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| Entered date | Event description | |
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| ENS 49974 | 31 March 2014 09:07:00 | The following summary was excerpted from GE Hitachi Interim Part 21 Report received via email: A potential analysis error has been identified that is associated with the ABWR (Advanced Boiling Water Reactor) hydrodynamic loads determined by using the Technical Specification Suppression Pool High Water Level (HWL) as an analysis input condition. Vessel coolant inventory is transferred into the containment Suppression Pool during a postulated LOCA blowdown, thereby increasing the Suppression Pool water level. The correction in the analysis may lead to a Suppression Pool water level greater than what is currently assumed in structural analyses which apply the containment hydrodynamic loads generated during a postulated LOCA event. Facility Identification: South Texas Project Units 3 and 4, Clinton ESP Site, Grand Gulf ESP Site, North Anna ESP Site, and includes the ESP application for the PSEG Site and Victoria County Station ESP application. If you have any questions, then contact: Dale E. Porter, GE-Hitachi Nuclear Energy Americas LLC, Ph. #(910) 819-4491.
June 26, 2014 MFN 14-013 R1 This letter provides supplemental information concerning an evaluation being performed by GE Hitachi Nuclear Energy (GEH) regarding the potential increase in hydrodynamic loads that may be experienced by containment structures during a postulated Loss of Coolant Accident (LOCA) associated with Reference 1, and requests additional time to complete the evaluation for the determination of reportability of this condition. A potential analysis error has been identified that is associated with the ABWR hydrodynamic loads determined by using the Technical Specification Suppression Pool High Water Level (HWL) as an analysis input condition. Vessel coolant inventory is transferred into the containment Suppression Pool during a postulated LOCA blowdown, thereby increasing the Suppression Pool water level. The correction in the analysis may lead to a Suppression Pool water level greater than what is currently assumed in structural analyses which apply the containment hydrodynamic loads generated during a postulated LOCA event. For example, a postulated Feedwater Line Break (FWLB) may transfer a large quantity of FW liquid into the Suppression Pool with a notable increase in pool water level, even assuming a portion of the discharged fluid spills over into the lower drywell region of the ABWR containment. A higher Suppression Pool water level may result in increased hydrodynamic loads acting on the submerged walls and structures in the containment. The higher Suppression Pool water level can extend the wetted regions of the Suppression Pool walls and the ABWR access tunnel, as well as result in wetted submerged structure segments that were not previously considered wetted. This potential analysis error affects the LOCA containment hydrodynamic loads including condensation oscillation (CO) and chugging, as well as Safety Relief Valve (SRV) actuation loads. Assessing the overall impact of increased hydrodynamic loads calculated with higher Suppression Pool water level requires an evaluation of the containment structural components' design bases. GEH is in the process of examining revised containment loads, and determining available margin in the ABWR containment component design specifications to accommodate potentially increased load source forcing functions. ABWR plants may then compare affected plant-specific containment structural design bases to these specifications for relative margin. An extended time period is needed in order to complete the revised containment load determination and evaluate the impact on containment structures. GEH is requesting additional time to complete the analysis previously noted in Reference 1. The information required for this GEH 60-Day Interim Report Notification per �21.21(a)(2) is provided in Attachment 1. The commitment for follow-on actions is provided in Attachment 1, item (vii). If you have any questions, please call me at (910) 819-4491. Sincerely, Dale E. Porter Safety Evaluation Program Manager GE-Hitachi Nuclear Energy Americas LLC Notified R2DO (Rich) and Part 21 Reactor Group via email.
August 29, 2014, MFN 14-013 R2, Specification 000N7289-R2 When considered with the realistic assumptions and the lowered scale factors, the condition reported in Reference 1 (of the final report) is determined as non-reportable, and there is no Substantial Safety Hazard nor will it lead to exceeding a Technical Specification Safety Limit for the affected plants and plant designs. The GEH evaluation within 10 CFR Part 21 is now closed. If you have any questions, please call; Dale E. Porter Safety Evaluation Program Manager GE-Hitachi Nuclear Energy Americas LLC Ph. (910) 819-4491. Notified R2DO (Rose) and Part 21 Reactor Group via email. |
| ENS 48735 | 8 February 2013 16:32:00 | 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. |
| ENS 48350 | 27 September 2012 11:01:00 | 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. |
| ENS 47630 | 1 February 2012 15:33:00 | 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). |
| ENS 46429 | 19 November 2010 08:33:00 | 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 12:54:00 | 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. |
| ENS 45927 | 14 May 2010 14:11:00 | 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 2010 09:27:00 | 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 |