IR 05000528/2016003
| ML16315A375 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 11/10/2016 |
| From: | Geoffrey Miller NRC/RGN-IV/DRP/RPB-D |
| To: | Bement R Arizona Public Service Co |
| GEOFF MILLER | |
| References | |
| IR 2016001, IR 2016003 | |
| Download: ML16315A375 (33) | |
Text
November 10, 2016
SUBJECT:
PALO VERDE NUCLEAR GENERATING STATION - NRC INTEGRATED INSPECTION REPORT 05000528/2016003, 05000529/2016003, AND 05000530/2016003, AND INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) INSPECTION REPORT 07200044/2016001
Dear Mr. Bement:
On September 30, 2016, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Palo Verde Nuclear Generating Station Units 1, 2, and 3. On October 11, 2016, the NRC inspectors discussed the results of this inspection with Ms. M. Lacal 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.
The finding involved a violation of NRC requirements. The NRC is treating this violation as a non-cited violation (NCV) consistent with Section 2.3.2.a of the NRC Enforcement Policy.
If you contest the violation or significance of the NCV, 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 a 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 NRCs 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/
Geoffrey B. Miller, Branch Chief Project Branch D Division of Reactor Projects Docket Nos. 05000528, 529, 530, 07200044 License Nos. NPF 41, NPF 51, NPF 74
Enclosure:
Inspection Report 05000528/2016003, 05000529/2016003, and 05000530/2016003, and 07200044/2016001 w/ Attachment : Supplemental Information
REGION IV==
Docket: 05000528, 05000529, 05000530, 07200044 License: NPF-41, NPF-51, NPF-74 Report: 05000528/2016003, 05000529/2016003, 05000530/2016003, and 07200044/2016001 Licensee: Arizona Public Service Company Facility: Palo Verde Nuclear Generating Station Location: 5801 South Wintersburg Road Tonopah, AZ 85354 Dates: July 1 through September 30, 2016 Inspectors: C. Peabody, Senior Resident Inspector D. Reinert, PhD, Resident Inspector D. You, Resident Inspector L. Brookhart, Senior ISFSI Inspector G. Guerra, Emergency Preparedness Inspector J. Reynoso, Resident Inspector Approved Geoffrey B. Miller By: Chief, Project Branch D Division of Reactor Projects 1 Enclosure
SUMMARY
IR 05000528, 529, 530/2016003, 07200044/2016001; 07/01/20106 - 09/30/2016; PALO
VERDE NUCLEAR GENERATING STATION INTEGRATED INSPECTION REPORT; PLANT MODIFICATIONS.
The inspection activities described in this report were performed between July 1 and September 30, 2016, by the resident inspectors at Palo Verde and inspectors from the NRCs Region IV office and other NRC offices. One finding of very low safety significance (Green) is documented in this report. This finding involved a violation of NRC requirements. 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,
Aspects 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: Initiating Events
- Green.
The inspectors identified a Green, non-cited violation of 10 CFR Part 50,
Appendix B, Criterion V, Instructions, Procedures, and Drawings, because the licensee failed to accomplish activities affecting quality in accordance with documented procedures.
Specifically, the inspectors identified multiple examples of design changes performed under the Advanced Work Authorization process which were placed into service prior to the completion of the associated engineering work orders. As an immediate corrective action, the licensee instituted a requirement for the design engineering director to approve all Advance Work Authorizations to ensure the in service point is clearly identified and understood. The licensee entered this issue into the corrective action program as Condition Report 16-09965.
The failure to establish adequate constraints to ensure that final engineering approval of advance work is completed prior to placing modified systems in service was a performance deficiency. This performance deficiency was more than minor, and therefore a finding, because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, routinely failing to implement the requirements of the engineering design change advance work authorization process could result in equipment being placed in service without an approved design configuration. In accordance with Inspection Manual Chapter 0609, Significance Determination Process, Attachment 4,
Initial Characterization of Findings, dated June 19, 2012, Table 2, reactor coolant system boundary issues are considered under the Initiating Event Cornerstone. Using Table 3, the inspectors determined the finding pertained to an event or degraded condition while the plant was shutdown and, therefore, used Inspection Manual Chapter 0609, Appendix G Shutdown Operations Significance Determination Process, dated May 9, 2014, for significance determination. The inspectors reviewed Appendix G, Attachment 1, Exhibit 2,
Initiating Events Screening Questions. The inspectors answered No to Question A.1, and found all other questions to be not applicable and therefore concluded that the finding was of very low safety significance (Green). The inspectors determined that this finding had a human performance crosscutting aspect associated with work management, because the licensee did not coordinate with all affected work groups so that operations personnel understood the constraints prior to placing the modified system back in service [H.5].
(Section 1R18)
PLANT STATUS
Unit 1 entered the inspection period at full power. From July 29-30, 2016, operators reduced power to 12 percent rated thermal power and took the generator offline for planned switchyard equipment repairs. On September 7, 2016, a faulted instrument power bus caused the main pressurizer spray valve to fail open, resulting in a loss of pressurizer pressure control and reactor depressurization. The condition required reactor operators to trip the reactor and secure the reactor coolant pumps in order to regain pressure control. Unit 1 entered a forced outage to repair the pressurizer spray valve as well as the voltage regulator on the instrument power bus.
Unit 1 was restarted on September 12, 2016, and returned to full power for the remainder of the inspection period.
Unit 2 operated at full power for the entire inspection period.
Unit 3 entered the inspection period at full power. On September 19, 2016, a protective system hardware failure on the control rod motor generator set sent a false positive trip signal to the reactor power cutback system, resulting in a turbine trip and an automatic power reduction to 45 percent rated thermal power. Operators subsequently shutdown the reactor to perform troubleshooting and repairs offline. The control system was repaired and Unit 3 restarted on September 22, 2016, and returned to full power for the remainder of the inspection period.
REPORT DETAILS
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R04 Equipment Alignment
Partial Walk-Down
a. Inspection Scope
The inspectors performed partial system walk-downs of the following risk-significant systems:
- July 7, 2016, Unit 2 control building essential ventilation system B
- July 20, 2016, Unit 3 low pressure safety injection system A
- July 27, 2016, Unit 2 containment spray system B
- September 28, 2016, Unit 1 4.16kV Class 1E power system A The inspectors reviewed the licensees procedures and system design information to determine the correct lineup for the systems. They visually verified that critical portions of the systems or trains were correctly aligned for the existing plant configuration.
These activities constituted four partial system walk-down samples as defined in Inspection Procedure 71111.04.
b. Findings
No findings were identified.
1R05 Fire Protection
.1 Quarterly Inspection
a. Inspection Scope
The inspectors evaluated the licensees fire protection program for operational status and material condition. The inspectors focused their inspection on six plant areas important to safety:
- July 7, 2016, Unit 2 diesel fuel oil day tank B room, fire area 23B
- July 12, 2016, Unit 1 auxiliary building 100 east corridor, fire area 42C
- July 12, 2016, Unit 1 auxiliary building 100 west corridor, fire area 42D
- July 14, 2016, Unit 3 reactor trip switchgear and control element drive mechanism control room, fire area 54
- August 25, 2016, Unit 2 upper cable spreading room, fire area 20
- August 30, 2016, Unit 3 essential chiller room A, fire area 1 For each area, the inspectors evaluated the fire plan against defined hazards and defense-in-depth features in the licensees 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 six quarterly inspection samples, as defined in Inspection Procedure 71111.05.
b. Findings
No findings were identified.
.2 Annual Inspection
a. Inspection Scope
On September 2, 2016, the inspectors completed their annual evaluation of the licensees fire brigade performance. This evaluation included observation of an unannounced fire drill for quarterly proficiency on August 29, 2016.
During this drill, the inspectors evaluated the capability of the fire brigade members, the leadership ability of the brigade leader, the brigades use of turnout gear and fire-fighting equipment, and the effectiveness of the fire brigades team operation. The inspectors also reviewed whether the licensees fire brigade met NRC requirements for training, dedicated size and membership, and equipment.
These activities constituted one annual inspection sample, as defined in Inspection Procedure 71111.05.
b. Findings
No findings were identified.
1R06 Flood Protection Measures
a. Inspection Scope
On July 25, 2016, the inspectors completed an inspection of underground bunkers susceptible to flooding. The inspectors selected two underground vaults that contained risk-significant or multiple-train cables whose failure could disable risk-significant equipment:
- Unit 3 spray pond electrical cabling vault (3EZV08BKEM12)
- Unit 3 spray pond electrical cabling vault (3EZYS1NKFM63)
The inspectors observed the material condition of the cables and splices contained in the bunkers and looked for evidence of cable degradation due to water intrusion. The inspectors verified that the cables and vaults met design requirements.
These activities constituted completion of one bunker/manhole sample, as defined in Inspection Procedure 71111.06.
b. Findings
No findings were identified.
1R11 Licensed Operator Requalification Program and Licensed Operator Performance
.1 Review of Licensed Operator Requalification
a. Inspection Scope
On August 2, 2016, the inspectors observed a portion of an evaluated simulator scenario 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 requalification activities.
These activities constituted 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 July 29, 2016, the inspectors observed the performance of on-shift licensed operators in the plants main control room. At the time of the observations, the plant was in a period of heightened risk due to planned down-power of Unit 1 for switchyard insulator repairs. Additionally, the inspectors observed the operators performance of the following activities:
- September 12, 2016, Unit 1 reactor startup following forced outage for a stuck-open pressurizer spray valve
- September 19, 2016, Unit 3 response to a turbine trip/reactor power cutback and subsequent shutdown
- September 22, 2016, Unit 3 reactor start up following forced outage for a turbine trip In addition, the inspectors assessed the operators adherence to plant procedures, including power operations procedure and other operations department policies.
These activities constituted 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):
- August 11, 2016, Unit 1 high pressure safety injection valve 616, bonnet seal weld leakage
- September 20, 2016, Unit 1 diesel generator B turning gear interlock, this sample included a review of applicable quality control criteria The inspectors reviewed the extent of condition of possible common cause SSC failures and evaluated the adequacy of the licensees corrective actions. The inspectors reviewed the licensees work practices to evaluate whether these may have played a role in the degradation of the SSCs. The inspectors assessed the licensees 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 five 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:
- July 1, 2016, online risk assessment for Unit 3 during pressurizer heater scheduled electrical maintenance
- August 2, 2016, weekly assessment for Unit 2
- August 9, 2016, weekly assessment for Unit 2
- September 14-16, 2016, online risk assessment for Unit 3 during remote shutdown disconnect switch operability test
- September 26, 2016, online risk assessment for Unit 1 during diesel generator B maintenance 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 licensees risk assessments and verified that the licensee implemented appropriate risk management actions based on the results of the assessments.
Additionally, on July 29, 2016, the inspectors observed portions of Unit 1 switchyard insulator repair that had the potential to cause an initiating event:
The inspectors verified that the licensee appropriately developed and followed a work plan for these activities. The inspectors verified that the licensee took precautions to minimize the impact of the work activities on unaffected SSCs.
These activities constituted completion of six 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 eight operability determinations and functionality assessments that the licensee performed for degraded or nonconforming SSCs:
- July 10, 2016, functionality assessment of Unit 3 diesel generator A HVAC panel
- July 12, 2016, operability determination of Unit 2 B balance of plant engineered safety features actuation system 15 volt power supply failure
- July 15, 2016, operability determination of Unit 1 loss of audible alarms on boards 6 and 7
- July 29, 2016, operability determination of Unit 1 diesel generator A degraded generator rotor pole
- August 23, 2016, functionality assessment of Unit 1 auxiliary building roof drainage capacity
- August 31, 2016, operability determination of Unit 3 auxiliary feedwater bypass steam supply valve isolation capability
- September 20, 2016, operability determination of Unit 3 incorrect bill of materials justification for replacement reactor coolant system wide range pressure transmitter
- September 27, 2016, operator work around of Unit 2 low pressurizer pressure setpoint in channel A failure to reset to a lower value The inspectors reviewed the timeliness and technical adequacy of the licensees evaluations. Where the licensee determined the degraded SSC to be operable or functional, the inspectors verified that the licensees 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 SSC.
The inspectors reviewed operator actions taken or planned to compensate for degraded or nonconforming conditions. The inspectors verified that the licensee effectively managed these operator workarounds to prevent adverse effects on the function of mitigating systems and to minimize their impact on the operators ability to implement abnormal and emergency operating procedures.
These activities constituted completion of eight operability and functionality review samples, which included one operator work-around sample, as defined in Inspection Procedure 71111.15.
b. Findings
No findings were identified.
1R18 Plant Modifications
a. Inspection Scope
On September 21, 2016, the inspectors completed a limited review of recent plant design changes performed under the licensees advance work authorization (AWA)process.
The inspectors reviewed the design and implementation of the modifications. 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 functionality of the SSC as modified.
These activities constituted completion of one sample of permanent modifications, as defined in Inspection Procedure 71111.18.
b. Findings
Introduction.
The inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because the licensee failed to accomplish activities affecting quality in accordance with station procedures.
Specifically, the inspectors identified multiple examples of design changes performed under the Advanced Work Authorization process which were placed into service prior to the completion of the associated engineering work orders.
Description.
Palo Verde Procedure 81DP-0EE10, Design Change Process, describes the licensees process for controlling, developing, installing, and processing plant design changes. Appendix G of this procedure defines an Advanced Work Authorization (AWA)as a component of the design change process that provides authorization for maintenance to proceed with work activities in the field without issuance of a completely approved design. An AWA is used to allow field work to start while engineering personnel are completing the required engineering documentation. The use of an AWA is beneficial for time sensitive activities, such as emergent repairs to plant equipment during outages. The field work is performed immediately, although the final engineering approved design might require additional changes or rework, before placing the modified equipment back in service.
Procedure 81DP-0EE10 Appendix G requires that the advance field work does not affect any in-service equipment. Furthermore, step G.1.1.4 states that the scope of the AWA must not communicate or interact with any operating structures, system, or components.
Step G.1.1.4 lists examples including: no heat transfer, fluid flow, pressure exchange, electrical interaction, or unbounded structural impact. Lastly, step G.1.1.7 requires that the modified systems not be placed back in service until after the approval of the engineering work orders.
The inspectors identified three examples in which the licensee had not developed adequate constraints to ensure that the final engineering approval was complete prior to placing modified components back in service.
- During the Spring 2016, Unit 1 refueling outage, a visual inspection by licensee engineers identified evidence of pressure boundary leakage at a reactor coolant pump cold leg temperature instrument nozzle. The licensee addressed the leakage by installing a mechanical nozzle seal assembly clamp under the AWA process. The reactor coolant system was re-filled prior to the approval of the engineering change documentation.
- During the Spring 2016, Unit 1 refueling outage, the licensee discovered small cracks on the inlet partition plate of the letdown heat exchanger. The licensee used an advanced work authorization to perform a weld repair of the damaged partition plate and then re-filled the heat exchanger in preparation for restart.
- At the start of the Spring 2015, Unit 3 refueling outage, licensee engineers observed evidence of pressure boundary leakage at a reactor coolant pump suction pressure instrument nozzle. The licensee used the AWA process to repair the cracked nozzle with a partial nozzle replacement. The reactor coolant system was re-filled prior to the approval of the engineering work orders.
In each case, contrary to station procedure requirements, the subject plant equipment was placed in service prior to the design being approved by the engineering department.
The inspectors reviewed each design change and confirmed that the eventual post-maintenance testing was adequate to demonstrate the operability or functionality of the modified components.
The issue was entered into the licensees corrective action program as CR 16-09965.
The licensee determined that operations personnel did not have a clear understanding of the in service terminology used in the 81DP-0EE10 Design Change Process procedure. As an interim corrective action, the licensee instituted a requirement for the design engineering director to approve all AWAs to ensure the in service point is clearly identified and understood. The licensee is evaluating additional controls to prevent components modified under the AWA process from prematurely being placed in service prior the approval of the engineering work orders.
Analysis.
The inspectors determined that the licensees failure to establish adequate constraints to ensure that final engineering approval of advance work is completed prior to placing modified systems in service was a performance deficiency. This performance deficiency was more than minor, and therefore a finding, because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, routinely failing to implement the requirements of the engineering design change advance work authorization process could result in equipment being placed in service without an approved design configuration. In accordance with Inspection Manual Chapter 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings, dated June 19, 2012, Table 2, reactor coolant system boundary issues are considered under the Initiating Event Cornerstone. Using Table 3, the inspectors determined the finding pertained to an event or degraded condition while the plant was in shutdown and, therefore, used Inspection Manual Chapter 0609, Appendix G Shutdown Operations Significance Determination Process, dated May 9, 2014, for significance determination. The inspectors reviewed Appendix G, Attachment 1, Exhibit 2, Initiating Events Screening Questions. The inspectors answered No to Question A.1, and found all other questions to be not applicable and therefore concluded that the finding was of very low safety significance (Green). The inspectors determined that this finding had a human performance crosscutting aspect associated with work management, because the licensee did not coordinate with all affected work groups so that operations personnel understood the constraints prior to placing the modified system back in service [H.5].
Enforcement.
Title 10 CFR, Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be accomplished in accordance with documented instructions, procedures, or drawings, of a type appropriate to the circumstances. Licensee procedure 81DP-0EE10, Design Change Process, Revision 39, an Appendix B quality related procedure, provides instructions for implementing advanced work authorization engineering design changes on safety-related plant structures, systems, and components. Procedure 81DP-0EE10, Appendix G, Step G.1.1.7 requires that systems modified under the advanced work authorization process not be placed back into service until after the approval of the engineering work order. Contrary to the above, prior to June 17, 2016, systems modified under the advanced work authorization process were placed back in service prior to the approval of the engineering work order. Specifically, the inspectors identified three examples in which the licensee placed systems modified under the Advance Work Authorization process back in service prior to the approval of the associated engineering work order. The licensee entered this condition into their corrective action program as Condition Report 16-09965. The immediate corrective action taken to restore compliance was to require design engineering director approval prior to the development of all advance work authorizations to ensure the in service point is clearly identified and understood. Because this violation was of very low safety significance and the licensee entered the issue into their corrective action program, this violation was treated as an NCV, consistent with Section 2.3.2.a of the NRC Enforcement Policy: NCV 05000528, 529, 530/2016003-01, Failure to Follow Advance Work Authorization Procedure.
1R19 Post-Maintenance Testing
a. Inspection Scope
The inspectors reviewed six post-maintenance testing activities that affected risk-significant SSCs:
- August 2, 2016, Unit 1 diesel generator A following degraded generator rotor pole repair
- August 24, 2016, Unit 2 C channel plant protection system post maintenance testing after repairing a cracked card edge connector
- August 30-31, 2016, Unit 3 auxiliary feed water pump AFA-P01, post-maintenance testing of flow control valve HV-33 and feed isolation valve UV-37, following static testing, inspection, and lubrication
- August 31-September 1, 2016, Unit 1 vital 480/120 VAC voltage regulator 1E-PNC-V27, post maintenance testing following replacement of transformers, chokes, and capacitors
- September 9, 2016, Unit 1 repair and testing of pressurizer spray valve (PV-100F)
- September 11, 2016, Unit 1 NN111 electrical transfer switch testing The inspectors reviewed licensing- and design-basis documents for the SSCs 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 SSCs.
These activities constituted completion of six 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 Unit 1 and Unit 3 short notice maintenance outages that concluded on September 12, 2016, and September 23, 2016, respectively, the inspectors evaluated the licensees 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:
- Monitoring of shut-down and cool-down activities
- Review of the licensees troubleshooting plans
- Verification that the licensee maintained defense-in-depth during outage activities
- Monitoring of heat-up and startup activities These activities constituted completion of two outage activities samples, as defined in Inspection Procedure 71111.20.
b. Findings
No findings were identified.
1R22 Surveillance Testing
a. Inspection Scope
The inspectors observed six 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:
- August 5, 2016, Unit 3 atmospheric dump valve 179 inservice test
- August 15, 2016, Unit 3 high pressure safety injection pump B inservice test
- August 25, 2016, Unit 2 containment spray pump B inservice test
- August 29, 2016, Unit 3 high pressure safety injection pump A inservice test
- September 30, 2016, Unit 3 spray pond manual valve inservice test Other surveillance tests:
- August 3, 2016, Unit 2 radiation monitoring quarterly functional test A 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 SSCs following testing.
These activities constituted completion of six surveillance testing inspection samples, as defined in Inspection Procedure 71111.22.
b. Findings
No findings were identified.
Cornerstone: Emergency Preparedness
1EP4 Emergency Action Level and Emergency Plan Changes
a. Inspection Scope
The inspector performed an in-office review of Palo Verde Emergency Plan, Revision 57.
This revision included updates to organizational titles for offsite response organizations, population, and demographic updates to data and tables, and numerous editorial changes.
This revision was compared to its previous revision, 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, 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 revision did not reduce 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, this revision is subject to future inspection.
These activities constitute completion of one emergency action level and emergency plan changes sample as defined in Inspection Procedure 71114.04.
b. Findings
No findings were identified.
1EP6 Drill Evaluation
Training Evolution Observation
a. Inspection Scope
On August 2, 2016, the inspectors observed simulator-based licensed operator requalification training that included implementation of the licensees emergency plan.
The inspectors verified that the licensees emergency classifications, off-site notifications, and protective action recommendations were appropriate and timely.
The inspectors verified that any emergency preparedness weaknesses were appropriately identified by the evaluators and entered into the corrective action program for resolution.
These activities constituted completion of one training observation sample, as defined in Inspection Procedure 71114.06.
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 Reactor Coolant System Specific Activity (BI01)
a. Inspection Scope
The inspectors reviewed the licensees reactor coolant system chemistry sample analyses for the period of July 1, 2015, through September 30, 2016, to verify the accuracy and completeness of the reported data. The inspectors observed a chemistry technician obtain and analyze a reactor coolant system sample on September 21, 2016.
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 reactor coolant system specific activity performance indicator for Units 1, 2, and 3, as defined in Inspection Procedure 71151.
b. Findings
No findings were identified
.2 Reactor Coolant System Total Leakage (BI02)
a. Inspection Scope
The inspectors reviewed the licensees records of reactor coolant system total leakage for the period of July 1, 2016, through September 30, 2016, to verify the accuracy and completeness of the reported data. The inspectors observed the performance of a reactor coolant system water inventory determination in accordance with station procedure 40ST-9RC02 on August 17, 2016. 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 reactor coolant system leakage performance indicator for Units 1, 2, and 3, as defined in Inspection Procedure 71151.
b. Findings
No findings were identified.
4OA2 Problem Identification and Resolution
Routine Review
a. Inspection Scope
Throughout the inspection period, the inspectors performed daily reviews of items entered into the licensees corrective action program and periodically attended the licensees 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 licensees problem identification and resolution activities during the performance of the other inspection activities documented in this report.
b. Findings
No findings were identified.
4OA3 Follow-up of Events and Notices of Enforcement Discretion
.1 NRC Event Number 52226: Manual Reactor Trip due to Stuck-Open Pressurizer Spray
Valve On September 7, 2016, at 9:32 p.m., Unit 1 initiated a manual reactor trip due to a stuck-open pressurizer spray valve. Prior to the trip, Unit 1 was at full power and establishing maintenance conditions by transferring a 120 VAC non-class instrument distribution panel to its alternate power supply. The load transfer attempt failed. At that time, control room operators noted that the main spray valve was open and attempted to close the main spray valve in accordance with alarm response procedures. Operators manually tripped the reactor when their attempts to close the main spray valve were unsuccessful. Additionally, operators secured all reactor coolant pumps to terminate spray flow to the pressurizer. No emergency safety features actuations occurred. The offsite power grid was stable. Unit 1 stabilized at normal operating temperature and pressure in Mode 3 with all reactor coolant pumps secured and on natural circulation.
Plant pressure was maintained via pressurizer heaters and auxiliary spray.
The inspectors responded to the event to verify that all safety systems responded as expected, no emergency classification was required per the emergency plan, no further Agency response was warranted, and no findings were identified.
These activities constitute completion of one event follow-up sample, as defined in Inspection Procedure 71153.
.2 Unit 3 Automatic Main Turbine Trip and Subsequent Manual Reactor Trip at Operator
Discretion On September 19, 2016, at 2:34 p.m., Unit 3 experienced a main turbine trip and a subsequent reactor power cutback to approximately 40 percent power. The control room operators carried out actions in accordance with the Load Rejection Abnormal Operating Procedure. The operations crew commenced a controlled reactor down power to approximately 12 percent in accordance with station procedures. At 3:54 p.m. the licensee decided to perform a manual reactor shutdown, and operators initiated a manual trip at 35 percent power. The decision to shutdown was based on the reactor core dynamics that would be expected to occur following a large power change at the end of the operating cycle. The inspectors responded to the Unit 3 control room following the main turbine trip. The inspectors monitored plant parameters and verified the operators actions were conducted in accordance with station procedures and Technical Specifications.
Following the manual trip, the inspectors verified that all safety systems responded as expected. The inspectors review determined that no further Agency actions were required, and no findings were identified.
These activities constitute completion of one event follow-up sample, as defined in Inspection Procedure 71153.
4OA5 Other Activities
(Closed) Unresolved Item (URI) 07200044/2015001-01, VCT Lifting Brackets Compliance with ANSI N14.6 The licensee operates its ISFSI under a general license as allowed in 10 CFR Part 72.
Palo Verde utilizes the NAC, Universal Multi-Purpose Canister System (UMS),approved under Certificate of Compliance No. 1015, Amendment 5 and the NAC UMS Final Safety Analysis Report (FSAR), Revision 10 to load and store spent fuel at their Independent Spent Fuel Storage Installation (ISFSI). To place a loaded cask at the ISFSI, the licensee utilized a Vertical Cask Transporter (VCT) and lift brackets to lift and transport the cask. During Palo Verdes last routine ISFSI inspection conducted on February 22-26, 2015, an Unresolved Item (URI) was identified and documented in NRC ISFSI Inspection Report 05000528/2015007, 05000529/2015007, 05000530/2015007, and 07200044/2015001 (ML15086A433).
The NAC FSAR Table 1.5-1, stated devices for lifting the systems that are important to safety must comply with American National Standards Institute (ANSI) N14.6. The licensees annual maintenance records of the VCT lift brackets were reviewed by the inspector during the routine inspection. The licensees maintenance practices were found to not fully comply with the requirements of ANSI N14.6, Section 6.3.2 for the annual inspection of the lifting devices. The licensee had been performing annual visual inspections, but neither dimensional nor non-destructive examination inspections were being performed. Several of the special lifting devices were discussed in detail in the FSAR, including the transfer cask lift yoke and the lift trunnions. However, the VCT lift brackets were not discussed in the FSAR and it was unclear if the FSAR Table 1.5-1 statement applied to the VCT lift brackets. The FSAR section 8.1.3 allowed a license to utilize a mobile lifting frame (or transporter), when the lift height is limited to 24 inches and the mobile lifting device followed the requirements of Section A5.6.c of the Technical Specifications. The NAC FSAR does not specifically address the type of mobile lifting device or the lift brackets that could be used to lift the cask and transport it to the ISFSI.
Use of Palo Verdes VCT and lift brackets were implemented through the licensees 10 CFR 72.48 and 10 CFR 72.212 evaluation processes. The NRC inspectors opened an URI and conferred with NRC Headquarters Division of Spent Fuel Management (DSFM) for clarification.
After an in-office review and consultation with DSFM staff, the inspectors concluded that annual maintenance of the VCT lift brackets in accordance with the ANSI N14.6 standard was not required. Per the NAC Technical Specification A5.6.c, the VCT lift brackets were only required to be designed in accordance with NUREG 0612, Control of Heavy Loads, at Nuclear Power Plants, which invoked the ANSI N14.6 standard. The Technical Specification did not specifically require the lifting brackets to be maintained and inspected in accordance with the NUREG-0612/ANSI N14.6 requirements.
The NRC inspector noted that the VCT lift brackets were visually inspected annually, designed in accordance with the ANSI N14.6 requirements, and were originally load tested prior to first use, in accordance with the standard. The NAC UMS FSAR Section 11.2.4 documented that the Vertical Concrete Cask was analyzed for a 24-inch drop onto a concrete storage pad. The analysis concluded that a 24-inch drop would not result in a breach of the canister, nor result in damage to the canister or concrete cask that would affect the cask systems design functions. Palo Verdes loading procedures did require the VCT transportation operations to limit the casks lift height to below 24 inches in accordance with the drop-analysis and Technical Specification Table A5-1 requirements.
The inspectors concluded that no performance deficiency existed. No additional deficiencies were identified during review of this URI.
Unresolved item 07200044/2015001-01 is closed.
4OA6 Meetings, Including Exit
Exit Meeting Summary
On August 25, 2016, the inspectors conducted a telephonic exit meeting to present the results of the in-office review of Unresolved Item URI 07200044/2015001-01 to Mr. M. McGhee, Regulatory Affairs Department Leader, and other members of the licensees staff. The licensee acknowledged the information presented. The inspectors asked the licensee whether any materials examined during the inspection should be considered propriety. No propriety information was identified.
On September 21, 2016, the inspector presented the results of the in-office inspection of the emergency preparedness plan change, to Mr. R. Davis, Director, Nuclear Security and 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 October 11, 2016, the inspectors presented the inspection results to Ms. M. Lacal 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.
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
- J. Cadogan, Sr. Vice President of Nuclear Operations
- M. Lacal, Sr. Vice President of Regulatory and Oversight
- B. Rash, Vice President, Engineering
- C. Kharrl, Plant General Manager for Operations
- M. McLaughlin, Plant General Manager of Site Support
- H. Ridenour, Director Maintenance
- G. Andrews, Director Regulatory Affairs
- R. Davis, Director, Nuclear Security and Emergency Preparedness
- D. Wheeler, Director Performance Improvement
- K. Graham, Director Plant Engineering
- K. House, Director Design Engineering
- C. Moeller, Director, Technical Support (Acting)
- J. Fearn, Manager, Emergency Preparedness
- R. Routolo, Manager, Radiation Protection (Acting)
- M. McGhee, Department Leader, Nuclear Regulatory Affairs
- M. Radspinner, Department Leader, System Engineering
- D. Elkinton, Section Leader, Compliance
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened and Closed
- 05000528,529,530/ NCV Failure to Follow Advance Work Authorization Procedure 2016003-01 (Section 1R18)
Closed
200044/2015001-01 URI VCT Lift Brackets Compliance with ANSI N14.6 (Section 4OA5)
Attachment