown to have high residual stresses as a result of the welding process. Additional stress can result from machining, grinding, or reaming operations. In thin wall products, this area could also have an altered microstructure from welding. It has been established that PWSCC can occur in materials provided that three conditions exist:
1. susceptible material 2. high tensile stress, and LER NUMBER (6) 3. an aggressive environment.
Almost all small-bore alloy 600 nozzle (including thermocouple and CRDM nozzles) attached with a partial penetration weld possesses these above characteristics. In PWR applications, numerous small bore alloy 600 nozzles and pressurizer heater sleeves have experienced leaks attributed to PWSCC.
These components in B&W designed plants are normally exposed to 600 degree F or higher temperatures and primary coolant water, as are the TMI Unit 1 RPV penetration nozzles.
CORRECTIVE ACTIONS
Immediate:
Based on prior RPV nozzle leakage found at other Babcock & Wilcox plants, 1MI Unit 1 organized a project team in January 2001 to deal with establishing personnel and equipment to inspect and subsequently repair any leaking RPV nozzles.
Subsequent:
Prior to exiting TMI Unit 1 refueling outage, the 14 identified RPV nozzle.: k.E. TIC and 6 CRDM) were repaired.
1) For the CRDM nozzles the initial repair action was to roll the nozzle above the J-groove weld, then to machine the lower portion of the CRDM nozzle including portions of the J-groove weld. The final surface was PT examined prior to ti re weld repair process. A new pressure boundary weld was formed between the CRDM nozzle and the RPV head low alloy steel at a location above the previous J-groove weld and below the rolled nozzle area. Protective compression surface stress remediation was performed after the repair was PT and UT examined.
2) For the thermocouple nozzles, the nozzles were cut approximately 1 inch from the outside surface of the RPV head. The remaining nozzle portion inside the RPV head was machined out of the DOCKET (2) LER NUMBER (6) head. Six TIC nozzles were plugged by installing an Inconel 690 nozzle plug in the RPV head bore. After the nozzle plug was tack welded, a nozzle plug weld dam was inserted into the cavity in the top of the nozzle plug and tack welded to the nozzle plug. A weld pat.; build up of Inconel 152 was welded over the nozzle plug and weld dam using a temper bead welding process. The other two TIC locations that are used for operation of the RCITS and Reactor High Point Vent had an Inconel 690 sleeve installed into-the RPV head bore. The sleeve .ras secured with a weld pad build up of Inconel 152. Afterwards, !" nozzle assembly was inserted into the sleeve and the circumference of the nozzle assembly was welded to the bottom of the sleeve. A replacement TIC flange assembly (consisting of a new flange, % inch schedule 160 stainless steel pipe welded to the flange and an Inconel tube welded to the stainless steel pipe) was inserted into the Inconel sleeve and welded in place.
In addition, a video inspection of the RPV head surface was 3orripleteJ after cleaning activities to provide a baseline for future visual inspections.
Planned:
Although repairs have been completed for the 14 identified RPV head nozzles, the potential for future leakage events of alloy 600 CRDM nozzles or the existing RPV head due to PWSCC will be addressed through continued RPV head nozzle inspections and repairs as necessary. Corrective actions are in place to re-examine CRDM nozzles 11 and 65 during the next refueling outage if they remain in service after the next refueling outage. In the long term, the RPV head may be replaced to prevent recurrence of this event.
TMI-1 plans to continue to participate in industry activities regarding PWSCC.
DOCKET (2) LER NUMBER (6) 2001 -- 002 -- 00
SAFETY ANALYSIS
The degraded condition of the eight TIC and six CRDM nozzles did not represent a challenge to the nuclear safety of the plant. The cracks were primarily in the J-groove weld metal. No circumferential cracking was found either above or below the J-groove weld therefore eliminat;ng any CRDM nozzle ejection. These cracks propagated until they resulted in leaks that were detected during a planned RPV head surveillance.
Primary coolant leakage from the cracks was minimal due to the relatively tight cracks. The total leakage from the 13 RPV head nozzles was significantly less than the Technical Specification limits for unidentified RCS inventory loss nor were there obvious incraas6s in Reactor Building (RB) normal sump levels or increases in Reactor Building air sampling radiation activity. No RB or area radiation monitors alarms sounded. The small amounts of boric acid deposits that were observed around the RPV head nozzles had caused' no observable corrosion to the RPV head.
Inspections of the top surface of the RPV head are performed at each scheduled refueling outage.
The results of the 1R14 RPV head inspection supports the conclusion that the CRDM nozzles would leak first and be discovered by station personnel during normal RPV head inspections. Finally, fracture mechanics analysis has shown that the lavozzle flaw on CRDM # 11 and #65 will be acceptable for at least 5 Effective Full Power Years of operation.
In conclusion, PWSCC of the TMI-1 nozzles did not pose a nuclear safety risk. The basis for RPV head nozzle cracking not being a safety risk include:
1. Leak rates from cracks within the annulus region of the nozzle are low, 2. No circumferential cracks were found either above or below the J-groove weld which also makes the potential for complete nozzle failure and control rod ejection a low probability event, 3. Leakage from cracked nozzles will result in visible boric acid deposits around the nozzle that would be discovered during normal refueling inspections, DOCKET (2) LER NUMBER (6) 2001 -- 002 -- 00 4. The RPV head has been cleaned in order to enhance the ability to identify any nozzle leaks that may develop in the future.
ADDITIONAL INFORMATION
There were no releases of radioactive n-r.terials, radiation exposures, cr personnel injuries associated with this event.
This event is considered reportable under the Equipment Performance and Information Exchange (EPIX) program.
SIMILAR EVENTS
TMI Unit 1 has had no LERs over the past 3 years that reported PWSCC of alloy 600 components or leaks involving RPV head penetration nozzles.
PWSCC is not new either to the domestic cc worldwide nuclear industry. As evidenced from the similar PWSCC discoveries at the other B&V/ nuclear stations and TMI Unit 1, TMI will remain susceptible to future PWSCC cracking of ecy 600 components. Until a planned corrective action to replace the RPV head is implemented, this type of event may be expected to recur.
- The Energy Industry Identification System (EIIS), System Identification (SI) and Component Function Identification (CFI) Codes are included in brackets, [SI/CFI] where applicable, as required by 10 CFR 50.73 (b)(2)(ii)(F).