IR 05000528/2014003
| ML14225A213 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 08/13/2014 |
| From: | Drake J F NRC/RGN-IV/DRP/RPB-D |
| To: | Edington R K Arizona Public Service Co |
| DRAKE, JAMES F | |
| References | |
| EA 13-232 IR-14-003 | |
| Preceding documents: |
|
| Download: ML14225A213 (57) | |
Text
August 13, 2014
EA #13-232 Randall K. Edington, Executive Vice President, Nuclear/CNO Mail Station 7602 Arizona Public Service Company P.O. Box 52034 Phoenix, AZ 85072-2034
SUBJECT: PALO VERDE NUCLEAR GENERATING STATION - NRC INTEGRATED INSPECTION REPORT 05000528/2014003, 05000529/2014003, 05000530/2014003, AND EXERCISE OF ENFORCEMENT DISCRETION
Dear Mr. Edington:
On June 30, 2014, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at the Palo Verde Nuclear Generating Station Units 1, 2, and 3. On July 8, 2014, the NRC inspectors discussed the results of this inspection with Mr. D. Mims and other members of your staff. Inspectors documented the results of this inspection in the enclosed inspection report.
NRC inspectors documented one finding of very low safety significance (Green) in this report. This finding involved a violation of NRC requirements. Further, inspectors documented a licensee-identified violation, which was determined to be of very low safety significance, in this report. The NRC is treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2.a of the NRC Enforcement Policy. Additionally, the inspectors reviewed Licensee Event Report 05000530/2013001-01, "Leakage on Reactor Vessel Bottom Mounted Instrumentation Nozzle 3." The result of this review was identification of a technical specification violation without an associated performance deficiency. As discussed in Section 2.2.4.d of the Enforcement Policy, a violation involving no performance defiency is considered an exception to using only the operating reator assessment program. Therefore, in consultation with the Office of Enforcement, the NRC has concluded that the exercise of enforcement discretion is warranted in accordance with Section 3.5 of the Enforcement Policy, because the violation resulted from matters not within the licensee's control. Accordingly, this violation will not be documented or considered in the NRC's assessment process.
If you contest these violations or significance of these non-cited violations, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region IV; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC resident inspector at the Palo Verde Nuclear Generating Station.
If you disagree with the cross-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region IV; and the NRC resident inspector at the Palo Verde Nuclear Generating Station.
In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, "Public Inspections, Exemptions, Requests for Withholding," a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRC's Public Document Room or from the Publicly Available Records (PARS) component of the NRCs Agencywide Documents Access and Management System (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,/RA/
James F. Drake, Chief Project Branch D Division of Reactor Projects Docket Nos.: 50-528, 50-529, 50-530 License Nos: NPF-41, NPF-51, NPF-74
Enclosure:
Inspection Report 05000528/2014003, 05000529/2014003, 5000530/2014003 w/
Attachment:
Supplemental Information
SUMMARY
IR 05000528, 529, 530/2014003; 04/01/2014 - 06/30/2014; Palo Verde Nuclear Generating Station; Problem Identification and Resolution. The inspection activities described in this report were performed between April 1, 2014, and June 30, 2014, by the resident inspectors at Palo Verde Nuclear Generating Station and inspectors from the NRC's Region IV office. One finding of very low safety significance (Green) is documented in this report. This finding involved a violation of NRC requirements. Additionally, NRC inspectors documented in this report one licensee-identified violation of very low safety significance. The significance of inspection findings is indicated by their color (Green, White, Yellow, or Red), which is determined using Inspection Manual Chapter 0609, "Significance Determination Process." Their cross-cutting aspects are determined using Inspection Manual Chapter 0310, "Components Within the Cross-Cutting Areas." Violations of NRC requirements are dispositioned in accordance with the NRC Enforcement Policy. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process."
Cornerstone: Mitigating Systems
- Green.
The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," for the failure of operations personnel to follow station procedures to ensure that appropriate contingency actions are entered in the operator challenge listing in the control room. Specifically, upon discovery that the pressurizer master level controller could not be placed into manual mode on February 20, 2014, the licensee did not prescribe appropriate contingency actions for operation of the pressurizer level control system. As a result, on March 15, 2014, Unit 3 exceeded the pressurizer maximum level mandated by Technical Specification 3.4.9. The licensee subsequently replaced the faulty controller and has entered this issue in their corrective action program as Palo Verde Action Request 4540981.
The failure of operations personnel to follow station procedures for identifying, documenting, and tracking operator challenges was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it adversely affected the configuration control attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to document an operator work around in the pressurizer level control system allowed operators to place the system in a configuration that challenged the availability, reliability, and capability of the pressurizer to respond to reactor coolant system pressure transients. The inspectors evaluated the significance of the issue under the Significance Determination Process, as defined in Inspection Manual Chapter 0609.04, "Initial Characterization of Findings," and Inspection Manual Chapter 0609, Appendix A, "The Significance Determination Process (SDP) for Findings at-Power." The inspectors concluded the finding was of very low safety significance (Green) because all questions in Exhibit 2 could be answered in the negative. The inspectors determined that the finding had a cross-cutting aspect in the area of human performance because the licensee did not challenge the uncertain condition of the pressurizer level controller. Specifically, after identifying the unexpected failure of the pressurizer level controller, operations personnel did not fully evaluate and manage the risks associated with the degraded condition before proceeding [H.11]. (Section 4OA2)
Licensee-Identified Violations A violation of very low safety significance that was identified by the licensee has been reviewed by the inspectors. Corrective actions taken or planned by the licensee have been entered into the licensee's corrective action program. This violation and the associated corrective action tracking number is listed in Section 4OA7 of this report.
PLANT STATUS
Unit 1 began the inspection period at essentially full power. On May 3, 2014, operators reduced power to approximately 11 percent, for planned maintenance to troubleshoot and repair condenser hotwell tube leakage and a main transformer neutral bushing oil leak. The licensee completed the repairs and returned the unit to full power on May 8, 2014. Unit 1 operated at essentially full power for the remainder of the inspection period. Unit 2 began the inspection period at essentially full power. Operators shut down Unit 2 on April 5, 2014, for Refueling Outage 2R18. The licensee completed the outage and started up Unit 2 on May 2, 2014. On May 5, 2014, operators reduced power from 69 percent to 59 percent to repair a main feedwater pump turbine trip valve. The licensee completed repairs and operators returned Unit 2 to essentially full power on May 7, 2014. Unit 2 operated at essentially full power for the remainder of the inspection period. Unit 3 operated at essentially full power during the inspection period.
REPORT DETAILS
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R01 Adverse Weather Protection
Summer Readiness for Offsite and Alternate AC Power Systems
a. Inspection Scope
On June 3, 2014, the inspectors completed an inspection of the station's off-site and alternate-ac power systems. The inspectors inspected the material condition of these systems, including transformers and other switchyard equipment to verify that plant features and procedures were appropriate for operation and continued availability of off-site and alternate-ac power systems. The inspectors reviewed outstanding work orders and open condition reports for these systems. The inspectors walked down the switchyard to observe the material condition of equipment providing off-site power sources. The inspectors also reviewed corrective action program items to verify that the licensee was identifying issues at an appropriate threshold and entering them into their corrective action program in accordance with station corrective action procedures. The inspectors verified that the licensee's procedures included appropriate measures to monitor and maintain availability and reliability of the off-site and alternate-ac power systems.
These activities constituted one sample of summer readiness of off-site and alternate-ac power systems, as defined in Inspection Procedure 71111.01.
b. Findings
No findings were identified.
1R04 Equipment Alignment
===.1 Partial Walkdown
a. Inspection Scope
The inspectors performed a partial system walkdown of the following risk-significant system: April 11, 2014, Unit 2, emergency diesel generator B April 22, 2014, Unit 2, shutdown cooling, train B June 24, 2014, Unit 1, auxiliary feedwater, train A The inspectors reviewed the licensee's procedures and system design information to determine the correct lineup for the systems. They visually verified that critical portions of the train were correctly aligned for the existing plant configuration. These activities constituted three partial system walk-down samples as defined in Inspection Procedure 71111.04.
b. Findings
No findings were identified.
.2 Complete Walkdown
a. Inspection Scope
On April 23, 2014, the inspectors performed a complete system walkdown inspection of the Unit 2 spent fuel pool cooling system, train B. The inspectors reviewed the licensee's procedures and system design information to determine the correct spent fuel pool cooling system lineup for the existing plant configuration. The inspectors also reviewed condition reports, and other open items tracked by the licensee's operations and engineering departments. The inspectors then visually verified that the system was correctly aligned for the existing plant configuration. These activities constituted one complete system walkdown sample, as defined in Inspection Procedure 71111.04.
b. Findings
No findings were identified.
1R05 Fire Protection
=
Quarterly Inspection
a. Inspection Scope
The inspectors evaluated the licensee's fire protection program for operational status and material condition. The inspectors focused their inspection on four plant areas important to safety:
April 11, 2014, Unit 2, emergency diesel generator B, all elevations April 15, 2014, Unit 2, auxiliary building, 56 feet elevation April 23, 2014, Unit 2, fuel building, 100 feet elevation April 23, 2014, Unit 2, control building, 100 feet elevation For each area, the inspectors evaluated the fire plan against defined hazards and defense-in-depth features in the licensee's fire protection program. The inspectors evaluated control of transient combustibles and ignition sources, fire detection and suppression systems, manual firefighting equipment and capability, passive fire protection features, and compensatory measures for degraded conditions. These activities constituted four quarterly inspection samples, as defined in Inspection Procedure 71111.05.
b. Findings
No findings were identified.
1R06 Flood Protection Measures
a. Inspection Scope
On June 17, 2014, the inspectors completed an inspection of the station's ability to mitigate flooding due to internal causes. After reviewing the licensee's flooding analysis, the inspectors chose one plant area containing risk-significant structures, systems, and components that were susceptible to flooding: Unit 3, control building, 100 feet elevation The inspectors reviewed plant design features and licensee procedures for coping with internal flooding. The inspectors walked down the selected area to inspect the design features, including the material condition of seals, drains, and flood barriers. The inspectors evaluated whether operator actions credited for flood mitigation could be successfully accomplished. These activities constitute completion of one flood protection measures sample, as defined in Inspection Procedure 71111.06.
b. Findings
No findings were identified.
1R08 Inservice Inspection Activities
The activities described in subsections 1 through 4 below constitute completion of one inservice inspection sample, as defined in Inspection Procedure 71111.08.
.1 Non-destructive Examination (NDE) Activities and Welding Activities
a. Inspection Scope
The inspectors directly observed the following nondestructive examinations: SYSTEM WELD IDENTIFICATION EXAMINATION TYPE Safety Injection 76-8 Liquid Penetrant Safety Injection 76-8 Ultrasonic The inspectors reviewed records for the following nondestructive examinations: SYSTEM WELD IDENTIFICATION EXAMINATION TYPE Safety Injection 76-13 Liquid Penetrant Safety Injection 76-17 Liquid Penetrant Safety Injection 76-18 Liquid Penetrant Safety Injection 76-13 Ultrasonic Safety Injection 76-17 Ultrasonic Safety Injection 76-18 Ultrasonic During the review and observation of each examination, the inspectors observed whether activities were performed in accordance with the ASME Code requirements and applicable procedures. The inspectors also reviewed the qualifications of all nondestructive examination technicians performing the inspections to determine whether they were current.
The inspectors directly observed a portion of the following welding activities: SYSTEM WELD IDENTIFICATION WELD TYPE EXAMINATION TYPE Auxiliary Feedwater 4418178-19 Gas Tungsten Arc Weld Radiographic Auxiliary Feedwater 4418178-20 Gas Tungsten Arc Weld Radiographic The inspectors reviewed records for the following welding activities: SYSTEM WELD IDENTIFICATION WELD TYPE EXAMINATION TYPE Auxiliary Feedwater 4418171-3 Gas Tungsten Arc Weld Radiographic Auxiliary Feedwater 4418171-12 Gas Tungsten Arc Weld Radiographic Auxiliary Feedwater 4418178-3 Gas Tungsten Arc Weld Radiographic Auxiliary Feedwater 4418178-6 Gas Tungsten Arc Weld Radiographic The inspectors reviewed whether the welding procedure specifications and the welders had been properly qualified in accordance with ASME Code,Section IX, requirements. The inspectors also determined whether essential variables were identified, recorded in the procedure qualification record, and formed the bases for qualification of the welding procedure specifications.
b. Findings
No findings were identified.
.2 Vessel Upper Head Penetration Inspection Activities
a. Inspection Scope
The inspectors reviewed the results of the licensee's bare metal visual inspection of the reactor vessel upper head penetrations to determine whether the licensee identified any evidence of boric acid challenging the structural integrity of the reactor head components and attachments. The inspectors also verified that the required inspection coverage was achieved and limitations were properly recorded. The inspectors reviewed whether the personnel performing the inspection were certified examiners to their respective nondestructive examination method.
b. Findings
No findings were identified.
.3 Boric Acid Corrosion Control (BACC) Inspection Activities
a. Inspection Scope
The inspectors reviewed the licensee's implementation of its boric acid corrosion control program for monitoring degradation of those systems that could be adversely affected by boric acid corrosion. The inspectors reviewed the documentation associated with the licensee's boric acid corrosion control walk-down as specified in Procedure 73DP-9ZC01, "Boric Acid Corrosion Control Program," Revision 5, and Procedure 70TI-9ZC01, "Boric Acid Walkdown Leak Detection," Revision 17. The inspectors reviewed whether the visual inspections emphasized locations where boric acid leaks could cause degradation of safety significant components, and whether engineering evaluation used corrosion rates applicable to the affected components and properly assessed the effects of corrosion-induced wastage on structural or pressure boundary integrity. The inspectors observed whether corrective actions taken were consistent with the ASME Code and 10 CFR Part 50, Appendix B, requirements.
b. Findings
No findings were identified.
.4 Steam Generator Tube Inspection Activities
a. Inspection Scope
The inspectors reviewed the steam generator tube examination scope and expansion criteria to determine whether the criteria met technical specification requirements, Electric Power Research Institute guidelines, and commitments made to the NRC. The inspectors also reviewed whether the eddy current test inspection scope included areas of degradations that were known to represent potential eddy current test challenges such as the top of tube sheet, tube support plates, and U-bends. The inspectors also reviewed the results of secondary sludge lancing and foreign object search and retrieval inspections, which the licensee performed in both steam generators. The inspectors observed portions of the eddy current testing being performed to determine whether: (1) the appropriate probes were used for identifying the expected types of degradation, (2) calibration requirements were followed, and (3) probe travel speed was in accordance with procedural requirements. The inspectors reviewed the site-specific qualifications for the techniques being used to determine whether eddy current test data analyses were adequately performed per Electric Power Research Institute and site-specific guidelines. The scope of the licensee's steam generator examinations included: 100 percent full-length bobbin testing 100 percent +Point inspection of bobbin flaw-like signals at tube support structures with bobbin indicated depth greater than 15 percent through wall Rotating pancake coil "boxing" of confirmed potential loose parts and observed loose part wear signals Special interest +Point testing of non-resolved freespan bobbin signals and foreign object locations identified by foreign object search and retrieval 100 percent +Point inspection of dent signals greater than two volts at tube supports 100 percent +Point inspection of freespan dent signals greater than five volts 100 percent +Point inspection of peripheral tube freespan dent signals greater than two volts, located within two inches of the top of tubesheet 100 percent +Point inspection of non-expanded tube sites or over expanded tube sites 100 percent tube plug visual inspection in all steam generators Tubesheet periphery and blowdown tube lane foreign object search and retrieval in all steam generators Steam drum upper internals visual inspection of all steam generators In-bundle visual inspection of the top of tube sheet in the low flow kidney regions inclusive of the central cavity region Visual inspection in all steam generators channel head primary side hot leg and cold leg The inspectors reviewed the licensee's identification of the following tube degradation mechanisms: Mechanical wear at tube support structures Foreign object/loose parts induced tube wear The inspectors reviewed the licensee's actions in response to identified loose parts. All loose parts identified were removed during the foreign object search and retrieval inspection. The licensee identified that three adjacent steam generator tubes on the periphery of the tube bundle were subject to near 40 percent through-wall foreign object wear indication. The licensee elected to stabilize and plug the tubes in accordance with the steam generator inspection program.
b. Findings
No findings were identified.
1R11 Licensed Operator Requalification Program and Licensed Operator Performance
(71111.11)
.1 Review of Licensed Operator Requalification
a. Inspection Scope
On June 3, 2014, the inspectors observed simulator training performed by an operating crew. The inspectors assessed the performance of the operators and the evaluators' critique of their performance. The inspectors also assessed the modeling and performance of the simulator during the training activities. These activities constitute completion of one quarterly licensed operator requalification program sample, as defined in Inspection Procedure 71111.11.
b. Findings
No findings were identified.
.2 Review of Licensed Operator Performance
a. Inspection Scope
On April 24, 2014, the inspectors observed the performance of on-shift licensed operators in the Unit 2 main control room. At the time of the observations, the plant was in a period of heightened risk due to mid-loop operations. The inspectors observed the operators' performance of the following activities: Preparation, control and monitoring of reduced reactor coolant system inventory and mid-loop operations, including the pre-job brief In addition, the inspectors assessed the operators' adherence to plant procedures, including conduct of operations procedure, and other operations department policies. These activities constitute completion of one quarterly licensed operator performance sample, as defined in Inspection Procedure 71111.11.
b. Findings
No findings were identified.
1R12 Maintenance Effectiveness
a. Inspection Scope
The inspectors reviewed two instances of degraded performance or condition of safety-related structures, systems, and components (SSCs): May 22, 2014, Unit 1, Unit 2, and Unit 3; auxiliary feedwater discharge pressure instruments, pump A May 29, 2014, Unit 1, Unit 2, and Unit 3, station blackout generators The inspectors reviewed the extent of condition of possible common cause structure, system, and component failures and evaluated the adequacy of the licensee's corrective actions. The inspectors reviewed the licensee's work practices to evaluate whether these may have played a role in the degradation of the structures, systems, and components. The inspectors assessed the licensee's characterization of the degradation in accordance with 10 CFR 50.65 (the Maintenance Rule), and verified that the licensee was appropriately tracking degraded performance and conditions in accordance with the Maintenance Rule.
These activities constituted completion of two maintenance effectiveness samples, as defined in Inspection Procedure 71111.12.
b. Findings
No findings were identified.
1R13 Maintenance Risk Assessments and Emergent Work Control
a. Inspection Scope
The inspectors reviewed four risk assessments performed by the licensee prior to changes in plant configuration and the risk management actions taken by the licensee in response to elevated risk: April 10, 2014, Unit 2, refueling outage 2R18 May 27, 2014, Unit 1, racking out of pressurizer backup heater breaker for calibration of analog isolation amplifier K-SDA-QY-0017 May 28, 2014, Unit 1, backup voltage regulator PNC removed from service for maintenance June 3, 2014, Unit 2, emergency diesel generator A super outage The inspectors verified that these risk assessments were performed timely and in accordance with the requirements of 10 CFR 50.65 (the Maintenance Rule) and plant procedures. The inspectors reviewed the accuracy and completeness of the licensee's risk assessments and verified that the licensee implemented appropriate risk management actions based on the result of the assessments. Additionally, the inspectors also observed portions of one emergent work activity that had the potential, to affect the functional capability of mitigating systems, May 27, 2014, Unit 1, Unit 2, and Unit 3, station blackout generator 2 non-functional due to surveillance test failure The inspectors verified that the licensee appropriately developed and followed a work plan for this activity. The inspectors verified that the licensee took precautions to minimize the impact of the work activity on unaffected SSCs.
These activities constitute completion of five maintenance risk assessments and emergent work control inspection samples, as defined in Inspection Procedure 71111.13.
b. Findings
No findings were identified.
1R15 Operability Determinations and Functionality Assessments
a. Inspection Scope
The inspectors reviewed three operability determinations and three functionality assessments that the licensee performed for degraded or nonconforming SSCs:
April 29, 2014, Unit 3, operability determination of reactor vessel head inner seal leakage April 30, 2014, Unit 3, operability determination of air voiding in containment spray system piping, train A May 6, 2014, Unit 1, functionality assessment of non-seismically qualified electrical conduit supports installed in control building May 14, 2014, Unit 1, Unit 2, and Unit 3, functionality assessment of procedural non-compliance with Generic Letter 96-06 commitments May 21, 2014, Unit 1, Unit 2, and Unit 3, functionality assessment of emergency preparedness RADDOSE dose assessment software isssues June 16, 2014, Unit 2, operability determination of emergency diesel generator A fuel oil cooler cracking The inspectors reviewed the timeliness and technical adequacy of the licensee's evaluations. Where the licensee determined the degraded structure, system, or component to be operable or functional, the inspectors verified that the licensee's compensatory measures were appropriate to provide reasonable assurance of operability or functionality. The inspectors verified that the licensee had considered the effect of other degraded conditions on the operability or functionality of the degraded structure, system, or component. These activities constitute completion of six operability and functionality review samples, as defined in Inspection Procedure 71111.15.
b. Findings
No findings were identified.
1R17 Evaluations of Changes, Tests and Experiments, and Permanent Plant Modifications
a. Inspection Scope
The inspectors reviewed the effectiveness of the licensee's implementation of evaluations performed in accordance with 10 CFR 50.59, "Changes, Tests, and Experiments," and changes, tests, experiments, or methodology changes that the licensee determined did not require 10 CFR 50.59 evaluations.
The inspectors reviewed 12 evaluations required by 10 CFR 50.59; 26 changes, tests, and experiments that were screened out by licensee personnel; and 16 permanent plant modifications. Documents reviewed are listed in the attachment. The inspectors verified that, when changes, tests, or experiments were made, evaluations were performed in accordance with 10 CFR 50.59 and licensee personnel had appropriately concluded that the change, test, or experiment can be accomplished without obtaining a license amendment. The inspectors also verified that safety issues related to the changes, tests, or experiments were resolved. The inspectors reviewed changes, tests, and experiments that licensee personnel determined did not require evaluations and verified that the licensee personnel's conclusions were correct and consistent with 10 CFR 50.59. The inspectors also verified that procedures, design, and licensing basis documentation used to support the changes were accurate after the changes had been made. In the inspection of modifications, the inspectors verified that supporting design and license basis documentation had been updated accordingly and was still consistent with the new design. The inspectors verified that procedures, training plans, and other design basis features had been adequately accounted for and updated. Specific documents reviewed during this inspection are listed in the attachment. These activities constitute completion of 12 samples of evaluations, 26 samples of changes, tests, and experiments that were screened out by licensee personnel; and 16 samples of permanent plant modifications as defined in Inspection Procedure 71111.17.
b. Findings
No findings were identified.
1R18 Plant Modifications
a. Inspection Scope
On June 19, 2014, the inspectors reviewed a permanent modification to the compensatory measures for a potential loss of essential ventilation in the engineered safety features equipment rooms. The inspectors reviewed the design and implementation of the modification. The inspectors verified that work activities involved in implementing the modification did not adversely impact operator actions that may be required in response to an emergency or other unplanned event. The inspectors verified that post-modification testing was adequate to establish the operability of the structure, system, or component as modified. These activities constitute completion of one sample of permanent modifications, as defined in Inspection Procedure 71111.18.
b. Findings
No findings were identified.
1R19 Post-Maintenance Testing
a. Inspection Scope
The inspectors reviewed four post-maintenance testing activities that affected risk-significant SSCs: April 15, 2014, Unit 2, integrated safeguards testing of engineered safety feature relays April 24, 2014, Unit 2, train A shutdown cooling valve SI-651 April 25, 2014, Unit 2, inverter PNA-N11 April 29, 2014, Unit 2, reactor coolant pump 1B The inspectors reviewed licensing and design-basis documents for the structures, systems, or components, and the maintenance and post-maintenance test procedures. The inspectors observed the performance of the post-maintenance tests to verify that the licensee performed the tests in accordance with approved procedures, satisfied the established acceptance criteria, and restored the operability of the affected structures, systems, or components. These activities constitute completion of four post-maintenance testing inspection samples, as defined in Inspection Procedure 71111.19.
b. Findings
No findings were identified.
1R20 Refueling and Other Outage Activities
a. Inspection Scope
During the station's Unit 2 Refueling Outage 2R18, that concluded on May 4, 2014, the inspectors evaluated the licensee's outage activities. The inspectors verified that the licensee considered risk in developing and implementing the outage plan, appropriately managed personnel fatigue, and developed mitigation strategies for losses of key safety functions. This verification included the following: Review of the licensee's outage plan prior to the outage Monitoring of shut-down and cool-down activities Verification that the licensee maintained defense-in-depth during outage activities Observation and review of reduced-inventory and mid-loop activities Observation and review of fuel handling activities Monitoring of heat-up and startup activities These activities constitute completion of one refueling outage sample, as defined in Inspection Procedure 71111.20.
b. Findings
No findings were identified.
1R22 Surveillance Testing
a. Inspection Scope
The inspectors observed five risk-significant surveillance tests and reviewed test results to verify that these tests adequately demonstrated that the SSCs were capable of performing their safety functions:
In-service tests: April 30, 2014, Unit 1, pump A spray pond inservice test Reactor coolant system leak detection tests: May 19, 2014, Unit 3, reactor coolant system leak detection surveillance Other surveillance tests: May 1, 2014, Unit 2, atmospheric dump valve 30 percent partial stroke testing May 8, 2014, Unit 2, containment penetration 45 isolation valve leak-rate testing June 16, 2014, Unit 3, emergency diesel generator 3B 24-hour load testing and full-load reject testing The inspectors verified that these tests met technical specification requirements, that the licensee performed the tests in accordance with their procedures, and that the results of the test satisfied appropriate acceptance criteria. The inspectors verified that the licensee restored the operability of the affected structures, systems, and components following testing. These activities constitute completion of five surveillance testing inspection samples, as defined in Inspection Procedure 71111.22.
b. Findings
No findings were identified.
Cornerstone:
1EP4 Emergency Action Level and Emergency Plan Changes
a. Inspection Scope
The inspector performed an in-office review of changes to Palo Verde Nuclear Generating Station Emergency Plan, Revision 52, and Procedure EP-901, "Classifications," Revision 5. The revision to these documents describe approved changes made to Technical Specification reactor coolant system specific activity limits. Specifically, the licensee will monitor Dose Equivalent Xe-133 vice Coolant Gross Specific activity. These revisions were compared to previous revisions, to the criteria of NUREG-0654, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants," Revision 1, to Nuclear Energy Institute Report 99-01, "Emergency Action Level Methodology," Revision 5, and to the standards in 10 CFR 50.47(b) to determine if the revision adequately implemented the requirements of 10 CFR 50.54(q)(3) and 50.54(q)(4). The inspector verified that the revisions did not decrease the effectiveness of the emergency plan. This review was not documented in a safety evaluation report and did not constitute approval of licensee-generated changes; therefore, these revisions are subject to future inspection. These activities constitute completion of two emergency action level and emergency plan change samples as defined in Inspection Procedure 71114.04.
b. Findings
No findings were identified.
OTHER ACTIVITIES
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency Preparedness, Public Radiation Safety, Occupational Radiation Safety, and Security
4OA1 Performance Indicator Verification
.1 Safety System Functional Failures (MS05)
a. Inspection Scope
For the period of April 1, 2013, through March 31, 2014, the inspectors reviewed licensee event reports (LERs), maintenance rule evaluations, and other records that could indicate whether safety system functional failures had occurred. The inspectors used definitions and guidance contained in Nuclear Energy Institute Document 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 7, and NUREG-1022, "Event Reporting Guidelines: 10 CFR 50.72 and 50.73," Revision 3, to determine the accuracy of the data reported.
These activities constituted verification of the safety system functional failures performance indicator for Unit 1, Unit 2, and Unit 3 as defined in Inspection Procedure 71151.
b. Findings
No findings were identified.
.2 Mitigating Systems Performance Index: Emergency AC Power Systems (MS06)
a. Inspection Scope
The inspectors reviewed the licensee's mitigating system performance index data for the period of April 1, 2013, through March 31, 2014, to verify the accuracy and completeness of the reported data. The inspectors used definitions and guidance contained in Nuclear Energy Institute Document 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 7, to determine the accuracy of the reported data. These activities constituted verification of the mitigating system performance index for emergency ac power systems for Unit 1, Unit 2, and Unit 3, as defined in Inspection Procedure 71151.
b. Findings
No findings were identified.
.3 Mitigating Systems Performance Index: High Pressure Injection Systems (MS07)
a. Inspection Scope
The inspectors reviewed the licensee's mitigating system performance index data for the period of April 1, 2013, through March 31, 2014, to verify the accuracy and completeness of the reported data. The inspectors used definitions and guidance contained in Nuclear Energy Institute Document 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 7, to determine the accuracy of the reported data. These activities constituted verification of the mitigating system performance index for high pressure injection systems for Unit 1, Unit 2, and Unit 3, as defined in Inspection Procedure 71151.
b. Findings
No findings were identified.
4OA2 Problem Identification and Resolution
.1 Routine Review
a. Inspection Scope
Throughout the inspection period, the inspectors performed daily reviews of items entered into the licensee's corrective action program and periodically attended the licensee's condition report screening meetings. The inspectors verified that licensee personnel were identifying problems at an appropriate threshold and entering these problems into the corrective action program for resolution. The inspectors verified that the licensee developed and implemented corrective actions commensurate with the significance of the problems identified. The inspectors also reviewed the licensee's problem identification and resolution activities during the performance of the other inspection activities documented in this report.
b. Findings
No findings were identified.
.2 Semiannual Trend Review
a. Inspection Scope
To verify that the licensee was taking corrective actions to address identified adverse trends that might indicate the existence of a more significant safety issue, the inspectors reviewed corrective action program documentation associated with the following licensee-identified trend:
Declining trend in human performance resulting in two site clock resets and three OSHA recordable injuries within a month Because the licensee identified emergent cross-cutting themes in H.13 ("Consistent Process: Individuals use a consistent, systematic approach to make decisions. Risk insights are incorporated as appropriate") and P.5 ("Operating Experience: The organization systematically and effectively collects, evaluates, and implements relevant internal and external operating experience in a timely manner"), the inspectors reviewed the licensee's response to those themes, to verify that the licensee had taken, was taking, and/or planned to take appropriate actions to address it.
The specific documents reviewed during this trend review are listed in the Attachment. These activities constitute completion of one semiannual trend review sample, as defined in Inspection Procedure 71152. b. Observations and Assessments The inspectors' review of the trends identified above produced the following observations and assessments: For the declining trend in human performance, the licensee performed a high-level review of data over a six-month period to determine if a cause beyond those identified in the event-specific causal analyses could be identified, and if so, identify actions to address the trend. The licensee reviewed clock reset data, prompt human performance interview summaries, corrective action program database documents, and maintenance re-work events to draw insight into current station performance. The licensee identified an unexpected difference in online and outage performance. Specifically, the percentage of department and team clock resets (about 90 percent) was higher during online periods than during outages. Considering the large number of person-hours worked during the outages, the licensee expected about 30-40 percent of the events should occur during the outages. Upon further investigation, the licensee determined that significant efforts had been made in recent years to improve outage performance (suspension of routine meetings, increased leadership oversight of field activities, focused observation program, etc) which were not in place during online periods. The licensee concluded that success during outage periods was due to director presence in the field coaching, mentoring, and reinforcing expectations and fundamental behaviors. The investigation concluded that in the absence of director presence, leaders do not always use teamwork or exhibit a conservative bias to recognize and mitigate risk, resulting in errors or events. The cause of the trend was leaders functioning at one or more levels below that delineated in the leadership model. The licensee initiated corrective actions to establish periodic meetings between directors and their applicable leaders to reinforce the leadership model and appropriate leadership behaviors, as well as, promote direct field observations and worker engagement. Additionally, site directors plan to generate a roles/responsibilities document for all leader levels within their respective departments.
The inspectors considered that, in response to this trend, the licensee had completed an appropriate evaluation and had developed appropriate corrective actions.
For the cross-cutting theme in H.13, the licensee initiated Condition Report/Disposition Request (CRDR) 4520275 soon after the NRC issued the fourth finding during the current assessment cycle that had a cross-cutting aspect in H.13. The licensee performed a common-cause analysis and also reviewed the actions taken in response to each individual finding. The licensee's analysis included additional events since 2012, including site clock resets, injuries, and specific department clock resets. The licensee determined that each individual issue had been adequately addressed in the corrective action program and some correlation existed between the human performance adverse trend, described above, and the H.13 cross-cutting theme. The analysis concluded that decision- making processes did not consistently emphasize compliance. Also, employees sometimes inappropriately accepted risk in the field, or did not clearly understand the risk during field activities. Additionally, the licensee identified process related weaknesses, including latent flaws in some processes and personnel not fully understanding and following the processes. Corrective actions include revision to guidance documents, such as the integrated risk procedure and operability determination procedure, to increase emphasis on compliance; use of a "designated skeptic" during shared risk decision-making; specific focus on process compliance during leadership training; and the implementation of dynamic learning activities that focus on those areas that challenge station performance.
So, for this cross-cutting theme, the inspectors determined that the licensee had entered the theme into their corrective action program in a timely manner, completed an appropriate evaluation of the theme, and developed and scheduled appropriate corrective actions to address identified weaknesses and areas for improvement.
For the cross-cutting theme in P.5, the licensee initiated CRDR 4496637 soon after the NRC issued the fourth finding during the current assessment cycle that had a cross-cutting aspect in P.5. The licensee performed a common cause analysis and also reviewed the actions taken in response to each individual finding. The licensee also reviewed actions implemented in response to previously identified weaknesses, from 2003 - 2010, with identification, evaluation, implementation of lessons learned from operating experience. The licensee concluded that, while previous actions to institutionalize effective use of operating experience had been robust, the station had not consistently used operating experience as an effective event prevention tool, and causal evaluations had not effectively captured missed opportunities from an operating experience preventable perspective. Corrective actions include establishing an operating experience preventable process; including applicable external and internal operating experience into continuing leader training; revising cause evaluator training and qualifications to include the operating experience preventable process; implementation of informal operating experience training for key station personnel; and assessment of previously reviewed operating experience from 2003 - 2010. So, for this cross-cutting theme, the inspectors determined that the licensee had entered the theme into their corrective-action program in a timely manner, completed an appropriate evaluation of the theme, and developed and scheduled appropriate corrective actions to address identified weaknesses and areas for improvement.
c. Findings
No findings were identified.
.3 Annual Follow-up of Selected Issues
a. Inspection Scope
The inspectors selected two issues for an in-depth follow-up: On June 17, 2014, the inspectors reviewed the licensee's root cause evaluation of the Unit 3 dropped control element assembly that occurred on December 2, 2013.
The inspectors assessed the licensee's problem identification threshold, cause analyses, extent of condition reviews and compensatory actions. The inspectors verified that the licensee appropriately prioritized the completed and planned corrective actions, and that these actions appeared adequate to correct the condition and prevent recurrence.
On June 26, 2014, the inspectors reviewed the Unit 3 Technical Specification 3.4.9 entry due to high pressurizer level, and licensee use of operator work-arounds. The inspectors assessed the licensee's problem identification threshold, cause analyses, extent of condition reviews and compensatory actions. The inspectors verified that the licensee appropriately prioritized the planned corrective actions and that these actions appeared adequate to correct the condition. These activities constitute completion of two annual follow-up samples, which included one operator work-around sample, as defined in Inspection Procedure 71152.
b. Findings
Introduction.
The inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," for the failure of operations personnel to follow station procedures to ensure that appropriate contingency actions are entered in the operator challenge listing in the control room. Specifically, upon discovery that the pressurizer master level controller could not be placed into manual mode on February 20, 2014, the licensee did not prescribe appropriate contingency actions for operation of the pressurizer level control system.
Description.
The pressurizer level control system is designed to maintain pressurizer water level within specific limits during power operations. Technical Specifications require that the pressurizer water level be greater than 27 percent and less than 56 percent during power operations. Deviations outside of the level limits adversely affect the ability of the pressurizer to respond to reactor coolant system pressure transients. Excessively high water levels reduce the pressurizer steam space volume, making the pressurizer spray valves less effective for reducing pressure.
The pressurizer water level is controlled by positioning the letdown control valves. The pressurizer master level controller compares the actual level to the desired level setpoint and re-positions the letdown control valves as necessary. There are three means of generating the desired level setpoint. Operators may select either remote-automatic mode, local-automatic mode, or manual mode. In remote-automatic mode, the normal mode of operation, the setpoint is calculated automatically and water level is maintained at a programmed valve that varies with the water temperature. In local-automatic mode, the operator generates the setpoint by turning a thumbwheel on the control board. Lastly, in manual mode, the controller output to the letdown control valves is manipulated manually by the control room operator positioning a sliding switch. Normally, during full power operations, the pressurizer master level controller is operated in remote-automatic mode. On February 20, 2014, during a periodic instrument calibration activity, instrumentation technicians discovered that the Unit 3 pressurizer master level controller, RCN-LIC-110, could not be placed into manual mode. The licensee performed an operability determination and concluded that the controller should be classified as operable-degraded. Although the controller did not have its full capability, operations personnel reasoned that the pressurizer level could continue to be maintained with the controller in remote-automatic mode. The licensee concluded that no compensatory measures were required, and they initiated a corrective maintenance work order to conduct future troubleshooting of the faulty controller. On March 15, 2014, Unit 3 operators were attempting to lower pressurizer level in preparation for main turbine control valve testing. Due to the previously identified problem with the pressurizer master level controller, operators were not able to place the controller in manual. Operators were aware that when shifting from remote-automatic mode to local-automatic mode for level adjustments, they normally first go into manual mode using the manual pushbutton. However, the manual mode pushbutton was unavailable. With the controller in remote-automatic mode, operators matched the remote-automatic and local-automatic setpoints, and then placed the controller in local-automatic mode. Control room operators were unaware that the transition from remote-automatic directly to local-automatic is a "break before make" connection. That is, the control system must sense a deviation between the actual level and the setpoint before the local-automatic setpoint becomes the controlling input to the level controller. The switch from remote-automatic mode directly to local-automatic mode produced a sudden, unexpected response from the letdown control valves. The letdown control valves immediately throttled closed and the pressurizer level started to increase. Operators attempted to lower the level setpoint with the local-automatic thumbwheel, but the system did not respond. Operators then returned the controller back into remote-automatic mode, but this action also did not produce any change to the letdown control valve position.
Operators then placed the controller back in local-automatic mode a second time, and lowered the setpoint to 10 percent below the actual pressurizer level. This action resulted in a response by the pressurizer level control system and operators were able to gradually lower pressurizer level. During the transient, pressurizer level exceeded the Technical Specification limit of 56 percent for approximately 17 minutes. Palo Verde Operating Procedure 40DP-9OP15, "Operator Challenges and Discrepancy Tracking," Revision 26, describes the methods used by operations personnel for identifying, documenting, and tracking operator challenges. Section 6.0 defines an operator challenge as an equipment or program deficiency that provides an obstacle to safe plant operations by requiring operators to perform contingency actions.
Furthermore, any operator challenge that affects transient plant operation requiring operators to perform contingency actions in order to comply with an emergency operating procedure or abnormal operating procedure is further defined as an operator workaround. Step 3.4.1 requires that for all operator challenges, the operations work control supervisor shall ensure that appropriate contingency actions for operators are entered in the operator challenge listing in the control room and on the area operator's operator challenge tracking sheet. Step 3.4.4 requires that operator workarounds include a description of the abnormal operating procedure impact including alternate actions or the inability to complete the required abnormal operating procedure response.
Abnormal Operating Procedure 40AO-9ZZ05, "Loss of Letdown," Step 3 requires operators to place the pressurizer master level controller into manual control mode. Upon discovering that the controller would not go into manual mode on February 20, 2014, the licensee should have identified the condition as an operator workaround. However, in their operability determination, the licensee did not classify the condition as an operator workaround and did not prescribe any alternate action for accomplishing the abnormal operating procedure step for establishing manual pressurizer level control. If the licensee had assessed the impact and had prescribed appropriate contingency action for the operator workaround, the March 15, 2014, pressurizer level transient would not have occurred. Following the event on March 15, 2014, the licensee replaced the faulty controller. The licensee entered this issue in the corrective action program as Palo Verde Action Request 4540981.
Analysis.
The inspectors determined that the failure of operations personnel to follow station procedures for identifying, documenting, and tracking operator challenges was a performance deficiency. The performance deficiency was more than minor, and therefore was a finding, because it adversely affected the configuration control attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to establish appropriate contingency actions for an operator work around in the pressurizer level control system allowed operators to place the system in a configuration that challenged the availability, reliability, and capability of the pressurizer to respond to reactor coolant system pressure transients. The inspectors evaluated the significance of the issue under the Significance Determination Process, as defined in Inspection Manual Chapter 0609.04, "Initial Characterization of Findings," and Inspection Manual Chapter 0609, Appendix A, "The Significance Determination Process (SDP) for Findings at-Power." The inspectors concluded the finding was of very low safety significance (Green) because all questions in Exhibit 2 could be answered in the negative. The inspectors determined that the finding had a cross-cutting aspect in the area of human performance because the licensee did not challenge the uncertain condition of the pressurizer level controller. Specifically, after identifying the unexpected failure of the pressurizer level controller, operations personnel did not fully evaluate and manage the risks associated with the degraded condition before proceeding [H.11].
Enforcement.
Title 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," requires, in part, that activities affecting quality be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and be accomplished in accordance with these instructions, procedures, or drawings. Procedure 40DP-9OP15, "Operator Challenges and Discrepancy Tracking," Revision 26, Step 3.4.1, required that appropriate contingency actions for operators be entered in the operator challenge listing in the control room. Contrary to the above, between February 20, 2014, and March 15, 2014, plant personnel failed to accomplish an activity affecting quality in accordance with prescribed instructions, procedures, and drawings. Specifically, operations work control did not enter appropriate contingency actions for operation of the pressurizer level control system. As a result, on March 15, 2014, Unit 3 exceeded the pressurizer maximum level mandated by Technical Specification 3.4.9. The licensee subsequently replaced the faulty controller, and initiated additional training for operators to reinforce compliance with operator challenges and discrepancy tracking procedures. Because this finding is of very low safety significance and has been entered into the licensee's corrective action program as Palo Verde Action Request 4540981, this violation is being treated as a non-cited violation in accordance with Section 2.3.2 of the Enforcement Policy: NCV 05000530/2014003-01, "Failure to Follow Operator Challenges Procedure."
4OA3 Follow-up of Events and Notices of Enforcement Discretion
These activities constitute completion of four event follow-up samples, as defined in Inspection Procedure 71153.
.1 (Closed) Licensee Event Report 05000528;529/2013-002-00, "Inoperable Systems Due to Defective Relays Reported under Part 21" Between January 22, 2012, and March 31, 2013, the licensee experienced failures of five model ARD660UR control relays.
On April 8, 2013, Westinghouse reported the common cause of the five relay failures, pursuant to 10 CFR Part 21. The common cause of the failures was a manufacturing defect that resulted in failures of the relays to actuate on demand during testing activities. Upon discovery of the common cause failure mechanism, the licensee issued the licensee event report because each of the failures resulted in a condition prohibited by Technical Specifications. The inspectors previously reviewed this issue following the fourth ARD660UR relay failure that occurred on August 15, 2012. The inspectors documented an NRC identified violation in NRC Integrated Inspection Report 05000528;529;530/2012004 for the licensee's failure to consider that the reliability of the model ARD660UR relay population may be reduced due to the multiple failures. The licensee initiated a preventative maintenance program to cycle all relays covered by the scope of the Part 21 report to verify they will not be subject to the common cause failure mechanism. To address the manufacturing problem, Westinghouse has modified the manufacturing process to eliminate the common cause failure mode. The inspectors identified no further issues. This licensee event report is closed.
.2 (Closed) Licensee Event Report 05000528/2013-004-00, "Condition Prohibited by Technical Specification 3.7.2 Due to an Inoperable Main Steam Isolation Valve (MSIV)" On November 6, 2013, Unit 1, operators inappropriately exited Technical Specifications Limiting Condition for Operation 3.7.2.
While recovering from a major failure of the hydraulic actuation system of MSIV-170, the licensee closed, deactivated, and declared operable MSIV-170. The licensee issued the licensee event report to report a condition prohibited by technical specifications. The licensee's operability determination had inappropriately concluded the main steam isolation valve was operable because the valve's safety function was met with the valve closed and deactivated. The licensee concluded the root cause of this event was that the operability determination was overly focused on the ability of the main steam isolation valve to perform its specified safety function and did not adequately consider the definition of operability and compliance with the technical specifications. The licensee has initiated a corrective action to revise its operability determination guidance to explicitly require that evaluation of operability determination technical conclusions support compliance with the technical specification limiting conditions for operation. The inspectors previously dispositioned this issue as an NRC identified violation in Section 1R15 of NRC Integrated Inspection Report 05000528;529;530/2014002. This licensee event report is closed.
.3 (Closed) Licensee Event Report 05000529/2013-002-00, "Automatic Actuation of Unit 2 Reactor Protection System Resulting from Reactor Coolant Pump Trip" On December 2, 2013, Unit 2, reactor coolant pump 1A tripped, resulting in an automatic actuation of the Unit 2 reactor protection system and subsequent reactor trip.
The licensee issued the licensee event report to report the automatic actuation of the reactor protection system.
The licensee determined that the pump motor breaker tripped on a protective relay for phase differential current. The licensee concluded the root cause of the pump trip was a manufacturing defect in the original coil construction. Specifically, the dimensions which created a coil tilt and non-linearity along the slot caused uneven seating of the coil in the stator slots and resulted in coil abrasion and ground-wall insulation pitting. To prevent recurrence, the licensee initiated actions to revise and update procurement documents to include additional reviews by plant engineering of coil dimensional tolerances and additional inspections following coil rewinds. Also, the licensee initiated actions to perform additional visual inspections and testing of the reactor coolant pumps currently in operation to identify any similar defects. The inspectors reviewed the licensee event report and did not identify any additional concerns. This licensee event report is closed.
analysis. Specifically, the original design of ammeter circuits which provide control room current indication for the train B and D, Class 1E batteries and battery chargers did not include overcurrent protection features to limit fault current. In the postulated event, a fire in the control room could cause a ground loop through unprotected ammeter wiring and potentially result in excessive current flow and heating to the point of causing a secondary fire outside the control room in the cable raceways. The postulated secondary fire could affect the availability of equipment needed to place the plant in a safe shutdown condition during a control room fire event. The licensee did not analyze this scenario in accordance with 10 CFR Part 50, Appendix R, Section III.G. The licensee implemented compensatory fire watch measures for the affected fire zones and initiated corrective actions to install fuses to provide overcurrent protection. The inspectors reviewed the licensee event report and documented a licensee-identified finding in Section
4OA7 of this report.
This licensee event report is closed.
.5 (Closed) Licensee Event Report 05000530/2013-001-00 and Supplement
05000530/2013-001-01, Leakage on Reactor Vessel Bottom-Mounted Instrumentation Nozzle 3 On October 6, 2013, during an examination of the bottom mounted instrument (BMI) nozzles on the reactor vessel of Unit 3, white residue was discovered at the annulus region of Nozzle 3. The residue was collected for testing and identified as boron, lithium and trace amounts of primary water radionuclides, which was indicative of pressure boundary leakage during the operating cycle. Pressure boundary leakage is prohibited by Technical Specification 3.4.14. Non-destructive examination of Nozzle 3 identified axial cracking in the nozzle tube and a near-surface weld flaw in the J-groove weld that connects the nozzle to the reactor vessel. This allowed for a reactor coolant leak path in the pressure boundary. Corrective actions included completion of an American Society of Mechanical Engineering (ASME) Code approved half-nozzle repair and increasing the frequency of bottom mounted instrument visual examinations to every refueling outage. An extent of condition evaluation found no indication of unacceptable flaws or leakage in the remaining 60 BMI nozzle assemblies in Unit 3. Additionally, review of past inspections did not identify any evidence of leakage from bottom mounted instrument nozzles in Units 1 or 2.
Destructive testing determined the cause of the event was primary water stress corrosion cracking of the bottom mounted instrument nozzle due to a weld defect that went undetected during fabrication. Inspectors reviewed all available causal information to assess if there was an opportunity to correct this condition prior to failure of the pressure boundary. Inspectors determined that initial installation and examinations of the affected weld were completed within the specifications used at the time of fabrication. Only surface examinations of the root weld pass, and inspections at every 0.5 in of weld thickness were required, and there was an allowance for cold straightening of nozzles, if required. As such, the defect associated with the weld during fabrication was not reasonably within the licensee's ability to foresee, prevent, and correct. Therefore, the inspectors determined that no performance deficiency occurred.
The issue is considered within the traditional enforcement process because there was no performance deficiency associated with the violation of NRC requirements. Inspection Manual Chapter 0612, Power Reactor Inspection Reports, Section 0612-09 states, in part, that such violations are dispositioned using traditional enforcement and may warrant enforcement discretion. The NRC Enforcement Policy, Section 6.1 ("Reactor Operations"), was reviewed to evaluate the significance of this violation. This violation was more than minor and best characterized at Severity Level IV (very low safety significance) because it is similar to Enforcement Policy Section 6.1.d.1.
Additionally, a qualitative assessment of the observed RCS leakage condition concluded the risk was of very low safety significance (Green). The basis for this qualitative risk determination was that the leakage rate was very small with little boron residue accumulation on the lower reactor vessel head and no appreciable accumulation on the structures beneath the vessel. Any leakage was within the capability of RCS makeup systems. Additionally, detailed inspections did not reveal any loss of vessel material. The NRC decided to exercise enforcement discretion in accordance with Section 3.5 of the NRC Enforcement Policy and refrain from issuing enforcement action for the violation of Technical Specification 3.4.14.a (EA-13-232) for the following reasons: this issue is of very low safety significance (Green); it was determined that this issue was not within the licensee's ability to foresee and correct; the licensee's actions did not contribute to the degraded condition; and the actions taken were reasonable to identify and address this matter. Further, because the licensees actions did not contribute to this violation, it will not be considered in the assessment process or the NRC's Action Matrix. Specific documents reviewed during this inspection are listed in the attachment. This licensee event report and its supplement are closed.
.6 Unusual Event for Seismic Event
a. Inspection Scope
On Saturday, June 28, at 10:17 p.m., the Palo Verde Unit 1 shift manager declared an Unusual Event for a seismic event, per EAL HU1.1. Ground motion was felt by personnel in the Unit 2 and Unit 3 Operations Support Building, and the Unit 1 shift manager was notified. The Unit 1 shift manager confirmed, per the United States Geological Survey website, that a magnitude 5.2 earthquake had occurred at 9:59 p.m., approximately 31 miles northwest of Lordsburg, NM, and approximately 178 miles East-Southeast of Phoenix. In accordance with EAL HU1.1, two of the three requirements were met for declaration of an Unusual Event. The inspectors responded to the site and monitored the licensee's response. The licensee performed walkdowns of all three units and the Independent Spent Fuel Storage Installation (ISFSI) and did not identify any issues. The licensee subsequently terminated the Unusual Event at 1:12 a.m. on Sunday, June 29, 2014.
b. Findings
No findings were identified.
4OA6 Meetings, Including Exit Exit Meeting Summary On April 18, 2014, resident inspectors presented the inspection results to Mr. B. Bement, Senior Vice President, Nuclear Operations, and other members of the licensee staff.
The licensee acknowledged the issues presented. The licensee confirmed that any proprietary information reviewed by the inspectors had been returned or destroyed. On May 28, 2014, a regional inspector conducted a telephonic exit meeting to present the results of the in-office inspection of changes to the licensee's emergency plan and emergency action levels to Ms. A. Krainik, Manager, Emergency Preparedness, and other members of the licensee staff. The licensee acknowledged the issues presented. The licensee confirmed that any proprietary information reviewed by the inspectors had been returned or destroyed.
On June 26, 2014, regional inspectors presented the inspection results to Mr. J. Cadogan, Vice President, Nuclear Engineering, and other members of the licensee's staff, on the results of the evaluation of changes, tests, and experiments, and permanent plant modification. The licensee acknowledged the results as presented. Proprietary information was provided to the inspectors and all proprietary information was returned to the licensee.
On July 8, 2014, regional inspectors presented the inspection results to Mr. D. Mims, Senior Vice President , Regulatory Assessment, and other members of the licensee staff. The licensee acknowledged the issues presented. The licensee confirmed that any proprietary information reviewed by the inspectors had been returned or destroyed.
4OA7 Licensee-Identified Violations The following violation of very low safety significance (Green) was identified by the licensee and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy for being dispositioned as a non-cited violation.
Palo Verde Nuclear Generating Station Unit 1 Operating License Condition 2.C(7), Unit 2 Operating License Condition 2.C(6), and Unit 3 Operating License Condition 2.F required, in part, that the licensee implement and maintain, in effect, all provisions of the approved fire protection program as described in the Final Safety Analysis Report for the facility, as supplemented and amended, and as approved in the Safety Evaluation Report (SER), through Supplement 11. Palo Verde Nuclear Generating Station Updated Final Safety Analysis Report, Section 9.5.1.3, stated, in part, that fire protection has been achieved consistent with 10 CFR Part 50, Appendix R, Part III, Sections G, J, and O. Appendix R,Section III.G.2, states, in part, that where cables or equipment, including associated non-safety circuits that could prevent operation or cause maloperation due to hot shorts, open circuits, or shorts to ground, of redundant trains of systems necessary to achieve and maintain hot shutdown conditions, are located within the same fire area outside of primary containment, one of three means of protecting cables to ensure that one of the redundant trains is free of fire damage shall be provided. Contrary to the above, on October 4, 2013, the licensee identified that non-safe shutdown cables that shared common electrical cable trays with safe shutdown cables were not electrically protected, and therefore, did not provide means of protecting the cables to ensure that the credited train of safe shutdown equipment would be free of fire damage. Specifically, the licensee identified that the battery ammeter circuits which provide control room current indication for the train B and D, class 1E batteries and battery chargers do not include overcurrent protection features to limit fault current. During a postulated fire event in the control room, fire-induced failures could have damaged the ammeter circuits, resulting in a secondary fire and damage to other safe shutdown cables that are in direct physical contact with these cables, which could affect the availability of equipment needed to place the plant in a safe shutdown condition. A senior reactor analyst performed a detailed risk evaluation and the bounding change to the core damage frequency was less than 3.6E-7/year. The dominant core damage sequences involved a control room fire initiating event in Panel B01, a secondary cable fire in a cable tray associated with one train of safety related equipment, and having the alternate train of safety related equipment out of service for maintenance. The low fire frequency and the normal train separation and protection, that are required by the fire protection program, helped to minimize the significance. The licensee entered this violation into its corrective action program as Condition Report/Disposition Request 4458522. The licensee established compensatory measures in the affected fire zones and initiated corrective actions to install fuses in the affected circuits. The licensee submitted Licensee Event Report 05000528;529;530/2013-003-00 to report this issue. Refer to Section
4OA3 of this inspection report for the review and closure of the
licensee event report.
A-1 Attachment
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
- N. AaronsCooke, Engineer, Nuclear Regulatory Affairs
- B. Bement, Senior Vice President, Site Operations
- M. Brannin, ISI Program Owner, Engineering Programs
- M. Brutcher, Senior Engineer, Predictive Maintenance and Cause Analysis
- N. Byrnes, Section Leader, Minor Modifications
- J. Cadogan, Vice President, Nuclear Engineering
- M. DiLorenzo, Department Leader, Engineering Programs
- S. Dornseif, Engineer, Nuclear Regulatory Affairs
- R. Doyle, Senior Engineer, Design Electrical
- E. Fernandez, Senior Engineer, Engineering Programs
- C. Fox, Senior Engineer, System Engineering - Electrical and I and C
- K. Geiszler, Engineer III, Transient Analysis
- D. Hansen, Engineer, Engineering Programs
- D. Haug, Section Leader, Plant Modification
- R. Hicks Jr., Senior Engineer, Transient Analysis
- R. Hunzelman, Shift Manager, Operations Work Control
- H. Hurley, Chemistry Operations Supervisor, Chemistry Work Management
- N. Illovits, Senior Engineer, Mechanical Procurement Engineer
- D. Kelsey, Supervisory Leader, Nuclear Regulatory Affairs
- A. Krainik, Manager, Emergency Preparedness
- S. Lopez, Section Leader, Predictive Maintenance and Cause Analysis
- L. Martinez, Shift Manager, Operations Work Control
- H. McKaig III, Nuclear Engineering Plant Department Leader, Equipment Reliability Engineering
- D. Mims, Senior Vice President, Regulatory and Oversight
- D. Morrow, Senior Engineer, Mechanical Procurement Engineer
- D. Naughton, Supervisory Leader, Engineering Programs
- C. Noack, Senior Engineer, Plant Modification
- F. Oreshack, Consultant, Regulatory Affairs
- R. Roehler, Senior Engineer, Nuclear Regulatory Affairs
- C. Stephenson, Engineer, Nuclear Regulatory Affairs/Licensing
- N. Sylvester, Senior Engineer, Digital Projects Project Management
- D. Teal, Chemistry Technician, Chemistry Work Management
- M. Waldo, Senior Engineer, Minor Modifications
- C. Wax, Engineer, Engineering Programs
- T. Weber, Department Leader, Nuclear Regulatory Affairs
- C. Williams, Engineer III, Plant Modification
- T. Williams, Section Leader, Emergency Preparedness
- J. Wood, Engineer III, Plant Modification
- A. Wullbrandt, Senior Engineer, Fuel Cycle Services
NRC Personnel
None
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened and Closed
- 05000530/2014003-01 NCV Failure to Follow Operator Challenges Procedure (Section 4OA2)
Closed
- 05000528; 529/2013-002-00 LER Inoperable Systems Due to Defective Relays Reported under Part 21 (Section 4OA3)
- 05000528/LER-2013-004-00 LER Condition Prohibited by Technical Specification 3.7.2 Due to an Inoperable Main Steam Isolation Valve (MSIV) (Section 4OA3)
- 05000529/LER-2013-002-00 LER Automatic Actuation of Unit 2 Reactor Protection System Resulting from Reactor Coolant Pump Trip (Section 4OA3)
- 05000528; 529; 530/2013-003-00 LER Appendix R Unanalyzed Condition - Direct Current Ammeter Circuits Without Overcurrent Protection (Section 4OA3)
LIST OF DOCUMENTS REVIEWED
Section 1R01: Adverse Weather Protection Procedures Number Title Revision 40DP-9OP34 Switchyard Administrative Control 20 40OP-9ZZ19 Hot Weather Protection 5 93DP-0LC12 Administrative Control and Compliance of
Section 1R04: Equipment Alignment Miscellaneous Documents Number Title Date
- System Health Report:
- PC-Fuel Pool Cooling and Clean Up for August 1, 2013 to January 31, 2014 March 19, 2014
- 2920331 TSCCR
- 3996661 CM
Procedures
- Number Title Revision 40OP-9AF01 Essential Auxiliary Feedwater System 59 40OP-9DG02 Emergency Diesel Generator B 68 40OP-9PC01 Fuel Pool Cooling 12 40OP-9SI01 Shutdown Cooling Initiation 53 73DP-9ZZ26 MOV Testing With Quicklook 2 73ST-9AF04
- AFA-P01 Full Flow - Inservice Test 18
- Palo Verde Action Requests
- 3160836
- 3885854
- 3992331
- 4268384
- 4407237
- 4493909
- 4518435
- 4519060
- 4547506
- 4547915
Condition Report
/ Disposition Requests 4408663
Section 1R05: Fire Protection Procedures Number Title Revision
- PVNGS Pre-Fire Strategies Manual, Control Building Elev. 100'-0" 24
- PVNGS Pre-Fire Strategies Manual, Fuel Building, Elev. 100'-0" 24
- Pre-Fire Strategies Manual, Auxiliary Building Elev. 51'6" 24 14-DP-0FP33 PVNGS Pre-Fire Strategies Manual 25
- Palo Verde Action Requests
- 4522053
Calculations
- 13-MC-FP-808
Section 1R06: Flood Protection Measures Calculations Number Title Revision 13-MC-EC-0511 Essential Chilled Water System 6 13-MC-ZJ-0200 As Built Control Building Flooding Calculation 8 31-MC-ZA-0809 As Built Auxiliary Building Flooding Calculation 7
- Palo Verde Action Requests 3590255
Section 1R08: Inservice Inspection Activities (71111.08) Procedures Number Title Revision 01PR-0AP04 Corrective Action Program 8 30DP-9MP03 System Cleanliness and Foreign Material Exclusion Controls 20 70TI-9ZC01 Boric Acid Walkdown Leak Detection 17 73DP-9WP01 Welder and Procedure Qualification 6 73DP-9WP04 Welding and Brazing Control 15 73DP-9ZC01 Boric Acid Corrosion Control Program 5 73TI-0ZZ13 Radiographic Examination 17 73TI-9RC01 Steam Generator Eddy Current Examinations 29 73TI-9RC10 Bare Metal Visual Examination of Reactor Vessel Bottom Head 2 73TI-9RC10 Bare Metal Visual Examination of Reactor Vessel Bottom Head 2 73TI-9ZZ07 Liquid Penetrant Examination 14 73TI-9ZZ22 Visual Examination for Leakage - Interval 3
- 7 73WP-0ZZ07 Welding of Stainless and Nickel Alloys 16 73WP-0ZZ20 Visual Inspection of Code Welds 8 81DP-9RC01 PVNGS Steam Generator Management Program 14 MN725-A02017 Rolled Mechanical Tube Plugging and Stabilizer Installation Palo Verde Nuclear Generating Station 0 MN725-A02019 Position Verification Procedure 0
Procedures
- Number Title Revision MN756-A00002
- PDI-UT-2 Generic EPRI Procedure for the Ultrasonic Examination of Austenitic Pipe Welds 0
- Steam Generator Program Documents Number Title Revision
- SG-SGMP-14-2 Palo Verde U2R18 Steam Generator Degradation Assessment 0 02-MS-C035 Steam Generator Condition Monitoring Evaluation, Unit 2, Cycle 16 0
- NDE Reports Numbers14-190 14-321 14-316 14-310 14-UTE-2027 14-UTE-2053 14-UTE-2031 14-UTE-2032 14-PT-2024 14-PT-2026 14-PT-2027 14-PT-2037
- Palo Verde Action Requests
- 3556683
- 3865668
- 3880687
- 4096911
- 4231874
- 4294303
- 4294626
- 4373587
- 4420201
- 4420265
- 4524789
- 4503237
- 4515760
- 4515762
- 4515824
- 4515827
- 4515830
- 4515831
- 4515836
- 4515838
- 4515842
- 4515843
- 4515848
- 4515850
- 4515851
- 4515854
- 4515858
- 4515860
- 4520962
- 4520970
- 4520984
- 4520990
- 4520996
- 4521613
- 4522699
- 4522206
- 4523409
- 4523921
- 4523983
- 4524297
- 4523923
- 4523926
- 4523934
- 4524516
- 4519930
- 4524694
Work Orders
- 4418121
- 4418171
- 4418178 4335568
Section 1R11: Licensed Operator Requalification Program and Licensed Operator Performance Miscellaneous Documents Number Title
- NLR 14S030201 Simulator Scenario, "Loss of Instrument Air, Loss of Letdwn, Functional Recovery Procedure, Inadvertent PPS"
Procedures
- Number Title Revision 40DP-9OP02 Conduct of Shift Operations 61 10OP-9ZZ16 RCS Drain Operations 76
Section 1R12: Maintenance Effectiveness Miscellaneous Documents
- Title Date
- MRule Function Scoping - GT System January 31, 2013
Procedures
- Number Title Revision 70DP-0MR01 Maintenance Rule 35 73DP-0EE05 Engineering Preventive Maintenance Program 7 86DP-0EE01 Reliability Centered Maintenance (RCM) Based System Reviews 7
- Palo Verde Action Requests
- 4505419
- 4506053
- 4507377
- 4510327
- 4516644
- 4517046
- 4539453
- 4546764
Condition Report
- Action Items
- 4424679
- 4488739
- 4506667
Condition Report
/ Disposition Requests
- 4421849
- 4488738
- 4506666
- 4507484
- 4508753
- 4511410
- 4511560
- 4517075
Work Orders
- 4505862
Section 1R13: Maintenance Risk Assessments and Emergent Work Control Miscellaneous Documents
- Title Date
- Shutdown Safety Function Assessment April 10, 2014
Procedures
- Number Title Revision 40DP-9AP21 Protected Equipment 5 40DP-9RS02 Shutdown Risk Management 1 70DP-0MR01 Maintenance Rule 35 70DP-0RA05 Assessment and Management of Risk When Performing Maintenance in Modes 1 and 2 20
- Palo Verde Action Requests
- 4356658
- 4539453
- 4539476
Work Orders
- 4373889 4539453
Section 1R15: Operability Determinations and Functionality Assessments Miscellaneous Documents Number Title Revision
- ODMI Evaluation / Implementation Plan - Unit 3 Reactor Vessel Flange Indication of Leakage from the Inner O-Ring 0 13-EN-0304 Installation Specification for Electrical Conduit and Junction Box/Equipment at Palo Verde Nuclear Generating Station, Units 1, 2, 3 8
Procedures
- Number Title Revision
- EP-0905 Protective Actions 4
Procedures
- Number Title Revision 40DP-9OP26 Operations PVAR Processing and Operability Determination/Functional Assessment 31 93-DP-0LC08 Regulatory Commitment Tracking 3
- Palo Verde Action Requests
- 4036293
- 4374138
- 4522316
- 4528726
- 4529761
- 4533443
- 4536490
- 4542500
- 4546734
Condition Report
/ Disposition Requests
- 4526959
- 4532316
Work Orders
- 3821466
- 3987052
- 4174561
- 4212006 4543900
Section 1R17: Evaluations of Changes, Tests and Experiments, and Permanent Plant Modifications Calculations Number Title Revision 13-MC-SI-0016 Tri-Sodium Phosphate Basis Calculation 7 13-MC-SI-0215
- HPSI System Performance Evaluation and Surveillance Requirement Basis Calculation 8 13-MC-SI-0215 HPSI System Performance Evaluation and Surveillance Requirement Basis Calculation, Appendix C HPSI Pump Maximum Recirculation Evaluation 7 02-MS-C027 An Evaluation for the Higher 2R16 Mode 4 Entry SFP Decay Heat Load 0 13-JC-CH-0209 Refueling Water Tank Level Measurement 12 13-MC-CH-0201 Refueling Water Tank (RWT), Hold-Up Tank (HT), and Reactor Make-Up Water Tank (RMWT) Sizing 8 13-MC-HA-0052 Auxiliary Building Essential Cooling System Heat Load Calculation (Appendix 'A') 9 13-MC-HJ-0003 Control Building (HJ) System Heat Load and Equipment Selection Calculation 8 13-MC-EC-0252 EC System Water Requirements and Chiller Sizing 12
Calculations
- Number Title Revision 13-MC-SP-307 SP/W System Thermal Performance Design Bases Analysis 9 13-MC-FP-0315 10CFR50 Appendix R Safe Shutdown Equipment List 13 13-CC-ZS-0202 Miscellaneous Cabinet and Panel Support Calculation 1 13-EC-NA-0222 Electrical Penetration Protection 8 13-EC-NG-0111 AC Equipment Protection (480V LC) Non-Class 1E 4 13-EC-FP-0004 10CFR50 Appendix R Safe Shutdown cable identification Analysis 8
Drawings
- Number Title Revision 01-CH-952-H-006 Pipe Support Assembly 0 01-CH-952-H-005 Pipe Support Assembly 0 01-CH-952-H-004 Pipe Support Assembly 0 01-CH-952-H-003 Pipe Support Assembly 0 01-CH-952-H-002 Pipe Support Assembly 0 01-CH-952-H-001 Pipe Support Assembly 0 03-E-NKB-002 125V DC Non-Class 1E Power System 480V Breaker for Battery Chargers 3E-NKNH17, 3E-NKN-H18, 3E-NKN-H20, and 3E-NKN-H21 5 01-E-NKB-002 125V DC Non-Class 1E Power System 480V Breaker for Battery Chargers 1E-NKNH17, 1E-NKN-H18, 1E-NKN-H20, and 1E-NKN-H21 6 03-E-PNA-001 120V AC Class 1E Power System Ungrounded Vital Instrument and Control Distribution Panels 3E-PNA-D25 and 3E-PNC-D27 15 03-E-PKA-001 125V DC Class 1E and 120V AC Vital Instrument Power System 5 13-M234A-00163 Hi-Pro Manifold Double Block and Bleed Valve Assembly Drawing 1 01-M-SGP-001 Main Steam System 69
Drawings
- Number Title Revision 02-M-SGP-001 Main Steam System 75 01-E-PGA-006 480V Class 1E Power System Load Center 1E -PGB-L36 7 01-E-PGA-002 480V Class 1E Power System Load Center 1E-PGB-L32 11 01-E-PGA-001 480V Class 1E Power System Load Center 1E-PGA-L31 9 13-J200-00075 Main Control Panel Cutout Details
- MK-43 thru
- MK-61 13 13-J200-00048 Main Control Panel Wiring Terminal Boards CVCS Panel BO3 32 01-M-ECP-001 Essential Chilled Water System 35 13-A-ZZD-002 Typical Penetration Seal Details Conduits 28 01-E-PBB-001 4.16KV Class 1E Power System Switchgears 1E-PBA-S03 and 1E-PBB-S04 4.16KV Normal Supply Breaker 7 01-E-PBA-002 4.16 KV Class 1E Power System Switchgear 1E-PBB-S04 12 01-E-RCF-017 Control Wiring Diagram Reactor Coolant System Pressurizer Level Control 1 01-E-RCB-017 Elementary Diagram Reactor Coolant System Pressurizer Level Control 3 01-E-NHB-011 Elementary Diagram, 480V Non-Class Power System 480V Feeder Breakers Fed from 480V MCC with Shunt Trip Coil Typical 4 03-E-PEF-001, Sh.2 Control Wiring Diagram Stand-By Generation System, Diesel Generator 3E-PEA-G01 4.16KV Breaker 6 03-E-PEF-003, Sh.1 Control Wiring Diagram Stand-By Generation System, Diesel Generator 3E-PEA-G01 Tripping and Voltage Regulation 3 13-M234A-0011 FWIV Hydraulic Tubing Schematic 5 13-M234A-0124 Schematic For Anchor/Darling Self Contained Hydraulic Actuator 7 13-M234A-0067 Schematic For A/DV Self Contained Hydraulic Actuator Assembly For FWIV 12 13-M234A-0121 A/DV Hydraulic Actuator Assy. for Feed Water Isolation Valve 9 13-M234A-0122 A/DV Hydraulic Actuator Assy. Part List Feed Water Isolation Valve 11 01-M-SGP-002 P and I Diagram Main Steam System 48 13-M234A-0109 Hydraulic Tubing Schematic 5
Drawings
- Number Title Revision 13-M234A-0123 Schematic For Anchor/Darling Self Contained Hydraulic Actuator 7 13-M234A-0066 Schematic For A/DV Self Contained Hydraulic Actuator - MSIV, Sh. 2 12
- Evaluations Number Title Revision E-07-0006 The proposed activity is to implement DMWO 2859190. In summary the activity will remove the existing 90VR panel which includes the Automatic and Manual Voltage Regulators, the Motor Operated Potentiometers (MOP) mounted adjacent to the 90VR and install a new 90VR panel which contains two AVRs and two Digital Reference Units (DRU) as replacement for the MOPs. 3 E-07-0006 The proposed activity is to implement DMWO 2859190. In summary the activity will remove the existing 90VR panel which includes the Automatic and Manual Voltage Regulators, the Motor Operated Potentiometers (MOP) mounted adjacent to the 90VR and install a new 90VR
panel which contains two AVRs and two Digital Reference Units (DRU) as replacement for the MOPs. 4 E-07-0013 10
- CFR 50.59 Evaluation of CENTS Code Version 06100 and
- CN-OA-04-24, Revs. 00,
- 01, and 02 2 E-10-0008 Upgrade of the Control Element Drive Mechanism Control System (CEDMCS) by replacing various Printed Circuit Boards (PCB), Power Supply Assemblies, and Circuit Breakers. 2 E-11-0001 This 10
- CFR 50.59 screening/evaluation covers the changes that are not AD-out by the applicability determination (AD) in the following documents:
- Revision 01 of Analysis Of Record (AOR) "Total Loss Of Reactor Coolant Flow" (LOFTA, Reference 1), and LDCR 2011-F004: updates to the associated UFSAR (Reference 2) Chapter 15.3.1. 0 E-11-0003 The proposed activity issues Engineering Study 02-MS-C027, Revision 0 which evaluates increasing the administrative limit for decay heat load from 12.6E6 BTU/hr to 13.5E6 BTU/hr for Mode 4 entry during 2R16. 0
- Evaluations Number Title Revision E-11-0007 The purpose of this 10CFR50.59 screening/evaluation is to address changes to the evaluation of Double-Ended Break Of A Letdown Line Outside Containment (DBLOCUS, or LDLB herein) (LDLB-15.6.2-TA, Ref. S-1) described in PVNGS UFSAR Chapter 15.6.2 (Ref. S-2). 0 E-12-0001 The proposed activity involves temporarily lowering the Pressurizer Pressure Control System (PPCS) master setpoint in response to Pressurizer Safety Valve (PSV) leakage and includes changes to supporting design documents. 1 E-12-0004 The activity is a compensatory action/procedure change to place the automatic static transfer switch for the Class 1E instrument inverter 3E-PNA-N11 in a manual mode. 0 E-12-0006 This evaluation reviews the compensatory measures to be taken in the event of a loss of cooling function of the Essential Safety Feature (ESF) Equipment Room Air Handling Units (AHUs)
- HJA-Z04 or
- HJB-Z04 or the Electrical Penetration Room AHUs
- HAA-Z06 or
- HAB-Z06. These compensatory actions are identified in the Loss of HVAC procedure 40AO-9ZZ20. 0 E-13-0001 The Legacy Honeywell Plant Computer (PC) System and associated software are being replaced with a new Rolls Royce System. The proposed change is applicable to Unit 2 and 3 only.
- Changes to include Unit 1 will be provided in the next revision. 0 E-13-0002 This document is a revision to the original 10
- CFR 50.59 Screening (S-09-0166) that was performed as part of the Legacy Core Monitoring Computer replacement (DMWO 2579403) portion of the Plant Monitoring System (PMS).
- This revision is a result of
- CRDR 4374602 that was initiated following Evaluation E-13-0001 that was performed as part of the Legacy Plant Computer (PC) replacement portion of the PMS. 0
- Evaluations Number Title Revision/Date
- 2461811 Replace the existing General Electric Ground Fault Relays with more reliable Westinghouse GFRs in the indicated "key" component applications. January 22, 2003
- 2587598 Using Vendor supplied modification installation kits for conversion of valves from rotating; rising stems to non-rotating, rising design Change CH Systems MOVs From Rising-Rotating Stem to Non-Rising Rotating Stem Valves. April 4, 2013
- 2772517 Process DCR to Implement MOD For Power Distribution System In the Containment March 13, 2009
- 2909756 Replace Actuator Gears on 13J-SI-652 and 23J-SI-651 to achieve an OAR of 186.40 and reset the torque switches on 13J-SI-652 and 13J-SI-651 per the associated EDC of this DMWO package. January 12, 2012
- 2938489 Potential exists for air entrainment in ECCS piping during suction switchover from RWT to ECCS sumps. October 27, 2011
- 3095435 In conjunction with the planned replacements of the RVHs, there exists the opportunity to design and install an innovative, maintenance-friendly RRP, which significantly reduces the duration for reactor destack and restack (66 hours7.638889e-4 days <br />0.0183 hours <br />1.09127e-4 weeks <br />2.5113e-5 months <br />) by reducing the number of crane picks from 24 to 3.
- The typical duration of the destack/restack activities and demand for the polar crane is significant and affects the overall outage schedule duration.
- This effort also significantly reduces the opportunities for negative impact to the outage schedule frequently caused by equipment failures and human performance errors during the destack/restack process. December 15, 2011
- 3133718 Replace the negative sequence (46) relay, 3JDGNB02A*746*RELAYX for EDG A. February 19, 2012
- 3345923 This modification will lower the low cooler refrigerant temperature trip setpoint, for analog temperature controllers. June 22, 2012
- 3497433 Amend back to have a DMWO generated for modification to install a new Essential Cooling Water Heat Exchanger. March 1,2013
- 3560653 This PVAR is to request a plant modification to replace the underground duct banks and associated cabling from all three EDG buildings to the six DFOST vaults with direct buried submersible cables. February 8, 2014
- 4304156 FLEX Primary Mechanical Connections, Reactor Coolant System MAKE UP April 22, 2014
- 4345882 FLEX Alternates Mechanical Connections, Steam Generator MAKE UP April 22, 2014
- Evaluations Number Title Revision/Date MEE04090 Item Type Technical Evaluation for Commercial Grade Dedication of Tri-Sodium Phosphate (TSP) Used in Containment 2 MEE04397 Part Substitution Evaluation for material change on coupling-key
- SAE 1020 carbon steel (original material) ASTM A108 Grade 1018 Carbon Steel replaces the original material of construction due to higher strength. Replacement
- APN 00126895 is approved for use in the Fuel Pool Cooling Pump (PC) System. 0 MEE04408 Parts Declassification for the Diesel Generator Exhaust Silencer Model 32 IN M-41, IPS R281-2 Group PM01.
- This MEE superseded SPQCE R281-002 Revision 00. Reason for Declassification:
- Some piece parts are not required to support Safety Function of Components. 0 MEE04458 Replacement Grease for Alvania EP series products. 1
Miscellaneous Documents
- Number Title Revision/Date
- CENTS-06100-LIMITATION CENTS 06100 limitations of use 0
- CENTS-LTR-OA-06-124 CENTS release notes for version 06100 0
- SEATS v4.2.0 Software, Software Error & Activity Tracking System, Software Configuration Control, Limitations of Use for CENTS Version 06100 January 31, 2014
- CENPD-282-P-A Technical Description Manual for the Cents Code Volume 1 1 2
- CENPD-282-P-A Technical Description Manual for the Cents Code Volume 4 1 2 ASTM Designation: D217-10 ASTM
- Standard Test Methods for Cone Penetration of Lubricating Grease
- ASTM Designation: D6185-11 Standard Practice for Evaluating Compatibility of Binary Mixtures of Lubricating Greases
- EPRI Nuclear Maintenance Applications Center:
- Lube Notes Compilation, 1989-2007
Miscellaneous Documents
- Number Title Revision/Date
- PCP 3926066 Pending Change Package:
- Add PO NONE
- APN 125592 to BOM R281-2 PM01. Add
- CRAI 4191022 Perform jar testing of Bulab 6002 with selected Closed Cooling systems to determine if there are any foaming issues which may preclude the products use in the systems. October 19, 2012
- 3023945 CRAI
- 3981525 CRDR
- 3332883 EWR
- 3989295 CRDR
- 4546624 CRDR
- SWMS
- 4481807 10
- CFR 50.59 Program Self-Assessment May 16, 2014
- VTD-O988-00006 USERS MANUAL for
- HI-Q118C006
- March 17, 2010 JN1011-A00003 Applications Detailed Design (ADD) for CMC (Core Monitoring Computer) 1 13-MS-B094 Operator Action Time for RWT Isolation After RAS 1 NNI1211001 Operations Training Department Area 1 and Area 2 Job Qualification Card April 12, 2013 NNR13C030400 Non Licensed Operator Continuing Training - Auxiliary Feedwater (AF) April 23, 2013 13-MS-A082 HVAC Environmental Design Parameters 1 02-MS-C027 An Evaluation for the Higher 2R16 Mode 4 Entry SFP Decay Heat Load 0 13-MC-SI-0984 Revise 13-MC-SI-0984, Rev 0 to add Fukushima FLEX RC Primary and Alternate relief valves (1,2,3JSIEPSV0476 and 1,2,3JSIEPSV0475) 0
- 3548453 Installation of Temporary Pressure Transducers and Data Acquisition Equipment for CVCS Charging Pump Pressure Pulsation Measurements in Unit 3 0 5875-9, Sh. 3 Main Control panel cutout Detail 15 5875-12 Main Control Panel Wiring Diagram- CVCS Panel - BO9 16
Procedures
- Number Title Revision 81DP-0CCP05 Design and Technical Document Control 45 91DP-0EN32 Management of Sumps and Manholes 1 73ST-9CL05 Containment Air Lock Leakage Rate Test 26 73ST-9CL04 Personnel Air Lock Interlock Test 14 73ST-9CL03 Containment Airlock Door Seal Leak Test 26 32MT-9ZZ58 Preventive Maintenance of Elgar Inverters 32 40AL-9RK1A Panel B01A Alarm Responses 0 40OP-9PN01 120V AC Class 1E Instrument Channel "A" 11 40OP-9ZZ16 RCS Drain Operations 76 40OP-9SI01 Shutdown Cooling Initiation 54 40OP-9SI02 Recovery from Shutdown Cooling to Normal Operating Lineup 104 40DP-9ZZ04 Time Critical Action (TCA) Program 9 40EP-9EO10 Standard Appendices 83 81DP-0CC28 Classification of Structures, Systems, and Components 12 81DP-9ZZ01 Penetration Seal Determinations 4 40ST-9ZZ20 Remote Shutdown Disconnect Switch and Control Circuit Operability 19 40ST-9ZZ20 Remote Shutdown Disconnect Switch and Control Circuit Operability - Modes 3, 4, 5, 6, or Defueled 24 87DP-0MC06 Material Engineering Evaluation 25 81DP-0EE10 Design Change Process 32 81DP-0DC13 Deficiency (DF) Work Order 28 93DP-0LC07 10
- CFR 50.59 and 72.48 Screenings and Evaluations 24 93DP-0LC07-01 10
- CFR 50.59 and 72.48 Administrative Guideline 0
- Screens Number Title Revision S-08-0374 The purpose of this DMWO is to permanently install three (3) 480V power panels in each Containment Building to support refueling outage work activities. 0 S-09-0119 Revision 2 - 10CFR50.59 screening S-09-0119 is revised to revision 2 supporting revision 1 of DMWO 2750245. Revision 2 to this screening is performed to provide clarification related to the protective function settings used in the new Beckwith relays.
- Additionally, administrative corrections are made and the references section is updated.
- Revision 1 - 10
- CFR 50.59 screening S-09-0119 is revised to revision 1 and supports revision 1 of DMWO
- 2750245 which changed the type, model number, and manufacturer for the new digital relay unit to be installed in the station blackout generators (SBOG). 2 S-10-0042 This 10
- CFR 50.59 Screening is to cover the upgrades to the Containment Personnel Air Locks (PALs) located on elevations 100' and 140' of the Containment Building in all 3 units. 1 S-10-0062 Modification Project ID #CH-642 proposed activity installs a vendor supplied valve kit on Motor Operated Valve (MOV)
- DCIDs JCHBHV0255 and JCHAHV0524 to convert the valves from a rising, rotating stem design; to a rising, nonrotating stem design. 0 S-10-0137 This 50.59 Screening revision 04 is in support of revision 1 of DMWO
- 3479906, DISPO Pen and Ink #1 which removes various unused, maintenance intensive, NSSS vent valves and replace with ASME socket weld caps. 4 S-10-0300 Calculation 13-MC-HJ-0003 (Rev. 7) is being issued to account for slight heat load increases as a result of
- CRAI 3404341 and 2608317.
- This revision also includes impacts from IandC updates to Revision 10 of 13-JCHJ-0203 via approved EDCs 2007-00259 and 2009-00421.
- Revision 7 evaluates the impact of this heat load increase and the resulting increase of temperature in the essential Control Room Envelope and other areas as well as the capacities of the Air handling Units (AHU's). 0
- Screens Number Title Revision S-11-0004 The Palo Verde Nuclear Generating Station (PVNGS) is equipped with two (2) Essential Cooling Water (EW) heat exchangers per unit.
- The Unit 2 train A EW heat exchanger (2MEWAE01**HTEXCH) has experienced tube degradation to the point where Palo Verde plant management has determined it is necessary to replace the heat exchanger with a new "like-for-like" heat exchanger purchased to the same performance requirements as the original heat exchanger.
- This DMWO
- 3497433 replaces the existing Unit 2 train A Essential Cooling Water heat exchanger manufactured by Struthers with one manufactured by Babcock and Wilcox. 1 S-11-0015 Isolate faulted cable 1ENG04NC3RB which is located between Load Center 1ENGNL26 and 1ENANS02H.
- Install a temporary cable between Load Centers 1ENGNL26 and 1ENGNL28 which will retain the existing protection circuit design. 1 S-11-0025 This activity corrects a wiring deficiency in the Emergency Diesel Generator (EDG) control circuits for the 'B' EDG Output Breaker and the Essential Exhaust Fan Breaker.
- In the event of a Control Room fire, faults could cause a loss of breaker control power requiring Operators to manually close the breakers if required. 0 S-11-0026 S-11-0026 has been revised.
- During U2R17 implementation of DIWO
- 3593945 for MSIV 2JSGEUV170, an interference with electrical conduit 2EZC2EARR16 prevented the installation of the double block and bleed valve within the allowance of EQCF D2012-0016.
- This revision reworks electrical conduit 2EZC2EARR16 and its associated cables 2ESG23AC1RC, 2ESG23AC1RD, and 2ESG23AC1RH to remove and avoid additional interferences with the double block and bleed valve, the valve filter, and junction box 2ESGAJ15. 2 S-11-0028 DMWO
- 3348483 will replace the existing three-position charging pump selector hand-switch, 1/2/3JCHNHS0004 (HS-4), with a four position hand switch. 0 S-11-0032 TMOD
- 3548453 installs pressure and acceleration instrumentation, and the associated data acquisition system on the CVCS system in the Unit 3 Auxiliary Building. 0
- Screens Number Title Revision S-11-0037 Calculation 13-MC-EC-0252, Rev 9, "EC System Water Requirements and Chiller Sizing" is being revised to account for overall heat load increase as a result of input calculations 13-MC-HA-0052, Auxiliary Bldg. Essential Cooling system heat load calculation, and 13-MC-HJ-0003, HJ System heat load and Equipment selection calculation. 0 S-11-0041 Six new sections were added to procedure 40ST 9ZZ20 to test Switch
- PBB-S04 feeder breaker and class 1E train B load center supply breaker. 0 S-11-0063 To implement VDP A17788 and
- EDC 2011-00319 for revision to Structural Integrity Associates Inc. (SIA) design calculations on the full structural weld overlay modification to the pressurizer and hot leg dissimilar metal welds.
- This revision was initiated by SIA to correct self-identified calculation errors which affect the fatigue design of the overlay modification. (PVAR 3438711) 0 S-11-0180 Revision of Calculation 13-MC-HD-0053 (Rev. 3) to include incorporation of the evaluation regarding PVNGS site outside environment minimum and maximum design basis temperatures for Diesel Generators HVAC system. 0 S-12-0003 Minor Modification
- DG-1456, DMWO
- 3133718, replaces the negative sequence relay for each Emergency Diesel Generator (EDG) except for the 3B EDG. 0 S-12-0054 Issue Calculation 13-MC-SI-0215 Revision 7 and implement associated changes to Procedure 40OP-SI02 that aligns the High Pressure Safety Injection (HPSI) systems for boration of the cold-leg injection flow paths. 0 S-12-00106 The proposed activity is to issue Calculation 13-MC-SI-0250, Revision 1.
- This screening also addresses proposed revisions to the PVNGS Technical Specification 3.4.8 and 3.9.5 BASES initiated by LDCR 2012-B006 and proposed revision to UFSAR 5.4.15 initiated by LDCR 2013-F008. 1 S-12-0131 This 50.59 Screening Rev 0 is in support of Rev 0 of DMWO
- 2660297 which will permanently relocate the blind flange downstream of letdown drain valve(s) 13PCHEV853**VALVEX. 0
- Screens Number Title Revision S-12-0191 The proposed change adds Bulab 6002 as a biocide to be used in the Closed Cooling Water Systems (i.e., Essential Chilled Water (EC), Essential Cooling Water (EW), Diesel Generator Cooling Water Jackets (DG), Nuclear Cooling Water (NC), Turbine Cooling Water (TC) and Normal Chilled Water (WC). 2 S-13-0046 Create a new Test Instruction, 40TI-9AF01, Rev. 00, that will allow for manual operation of the
- AFA-P01 Turbine Driven Auxiliary Feed Pump without the 'A' train DC power supply available.
- This test instruction will be performed to provide verification that the class, Turbine Driven Auxiliary Feed Pump can be controlled by manual Operator action while in Mode 3, while intentionally simulating a loss of DC control power to the pump and associated valves. 0 S-13-0170 The proposed activity is to modify the Technical Requirements Manual (TRM) to remove plant shutdown requirements for various TRM Limiting Conditions for Operation (TLCOs).
- Specifically, TLCOs T3.1.103, Charging Pumps - Operating, T3.4.101, RCS Chemistry, T3.4.102, Pressurizer Heatup and Cooldown Limits, T3.5.201, Shutdown Cooling System, and T3.6.200,
- Pre-Stressed Concrete Containment Tendon Surveillance. Revision 1 of this Screening addresses feedback from System Engineering to correct typographical and format issues and to simplify the response to question 1.
- The changes are identified with change bars in the margin. 1 S-14-0004 The change to the subchannel uncertainties is also used as input to calculation 13-JC-RJ-0205, "Core Operating Limit Supervisory System (COLSS) and Core Protection Calculator (CPC) Measurement Channel Uncertainty". The purpose of this calculation is to serve as an interface between NFM and I and C Design Engineering setpoint and uncertainty calculations.
- The change that occurred in 13-JC-RJ-0205 as a result to the change that occurred in 13-JC-SE-0202 subchannel uncertainties is not adverse.
- Any changes that occur to SABD or Transient Analysis as a result of the change to 13-JC-RJ-0205 will be screened accordingly by NFM. 1 S-14-0022 This 10
- CFR 50.59 Screening addresses five input parameters changes made by the Analysis of Record (AOR) Feedwater Line Break (FWLB) calculation and an update to the UFSAR which required further review as determined by the Applicability Determination made in Appendix A of
- FWLB-TA, Rev. 1. [Reference 2] 1
Section 1R18: Plant Modifications Miscellaneous Documents Number Title Date
- E-12-0006 10
- CFR 50.59 Screening for 40ST-9EC03 and 40AO-9ZZ20 December 23, 2013
Procedures
- Number Title Revision 40AO-9ZZ20 Loss of HVAC 6 40ST-9EC03 Loss of Essential Ventilation 15
Section 1R19: Post-Maintenance Testing Procedures Number Title Revision 32MT-9ZZ58 Preventive Maintenance of Elgar Inverters 32 32MT-9ZZ84
- AC Motor Operational Testing 26 73ST-9DG01 Class 1 E Diesel Generator and Integrated Safeguards Test, train A 25
- Palo Verde Action Requests
- 4519593
- 4519570
- 4519595
- 4519647
- 4519771
- 4522602
- 4524945
- 4525168
- 4528404
Work Orders
- 4279199
- 4316518
- 4342685
- 4481999
- 4495619
- 4522618 4524934
Section 1R20: Refueling and Other Outage Activities Miscellaneous Documents
- Title Date
- 2R18 Revision "C" Schedule March 5, 2014
- 2R18 3-Day Critical Path April 5, 2014
Procedures
- Number Title Revision 40OP-9ZZ11 Mode Change Checklist 90 40OP-9ZZ23 Outage
- GOP 66 40ST-9RC01 RCS and Pressurizer Heatup and Cooldown Rates 17 40ST-9ZZ09 Containment Cleanliness Inspection 23
- Palo Verde Action Requests
- 4516797
- 4518429
- 4522877
- 4522951
- 4524452
- 4524455
- 4524992
- 4525636
- 4525777
- 4526618
Work Orders
- 4343947 4418405
Section 1R22: Surveillance Testing Miscellaneous Documents Number Title Date 13-MS-A119 Evaluation of Atmospheric Dump Valves' Optimum Trip and Body Height Difference September 9, 2011
Procedures
- Number Title Revision 40ST-9RC02 ERFDADS (Preferred) Calculation of Water Inventory 53 73ST-9CL01 Containment Leakage Type "B" and "C" Testing 40 73ST-9DG08 Class 1E Diesel Generator Load Rejection, 24 Hour Rated Load and Hot Start Test train B 9 73TI-9SG03
- ADV 30% Partial Stroke Test 7 73ST-9SP01 Essential Spray Pond Pumps - Inservice Test 40 73ST-9SP02 Essential Spray Pond Pumps - Comprehensive Pump Test 11 73ST-9XI20 ADVs - Inservice Test 37
- Palo Verde Action Requests
- 4161870
- 4316302
- 4529068
- 4544984
- 4544989
Condition Report
- Action Items
- 3565262
Condition Reports
/ Disposition Requests
- 3123174
- 3557255
- 4164816
- 4529330
Work Orders
- 4330332
- 4342703
- 4344384
- 4348373
- 4384782
- 4529080 4529081
Section 4OA1: Performance Indicator Verification Miscellaneous Documents Number Title
- PVNGS Control Room Operator Logs
- 2013-002-00 Licensee Event Report
- 2013-004-00 Licensee Event Report
Procedures
- Number Title Revision 93DP-0LC09 Data Collection and Submittal Using INPO's Consolidated Data Entry System 11 93DP-0LC10 SSFF Mitigating System Performance Indicator 4
- Palo Verde Action Requests
- 4432918
- 4493929
- 4496292
- 4498205 4530668
Section 4OA2: Problem Identification and Resolution Miscellaneous Documents
- Title Date
- CARB Comments for
- CRDR 4481512 June 5, 2014
- PVNGS Monthly Trend Report March 2014
- PVNGS Monthly Trend Report April 2014
- Significant Condition Investigation Review Checklist for
- CRDR 4481512 May 21, 2014
Miscellaneous Documents
- Title Date
- Training Review Group Meeting for Site Leader Training June 10, 2014
Procedures
- Number Title Revision 40AO-9ZZ05 Loss of Letdown 24 40DP-9OP15 Operator Challenges and Discrepancy Tracking 26
Condition Reports
/ Disposition Requests
- 4481512
- 4506573
- 4513596
- Palo Verde Action Requests
- 4487622
- 4489320
- 4498289
- 4505456
- 4513043
- 4513069
- 4529461
- 4532269 4540981
Section 4OA3: Follow-up of Events and Notices of Enforcement Discretion Miscellaneous Documents Number Title
- 50237 Event Notification Worksheet
Procedures
- Number Title Revision 40AO-9ZZ21 Acts of Nature 31 79IS-9SM01 Analysis of Seismic Event 25
Condition Reports
/ Disposition Requests
- 4458522 4480904