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| Title | Quarter | Description | |
|---|---|---|---|
| 05000528/FIN-2004003-02 | Containment Purge Penetration Nonconformance | 2004Q2 | A Severity Level IV noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified for the failure to correct a nonconforming condition in a timely manner. Specifically, since June 2001, the licensee discontinued implementation of required Technical Specification surveillance testing for the containment purge valves by declaring the valves inoperable and installing blind flanges. This issue was entered into the corrective action program as CRDR 2711167 The finding is greater than minor because the licensee's failure to submit a license amendment to correct the nonconforming condition impacted the NRC's ability to perform its regulatory function. Therefore, this finding was considered applicable to traditional enforcement. Although the significance determination process is not designed to assess the significance of violations that potentially impact or impede the regulatory process, the finding can be assessed using the significance determination process. Using the Phase 1 worksheet in Manual Chapter 0609, "Significance Determination Process," the finding is determined to have very low safety significance because it only affected the barrier integrity cornerstone and the installation of blind flanges adequately maintained containment integrity. |
| 05000528/FIN-2004003-06 | Failure to Perform a Complete Shut Down Cooling Heat Exchanger Temperature LOOP Channel Calibration | 2004Q2 | A Severity Level IV noncited violation of Technical Specification 3.3.11 was identified for the failure to include the resistance temperature detectors in the channel calibration for the shutdown cooling heat exchanger temperature instruments. Specifically, prior to the implementation of Improved Technical Specifications, the licensee did not perform testing of the resistance temperature detectors. Following the implementation of Improved Technical Specifications, the licensee did not perform an in-place qualitative assessment of the resistance temperature detectors' behavior. This issue was entered into the corrective action program as CRDR 280178 The failure to perform a complete shutdown cooling heat exchanger temperature loop channel calibration is determined to have greater than minor significance because the licensee's failure to report the condition impacted the NRC's ability to perform it's regulatory function. Therefore, this finding was considered applicable to traditional enforcement. Although the significance determination process is not designed to assess the significance of violations that potentially impact or impede the regulatory process, the finding can be assessed using the significance determination process. Using the Phase 1 worksheet in Manual Chapter 0609, "Significance Determination Process," this finding is determined to be of very low safety significance because it only affected the mitigating system cornerstone and the resistance temperature detectors were found to be within calibration. |
| 05000528/FIN-2004006-02 | Failure to Provide an Evaluation of a Change to the Facility as Described in the UFSAR, Under 10 CFR 50.59 Requirements | 2004Q2 | The team identified a Severity Level IV violation of 10 CFR 50.59 requirements for failing to evaluate a modification to spent fuel storage in the spent fuel pools. The team reviewed CRDR 2524176, regarding the lack of a criticality analysis to support the use of rod capture tubes, which hold individual harvested fuel pins, in the spent fuel rack. The team reviewed the licensee's process of storing individual fuel pins, removed from a parent fuel assembly, and placed in rod capture tubes to be located in guide tubes of another host assembly. This resulted in a component that had nuclear fuel pins, of varying enrichment and depletion, stored as a regular fuel assembly in the spent fuel pools. The team noted that Section 9.1 of the UFSAR specifically described the storage of spent fuel in regions based upon fuel assembly initial enrichment, actual burnup, and actual decay time. The UFSAR does not describe the storage of individual pins in these regions. The licensee previously interpreted this as meaning the UFSAR did not prohibit such storage, and would not require consideration of enrichment, burnup, and decay of individual pins. The licensee failed to provide an evaluation of a change to the facility as described in the UFSAR, under 10 CFR 50.59 requirements. The licensee subsequently performed an evaluation of the criticality under station procedure 72DP-9NF01, "Control of SNM Transfer and Inventory," which was found acceptable The issue was determined to be more than minor, through Inspection Manual Chapter 0612, Appendix B, in that it affected the barrier integrity cornerstone attribute of human performance, and could have represented a more significant issue if left uncorrected. In accordance with the NRC Enforcement Manual, violations of 10 CFR 50.59 are not processed through the significance determination process. Therefore, this issue was considered applicable to traditional enforcement. Although the significance determination process is not designed to assess significance of violations that potentially impact or impede the regulatory process, the result of a 10 CFR 50.59 violation can be assessed significance through the significance determination process. The team leader and the Region IV senior reactor analyst discussed the significance of this finding. An SDP Phase 1 screening was performed and the finding was determined to have very low safety significance because there was no actual loss of the barrier integrity function. The licensee entered this issue into its corrective action program as CRDR 2711241. |
| 05000528/FIN-2004014-01 | Failure to Maintain Design Control of Containment Sump Recirculation Piping | 2004Q4 | The team identified an apparent violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to establish measures to assure design basis information was translated into specifications, drawings, procedures, and instructions. Specifically, the licensee failed to maintain the safety injection sump suction piping full of water in accordance with the Updated Final Safety Analysis Report. This nonconformance had the potential to significantly affect the available net positive suction head described in the Updated Final Safety Analysis Report for the high pressure safety injection and containment spray pumps, since the analysis assumed the piping would be maintained full of water. This finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely affects the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. The finding has a potential safety significance greater than very low significance (i.e., Greater than Green) based on the results of a Significance Determination Process, Phase 3 analysis |
| 05000528/FIN-2004014-03 | Failure to Perform Written Safety Evaluation in Accordance with 10 CFR 50.59 Requirements | 2004Q4 | The team identified three examples of a noncited, Severity Level IV violation of 10 CFR 50.59 requirements involving the failure to perform written safety evaluations prior to implementing changes to the facility. The first example involved a change for using manual actions in lieu of automatic actions as compensatory measures to support the safety functions of the high pressure safety injection and containment spray systems during postulated design basis loss-of-coolant accident conditions following a recirculation actuation signal. The second example involved operation of emergency core cooling systems with a 10-20 cubic foot void in the suction piping. The third example involved the failure to perform a written safety evaluation for changes involving filling the containment sump with borated water to a level above the containment sump safety injection recirculation piping. These changes were implemented in response to identifying that the safety injection system was not being maintained full of water In accordance with Inspection Manual Chapter 0612, Appendix B, \"Issue Disposition Screening,\" the team determined that traditional enforcement applied because this finding may have impacted the NRC\'s ability to perform its regulatory function. The severity level of this finding was assessed as having very low safety significance reflective of a Severity Level IV violation. This determination was based in part on use of the significance determination process. |
| 05000528/FIN-2004014-04 | Failure to Obtain Prior NRC Approval for a Change to the Facility Involving Maintaining a Significant Segment of Containment Sump Safety Injection Recirculation Piping Void of Water | 2004Q4 | The team identified an apparent violation of 10 CFR 50.59 requirements for the licensee\'s failure to perform a written safety evaluation and receive NRC approval prior to implementing changes to the facility in 1992 which involved draining, and maintaining drained, a significant segment of containment sump safety injection recirculation piping during normal plant operations. This change resulted in the failure to maintain the safety injection piping full of water in accordance with the Updated Final Safety Analysis Report. This represented an unreviewed safety question since it increased the probability of a malfunction of equipment important to safety previously evaluated in the safety analysis report. In accordance with Inspection Manual Chapter 0612, Appendix B, Issue Disposition Screening, the team determined that traditional enforcement applied because this finding may have impacted the NRCi12s ability to perform its regulatory function. This is an apparent violation pending the results of a predecisional enforcement conference |
| 05000528/FIN-2005002-06 | Failure to Obtain Prior NRC Approval for a Design Change to the Facility | 2005Q1 | A Severity Level IV non-cited violation of 10 CFR 50.59 requirements was identified for the failure to obtain a license amendment for a permanent modification to all six station emergency diesel generators. The inspectors determined that there were two modifications performed on the jacket water system of each emergency diesel generator. Condition Report/Disposition Request (CRDR) 130208, in 1993, directed the abandonment of the jacket water surge tank makeup valves on both emergency diesel generators of all three units. A recent modification, Design Modification Work Order 220055 in 2003, removed the surge tank low level alarm on both emergency diesel generators of all three units. The licensee replaced these two automatic actions (automatic makeup and low level alarm) with a manual operator action to fill, as necessary, every 12 hours during rounds. The inspectors reviewed the updated final safety analysis report (UFSAR) and design basis documents, and found that the automatic jacket water surge tank makeup, and the low level alarm, were both shown in UFSAR descriptions, drawings, and design value tables. The issue was determined to be more than minor, through Inspection Manual Chapter 0612, Appendix B, in that it affected the mitigating systems cornerstone attribute of equipment performance, and was repeated for all of the station emergency diesel generators. The issue was determined to result in more than a minimal increase in the consequences of a malfunction of an structure, system, or component important to safety evaluated in the UFSAR, since jacket water leakage could go undetected for up to 12 hours and affect diesel operability. Thus, a license amendment was required. In accordance with the NRC Enforcement Manual, violations of 10 CFR 50.59 are not processed through the significance determination process. Therefore, this issue was considered applicable to traditional enforcement. Although the significance determination process is not designed to assess significance of violations that potentially impact or impede the regulatory process, the result of a 10 CFR 50.59 violation can be assessed significance through the significance determination process. The lead inspector and the Region IV senior reactor analyst discussed the significance of this finding. An SDP Phase 1 screening was performed and the finding was determined to have very low safety significance because there was no actual loss of the mitigating system safety function. The licensee entered this issue into its corrective action program as CRDR 2711244. |
| 05000528/FIN-2005004-08 | Incomplete and Inaccurate Information Associated with the Ex-Core Safety Channels | 2005Q3 | The inspectors identified a noncited Severity Level IV violation of 10 CFR 50.9 for providing incomplete or inaccurate information to the NRC. Specifically, the licensee provided incomplete and inaccurate information regarding the design control of ex-core safety channel log power instrument setpoints. This information was determined to be material in that it affected the NRC's ability to determine compliance with NRC requirements. This issue was entered into the licensee's corrective action program as Condition Report/Disposition Request 2829051 This finding was not assessed via NRC Manual Chapter 0609, Significance Determination Process, because the licensees actions impeded the regulatory process. Therefore, this finding was assessed in accordance with the NRC Enforcement Policy. The finding is associated with the mitigating systems cornerstone. The inspectors determined that engineering personnel had additional information, including the subsequently corrected revision of the calculation going through final verification, and additional explanatory setpoint procedures, which were not referenced or provided during the original correspondence by the licensee. Had the complete and accurate information been supplied at the time of the original request in 2003, the NRC would have identified a design control violation at that time. The safety consequence of this issue is of very low safety significance, in that there was no actual loss of a safety function. |
| 05000528/FIN-2005005-04 | Failure to Submit LER to Report Shutdown Required by Technical Specifications | 2005Q4 | The inspectors identified a noncited Severity Level IV violation of 10 CFR 50.73 for the failure to submit a licensee event report within 60 days to report the completion of a plant shutdown required by the Technical Specifications. A second similar example of a violation of the same regulation was identified by the licensee. Specifically, the licensee was required to submit a licensee event report by May 17, 2005, to report the completion of a plant shutdown required by the Technical Specifications that occurred on March 18, 2005. This licensee event report was submitted on November 7, 2005. Additionally, the licensee was required to submit a licensee event report by April 10, 2005, to report the completion of a plant shutdown that occurred on February 9, 2005. A revised licensee event report was submitted on January 6, 2006. This issue was entered into the licensee's corrective action program as Condition Report/Disposition Requests 2829976 and 2844019. The finding was determined to be applicable to traditional enforcement because the NRCs ability to perform this regulatory function was potentially impacted by the licensees failure to report the event. The finding was determined to be a Severity Level IV violation in accordance with Section D.4 of Supplement I of the NRC Enforcement Policy. The finding is not suitable for evaluation using the significance determination process, but has been reviewed by NRC management and is determined to be a finding of very low safety significance. The cause of the finding is related to the crosscutting element of problem identification and resolution in that the transportability review, conducted by regulatory affairs personnel, failed to identify an additional example of a missed reportable event that was subsequently identified by the NRC. |
| 05000528/FIN-2005012-05 | Summary Finding. 95002 Inspectors Assessment of IR2004-14 Severity Level Iii Violation for 50.59 Issue | 2005Q4 | The U.S. Nuclear Regulatory Commission (NRC) performed this supplemental inspection, in part, to assess the licensee's evaluation and corrective actions associated with an inappropriate change to an emergency core cooling system procedure without prior NRC approval. This procedure change rendered portions of the system inoperable because of voiding. This performance issue was previously characterized as a Severity Level III violation of 10 CFR 50.59 and was originally identified in NRC Inspection Report 05000528; 529; 530/2004014. During this supplemental inspection, performed in accordance with Inspection Procedure 95002, the inspectors determined that the licensee's evaluation identified the primary root causes of the performance issue to be: (1) The site procedure revision process (01AC-0AP02) was inadequate, in that, the procedure allowed pre-screening' of changes that could potentially bypass performing a 10 CFR 50.59 screening for changes to the facility as described in the licensing basis; and (2) The corrective action program implementation was ineffective. The licensee also identified overlap and interface problems between the corrective action program, the engineering evaluation request program, and the instruction change request program. These issues, in conjunction with inadequate training to recognize a corrective action condition, contributed to the failure of station personnel to initiate a corrective action program input document in 1992 for the potential pipe voiding concern. The inspectors concluded that the licensee's evaluation and implemented corrective actions were appropriate to reasonably prevent repetition of the 10 CFR 50.59 violation. Given the licensee's acceptable performance in addressing the inappropriate procedure change and 10 CFR 50.59 program deficiencies, the Severity Level III violation is closed. |
| 05000528/FIN-2005012-06 | Summary Finding. 95002 Inspectors Assessment of IR2004-14 (Yellow) 10CFR50, App B, Criterion Iii Violation | 2005Q4 | The NRC performed this supplemental inspection, in part, to assess the licensee's evaluation and corrective actions associated with potential air entrainment into the emergency core cooling system. The licensee failed to incorporate original design requirements into the plant to maintain piping between the containment sump isolation valves filled with water. This performance issue was previously characterized as a 10 CFR 50, Appendix B, Criterion III, violation having substantial safety significance (Yellow), and was originally identified in NRC Inspection Report 05000528; 529; 530/2004014. The inspectors determined that the licensee's evaluation identified a direct cause, nine root causes, and nine contributing causes of the performance issue. The evaluation was also used to develop an extensive list of corrective actions. The inspectors found the licensee's methods of evaluation to be appropriate. The NRC concluded that, while the licensee performed an adequate root cause evaluation of the Design Control violation, certain corrective actions were incomplete at the time of this inspection. Specifically, the team determined that for each of the root and contributing causes, not all corrective actions were sufficiently developed to ensure that the identified performance deficiencies were adequately addressed. In addition, some of the corrective actions were narrowly focused, or the implementation of those actions was not fully effective. Also, the team concluded that criteria and reviews were not established, for auditing or followup, to ensure that corrective actions were effective in improving performance in the affected areas. Consequently, the team did not have assurance that the planned corrective actions were sufficient to address the causes for the performance deficiencies associated with the violation. Therefore, the (Yellow) violation (VIO 2004/014-01) will remain open for further NRC review. |
| 05000528/FIN-2006004-06 | Failure to Submit Complete Revisions to the Updated Final Safety Analysis Report for Temporary Modifications | 2006Q3 | The inspectors identified a noncited violation of 10 CFR 50.71(e)(4) for the failure to file revisions to the Updated Final Safety Analysis Report. Specifically, Procedure 93DP-0LC03, Licensing Document Maintenance, Revision 13, Step 3.5.6, required that temporary modifications that are in place for greater than 24 months be incorporated into the Updated Final Safety Analysis Report. Temporary modifications for heated junction thermocouples were installed for greater than 24 months and a revision to the Updated Final Safety Analysis Report was not made. This issue was entered into the licensee\'s corrective action program as Condition Report/Disposition Request 2894741. The performance deficiency associated with this finding involved the failure of licensee personnel to submit revisions to the Updated Final Safety Analysis Report reflecting temporary modifications installed in Unit 3 for more than 24 months. The finding was determined to be applicable to traditional enforcement because the NRCs ability to perform its regulatory function was potentially impacted by the licensees failure to revise the Updated Final Safety Analysis Report in a timely manner. The finding was determined to be a Severity Level IV violation in accordance with Section D.4 of Supplement I of the NRC Enforcement Policy. The finding is not suitable for evaluation using the significance determination process, but has been reviewed by NRC management and is determined to be a finding of very low safety significance. This finding has a crosscutting aspect in the area of human performance associated with work practices because not following established procedures led to an inaccurate Updated Final Safety Analysis Report (Section 4OA2). |
| 05000528/FIN-2006004-07 | Failure to Submit Complete Revisions to the Updated Final Safety Analysis Report for Permanent Modifications | 2006Q3 | The inspectors identified a violation of 10 CFR 50.71(e)(4), for which enforcement discretion was exercised, that involved the failure to file revisions to the Updated Final Safety Analysis Report. Specifically, for the reporting period between January 2003 and December 2005, licensing personnel failed to submit a revision to the Updated Final Safety Analysis Report reflecting core protection calculator system modifications. The issued was entered into the licensees corrective action program as Condition Report/Disposition Request 2894635. The performance deficiency associated with this finding involved the failure of licensee personnel to submit revisions to the Updated Final Safety Analysis Report reflecting modifications installed in Unit 2 for more than 24 months. The finding was determined to be applicable to traditional enforcement because the NRCs ability to perform its regulatory function was potentially impacted by the licensees failure to revise the Updated Final Safety Analysis Report in a timely manner. Normally, the violation would be categorized at Severity Level IV in accordance with Section D.4 of Supplement I of the NRC Enforcement Policy. However, in accordance with Section VII.B.6 of the NRC Enforcement Policy, the NRC is refraining from taking enforcement action because of the NRC action taken in 1993 to issue and then retract a similar occurrence and the low safety significance of the finding (EA-06-267). The finding is not suitable for evaluation using the significance determination process, but has been reviewed by NRC management and is determined to be a finding of very low safety significance. |
| 05000528/FIN-2006010-01 | Summary Finding. 95002 Team'S Assessment of IR 2004-14 (Yellow) 10 CFR Part 50, Appendix B, Criterion Iii, Violation | 2006Q3 | The NRC performed a followup supplemental inspection to assess the licensees corrective actions associated with a Yellow design control finding involving the potential for air entrainment into the emergency core cooling system. The team concluded that the technical issues specifically associated with the voided emergency core cooling system piping have been addressed. However, the Yellow finding will remain open because the licensee did not implement effective corrective actions for all of the causes associated with the Yellow finding. Specifically, the licensees actions to improve questioning attitude, technical rigor, and technical review were not fully effective. Also, the implementation of performance measures and metrics to monitor the effectiveness of corrective actions associated with the Yellow finding were not adequate to assess effectiveness. This performance issue was previously characterized as a 10 CFR Part 50, Appendix B, Criterion III, violation having substantial safety significance (Yellow), and was originally identified in NRC Inspection Report 05000528; 05000529; 05000530/2004014. The licensees corrective actions taken in response to the root causes and related programmatic concerns involving questioning attitude, technical rigor, and technical review have not been completely effective. Specifically, following implementation of corrective actions between September 2005 and March 2006, the licensee: (1) continued to conduct inadequate technical reviews of emerging issues; (2) did not routinely question the validity of engineering assumptions used to support operability decisions; (3) did not consistently implement a qualify, validate, and verify process; and (4) did not consistently notify operations personnel of immediate operability concerns. The team concluded that adequate qualitative or quantitative measures for determining the effectiveness of the corrective actions to prevent recurrence have not been established. For example, not all relevant performance data was considered when performance monitoring measures were developed to assess the effectiveness of corrective actions. When the pertinent data was considered, or otherwise clarified, the performance measures suggested declining rather than improving performance in some areas. The team also concluded that the licensee had not completed adequate reviews of the effectiveness of corrective actions prior to their notifying the NRC of their readiness for inspection of the Yellow finding. Specifically, several assessments were completed after the requested dated of the inspection (June 2006). Several of the assessments noted that insufficient progress in resolving some of the root and contributing causes had been made. Additionally, a standard guideline for metrics was not issued and implemented until July 2006. |
| 05000528/FIN-2006011-03 | 10 CFR 50.59 Reviews not performed or Inadequate for multiple changes to spray pond chemistry control procedure | 2006Q3 | A noncited violation of 10 CFR 50.59 was identified for making nine revisions to Procedure 74DP-9CY04, System Chemistry Specification, a procedure described in the Updated Final Safety Analysis Report between 1998 and 2004. Specifically, the licensee failed to perform evaluations for Revisions 3, 6, 8, 10, 12, 24, 28, 32, and 36 and performed inadequate evaluations for Revisions 10 and 36, to assess the potential impact of the changes on the safety-related components in the spray pond system. Each of these changes revised spray pond chemistry parameter limits which were subsequently determined to have contributed to heat exchanger fouling. Failure to adequately evaluate the impact of changes to the Chemistry Control Program was a performance deficiency. Because this violation had the potential to impact the NRC?s regulatory function, and because the associated significance was determined to be Green using Phase 3 of the significance determination process, this violation is being treated as a Severity Level IV violation. This issue was entered into the Corrective Action Program under CRDR 2902498. Because this violation was of very low safety significance and has been entered into the corrective action program, it is being treated as a noncited violation consistent with Section VI.A of the Enforcement Policy: NCV 05000528; 05000529; 05000530/2006011-03, 50.59 Reviews Not Performed or Inadequate for Multiple Changes to Spray Pond Chemistry Control Procedure. |
| 05000528/FIN-2006012-03 | Failure to Implement the Operability Determination Process | 2006Q4 | The team identified two examples of a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to perform operabilty determinations. In both examples, the licensee failed to perform an operability determination following identification of a degraded condition that had the potential to adversely affect the safety function of all emergency diesel generators. Specifically, an operability determination was not performed after identifying the failure of the Unit 3 Train A emergency diesel generator on July 25, 2006, was potentially the result of plastic debris affecting proper auxiliary contact operation of a K-1 relay. The licensee determined the debris most likely originated from a modification performed on all emergency diesel generator K-1 relays during initial plant startup. Following another failure of the Unit 3 Train A emergency diesel generator on September 22, 2006, an operability determination was not performed after identifying the failure was the result of the K-1 relay actuating arm not providing adequate compression of the auxiliary contacts. The licensee determined this degraded condition most likely originated during implementation a modification done to all emergency diesel generator K-1 relays during initial plant startup. This finding is greater than minor because the failure to follow the operability determination process, if left uncorrected, would become a more significant safety concern in that degraded or nonconforming conditions would not be properly evaluated. Using the Phase 1 worksheet in NRC Inspection Manual Chapter 0609, Significance Determination Process, the finding was determined to have very low safety significance because unreliable K-1 relay operation resulted in no actual loss of safety function of the other five emergency diesel generators prior to corrective actions being implemented, and the finding did not represent a potential risk significant condition because of a seismic, flooding, or severe weather event. This issue is documented in the licensees corrective action program as Condition Report/Disposition Requests 2928389 and 2940558. The cause of this finding is related to the crosscutting element of problem identification and resolution in that engineering personnel failed to properly evaluate and perform operability determinations for identified degraded conditions affecting the emergency diesel generators (Section 4.0). |
| 05000528/FIN-2007011-06 | Licensee-Identified Violation | 2007Q2 | 10 CFR 50.63(a)(1) requires that a licensed nuclear power plant must be able to withstand and recover from an station blackout event. To meet this requirement, Section 8.3.1.1.10 of the Palo Verde Nuclear Generating Station Updated Final Safety Analysis Report (UFSAR) states that the alternate ac power system is capable of energizing the required loads within one hour of the onset of an station blackout. The UFSAR also states that a study was performed to demonstrate that Palo Verde Nuclear Generating Station is capable of coping with a station blackout for that initial one-hour period. Contrary to this requirement, the licensee determined, as the result of five tests, that it took from 61 minutes 30 seconds to 67 minutes 30 seconds to energize the required loads. This issue is documented in the licensees corrective action program as Palo Verde Action Request PVAR 2970059. This finding is of very low safety significance because testing has demonstrated that, even at the most limiting time of 67 minutes, 30 seconds, Palo Verde Nuclear Generating Station could withstand an station blackout |
| 05000528/FIN-2007012-18 | Failure to Periodically Update the Updated Final Safety Analysis Report | 2007Q4 | The team identified a Severity Level IV NCV of 10 CFR 50.71(e) for the failure of the licensee to periodically update the UFSAR with all changes made in the facility or procedures. While conducting a review of the Unit 2 liquid radiological waste system, the team found that the system was not being operated in accordance with the description provided in the UFSAR. Specifically, evaporator concentrate was being pumped to one of the high total dissolved solids (TDS) holdup tanks rather than the concentrate monitor tanks as specified in Section 11.2.2 of the UFSAR. The licensee stated that the Unit 2 concentrate monitor system had been out of service since 2002. The teams review of corrective action documents related to the system determined that the concentrate monitor tanks were not being used because of equipment/maintenance issues with the concentrate monitor system. The UFSAR stated in Section 11.2.2.4.1.2, that flow from the high TDS holdup tank can be terminated or diverted to an alternate path by operator action based on evaporator or holdup pump malfunction, high-pressure drop across the adsorption bed or ion exchangers, an exhausted resin bed, or when the radiological waste section leader determines it is necessary. The UFSAR did not specify the alternate flow path nor the allowed duration. The team concluded that operating outside of the UFSAR design basis for approximately 5 years was not the intent of UFSAR Section 11.2.2.4.1.2. Analysis: The team determined that the failure to update the UFSAR to reflect changes made to the facility was a performance deficiency. This issue was subject to traditional enforcement because it had the potential for impacting the NRCs ability to perform its regulatory function. The finding is characterized as a Severity Level IV violation because the erroneous information in the UFSAR was not used to make an unacceptable change to the facility or procedures. The cause of this finding had a crosscutting aspect associated with resources of the human performance area in that the licensee failed to ensure that personnel and equipment were available and adequate to maintain radiological safety by minimization of long-standing equipment issues (H.2.(a)). Enforcement: 10 CFR 50.71(e) requires that the licensee periodically update the USFAR with all changes made in the facility or procedures. Contrary to the above, in 2002 the licensee made a change to the facility and procedures as described in the UFSAR and failed to update the UFSAR. Specifically, the licensee began operating the Unit 2 liquid radiological waste system in a manner different than that specified by UFSAR when they commenced pumping evaporator concentrate to the high TDS holdup tanks rather than the concentrate monitor tanks as specified in UFSAR Section 11.2.2. The failure to update the UFSAR was characterized as a Severity Level IV violation. The finding was of very low safety significance because the change in operation of the total dissolved solids holdup tanks did not result in an increase in the likelihood of a release of radioactive material. This issue was entered in the licensees CAP as PVAR 3075089. This violation was treated as an NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV 05000529/2007012-18, Failure to Periodically Update the Updated Final Safety Analysis Report |
| 05000528/FIN-2007012-19 | Routine Heavy Use of Overtime | 2007Q4 | The team identified an unresolved item (URI) associated with Technical Specification 5.2.2.d. for the routine use of heavy amounts of overtime for operations personnel. Interviews with frontline personnel and managers in operations indicated that shortages of licensed operators and operator training personnel were perceived to be the most significant issue facing the operations organization. Interviewees reported that the licensed operator training pipeline was interrupted several times after 2000 with a resulting net loss of 20 licensed operators by 2007 (see chart below). This loss occurred concurrently with a reduction from 6 operator shifts to 5 self-relieving shifts (i.e., shift crews that have sufficient numbers of personnel to ensure that regulatory and administrative control room staffing requirements can be met without overtime or assigning a member of another shift crew to cover for an individuals absence). The continued loss of operators reduced shift staffing to a point where 13 of 15 shifts were not self-relieving. This meant that most control room shifts did not have a sufficient number of operators to make up for a temporary absence or permanent loss of either a reactor operator (RO) or SRO. The reductions had the effect of requiring personnel to work additional overtime and limited most licensed operators activities to standing watch in the control room. Interviewees indicated that career advancement opportunities for licensed operators were limited because of pressures to maintain shift crews; thereby, limiting the ability of licensed operators to integrate an operations perspective into other site activities. The team reviewed operations payroll data that summarized the cumulative regular and overtime hours for each operations department position and calculated the annual overtime rate for select positions. Since 2003, overtime, as a percent of regular hours worked, has increased steadily and substantively for control room and auxiliary operators. The team noted that the increase in overtime rates for operations department positions appeared to be largely the result of a decrease in staffing, rather than the result of an increase in the total number of person-hours expended. Specifically, from 2003 through 2006, the total number of hours worked annually by personnel in the control room supervisor (CRS), SRO, RO, and auxiliary operator (AO) positions remained relatively constant, or decreased, while the percentage of those total hours that were worked as overtime increased. As a result, the payroll data indicated that the licensee increasingly relied on the use of overtime to provide the person-hours necessary to operate the three units. Technical Specification 5.2.2.d requires administrative procedures to be developed and implemented to limit the working hours of unit staff that perform safety-related functions (e.g., licensed SROs, licensed ROs, radiation protection technicians, auxiliary operators and key maintenance personnel). The Technical Specifications further requires that the controls shall include guidelines on working hours that ensure adequate shift coverage shall be maintained without routine heavy use of overtime. Pending the completion of a review of the actual work hours by operations personnel, this issue is identified as URI 05000528, 05000529, 05000530/2007012-19, Routine Heavy Use of Overtime. |
| 05000528/FIN-2008002-01 | Failure to Establish Preventative Maintenance Procedures for Emergency Diesel Generator Fuel Oil Injection Pump O-rings (Section 1R15) | 2008Q1 | The inspectors identified a non-cited violation of Technical Specification 5.4.1.a for the failure of operations and engineering personnel to establish and implement maintenance procedures for inspection and replacement of items that have a specific lifetime. Specifically, between February 12, 2007 and March 7, 2008, operations and engineering personnel failed to inspect or replace the emergency diesel generators fuel oil injection pump upper O-rings prior to the end of their service life resulting in fuel leakage and increased unavailability and unreliability of Unit 1 Train A, Unit 2 Train B, and Unit 3 Train B emergency diesel generators. This issue was entered into the licensee\\\'s corrective action program as Palo Verde Action Request 3143422. This finding is greater than minor because it is associated with the equipment performance attribute of the mitigating systems cornerstone and affects the cornerstone objective of ensuring the availability and reliability 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 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 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 operating experience because the licensee failed to use available operating experience, including vendor recommendations, to implement and institutionalize operating experience through changes to station processes, procedures, equipment, and training programs (P.2(b)). (Section 1R15) |
| 05000528/FIN-2008002-02 | Two Examples of a Failure to Properly Implement the Systematic Troubleshooting Process (Section 1R19) | 2008Q1 | The inspectors identified two examples of a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, \\\"Instructions, Procedures and Drawings,\\\" for the failure of operations, engineering, and maintenance personnel to follow procedures for troubleshooting failures of safety-related components. Specifically, between January 8 and January 13, 2008, operations, engineering, and maintenance personnel failed to incorporate the adequate level of detail into their troubleshooting plans for the Unit 3 auxiliary feedwater trip and throttle Valve AFA-HV-0054 when it failed to fully close upon demand from the control room hand switch, and for the Unit 3 log power Channel A when induced noise was present. These issues were entered into the licensee\\\'s corrective action program as Palo Verde Action Requests 3120075 and 3118744. This finding is greater than minor because it 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 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 screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. Both examples have a crosscutting aspect in the area of human performance associated with decision-making because the licensee did not obtain appropriate interdisciplinary input and reviews on safety-significant or risk-significant decisions (H.1(a)). (Section 1R1 |
| 05000528/FIN-2008002-03 | Inadequate Procedure to Evaluate Foreign Material in the Spent Fuel Pool (Section 4OA2) | 2008Q1 | The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, \\\"Instructions, Procedures, and Drawings,\\\" for the failure of fuels services personnel to evaluate leaving foreign material in the Unit 2 spent fuel pool in accordance with procedures, and failed to ensure those procedures included appropriate quantitative and qualitative acceptance criteria. Specifically, between October 13, 2006, and January 31, 2008, fuels services personnel used Procedure 30DP-9MP03, \\\"System Cleanliness and Foreign Material Exclusion Controls,\\\" Revision 6, which did not specify acceptance criteria for time to perform a functional assessment of foreign material in the spent fuel pool, resulting in foreign material being left in the spent fuel pool for greater than one year without an evaluation on affected safety systems. This issue was entered into the licensee\\\'s corrective action program as Palo Verde Action Request 3126308. This finding is greater than minor because it is associated with the structure, systems, and component performance and human performance attributes of the barrier integrity cornerstone and affects 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. 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 result in loss of cooling to the spent fuel pool; the finding did not result from fuel handling errors that caused damage to the fuel clad integrity or a dropped assembly; and the finding did not result in a loss of spent fuel pool inventory greater than ten percent of the spent fuel pool volume. This finding has a crosscutting aspect in the area of human performance associated with decision-making because the licensee failed to use conservative assumptions when evaluating degraded and nonconforming conditions (H.1.(b)). (Section 4OA2 |
| 05000528/FIN-2008002-04 | Failure to Maintain Adequate Staffing Levels Results in Heavy Use of Overtime to Maintain Adequate Shift Coverage (Section 4OA2) | 2008Q1 | The inspectors identified a non-cited violation of Technical Specification 5.2.2.d involving the routine use of excessive overtime for operations personnel that performed safety-related functions. Specifically, between January 1 and December 31, 2007, operations personnel routinely used excessive overtime. This issue was entered into the licensees corrective action program as Condition Report/Disposition Request 3112231. The finding is greater than minor because if left uncorrected the finding would become a more significant safety concern in that the routine use of excessive work hours increases the likelihood of operator errors. Using the IMC 0609, \\\"Significance Determination Process,\\\" Phase 1 Worksheets, the finding is determined to have very low safety significance because no specific human performance issues due to personnel fatigue were identified that resulted in the degradation or loss of safety function of equipment important to safety. The finding has a crosscutting aspect in the area of human performance associated with resources because the licensee failed to maintain sufficient qualified operations personnel to maintain working hours within guidelines without the excessive use of overtime (H.2(b)). (Section 4OA2 |
| 05000528/FIN-2008002-05 | Failure to Properly Implement Corrective Action Process for Potential Operability Issues with the Class 1E 125 V DC System (Section 4OA2) | 2008Q1 | The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, \\\"Instructions, Procedures, and Drawings,\\\" for the failure of engineering personnel to ensure that potentially nonconforming conditions associated with the Class 1E 125 Vdc system were reviewed for operability. Specifically, between September 29, 2007 and March 7, 2008, engineering personnel failed to ensure all relevant information was reviewed for operability when it was determined that vendor recommended preventative maintenance tasks were not being performed on the Class 1E 125 Vdc system. This issue was entered into the licensee\\\'s corrective action program as Palo Verde Action Request 3144707. This finding is greater than minor because it is associated with the equipment performance attribute of the mitigating systems cornerstone and affects the cornerstone objective to ensure the availability and reliability 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 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 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 human performance associated with decision-making because safety-significant decisions were not verified to validate underlying assumptions and identify unintended consequences (H.1(b)). (Section 4OA2 |
| 05000528/FIN-2008002-06 | Failure to Follow Procedures Resulted in Water Transfer from the Spent Fuel Pool (Section 4OA3) | 2008Q1 | A self-revealing non-cited violation of Technical Specification 5.4.1.a was identified for the failure of operations personnel to follow procedures. Specifically, on January 13, 2008, operations personnel failed to properly implement Procedure 40OP-9PC06, \\\"Fuel Pool Cleanup and Transfer,\\\" Revision 41, for operating the pool cooling cleanup system, resulting in pool cooling cleanup Filter PCN-F01B bypass Valve PCN-V061 being improperly aligned. This resulted in the inadvertent transfer of 300 gallons of spent fuel pool water to the refueling water tank. This issue was entered into the licensee\\\'s corrective action program as Condition Report/Disposition Request 3121713. The finding is greater than minor because it is associated with the configuration control and human performance attributes of the barrier integrity cornerstone and affects 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. 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 result in loss of cooling to the spent fuel pool; the finding did not result from fuel handling errors that caused damage to the fuel clad integrity or a dropped assembly; and the finding did not result in a loss of spent fuel pool inventory greater than ten percent of the spent fuel pool volume. This finding has a crosscutting aspect in the area of human performance associated with work practices because the licensee failed to use adequate human error prevention techniques, such as pre-job briefings, to ensure that the pool cooling cleanup system activity was performed safely (H.4(a)). (Section 4OA3 |
| 05000528/FIN-2008002-07 | Failure to Identify Inoperable Feedwater Isolation Valve Exceeds Technical Specification Allowed Outage Time (Section 4OA3) | 2008Q1 | A self-revealing non-cited violation of Technical Specification 3.7.3.c was identified for the failure of operations personnel to perform the actions required for an inoperable main feedwater isolation valve. Specifically, on July 17, 2006, operations personnel failed to perform actions to place the unit in Mode 3 within 6 hours and Mode 5 within 36 hours, as required by Technical Specification 3.7.3.c, for an inoperable main feedwater isolation valve that had not been closed or isolated in 72 hours, as required by Technical Specification 3.7.3.a. This resulted in main feedwater isolation Valve 2JSGAUV0174 to steam Generator A exceeding the Technical Specification 3.7.3 allowed outage time. This issue was entered into the licensee\\\'s corrective action program as Condition Report/Disposition Request 2915450. This finding is greater than minor because it is associated with the equipment performance attribute of the mitigating systems cornerstone and affects the cornerstone objective to ensure the availability and reliability of systems that respond to initiating events to prevent undesirable consequences. A Phase 2 analysis was required because the Manual Chapter 0609, \\\"Significance Determination Process,\\\" Phase 1 Worksheets, determined that there was a loss of main feedwater isolation of a single train to steam Generator A for greater than the technical specification allowed outage time. Using the Phase 2 Worksheets associated with a steam generator tube rupture without steam generator isolation, the finding is determined to have very low safety significance since all remaining mitigation capability was available or recoverable. (Section 4OA3 |
| 05000528/FIN-2008002-08 | Licensee-Identified Violation | 2008Q1 | 10 CFR 50.54(q) of Title 10 of the Code of Federal Regulations requires the licensee to follow their emergency plan. Contrary to the above, between 2002 and 2007, training personnel did not administer annual emergency preparedness training to all employee site badge holders, as required by Section 8.1.1 of the Emergency Plan. The finding was entered into the CAP as CRDR 2966025. The finding is of very low safety significance because it is associated with Planning Standards 50.47(b)(7) and 50.47(b)(15), is not a functional failure of the planning standards because all employees received initial general emergency preparedness training, and means existed to inform holders of site badges about the actions they should take during an emergency. |
| 05000528/FIN-2008002-09 | Licensee-Identified Violation | 2008Q1 | Technical Specification Surveillance Requirement 3.3.11.2 requires that each remote shutdown system disconnect switch and control circuit is verified capable of performing the intended function. Contrary to the above, between January 20, 2008 and March 15, 2008, Procedure 40ST-9ZZ20, \\\"Remote Shutdown Disconnect Switch and Control Circuit Operability,\\\" Revision 10, did not verify all circuit paths associated with each disconnect switch were adequately tested. This issue affected all the disconnect switches to the remote shutdown panel. The licensee entered into TS Surveillance Requirement 3.0.3 for a missed surveillance, performed a risk evaluation, and tested the most risk-significant disconnect switches to verify that these disconnect switches could perform their intended function. Of the risk-significant disconnect switches tested, the licensee identified that one disconnect switch associated with Unit 1 AFW pump to SG 1 block Valve AFB-UV-34 would not have been capable of performing it\\\'s intended function due to an electrical jumper installed in the closing circuit. This valve is in the flow path from the motor driven AFW pump to SG 1. However, the potential failure of this valve would not have affected the ability to maintain a shutdown condition, because the flowpath to the SG 2 was not affected. The finding was entered into the CAP as PVARs 3129077, 3135575, 3136664, 3138937 and 3144595. Using Manual Chapter 0609, \\\"Significance Determination Process,\\\" Appendix F, \\\"Fire Protection Significance Determination Process,\\\" the finding is determined to have very low safety significance because at Step 1.3, Qualitative Screening Approach, the finding only affected the ability to reach and maintain a cold shutdown condition. |
| 05000528/FIN-2008003-01 | Inadequate Work Instructions for Reinstallation of Constant Support Hanger | 2008Q2 | The inspectors identified a noncited violation of Technical Specification 5.4.1.a, \"Procedures,\" for the failure to establish and implement adequate maintenance procedures. These inadequate instructions resulted in the failure to install required washers during installation of a constant support spring hanger for a main steam line on May 14, 2008. This issue was entered into the licensee corrective action program as Condition Report/Disposition Request 3177622. The finding is greater than minor because it is associated with the procedure quality attribute of the mitigating systems cornerstone and affects the cornerstone objective of ensuring the 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 a very low safety significance because the finding did not result in 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 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 human performance associated with resources because the licensee failed to ensure work packages were complete, accurate and included up-to-date design documentation to assure nuclear safety (H.2(c)) (Section 1R12.1) |
| 05000528/FIN-2008003-02 | Failure to Resolve Discrepancies Between Installed Equipment and Work Instructions Results in Mispositioning Event | 2008Q2 | A self-revealing finding was identified for the failure of operations and maintenance personnel to follow Procedure 01DP-9ZZ01, \"Systematic Troubleshooting,\" and resolve a discrepancy with a work instruction prior to proceeding with troubleshooting. Specifically, maintenance and operations personnel did not resolve an error in Work Order 3174332 when troubleshooting Breaker NBN-S01A that failed to trip, resulting in a loss of the non-vital electrical bus that supplied power to the nuclear cooling water and normal chilled water systems. This issue was entered into the licensee\'s corrective action program as Palo Verde Action Request 3174647. The finding is greater than minor because it is associated with the initiating events cornerstone attribute of configuration control and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown and power operations. Using the Manual Chapter 0609 Appendix G, \"Shutdown Operations Significance Determination Process,\" the finding is determined to have very low safety significance because the finding did not result in a loss of shutdown safety functions. This finding has a crosscutting aspect in the area of human performance associated with work practices because maintenance and operations personnel proceeded in the face of uncertainty or unexpected circumstances (H.4(a)) (Section 1R12.2) |
| 05000528/FIN-2008003-03 | Inadvertent Decrease in Reactor Water Level Due to Personnel Error | 2008Q2 | A self-revealing noncited violation of Technical Specification 5.4.1, \"Procedures,\" was identified for the failure of operations personnel to adequately implement Procedure 40DP-9OP19, \"Locked Valve, Breaker, and Component Tracking.\" Specifically, on May 14, 2008, Valve SIA-V421 was found out of its locked closed position one and one-half turns open resulting in approximately 930 gallons of water being inadvertently transferred from the reactor coolant system to the refueling storage water tank. This issue has been entered into the licensee\'s corrective action program as Palo Verde Action Request 3174527. The failure to ensure the valve was properly closed resulted in an inadvertent reactor vessel level decrease. The finding is more than minor because it is associated with the configuration control attribute of the initiating events cornerstone and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown operations. A Phase 2 analysis was required because using Manual Chapter 0609, Appendix G, \"Shutdown Operations Significance Determination Process,\" Attachment 1, the inspectors determined that the finding actually resulted in a loss of reactor coolant system inventory. Using the Phase 2 worksheets in Attachment 2, this was determined to be a loss of level control precursor event. The initiating event likelihood for this finding was determined from Table 1 of the worksheet and the resultant core damage frequency was determined to be 1E-8, therefore the finding screened as having very low safety significance. The finding has a crosscutting aspect in the area of human performance associated with work practices because the licensee failed to use human error prevention techniques such as self-checking (H.4(a)) (Section 1R22) |
| 05000528/FIN-2008003-04 | Failure to Prevent Recurrence of a Significant Condition Adverse to Quality for the Feedwater Isolation Valves | 2008Q2 | A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, \"Corrective Actions,\" was identified for the failure of engineering personnel to implement adequate corrective actions to preclude recurrence of a significant condition adverse to quality. Specifically, between June 28, 1998 and July 17, 2006, on several occasions, the four-way \'N\' valve for an economizer main feedwater isolation valve became lodged in the center blocked position, preventing fast closure of the main feedwater isolation valve upon receipt of a main steam isolation signal. This issue was entered into the licensee\'s corrective action program as Condition Report/Disposition Request 2915450. This finding is greater than minor because it is associated with the equipment performance attribute of the mitigating systems cornerstone and affects the cornerstone objective of ensuring the availability and reliability of systems that respond to initiating events to prevent undesirable consequences. A Phase 2 analysis was required because using Manual Chapter 0609.04, \"Phase 1 Initial Screening and Characterization of Findings,\" there was a loss of main feedwater isolation of a single train to Steam Generator 1 for greater than the Technical Specification allowed outage time. Using the Phase 2 worksheets associated with a steam generator tube rupture without steam generator isolation, the finding is determined to have very low safety significance since all remaining mitigation capability was available or recoverable. This finding was evaluated as not having a crosscutting aspect because the performance deficiency is not indicative of current performance (Section 4OA2) |
| 05000528/FIN-2008003-05 | Fire in Pressurizer Cubicle Due to Poor Work Practices | 2008Q2 | A self-revealing noncited violation of License NPF-51, Condition 2.C. (6), was identified involving the failure to follow procedures for proper control of ignition sources. Specifically, contract welding personnel failed to deenergize welding equipment and properly secure the welding rod electrodes, resulting in a fire in the Unit 2 pressurizer cubicle inside containment. This issue was entered into the licensee\'s corrective action program as Condition Report/Disposition Request 3170965. The finding is greater than minor because it is associated with the external factors attributes of the initiating events cornerstone and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Manual Chapter 0609, \"Significance Determination Process,\" Appendix M, \"Significance Determination Process Using Qualitative Criteria,\" was used since the Manual Chapter 0609, Appendix F, \"Fire Protection Significance Determination Process,\" does not address the potential risk significance of fire protection findings during shutdown conditions. The finding was determined to be of very low safety significance by NRC management review because the finding occurred while the unit was already in a cold shutdown condition and the finding did not affect equipment necessary to maintain safe shutdown. This finding has a crosscutting aspect in the area of human performance associated with work practices because the licensee did not ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety is supported (H.4(c)) (Section 4OA3) |
| 05000528/FIN-2008003-06 | Failure to Adequately Implement Procedural Requirements for Open Doors, Hatches, and Floor Plugs | 2008Q2 | A self-revealing noncited violation of Technical Specification 5.4.1.a, \"Procedures,\" was identified for the failure of maintenance personnel to adequately implement procedural guidance. Specifically, on May 9, 2008, maintenance personnel failed to ensure the permit requirements of Procedure 0DP-9ZZ17, \"Control of Doors, Hatches, and Floor Plugs,\" were complete while accessing the tendon gallery access shaft, resulting in the control room determining that both trains of the pump room exhaust air cleanup system had been inoperable. This issue was entered into the licensee\'s corrective action program as Palo Verde Action Request 3172712 and as significant Condition Report/Disposition Request 3173930. The finding is greater than minor because it is associated with the barrier performance attribute associated with maintaining radiological barrier functionality for the auxiliary building and affects the cornerstone objective to provide reasonable assurance that the physical design barriers protect the public from radio nuclide releases caused by accidents or events. Using Manual Chapter 0609.04, \"Phase 1 Initial Screening and Characterization of Findings,\" the finding is determined to have very low safety significance because it only affected the barrier integrity cornerstone and only represented a degradation of the radiological barrier function of the auxiliary building. This finding has a crosscutting aspect in the area of human performance associated with work practices because the licensee did not ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety is supported (H.4(c)) (Section 40A3) |
| 05000528/FIN-2008003-07 | Failure to Take Timely Corrective Actions for a Condition Adverse to Quality Resulting in SIT 1A Being Declared Inoperable | 2008Q2 | The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, \"Corrective Action,\" for the failure of operations and maintenance personnel to promptly identify and correct a condition adverse to quality. Specifically, from August 2007 till June 2008, operations and maintenance personnel failed to ensure that work management process procedures were followed for a degraded condition affecting Safety Injection Tank 1A. This issue was entered into the licensee\'s corrective action program as Condition Report/Disposition Request 3185716. The finding is greater than minor because it is associated with the equipment performance attribute of the mitigating systems cornerstone and affects the cornerstone objective of ensuring the reliability, availability 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 a very low safety significance because the finding did not result in 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 screen as potentially risksignificant due to a seismic, flooding, or severe weather initiating event. This finding has a crosscutting aspect in the area of human performance associated with work control because the licensee failed to plan work activities to support long-term equipment reliability by limiting operator work-arounds, safety systems unavailability, and reliance on manual actions (H.3 (b)) (Section 40A3) |
| 05000528/FIN-2008003-08 | Failure to Evaluate Design Change Leads to Manual Reactor Trip | 2008Q2 | A self-revealing finding of Procedure 81DP-0DC13, \"Deficiency Work Order,\" Revision 13, was identified for the failure of engineering personnel to ensure modifications do not inadvertently affect design basis plant conditions. Specifically, between January 23, 2001 and October 6, 2007, engineering personnel failed to ensure material compatibility of the condenser air removal system seal water cooler tube plugs to prevent corrosion. This resulted in sodium ingress into the condenser hotwell and steam generators due to a corroded tube plug that failed in the condenser air removal system D seal water cooler, and consequently a manual reactor scram. This issue was entered into the licensee\'s corrective action program as Condition Report/Disposition Request 3074272. The finding is greater than minor because it is associated with the design control attribute of the initiating events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown and power operations. Using Manual Chapter 0609.04, \"Phase 1 Initial Screening and Characterization of Findings,\" the finding is determined to have very low safety significance because the finding did not result in exceeding the technical specification limit for identified reactor coolant system leakage, did not affect other mitigation systems, did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available; and did not increase the likelihood of a fire or internal/external flood. This finding was evaluated as not having a crosscutting aspect because the performance deficiency is not indicative of current performance (Section 4OA3) |
| 05000528/FIN-2008003-09 | Failure to Correct a Risk Significant Planning Standard | 2008Q2 | The inspectors identified a noncited violation (NCV) of 10 CFR 50.54(q) and 10 CFR Part 50, Appendix E.IV.F.2.g, for the licensees failure to correct an identified risk significant planning standard weakness between May 2, 2007 and October 28, 2007. Specifically, the licensee failed to implement adequate corrective actions for identified weaknesses in the ability to correctly make a Site Area Emergency declaration for a steam generator tube rupture event. This issue was entered into the licensees correction action program as Palo Verde Action Request 3083911. The NRC determined that the inability to consistently implement an Emergency Action Level was a performance deficiency within the licensees control. This finding is more than minor because it was associated with the Emergency Preparedness attribute of emergency response organization performance and affected the cornerstone objective to implement adequate measures to protect the health and safety of the public because the inability to properly recognize and classify an emergency condition affects the licensees ability to implement adequate protective measures. This finding was preliminarily determined to be of low to moderate safety significance. After consideration of information provided during and after a Regulatory Conference held on March 25, 2008, the NRC has concluded that the knowledge deficiency identified among senior operators would not likely result in an incorrect emergency classification during a steam generator tube rupture event, and the NRC has concluded the significance of the inspection finding is appropriately characterized as Green (i.e., a finding of very low safety significance). This violation is being treated as an NCV, consistent with Section VI of the NRC Enforcement Policy. The cause of this finding has crosscutting aspects associated with the corrective action aspect of the problem identification and resolution area in that the licensee failed to thoroughly evaluate problems such that resolutions ensured correcting problems (P.1.(c)). The cause of this finding was also related to the safety culture component of accountability in that the licensee failed to demonstrate a proper safety focus and reinforce safety principles (O.1.(c)) (Section 4OA5) |
| 05000528/FIN-2008003-10 | Licensee-Identified Violation | 2008Q2 | Title 10 CFR Part 50, Appendix B, Criterion III, states, \"measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in 10 CFR Part 50.2 and as specified in the license application, for those SSCs to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions.\" The licensee identified that the specification for installation of main steam line constant support hangers was not followed and the allowable deviation from vertical was exceeded when the main steam line was at normal operating temperature and pressure. This event has been documented in the licensee\'s CAP as CRDR 3153607. The finding is of very low safety significance because it did not result in a loss of main steam line operability as defined in NRC Inspection Manual, Part 9900, Technical Guidance, \"Operability Determination Process for Operability and Functional Assessment.\ |
| 05000528/FIN-2008003-11 | Licensee-Identified Violation | 2008Q2 | Title 10 CFR Part 50, Appendix B, Criterion V, \"Instructions, Procedures and Drawings,\" requires that activities affecting quality shall be prescribed by instructions, procedures, or drawings, and shall be accomplished in accordance with those instructions, procedures, or drawings. The licensee identified that operations personnel did not follow procedures to promptly evaluate a degraded condition identified for the SIT 1A nitrogen leak. This issue has been entered into the licensee\'s CAP as PVAR 3185480, CRDR 3186791 and significant CRDR 3185716. The finding is of very low safety significance because it did not result in a loss of system safety function, an actual loss of safety function of a single train for greater than its TS allowed outage time, or screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event |
| 05000528/FIN-2008004-01 | Inadequate Procedural Requirements to Implement Technical specification 5.5.2.b | 2008Q3 | The inspectors identified a noncited violation of Technical Specification 5.5.2.b, \"Primary Coolant Sources Outside Containment,\" for the failure of engineering and maintenance personnel to implement a program to verify integrated leak test requirements for abandoned valves still connected to an active system. Specifically, between January 8, 1993, and September 30, 2008, engineering personnel failed to ensure portions of the containment spray system, which could be in contact with radioactive fluids outside containment, were included in the integrated leak test requirements. This issue was entered into the licensee\'s corrective action program as Condition Report/Disposition Request 3170965. The performance deficiency associated with this finding was the failure of engineering and maintenance personnel to implement a program to verify integrated leak test requirements for abandoned valves still connected to an active system. The finding is greater than minor because it is associated with the design control and procedural quality attribute associated with maintaining radiological barrier functionality for the auxiliary building of the barrier integrity cornerstone and affects the cornerstone objective to provide reasonable assurance that the physical design barriers protect the public from radio nuclide releases caused by accidents or events. Using Manual Chapter 0609.04, \"Phase 1 Initial Screening and Characterization of Findings,\" the finding is determined to have very low safety significance because it only represented a degradation of the radiological barrier function of the auxiliary building. This finding was evaluated as not having a crosscutting aspect because the performance deficiency is not indicative of current performance (Section 1R04). |
| 05000528/FIN-2008004-02 | Potentially Degraded fire Penetration Seals | 2008Q3 | The inspectors identified an unresolved item (URI) associated with License NPF-51, Condition 2.C. (7) for the failure of engineering and maintenance personnel to periodically inspect or test fire penetration seals at regular intervals. The licensee performed inspections of all penetrations and barriers in response to Information Notice 2007-01, \"Recent Operating Experience Concerning Hydrostatic Barriers,\" to verify adequate internal flood protection for safety-related equipment. The inspectors reviewed Information Notice 2007-01 in February 2007, and determined that engineering personnel credited fire protection inspection Procedure 14FT-9FP70, \"Appendix R & Former Technical Specification Penetration Seal Surveillance,\" Revision 7, for flood inspections even though the acceptance criteria for fire protection allows small gaps in seals while flood protection does not. During the penetration and barrier inspections, the licensee found 19 seals that were nonfunctional for both flood and fire functions. The inspectors reviewed Technical Requirements Manual, Section 3.11.107, which stated, in part, required fire-related assemblies and penetration seals will be inspected every 15 years. Procedure 14FT-9FP70 stated the Appendix R and Former Technical Specification penetration seals shall be sampled such that all of the seals are tested every 15 years. Inspectors questioned why the licensee was finding nonfunctional fire seals when they currently had a program and procedure in place to periodically inspect the fire seals. The licensee could not verify that these seals had been inspected within the past 15 years and wrote Palo Verde Action Request (PVAR) 3221773 to determine the date they were last inspected. The licensee has inspected 96 percent of seals in Units 1 and 2, to ensure the fire and flood protection functions can be met. The licensee plans to inspect the additional seals in Units 1 and 2, and all of the seals in Unit 3 by the end of October 2008. This URI is being opened to determine the extent of nonfunctional fire penetration seals at Palo Verde Nuclear Generating Station and the potential aggregate effect of more than one seal penetration failure: URI 05000528;529;530/2008004-02, Potentially Degraded Fire Penetration Seals. |
| 05000528/FIN-2008004-03 | Potentially Degraded Flood Penetration Seals | 2008Q3 | The inspectors identified a URI associated with 10 CFR Part 50, Criterion XVI, \"Corrective Action,\" for failure of operations and engineering personnel to promptly identify and correct a condition adverse to quality involving internal flood protection for safety-related equipment. The inspectors reviewed Information Notice 2007-01, \"Recent Operating Experience Concerning Hydrostatic Barriers,\" and its applicability to Palo Verde. The review was performed to verify that recent operating experience involving degraded foam penetration seals was evaluated to ensure adequate internal flood protection for safety-related equipment. The licensee\'s review of Information Notice 2007-01, in February 2007, determined that the penetration seal material discussed in Information Notice 2007-01 was similar to the material used at Palo Verde, that the seals are not periodically inspected for degradation, and that Palo Verde has previously had unsealed conduits allow flooding into safety-related areas. Inspectors determined that the licensee credited their fire protection inspection Procedure 14FT-9FP70, \"Appendix R and Former Technical Specification Penetration Seal Surveillance,\" Revision 7, for flood inspections even though the acceptance criteria for fire protection allowed small gaps in seals while flood protection did not. In December 2007, when the inspectors reviewed Information Notice 2007-01, the inspectors questioned what actions had been taken to ensure there were no degraded or nonconforming hydrostatic seals in the plant. The inspectors determined that maintenance and engineering personnel had not taken corrective action to inspect the penetration seals and to monitor the seals in accordance with the maintenance rule program, even after reviewing Information Notice 2007-01 in February 2007. After reviewing the fire inspection Procedure 14FT-9FP70, flood design calculations, and the maintenance rule scoping document Procedure 81DP-0ZZ01, \"Civil System, Structure, and Component Monitoring Program,\" Revision 13, the inspectors determined that the licensee was not verifying the adequacy of design of their hydrostatic seals in accordance with these documents. After discussion with the inspectors, the licensee developed acceptance criteria to inspect the barriers, and began inspecting all of the approximately 1,500 seals located in Units 1, 2, and 3 in July 2008. The licensee has inspected 96 percent of seals in Units 1 and 2, and has declared 19 hydrostatic seals nonfunctional. The licensee plans to inspect the additional seals in Units 1 and 2, and all of the seals in Unit 3 by the end of October 2008. This unresolved issue is being opened to determine the extent of degraded flood penetration seals at Palo Verde and the potential aggregate effect of more than one seal penetration failure: URI 05000528;529;530/2008004-03, Potentially Degraded Flood Penetration Seals. |
| 05000528/FIN-2008004-04 | Failure to Provide an Adequate procedure to Control Essential Spray Pond Missile Hazards | 2008Q3 | 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 to ensure evaluation and approval of transient missile structure hazards that have an effect on the operability of the essential spray ponds. Specifically, since January 15, 1997, civil engineering personnel failed to develop an adequate procedure to verify missile density criteria are not exceeded to ensure operability of the essential spray ponds during severe weather. This resulted in approximately 40 transient missile hazards being placed around Unit 1 spray Pond A without an approval or evaluation to ensure continued operability of the essential spray ponds. The licensee determined the spray pond was operable following a walkdown and evaluation of the missile hazards. This issue was entered into the licensee\'s corrective action program as Condition Report/Disposition Request 3224028. The finding is greater than minor because it is associated with the external factors attribute of the mitigating systems cornerstone and affects the cornerstone objective of ensuring the reliability 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 a very low safety significance because the finding did not result in 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 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 because appropriate corrective actions were not taken to address safety issues and adverse trends in a timely manner, commensurate with their safety significance and complexity (P.1(d)) (Section 1R13). |
| 05000528/FIN-2008004-05 | Failure to Perform an Operability Determination for High Chlorine in the Essential Spray Pond | 2008Q3 | The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, \"Instructions, Procedures, and Drawings,\" for the failure of operations and chemistry personnel to follow the corrective action program to ensure that potentially nonconforming conditions associated with the essential spray pond system were reviewed for operability. Specifically, between July 10, 2008, and July 11, 2008, operations and chemistry personnel failed to ensure all relevant information was reviewed for operability when the Unit 2 essential spray Pond A hypochlorite addition Valve 2-SPN-V494 was found open. This resulted in the essential spray pond chemistry pH and chlorine samples being delayed to the extent that the sample results were not reliable to assess operability. This issue was entered into the licensee\'s corrective action program as Condition Report/Disposition Request 3206115. The finding is greater than minor because it is associated with the equipment performance attribute of the mitigating systems cornerstone and affects the cornerstone objective of ensuring the reliability 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 a very low safety significance because the finding did not result in 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 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 human performance associated with decision-making because safety-significant decisions were not verified to validate underlying assumptions and identify unintended consequences (H.1(b)) (Section 1R15). |
| 05000528/FIN-2008004-06 | Failure to Correct a Condition Adverse to Quality with the Refueling Water Tank Instruments in a Timely Manner | 2008Q3 | The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, \"Corrective Action,\" for the failure of the licensee to correct a deficiency associated with the refueling water tank instrument pit in a timely manner. Specifically, between June 16, 2006, and July 2, 2008, maintenance and engineering personnel failed to ensure the openings of the pit covers were adequately sealed to prevent rain water intrusion. This issue was entered into the licensees corrective action program as Palo Verde Action Request 3194904. The performance deficiency associated with this finding involved the failure of maintenance personnel to correct a condition adverse to quality in a timely manner. The finding is greater than minor because it is associated with the protection against external factors cornerstone attribute of the mitigating systems cornerstone and affects the associated cornerstone objective to ensure the reliability and availability of systems that respond to initiating events. Using Manual Chapter 0609.04, \"Phase 1 Initial Screening and Characterization of Findings,\" the finding required a Phase 3 analysis by a senior reactor analyst, since the finding is potentially risk significant due to external initiating event core damage sequences. Based on the analysis performed, the analyst concluded that the finding had very low safety significance (Green) because of the very small probability of a large rainfall event and a loss of coolant accident occurring at the same time. This finding was evaluated as not having a crosscutting aspect because the performance deficiency is not indicative of current performance (Section 1R15). |
| 05000528/FIN-2008004-07 | Licensee-Identified Violation | 2008Q3 | Technical Specification 3.3.12 requires that two channels of boron dilution alarm system shall be operable. Technical Specification 3.3.12 required Actions A.1, B.1, and C.1 are to be performed upon meeting the requirements of the limiting condition for operation. Contrary to the above, between May 8 and May 21, 2008, on multiple occasions, operations personnel did not follow Procedure 40AL-9RK3A, \"Panel B03A Alarm Response Procedure,\" Revision 20, when they failed to reset the boron dilution alarm system alarms and rendered both boron dilution alarm system channels inoperable. The boron dilution alarm system alarms were allowed to remain in fast flash for extended periods and the Technical Specification required actions were not completed. The issue has been entered into the licensees corrective action program as significant CRDR 3178553. The finding is of very low safety significance because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available. |
| 05000528/FIN-2008004-08 | Licensee-Identified Violation | 2008Q3 | Technical Specification 3.7.2 requires that four main steam isolation valves and their associated actuator trains shall be operable. Technical Specification 3.7.2, Limiting Condition for Operation, Condition A, requires that with one main steam isolation valve with a single actuator train inoperable, that the main steam isolation valve actuator train be restored to operable status within 7 days. Contrary to the above, between July 31, and October 27, 2007, main steam isolation Valve 181 actuator Train A was found to be inoperable during surveillance test Procedure 73ST-9SG01, \"MSIVs Inservice Testing,\" Revision 32. The licensee identified air leakage between the four-way valve and the main steam isolation valve actuator due to maintenance personnel failing to install an air port O-ring. This issue has been entered into the licensee\'s corrective action program as PVAR 3083549, and significant CRDR 3087163. The finding is of very low safety significance because it did not result in 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 since actuator Train B was available, or screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. |
| 05000528/FIN-2008005-01 | Failure to Promptly Identify and Correct Degraded Hydrostatic Penetration Seals | 2008Q4 | The inspectors identified a finding at Palo Verde Nuclear Generating Station Procedure 01DP-0AP10, Corrective Action Program, Revision 1, for the failure of operations and engineering personnel to promptly identify and correct a condition adverse to quality. Specifically, between February 13, 2007 and July 18, 2008, operations and engineering personnel failed to identify and correct degraded hydrostatic flood penetration seals which provide protection to safety-related equipment during internal flooding events. This resulted in over100 hydrostatic penetration seals in the control, diesel, and main steam support structure buildings being left degraded for greater than 12 months. This issue was entered into the licensee\'s corrective action program as Palo Verde Action Request 3264501. The finding is greater than minor because it is associated with the protection against external factors (i.e. flood hazard) attribute of the mitigating systems cornerstone and affects the cornerstone objective of ensuring the 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 a very low safety significance because the finding did not result in 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 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 operating experience because operations and engineering personnel failed to implement and institutionalize operating experience through changes to station processes, procedures, equipment, and training programs P.2(b) (Section 1R06) |
| 05000528/FIN-2008005-02 | Failure to Adequately Implement Procedure Requirements for Refueling Machine Operation | 2008Q4 | A self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, was identified for the failure of refueling services personnel to follow procedures to address refueling machine fault indications. Specifically, during the Unit 1 refueling outage core offload, refueling services personnel had overridden interlocks that protect the fuel from damage. This issue has been entered into the licensees corrective action program as Palo Verde Action Request 3235153 and Condition Report Disposition Request 3237465.The finding is greater than minor because it is associated with the human performance attribute of the barrier integrity cornerstone and affects 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. Manual Chapter 0609, Appendix M, Significance Determination Process Using Qualitative Criteria, was used since the Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, does not address the potential risk significance of refueling machine operation errors. The finding is determined to have very low safety significance because there was no apparent damage done to the fuel barrier and no radioactive release occurred. This finding has a crosscutting aspect in the area of human performance associated with decision making because refueling services personnel did not use a systematic process to make a risk significant decision when faced with uncertain or unexpected plant conditions H.1(a) (Section 1R20) |
| 05000528/FIN-2009002-01 | Failure to Correct Deficient Condition for the Essential Spray Pond Chemical Addition System Valves High Failure Rate | 2009Q1 | The inspectors identified a finding for the failure of engineering and maintenance personnel to adequately implement timely corrective actions for deficiencies associated with the essential spray pond sodium hypochlorite chemical addition system. Specifically, between May 2006 and March 2009, corrective actions to replace degraded sodium hypochlorite valves with a more reliable chemical addition system were not taken resulting in the Unit 2 spray pond Train A chemistry pH level being out of specification high on two occasions. This issue was entered into the licensees corrective action program as Palo Verde Action Request 3277070.The finding is more than minor because it is associated with the equipment performance attribute of the mitigating systems cornerstone and affects the cornerstone objective of ensuring the reliability 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 a very low safety significance because the finding did not result in 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 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 human performance associated with decision making because the licensee did not communicate bases for decisions to personnel with a need to know such that work is performed safely in a timely manne |