12-16-2005 | On October 18, 2005 a prompt notification to the NRC was made per 10CFR50.72(b)(3)(ii)(A) to report discovery of flaws in the pressurizer spray nozzle [AB] to safe-end bimetallic weld region. With the plant in Mode 6 (Refueling), engineering personnel performing preplanned radiography ( RT) examinations in accordance with the Alloy 600 weld inspection program (as committed to in response to NRC Bulletin 2004-01), discovered two linear indications. In an effort to better characterize the indications, additional ultrasonic examinations were performed which identified three indications. The indications were conservatively identified as degradation of the reactor coolant system pressure boundary in that corrective action to restore the barrier's capability was necessary. A request was made and the NRC granted approval for Millstone to utilize an alternate repair method from that specified in the ASME Code. The weld was repaired utilizing a weld overlay process.
This event/condition is being reported pursuant to 50.73(a)(2)(ii)(A), as a condition that resulted in the nuclear power plant, including its principal safety barriers, being degraded. |
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1. Event Description On October 18, 2005 a prompt notification to the NRC was made per 10CFR50.72(b)(3)(ii)(A) to report discovery of flaws in the pressurizer spray nozzle [AB] to safe-end bimetallic weld region. With the plant in Mode 6 (Refueling), engineering personnel performing preplanned RT examinations in accordance with the Alloy 600 weld inspection program (as committed to in response to NRC Bulletin 2004-01), discovered two linear indications. In an effort to better characterize the indications, additional ultrasonic (UT) examinations were performed which identified three indications. There was no evidence of any past RCS leakage from any of the indications.
The indications are located at or near the inner diameter surface of the pressurizer spray nozzle to safe-end bimetallic weld region, which is part of the RCS pressure boundary. Though it cannot be conclusively stated that the indications break the ID surface, all three indications are characterized by the rules'of IWA-3310 (b) as surface planar indications due to their proximity to the ID surface. This rule states that even if the indication was characterized as a sub-surface indication, if any portion of the indication is less than 0.4d (d = indication depth) from the nearest surface of the component to the indication it must be classified as a surface indication.
Indication Number 1 Indication number 1 is a lack of fusion between the inconel weld at the fusion line between the structural weld and the nozzle butter. The length of this indication is 16.75 inches with a measured through wall depth of 0.208" (- 24% of wall thickness). This indication is a circumferential indication between the carbon steel nozzle and the nozzle butter. This indication is essentially comprised of 5 indications, which due to their proximity to one another, must be considered as one continuous circumferential indication. However the specifics of the five locations are: 3.05" to 5.85", 6.90" to 12.55", 12.95" to 13.85", 14.25" to 18.25", and 19.2" to 22.6". The distance between these indications is such that all indications would require evaluation as a single indication whether classified as parallel planar indications or multiple nonaligned coplanar indications.
Indication Number 2 Indication number 2 is located on the opposite side of the weld at the interface between the weld and the safe-end and is also a lack of fusion. The length of this indication is 7.7" with a measured through wall depth of 0.219" (- 24% of wall thickness). This indication is a circumferential indication between the inconel weld and the stainless steel safe-end. The separation between the circumferential indications along the long axis of the component is the width of the butter material, which is calculated to be approximately 0.7".
Indication Number 3 Indication number 3 is an axial indication originating at the inside surface and located wholly in the nozzle weld butter region. The length of this indication is 0.25" with a measured through wall depth of 0.214" (- 24% of wall thickness). Note: This is an estimated value only as through wall and length sizing of axially orientated indications is not qualified through the Performance Demonstration Initiative (PDI) program for Appendix VIII, Supplement 10 of Section XI of the ASME Code. Axial indications are only qualified for detection.
Previous examination results were reviewed. Neither the pre-service examination performed in 1985 nor the in- service examination conducted in 1991 noted any recordable indications. It should be noted that the UT technique utilized for the pre-service and the 1991 examinations were not qualified through the PDI program for Appendix VIII, Supplement 10 of Section XI of the ASME Code, as they predated the development of the PDI program.
A request was made and the NRC granted approval for Millstone to utilize an alternate repair method from that specified in the ASME Code. A structural weld overlay was applied to restore the structural integrity of the pressurizer spray nozzle.
This event/condition is being reported pursuant to 50.73(a)(2)(ii)(A), as an event or condition that resulted in the nuclear power plant, including its principal safety barriers, being degraded.
2. Cause The cause of these weld indications was attributed to defects in the weld that occurred during original construction.
Improvements in the NDE techniques over those in place at the time of the original construction, allowed these indications to be found at this time. The indications are in the transition area between the pressurizer spray nozzle and the safe-end weld butter. Dominion NDE personnel reviewed the original RTs. Since the existence of the indications was known (as a result of the additional UTs), the indications could then be 'seen' in the original RTs.
However, the NDE personnel indicated that had they been reviewing the RTs for the first time without the benefit of knowing the indications existed, they would have also accepted the RTs as satisfactory. The spray nozzle had radiography performed to meet the ASME Section III (construction) requirements and a pre-service UT was performed using shear waves. An in-service inspection was performed in 1991 utilizing a 45 degree refracted longitudinal wave transducer based upon industry experience in dissimilar metal (DM) welds (Hope Creek). Since the last in-service inspection performed in 1991, Appendix VIII, Supplement 10 of ASME Section XI has become mandatory (November 2002) for DM welds through the PDI Program. The PDI Program has identified through procedure, personnel and equipment qualification that previous inspections performed to DM welds were inadequate for detection of flaws. It should also be noted that differences in the RT technique is what enabled the indications to be seen more clearly on the 3R10 radiographs. The construction radiographs used Kodak AA film, which is a grainy high-speed film typically used for construction in order to reduce the exposure time. The radiographs taken during 3R10 utilized D4 film, which is a fine grain film, used to ensure high quality radiographs.
3. Assessment of Safety Consequences An evaluation in accordance with ASME XI (IWB-3600) showed that the indication sizes did not currently exceed code limits. Therefore the as found structural integrity of the weld was adequate for all design loads, and the pressure boundary integrity and structural loading capability of the weld was not impaired below acceptable limits.
The pressure and structural integrity was also adequate for all prior plant operations. However, future adequacy of the flawed weld could not be assured without completion of the weld overlay repair. A request was made and the NRC granted approval for Millstone to utilize an alternate repair method from those specified in the ASME Code. A structural weld overlay was designed and installed via DCN DM3-00-0422 and AWO M3-05-14477. A structural weld overlay was applied to ensure future structural integrity of the pressurizer spray nozzle.
4. Corrective Action The structural weld overlay provided structural integrity for the pressurizer spray nozzle to address the identified weld indications. Additionally, in response to NRC Bulletin 2004 -01, a plan has been developed to provide direction for administrating Alloy 600/82/182 inspections, and will be fully implemented over several outages.
There is high confidence that the ISI program, implemented in accordance with Section XI of the ASME code, would be capable of detecting similar weld defects.
6. Previous Occurrences None Energy Industry Identification System (EDS) codes are identified in the text as [xx].
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05000328/LER-2005-001 | Unit 2 Reactor Trip Following Closure of Main Feedwater Upon Inadvertent Opening of Control Breakers | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function | 05000388/LER-2005-001 | DDegradation of Primary Coolant Pressure Boundary due to Recirculation Pump Discharge Valve Bonnet Vent Connection Weld Flaw | 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded | 05000423/LER-2005-001 | | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000455/LER-2005-001 | Unit 2 Automatic Reactor Trip Due to Low Steam Generator Level resulting from a Software Fault on the Turbine Control Power Runback Feature | | 05000370/LER-2005-001 | Automatic Actuation of Motor Driven Auxiliary Feedwater Pumps During Outage | | 05000244/LER-2005-001 | Failure of ADFCS Power Supplies Results in Plant Trip | | 05000247/LER-2005-001 | 0Technical Specification Prohibited Condition Due to Exceeding the Allowed Completion Time for One Inoperable Train of ECCS Caused by an Inoperable Auxiliary Component Cooling Water Check Valve | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000529/LER-2005-001 | REACTOR HEAD VENT AXIAL INDICATIONS CAUSED BY DEGRADED ALLOY 600 COMPONENT | 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded | 05000336/LER-2005-001 | | | 05000266/LER-2005-001 | | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition | 05000269/LER-2005-001 | | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000289/LER-2005-001 | | | 05000293/LER-2005-001 | | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | 05000298/LER-2005-001 | Reactor Scram due to Reactor Level Transient and Inadvertent Rendering of High Pressure Coolant Injection Inoperable | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | 05000331/LER-2005-001 | | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition | 05000315/LER-2005-001 | Reactor Trip Following Intermediate Range High Flux Signal | | 05000316/LER-2005-001 | Reactor Trip from RCP Bus Undervoltage Signal Complicated by Diesel Generator Output Breaker Failure | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000317/LER-2005-001 | Main Feedwater Isolation Valve Inoperability Due to Handswitch Wiring | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000323/LER-2005-001 | TS 3.4.10 Not Met During Pressurizer Safety Valve Surveillance Testing Due to Random Lift Spread | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000333/LER-2005-001 | Inoperable Offsite Circuit In Excess of Technical Specifications Allowed Out of Service Time | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000352/LER-2005-001 | Loss Of Licensed Material In The Form Of A Radiation Detector Calibration Source | | 05000353/LER-2005-001 | Core Alterations Performed With Source Range Monitor Alarm Horn Inoperable | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000306/LER-2005-001 | | 10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000361/LER-2005-001 | | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000362/LER-2005-001 | Emergency Diesel Generator (EDG) 3G003 Declared Inoperable Due to Loose Wiring Connection on Emergency Supply Fan | | 05000263/LER-2005-001 | | | 05000456/LER-2005-001 | Potential Technical Specification (TS) 3.9.4 Violation Due to Imprecise Original TS and TS Bases Wording | | 05000454/LER-2005-001 | Failed Technical Specification Ventilation Surveillance Requirements During Surveillance Requirement 3.0.3 Delay Period | | 05000282/LER-2005-001 | | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v), Loss of Safety Function | 05000286/LER-2005-001 | Plant in a Condition Prohibited by Technical Specifications due to Error Making Control Room Ventilation System Inoperable | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000400/LER-2005-001 | Reactor Auxiliary Building Emergency Exhaust System Single Failure Vulnerability | | 05000395/LER-2005-001 | Emergency Diesel Generator Start and Load Due To A Loss Of Vital Bus | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000382/LER-2005-001 | RCS Pressure Boundary Leakage Due to Primary Water Stress Corrosion Cracking of a Pressurizer Heater Sleeve | 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded | 05000305/LER-2005-001 | | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000369/LER-2005-001 | Reactor Coolant System Leakage Detection Instrumentation Inoperable | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000266/LER-2005-002 | | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition | 05000255/LER-2005-002 | | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000361/LER-2005-002 | Missing Taper Pins on CCW Valve Cause Technical Specification Required Shutdown | 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown | 05000370/LER-2005-002 | Ice Condenser Lower Inlet Door Failed Surveillance Testing | | 05000353/LER-2005-002 | High Pressure Coolant Injection System Inoperable due to a Degraded Control Power Fuse Clip | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | 05000263/LER-2005-002 | | | 05000454/LER-2005-002 | One of Two Trains of Hydrogen Recombiners Inoperable Longer Than Allowed by Technical Specifications Due to Inadequate Procedure | | 05000244/LER-2005-002 | Emergency Diesel Generator Start Resulting From Loss of Off-Site Power Circuit 751 | | 05000362/LER-2005-002 | Emergency Containment Cooling Inoperable for Longer than Allowed by Technical Specifications | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000247/LER-2005-002 | DTechnical Specification Prohibited Condition Due to Exceeding the Allowed Completion Time for One Inoperable Train of ECCS Caused by Gas Intrusion from a Leaking Check Valve | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000306/LER-2005-002 | | 10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown | 05000265/LER-2005-002 | Main Steam Relief Valve Actuator Degradation Due to Failure to Correct Vibration Levels Exceeding Equipment Design Capacities | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown | 05000286/LER-2005-002 | • Entergy Nuclear Northeast Indian Point Energy Center 450 Broadway, GSB P.O. Box 249Entergy Buchanan. NY 10511-0249 Tel 914 734 6700 Fred Dacimo Site Vice President Administration July 5, 2005 Indian Point Unit No. 3 Docket Nos. 50-286 N L-05-078 Document Control Desk U.S. Nuclear Regulatory Commission Mail Stop O-P1-17 Washington, DC 20555-0001 Subject:L Licensee Event Report # 2005-002-00, "Automatic Reactor Trip Due to 32 Steam Generator Steam Flow/Feedwater Flow Mismatch Caused by Low Feedwater Flow Due to Inadvertent Condensate Polisher Post Filter Bypass Valve Closure." Dear Sir: The attached Licensee Event Report (LER) 2005-002-00 is the follow-up written report submitted in accordance with 10 CFR 50.73. This event is of the type defined in 10 CFR 50.73(a)(2)(iv)(A) for an event recorded in the Entergy corrective action process as Condition Report CR-IP3-2005-02478. There are no commitments contained in this letter. Should you or your staff have any questions regarding this matter, please contact Mr. Patric W. Conroy, Manager, Licensing, Indian Point Energy Center at (914) 734-6668. Sincerely, 4F-/t R. Dacimo Vice President Indian Point Energy Center Docket No. 50-286 NL-05-078 Page 2 of 2 Attachment: LER-2005-002-00 CC: Mr. Samuel J. Collins Regional Administrator — Region I U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Resident Inspector's Office Resident Inspector Indian Point Unit 3 Mr. Paul Eddy State of New York Public Service Commission INPO Record Center NRC FORM 3660 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 EXPIRES: 06/30/2007 (6-2004) Estimated burden per response to comply with this mandatory collection request 50 hours.RReported lessons teamed are incorporated into the licensing process and fed back to Industry. Send comments regarding burden estimate to the Records and FOIA/Privacy Service Branch (T-5 F52), U.S. Nuclear Regulatory Commission, Washington, DC 29555-0001, or by InternetLICENSEE EVENT REPORT (LER) e-mail to Infocoilectsenrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-l0202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person Is not required to respond to, the Information collection. 1. FACIUTY NAME 2. DOCKET NUMBER 3. PAGE INDIAN POINT 3 05000-286 10OF06 4. TITLE Automatic Reactor Trip Due to 32 Steam Generator Steam Flow/Feedwater Flow Mismatch Caused by Low Feedwater Flow Due to Inadvertent Condensate Polisher Post Filter Bypass Valve Closure | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000287/LER-2005-002 | Unit 3 trip with ES actuation due to CRD Modification Deficiencies | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000336/LER-2005-002 | | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
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