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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| 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 |