ML11292A192
| ML11292A192 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 01/11/2012 |
| From: | Mark King NRC/NRR/DIRS/IOEB |
| To: | |
| King, Mark | |
| References | |
| Download: ML11292A192 (54) | |
Text
Page 1 of 54 RECENT NRC INSPECTION FINDINGS RELATED TO SEISMIC RELATED ISSUES -
issued 1/1/2009 to 4/1/2011 - (72 findings)
NOTE: This NRC Inspection findings document is referenced from and supports Information Notice 2012-01 SEISMIC CONSIDERATIONSPRINCIPALLY ISSUES INVOLVING TANKS issued under Agencywide Documents and Management System (ADAMS) Accession No. ML1129A175 Finding Cross Cutting Areas:
- SCWE - Safety Conscious Work Environment
- HP - Human Performance
- PIR - Problem Identification and Resolution Active hyperlinks are provided to the inspection reports discussing each finding in more detail.
1)
Mitigating 09/30/2009 FARLEY Green *SCWE: N *HP: Y *PIR: N Systems Docket/Status: 05000348 (C)
Open: 2009004 (PIM) Failure to Implement Procedures Used to Prevent Seismic Interaction with Safety-Related Equipment Green. A self-revealing finding was identified for a failure to implement procedures to ensure that temporary equipment carts were immobilized in order to prevent inadvertent contact with safety related equipment. Specifically, the Unit 1 H bus protective relay cabinet resulting in inoperability of the 1C Emergency Diesel Generator (EDG). This finding was entered into the licensees CAP as CR 2009101710. Failure to implement a procedure to ensure temporary equipment carts were immobilized to prevent inadvertent contact with the Unit 1 H bus protective relay cabinet was a performance deficiency. This finding was greater than minor because it adversely affected the equipment reliability attribute of the mitigating systems cornerstone objective to ensure the availability, reliability and capability of systems responding to initiating events to prevent undesirable consequences (i.e., core damage). This finding was assessed using the Phase 1 screening worksheet of the SDP and determined to be of very low safety significance (Green) because it did not result in an actual loss of safety function of a single train for greater than the TS allowed outage time. The finding did not involve a total loss of any safety function, as identified by the licensee through a Probabilistic Risk Assessment (PRA), Individual Plan Examination of External Events (IPEEE), or similar analysis, contributing to external event-initiated core damage accident sequences (i.e., initiated by a seismic, flooding, or severe weather event). The inspectors identified a cross-cutting aspect in the human performance area of work practices component (H.4(b)). The licensee had established a procedure requiring all wheeled items left in safety-related areas be made incapable of rolling and personnel did not follow the procedure. The procedure the licensee failed to implement was not safety related, therefore, the performance deficiency did not result in a violation of regulatory ML11292A192
Page 2 of 54 requirements.
2)
Initiating Events 09/30/2009 BYRON Green *SCWE: N *HP: Y *PIR: N Docket/Status: 05000454 (C) , 05000455 (C)
Open: 2009004 (PIM) INADEQUATE EVALUATION OF SEISMIC RESTRAINT ON THE FHB CRANE TROLLEY A finding of very low safety-significance and associated Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for failure to perform an adequate evaluation of seismic restraint on the Fuel Handling Building (FHB) crane trolley. Specifically, for evaluation of the seismic restraint in their single failure proof trolley analysis, the licensee failed to use adequate seismic acceleration values and failed to evaluate the connections for resulting reaction forces. Subsequent review found that the restraint was inadequate. The licensee documented the condition in Issue Report (IR) 934467 and initiated actions for calculation revision and installation of a field modification. The inspectors determined that the failure to perform an adequate analysis for the seismic restraint and its connections for seismic loads was contrary to American Society of Mechanical Engineers (ASME) NOG-1-2004, requirements and was a performance deficiency. The FHB crane is designed to Seismic Category I requirements and the licensee used compliance with ASME NOG-1-2004, as the design basis for their upgrade to a single failure proof crane. The finding was more than minor because it was associated with the Initiating Events cornerstone attribute of Equipment Performance, Refueling/Fuel Handling equipment, and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors evaluated the finding using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, and based on a No answer to all the questions in the Initiating Events column of Table 4a, determined the finding to be of very low safety-significance (Green). This finding has a cross-cutting aspect in the area of Human Performance, Work Practices because the licensee did not provide adequate oversight of work activities, including contractors, such that nuclear safety is supported. H.4(c) 3)
Initiating 12/31/2009 BRAIDWOOD Green *SCWE: N *HP: Y *PIR: N Events Docket/Status: 05000456 (C) , 05000457 (C)
Open: 2009005 (PIM) FAILURE TO FULLY IMPLEMENT ABNORMAL OPERATING PROCEDURES ML11292A192
Page 3 of 54 FOLLOWING A SEISMIC EVENT The inspectors identified a Green finding and an associated Non-Cited Violation of Technical Specification 5.4.1 for the failure to fully implement an abnormal procedure following a seismic event. Specifically, on April 18, 2008, following a seismic event, the licensee chose to perform field walkdowns to verify that sulfuric acid and sodium hypochlorite tanks were intact rather than to isolate control room ventilation as required by Procedure 0BwOA ENV-4, Earthquake. As a corrective action, the licensee performed training activities to clarify when procedural deviations are allowed. The finding was determined to be more than minor because it impacted the procedure quality attribute of the Initiating Events Cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors evaluated the finding in accordance with IMC 0612, Appendix B, Issue Screening. The inspectors performed a significance evaluation in accordance with IMC 0609, Attachment 4, Determining the Significance of Reactor Inspection Findings for At-Power Situations. The inspectors answered No to the external event initiators question in the Initiating Events Cornerstone column of Table 4a and the issue screened as one of very low safety significance. This finding is associated with the cross-cutting attribute of decision making in the Human Performance cross-cutting component (H.1(b)). Specifically, the licensee did not use conservative assumptions in the decision to send an operator to locally verify rather than perform a procedural step from the control room as written. In the event the sulfuric acid and sodium hypochlorite tanks were damaged, the control room operators could have been impacted with chlorine gas prior to receiving verification from the locally dispatched operator since the licensee elected not to isolate control room ventilation.
4)
DIABLO Events 12/31/2009 Green *SCWE: N *HP: N *PIR: Y CANYON Docket/Status: , 05000323 (C)
Open: 2009005 (PIM) Less Than Adequate Replacement Reactor Head Modification Design Control The inspectors identified a noncited violation of Title 10 CFR, Part 50, Appendix B, Criterion III, Design Control, after the design contractor failed to perform adequately calculations demonstrating that the replacement reactor head met ASME Code acceptance criteria. The contractor failed to use the critical seismic damping values specified in the plant design basis for the design of the integrated head assembly and the control rod drive mechanism housing assembly and when calculating component stress during a postulated design basis earthquake.
The licensee entered this condition into the corrective action program as Notifications 50276107 and 50276288. The inspectors concluded that the failure to properly implement the plant design basis in the replacement head design was a performance deficiency. The finding is more than minor because the performance deficiency is associated with the Initiating Events Cornerstone design control attribute and adversely affected the cornerstone objective to limit the likelihood of loss of a coolant accident during a seismic event. The inspectors determined the finding is of ML11292A192
Page 4 of 54 very low safety significance because assuming worst case degradation, the finding would not result in exceeding the Technical Specification limit for reactor coolant system leakage nor have likely affected other mitigation systems resulting in a total loss of their safety function. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not identify the use of improper damping values with a low threshold for identifying issues during oversight of contractor activities and design reviews P.1(a).
5)
Initiating DIABLO 12/31/2009 SL-IV *SCWE: N *HP: N *PIR: Y Events CANYON Docket/Status: , 05000323 (C)
Open: 2009005 (PIM) Lees Than Adequate Change Evaluation to the Facility as Described in the Final Safety Analysis Report Update The inspectors identified a Severity Level IV noncited violation of 10 CFR 50.59 after the licensee failed to perform an adequate evaluation to demonstrate that prior NRC approval was not required before making changes to the facility as described in the Final Safety Analysis Report Update. In October 2009, the inspectors identified that the replacement reactor head contractor used incorrect damping values in the replacement head design. The contractor was unable to demonstrate that the design met ASME Code using the damping values specified in the plant design basis. On November 5, 2009, the licensee incorporated the new damping values and revised the method for determining the seismic response spectra. Using NEI 96-07, Guidelines for 10 CFR 50.59 Evaluations, Revision 1, the inspectors concluded that these changes resulted in a departure from a method of evaluation described in the Final Safety Analysis Report Update establishing the facility design bases. The licensees 50.59 evaluation, Licensing Basis Impact Evaluation LEBE 2009-021, Integrated Head Assembly, was less than adequate to conclude that prior NRC approval was not required for the changes. The licensee entered this issue into their corrective action program as 50276288. The failure of Pacific Gas and Electric to perform an adequate 10 CFR 50.59 evaluation prior to changing the facility as described in the Final Safety Analysis Report Update is a performance deficiency. The inspectors evaluated this issue using the traditional enforcement process because the performance deficiency had the potential for impacting the NRCs ability to perform its regulatory function. The inspectors concluded that the issue was more than minor because of a reasonable likelihood the change to the facility would require Commission review and approval prior to implementation. The inspectors also evaluated this issue using the Significance Determination Process. The inspectors concluded that the violation affected the Initiating Events Cornerstone because the change potentially decreased the structural integrity of the control rod drive mechanism reactor coolant pressure barrier and screened Green because assuming worst case degradation, the finding would not result in exceeding the technical specification limit for reactor coolant system leakage nor have a likely effect on other mitigation systems resulting in a total ML11292A192
Page 5 of 54 loss of their safety function. The inspectors concluded that the violation was a Severity Level IV because the issue screened Green under the Significance Determination Process. The finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not thoroughly evaluate the original problem associated with the replacement reactor head design such that the resolutions address causes and extent of conditions, as necessary P.1(c).
6)
Initiating DUANE 03/31/2010 Green *SCWE: N *HP: N *PIR: N Events ARNOLD Docket/Status: 05000331 (C)
Open: 2010002 (PIM) INADEQUATE EVALUATIONS FOR CRANE AND SPECIAL LIFTING DEVICES.
A finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for deficiencies in the design documents for the reactor building crane and the special lifting devices. Specifically, the crane bridge girder rails supporting the trolley were not evaluated for the design basis seismic loads.
In the reactor vessel head special lifting device calculation, the licensee did not evaluate the hook pins and the calculated safety factors did not meet the design criteria. In the dryer/separator special lifting device calculation, the licensee used incorrect stress allowable values. The licensee documented the condition in their Corrective Action Programs (CAPs) as CAPs 072917, 072568, 072885 and 072880, and initiated actions for calculation revisions and/or modifications. The inspectors determined that not evaluating bridge girder rails for seismic loads in accordance with NUREG 0554, not evaluating the hook pins and accepting safety factors not meeting the design criteria and American National Standards Institute (ANSI)
Standard N14.6 on the reactor vessel head special lifting device, and the inadequate calculation of safety factors on the dryer/separator special lifting device in accordance with ANSI N14.6 was a performance deficiency. The finding was more than minor because it was associated with the Initiating Events Cornerstone attribute of Equipment Performance and affected the cornerstone objective to limit the likelihood of those events that upset the plant stability and challenge critical safety functions during shutdown as well as power operations. For the item associated with the crane rail, the Region III Senior Risk Analyst (SRA) performed an SDP Phase 3 risk assessment for estimating the frequency of occurrence of an Operating Basic Earthquake (OBE) or higher seismic event during use of reactor building crane and concluded that the issue was of very low risk significance (Green). For the item associated with the special lifting devices, the inspectors evaluated the finding using IMC 0609.04, Phase 1 - Initial Screening and Characterization of Findings, and based on a No answer to all the questions in the Initiating Events column of Table 4a, as the licensee demonstrated adequate safety factors on all components through subsequent evaluations, determined the finding to be of very low safety significance (Green).
The inspectors did not identify any cross cutting aspects associated with this finding because, based on the age of the performance deficiencies, it was not reflective of the current licensee ML11292A192
Page 6 of 54 performance.
7)
Initiating THREE MILE 09/30/2010 Green *SCWE: N *HP: Y *PIR: N Events ISLAND Docket/Status: 05000289 (C)
Open: 2010004 (PIM) Deficient Control of Transient Material in Seismic Class I Buildings The inspectors identified a Green non-cited violation (NCV) of Technical Specification 6.8.1 for failure to properly control and store transient material within seismic Class I buildings such that the equipment did not pose a hazard to nuclear safety or safe plant operation. Specifically, an extension ladder and a maintenance tool cart were left unattended and unsecured in close proximity to the spent fuel pool cooling piping within the fuel handling building and near intermediate cooling pump IC-P-1A and intermediate cooling supply valve IC-V-4 in the auxiliary building, respectively. Operators promptly initiated actions to remove the subject material.
During subsequent plant tours the inspectors identified numerous additional examples of improperly controlled transient material. The licensee promptly corrected the identified individual discrepancies and initiated issue reports (IRs) 1095403 and 1122633 to address this performance deficiency. The transient material posed a potential hazard to safe shutdown and safety related equipment operation during a seismic event. Cooling water supplies to the spent fuel pool, the reactor coolant pump (RCP) thermal barriers, and control rod drive mechanisms (CRDM) were potentially affected. The dominant risk associated with this performance deficiency is the increased likelihood of a loss of coolant accident or forced plant shutdown. This finding is more than minor because it affected the equipment performance attribute of the Initiating Events cornerstone. The issue was also similar to IMC 0612, Appendix E, Examples of Minor Issues, example 4.k which stated the issue was more-than-minor because it involved a credible (seismic) scenario in which the transient materials could affect equipment important to safety. This finding was of very low safety significance because it did not involve loss or degradation of equipment specifically designed to mitigate a seismic event, and did not involve total loss of a safety function that contributes to external event-initiated core damage accident sequences. The finding had a cross-cutting aspect in the area of Human Performance, Work Practices component because station personnel did not follow procedures for equipment storage and housekeeping within seismic Class I buildings H.4(b).
8)
Initiating 10/05/2010 SAN ONOFRE Green *SCWE: N *HP: Y *PIR: N Events Docket/Status: , 05000361 (C)
Open: 2010005 (PIM) Failure to Properly Store C-Panels in the Radwaste Building ML11292A192
Page 7 of 54 The team identified a Green noncited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure of plant personnel to follow site procedures that controlled equipment storage in the radwaste building. Specifically, on October 5, 2010, inspectors identified that plant personnel failed to follow Procedure SO23-XX-31, Control of Work and Storage Areas within the Protected Area during Unit Outages at SONGS 2 and 3, Revision 0, by improperly storing portable electrical equipment panels outside an approved laydown area. The portable electrical equipment panels were tied-off near a hydrogen supply line which could have been damaged during a seismic event. Consequently, a hydrogen fire could have damaged trains A and B safety related equipment cables in the overhead, but sufficient train A cables were free of the area to permit a safe shutdown. A hydrogen fire was not analyzed in the San Onofre Units 2 and 3 Fire Hazards Analysis Report, because the hydrogen line was designed to withstand a seismic event. The licensee captured this performance deficiency in their corrective action program as Nuclear Notifications NNs 201142972 and 201140052. This performance deficiency is more than minor because it could adversely affect the protection against fires attribute of the initiating events cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations, and is therefore a finding. The inspectors performed the initial significance determination for the finding using the NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The inspectors transitioned to NRC Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process. However, this guidance was not well suited for this finding. A Region IV senior reactor analyst completed a Phase 3 significance determination and found that the finding was of very low safety significance. The bounding change to core damage frequency was 4E-8/year. The dominant core damage sequence included a loss of offsite power initiating event and failure of a safety relief valve to seat. The relatively low frequency of a seismic induced loss of offsite power event coupled with the remaining available equipment helped to limit the findings significance. The finding had a crosscutting aspect in the area of human performance associated with the work practices component and the self-checking theme, because personnel failed to properly check the procedural requirements prior to staging C-panels near the hydrogen line H.4(a) (Section 4OA5).
9)
Initiating 12/31/2010 LASALLE Green *SCWE: N *HP: Y *PIR: N Events Docket/Status: 05000373 (C) , 05000374 (C)
Open: 2010005 (PIM) Failure to Perform Adequate Evaluation for Reactor Building Crane Upgrade During an inspection of pre-operational testing activities of an independent spent fuel storage installation (ISFSI) at the LaSalle County Station, the inspectors identified a finding of very low safety significance with an associated Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix ML11292A192
Page 8 of 54 B, Criterion III, Design Control, for the licensees failure to perform adequate evaluations to upgrade the single failure proof crane. Specifically, the inspectors identified five examples where the licensee failed to perform adequate evaluations in accordance with American Society of Mechanical Engineers (ASME) NOG-1-2004, Rules for Construction of Overhead and Gantry Cranes (Top Running and Bridge, Multiple Girder), requirements. The Reactor Building crane is designed to Seismic Category I requirements and the licensee used compliance with ASME NOG-1-2004 as the design basis for their crane upgrade to a single failure proof crane. The inspectors determined that the failure to perform adequate evaluations was contrary to ASME NOG-1-2004 requirements and was a performance deficiency. The licensee documented the conditions in Issue Report (IR) 957014, IR 1093028, and IR 1098435 and initiated actions for calculation revisions and field modifications. The finding was of more than minor significance because the failure to perform adequate evaluations affected the licensees ability to provide reasonable assurance that loads would not be dropped during critical lifts. The inspectors evaluated the finding using Inspection Manual Chapter (IMC) 0609.04, Phase 1 - Initial Screening and Characterization of Findings, and based on a No answer to all of the questions in the Initiating Events column of Table 4a, determined the finding to be of very low safety-significance (Green). This finding has a cross-cutting aspect in the area of Human Performance, Work Practices because the licensee did not ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety is supported. (IMC 0310, H.4(c))
(Section 4OA5) 10)
Mitigating 03/28/2009 SAN ONOFRE Green *SCWE: N *HP: N *PIR: N Systems Docket/Status: , 05000361 (C) , 05000362 (C)
Open: 2009003 (PIM) Failure to Establish Adequate Scaffolding Erection Procedures The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure of engineering personnel to establish adequate procedures for scaffolding erection in safety-related areas. Specifically, Procedure SO123-I-1.34 required a minimum separation distance of 1 inch from safety-related equipment which only considered the seismic displacements of scaffolding and not other movements, such as thermal expansion of piping, equipment vibrations, or component operation. Insufficient scaffolding to component separation could result in interactions that adversely affect the safety functions of safety-related equipment. This finding was entered into the licensees corrective action program as Nuclear Notification 200366460. The finding is greater than minor because, if left uncorrected, it would have the potential to lead to a more significant safety concern. The inspectors concluded this finding was associated with the Mitigating Systems Cornerstone.
Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheets, the finding is determined to have very low safety significance because the finding did not affect both trains of any single mitigating system or represent an actual loss of a safety function. A ML11292A192
Page 9 of 54 crosscutting aspect is not assigned since the cause of the performance deficiency is not indicative of current performance.
11)
Mitigating 03/28/2009 COLUMBIA Green *SCWE: N *HP: Y *PIR: N Systems Docket/Status: , 05000397 (C)
Open: 2009002 (PIM) Failure to perform engineering evaluation to determine seismic qualification of safety-related equipment The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for Energy Northwests failure to follow procedure PPM 10.2.53, Seismic Requirements for Scaffolding, Ladders, Man-Lifts, Tool Gang Boxes, Hoists, Metal Storage Cabinets, and Temporary Shielding Racks, Revision 26. Specifically, the position of equipment is required to meet specific criteria to prevent damage to safety related equipment during a seismic event. Contrary to this procedure, the inspectors identified that equipment was routinely positioned next to safety-related equipment without a supporting engineering evaluation. This finding is greater than minor because it was a human performance error which affected the mitigating systems cornerstone objective to ensure the availability and reliability of systems that respond to initiating events to prevent undesirable consequences. This was determined to be consistent with NRC Manual Chapter 0612, Power Reactor Inspection Reports, Appendix E, Example 4.a. for being more than minor risk significance because Energy Northwest had routinely failed to perform the requisite engineering evaluation. The finding was determined to be of very low risk significance (Green) because no actual loss of safety function occurred and the finding did not screen as potentially risk significant due to external events.
Specifically, the as-found position of the equipment was determined to not adversely affect seismic qualification of the affected safety-related components. A crosscutting aspect in human performance with a work control component was identified in that Energy Northwest failed to appropriately plan work on multiple occasions, resulting in job site conditions which may have impacted plant components H.3.a]..
12)
Mitigating 03/31/2009 BRAIDWOOD Green *SCWE: N *HP: Y *PIR: N Systems Docket/Status: , 05000457 (C)
Open: 2009002 (PIM) SAFETY INJECTION PIPE SUPPORT DEFICIENCIES A finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified by the inspectors for the licensee's failure to ML11292A192
Page 10 of 54 properly evaluate the addition of lead shielding to Unit 2 safety injection piping. Specifically, the licensee did not have sufficient rationale and incorrectly concluded that sufficient margin existed in the pipe support design with the additional weight. The licensee entered the issue into their corrective action program, revised associated calculations, and planned modifications as needed to restore required design margins. The finding was determined to be more than minor because compliance with Seismic Category I design requirements was necessary to ensure the Subsystem 2SI06 pipe supports would function as required during a Seismic Category I design basis event. The finding screened as having very low safety significance because the design deficiency was confirmed not to result in loss of operability of the safety injection pipe supports.
The cause of the finding is related to the cross-cutting component of Human Performance, Resources, because the licensee did not maintain adequate design margins (H.2(a)).
13)
Mitigating 03/31/2009 BYRON Green *SCWE: N *HP: N *PIR: N Systems Docket/Status: 05000454 (C)
Open: 2009002 (PIM) FAILURE TO ADHERE TO SCAFFOLD PROCEDURES The inspectors identified a finding of very low safety significance and a non-cited violation of Technical Specification 5.4, Procedures, during a routine inspection of the Auxiliary Building on February 21. The inspectors observed scaffold construction in the containment purge area of Unit 1 that was in close proximity to a safety-related containment pressure instrument. The scaffold construction was determined to be contrary to seismic clearance procedural requirements. As part of their immediate corrective actions, licensee personnel modified the affected scaffolding. The finding was more than minor because it was associated with the Protection against External Factors attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events in order to prevent undesirable consequences. Specifically, the finding was determined to have placed scaffolding near safety related equipment in an unacceptable seismic configuration. The finding was determined to be of very low safety significance because it was determined not to represent a loss of safety function.
14)
Mitigating 03/31/2009 FERMI Green *SCWE: N *HP: N *PIR: Y Systems Docket/Status: , 05000341 (C)
Open: 2009002 (PIM) Inadequate Procedural Controls Over Construction of Storage Racks and Storage Areas ML11292A192
Page 11 of 54 A finding of very low safety significance and an associated NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the licensees failure to include criteria in procedures for evaluation of storage areas and storage racks built in the power block. Licensee procedure MOP11, Combustible Material, placed controls on the storages areas and storage racks to ensure that combustible loading remained acceptable but failed to incorporate adequate guidance for designating the storage area and constructing the racks to ensure nearby safety-related equipment would not be adversely affected during a plant transient or seismic event. After the issue was raised, modifications to the scaffold storage locations were completed, as needed. The finding was more than minor because it was associated with the Mitigating Systems Cornerstone attribute of design control (plant modifications) and it adversely impacted the cornerstone objectives. As a result of not evaluating the storage areas, safety-related components, systems or structures could have been affected. This finding was determined to be of very low safety significance because it did not result in loss of operability or functionality. The inspectors determined that the finding had an associated cross-cutting aspect of Problem Identification and Resolution, Corrective Action Program, Corrective Action (P.1 (d).
15)
Mitigating VERMONT 03/31/2009 Green *SCWE: N *HP: Y *PIR: N Systems YANKEE Docket/Status: 05000271 (C)
Open: 2009002 (PIM) Failure to perform procedurally required engineering evaluations for scaffolding.
The inspectors identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for Vermont Yankees failure to routinely perform procedurally required engineering evaluations for scaffold bracing attached to pipe supports.
Specifically, Vermont Yankee failed to perform engineering evaluations on 27 out of 32 scaffolds with horizontal bracing attached to safety related pipe supports. Subsequently, each scaffold was evaluated and documented by Vermont Yankee engineering and no immediate safety issues were found. This NCV has been entered into the Vermont Yankee corrective action program (CAP). The performance deficiency was more than minor because if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern.
Specifically, installing scaffold bracing on pipe supports without engineering approval could place a pipe support in an unanalyzed seismic condition, which could lead to failure in a seismic event. The finding had a cross-cutting aspect in the Work Practices component of the Human Performance cross-cutting area because Vermont Yankee did not implement adequate management oversight of contractor work activities regarding scaffold procedural compliance.
H.4(c). (Section 4OA2).
Page 12 of 54 16)
Mitigating GRAND 04/02/2009 Green *SCWE: N *HP: N *PIR: N Systems GULF Docket/Status: 05000416 (C)
Open: 2009006 (PIM) Inadequate Corrective Actions fo rReplacement of Safety-Related Batteries The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions," for failure to identify and correct a condition adverse to quality related to the seismic qualification of the Division III High Pressure Core Spray safety-related battery.
Specifically, the licensee failed to identify an incorrectly installed end bracket after replacement of the Division III safety-related battery in 2002 using procedures, work instructions, and drawings that were supposed to have been corrected after this same issue was identified during a 1997 battery replacement activity. The licensee has entered this into their corrective action program as CR-GGN-2009-00830. This finding was more than minor because it affected the mitigating systems cornerstone attribute of external events for ensuring the availability, reliability, and capability of systems that respond to initiating events. Also, using Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix B, Section 1-3, "Screen for More than Minor - ROP," question 2, the finding is more than minor because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheets, the finding was determined to have very low safety significance (Green) because it was confirmed to not result in a loss of operability or functionality. The finding was reviewed for cross-cutting aspects and none were identified.
17)
Mitigating 04/17/2009 SAN ONOFRE Green *SCWE: N *HP: Y *PIR: N Systems Docket/Status: , 05000362 (C)
Open: 2009003 (PIM) Improper Controls for Electrical Test Equipment The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure of maintenance planning personnel to provide adequate work instructions to control the connection of electrical monitoring devices on operable plant equipment. Specifically, the work instructions failed to require verification and functional testing after installation and removal, compliance with seismic requirements, and controls to ensure removal within the allowed time limit for a temporary installation. This finding was entered into the licensees corrective action program as Nuclear Notification 200396106.
Page 13 of 54 The finding is greater than minor because the improper controls for installation of test equipment is associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheets, the finding is determined to have very low safety significance because the finding did not affect both trains of any single mitigating system or represent an actual loss of a safety function. The finding has a crosscutting aspect in the area of human performance associated with work practices because maintenance planning personnel failed to follow procedures to develop adequate work instructions for safety-related maintenance H.4(b).
18)
Mitigating DUANE 05/22/2009 Green *SCWE: N *HP: N *PIR: Y Systems ARNOLD Docket/Status: 05000331 (C)
Open: 2009007 (PIM) FAILURE TO PROMPTLY IDENTIFY AND EVALUATE THE DEGRADED CONDITION ASSOCIATED WITH THE 'D' RWS PUMP MOUNTING BASE BOLTED CONNECTORS.
A finding of very low safety significance and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified by the inspectors for a failure of the licensee to promptly identify and correct a condition adverse to quality (CAQ) associated with the D river water supply (RWS) pump mounting base bolted connectors. The licensees failure to evaluate the operability of the D RWS pump due to the degraded bolting was considered a performance deficiency. By not examining the thread degradation documented on the overtorqued D RWS pump mounting base bolted connectors, the licensee was unable to adequately identify the as-left condition of the stud threads, evaluate the impact that condition had on the seismic qualification of the pump, and implement appropriate corrective actions to resolve the degraded condition. The failure to promptly identify and correct a CAQ associated with the safety-related D RWS pump was a violation of NRC requirements specified in 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action. The licensee entered this issue into the Corrective Action Program (CAP Item 067412), examined the pump mounting connectors, and initiated a prompt operability determination to evaluate the seismic qualification. Based on this evaluation, the D RWS pump was declared Operable but degraded. The performance deficiency was determined to be more than minor because the issue was associated with the Mitigating Systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated this finding using the Significance Determination Process (SDP) and determined the finding was of very low safety significance (Green) because this finding was a design or qualification deficiency that did not result in a loss of operability of the safety component. The inspectors also determined that this finding had a cross-cutting aspect in the area of Problem Identification and ML11292A192
Page 14 of 54 Resolution, Corrective Action Program, because the licensee did not promptly and completely identify an adverse condition in the CAP in a timely manner commensurate with its safety significance (P.1.a).
19)
Mitigating GRAND 06/23/2009 Green *SCWE: N *HP: N *PIR: N Systems GULF Docket/Status: 05000416 (C)
Open: 2009003 (PIM) Failure to Incorporate Design Changes to Protect the Standby Service Water Slab The inspectors identified a Green noncited violation of 10 CFR 50 Appendix B, Criterion III involving the failure to incorporate design changes required to limit dynamic loads on the standby service water basin slab. In 1997, the plant experienced damage to the standby service water basin slabs resulting from unanalyzed dynamic loads. During a standby service water system inspection on April 18, 2009, inspectors observed several different tire tracks on the seismically-designed concrete slab that covers and is integral to the safety-related standby service water basin. The inspectors also noted small placards attached to the basin slabs which prohibited moving vehicles on the slabs, and other signs requiring protective mats under any items placed on the slabs. Plant personnel evaluated the loading of the vehicle and determined that the load limits on the basin slab had not been exceeded. The licensee entered this issue into their corrective action program as Condition Report CR GGN 2009 002087. The inspectors determined this finding affected the design control attribute of the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In addition, the finding was more than minor because the failure to prevent dynamic loads on the standby service water basin slabs, if left uncorrected, could become more significant safety concern. Using the Manual Chapter of 0609, Significance Determination Process, Phase 1 Worksheet, this finding was determined to have very low safety significance, because it did not represent an actual loss of a safety function of the standby service water system. There is no crosscutting aspect associated with this performance deficiency since the cause of this issue occurred several years ago and does not reflect current licensee performance (Section 1R04).
20)
Mitigating THREE MILE 06/26/2009 Green *SCWE: N *HP: N *PIR: N Systems ISLAND Docket/Status: 05000289 (C)
Open: 2009006 (PIM) Failure to Assess Seismic Qualification of Stop Logs The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control.
Page 15 of 54 The team determined Exelon did not evaluate the adequacy of the river water stop logs in the Intake Screen and Pump House (ISPH) structure, to ensure that logs would not fail when exposed to seismically-induced loads. The team determined that failure of the logs would impact the capability of the safety related nuclear river water, decay river water, and reactor river water pumps to perform their design function following the seismic event. FSAR section 5.1.1 describes the ISPH and the river water systems as a Seismic Class I structure systems and components and states that this equipment should be evaluated in accordance with the methodologies decribed in the FSAR. The licensee entered this issue into the corrective action program and performed analysis which indicated the stop logs would remain in place following a seismic event. This finding is more than minor because it affected the design control attribute of the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very low safety significance (Green) because the design/qualification deficiency did not result in a loss of function. This finding was not assigned a cross-cutting aspect because the underlying cause was not indicative of current performance.
21)
Mitigating DUANE 06/30/2009 Green *SCWE: N *HP: N *PIR: Y Systems ARNOLD Docket/Status: 05000331 (C)
Open: 2009003 (PIM) FAILURE TO PROMPTLY IDENTIFY AND CORRECTA NONCONFORMING CONDITION ON A HPCI SUPPRESSION POOL SUCTION LINE SEISMIC RESTRAINT.
A finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified by the inspectors for a failure of the licensee to promptly identify and correct a condition adverse to quality (CAQ) associated with a seismic restraint on the High Pressure Coolant Injection (HPCI) Suppression Pool suction line. The licensees failure to promptly identify and correct the nonconforming condition during engineering walkdowns of the HPCI system was considered a performance deficiency. The licensee entered this issue into the Corrective Action Program (CAP) as items CAP 066713 and CAP 066750, declared the HPCI system inoperable, and isolated the HPCI Suppression Pool suction line. The seismic restraint was the repaired to return it to a fully operable condition. The performance deficiency was determined to be more than minor because the issue was associated with the Mitigating Systems Cornerstone attribute for protection against external events and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesired consequences. The inspectors evaluated this finding using the SDP and determined the finding was of very low safety significance (Green) because this finding was a design deficiency that did not result in a loss of operability of the HPCI System. The inspectors also determined that this finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because the licensee did not promptly identify an adverse condition in the CAP in a ML11292A192
Page 16 of 54 timely manner commensurate with its safety significance.
22)
Mitigating POINT 06/30/2009 Green *SCWE: N *HP: Y *PIR: N Systems BEACH Docket/Status: 05000266 (C) , 05000301 (C)
Open: 2009003 (PIM) Inadequate Seismic Assessment Of Temporary Cable Installations Above Motor-Driven Auxiliary Feedwater Pumps The inspectors identified a finding of very low safety significance (Green) and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure of the licensees modification process to ensure that new 4160-volt cables installed for proposed auxiliary feedwater (AFW) pump motor replacements were installed in accordance with applicable regulatory requirements. Specifically, no seismic design evaluation was completed prior to the installation of the cable coils suspended above the existing motor-driven AFW pumps for over 6 months. In response to the issue, the licensee installed a new restraint designed to meet seismic criteria and completed calculations that showed the as-left condition of the modification did not challenge operability. This performance deficiency was more than minor because it was associated with the Mitigating System Cornerstone attribute of design control and adversely affected the cornerstone objectives of ensuring the availability, reliability, and capability of systems to respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, once identified, the modification required rework to comply with applicable design requirements. The inspectors determined the finding was of very low safety significance (Green) because the issue did not result in the actual loss of a safety function. The inspectors also determined the finding has a cross cutting aspect in the area of human performance, work control, because the licensee failed to incorporate risk insights and planned contingencies into work plans (H.3(a)).
23)
Mitigating AKANSAS 07/31/2009 Green *SCWE: N *HP: N *PIR: N Systems NUCLEAR ONE Docket/Status: 05000313 (C)
Open: 2009007 (PIM) Inadequate Design Control for Class 1E Batteries and Battery Racks Green. The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, that design changes, including field changes, shall be subject to design control measures commensurate with those applied to the original design.
Contrary to the above, the licensee failed to adequately perform a seismic evaluation for a modification to the Unit 2 safety related 125 Vdc battery racks. Specifically, on June 17, 1986, a ML11292A192
Page 17 of 54 design change was made to the battery racks to add hand hold and step on rails for ease of maintenance and inspection of the battery cells. The seismic evaluation for these rails addressed the impact to the battery rack seismic rating, and determined that the bolts for the rails must not be tightened to a specified torque value, but installed hand tight only. However, the seismic evaluation failed to address the potential for the rails to fall because the bolts were only hand tight. The licensee has entered this into their corrective action program as Condition Report CR ANO 2009 01573. The failure to perform a seismic evaluation for a modification to the Class 1E battery racks was a performance deficiency. The finding is more than minor because it is similar to Example 3.a of Inspection Manual Chapter 0612, Power Reactor Inspection Reports, Appendix B, Section 1-3, Screen for More than Minor - ROP, and it also affected the Mitigating Systems cornerstone attribute of design control to ensure the availability, reliability, and capability of safety systems that respond to initiating events to prevent undesirable consequences, and adversely affected the cornerstone objective because actions were required to be taken to ensure the hand tight bolts and rail met seismic qualifications.
Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheets, the finding was determined to have very low safety significance (Green) because it was a design issue that did not result in loss of operability or function. The inspectors reviewed the finding for cross cutting aspects and none were identified because the finding was not indicative of current performance.
24)
Mitigating ARKANSAS 09/23/2009 Green *SCWE: N *HP: N *PIR: N Systems NUCLEAR ONE Docket/Status: , 05000368 (C)
Open: 2009004 (PIM) FAILURE TO MAINTAIN SEISMIC DESIGN BASES CONTROL Green. The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to assure that applicable design basis for applicable structures, systems, and components were correctly translated into specifications, procedures, and instructions. Specifically, the licensee approved a nonconservative engineering calculation which led to operating procedure changes that allowed the removal of safety related, motor operated valve actuator rigid seismic restraints in the support of maintenance without verifying conformance to meet seismic design basis requirements. The issue was entered into the licensee's corrective action program as Condition Report ANO C 2009 0710. The performance deficiency was determined to be more than minor because it was associated with the protection against external events attribute of the Mitigating Systems Cornerstone, and affected the associated cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences, and is therefore a finding.
Specifically, the engineering calculation used to support removal of rigid seismic restraints and maintain operability only analyzed the deadweight of the motor operated valve actuator, not any dynamic seismic loading. Using NRC Manual Chapter 0609, Significance Determination ML11292A192
Page 18 of 54 Process, Phase 1 Worksheets, Mitigating Systems Cornerstone, the finding was determined to have very low safety significance because it did not represent an actual loss of safety function and did not screen as potentially risk significant due to a seismic initiating event. This finding did not have a crosscutting aspect because the engineering calculation used to determine the acceptability of removal of motor operated valve actuator seismic restraints to support maintenance and maintain system operability was made in 1994 and was not indicative of current plant performance.
25)
Mitigating 09/30/2009 BYRON Green *SCWE: N *HP: N *PIR: N Systems Docket/Status: 05000454 (C) , 05000455 (C)
Open: 2009004 (PIM) DIESEL OIL STORAGE VENTS DO NOT SEISMICALLY QUALIFIED OR TORNADO RESISTANT A finding of very low safety significance and associated NCV of 10 CFR 50, Appendix A, Criterion 2, Design basis for protection against natural phenomena, and Criterion 4, Environmental and natural effects design bases, was identified by the inspectors for the failure to seismically support and protect from tornado generated missiles the DG fuel oil storage tank vent lines. Specifically, the licensee installed the vent lines as non-safety related and as such they were not seismically supported nor protected from tornado generated missiles. In response to the issue, the licensee performed an operability determination and concluded that the DGs remained operable. This performance deficiency was more than minor because it was associated with the Mitigating Systems Cornerstone attribute of equipment performance and adversely affected the cornerstone objective of ensuring availability of the DG to respond to initiating events to prevent undesirable consequences. This finding was of very low safety significance (Green) because the inspectors determined that the finding was a design deficiency confirmed not to result in loss of operability or functionality and the finding screened as Green using the Significance Determination Process Phase 1 screening worksheet. The inspectors did not identify a cross cutting aspect associated with this finding because the performance deficiency occurred over 30 years ago and was not current.
26)
Mitigating 09/30/2009 CALLAWAY Green *SCWE: N *HP: N *PIR: N Systems Docket/Status: 05000483 (C)
Open: 2009009 (PIM) Failure to Maintain an Adequate Lubrication Procedure for Valve FCHV0312 The team identified a noncited violation of Technical Specification 5.4.1.a, Procedures, for the ML11292A192
Page 19 of 54 failure to provide adequate procedural guidance for the lubrication of auxiliary feedwater pump turbine trip throttle valve FCHV0312. The inspectors found that 2002 corrective actions to improve the lubrication procedure were not fully developed and the procedure lubrication guidance was ambiguous in that it did not specify the amount of lubricant to apply or what valve subcomponents to lubricate. The licensee entered this deficiency in their corrective action program as Callaway Action Request 200905032. This finding is greater than minor because it was associated with the Mitigating Systems Cornerstone attribute of procedural quality and it affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the issue screened as very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than the technical specification allowed outage time and did not affect seismic, flooding, or severe weather initiating events. This finding did not have a crosscutting aspect since the 2003 lubrication procedure revision was not reflective of current licensee performance.
27)
Mitigating DUANE 09/30/2009 Green *SCWE: N *HP: Y *PIR: N Systems ANRNOLD Docket/Status: 05000331 (C)
Open: 2009004 (PIM) FAILURE TO PERFORM AN IMMEDIATE OPERABILITY DETERMINATION FOR THE
'B' STANDBY DIESEL GENERATOR.
A finding of very low safety significance and associated NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for a failure of the Shift Manager to perform an Immediate Operability Determination (IOD) of the B Standby Diesel Generator (SBDG) after being notified by engineers of a concern with the seismic adequacy of the B SBDG normal air start system. The Shift Managers failure to follow procedure EN-AA-203-1001, Operability Determinations/Functionality Assessments, and Administrative Control Procedure (ACP) 110.1, Conduct of Operations, was considered a performance deficiency. The licensee entered this issue into the Corrective Action Program (CAP) as item CAP 070061, and isolated the B SBDG normal air start system from the emergency air start system. A detailed seismic analysis was performed on the B SBDG normal air start system to fully evaluate operability of the system during the design basis earthquake.
The performance deficiency was determined to be more than minor because if left uncorrected, the failure to adequately implement the operability procedures could result in safety-related components being incorrectly declared operable rather than inoperable or operable but non-conforming (a more significant safety concern). The inspectors evaluated this finding using the SDP and determined the finding was of very low safety significance (Green) because it did not represent an actual loss of safety function of a single train for longer than its Technical Specification (TS) allowed outage time. The inspectors also determined that this finding has a ML11292A192
Page 20 of 54 cross-cutting aspect in the area of Human Performance, Decision-Making, because the licensee failed to make a safety significant or risk-significant decision using a systematic process, especially when faced with uncertain or unexpected plant conditions, and thereby demonstrate that nuclear safety is an overriding priority. Specifically, the licensee did not make and document an IOD for the B SBDG once an adverse condition affecting a SBDG support system was identified.
28)
Mitigating PALO 09/30/2009 Green *SCWE: N *HP: N *PIR: Y Systems VERDE Docket/Status: 05000528 (C) , 05000529 (C) , 05000530 (C)
Open: 2009008 (PIM) Failure to Perform an Adequate Operability Evaluation for the Condensate Storage Tank The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to perform an adequate operability evaluation for the condensate storage tank as required by site procedures. Specifically, upon discovery of the condition, the licensee performed an immediate operability determination evaluation based on concerns with the capability of the loop seal to provide protection from vacuum conditions. Subsequently, the licensee performed additional assessments of their overall program which included the specified operability evaluation in a component design bases review and closure of a confirmatory action letter and failed to identify the inadequacy.
During the inspection, the team reviewed the operability determination and identified that the licensee failed to consider or identify concerns with the ability of the condensate storage tank pressure relief valves to operate after a design basis earthquake. The licensee entered this issue into their corrective action program as Palo Verde Action Request 3353683. This finding is more than minor because it is associated with the protection against external events (seismic) attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The risk significance of this finding was determined using Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings. The finding is of very low safety significance (Green) since the finding did not result in a loss of operability, a loss of system safety function, an actual loss of safety function of a single train for greater than its technical specification allowed outage time, or an actual loss of safety function for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the finding did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program since the licensee failed to properly evaluate for operability.
29)
Mitigating RIVER 09/30/2009 Green *SCWE: N *HP: Y *PIR: N Systems BEND ML11292A192
Page 21 of 54 Docket/Status: 05000458 (C)
Open: 2009004 (PIM) Failure to Control Scaffold Construction The inspectors identified a Green noncited violation of Technical Specification 5.4.1.a for the failure of maintenance personnel to control scaffold erection per procedure. This failure resulted in the licensee installing 31 scaffolds in safety related areas that required either rework or an engineering evaluation to resolve as built deviations from the minimum seismic separation requirements. As a result, the design function of the safety related equipment was potentially adversely affected. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2009-3963. The failure to erect scaffolds in accordance with procedures is a performance deficiency. This finding is more than minor because it is similar to Inspection Manual Chapter 0612, Appendix E, Examples of Minor Issues, Section 4, Example a, because Entergy had routinely failed to perform the requisite engineering evaluation and because it was associated with the protection against external events attribute of the mitigating systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events in order to prevent undesirable consequences. The finding was determined to be of very low risk significance (Green) because no actual loss of safety function occurred and the finding did not screen as potentially risk significant due to external events. This finding has a crosscutting aspect in the area of human performance, work practices, because the licensee failed to ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety is supported H.4(c).
30)
Mitigating RIVER 09/30/2009 Green *SCWE: N *HP: N *PIR: N Systems BEND Docket/Status: 05000458 (C)
Open: 2009004 (PIM) Failure to Maintain Reactor Core Isolation Cooling System Seismic Design The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for failure to implement measures to ensure that the seismic design basis for the reactor core isolation cooling turbine governor hydraulic system was correctly translated into the specifications, drawings, procedures, or instructions. This resulted in work to reroute the piping and an engineering evaluation to resolve seismic concerns. The licensee entered this issue into their corrective action program as Condition Report CR RBS 2009 3747. The failure to implement design control features for the seismic design of the reactor core isolation cooling system is a performance deficiency. This finding was more than minor because it is similar to Inspection Manual Chapter 0612, Appendix E, Examples of Minor Issues, Section 5, Example ML11292A192
Page 22 of 54 a, in that the reactor core isolation cooling turbine was returned to service without the seismic spacing required by the original design or completion of an evaluation for the as left condition.
This resulted in rework and additional engineering analysis to correctly resolve the seismic qualification concerns. The performance deficiency also affected the mitigating systems cornerstone attribute of external events and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events in order to prevent undesirable consequences. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Attachment A, Phase 1 - Initial Screening and Characterization of Findings, for the mitigating systems cornerstone. After answering no to all five questions in the mitigating systems cornerstone column of Table 4a, Characterization Worksheet for Initiating Events, Mitigating Systems, and Barrier Integrity Cornerstones, the inspectors concluded that the finding was of very low safety significance. This finding does not have a crosscutting aspect because the performance deficiency occurred in 1989 and is not reflective of current plant performance.
31)
Mitigating RIVER 09/30/2009 Green *SCWE: N *HP: N *PIR: Y Systems BEND Docket/Status: 05000458 (C)
Open: 2009004 (PIM) Failure to Ensure Standby Liquid Control System Test Tank Remained Drained The inspectors identified a Green noncited violation of Technical Specification 5.4.1.a for the failure of operations personnel to provide adequate procedural guidance to preclude water intrusion into the nonseismically qualified standby liquid control system test tank which resulted in the degradation of both trains of the standby liquid control system. The licensee entered this issue into their corrective action program as Condition Report CR RBS 2009 3862. The failure to provide appropriate procedures to keep the standby liquid control test tank drained is a performance deficiency. The finding is more than minor because it affects the protection against external events attribute of the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems responding to initiating events to prevent undesirable consequences. The inspectors determined that the finding was of very low safety significance because the finding was not a design or qualification deficiency, did not represent a loss of a system/train safety function, and did not screen as potentially risk significant due to external events. This finding has a crosscutting aspect in the area of problem identification and resolutions corrective action program because the licensee failed to take appropriate corrective actions to address safety issues and adverse trends in a timely manner, commensurate with their safety significance and complexity. Specifically, the licensee failed to address the cause of inadvertent water intrusion into the standby liquid control test tank in a timely manner to prevent the common mode failure of both trains of standby liquid control P.1(d).
Page 23 of 54 32)
Mitigating HOPE 10/09/2009 Green *SCWE: N *HP: N *PIR: N Systems CREEK Docket/Status: 05000354 (C)
Open: 2009007 (PIM) EDG OVERHEAD CRANES NOT SEISMICALLY RESTRAINED The team identified a finding of very low safety significance (Green) involving a non-cited violation of 10 CFR 50 Appendix B Criterion III, Design Control, in that PSEG design control measures had not verified the adequacy of design with respect to ensuring adequate two-over-one seismic protection existed for the emergency diesel generators (EDG). Specifically, PSEG had not performed design reviews, calculations or testing to ensure the existing field crane configuration would not adversely impact the EDG function for a design basis safe shutdown earthquake (SSE) event. PSEG entered this issue into their corrective action program, performed Technical Evaluation (TE) 70102445-0050, Diesel Generator Underhung Crane Seismic II/I Evaluation, to calculate the seismic response of the diesel cranes and assess the as-found condition (e.g., crane seismic restraints not installed) and implemented appropriate compensatory measures. The finding was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team determined the finding was of very low safety significance (Green) because it was a design deficiency subsequently confirmed not to result in a loss of operability or functionality. This finding did not have a cross-cutting aspect because the issue was not considered to be indicative of current licensee performance.
33)
Mitigating WOLF 10/15/2009 Green *SCWE: N *HP: Y *PIR: N Systems CREEK Docket/Status: 05000482 (C)
Open: 2009005 (PIM) Unevaluated Scaffold Against Component Cooling Water Piping The inspectors identified a noncited violation of Technical Specification 5.4.1.a for failure to properly implement Procedure AP 14A-003, Scaffold Construction and Use, when scaffolding was erected against operable safety-related equipment. On October 15, 2009, the inspectors walked down containment and identified scaffolding in contact with component cooling water piping. The tag on the scaffold explicitly stated that it was not seismically qualified. At the time, both steam generators were inoperable and both trains of residual heat removal were required ML11292A192
Page 24 of 54 to be operable. The inspectors reviewed the bases for Technical Specification 3.4.7, RCS Loops - Mode 5, Loops Filled, which required an operable heat sink path from residual heat removal to component cooling water to essential service water. This issue was entered into the corrective action program as Condition Report 22464. The construction of an unqualified scaffold against operable component cooling water piping was a performance deficiency. The inspectors determined that this finding was more than minor because it is associated with the equipment performance attribute for the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, this issue relates to the availability and reliability examples of the equipment performance attribute because a latent failure mechanism was not evaluated. The inspectors evaluated the significance of this finding using Inspection Manual Chapter 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Operational Checklists for Both PWRs and BWRs. The inspectors determined that Checklist 3 was applicable because the unit was in cold shutdown with the refueling cavity level less than 23 feet. Using Appendix G, Attachment 1, Checklist 3, Phase 2 analysis was not needed and the finding was of very low safety significance (Green) because the licensee was able to demonstrate that the seismically unqualified scaffolding would not have resulted in a loss of safety function. The inspectors determined the cause of the finding had a human performance aspect in the area of resources.
Specifically, Procedure AP 14A-003 was inadequate because it had conflicting guidance that allowed seismically unqualified scaffolds in Modes 5 and 6 H.2(c).
34)
Mitigating 11/18/2009 DRESDEN Green *SCWE: N *HP: N *PIR: N Systems Docket/Status: , 05000237 (C)
Open: 2009007 (PIM) Unit 2 SBLC Tank Thickness Calculation Errors (4OA5.1.b(2))
A finding of very low safety-significance (Green) and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the failure to accurately translate the design bases for the Standby Liquid Control (SBLC) tank into specifications, drawings, procedures, and instructions. Specifically, the SBLC tank wall thickness used in a design basis calculation was incorrect. The licensee initiated IR 983037 to address deficiencies in the calculation. The finding was determined to be more than minor because the finding was associated with the mitigating systems cornerstone attribute of design control and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
Specifically, the design basis calculations did not demonstrate that the tank will remain available following design basis seismic events. This finding is of very low safety-significance (Green) because it did not result in a loss of operability. The inspectors did not identify a cross-cutting aspect associated with this finding as it was not indicative of current performance.
Page 25 of 54 35)
Mitigating 12/04/2009 MONTICELLO Green *SCWE: N *HP: N *PIR: Y Systems Docket/Status: 05000263 (C)
Open: 2009007 (PIM) Emergency Service Water Piping Supports Did Not Meet Seismic Category 1 Design Basis Requirements The inspectors identified a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, having very low safety significance for the failure to restore the emergency service water (ESW) piping supports to their design specifications. Specifically, although the licensee identified the existence of gaps between the ESW piping supports and the baseplates, the licensee failed to recognize that this condition did not meet seismic Category 1 design basis requirements. As a result, corrective actions were not implemented. The licensee entered this issue into its corrective action program and restored the supports to their design specifications.
The finding was more than minor because it was associated with the Mitigating Systems cornerstone attribute of protection against external events and affected the cornerstone objective of ensuring the availability of the ESW system, and ultimately the emergency diesel generators (EDGs), to respond to initiating events to prevent undesirable consequences. This finding is of very low safety significance (Green) because the design deficiency was confirmed not to result in loss of operability or functionality. This finding has a cross-cutting aspect in the area of problem identification and resolution because the licensee did not properly prioritize and evaluate an identified problem.
36)
Mitigating 12/07/2009 SAN ONOFRE Green *SCWE: N *HP: Y *PIR: N Systems Docket/Status: , 05000361 (C)
Open: 2009005 (PIM) Failure to Adequately Identify Problems in Corrective Action Program The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for failure of engineering personnel to adequately identify for correction conditions adverse to quality between November 10 and December 1, 2009. Specifically, the inspection of potential degradation associated with the support welds and embedded wall plates for safety related seismic pipe restraints for emergency core cooling piping was inadequate, in that, standing water and corrosion product interference was not removed to enable an adequate ML11292A192
Page 26 of 54 inspection and evaluation of the structural material. This finding was entered into the licensees corrective action program as Nuclear Notification NN 200743417. The finding is greater than minor because the failure to adequately identify for correction conditions adverse to quality on safety related equipment, if left uncorrected, would have the potential to lead to a more significant safety concern. Additionally, the finding is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheets, the finding is determined to have very low safety significance because it did not represent an actual loss of safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The finding has a crosscutting aspect in the area of human performance associated with decision making because engineering personnel failed to use conservative assumptions for operability decision making when inspecting degraded and nonconforming conditions H.1(b) (Section 1R06).
37)
Mitigating ARKANSAS *HP: *PIR:
12/31/2009 Green *SCWE: N Systems NUCLEAR ONE Y N Docket/Status: , 05000368 (C)
Open: 2009005 (PIM) Failure to Correct a Condition Adverse to Quality Associated With Removal of a Rigid Seismic Restraint Green. The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," for the failure of licensee personnel to correct a condition adverse to quality - removal of rigid seismic restraint for valve 2CV 5672 1, containment spray pump 2P-35B minimum recirculation valve, in the support of motor operated valve actuator maintenance with an invalid engineering change to support the containment spray system's seismic operability licensing basis. This condition should have caused Unit 2 to enter Technical Specification 3.0.3 for 31 minutes on October 15, 2009. The inspectors had previously identified that the licensee was incorrectly applying ASME Code,Section III, Appendix F allowables to maintain operability for planned preventative maintenance. This issue was originally entered into the corrective action program as Condition Report CR-ANO-C-2009-1408. The licensee took action to cancel several engineering change documents, but did not review previously approved work orders to ensure that the removal of rigid seismic restraints would be prevented. This issue has been entered into the licensee's corrective action program as Condition Reports CR ANO C 2009 2193, CR ANO 2 2009 3356, and CR ANO 2 2009 3794. The failure to correct a condition adverse to quality associated with the removal of motor-operated valve actuator seismic restraints without a valid engineering evaluation was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the protection against external events attribute of the Mitigating Systems Cornerstone and ML11292A192
Page 27 of 54 directly affected the associated cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences and is therefore a finding. Specifically, the engineering change used to justify seismic operability was invalid and should not have been used to support continued operability and had been cancelled for future use. Using NRC Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheets, Mitigating Systems Cornerstone, the finding was determined to have very low safety significance because it did not represent an actual loss of safety function and did not screen as potentially risk significant due to a seismic initiating event. The cause of this finding was determined to have a crosscutting aspect in the area of human performance associated with resources H.2(c) in that the licensee failed to have complete and accurate procedures to prevent engineering changes that had been cancelled from being used in work orders that had been previously planned and approved for work.
38)
Mitigating 12/31/2009 ANO Green *SCWE: N *HP: N *PIR: Y Systems Docket/Status: , 05000368 (C)
Open: 2009005 (PIM) Failure to Follow Procedure Results in an Inadequate Operability Determination Green. The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instruction, Procedures, and Drawing, regarding the licensees failure to follow the requirements of Procedure EN OP 104, Operability Determination Process, Revision 4.
Specifically, on October 15, 2009, following removal of a seismic restraint from the train B containment spray valve 2CV-5672-1 for preventive maintenance purposes, the inspectors identified that the shift manager approved and documented an operability determination using a cancelled engineering change document. The licensee entered this into their corrective action program as Condition Report CR ANO 2 2009 3794. The failure of the licensee to follow the requirements of Procedure EN OP 104, Operability Determination Process, Revision 4, and approve an adequate basis for operability was a performance deficiency. The performance deficiency was determined to be more than minor because the condition of not performing adequate operability determinations could become more significant if left uncorrected and is therefore a finding. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to have very low safety significance because it did not result in the loss of safety function of any technical specification required equipment. It was determined that the finding had a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program P.1(c), in that, the licensee failed to thoroughly evaluate problems such that the resolutions addressed causes and extent of conditions, as necessary.
Page 28 of 54 39)
Mitigating 12/31/2009 FARLEY Green *SCWE: N *HP: N *PIR: Y Systems Docket/Status: 05000348 (C) , 05000364 (C)
Open: 2009005 (PIM) Failure to Implement Performance Monitoring of SWP Seismic Supports Self-revealing NCV of 10CFR50.65(a)(1) identified for failure to perform monitoring of SWIS seismic rings resulting in inability of 2A,B,C&E SWP seismic rings to perform their func.b/c of fastener degradation. Licensee entered CR into CAP as 2009109700 and completed corrective actions to restore ring function. The ring failure is a performance deficiency. Finding greater than minor b/c adversely affected equip reliability attribute of MS cornerstone objective ensuring availability, reliability, capability of systems reponding to IE preventing undesired conseq. Phase 1 Screen Wksht of SDP used to assess finding. SWP determined not degraded. Finding associated w/cross-cutting aspect in CAP component for PI&R area (P.1)(d)).
40)
Mitigating FORT CALHOUN *HP:
12/31/2009 Green *SCWE: N *PIR: N Systems STATION N Docket/Status: 05000285 (C)
Open: 2009006 (PIM) Inadequate Assessment of Seismic Qualification of Raw Water Pumps Green. The team identified a Green, noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, from February 1992 to September 8, 2009, the licensee failed to adequately evaluate the seismic qualification of the raw water pumps to ensure that the pumps anchor bolts imbedded in the floor would meet Seismic Class I standards. The team determined that the February 1992 seismic analysis was not conservative for the following reasons: (1)The weight distribution of the pump/motor assembly in the analysis did not correctly apply the center of gravity of the pump to the loading analysis. (2)The stress analysis of the anchors did not include the weight of the water in the piping. (3)The stress analysis did not include the nozzle loads applied to the pump due to the weight of the discharge piping. The licensee entered the issue into their corrective action program as CR 2009-3977, and performed a preliminary operability evaluation of the support components which determined that the pumps would remain operable following a safe shutdown earthquake. The team reviewed the evaluation, and concurred with the operability evaluation. The finding is more than minor because it adversely affected the design control attribute of the Mitigating Systems Cornerstone ML11292A192
Page 29 of 54 objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was of very low safety significance (Green) because it was a design deficiency that did not result in actual loss of safety function.
This finding was not assigned a crosscutting aspect because the underlying cause was not indicative of current performance (Section 1R21.2.15).
41)
Mitigating FORT CALHOUN 12/31/2009 SL-IV *SCWE: N *HP: N *PIR: N Systems STATION Docket/Status: 05000285 (C)
Open: 2009006 (PIM) Failure to Maintain Quality Records of the Intake Structure Design SL-IV. The team identified a Severity Level IV, noncited violation of 10 CFR Part 50, Appendix B, Criterion XVII, Quality Assurance Records, for failure to maintain original records of the seismic and tornado analysis of the intake structure. Specifically, in 2005, the licensee could not retrieve the original design documentation of the seismic and tornado analysis of the intake structure. This condition was documented in CR 200504345. After the licensee determined the documentation was not retrievable, the licensee reconstituted the seismic and tornado analysis of the intake structure. These analyses were available during the teams inspection. This finding is assessed through traditional enforcement because the finding has the potential for impacting the NRCs ability to perform its regulatory function. Using Inspection Manual Chapter 0612, Appendix E, the finding is more than minor because the records were not retrievable. Using Supplement I of the NRC Enforcement Policy, this finding will be treated as a Severity Level IV violation. This finding was not assigned a crosscutting aspect because the underlying cause was not indicative of current performance (Section 4OA5.1).
42)
Mitigating SAN 03/18/2010 Green *SCWE: N *HP: N *PIR: N Systems ONOFRE Docket/Status: , 05000361 (C)
Open: 2010003 (PIM) Failure to Assure Circuit Breakers Were Qualified for Installation A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion VII, Control of Purchased Material, Equipment, and Services, was identified for the failure of engineering personnel to assure that 4 kV vacuum circuit breakers supplied by NLI/Square D conformed to the procurement documents prior to installation in Unit 2 bus 2A06 train B. Specifically, on December 18, 2009, 4 kV bus 2A06 was restored to operable status following installation of 4 kV vacuum circuit breakers supplied by NLI/Square D that did not conform to the design requirements specified in the procurement documents. Engineering personnel failed to assure ML11292A192
Page 30 of 54 that 4 kV vacuum circuit breakers conformed to the requirements of Specification SO23-302-02A, 4kV Roll-in Replacement Circuit Breakers, Revision 1, and failed to identify that the vendor completed seismic qualification test deviated from the procurement specifications prior to installation in the plant. On March 18, 2010, an unexpected trip of component cooling water pump circuit breaker 2A0605 prompted an investigation that identified the design inadequacies.
Operations personnel declared the associated circuit breakers inoperable following identification of the design inadequacies. Immediate actions to eliminate the design inadequacies were completed to return 4 kV bus 2A06 to operable on March 25, 2010. Apparent Cause Evaluation ACE 200845084 was initiated to identify additional corrective actions. This issue was entered into the licensees corrective action program as Nuclear Notification NN 200842716. The performance deficiency is more than minor because it is associated with the design control attribute of the Mitigating Systems Cornerstone and affects the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences, and is therefore a finding. Using the Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, Phase 1 guidance, the finding is determined to have very low safety significance because the finding did not result in an increase in the likelihood of a loss of reactor coolant system inventory, degrade the ability to add reactor coolant system inventory, or degrade the ability to recover decay heat removal.
Since the lack of questioning attitude that contributed to an overreliance on the specifications occurred in 2005, and Procurement Specification Training was conducted in 2008 to close an identified gap in specification review and implementation, the inspectors determined that this was not reflective of current performance and therefore did not have a crosscutting aspect associated with it.
43)
Mitigating DIABLO 03/27/2010 Green *SCWE: N *HP: N *PIR: Y Systems CANYON Docket/Status: 05000275 (C) , 05000323 (C)
Open: 2010002 (PIM) Failure to Effectively Implement the Seismically-induced Systems Interaction Program The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, after Pacific Gas and Electric personnel failed to effectively implement the Seismically Induced System Interaction Program. The Seismic Interaction Program is part of the design basis mitigation strategy for a potential 7.5 magnitude Hosgri earthquake and is required by Procedure AD4.ID3, SISIP Housekeeping Activities. The inspectors identified three examples of transient equipment and materials improperly staged in seismically induced system interaction target areas. Pacific Gas and Electric had not analyzed the transient equipment to assess the risk to safety related components as required by plant procedures. Pacific Gas and Electric entered this finding into the corrective action program as Notification 50299740. The finding is more than minor because the failure to follow the Seismically Induced System Interaction Program is associated with the Mitigating Systems ML11292A192
Page 31 of 54 Cornerstone external events protection attribute and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors concluded that the finding had very low safety significance because none of the examples of improperly staged equipment resulted in an actual loss of a system safety function or equipment required by technical specifications, or involve the loss or degradation of equipment specifically designed to mitigate a seismic, flooding, or severe weather initiating event, and did not involve the total loss of any safety function that contributes to an external event initiated core damage accident sequence.
The inspectors concluded this finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensees past actions to address Seismically Induced System Interaction Program deficiencies were not effective P.1(d).
44)
Mitigating 03/31/2010 MCGUIRE Green *SCWE: N *HP: Y *PIR: N Systems Docket/Status: 05000369 (C) , 05000370 (C)
Open: 2010002 (PIM) Failure to flow test nuclear service water "A" train standby nuclear service water pond (SNSWP) supply header at maximum design.
A self-revealing Green NCV of 10 CFR 50, Appendix B, Criterion XI, Test Control, was identified for the licensees failure to flow test the Nuclear Service Water System (NSWS) A Train Standby Nuclear Service Water Pond (SNSWP) unit common supply header at maximum design flow. The licensee entered this issue into their corrective action program as PIP M 2216 and has taken corrective actions to increase the minimum required flow velocity, frequency, and duration of the A Train SNSWP unit common supply header test procedure.
The finding was more than minor because it affected the cornerstone attributes of protection against external events and equipment performance and the Mitigating Systems objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, inadequate flushing of the A Train SNSWP unit common supply header led to ineffective flushes and the accumulation of corrosion products which challenged the design function of the NSWS system. This finding was evaluated using IMC 0609, Attachment 4, Phase I - Initial Screening and Characterization of Findings, to determine the safety significance. Since the finding was related to a seismic initiating event, a Phase III was required to be performed by an NRC Senior Risk Analyst. The Phase III analysis calculated the risk increase to be less than 1E-7 for both conditional core damage probability and conditional large early release probability, resulting in a determination of very low risk significance (Green). This performance deficiency was associated with the cross-cutting aspect of complete, accurate and up-to-date design documentation and procedures H.2(c) as described in the Resources component of the Human Performance cross-cutting area. (Section 4OA3.1)
Page 32 of 54 45)
Mitigating NORTH 03/31/2010 Green *SCWE: N *HP: Y *PIR: N Systems ANNA Docket/Status: 05000338 (C)
Open: 2010002 (PIM) Failure to Establish an Adequate Post-Modification Test Program for Piping Supports A Green, non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, was identified by the NRC for the failure to establish an adequate postmodification test program for piping supports affected by piping design changes (modifications). The licensee entered this problem into their corrective action program as condition report 357450. The inspectors determined that the failure to establish an adequate post-modification test program for piping supports affected by piping modifications as required by 10 CFR 50, Appendix B, Criterion III, was a performance deficiency (PD). This PD had a credible impact on safety due to a programmatic deficiency that resulted in safetyrelated piping supports adversely affected by modifications. The PD was more than minor because if left uncorrected it would have the potential to result in a more significant event involving inoperable, unidentified safety-related piping supports with consequent adverse impact on the respective system during a seismic event. In accordance with NRC Inspection Manual Chapter 0609, Significant Determination Process, the inspectors performed a Phase 1 analysis and determined the finding was of very low safety significance or Green due to a design deficiency confirmed not to result in a loss of operability or functionality. This finding involved the crosscutting area of human performance, the component of the resources, and the aspect of complete, accurate and up-to-date procedures, H.2(c), because the licensee failed to establish up-to-date program procedures to ensure adequate post-modification testing of piping supports.
46)
Mitigating 03/31/2010 PALISADES Green *SCWE: N *HP: Y *PIR: N Systems Docket/Status: 05000255 (C)
Open: 2010002 (PIM) Improper Construction of Scaffolding A finding of very low safety significance (Green) and associated NCV of Palisades Technical Specification (TS) 5.4.1, Procedures, was identified by the inspectors for failing to adequately implement a procedure to construct a scaffold near the 1 2 emergency diesel generator (EDG).
Specifically, a fire sprinkler was impaired without the proper fire protection evaluation; and ML11292A192
Page 33 of 54 required seismic evaluations were not performed despite being in close proximity to safety related equipment. The issue was entered into the licensees corrective action program and the scaffold was modified. The issue is more than minor because it affects the Protection Against External Events attribute of the Mitigating Systems Cornerstone in that it affects the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. Specifically a fire protection feature (sprinkler) in a safety related area was affected without compensatory measures. Additionally, the scaffold was in close proximity to safety related equipment, and the equipment could have been impacted by a seismic event. The finding screened as Green based on remaining sprinkler capability and the fact that only one EDG could be affected by the scaffold during a seismic event. The finding had an associated cross cutting aspect in the area of Human Performance (Planning) in that the licensee failed to appropriately plan work activities by incorporating the need for compensatory actions (H.3(a)).
(1R05) 47)
Mitigating 04/02/2010 BRAIDWOOD Green *SCWE: N *HP: N *PIR: N Systems Docket/Status: 05000456 (C) , 05000457 (C)
Open: 2010007 (PIM) DDAFW Pump Battery Racks were not restored to their Design Basis Seismic Category I The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, having very low safety significance for the licensees failure to restore the Diesel Driven Auxiliary Feedwater (DDAFW) battery racks to their design basis qualification, Seismic Category I. Specifically, although the licensee identified the existence of gaps between the wooden spacer blocks, batteries and end of racks in 2004 the licensee failed to provide adequate justification to demonstrate that the existing condition still met the Seismic Category I Design Basis requirements as specified in their design documents. The gaps between the wooden spacer blocks could affect the reliability of the DDAFW DC safety-related batteries being that this component was outside its design basis for over a period of six years. The licensee subsequently entered the issue into their corrective action program and restored the batteries racks to their design requirements. The finding was more than minor because it was associated with the Mitigating Systems cornerstone attribute of equipment performance, and affected the cornerstone objective of ensuring the availability of DDAFW batteries to perform their safety function in external events to prevent undesirable consequences. Specifically, the licensee did not assure that the wooden spacer blocks including the gap would provide adequate support to ensure that the seismically qualified battery rack will perform its safety function. This finding is of very low safety significance (Green) because the qualification deficiency was confirmed not to result in loss of operability or functionality. The inspectors determined that there was no cross-cutting aspect associated with this finding because the gaps between the wooden spacers and the DDAFW batteries were initially identified in 2004; therefore, the finding was not indicative of the plants current performance.
Page 34 of 54 48)
Mitigating 04/27/2010 WATTS BAR Green *SCWE: N *HP: N *PIR: N Systems Docket/Status: 05000390 (C)
Open: 2010006 (PIM) Inadequate assessment of seismic qualification of ERCW strainers The team identified a Green NCV of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure to update ERCW strainer mounting (seismic/structural) calculations to reflect the as-built conditions, a failure which was allowed to exist since commercial operations began. This calculation was then used in making acceptance conclusions for a modification installed in recent months. The licensee entered this condition into their corrective action program as Problem Evaluation Reports (PERs) 221018, 220754, and 223677 and took immediate actions to determine the seismic acceptability of the current installation, utilizing calculational conclusions of a similar installation at the licensees Sequoyah Nuclear Plant. The finding was determined to be more than minor because it was associated with the design control attribute within the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences, in that there was reasonable doubt as to the operability of the ERCW strainers as a result of the performance deficiency. The team evaluated the finding to be of very low safety significance (Green) utilizing IMC 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings worksheet, as it was a calculational error subsequently determined to not result in an operability issue. No cross-cutting aspect was identified since the issue was not reflective of current licensee performance. (Section 1R21.2.5) 49)
Mitigating TURKEY 05/21/2010 Green *SCWE: N *HP: N *PIR: N Systems POINT Docket/Status: , 05000250 (C) , 05000251 (C)
Open: 2010006 (PIM) Inadequate procedure implementation resulting in snubber failure.
The NRC identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V, for the licensees failure to implement procedures during a visual inspection of safety related seismically qualified snubber SN-4-1039. Specifically, the licensee failed to identify missing, detached, loosened support items, or full thread engagement of all mechanical connections that led to a snubber failure as prescribed in procedure 0-OSP-105.1, Visual Inspection, Removal and Reinstallation of Mechanical Shock Arrestors, section 7.2.1.3.d. The snubber would not have been able to perform its design function to arrest shocks of the main steam piping to the C Steam Generator during seismic events or transients, such as sudden isolation of the main steam isolation valve. The licensee implemented immediate corrective actions which included ML11292A192
Page 35 of 54 replacing the snubber in containment, adding specific instructions in procedure 0-OSP-105.1 to specifically inspect the locking ring and correct installation, and to include emphasis on FPL expectations from vendor provided snubber inspection services. The licensee documented this in condition report CR 2008-31372. The performance deficiency was more than minor because it affected the equipment performance attribute of the Mitigating Systems cornerstone in that the licensee did not ensure reliability of the snubber to respond to initiating events to prevent undesirable consequences in that the snubber would not have been able to perform its design function to arrest shocks of the main steam piping to the C Steam Generator during seismic events or transients. The finding was screened using Manual Chapter 0609.04, "Phase 1 -
Initial Screening and Characterization of Findings," and was determined to have a very low safety significance (Green) because the system remained operable and capable of meeting its design function with no loss of safety function of the C main steam system. This finding was reviewed for cross-cutting aspects and none were identified. (4OA2).
50)
Mitigating SAN\
06/17/2010 Green *SCWE: N *HP: Y *PIR: N Systems ONOFRE Docket/Status: , 05000361 (C)
Open: 2010006 (PIM) Lack of preventive maintenance results in valve failure and inoperable condensate storage tank.
The inspectors identified a noncited violation of Technical Specification 3.7.6, which requires, in part, that Condensate Storage Tank T-120 be operable. Specifically, the tank isolation valve 2HV5715 had been inoperable for a period greater than the allowed outage time of seven days while Unit 2 was in Modes 1, 2, and 3. The valve isolates nonseismic piping from the tank and is required to be manually closed within 90 minutes following a seismic event. The licensee had not performed preventive maintenance on the valve resulting in the valve failing to close during an in-service test on January 26, 2010. This finding was entered into the licensees corrective action program as Nuclear Notification 200765235. The licensees corrective actions included repairing the isolation valve. This finding is more than minor because it impacted the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Phase 1, Initial Screening and Characterization of Findings, a Phase 2 analysis was performed because the condensate storage, Tank T-120, was inoperable greater than that allowed in technical specifications. Phase 2 analysis resulted in a potential greater than Green issue therefore, a Phase 3 was performed. The analyst performed a Phase 3 using San Onofre seismic information and fragility data associated with the piping that could not be isolated because of the failed condition of valve 2HV5715. The frequency of a seismic event that would cause a pipe break and drain tank T-120 was estimated to be 2.7E-5/yr. Given a seismic event that causes a loss of offsite power (nearly 100 percent of seismic events that rupture the piping would also cause a loss of offsite power), operators are compelled by procedure to cool down ML11292A192
Page 36 of 54 and initiate shutdown cooling. The amount of water that is protected with valve 2HV5715 failed to open, which includes inventory from tank T-121 and water below the break line in tank T-120, given that operators close the working manual isolation valve within 30 minutes, is more than what is needed to get to shutdown cooling in natural circulation with only 1 of 2 steam generator atmospheric dump valves in operation, even if there is a 4-hour hold time at hot standby. The analyst estimated that the failure probability of operators to cool down and initiate shutdown cooling is 1.0E-2. Therefore, assuming a zero base case, the estimated delta- core damage frequency of the finding is 2.7E-5/yr. (1.0E-2) = 2.7E-7/yr. The inspectors also determined that the cause of the finding has a crosscutting aspect in the area of human performance associated with resources in that the licensee did not ensure that equipment was available and adequate to assure nuclear safety by minimization of long standing equipment issues in that the valve was not being maintained through a preventive maintenance program. H.2(a)
51)
Mitigating POINT 06/30/2010 Green *SCWE: N *HP: N *PIR: Y Systems BEACH Docket/Status: 05000266 (C) , 05000301 (C)
Open: 2010003 (PIM) Failure to Control the Design of Partially Installed Modifications for Seismic Requirements A finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the failure of the licensees modification process to ensure that new 480 volt cables, installed for the future repowering of various auxiliary feedwater (AFW) system motor operated valves, were installed in accordance with applicable regulatory requirements. Specifically, a seismic design evaluation was not completed prior to the installation of a cable coil suspended above the 2MS 2020 valve, 2P 29 turbine driven AFW pump steam supply. In response to this issue, the licensee installed more robust restraints that satisfied seismic acceptability criteria and performed an evaluation that showed the interim condition of the modification did not challenge operability. At the conclusion of this inspection period, the licensee was in the process of conducting a root cause evaluation. The inspectors also noted that a very similar issue at this site resulted in the issuance of a NCV in the second quarter of 2009. This performance deficiency was more than minor because it was associated with the Mitigating Systems Cornerstone attribute of design control and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, once identified, the modification required rework to comply with applicable design requirements. The inspectors determined the finding was of very low safety significance (Green) because the issue did not result in the actual loss of a safety function. The inspectors also determined the finding has a cross cutting aspect in the area of problem identification and resolution, corrective action program, because the licensee failed to implement appropriate corrective actions for a previous violation with the same performance deficiency (P.1(d)).
Page 37 of 54 52)
Mitigating POINT 06/30/2010 Green *SCWE: N *HP: N *PIR: Y Systems BEACH Docket/Status: 05000266 (C) , 05000301 (C)
Open: 2010003 (PIM) PROCEDURES WERE NOT APPROPRIATE TO ADEQUATELY VERIFY AND DOCUMENT THE DESIGN OF NEW OR MODIFIED SSCs WITH RESPECT TO SEISMIC II/I INTERACTIONS.
A finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the failure to provide procedures that were appropriate to verify and document the design of new or modified SSCs with respect to seismic II/I interactions. Specifically, the procedures used for seismic II/I interaction evaluations of new or modified SSCs did not provide guidance for evaluating equipment that was not represented in the earthquake experience or generic testing equipment classes under the scope of the Seismic Qualification Utility Group methodology. Also, no formal guidance was incorporated in modification and seismic procedures to document seismic II/I interaction evaluations. As a result, the licensee did not perform an evaluation that was in accordance with the licensing basis to verify the design of the B containment sump strainers of Units 1 and 2 with respect to potential seismic II/I interactions. The licensee entered this issue into its corrective action program. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of protection against external events and adversely affected the cornerstone objective of ensuring the capability of systems that respond to initiating events to prevent undesirable consequences. This finding was of very low safety significance (Green) because it was a qualification deficiency confirmed not to result in loss of operability or functionality. The inspectors determined that the finding had a cross cutting aspect in the area of problem identification and resolution, self and independent assessments, because the licensee did not conduct self assessments of the Seismic Qualification Utility Group program (P.3(a)).
53)
Mitigating WOLF 06/30/2010 Green *SCWE: N *HP: N *PIR: Y Systems CREEK Docket/Status: 05000482 (C)
Open: 2010003 (PIM) Failure to Maintain an Adequate Flooding Analysis for Auxiliary Feedwater Trains ML11292A192
Page 38 of 54 The inspectors identified a green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, after Wolf Creek failed to provide adequate design control measures for verifying the adequacy of the flooding analysis for the auxiliary feedwater pipe rooms 1206 and 1207. Wolf Creek failed to identify piping that was seismically unqualified and that if ruptured could potentially overwhelm the floor drains. Wolf Creek re-analyzed the piping and determined it would not rupture during an earthquake. Flooding of the room could have caused all three of the auxiliary feedwater pump suction pressure transmitters to fail and inhibit automatic swap to essential service water. The licensee placed this issue in their corrective action program as Condition Report 26050. The inspectors determined that the incorrect calculation assumption in the flooding analysis of record was the performance deficiency. This finding was determined to be more than minor because it impacted the Mitigating Systems Cornerstone attribute of the design control and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to have very low safety significance because the deficiency was confirmed not to result in loss of operability or functionality. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program because the licensees evaluation focused on the probability of equipment failure leading to a flooding event rather than the stated design basis of the facility P.1(c).
54)
Mitigating DIABLO 07/27/2010 Green *SCWE: N *HP: Y *PIR: N Systems CANYON Docket/Status: 05000275 (C) , 05000323 (C)
Open: 2010006 (PIM) Inadequate Design Control for the Emergency Diesel Generator The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, involving the failure to maintain adequate design control measures associated with the emergency diesel generating air system. Specifically, failure of non-seismically qualified air compressor unloader sensing lines during a seismic event could impact the safety function of the emergency diesel generators. Subsequent analysis of the nonconforming condition performed by the licensee determined the piping would not fail during a postulated seismic event. The licensee entered this issue into the corrective action program as Notifications 50307496, 50307497, 50307504, 50307670, 50308204, and 50308824. The finding was more than minor because it affected the mitigating systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Significance Determination Process (SDP) Phase 1 Screening Worksheet for the Initiating Events, Mitigating Systems, and Barriers Cornerstones the finding was potentially risk significant for a seismic initiating event requiring a Phase 3 analysis. The analyst estimated the nonrecovery probabilities for operators failing to isolate air between the receiver and the compressor prior to air pressure depletion, and operators failing to ML11292A192
Page 39 of 54 manually open fuel transfer valves to makeup to the diesel day tank. The final quantitative result was calculated to be 1.06 x 10-6. However, using a qualitative evaluation of the bounding assumptions, the analyst determined that the best available information indicated that the finding was of very low risk significance (Green). The team determined that the finding was reflective of current plant performance because it had been recently identified during the license renewal inspection and had a human performance crosscutting aspect related to decision making because the licensee did not use conservative assumptions when evaluating this nonconforming condition in previous evaluations.
55)
Mitigating SOUTH TEXAS 09/30/2010 Green *SCWE: N *HP: Y *PIR: N Systems PROJECT Docket/Status: , 05000499 (C)
Open: 2010004 (PIM) Failure to Repair Essential Cooling Water System Leak within the Technical Specification Allowed Outage Time The inspectors identified a Green noncited violation of Technical Specification 3.7.4 because the licensee had one independent loop of essential cooling water inoperable for longer than the allowed outage time of 7 days. Specifically, on October 27, 2009, the licensee failed to initiate actions to evaluate and repair a through-wall leak in the 30-inch essential cooling water return line from the Unit 2 train C component cooling water heat exchanger, as required by American Society of Mechanical Engineers Boiler and Pressure Vessel Code, and in accordance with guidance contained in NRC Generic Letter 90-05, Guidance for Performing Temporary Non-Code Repair of ASME Code Class 1, 2, and 3 Piping. The inspectors questioned the licensees reportability review and determined there was firm evidence that the through-wall leak caused the Unit 2 train C essential cooling water system to be inoperable for a period of 11 days instead of 8 days as initially concluded by the licensee. The licensees corrective actions were: (1) the leak was repaired, (2) a revised licensee event report was submitted, (3) training was provided to personnel performing these evaluations, and (4) procedures were updated to require that these types of evaluations must be performed. The finding was more than minor because the through-wall leak could have challenged the structural integrity of the piping and it was associated with the Mitigating Systems Cornerstone attribute of configuration control and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors performed the initial significance determination using NRC Inspection Manual 0609, Attachment 0609.04, dated January 10, 2008, Phase 1- Initial Screening and Characterization of Findings, because it affected the Mitigating Systems Cornerstone while the plant was at power, and determined a Phase 2 was required because it involved an actual loss of safety function of a single train. A Region IV senior reactor analyst performed a Phase 2 significance determination and found that the finding was potentially greater than Green. The senior reactor analyst then performed a bounding Phase 3 significance determination and found the finding to be of very ML11292A192
Page 40 of 54 low safety significance. The dominant core damage sequences included: seismic initiated loss of offsite power, failure of the essential cooling water train C, failure of the train A and B standby diesel generators, failure to recover offsite power and a standby diesel generator in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and an event initiated reactor coolant pump seal loss-of-coolant accident. Remaining mitigation equipment that helped to limit the significance of the finding included the remaining functional essential cooling water trains and the turbine-driven auxiliary feedwater pump. In addition, this finding had human performance crosscutting aspects associated with resources in that the licensee did not ensure that training of personnel about the requirements for properly characterizing Class 3 piping leaks was adequate to assure nuclear safety H.2(b).
56)
Mitigating WOLF 10/01/2010 Green *SCWE: N *HP: N *PIR: N Systems CREEK Docket/Status: 05000482 (C)
Open: 2010007 (PIM) Inadequate Design of Component Cooling Water Safety/Nonsafety Isolation The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Specifically, as of October 8, 2010, the licensee failed to incorporate design seismic requirements into the design calculations and actual system operation. This finding was entered into the licensees corrective action program as Condition Report 00028237. The team determined that the failure to adequately analyze the isolation between the safety related and nonsafety-related portions of the component cooling water system was a performance deficiency. This finding was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the design basis analysis did not ensure that the affected train of component cooling water would perform its required functions after the failure of nonsafety-related component cooling water piping. The inspectors evaluated the issue using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings. This finding affected the Mitigating Systems Cornerstone because seismic protection was degraded. The inspectors determined that this finding represented the degradation of equipment and functions specifically designed to mitigate a seismic event and that during an earthquake the deficiency would degrade one train of component cooling water, a system that supports a safety system or function. Therefore, this finding was potentially risk significant to seismic initiators and a Phase 3 analysis was required. A Region IV senior reactor analyst performed the Phase 3 significance determination. The change in core damage frequency was calculated to be 7.0 x 10 8 indicating that this finding was of very low safety significance (Green). The dominant risk sequence included a seismic initiating event, loss of offsite power, loss of reactor coolant pump seal cooling, and a failure of high pressure ML11292A192
Page 41 of 54 recirculation. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
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Mitigating 10/20/2010 COOPER Green *SCWE: N *HP: N *PIR: Y Systems Docket/Status: 05000298 (C)
Open: 2010007 (PIM) Inadequate Design Control The team identified seven examples of a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for failure to establish measures to ensure that applicable regulatory requirements and the design bases were correctly translated into specifications, drawings, procedures, and instructions. These measures shall include provisions to ensure that appropriate quality standards are specified and included in design documents and that deviations from such standards are controlled. Specifically, as of August 12, 2010, the licensee failed to correctly translate regulatory requirements and design bases information into specifications, drawings, procedures, and instructions involving emergency diesel generator frequency, service water pump, electrical cables for the residual heat removal pumps, seismic supports, the emergency diesel generator air start system testing, tornado and high wind impact on the emergency diesel generator fuel oil storage facilities and safety related Agast relay service life evaluations. This finding was entered into the licensees corrective action program as Condition Reports CNS-2010-05301, CNS-2010-5763, CNS-2010-05222, CNS-2010-05281, CNS-2010-5294, CNS-2010-5350, and CNS-2010-5438. The failure to correctly translate regulatory requirements and design bases information into specifications, drawings, procedures, and instructions for the emergency diesel generator frequency, service water pump, electrical cables for the residual heat removal pumps, emergency diesel generator room ventilation seismic supports, emergency diesel generator air start system testing, tornado and high wind impact on the emergency diesel generator fuel oil storage facilities and safety related Agast relay service life evaluations was a performance deficiency. This finding was more than minor because it was associated with the design control attribute of the mitigating systems cornerstone and impacted the cornerstone objective to ensure the availability, reliability, and capability of the affected system to respond to events and prevent undesirable consequences.
Using the Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the issue screened as having very low safety significance (Green) because it was not a design or qualification deficiency and did not represent a loss of safety function. The licensee performed evaluations which determined that the affected components and systems were capable of meeting their design functions. The finding had a crosscutting aspect in the area of problem identification and resolution, associated with operating experience because the licensee failed to properly evaluate and apply various industry events associated ML11292A192
Page 42 of 54 with the above systems and incorporate the information into plant procedures and training
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Mitigating 11/11/2010 GINNA Green *SCWE: N *HP: N *PIR: N Systems Docket/Status: 05000244 (C)
Open: 2010009 (PIM) Inadequate Evaluation of Breaker Coordination for Ampetector Type LSG Trip Unit Discriminator Feature The inspectors identified a finding of very low safety significance involving a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control. Specifically, Ginna had not verified the adequacy of their design with respect to the impact of the installed Amptector type long, short, and ground trip unit discriminator feature on breaker coordination. The discriminator circuit design had not been evaluated to ensure the 480V load center bus motor control center feeder breakers would maintain coordination and be capable of maintaining power to downstream safety-related components in response to design basis events such as seismic or steam line break transients. Ginna entered this into their correction action program to evaluate the adequacy of their design and ensure the feeder breakers remained operable. The finding was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of design control and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of the 480V busses to respond to initiating events to prevent undesirable consequences. The inspectors evaluated the finding in accordance with Inspection Manual Chapter 0609, Significance Determination Process, 609.04, Phase 1, "Initial Screening and Characeterization of Findings", Table 4a, for the Mitigating Systems Cornerstone. The inspectors determined the finding was of very low safety significance because it was a design deficiency confirmed not to result in a loss of operability. The inspectors did not identify a cross-cutting aspect with this finding because this was an old design issue and, therefore, was not reflective of current performance.
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Mitigating 12/31/2010 CALLAWAY Green *SCWE: N *HP: Y *PIR: N Systems Docket/Status: 05000483 (C)
Open: 2010005 (PIM) Failure to Follow Operability Determination Procedure ML11292A192
Page 43 of 54 The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for failure to follow Procedure APA ZZ 00500, Appendix 1, Operability and Functionality Determinations. On the morning of September 23, 2010, Callaway engineering was informed that a concern existed that the safety related portion of the component cooling water system safety function could be affected by a guillotine break at the nonsafety/nonseismic boundary for supply and return piping to the radwaste building. The inspectors determined that the licensee staff did not engage the shift manager early enough and the shift manager did not adequately challenge the basis describing the nonconforming condition as acceptable. The shift manager allowed the component cooling water system to be in an indeterminate state of operability for over two hours without putting compensatory measures in place as described in Procedure APA ZZ 00500, Appendix 1. This issue was entered into the licensees corrective action program as Callaway Action Request 201010739.
This finding was determined to be greater than minor because it impacted the mitigating systems cornerstone attribute of human performance and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this issue screened as requiring a Phase 3 analysis. The NRC senior risk analyst determined that because ?CDF was less than 1E-6 and
?LERF was not a significant contributor to risk, this finding was of very low safety significance, Green. This finding has a crosscutting aspect in the area of human performance associated with the decision making component because the licensee failed to use conservative assumptions when performing operability evaluations.
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Mitigating 12/31/2010 HARRIS Green *SCWE: N *HP: Y *PIR: N Systems Docket/Status: 05000400 (C)
Open: 2010005 (PIM) Failure to Comply with the Limiting Conditions for Operation, While the Refueling Water Storage Tank was Aligned to the Non-seismically Qualified Fuel Pool Purification System.
The inspectors identified a Green NCV of TS 3.1.2.6, Borated Water Sources, for the failure to comply with the limiting conditions for operation, while the Refueling Water Storage Tank (RWST) was aligned to the non-seismic Fuel Pool Purification system (FPPS) for purification, causing the RWST to be inoperable. Specifically, when FPPS was aligned to the RWST, the licensee did not declare the RWST inoperable. The licensee took corrective actions (AR
- 422180) and revised OP-116.1, FPPS, to remove the capability to purify the RWST in Modes 1 through 4. The failure to comply with the actions of TS Limiting Condition for Operation (LCO) 3.1.2.6 while the Refueling Water Storage Tank (RWST) was aligned to the nonseismic FPPS for purification on May 24, 2010, causing the RWST to be inoperable, was a performance deficiency. The performance deficiency was more than minor because it affected the Design Control attribute of the Mitigating System cornerstone objective of ensuring the availability, ML11292A192
Page 44 of 54 reliability and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, when the FPPS was aligned to the RWST, the licensee did not declare the RWST inoperable. The inspectors evaluated the significance of this finding Using Attachment 4 of IMC 0609, the significance of this finding was determined to be of very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in loss of operability or functionality, did not represent a loss of system safety function, did not represent actual loss of safety function of a single train for longer than its TS Allowed Outage Time, did not represent an actual loss of safety function of one or more non-TS Trains of equipment designated as risk-significant, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The finding had a cross-cutting aspect of Conservative Assumptions, as described in the Decision Making component of the Human Performance cross-cutting area because, assumptions used in the justification to support the procedure change (i.e. a license amendment was not deemed required to support the procedure change) to OP-116.01 were non-conservative and the review of the issue in May 2010 did not adequately validate the assumptions (H.1(b)).
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Mitigating NORTH 12/31/2010 Green *SCWE: N *HP: N *PIR: N Systems ANNA Docket/Status: 05000338 (C) , 05000339 (C)
Open: 2010005 (PIM) Inadequate Design Control Measures for Field Changes Affecting Station Battery Cables The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the failure to ensure that design control measures for field changes impacting the support of station battery cables were commensurate with those applied to the original design requirements. The licensee entered this problem into their corrective action program as condition report 358461. The inspectors determined that the failure to adhere to the requirements of Criterion III for field changes involving the support of station battery cables was a performance deficiency (PD). This PD had a credible impact on safety due to an increase in battery post loading not analyzed by the vendor for a seismic event impacting the unsupported cables. The PD was more than minor, because it impacted the mitigating systems cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences and the related attribute of design controls due to changes made to battery cable supports which created a condition adverse to quality. In accordance with NRC Inspection Manual Chapter (IMC) 0609, Significant Determination Process, the inspectors performed a Phase 1 analysis and determined that the finding was of very low significance (Green) because the design deficiency did not result in the loss of functionality. The finding had no cross-cutting aspects because it is not indicative of current licensee performance.
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Mitigating 12/31/2010 WATTS BAR Green *SCWE: N *HP: N *PIR: N ML11292A192
Page 45 of 54 Systems Docket/Status: 05000390 (C)
Open: 2010005 (PIM) Failure to adequately qualify molded-case circuit breakers to safety-related application through commercial grade dedication.
- Green. The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the failure to assure that appropriate quality standards were specified and included in design documents and that deviations from such standards were controlled.
Specifically, the licensee failed to demonstrate the necessary conditions for commercial grade dedication and seismic qualification of molded case circuit breakers to safety-related application within the station 120VAC vital instrumentation boards. Corrective actions for this issue are still being evaluated and has been entered into the licensees corrective action program as PER 171695. Failure to specify appropriate qualification standards in performing commercial grade dedication of a component-level commodity is a performance deficiency. This performance deficiency is more than minor and a finding because it affected the design control attribute of the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, adequate measures were not implemented to ensure the station 120VAC vital instrumentation boards were properly seismically qualified for their application. The inspector assessed the finding using the SDP and determined that the finding was of very low safety significance (Green) because the breaker panels had originally been qualified by testing a complete prototype panel, while the licensees processes replaced a component-level item within that panel utilizing the original make and model component through commercial grade dedication.
The inspectors concluded that overall operability was not brought into question. This finding was reviewed for cross-cutting aspects and none were identified, as it was determined not to reflect current licensee performance. (Section 4OA5.2) 63)
Mitigating 01/03/2011 LASALLE Green *SCWE: N *HP: N *PIR: N Systems Docket/Status: 05000373 (C) , 05000374 (C)
Open: 2010006 (PIM) Supporting Structure for Standby Liquid Control System Test Tank Non-Functional During Postulated Design Basis Earthquake (DBE).
The team identified a finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to have an adequate calculation to demonstrate the seismic qualification of the standby liquid control (SBLC) system test tanks. Specifically, the licensee could not ensure that the Units 1 and 2 SBLC test tanks, if filled with water, would not collapse and damage nearby safety-related ML11292A192
Page 46 of 54 components during a design basis event. The licensee entered this finding into their corrective action program and drained the water from the SBLC test tanks to restore seismic qualification.
The team determined that this finding was more than minor because it was associated with the Mitigating Systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability of the SBLC system to respond to initiating events to prevent undesirable consequences (i.e., core damage). This finding was determined to be of very low safety significance (Green) utilizing the Risk-Assessment Standardization Project Handbook based on the frequency of seismic events. The finding did not have a cross-cutting aspect because it was not reflective of current performance. (Section 1R21.3.b.(1))
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Barrier Integrity 01/30/2009 PERRY Green *SCWE: N *HP: Y *PIR: N Docket/Status: 05000440 (C)
Open: 2009006 (PIM) FAILURE TO ADHERE TO PROCEDURES FOR SCAFFOLD AFFECTING CONTAINMENT SYSTEMS A finding of very low safety significance and associated non-cited violation of Technical Specification 5.4, Procedures, was identified by the team for the failure to erect scaffolding in accordance with procedural requirements. Specifically, scaffold constructed in the Intermediate Building had seismic bracing attached to a safety related cable tray support and was connected to a duct support without an approved engineering document as specified in procedural requirements. Although the licensee was able to demonstrate that the cable tray support and duct support were operable, the finding was determined to be more than minor because there was reasonable doubt that the licensee routinely performed engineering evaluations on similar scaffold issues. The finding was determined to be of very low safety significance because it did not represent an actual open pathway in the physical integrity of reactor containment. This finding had a cross-cutting aspect in the area of human performance, work practices, because the licensee failed to ensure supervisory and management oversight of work activities such that nuclear safety is supported. Specifically, the licensee failed to provide effective oversight of the erected seismic scaffold to ensure compliance with procedural requirements H.4(c).
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Barrier POINT 03/31/2010 Green *SCWE: N *HP: N *PIR: N Integrity BEACH Docket/Status: 05000266 (C) , 05000301 (C)
Open: 2010002 (PIM) Failure To Evaluate Seismic Piping Interactions A finding of very low safety significance and an associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified by the inspectors for the licensees ML11292A192
Page 47 of 54 failure to evaluate seismic piping interactions. Specifically, for a plant configuration where the stem of a spent fuel pool cooling system valve contacted an adjacent service water pipe, the licensee's evaluation to demonstrate that the existing spent fuel pool cooling system piping and valves met the design basis acceptance criteria of United States of America Standard (USAS)
B31.1-1967 used a method of analysis that did not evaluate the dynamic effect of impact forces as specified by the design basis piping code. The licensee entered this issue into its corrective action program. The finding was determined to be more than minor because it was associated with the Barrier Integrity Cornerstone attribute of design control and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, compliance with the seismic Category I design basis requirements of United States of America Standard (USAS) B31.1-1967 was to ensure valve SF-2, the valve connection between two sections of spent fuel pool cooling system piping, would function as required during a seismic Category I design basis event. The finding screened as having very low safety significance (Green) because it was a design deficiency of the structural integrity of the spent fuel pool cooling piping system that: did not result in loss of cooling to the spent fuel pool; did not result from fuel handling errors that caused damage to fuel clad integrity or a dropped assembly; and did not result in loss of spent fuel pool inventory greater than 10 percent of spent fuel pool volume. The finding had no cross-cutting aspect because it was a legacy design issue, not reflective of current performance.
66)
PRAIRE Integrity 08/30/2010 Green *SCWE: N *HP: N *PIR: N ISLAND Docket/Status: 05000282 (C) , 05000306 (C)
Open: 2010006 (PIM) Inadequate Analysis Used to Determine PORV/LTOP Setpoint The inspectors identified a finding having very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to have adequate calculation used to ensure reactor vessel 10 CFR Part 50, Appendix G limits are not exceeded. Specifically, the design calculation performed by Westinghouse to determine the pressurizer power operated relief valve (PORV) lift setting for low temperature overpressure protection (LTOP) analysis failed to include the correct inputs for mass addition transient, and also failed to consider the seismic and environmental terms in the instrument uncertainty calculations. The licensee subsequently entered this finding into their corrective action program and performed an operability evaluation and determined the PORVs remained operable and capable of performing their LTOP functions. The inspectors determined that this finding was more than minor because it was associated with the Barrier Integrity cornerstone attribute of design control and affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. This finding was of very low safety significance (Green) because it did not result in non-compliance with LTOP TS and the ML11292A192
Page 48 of 54 licensees operability evaluation concluded that based on the last testing of the PORV opening stroke time, the predicted peak pressure was determined to be below the adjusted Appendix G pressure limit. Therefore, the PORVs remained operable and capable of performing their LTOP functions. The finding did not have a cross-cutting aspect because it was not reflective of current performance. (Section 1R21.3.b.(3))
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Barrier 09/30/2010 BYRON Green *SCWE: N *HP: Y *PIR: N Integrity Docket/Status: 05000454 (C) , 05000455 (C)
Open: 2010004 (PIM) FAILURE TO PERFORM ADEQUATE EVALUATION FOR CRANE UPGRADE A finding of very low safety significance and an associated Non-Cited Violation (NCV) of Title 10 Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion III, Design Control was identified by the inspectors for the licensees failure to perform adequate evaluations to upgrade the single failure proof crane. Specifically, the inspectors identified six examples where the licensee failed to perform adequate evaluations in accordance with American Society of Mechanical Engineers (ASME) NOG-1-2004, requirements. The licensee documented the conditions in Issue Report (IR) 1099897, and IR 1100062 and initiated actions for calculation revisions and field modifications. The Fuel Handling Building (FHB) crane is designed to Seismic Category I requirements and the licensee used compliance with ASME NOG-1-2004, as the design basis for their crane upgrade to a single failure proof crane. The inspectors determined that the failure to perform adequate evaluations was contrary to ASME NOG-1-2004 requirements and was a performance deficiency. The finding was more than minor as it was associated with the Barrier Integrity cornerstone, because a fuel handling building crane heavy load drop can damage the Spent Fuel Pool (SFP) Cooling System or spent fuel cladding. The inspectors evaluated the finding using Inspection Manual Chapter (IMC) 0609.04, Phase 1 -
Initial Screening and Characterization of Findings, and based on a No answer to all of the questions in the Barrier Integrity column of Table 4a, determined the finding to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of Human Performance, Resources because the licensee did not provide adequate oversight of work activities, including contractors, such that design documentation was accurate to support nuclear safety. H.2(c) (Section 4OA5) 68)
Emergency *HP: *PIR:
06/30/2009 KEWANUEE Green *SCWE: N Preparedness N Y Docket/Status: 05000305 (C)
Open: 2009003 (PIM) Seismic Monitoring System Repeatedly Fails Surveillance ML11292A192
Page 49 of 54 A finding of very low safety significance (Green) and an associated Non-Cited Violation were identified by the inspectors for the licensees failure to maintain radiation monitoring instrumentation operable that was required by its emergency plans to meet the standards set forth in 10 CFR 50.47(b). Specifically, seismic instrumentation needed for two Emergency Action Levels, HU1.1 and HA1.1, was not maintained operable such that a related Unusual Event notification and an Alert declaration could have been made under certain conditions.
Corrective actions were taken for this issue and included revising Emergency Action Level (EAL) requirements to values within the range of the instrumentation. The inspectors determined that the issue was more than minor because the finding, if left uncorrected, would become a more significant safety concern. Specifically, for a seismic event, the deficiency could lead to the failure to declare an Unusual Event for a Natural and Destructive Phenomena Affecting the Plant Protected Area, HU1.1, and an Alert for a Natural and Destructive Phenomena Affecting the Plant Vital Area, HA1.1. The inspectors determined the finding could be evaluated using the Significance Determincation Process (SDP) and concluded that the risk significant planning standard problem was not a functional failure, nor did it represent a degraded function and, therefore, screened as an issue of very low safety significance (Green).
The inspectors determined this was a Green risk significant planning standard problem, rather than degraded or failed risk significant planning standard function, because the process failure affected only one Unusual Event and one Alert emergency classification. The inspectors determined this issue had a cross-cutting aspect in the area of Problem Identification and Resolution, trending and assessment, because the licensee failed to perform aggregate assessments that could have identified and prevented this and related issues earlier (P.1(b)).
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Miscellaneous 03/31/2010 WATTS BAR SL-IV *SCWE: N *HP: N *PIR: N Docket/Status: , 05000391 (C)
Open: 2010602 (PIM) Failure to Specify Requirements for Safety-related Conduit Supports On July 31, 2009, the applicant received 100 safety-related, Seismic Category I conduit supports from an outside vendor that were purchased under PO 78077. Sixty of these supports were made from a Unistrut channel welded perpendicular to a square base plate. The inspectors reviewed PO 78077 and identified that the applicant failed to specify the required welding code for the supports. As described in Section 3.10.3.3.2 of the Safety Analysis Report, the welding code of record for these supports was AWS (American Welding Society) D1.1-1972, Structural Welding Code, with Revisions 1-73 and 1-74. The applicant used Bechtel procedure VT-AWS-D1.1, Bechtel Nondestructive Examination Standard Visual Examination, Revision 0, to provide the requirements for performing direct visual examination of welds in structural steel components where AWS D1.1 applied. Step 5.1.6 of this procedure stated that, undercut shall not be more than 0.01 deep when its direction is transverse to primary tensile stress in the part that is undercut, nor more than 1/32 for all other situations. The inspectors independently inspected 18 supports purchased under this PO and identified 15 examples of ML11292A192
Page 50 of 54 potential unacceptable weld undercut. One of those supports was installed under EDCR 52419 for electrical conduit containing safety-related cables for the safety injection system.
Subsequently, the inspectors identified PO 63330 that purchased an additional 100 similar safety-related, Seismic Category I conduit supports from the same vendor. The inspectors examined 69 supports from PO 63330 and identified 14 examples of potential unacceptable weld undercut. All of the inspected supports were either installed or released for use (i.e.
accepted by QC). The inspectors identified undercut concerns with the following supports:
- PO 78077: STR-11, STR-19, STR-9, STR-18, STR1-4, STR1-17, STR1-13, STR1-8, STR2-9, STR2-14, STR-8, STR2-30, STR2-10, STR2-12, and 89513534-1 (installed support).
- PO 63330: PL1-95, PL1-48, PL1-5, PL1-53, PL1-44, PL1-12, PL1-75, PL1-68, PL1-25, PL1-27, PL1-66, PL1-77, PL1-37, and PL1-89. The applicant entered the issues into their corrective action program as PER 219039. The applicant inspected all available (151 of 160) supports and found unacceptable undercut associated with 47. Of the 47 supports identified, 22 were installed and 25 were released for use. Subsequently, the applicant issued a stop work order on all work related to conduit supports, placed all vendor supports that were in the warehouse on QC hold, and initiated an investigation to determine the root cause of the related issues. The inspectors determined that the failure to include all applicable fabrication requirements for the fabrication of safety-related components in the procurement documents was contrary to both applicant procedures and NRC requirements. The finding was determined to be greater than minor in accordance with IMC 2517, Appendix C, Watts Bar Unit 2 Construction Inspection Program, because the finding represented an improper or uncontrolled work practice that could impact quality involving safetyrelated components. Specifically, the failure to specify all applicable fabrication requirements for safety-related, Seismic Category I conduit supports in the procurement documents contributed to the applicants failure to ensure that installed and released-foruse supports met those requirements. The cause of this finding was not directly related to any of the cross-cutting area components as defined in IMC 0310, Components Within The Cross-Cutting Areas.
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SAN Miscellaneous 06/17/2010 SL-IV *SCWE: N *HP: Y *PIR: N ONOFRE Docket/Status: , 05000362 (C)
Open: 2010006 (PIM) Failure to report conditions that could have prevented fulfillment of safety function.
The inspectors identified a Severity Level IV noncited violation of 10 CFR 50.73, Licensee Event Report System, in which the licensee failed to submit a licensee event report within 60 days following discovery of an event meeting the reportability criteria. On January 26, 2010, the valve which isolates nonseismic piping from condensate storage tank T-120 failed its in-service test when the hand wheel stem snapped after a leveraging device was used in an attempt to close the valve. This isolation valve, 2HV5715, must be closed within 90 minutes of an operating basis earthquake in order to prevent the loss of condensate storage tank T-120 water ML11292A192
Page 51 of 54 inventory from a line break in the nonseismic portion of the condensate system. The failure of this valve resulted in a condition prohibited by Technical Specification 3.7.6 and therefore was reportable. This finding was entered into the licensees corrective action program as Nuclear Notification 200888616, and the licensee was taking actions to send a licensee event report to the NRC for this event. The inspectors determined that traditional enforcement was applicable to this issue because the NRC's regulatory ability was affected. Specifically, the NRC relies on the licensee to identify and report conditions or events meeting the criteria specified in regulations in order to perform its regulatory function. The inspectors determined that this finding was not suitable for evaluation using the significance determination process, and as such, was evaluated in accordance with the NRC Enforcement Policy. The finding was reviewed by NRC management, and because the violation was determined to be of very low safety significance, was not repetitive or willful, and was entered into the corrective action program, this violation is being treated as a Severity Level IV noncited violation consistent with the NRC Enforcement Policy. This finding was determined to have a crosscutting aspect in the area of human performance in the decision-making component in that the licensee did not make safety-significant decision using a systematic process, especially when faced with uncertainty. H.1(a)
(Section 4OA2.5d) 71)
Mitigating PRAIRE 07/09/2009 White *SCWE: N *HP: Y *PIR: N Systems ISLAND Docket/Status: , 05000306 (C)
Open: 2009010 Discussed: 2009013 , 2010009 , 2010012 (PIM) Failure to Ensure Design Measures Were Appropriately Established for The Unit 2 Component Cooling Water System An inspector identified apparent violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified due to the licensees failure to establish design control measures to ensure that the design basis for the Unit 2 CCW system was correctly translated into specifications, drawings, procedures, and instructions. Specifically, the licensee failed to ensure that the safety related function of the CCW system was maintained following initiating events (such as high energy line break, seismic or tornado events) in the turbine building. This issue has been preliminarily determined to be of low to moderate safety significance (White). This issue was entered into the licensees corrective action program as corrective action document 1145695. Upon identifying this issue, the licensee immediately declared the Unit 2 CCW system inoperable and entered Technical Specification 3.0.3. The Technical Specification was exited following the closure of several system isolation valves approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> later. The closure of the isolation valves prevented the Unit 2 CCW system from being vulnerable to failure following events in the turbine building. This finding was determined to be more than minor because it impacted the design control and external events aspects of the Mitigating Systems Cornerstone. The finding also impacted the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to ML11292A192
Page 52 of 54 prevent undesirable consequences. The initiating events in the turbine building could cause the CCW piping to fail. Loss of CCW inventory affects both trains of CCW based on the piping arrangement. The loss of both trains of CCW required a phase 3 significance determination.
The results of the phase 3 assessment showed a delta core damage frequency of 3.2E-6, White. The cause of this finding was related to the cross cutting element of Human Performance, Decision Making because the licensee failed to make safety significant and risk significant decisions using a systematic process to ensure that safety was maintained (H.1(a)).
Since both the Unit 1 and Unit 2 cross-cutting aspects are from the same performance deficiency and are separated based on the risk determination, the aspect of H.1(a) counts as one cross-cutting aspect in this report. (Section 4OA5.1). Final determination letter issued with report number 2009-013 on September 3, 2009.
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Miscellaneous 06/30/2010 WATTS BAR SL-IV *SCWE: N *HP: Y *PIR: N Docket/Status: , 05000391 (O)
Open: 2010603 (PIM) Failure to Adequately Evaluate and Qualify Molded Case Circuit Breakers The applicant installed breakers into the 4 divisions of 120VAC Vital Instrument Power Boards.
The electrical loads in the power boards are safety related and necessary to shut down and maintain the plant in a safe condition. The following describes the violation with two examples resulting from the inspectors observations and review of the records and time line involved in the installation of the breakers. Example 1: The inspectors found that significant differences existed between the original qualification mounting and the actual mounting method of the breakers in the power boards. During the 1992 qualifications testing, (report S522-RP-02) the breakers were individually mounted to a plate with screws and no front cover. The actual breakers are mounted in the power boards by clamping 12 breakers with a 36 front cover against two horizontal angle iron supports in the rear and without additional screws. The actual mounting method used introduces different shock and random frequency impacts during a postulated seismic event to the breakers that were not simulated nor analyzed in the tested mounting method. Additionally, the actual mounting method exposes the bus bar to additional seismic loading not simulated in the qualification testing. The front cover plate is held in place with one screw in each corner and introduces an uncertain clamping pressure across the 12 breakers. The inspectors determined that the tested breaker mounting did not envelope the mounted condition of the breakers in the power boards. Example 2: The inspectors reviewed WO 08-816370-000 used to replace old and add new breakers into the WBN-2-BD-235-0003 120VAC vital instrument power board. The inspectors determined the applicant received breakers in which two attributes had been changed. The depth (Z dimension), a critical characteristic, was reduced and the auxiliary contact was moved from the outside to the inside of the breakers. The 1992 qualification report indicated a Z dimension of 3.75 and the manufacturers catalog now indicates 2.609 for the critical characteristic. The applicant recognized that the breaker with the smaller Z dimension would not mount into the power board.
Page 53 of 54 The inspectors determined that the applicant modified the breaker by attaching a Micarta plate to the rear of the breaker using 4 nuts and bolts to fit them in the power boards as described in example 1 without updating the qualification package. The inspectors determined the applicant did not analyze if any detrimental effects may have been introduced by the modification, demonstrate an adequate review for suitability of application, or the impact on other components. The inspectors reviewed calculation WCG-ACQ-1004, Rev. 1, used by the applicant to demonstrate qualification by analysis. The inspectors verified the calculations purpose was to qualify the breakers and the calculation concluded the breakers were qualified.
The calculation only determined the power board weight change and the shift in natural frequency of the power board and its effect on the power board floor anchor bolts. However, the calculation did not address the breaker performance during a seismic event. The inspectors determined that the conclusion that the breakers were qualified was not validated by the calculation. The inspectors reviewed problem evaluation report (PER) 227786 dated May 10, 2010 for the qualification findings identified by the inspectors. The inspectors determined that the PER was inadequate. The PER characterized the seismic evaluation calculation WCG-ACQ-1004, Rev. 1 as having evaluated and addressed the seismic qualification of the breakers.
The inspectors determined the calculation did not address the qualification of the breakers but rather the qualification of the floor anchor bolts of the power boards. The PER further stated that the breaker installation was in compliance with the 1992 qualification report S522-RP-02. The inspectors determined that the 1992 seismic qualification of the breakers did not envelope the breaker mounting method used for the breakers in the power boards. The 1992 seismic qualification report did not envelope the breaker mounting method used either before or after the attribute changes imposed by the manufacturer occurred. The PER stated that the breakers are acceptable and the integrity of the breakers in the power boards is maintained. No further action was required by the applicant in the corrective actions of the PER. The inspectors determined the PER did not address the effects that changes in attributes may impose on safety, verify the conclusions made in the above calculation, address the qualification of the breakers, or the different mounting method in the 1992 qualification report. The inspectors determined the PER failed to adequately address the findings identified and would not ensure that compliance was restored in a reasonable timeframe or prevent recurrence in accordance with the enforcement policy. Therefore, the criterion for a noncited violation was not met and a Notice of Violation is warranted, requiring a formal response from the applicant. Example 1 is more than minor because the finding represents a deviation that, if left uncorrected, could adversely affect the environmental or seismic qualification of a component. Example 2 is more than minor because the violation represents an inadequate process, procedure, or quality oversight function that, if left uncorrected, could adversely affect the quality of the fabrication, construction, testing, analysis, or records of a safety-related component. The cause of this finding was directly related to the decision-making component of the Human Performance cross-cutting area because the applicant failed to appropriately review safety significant decisions to verify the validity of the underlying assumptions and identify possible unintended consequences. H.1 (b) An unresolved item, URI-05000390/2009002-03, was identified on Unit 1 related to the adequacy of the same seismic qualification report for breakers associated with station 120VAC Vital Instrumentation Boards. Enforcement: 10 CFR 50, Appendix B, Criterion ML11292A192
Page 54 of 54 III, Design Control, states that measures shall be established for the review for suitability of application of materials, parts, and equipment that are essential to the safety-related functions of the structures, systems, and components (SSCs). The design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Where a test program is used to verify the adequacy of a specific design feature in lieu of other verifying or checking processes, it shall include suitable qualifications testing of a prototype unit under the most adverse design conditions. Contrary to the above, measures used to review for the suitability of application of materials, parts, and equipment essential to the safety related functions of molded case circuit breakers and measures to provide for the verification of checking the adequacy of design, such as, calculational methods, performing a suitable test program, including qualifications testing of a prototype unit under the most adverse conditions were not adequate in that: 1. On October 5, 2009, the applicant installed molded case breakers into the 120VAC vital instrument power boards, however, the test program used to qualify a prototype breaker failed to use a suitable mounting method that reflected the most adverse mounting condition. 2. On September 3, 2009, the applicant failed to perform an adequate review for suitability of application parts and material used to modify dimensional critical characteristic in molded case breakers, and further, the applicant failed to verify the adequacy of design for the modification and the effects on essential safety related functions of the breakers. This is identified as violation (VIO) 005000391/2010603-08, Failure to Adequately Evaluate and Qualify Molded Case Circuit Breakers.