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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 55314 | 18 February 2021 04:00:00 | 10 CFR 21.21(d)(3)(i), Failure to Comply or Defect | Part 21 Report | The following is a summary of a report received from the vendor via email: The vendor notified the NRC that the hydraulic loss coefficient used to calculate the pressure loss and flow rate into the Side Entry Orifice (SEO) at the fuel bundle entrance in BWR/6 plants may have been underpredicted. The SEO loss coefficient was underpredicted for some fuel bundle locations, which could result in an overprediction of MCPR margin in core monitoring applications. The overprediction was a result of not originally including in the assessment flow area restrictions associated with instrument support structures in the cross beams (structural supports underneath the core plate) in BWR/6 plant designs. BWR/2-5 plants built by GE have a different core support structure that is more open so that multiple SEO losses are not applied to evaluations for those plants. However, there are currently no US ABWR plants in operation that would potentially be affected by this evaluation. The potential error in the core monitoring system does not affect the NRC certified design of the ABWR or the GEH ABWR design certification renewal application currently under review. The unique core support structure design of the BWR6 and ABWR is not shared by earlier BWR plants or the ESBWR. Potentially Affected Plants: Grand Gulf, River Bend, Clinton, Perry Point of Contact: Michelle Catts GE Hitachi Nuclear Energy Safety Evaluation Program Manager 3901 Castle Hayne Road, Wilmington, NC 28401 | ||
| ENS 54756 | 28 April 2020 04:00:00 | 10 CFR 21.21(d)(3)(i), Failure to Comply or Defect | Part 21 Report - Non-Conservative Bwr/6 Side Entry Orifice Loss Coefficients | The following is a summary of information received from GE Hitachi Nuclear Energy: For core monitoring, Global Nuclear Fuel-Americas (GNF) applies different hydraulic loss coefficients to BWR/6 fuel bundles as a function of location. The Side Entry Orifice (SEO) loss coefficient varies in BWR/6 plants, depending on the orientation of the SEO relative to intersecting core support beams. The SEO loss coefficient for some bundle locations are under predicted, which results in a local over prediction of Minimum Critical Power Ratio (MCPR) margin. Specifically, the SEO loss coefficients for fuel bundle locations adjacent to Intermediate and Source Range Monitor (IRM/SRM) locations should be higher than the current "one beam" values that are applied. The MCPR effect on affected bundles under conditions where they are near limits can reduce margin by 0.02 in CPR. The condition is not a substantial safety hazard, but the MCPR effect is above the threshold that GNF applies for reportability. GNF has completed its evaluation and concluded that this issue is a reportable condition under 10 CFR 21.21(d). The basis for reportability is that the change in MCPR associated with this issue could contribute to the exceeding of a safety limit, as defined in the technical specifications of a license for operation issued under 10 CFR Part 50. US BWR Potentially Affected Plants: Grand Gulf, River Bend, Clinton, and Perry. There are currently no US ABWR plants in operation that would potentially be affected by this evaluation. The potential error in the core monitoring system does not affect the NRC certified design of the ABWR or the GEH ABWR design certification renewal application currently under review. The unique core support structure design of the BWR6 and ABWR is not shared by earlier BWR plants or the ESBWR. Affected plants should review the following paragraph and contact their GNF account representative for schedule and delivery dates for updated core monitoring databanks. Updated core monitoring databanks for the affected BWR/6 plants are scheduled to be completed by July 10, 2020. For affected plants that use GNF core monitoring software, the long-term corrective action is to implement an updated core monitoring databank with corrected SEO loss coefficient values. Affected plants should contact their GNF account representative for schedule and delivery dates. If necessary, a short-term corrective action such as an administrative MCPR penalty on IRM/SRM locations, should be applied to the core monitoring system. An effective short-term measure would be to assure that IRM/SRM locations are 0.02 less than the Maximum Fraction of Limiting CPR limit (e.g., MFLCPRlimit = 1.00). This condition is met if the core wide MFLCPR is 0.02 below the limit (e.g., MFLCPR < 0.98), because the IRM/SRM locations will be bounded. Note that this recommended short term or interim measure was developed on a standalone basis and does not supplant any existing MFLCPR limit reductions established for other purposes. For non-affected plants, there are no recommended actions. GNF recognizes that other non-GNF fueled plants, safety analysis methodologies, and core monitoring systems may have different design bases and address SEO losses by other means. For more information, please contact: Michelle Catts GE Hitachi Nuclear Energy Safety Evaluation Program Manager 3901 Castle Hayne Road, Wilmington, NC 28401 Michelle.Catts@GE.com (910) 200-9836 | ||
| ENS 52692 | 27 February 2017 04:00:00 | 10 CFR 21.21(a)(2), Interim Report for Comply or Defect in Component | Part 21 Report - Control Rod Drive Mechanisms Contaminated with Chlorides | Dale E. Porter GE-Hitachi Nuclear Energy Safety Evaluation Program Manager 3901 Castle Hayne Rd., Wilmington, NC 28401 (910) 819-4491 Dale.Porter@GE.Com The inappropriate addition of chlorinated water from container box desiccants into the CRDMs (Control Rod Drive Mechanisms) during leak testing after rebuild could potentially initiate Intergranular Stress Corrosion Cracking (IGSCC) or Transgranular Stress Corrosion Cracking (TGSCC). These two types of SCC could cause a separation of the stop piston or separation of the index tube contained within the CRDM. The stop piston separation could cause a slower scram speed and damage the drive so it could not be withdrawn. An index tube separation could result in a similar type of rod uncoupling event that would have the potential to result in a rod drop accident (RDA). The piston tube located within the CRDM is a reactor coolant pressure boundary (RCPB) and is an ASME component. There is a possibility of cracking causing a RCPB leak. SCC initiation on the Cylinder Tube and Flange (CTF) area of the CRDM could result in a separation that could prevent a scram or normal insertion of a CRDM. Reports have been issued to River Bend, LaSalle Unit 2, and Hatch Unit 2 providing the results of an evaluation that concludes that the condition will not create a substantial safety hazard or potentially cause a Technical Specification Safety Limit violation for a minimum of one operating cycle. The Browns Ferry Unit 2 drives were shipped but were not installed prior to recall, thus a short-term evaluation for Browns Ferry has not been completed. River Bend, Entergy, Shipped Date: 2017, Quantity Shipped: 15, Customer PO Number: 10478763 LaSalle Unit 2, Exelon, Shipped Date: 2017, Quantity Shipped: 24, Customer PO Number: 00414787-66 Hatch Unit 2, Southern Nuclear, Shipped Date: 2017, Quantity Shipped: 15, Customer PO Numbers: SNG50295-0001 & SNG50295-0002 Browns Ferry Unit 2, TVA, Shipped Date: 2017, Quantity Shipped: 32, Customer PO Number: 2424171
Pursuant to 10 CFR 21.21(d)(4), GEH is providing the final report with the conclusion that, in limited cases, the chloride contamination could create a substantial safety hazard. Attachment 1 identifies the potentially impacted plants and Attachment 2 contains the final report information. The enclosure provides additional details of the evaluation. Updated Attachment 1 notes: A portion of the CRDMs at the Hatch Plant are Reportable while the remainder are not. Summary of updates to Attachment 2: The inappropriate addition of chlorinated water from container box desiccants into the CRDMs during leak testing after rebuild could potentially initiate Intergranular Stress Corrosion Cracking (IGSCC) or Transgranular Stress Corrosion Cracking (TGSCC). These two types of SCC were initially considered for the potential to cause a separation of the stop piston or separation of the index tube contained within the CRDMs that are constructed of 304 Stainless steel. The completed evaluation indicates that a stop piston separation could cause a slower scram speed and damage the drive so it could not be withdrawn. The potential exists for the control rod to drift out. The piston tube located within the CRDM is a reactor coolant pressure boundary (RCPB) and is an ASME component. The possibility of cracking causing RCPB leakage was eliminated by the evaluation. An index tube separation was eliminated as a potential failure mode. Likewise, the potential for SCC initiation on the Cylinder Tube and Flange (CTF) area of the CRDM resulting in a separation that could prevent a scram or normal insertion/withdrawal of a CRD was eliminated. The long-term evaluation concluded there is no concern for TGSCC. This evaluation also determined the stop piston to piston tube separation is the only failure mechanism that could occur, and only if those components were manufactured from 304 SS. GEH initiated a Root Cause Evaluation (RCA) to determine why this event occurred and has implemented process changes to ensure that the condition does not reoccur. Actions to prevent recurrence, such as eliminating the desiccant material and flushing the closed loop water system, have been completed. For Hatch Unit 2, the 12 CRDMs that have the 304 SS piston tubes should be replaced prior to those CRDMs exceeding 10 years in service. See table below: CRDM S/N, Piston Tube SE0474, 304 SS A8737, 304 SS A9423, 304 SS A6791, 304 SS 3095, 304 SS A8729, 304 SS 7253, 304 SS A6786, 304 SS SE0368, 304 SS A5409, 304 SS 7080, 304 SS A9484, 304 SS Interim Reports were issued to River Bend, LaSalle Unit 2, and Hatch Unit 2 providing the results of an evaluation that concluded the condition would not create a substantial safety hazard or potentially cause a Technical Specification Safety Limit violation for a minimum of one operating cycle. The Browns Ferry Unit 2 drives were shipped but were not installed prior to recall, thus a short-term evaluation for Browns Ferry was not provided. GEH has completed the long-term CRDM evaluation with the following results: A Safety Information Communication is being issued to River Bend, LaSalle Unit 2, and Browns Ferry Unit 2 stating that the CRDMs exposed to the chloride intrusion will not cause a substantial safety hazard, or cause a Technical Specification Safety Limit violation and is therefore not reportable (see enclosure 1 for details). For Hatch Unit 2, the introduction of chlorides could cause a substantial safety hazard for the 12 CRDMs that were manufactured from 304 SS material, and is therefore a reportable condition per 10CFR 21.21(d); however, the 3 CRDMs manufactured from XM-19 material would be considered not reportable. The results of the short-term evaluations for all three plants where CRDMs were installed remains valid. Notified the R2DO (Bonser), R3DO (Peterson), R4DO (Proulx), and Part 21/50.55 Reactors Group by email. | Control Rod | |
| ENS 49610 | 9 October 2013 05:00:00 | 10 CFR 21.21(d)(3)(i), Failure to Comply or Defect | Part 21 - Failure of Relay Logic Card Modules in 2 Out of 4 Voters for Power Range Monitors | The observed aging characteristics of the Solid State Relays on two particular Voter Relay Logic Cards Modules present a future challenge to the Power Range Neutron Monitoring System's ability to reliably insert the Average Power Range Monitor High-High/INOP scram and/or the Oscillating Power Range Monitor Scram to the Reactor Protection System in plant conditions warrant such action. For Relay Logic Cards that have been in operation for over 10 years, Relay Logic Card Module replacement should be scheduled and accomplished at the first reasonable opportunity. For Relay Logic Cards in operation for less than 10 years, a program should be implemented to replace Relay Logic Card Modules before the 10 years is reached. Plant affected; Nine Mile Point 2, Fermi 2, Grand Gulf, Limerick 1 & 2, Peach Bottom 2 & 3, Susquehanna 1&2, Brunswick 1 & 2, Hatch 1 & 2, Browns Ferry 1, 2, & 3, Monticello, and Columbia. | Reactor Protection System | |
| ENS 48735 | 8 February 2013 05:00:00 | 10 CFR 21.21(d)(3)(i), Failure to Comply or Defect | Part 21 - Completed Evaluation for Potential Error in Main Steam Line High Flow Calculational Methodology | GEH recently discovered that some calculations of the choked flow rate in the Main Steam Lines (MSLs) of GEH BWRs were non-conservative, with potential effects on margins between choked flow conditions and existing MSL high-flow Nominal Trip Setpoints (NTSPs), Allowable Values (AVs), and Analytical Limits (ALs). GEH has now completed the evaluation of this condition and has determined this condition is not reportable under 10 CFR 21 for all U.S. BWR/2-6 plants. The effect of the discovered non-conservatisms in choked flow rate values was offset by unintended conservatisms in the GEH recommended formulation for calculating pressure drop across the MSL flow restrictor. As a result, GEH has determined that the flow-instrument pressure remain at conservative values (which would ensure that the associated NTSPs and AVs expressed in psid also remain at conservative values), and the MSL high-flow trip will function as designed. This update to the 60-day Interim Notification issued on December 12, 2012 (MFN 12-111 R1) will be sent to all US BWR/2-6 plants licensed using the GEH design basis and safety analysis. See previous NRC Event Report 48350. | Main Steam Line Main Steam | |
| ENS 48585 | 13 December 2012 05:00:00 | 10 CFR 21.21(a)(2), Interim Report for Comply or Defect in Component | 60-Day Interim Report Involving an Evaluation of a Design Change to Magne-Blast Circuit Breakers | The following information was received via fax: Summary GE Hitachi Nuclear Energy (GEH) is investigating the adequacy of a Design Change in AM 4.16-350-2C and AM 4.16-350-2H Magne-Blast Circuit Breakers as a result of a breaker failure at a BWR Licensee. GEH has not completed the evaluation of this condition to determine reportability under 10 CFR Part 21 and is therefore issuing this 60-day Interim Notification. GEH will close or issue an update on this matter on or before June 14, 2013. Given the early status of the evaluation, GEH has no recommended actions at this time. This 60-day Interim Notification is issued in accordance with 10 CFR Part 21.21(a)(2), and will be sent to all GE BWR/2-6 plants and all PWRs. Recommendation GEH advises licensees to take no action at this time. However, if licensees wish to determine if the extent of this issue is present at their respective plants, they may identify breaker models AM 4.16-350-2H/2C with a date of manufacture prior to October 6, 1971. Other AM breaker styles do not use the booster cylinder as designed in breaker model AM 4.16-350-2H/2C and are not included in this concern. Therefore the extent is limited to the AM 4.16-350-2H/2C models manufactured prior to October 6, 1971. US Plants Potentially Affected Nine Mile Point 1-2, Fermi 2, Columbia, Grand Gulf, River Bend, FitzPatrick, Pilgrim, Vermont Yankee, Clinton, Dresden 2-3, LaSalle 1-2, Limerick 1-2, Oyster Creek, Peach Bottom 2-3, Quad Cities 1-2, Perry 1, Duane Arnold, Cooper, Susquehanna 1-2, Brunswick 1-2, Hope Creek, Hatch 1 - 2, Browns Ferry 1-3, Monticello, Callaway, Palo Verde 1-3 , Calvert Cliffs 1-2, Ginna, Arkansas Nuclear One 1-2, Indian Point 2-3, Millstone 2, Millstone 3, North Anna 1-2, Palisades, Surry 1-2, Waterford 3, Catawba 1-2, Oconee 1-3, McGuire 1-2, Braidwood 1-2, Exelon Byron 1-2 , Three Mi le Island 1, Beaver Valley 1-2, Davis-Besse, Seabrook, St. Lucie 1-2, Turkey Point 3-4, Point Beach 1-2, DC Cook 1-2, Prairie Island 1-2, Fort Calhoun, Diablo Canyon 1-2, Crystal River 3, Robinson, Shearon Harris, Salem 1, Salem 2, Summer, South Texas Project 1-2, San Onofre 2-3, Farley 1-2, Vogtle 1-2, Sequoyah 1-2, Watts Bar 1, Comanche Peak 1-2, and Wolf Creek. If you have an questions, please call me at (910) 819-4491. Dale E. Porter Safety Evaluation Program Manager GE-Hitachi Nuclear Energy Americas LLC Hitachi Reference Number: MFN-128 R0
The following information was received via fax: (GE Hitachi Nuclear Energy) GEH has completed all testing and evaluations and has determined that the condition previously described in MFN 12-128 R0 is not a reportable condition under 10 CFR Part 21. This information will be sent to all GE BWR/2-6 plants and all PWRs that were previously notified under the 10CFR21.21(a)(2) communication. Notified R1DO (Dentel), R2DO (Ehrhardt), R3DO (Lara), R4DO (Gepford), and Part 21 Group via email. | ||
| ENS 48350 | 27 September 2012 15:01:00 | 10 CFR 21.21(a)(2), Interim Report for Comply or Defect in Component | Error in Main Steam Line High Flow Calculational Methodology | The following information was received by facsimile: GEH (General Electric Hitachi) has recently discovered that calculations of choked flow rate in the Main Steam Line (MSL) of GEH BWRs may not be conservative, with the potential impacts to be evaluated for existing MSL high-flow setpoints and Analytical Limits (ALs). GEH has not completed the evaluation of this condition to determine reportability under 10CFR Part 21 and is therefore issuing this 60-day Interim Notification. GEH will close or issue an update on this matter on or before December 12, 2012. Given the early status of the evaluation, GEH has no recommended actions at this time. This 60-day Interim Notification is issued in accordance with 10CFR Part 21.21(a)(2), and will be sent to all GE BWR/2-6 plants and ABWR plants. Affected plants include the following: Nine Mile 1-2, Fermi 2, Columbia, Grand Gulf, River Bend, FitzPatrick, Pilgrim, Vermont Yankee, Clinton, Dresden 2-3, LaSalle 1-2, Limerick 1-2, Oyster Creek, Peach Bottom 2-3, Quad Cities 1-2, Perry 1, Duane Arnold, Cooper, Susquehanna 1-2, Brunswick 1-2, Hope Creek, Hatch 1-2, Browns Ferry 1-3, and Monticello. | Main Steam Line | |
| ENS 47630 | 1 February 2012 05:00:00 | 10 CFR 21.21(a)(2), Interim Report for Comply or Defect in Component | Part 21 Report - Failure of Crd Collet Retainer Tube/Outer Tube Weld | The following information was received via facsimile: During a recent refurbishment of a Control Rod Drive (CRD) performed by GE Hitachi Nuclear Energy (GEH) for a domestic customer a 360 degree failure of the collet retainer tube fillet weld was identified. This weld is part of the CRD 919D258G003 Cylinder, Tube and Flange (CTF) assembly. The collet retainer tube fillet weld was performed in 1983 and subsequently assembled into a Group 003 part number 919D258G003 CTF. This G003 CTF assembly was assembled into a CRD in 1995 and placed into service in 1996. GEH continues to investigate the cause(s) of the failed fillet weld. Once the cause of the fillet weld failure is determined, GEH will review the extent of condition of this failure as well as the consequences to determine if a reportable condition exists. There were no adverse effects on the CRD's operation observed due to this failure. This 60-day interim notification, in accordance with 10CFR Part 21.21(a)(2), will be sent to all BWR/2-6 plants that utilize CRDs equipped with either 919D258G002 or 919D258G003 CTF assemblies. The affected plants are: Nine Mile Point 1-2, Fermi 2, Columbia, Grand Gulf, River Bend, Fitzpatrick, Pilgrim, Vermont Yankee, Clinton, Dresden 2-3, LaSalle 1-2, Limerick 1-2, Oyster Creek, Peach Bottom 2-3, Quad Cities 1-2, Perry 1, Duane Arnold, Cooper, Susquehanna 1-2, Brunswick 1-2, Hope Creek, Hatch 1 - 2, Browns Ferry 1-3, Monticello, and Millstone.
GEH has completed the evaluation of this condition and has determined that the failure of Control Rod Drive collet retainer tube fillet weld is not a Reportable Condition as defined by 10CFR Part 21. Notified R1DO (Cahill), R2DO (Widmann), R3DO (Passehl), R4DO (Gepford) and Part 21 Group (via email). | Control Rod | |
| ENS 46429 | 18 November 2010 05:00:00 | 10 CFR 21.21 | Potential for Reverse Polarity on Hpci Turbine Eg-R Hydraulic Actuators | GE Hitachi Nuclear Energy (GEH) has completed an evaluation of the 'Reverse Polarity on HPCI EG-R Hydraulic Actuators,' and has concluded that this is a Reportable Condition in accordance with the requirements of 10 CFR 21.21 (d). Discussion: GEH provided a refurbished HPCI turbine EG-R Hydraulic Actuator, (GEH Part number DD213A8527P003), as a safety related component, to a domestic BWR/4. When the customer installed the EG-R Hydraulic Actuator at the plant, calibration and post maintenance testing found that the turbine governor valves went to the full open position when the proper response was a fully closed position. Troubleshooting of the newly installed component revealed that the polarity of the component was reversed. An improperly configured EG-R Hydraulic Actuator cannot be utilized in the system because the reversed polarity causes the turbine governor control valves to operate in a manner opposite to the expected response, and calibration of the component by plant personnel cannot be completed. GEH contracted Engine Systems Incorporated (ESI) to perform the repair/refurbishment of this EG-R Hydraulic Actuator. This particular EG-R Hydraulic Actuator is identified as GEH part number DD213A8527P003. The specific EG-R Hydraulic Actuator that was identified with this defective condition was identified as serial number 2288717. Conclusion: This condition would change the operational characteristics of the HPCI system and would create a Substantial Safety Hazard or a violation of a Technical Specification Safety Limit. As such this condition has been determined to be a Reportable Condition within the context of 10 CFR Part 21.21 (d). ABWR and ESBWR Design Certification Documentation Applicability: The issues described above have been reviewed for applicability to documentation associated with 10CFR 52 and it has been determined that there is no affect on the technical information contained in either the ABWR certified design or the ESBWR design in certification. Recommended Action: GEH recommends that (the Hatch, Hope Creek and Peach Bottom) sites that have received EG-R Hydraulic Actuator(s) (GEH Part number DD213A8527P003), check warehouse inventory. If the EG-R Hydraulic Actuator remains 'in stock,' the potential exists that incorrect internal wiring could exist resulting in the EG-R Hydraulic Actuator not responding as expected. GEH recommends that if an EG-R Hydraulic Actuator (GEH Part number DD213A8527P003) is in warehouse stock, that the component be returned to GEH for verification of the internal wiring configuration. | ||
| ENS 46348 | 20 October 2010 04:00:00 | 10 CFR 21.21 | Part 21 - Crack Indications in Marathon Control Rod Blades | The following was received via facsimile: A recent inspection of near 'End-of-Life' Marathon Control Rod Blades (CRB) at an international BWR/6 has revealed crack indications. The CRB assemblies in question were manufactured in 1997. GE Hitachi Nuclear Energy (GEH) continues to investigate the cause(s) of the crack indications. Once the cause of the crack indications is determined, GEH will evaluate the nuclear and mechanical lifetime limits of the Marathon Control Rod Blade design in light of the new inspection data, and make revised lifetime recommendations, if necessary. This 60-day interim notification, in accordance with 10CFR Part 21.21(a)(2), is sent for all plants that are D lattice, BWR/2-4 or S lattice, BWR/6 plants. Since there have been no reported cracking occurrences in C lattice assemblies to date, these CRBs are tentatively eliminated from the investigation. C lattice, BWR/4-5 plants have been included on Attachment 2 for identification. Should the results of the investigation implicate the C lattice plants, the final resolution to this 10CFR Part 21 evaluation will include the C lattice plants. The D lattice and S lattice plants in the US that are affected by this notification include Nine Mile Point, Unit 1; Millstone, Unit 1; Fitzpatrick; Pilgrim; Vermont Yankee; Grand Gulf; River Bend; Clinton; Oyster Creek; Dresden, Unit 2; Dresden, Unit 3; Peach Bottom, Unit 2; Peach Bottom, Unit 3; Quad Cities, Unit 1; Quad Cities, Unit 2; Perry, Unit 1; Duane Arnold; Cooper; Monticello; Brunswick, Unit 1; Brunswick, Unit 2; Hatch, Unit 1; Hatch, Unit 2; Browns Ferry, Unit 1; Browns Ferry, Unit 2; and Browns Ferry, Unit 3.
In August 2010, GE Hitachi (GEH) performed the planned inspection of four near 'End-of-Life' CRBs at 'Plant O.' The inspection revealed crack indications on all four Control Rod Blades (CRBs). The observed cracks are much more numerous, and have more material distortion than previously observed. Further, the cracks occur at a much lower reported local B-10 depletion than previously observed, with cracking predominantly starting at approximately 40% local depletion, whereas previous inspections observed cracking only above 60% local depletion. The cracks at 'Plant O' are also more severe, in that they resulted in missing capsule tube fragments from two of the inspected CRBs. A lost parts analysis performed for 'Plant O' determined that there is no negative affect on plant performance due to the missing tube fragments. At this point in the investigation, no causal or contributing factors unique to the 'Plant O' CRBs, nor their operation, has been identified. Including the inspections at 'Plant O,' GEH has now completed the visual inspection of 97 irradiated Marathon CRBs, with 10 showing crack indications. As 'Plant O' is an S lattice design, all crack indications are still confined to D and S lattice applications, with no crack indications on C lattice designs. When considering only D and S lattice applications that are near 'End-of-Life' depletion limits, 10 of 23 control rod inspections have revealed crack indications. Notified R1DO (Schmidt), R2DO (Shaeffer), R3DO (Ring), R4DO (Powers) and Part 21 Group.
Subject: Part 21 Reportable Condition Notification: Design Life of D and S Lattice Marathon Control Blades GE Hitachi Nuclear Energy (GEH) has completed its evaluation of the cracking of Marathon Control Rod Blades (CRB) at an international BWR/6. This issue was initially reported on October 20, 2010 as GEH letter MFN 10-327 (Reference 1). Additional information was provided on December 1, 2010 as GEH letter MFN 10-351 (Reference 2). GEH has determined that the design life, of D and S lattice Marathon Control Blades may be less than previously stated. The design life if not revised, could result in significant control blade cracking and could, if not corrected, create a substantial safety hazard and is considered a reportable condition under 10 CFR Part 21.21 (d). Marathon C lattice Control Blades are not affected by this condition. The information contained in this document informs the NRC of the conclusions and recommendations derived from GEH's investigation of this issue. Notified R1DO (Ferdas), R2DO (McCoy), R3DO (Kozak), R4DO (Gaddy) and Part 21 Group. | Control Rod | |
| ENS 46230 | 3 September 2010 04:00:00 | 10 CFR 21.21 | Part 21 - Failure to Include Seismic Input in Reactor Control Blade Customer Guidance | The following is text of a facsimile submitted by the vendor: GE Hitachi Nuclear Energy (GEH) has identified that engineering evaluations that support the guidance provided in SC 08-05, Revision 1, do not address the potential impact of a seismic event on the ability to scram as it relates to the channel-control blade interference issue. Note that the seismic loads are not a consideration in the scram timing, but rather the ability to insert the control blades. In other words, the control blades must be capable of inserting during the seismic event, but not to the timing requirements of the Technical Specifications. GEH is evaluating the impact of the seismic loads between the fuel channel and the control blade associated with an Operating Basis Earthquake (OBE), and a Safe Shutdown Earthquake (SSE) on BWR/2-5 plants. The scram capability is expected to be affected due to the added seismic loads at low reactor pressures in the BWR/2-5 plants. The ability to scram for the BWR/6 plants is not adversely affected by the seismic events. Additional evaluation is required to determine to what extent the maximum allowable friction limits specified for the BWR/2-5 plants in SC 08-05 Revision 1 is affected by the addition of seismic loads. GEH issues this 60-Day Interim Report in accordance with the requirements set forth in 10 CFR 21.21 (a)(2) to allow additional time to for this evaluation to be completed. Affected US plants previously notified by vendor and recommended for surveillance program include: Nine Mile Point, Units 1 and 2; Fermi 2; Columbia; FitzPatrick; Pilgrim; Vermont Yankee; Grand Gulf; River Bend; Clinton; Oyster Creek; Dresden, Units 2 and 3; LaSalle, Units 1 and 2; Limerick, Units 1 and 2; Peach Bottom, Units 2 and 3; Quad Cities, Units 1 and 2; Perry, Unit 1; Duane Arnold; Cooper; Monticello; Brunswick, Units 1 and 2; Hope Creek; Hatch, Units 1 and 2; and Browns Ferry, Units 1and 2. Affected US plants previously notified by vendor and provided information include: Susquehanna, Units 1 and 2 and Browns Ferry, Unit 3.
The following update was received via fax: This letter provides a revision to the information transmitted on September 2, 2010 in MFN 10-245 concerning an evaluation being performed by GE Hitachi Nuclear Energy (GEH) regarding the failure to include seismic input in channel-control blade interference customer guidance. Two changes have been made in Revision 1: 1) A statement was added regarding the applicability of this issue to the ABWR and ESBWR design certification documentation. 2) The original MFN 10-245 referenced the Safety Communication SC 08-05 R1 that was transmitted to the US NRC via MFN 08-420. The references to SC 08-05 were changed to MFN 08-420 to prevent possible confusion. As stated herein, GEH has not concluded that this is a reportable condition in accordance with the requirements of 10CFR 21.21(d) and continued evaluation is required to determine the impact of a seismic event on the guidance contained in MFN 08-420. Notified the R1DO (Gray), R2DO (Hopper), R3DO (Orth), R4DO (Farnholtz), NRR EO (Lee) and Part 21 Group (via email).
The following update was received via fax: This letter provides information concerning an on-going evaluation being performed by GE Hitachi Nuclear Energy (GEH) regarding the failure to include seismic loads in the guidance provided in MFN 08-420. As stated herein, GEH has not concluded that this is a reportable condition in accordance with the requirements of 10CFR21.21(d) and continued evaluation is required to determine the impact of a seismic event on the guidance contained in MFN 08-420. GEH has not completed the evaluation of the impact of the seismic loads between the fuel channel and the control blade associated with an Operating Basis Earthquake (OBE), and a Safe Shutdown Earthquake (SSE) on BWR/2-5 plants. GEH expects the task to be completed by August 15, 2011. Notified the R1DO (Holody), R2DO (Henson), R3DO (Kozak), R4DO (Werner), NRR EO (Evans) and Part 21 Group (via email).
The following was received via fax: GE Hitachi Nuclear Energy (GEH) identified, in July 2010, that engineering evaluations did not address the potential impact of a seismic event on the ability to scram as it relates to the channel-control blade interference issue. GEH provided status of the on-going evaluation in (December 2010). GEH has not completed the evaluation of the impact of the seismic loads between the fuel channel and the control blade associated with a bounding Safe Shutdown Earthquake (SSE) on BWR/2-5 plants. The scram capability is expected to be affected due to the added seismic loads at low reactor pressures (less than 1000 psig) in the BWR/2-5 plants. Additional evaluations are required to determine to what extent the maximum allowable friction limits specified for the BWR/2-5 plants are affected by the addition of SSE seismic loads at low reactor pressures. GEH issues this 60-Day Interim Report in accordance with the requirements set forth in 10CFR 21.21 (a)(2) to allow additional time for this evaluation to be completed. The following sites are noted as having channel-control blade concerns: Region 1: Nine Mile Point, Fitzpatrick, Pilgrim, Vermont Yankee, Oyster Creek, Limerick, Peach Bottom, Susquehanna, and Hope Creek Region 2: Browns Ferry, Brunswick, Hatch, Region 3: Fermi, Clinton, Dresden, LaSalle, Quad Cities, Perry, Duane Arnold, Monticello Region 4: Columbia, Grand Gulf, River Bend, Cooper. Notified R1DO (Powell), R2DO (Hopper), R3DO (Dickson), R4DO (Farnholtz) and NRR Part 21 Grp via email.
The following is a summary of information received from GE Hitachi Nuclear Energy via e-mail of a letter, Reference MFN 10-245 R4, addressed to the NRC and dated September 26, 2011: GE Hitachi (GEH) has determined that the scram capability of the control rod drive mechanism in BWR/2-5 plants may not be sufficient to ensure the control rod will fully insert in a cell with channel-control rod friction at or below the friction limits specified in MFN 08-420 with a concurrent Safe Shutdown Earthquake (SSE). The plant condition for which incomplete control rod insertion might occur is when the reactor is below normal operating pressure (<900 psig) and a scram occurs concurrent with the SSE, for Mark I containment plants, and for the SSE with concurrent Loss-of-Coolant Accident (LOCA) and Safety Relief Valve (SRV) events for Mark II containment plants. In this scenario a Substantial Safety Hazard results because the affected control rods might not fully insert to perform the required safety function. GEH has determined that when channel-control blade interference is present at reduced reactor pressure and at friction levels considered acceptable in MFN 08-420, a simultaneously occurring Safe Shutdown Earthquake (SSE) may result in control rod friction that inhibits the full insertion of the affected control rods during a reactor scram from these conditions. This scenario was not explicitly considered in MFN 08-420. GEH has also quantified maximum allowable control rod friction for channel-control blade interference during the SSE with reactor system pressure greater than or equal to 900 psig. The previous conclusion regarding the scram capability for the BWR/2-5 plants, last communicated in MFN 10-245 R2, was based upon a reactor system pressure of 1000 psig. The updated evaluation at 900 psig has resulted in modifications to the guidance specified in MFN 08-420. The GE Hitachi Letter recommends testing with new allowable friction limits that will ensure control rods fully insert at low reactor pressure concurrent with an SSE (for Mark I containment plants) and SSE with concurrent LOCA (for Mark II containment plants). The enclosure in the GEH letter provides a description of the evaluation, with surveillance recommendations for BWR/2-5 plants. The recommended surveillance is intended to augment the surveillance requirements in the plant Technical Specifications and define populations of control rods to be tested, and the method for testing, until other actions that mitigate or limit the potential for channel control blade interference can be identified and implemented. Based upon the evaluation, GEH has concluded that a Reportable Condition under 10CFR Part 21 exists for BWR/2-5 plants. This determination does not apply to BWR/6 or ABWR plants or the ABWR/ESBWR Design Control Document's (DCD). The information contained in this document informs the NRC of the conclusions and recommendations derived from GEH's evaluation of this issue. The list of potentially affected plants has previously been noted in this Part 21 notification and have been previously notified by GE Hitachi of the concern. Notified R1DO (Doerflein), R2DO (Lesser), R3DO (Passehl), R4DO (Werner) and NRR Part 21 Grp via email.
GE Hitachi Nuclear Energy (GEH) provided an update to its guidance and supporting evaluations that were reported in MFN 10-245 R4 on September 26, 2011. Notified R1DO (Burritt), R2DO (Calle), R3DO (Giessner), R4DO (Camplbell) and Part 21 Group via email.
GE Hitachi Nuclear Energy (GEH) provided an update to its guidance and supporting evaluations that were reported in MFN 08-420 R0 on December 19, 2008 and MFN 10-245 R5 on February 7, 2011. Notified R1DO (Dimitriadis), R2DO (Rose), R3DO (Riemer), R4DO (Lantz) and Part 21 Group via email. | Safety Relief Valve Control Rod | |
| ENS 45927 | 15 April 2010 04:00:00 | 10 CFR 21.21 | Bent Fuel Spacer Flow Wing | During inspection of GNF2 reload fuel, a spacer flow wing on the corner rod position was discovered to be deformed (bent). A review of this condition and the associated root cause evaluation has determined that it could be present in previously manufactured GNF2 fuel that has been shipped for Fitzpatrick Cycle 19, Pilgrim Cycle 18, Vermont Yankee Cycle 28, Vermont Yankee GNF2 Lead Use Assemblies and Grand Gulf Cycle 18. It is not known that this condition exists in the GNF2 fuel for these plants, but it cannot be ruled out. A conservative assessment of thermal hydraulic impact of this condition resulted in a 0.01 OLMCPR (Operating Limit Minimum Critical Power Ratio) impact for these plants. An OLMCPR impact of 0.01 is at the threshold for reportability. | ||
| ENS 46060 | 1 July 2009 04:00:00 | 10 CFR 21.21 | Part 21 Report Concerning Failure of Turbine Overspeed Reset Control Valve Diaphragm | The information below is a summary of a report received via facsimile from GE Hitachi; Report MFN 10-192 dated July 1, 2010. Background: A diaphragm used in a 1" HPCI turbine stop valve / mechanical trip hold valve operator failed at a domestic BWR 4 in July 2009. The failure resulted in a HPCI turbine lube oil leak, which was the indication that the diaphragm had failed. The BWR 4 plant completed an Apparent Cause Evaluation and concluded that a material defect in the diaphragm allowed the diaphragm to tear after being installed for 2 years 8 months. The diaphragm that failed was a Robertshaw (RS) part number 25471-A2, and was installed in a Robertshaw model VC-210 diaphragm control valve operator. The diaphragm was made from Buna-n rubber and was designed to have two layers of Dacron reinforcement fabric over all pressure bearing surface areas of the diaphragm. The diaphragms are manufactured by Chicago-Allis using a 2-plate compression mold process. The diaphragms are purchased as commercial grade and are dedicated by GEH and supplied as safety related under GE part number Q25471-A2. The failed diaphragm was manufactured in 2006. Discussion: Reinforcement fabric is considered a critical design requirement that is essential to ensure durability, reliability, and prevents tearing of the diaphragm material when these diaphragms are used in the HPCI turbine lube oil system as turbine trip and reset valves. An inspection was performed on six diaphragms, three manufactured in 2006 and three manufactured in 2008. All six of these diaphragms were found to have areas without fabric reinforcement. Inspection of the three samples from 2006 found non-uniform reinforcement. Inspection of the three samples from 2008 found all diaphragms were void of reinforcement in the sidewalls and inspection indicates that the reinforcement fabric was torn away from the inner sidewall during the manufacturing process. The inspections identified no diaphragms that were in full compliance with the design requirements for two layers of reinforcing fabric over all pressure bearing surfaces of the diaphragm. Safety Analysis: The failure of the HPCI turbine over-speed reset control valve's diaphragm would result in a loss of HPCI turbine lube and control oil through the failed diaphragm. Depending on the amount of oil lost and the system demands, this loss could ultimately result in a failure of the HPCI System. Failure is not imminent, but cannot be precluded. Other safety related equipment is sufficient to mitigate design basis events in the event of a loss of HPCI. Conclusion: Because of the similarity of the defects in all diaphragms inspected, it is credible to believe that this type of deviation from technical requirement also exists in other diaphragms manufactured by Chicago Allis and sold by GE as part number Q25471-A2 and 25471-A2Q, and as part of Control Valve Assembly DD233A3600P001. The identified defective diaphragms were present in two lots; one manufactured in 2006 and one in 2008. Based on the observations it is reasonable to believe that other diaphragms manufactured in 2006 and 2008 have similar deviations. GEH has been unable to determine if the identified manufacturing deviation exists in diaphragms manufactured prior to 2006. Since GEH is not able to rule out defects in diaphragms manufactured prior to 2006, it is credible to believe that similar deviations existed in diaphragms manufactured prior to 2006. In order to determine the possible extent of condition, all diaphragms in service or in stock at plants as spare parts inventory are suspect. Since the diaphragms have a designated service life of 5 years, and a shelf life of 10 years, the extent of condition is bounded by replacement of all diaphragms purchased by plants since 1995. GEH has evaluated the consequences of the failure of this diaphragm and concluded that this type of failure could result in the HPCI system not performing its safety function. The HPCI system is considered an essential safety related system. Failure of the HPCI system is considered a major degradation of essential safety related equipment. Therefore this condition is determined to be a Substantial Safety Hazard and is a Reportable condition per 10CFR Part 21. Recommended Action: GEH has evaluated the consequences of the failure of this diaphragm and concluded that this type of failure could result in the HPCI system not performing its safety function. The HPCI system is considered an essential safety related system. Failure of the HPCI system is considered a major degradation of essential safety related equipment. Therefore this condition is determined to be a Substantial Safety Hazard and is a Reportable condition per 10CFR Part 21. US Plants With Affected Diaphragms: Fermi 2 Limerick Peach Bottom Duane Arnold Cooper Susquehanna Brunswick Hatch Browns Ferry |