IR 05000528/2019011
| ML19130A127 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 05/10/2019 |
| From: | Vincent Gaddy Region 4 Engineering Branch 1 |
| To: | Bement R Arizona Public Service Co |
| Gaddy V | |
| References | |
| IR 2019011 | |
| Download: ML19130A127 (38) | |
Text
SUBJECT:
PALO VERDE NUCLEAR GENERATING STATION, UNITS 1, 2, AND 3 -
DESIGN BASES ASSURANCE (TEAMS) INSPECTION REPORT 05000528/2019011, 05000529/2019011, AND 05000530/2019011
Dear Mr. Bement:
On April 17, 2019, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Palo Verde Nuclear Generating Station, Units 1, 2, and 3, and discussed the final results of this inspection with Mrs. M. Lacal and other members of your staff. The results of this inspection are documented in the enclosed report.
NRC inspectors documented three findings of very low safety significance (Green) in this report.
These findings involved violations of NRC requirements. The NRC is treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.
If you contest the violations or significance of these NCVs, 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; 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 U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region IV; and the NRC resident inspector at Palo Verde Nuclear Generating Station. This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with 10 CFR 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely,
/RA/
Vincent G. Gaddy, Chief Engineering Branch 1 Division of Reactor Safety Dockets: 50-528, 50-529, and 50-530 Licenses: NPF-41, NPF-51, and NPF-74 Enclosure:
Inspection Report 05000528/2019011, 05000529/2019011, and 05000530/2019011 w/attachment: Additional Document and Inspection Requests
U.S. NUCLEAR REGULATORY COMMISSION INSPECTION REPORT Docket Numbers: 05000528, 05000529, and 05000530 License Numbers: NPF-41, NPF-51, and NPF-74 Report Numbers: 05000528/2019011, 05000529/2019011, and 05000530/2019011 Enterprise Identifier: I-2019-011-0009 Licensee: Arizona Public Service Company Facility: Palo Verde Nuclear Generating Station Location: Tonopah, Arizona Onsite Inspection Dates: February 11, 2019, to March 01, 2019 Exit Date: April 17, 2019 Inspectors: G. George, Senior Reactor Inspector, Team Lead J. Braisted, Reactor Inspector S. Hedger, Emergency Preparedness Inspector N. Okonkwo, Reactor Inspector R. Deese, Senior Reactor Analyst S. Gardner, Contractor, Beckman and Associates J. Zudans, Contractor, Beckman and Associates Approved By: Vincent G. Gaddy, Chief Engineering Branch 1 Division of Reactor Safety Enclosure
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting a standalone report inspection at Palo Verde Nuclear Generating Station, Units 1, 2, and 3, in accordance with the Reactor Oversight Process. The Reactor Oversight Process is the NRCs program for overseeing the safe operation of commercial nuclear power reactors. Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information. Findings and violations being considered in the NRCs assessment are summarized in the table below.
List of Findings and Violations Failure to Evaluate Unit 1 MSIV Actuator Trains for 50.65(a)(1) Status in the Maintenance Rule Program Cornerstone Significance Cross-cutting Report Aspect Section Barrier Green [H.13] - 71111.21M Integrity NCV 05000528/2019011-01 Consistent Closed Process The inspectors identified a Green, non-cited violation of 10 CFR 50.65 (a)(1) for the failure to monitor the performance of Unit 1 main steam isolation valve (MSIV) actuators and failure to take corrective action when the performance did not meet established performance goals.
Failure to Correct MSIV Actuator Hydraulic Fluid Viscosity Sensitivity to Low Temperature Cornerstone Significance Cross-cutting Report Aspect Section Barrier Green [H.12] - Avoid 71111.21M Integrity NCV 05000528/2019011-02 Complacency NCV 05000529/2019011-02 NCV 05000530/2019011-02 Closed The inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B,
Criterion XVI, Corrective Action, for the licensees failure to correct a condition adverse to quality associated with a known latent condition of the MSIV actuator hydraulic oil viscosity sensitivity to low temperature.
Failure to Adequately Manage the Risk Involved with Various Risk-Significant Maintenance Windows Cornerstone Significance Cross-cutting Report Aspect Section Mitigating Green [H.1] - 71111.21M Systems NCV 05000528/2019011-03 Resources NCV 05000529/2019011-03 NCV 05000530/2019011-03 Closed The inspectors identified a Green, non-cited violation of 10 CFR 50.65 (a)(4) for the licensees failure to complete adequate risk assessments prior to performing maintenance activities because the probabilistic risk assessment (PRA) tools were not maintained such that its representation of the as-built, as-operated plant is sufficient to support the applications for which it is being used.
INSPECTION SCOPES
Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html.
Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.
REACTOR SAFETY
===71111.21M - Design Bases Assurance Inspection (Teams)
The inspectors evaluated the following components and listed applicable attributes, permanent modifications, and operating experience:
Design Review - Large Early Release Frequency (LERF) (IP Section 02.02) ===
From February 11, 2019, to March 1, 2019, the team inspected the following large-early-release-frequency components.
- (1) Containment Isolation Actuation System
- Component maintenance history and corrective action program reports to verify the monitoring of potential degradation
- Calculations for timing of equipment actuation
- Procedures for response time testing and inservice tests to determine consistency between the updated final safety analysis report, calculations, and testing acceptance criteria
- Component maintenance history, system health reports, and corrective action program reports to verify the monitoring of potential degradation
- Preventive maintenance bases to ensure activities conform to vendor requirements
- Inservice testing and system-level design bases documents
- Procedures for full-stroke open/close inservice testing
- Completed surveillance tests to ensure acceptance criteria have been met
- Trend data to ensure monitoring of potential degradation
- Valve stoke time adjustment evaluation to account for differences in static and dynamic testing
- Calculations for main steam line air volume and air accumulator pressure drop operability criteria Design Review - Risk-Significant/Low Design Margin Components (IP Section 02.02)
(5 Samples 1 Partial)
From February 11, 2019, to February 28, 2019, the team inspected the following components and listed applicable attributes.
- (1) 4160VAC Bus (PBA-S03) and Control Relays
- Component maintenance history and corrective action program reports to verify the monitoring of potential degradation
- Calculations for electrical distribution, system load flow/voltage drop, short-circuit, and electrical protection to verify that bus capacity and voltages remained within minimum acceptable limits
- Component walkdown and vendor document review to verify installed configuration, specifications and acceptance criteria, and design bases functions
- (2) Train A Emergency Safety Features Service Transformer S03
- Component maintenance history and corrective action program reports to verify the monitoring of potential degradation
- Component walkdown and vendor document review to verify installed configuration, specifications and acceptance criteria, and design bases functions
- Calculations for transformer, system load flow/voltage drop, short-circuit, and electrical protection to verify that transformer loading capacity and voltages remained within minimum acceptable limits
- (3) High Pressure Safety Injection B Long-Term Recirculation Isolation Valve (13JSIBHV0609)
- Completed work orders for activities such as static testing and electrical inspection, visual inspection, grease sampling, and stem lubrication
- Inservice testing bases and system-level design bases documents
- Procedures for full-stroke open/close inservice testing
- Component maintenance history, system health reports, and corrective action program reports to verify the monitoring of potential degradation
- Calculations for motor-operated valve torque and thrust, valve weak-link analysis, and maximum operating differential pressures, line pressures, temperature, and flow
- (4) High Pressure Safety Injection Pump (2MSIBP02) and Refueling Water Storage Tank
- Component maintenance history, system health reports, testing results, and corrective action program reports to verify the monitoring of potential degradation
- Calculations for design of component, seismic analyses, net positive suction head for the pump, hydraulic piping analyses, including runout conditions, refueling water storage tank capacity and capabilities to verify these components can support accident mitigation requirements
- Design bases documents, updated final safety analysis report, technical specifications and bases to assure licensing bases match component capabilities
- (5) Reactor Trip Circuit Breakers
- Component maintenance history and corrective action program reports to verify the monitoring of potential degradation
- Calculations for timing of equipment actuation in series with reactor protection system (RPS)
- Procedures for circuit breaker inspection and testing to compare maintenance practices against industry and vendor guidance
- Corrective actions associated with a non-cited violation involving reactor trip circuit breakers as documented in the 2009 Component Design Basis inspection report, (ADAMS Accession Number ML093240524)
- (6) Evaluation of Operator Procedures and Actions Related to Components (Partial Sample)
- Control room operator actions resulting from a simulated break in the letdown system piping. From the receipt of associated alarms, actions to isolate the leak are completed within 10 minutes.
- Control room operator actions resulting from a feedwater line break to the steam generators. Actions to restore feedwater to an unaffected steam generator are completed within 20 minutes.
- Control room operator actions resulting from a steam generator tube rupture as described in the updated final safety analysis report. Actions to feed the affected steam generator up to a level to cover its tubes is completed within 46 minutes, 19 seconds.
- Control room and auxiliary operator actions to restore feedwater to a steam generator using the main feedwater system after a loss of instrument air and its backup nitrogen system a. Auxiliary operators manually open feedwater downcomer isolation valves within 35 minutes.
b. Control room operators, in coordination with auxiliary operators, restore feedwater to a steam generator within 46 minutes.
Modification Review - Permanent Modifications (IP Section 02.03) (7 Samples)
From February 11, 2019, to March 1, 2019, the team inspected the following permanent modifications.
- (1) Design Equivalent Change DEC-00149, Replacement Spring Packs for Limitorque Actuators, Revision 2
- (2) Design Equivalent Change DEC-00945, Allow the use of filler metal welding of tube plugs inside the Shutdown Cooling Heat Exchangers (SCHEs), Revision 0
- (3) Engineering Change SI-1081, High Pressure Safety Injection (HPSI) Pump Bearing Isolator and Local Level Site Glass Modification
- (4) Design Equivalent Change DEC-00557, "Fuse Addition to the PS1, PS2, and PS3 Power Supplies Internal to MSFIS Logic Cabinets 1/2/3J-SGA/B-C01 to Enhance Existing Overcurrent Protection"
- (5) Design Equivalent Change DEC-01136, Stiffening of High Pressure Safety Injection Pump 2MSIBP02 Support Pedestal, Revision 2
- (6) Engineering Change PB-1655, Degraded Voltage Relay Modification, Revision 0
- (7) Design Equivalent Change DEC-01035, Replacement of Obsolete TEC Molded Case Circuit Breakers in Q Class Application" Review of Operating Experience Issues (IP Section 02.06) (3 Samples)
From February 11, 2019, to March 1, 2019, the team inspected the following operating experience issues.
- (1) NRC Information Notice 2006-31, Inadequate Fault Interrupting Rating of Breakers
- (2) Westinghouse Nuclear Safety Advisory Letter (NSAL) 12-4, Potential Non-Conservatism in COLSS/CPCS Setpoints Analyses, June 20, 2012
- (3) NRC Information Notice 2012-14, Motor-Operated Valve Inoperable Due to Stem-Disc Separation
INSPECTION RESULTS
Failure to Evaluate Unit 1 MSIV Actuator Trains for 50.65(a)(1) Status in the Maintenance Rule Program Cornerstone Significance Cross-cutting Report Aspect Section Barrier Integrity Green [H.13] - 71111.21M NCV 05000528/2019011-01 Consistent Closed Process The inspectors identified a Green, non-cited violation of 10 CFR 50.65 (a)(1) for the failure to monitor the performance of Unit 1 main steam isolation valve (MSIV) actuators and the failure to take corrective action when the performance did not meet established performance goals.
Description:
The function of the main steam isolation valves (MSIVs) is to assure the integrity of the containment by fast closing to prevent an inadvertent radioactive release caused by a faulted steam generator. Each MSIV is supported by two physically separate and electrically independent solenoid actuators to provide redundant means of valve operation. Technical Specification 3.7.2, Main Steam Isolation Valves, requires Four MSIVs and their associated actuator trains shall be OPERABLE, in the applicable modes 1, 2, 3, and 4 except when valves are in the closed position. To verify system operability, the licensee performs Surveillance Requirement 3.7.2.1, which requires, at entry into Mode 3, Verify closure time of each MSIV is within limits with each actuator train on an actual or simulated activation signal.
The inspectors reviewed the surveillance requirement verification of closure times associated with the Unit 1 MSIVs with each actuator train. Following the 1R19 refueling outage on May 12, 2016, MSIV 180A actuator train and MSIV 181A actuator train failed to close their supported MSIVs. During the subsequent 1R20 refueling outage on November 4, 2017, MSIV 180A actuator train failed to close its supported MSIV within acceptable closure time, and on November 5, 2017, MSIV 171A actuator train failed to close its supported MSIV.
Within an 18-month period, there were four failures of the actuator trains to perform their safety functions to close MSIVs within limits, as required by Surveillance Requirement 3.7.2.1.
Two of those failures were during tests of actuator MSIV 180A.
To comply with 10 CFR 50.65, the licensee implements Procedure 70DP-0MR01, Maintenance Rule, Revision 44. The procedure implements the licensees commitment to NRC Regulatory Guide 1.160, Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, Revision 3, which endorses NUMARC 93-01, Revision 4A, Industry Guideline for Monitoring Effectiveness of Maintenance at Nuclear Power Plants.
Section 9.4 of NUMARC 93-01 establishes the guidance for monitoring and goal setting of structures, systems, and components (SSCs), at the plant, system, train, or component level.
Section 9.4 states, The basis for technical specification, licensing commitments, and other regulation may be appropriately used for goal setting. For train level monitoring, Section 9.4.1.2, states, Risk-significant systems and standby systems that have redundant trains should have goals established for the individual trains. Furthermore, Train level goals provide a method to address degraded performance of a single train even though the system function is still available.
Palo Verde Nuclear Generating Station establishes the requirements for goals, monitoring, and reliability of the main steam system using, Maintenance Rule Performance Criteria Basis Worksheet for Main Steam, Revision 4. The steam generator function, SG-05, performance criteria is described as High Risk-Significant with the stated system function of, Isolating faulted steam generator, steam line, feedwater line, or non-class BOP equipment.
Monitoring for reliability, availability, and condition monitoring are established on a train level with the trains consisting of the physically separate solenoid actuators. The SG-05 performance criteria for MSIVs identifies a maintenance rule functional failure (MRFF) as an, Actuator train or valve fail to close on MSIS [main steam isolation signal]. The worksheets Unreliability Statement states, Performance is unacceptable if there are 3 or more MRFFs (failure to close) per 18 months per valve. Performance is unacceptable if there are 1 or more MRFFs (spurious closure). The worksheet establishes a Trigger Value of 2 failures to close.
The inspectors determined that the licensee failed to identify four actuator train failures of MSIVs 171A, 180A (twice), and 181A as MRFFs in accordance with the Palo Verde Nuclear Generating Station maintenance rule program Procedure 70DP-0MR01. The inspectors identified that the licensee personnel failed to meet Procedure 70DP-0MR01, Step 4.5.2.6.
Specifically, licensee personnel failed to identify whether the actuator failures were MRFFs; failed to determine if performance criteria were exceeded or another type of 10 CFR 50.65 (a)(1) issue had occurred; ensure SSCs were quarantined if required; and evaluate the MRFF to identify corrective actions. Additionally, for the two failures of actuator train MSIV 180A, licensee personnel failed to implement Step 4.5.1.6 to consider implementing the 10 CFR 50.65 (a)(1) process for a significant declining trend in performance of the SSC for a condition that results in exceeding a performance criteria trigger point.
Corrective Action: In response to this issue, the licensee entered the issue into the corrective action program to evaluate whether placement of the MSIVs and actuator trains warranted entry into 10 CFR 50.65(a)(1) status. This finding does not represent an immediate safety concern condition.
Corrective Action Reference: Condition Report 19-02987
Performance Assessment:
Performance Deficiency: The licensees failure to identify four MSIV actuator train failures as maintenance rule functional failures and monitor those failures against reliability performance criteria in accordance with Procedure 70DP-0MR01 was a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the SSC and Barrier Performance attribute of the Barrier Integrity cornerstone. Specifically, the failure to identify and monitor maintenance rule functional failures against reliability performance criteria resulted in the licensee not recognizing the ineffectiveness of the maintenance and taking adequate corrective actions to restore MSIVs to acceptable reliability goals.
Significance: The inspectors assessed the significance of the finding using Inspection Manual Chapter 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At -
Power Situations. Using Exhibit 3, Barrier Integrity Screening Questions, the inspectors determined the finding to be of very low safety significance (Green) because the finding does not represent an actual open pathway in the physical integrity of reactor containment or does not involve a reduction in function of hydrogen igniters in the reactor containment.
Cross-cutting Aspect: H.13 - Consistent Process: Individuals use a consistent, systematic approach to make decisions. Risk insights are incorporated as appropriate. This finding had a human performance cross-cutting aspect, associated with consistent process, because the licensee failed to use a consistent, systematic approach to make decisions. Specifically, individuals failed to demonstrate an understanding of the maintenance rule decision making process and use it consistently to monitor the reliability of the MSIVs.
Enforcement:
Violation: Title 10 CFR 50.65 (a)(1) requires, in part, that each holder of an operating license for a nuclear power plant shall monitor the performance or condition of SSCs, against licensee-established goals, in a manner sufficient to provide reasonable assurance that these SSCs are capable of fulfilling their intended functions. These goals shall be established, commensurate with safety and, where practical, take into account industrywide operating experience. When the performance or condition of a SSC does not meet established goals, appropriate corrective action shall be taken.
Contrary to the above, from May 12, 2016, to November 5, 2017, the licensee failed to monitor the performance or condition of SSCs, against licensee-established goals, in a manner sufficient to provide reasonable assurance that these SSCs are capable of fulfilling their intended functions. When the performance did not meet established goals, appropriate action was not taken. Specifically, the licensee failed to evaluate the MSIV actuator train failures for (a)(1) status after four Unit 1 MSIV actuator trains failed SR 3.7.2.1 in an 18-month period.
Additionally, the licensee did not take any corrective action to ensure actuator trains would be capable of meeting their intended functions.
Enforcement Action: This violation is being treated as a Non-Cited Violation, consistent with Section 2.3.2 of the Enforcement Policy.
Failure to Correct MSIV Actuator Hydraulic Fluid Viscosity Sensitivity to Low Temperature Cornerstone Significance Cross-cutting Report Aspect Section Barrier Integrity Green [H.12] - Avoid 71111.21M NCV 05000528/2019011-02 Complacency NCV 05000529/2019011-02 NCV 05000530/2019011-02 Closed The inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to correct a condition adverse to quality associated with a known latent condition of the main steam isolation valve (MSIV)actuator hydraulic oil viscosity sensitivity to low temperature.
Description:
The function of the MSIVs is to assure the integrity of the containment by fast closing to prevent an inadvertent radioactive release. Each MSIV is supported by two physically separate and electrically independent solenoid actuators to provide redundant means of valve operation. Technical Specification 3.7.2, Main Steam Isolation Valves, requires, Four MSIVs and their associated actuator trains shall be OPERABLE, in the applicable modes 1, 2, 3, and 4 except when valves are in the closed position. To verify system operability, the licensee performs Surveillance Requirement 3.7.2.1, which requires, at entry into Mode 3, to verify closure time of each MSIV is within limits with each actuator train on an actual or simulated activation signal.
While in Mode 3 during the 1R20 refueling outage on November 4, 2017, MSIV 180A actuator train failed to close its supported MSIV within acceptable closure time and on November 5, 2017, MSIV 171A actuator train failed to close its supported MSIV. Following the failures, the licensee completed an Engineering Evaluation 17-16133-005 to determine the cause and recommend corrective actions for the actuator train failures. The results of the evaluation stated, it can be concluded that the primary source of the MSIV stroke time issues (specifically MSIV 171A and 180A from 1R20) is a combination of low temperature (lower than vendor recommended limits), tolerances within the 4-way affected by this low temperature, and high viscosity of the oil. When these mechanisms are combined with not actuating the 4-ways after returning them to their normal operating band, the 4-ways should not be expected to shuttle appropriately. The licensees corrective action recommendation was to add a contingency, require Operations to contact engineering on initial slow strokes or failure to stroke and allow Operations to perform a slow stroke of the M 4-way in order to rectify the conditions created by low temperature effects experienced during outages Engineering Evaluation 17-16133-005 cited a previous 2006 evaluation that determined that the valve closed stroke time is directly dependent on the flow rate of the hydraulic fluid from the bottom of the actuator cylinders. The flow rate becomes slower with a higher viscosity attributed to colder temperature. This was supported by documentation that a feedwater isolation valve (FWIV) (with Fyrquel hydraulic oil) with stroke time of 9.31 seconds was heated for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to 85 F, retested, then had a stroke time of 7.79 seconds. This is consistent with the manufacturers Safety Data Sheet, Fyrquel 220MLT, that states, good flow rates are between 80 F and 100 F. The 2006 evaluation, CRDR 2856266, recommended performing the surveillance when accumulator temperature is greater than 60 F or heat the valves until the oil temperature is above this minimum and then perform the stroke testing. In 2006 the licensees corrective action was to place a caution in the MSIV test Procedure 73ST-9SG01, MSIV - Inservice Test, which stated, Stroke timing MSIVs when accumulator temperature is less than 60 F may adversely affect test results.
The inspectors determined that the licensees corrective action, to add a caution statement to the surveillance test procedure, did not correct the condition in which low oil temperature would cause adverse surveillance test results. This known latent condition was not corrected since discovered in 2006. Additionally, the licensee did not take corrective actions to ensure MSIV actuator trains would perform their intended design function prior to the plant entering the mode of applicability (Mode 4 when MSIVs are open) when tests failures occurred in November 2017. While the MSIVs were open in Mode 4, the latent condition did not ensure the operability of all eight MSIV actuator trains were maintained. Finally, following repeat November 2017 failures caused by the known latent condition, the licensee failed to reevaluate the original corrective actions and identify additional corrective actions to maintain MSIV actuator train design function.
Corrective Action: In response to this issue, the licensee initiated corrective actions to revise their mode change checklist procedure to prevent entry into the mode of applicability when MSIV actuator hydraulic oil would be below minimum temperature. This finding does not represent an immediate safety concern condition.
Corrective Action Reference: Condition Report 19-03044
Performance Assessment:
Performance Deficiency: The licensees failure to correct a condition adverse to quality, identified in 2006, to assure that the MSIV actuator trains would satisfactorily function in all applicable modes as required by 10 CFR Part 50, Appendix B, Criterion XVI, was a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, the failure to correct hydraulic oil viscosity sensitivity to low temperature would continue to cause repeat failures of MSIV actuator surveillance tests, would not ensure MSIV actuator trains remained operable during Mode 4, and would not ensure operability when changing to Mode 3 to complete surveillance tests.
Significance: The inspectors assessed the significance of the finding using Inspection Manual Chapter 0609, Appendix H, Containment Integrity SDP. Using Figure 4.1, the inspectors determined that the finding would not affect core-damage-frequency (CDF), and would affect large-early-release-frequency (LERF), therefore this was a Type B finding. Using Figure 7.2, the finding screens as having very low safety significance (Green) because the condition would only occur when the plant is in an outage for greater than 8 days.
Cross-cutting Aspect: H.12 - Avoid Complacency: Individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Individuals implement appropriate error reduction tools. This finding had a human performance cross-cutting aspect, associated with avoid complacency, because individuals failed to recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Specifically, the licensee failed to recognize and address a known latent condition identified in 2006, which subsequently caused MSIV actuator surveillance test failures in November 2017.
Enforcement:
Violation: Title 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requires, in part, conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected.
Contrary to the above, from 2006 to February 28, 2019, the licensee failed to correct a condition adverse to quality. Specifically, the licensee failed to correct a known latent condition of MSIV actuator hydraulic oil viscosity sensitivity to low temperature.
Enforcement Action: This violation is being treated as a Non-Cited Violation, consistent with Section 2.3.2 of the Enforcement Policy.
Failure to Adequately Manage the Risk Involved with Various Risk-Significant Maintenance Windows Cornerstone Significance Cross-cutting Report Aspect Section Mitigating Systems Green [H.1] - 71111.21M NCV 05000528/2019011-03 Resources NCV 05000529/2019011-03 NCV 05000530/2019011-03 Closed The inspectors identified a Green, non-cited violation of 10 CFR 50.65 (a)(4) for the licensees failure to complete adequate risk assessments prior to performing maintenance activities because the probabilistic risk assessment (PRA) tools were not maintained such that its representation of the as-built, as-operated plant is sufficient to support the applications for which it is being used.
Description:
The inspectors observed operators performing actions via the control room simulator and job performance measures (JPM) related to a scenario involving restoration of feedwater using the main feedwater pumps during a loss of instrument air and nitrogen. This scenario, documented as basic event 1AF-FWIV-2HR in PRA document 13-NS-B062 (Revision 13, November 16, 2016), has a risk achievement worth (RAW) value of 9.434, and is included in the licensees online risk model. The assumed total Human Error Probability (HEP) was 9.07E-03. The following time assumptions, based on table-top discussions with operators, were made about how long it took operators to perform certain actions:
- Time from reactor trip until an auxiliary operator is dispatched to manually operate downcomer isolation valves - 11 minutes
- Time for auxiliary operator to manually open downcomer isolation valves - 35 minutes
- Time to complete actions - 46 minutes
- Time limit to restore main feedwater flow to a steam generator to prevent core damage
- 75 minutes
- Time margin to complete the actions - 29 minutes Observation of operating crews in the control room simulator on February 27, 2019, showed that it took 27 minutes, 25 seconds from reactor trip to dispatch an auxiliary operator to manually operate downcomer isolation valves. During JPM performances on February 14 and 26, 2019, the inspectors confirmed that the time assumption associated with auxiliary operators manually opening downcomer isolation valves was valid.
With these observations, the inspectors asked the licensee if any time measurement data for this basic event was recorded in the past. On June 15, 2015, two operating crews were evaluated in the simulator performing actions associated with this basic event. The licensees records indicate that the two operating crews took 16 minutes and 23 minutes, respectively, from reactor trip to dispatch an auxiliary operator to manually operate downcomer isolation valves. On March 13, 2019, the licensee also evaluated another operating crew performing this basic event which took 22 minutes to reach the point where an auxiliary operator was dispatched. Averaging these times with those measured February 27, 2019, results in 22 minutes, 6 seconds to dispatch the auxiliary operator to manually operate downcomer isolation valves.
Procedure 40DP-9ZZ04, Time Critical Action (TCA) Programs, Section 4.4.8, required generating a condition report for an engineering review, additional validations of the action, and evaluation for a degrading TCA completion time, if a measured TCA exceeded 80 percent of its required action time. In this case, 60 minutes equals 80 percent of the required action time for basic event 1AF-FWIV-2HR. The inspectors inquired as to what licensee actions were taken after June 15, 2015, to evaluate the time assumptions associated with the basic event 1AF-FWIV-2HR. These results were evaluated in conjunction with August 11, 2014, JPM time measurements of auxiliary operators opening the downcomer isolation valves. Job performance measure completion times indicated that auxiliary operators were taking an average of 26 minutes, 12 seconds to manually open the downcomer isolation valves. The licensee determined no action was necessary because the completion time to dispatch the auxiliary operator, in conjunction with the 2014 JPM to manually open the downcomer isolation valves, did not result in exceeding 60 minutes.
The inspectors determined that, in 2015, the licensee failed to identify that the August 11, 2014, JPM measurements recorded time for operators to align equipment to allow opening of the downcomer isolation valves, but the time needed to stroke the valve open was not included. This would result in a completion time closer to the original 35-minute time assumption to complete the action.
The licensee maintains the Palo Verde Nuclear Generating Station PRA model consistent with standard ASME RA-Sa-2009, addenda to ASME/ANS RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications. Section 1-5.4, PRA Maintenance and Upgrades, of ASME RA-Sa-2009 states that, The PRA shall be maintained and upgraded, such that its representation of the as-built, as-operated plant is sufficient to support the applications for which it is being used. It further states, Changes in PRA inputs or discovery of new information identified pursuant to 1-5.3 shall be evaluated to determine whether such information warrants PRA maintenance or PRA upgrade. Changes that would impact risk-informed decisions should be incorporated as soon as practical. Section 1-5.3, Monitoring PRA Inputs and Collecting New Information, defines new information as changes to inputs that impact operating procedures, design configuration, initiating event frequencies, system or subsystem unavailability, and component failure rates.
The inspectors determined that the licensee failed to evaluate time assumption differences identified during the 2015 performance of actions associated with basic event 1AF-FWIV-2HR, to determine whether such information warrants PRA maintenance or upgrade. Additionally, the licensee failed to maintain or upgrade the Palo Verde Nuclear Generating Station Engineering Study 13-NS-B062, At-Power PRA Study for Human Reliability Analysis, when updated in 2016 to ensure the PRA included its representation of the as-operated plant to support applications for which it is being used. Additionally, the inspectors determined that the performance TCA evaluation process in procedure 40DP-9ZZ04, Time Critical Action (TCA)
Programs, was silent on evaluation of challenges to time assumptions in PRA basic event to determine if changes in the PRA model were needed. The TCA evaluation process used in 2015 was the same process in place at the time of on-site inspection.
Using the results of the operator performance observances, the inspectors determined that the average time from reactor trip to dispatch the auxiliary operator was 24 minutes. In conjunction with the 35-minute assumption to manually operate the valve, this resulted in a change in the total time to complete basic event 1AF-FWIV-2HR from 46 minutes to 59 minutes. This reduced the time margin to restore feedwater flow to a steam generator which, consequently, increased the HEP value for the basic event from 9.07E-03 to 1.8E-01.
Using the corrected HEP value, the overall core damage frequency to Palo Verde Nuclear Generating Station PRA model increased to 7.60E-06/year, or a change in overall core damage frequency of 5.23E-06/year.
Since the overall core damage frequency increased, the inspectors and Region IV Senior Reactor Analyst evaluated the impact on the Palo Verde Nuclear Generating Station online risk model used to perform maintenance risk assessments, pursuant to 10 CFR 50.65 (a)(4) of the maintenance rule. The analyst requested plant maintenance information and incremental core damage probability (ICDP) data to determine the most-risk significant maintenance windows for safety- and non-safety related auxiliary feedwater, high and low pressure safety injection, and diesel generator systems at Palo Verde Nuclear Generating Station, Units 1, 2, and 3 for the previous 5 years. The analyst compared the original ICDP associated with the longest maintenance windows for each system to the new ICDP if the PRA model was corrected for the new HEP. The data illustrated that the largest magnitude change in ICDP was associated with a 144-hour maintenance window associated with Unit 2 diesel generator B outage in September 2017. The ICDP was originally calculated to be 2.22E-08 over the 144-hour maintenance window, with a risk increase factor of 1.56. Incorporation of the new HEP for basic event 1AF-FWIV-2HR resulted in a new ICDP of 2.25E-07 for the same maintenance window, with a risk increase factor of 2.80. Since the new risk increase factor was greater than 2 and less than 10, the PRA error resulted in the licensee unknowingly entering a Yellow risk maintenance window.
Corrective Action: The licensee entered the issue in their corrective action program. In addition, reviews of how the TCA validation process interfaces with their PRA model maintenance program have been initiated. This is not an immediate safety concern.
Corrective Action Reference: Condition Report 19-04087
Performance Assessment:
Performance Deficiency: The licensees failure to maintain and upgrade the probabilistic risk assessment, such that its representation of the as-built, as-operated plant is sufficient to support the applications for which it is being used, in accordance with ASME/ANS RA-Sa-2009, Section 1-5.4, was a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, the use of a maintained PRA with an overall elevated plant risk would have put the plant into a higher licensee-established risk category as established by 10 CFR 50.65 (a)(4), similar to example 7.e of NRC Inspection Manual Chapter 0612, Appendix E, Examples of Minor Issues.
Significance: The inspectors assessed the significance of the finding using Inspection Manual Chapter 0609, Appendix K, Maintenance Risk Assessment and Risk Management SDP. The inspectors determined that the finding was not related to risk management actions only; therefore Flowchart 1, Assessment of Risk Deficit, was used to determine the significance of the performance deficiency. Using the new ICDP for the performance deficiency, the inspectors calculated the risk deficit using changes to the licensees online risk model. The licensee calculated a new ICDP for the performance deficiency using the longest maintenance outage windows in the previous 5 years for the risk-significant auxiliary feedwater, safety injection, and diesel generator systems. The inspectors determined that the risk deficit in ICDP was 2.03E-07. This ICDP was attributed to a 144-hour Unit 2 diesel generator outage in September 2017. Since the ICDP is less the 1E-06, the finding had very low safety significance (Green).
Cross-cutting Aspect: H.1 - Resources: Leaders ensure that personnel, equipment, procedures, and other resources are available and adequate to support nuclear safety.
This finding had a human performance cross-cutting aspect, associated with resources because the licensee failed to ensure that procedures and other resources are adequate to support nuclear safety. Specifically, the TCA validation process, documented in Procedure 40DP-9ZZ04, provides a means to evaluate operator action time measurements versus licensing basis time assumptions, but doesnt provide a mechanism to evaluate measurements versus the PRA assumptions to see if additional action is warranted. Since the process and resources used in this effort today are the same that were used for PRA time measurement evaluation in 2015, this is still reflective of present performance.
Enforcement:
Violation: Title 10 CFR 50.65(a)(4), requires, in part, before performing maintenance activities (including but not limited to surveillance, post-maintenance testing, and corrective and preventive maintenance), the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities. The scope of the assessment may be limited to SSCs that a risk-informed evaluation process has shown to be significant to public health and safety.
Contrary to the above, from June 15, 2015, to March 20, 2019, the licensee failed to assess and manage the increase in risk that may result from the proposed maintenance activities.
Specifically, the licensee failed to complete adequate risk assessments prior to performing maintenance activities because the PRA tools were not maintained such that its representation of the as-built, as-operated plant is sufficient to support the applications for which it is being used.
Enforcement Action: This violation is being treated as a Non-Cited Violation, consistent with Section 2.3.2 of the Enforcement Policy.
EXIT MEETINGS AND DEBRIEFS
The inspectors confirmed that proprietary information was controlled to protect from public disclosure.
- On February 28, 2019, the inspectors presented the initial results of the Design Bases Assurance (Team) Inspection to Mr. J. Cadogan, Senior Vice President - Site Operations, and other members of the licensee staff.
- On March 15, 2019, the inspectors presented a status of the results of the Design Bases Assurance (Team) Inspection to Mr. M. Lacal, Senior Vice President - Regulatory and Oversight, and other members of the licensee staff.
- On April 17, 2019, the inspectors presented the final results to Mr. M. Lacal, Senior Vice President - Regulatory and Oversight, and other members of the licensee staff.
DOCUMENTS REVIEWED
Calculations
Number Title Revision
01ECPH0255 120 VAC Control Circuits 2
13-EC-PB-0202 4160V Degraded Voltage Relay and Loss of Voltage Relay 5
Setpoint and Calibration Calculation
13-EC-PB-0204 AC Equipment Protection (4.16KV and 6
480V): Class 1E
13-EC-PG-0303 Protection Coordination Between PG Load Centers and PH 3
Motor Control Centers with Thermal Overloads in Bypass
due to an ESFAS
13-EC-PH-0254 Motor-Operated Valve Thermal Protection 13
13-JC-CH-0209 Refueling Water Tank (RWT) Level Measurement 12
13-JC-SB-0202 Acceptance Criteria for RPS and ESFAS Response Time 20
Testing
13-JC-ZZ-0307 Motor-Operated Valve Thrust and Torque Calculation for 3
13JSIAHV0604 and 13JSIBHV0609
13-JC-ZZ-0505 Motor-Operated Valve Torque Calculation for 2
13JSIAUV0673, 13JSIAUV0674, 13JSIBUV0675, and
13JSIBUV0676
13-MC-CH-0201 Refueling Water Tank Sizing 8
13-MC-SG-0316 MSIV and FWIV Pressure Drop of Air Reservoir 10
13-MC-SG-0807 Main Steam Line Volumes 6
13-MC-SI-0215 HPSI System Performance Evaluation and 8
Surveillance Requirement Basis Calculation
13-MC-SI-0222 HPSI Hot Leg Injection MOVs - Maximum Differential 5
Pressure, Line Pressure, Temp, FLOW
13-MC-SI-0250 Safety Injection, Containment Spray, and Shutdown 1
Cooling System Pump NPSH Evaluations
13-MC-SI-0316 HPSI Flow with Hot Leg Valve Fully Open 1
Calculations
Number Title Revision
13-MC-ZZ-0217 Gate Valve Open Thrust Required During Potential 7
Pressure Locking Conditions
13-MC-ZZ-0633 Consolidation of Jet Impingement/Pipe Whip Calculations 13
13-N1105-00015 Structural Integrity and Operability Analysis of HPSI 3
Pump 4X11ca-8
73DP-9ZZ13 Motor-Operated Valve - Thrust and Torque Calculations 5
N001-2101- Borg-Warner MOV Weak Link Analysis 10
00094
TA-02-C00- Steam Generator Tube Rupture with and without Loss of 1
2004-010 Offsite Power
Drawings
Number Title Revision
01-E-MAA-002 Unit Single Line Diagram 8
01-E-PBA-001 Single Line Diagram, 4.16 KV Class 1E Power System 17
Switchgear 1E-PBA-S03
01-E-PBA-002 Single Line Diagram, 4.16 KV Class 1E Power System 16
Switchgear 1E-PBA-S04
01-E-PBB-001, Elementary Diagram, 4.16KV Class 1E Power System 7
Sh. 1 Switchgears 1E-PBA-S03 and 1E-PBB-S04 4.16KV Normal
Supply Breaker
01-E-PBB-001, Elementary Diagram, 4.16KV Class 1E Power System 7
Sh. 2 Switchgears 1E-PBA-S03 and 1E-PBB-S04 4.16KV Normal
Supply Breaker
01-E-PBF-001, Control Wiring Diagram, 4.16KV Class 1E Power System 4
Sh. 1 of 10 Switchgears 1E-PBA-S03 4.16KV Normal Supply Breaker
01-E-PBF-001, Control Wiring Diagram, 4.16KV Class 1E Power System 4
Sh. 2 Switchgears 1E-PBA-S03 4.16KV Normal Supply Breaker
01-E-SIB-040, Elementary Diagram, Safety Injection Shutdown CLG 12
Sh. 1 System, HPSI Pump Long Term CLG-Valve 1J-SIB-HV-604
Drawings
Number Title Revision
01-E-SIB-040, Elementary Diagram, Safety Injection Shutdown CLG 12
Sh. 2 System, HPSI Pump Long Term CLG-Valve 1J-SIB-HV-609
01-M-SGP-001 Main Steam System 70
01-M-SGP-002 Main Steam System 51
01-M-SGP-002 P&I Diagram Main Steam System 51
104-08-F103681 SPLA Block Diagram C
13-E018-00526 O/L DIM of MCC E-PHA-M35, Sh. 5 of 10 18
13-E-MAA-001 Main Single Line Diagram 31
13-E-ZAC-080 Conduit and Tray Symbols and Details 28
13-J-ZAF-001 Instrument Location Plan Auxiliary 14
Building EL. 40-0 Level D ZADC
13-J-ZAF-002 Instrument Location Plan Auxiliary 13
Building EL. 40 Lvl D ZADC
13-M234A-00045 Main Steam Isolation Valve with A/DV Hydraulic Actuator 17
13-M234A-00123 Schematic for Anchor/Darling Self-Contained Hydraulic 9
Actuator
13-N001-1105- Sectional Assembly - HPSI Pump 5
00006
13-N1105-00045 HPSI Pump with Modified Thrust Bearing Housing 0A
Procedures
Revision
Number Title Date
01DP-0AP12 Condition Reporting Process 33
01DP-0CC01 Design Control Process 12
01DP-0CC01 Design Control Process 14
01DP-9ZZ01 Systematic Troubleshooting 11
01-E-ZZI-0003 Electrical Equipment Database 19
Procedures
Revision
Number Title Date
13-CS-A038 Buried Piping and Tanks Program 2
14-VT-1023 RWT Pressure Test 07/08/2014
15DP-0CC03 Simulator Load Control 3
30DP-0WM07 Controls for Use of Measuring and Test Equipment (M&TE) 5
30DP-0WM07 Controls for Use of Measuring and Test Equipment (M&TE) 7
30DP-9MP01 Conduct of Maintenance 49
31MT-9SIG4 Tube Plugging of the Shutdown Cooling Heat Exchangers 0
(SCHEs)
2MT-9NB01 ESF and Normal Service Transformer Maintenance 10
2MT-9NB01 ESF and Normal Service Transformer Maintenance 4
2MT-9NB01 ESF and Normal Service Transformer Maintenance 5
2MT-9ZZ34 Maintenance of AM-4.16-250-9H GE Magne-Blast 35
Circuit Breakers
2MT-9ZZ74 Molded Case Circuit Breaker Test 50
2MT-9ZZ82 Time Delay Relay Test 20
2ST-9ZZ03 Surveillance Test Procedure for Class 1E 4160 Volt Bus 21
Undervoltage Protection Relays
2ST-9ZZ74 Molded Case Circuit Breaker Surveillance Test 46
36ST-9SB41 Transmitter Response Time Test Inside Containment 22
36ST-9SB42 PPS Bistable and Bistable Relay Response Time Test 39
36ST-9SB44 RPS Matrix Relays to Reactor Trip Response Time Test 25
36ST-9SB51 Transmitter Response Time Test Outside Containment 15
40AL-9RK2B Panel B02B Alarm Responses 59
40AL-9RK2C Panel B02C Alarm Responses 7
40AL-9RK2D Panel B02D Alarm Responses 7
Procedures
Revision
Number Title Date
40AL-9RK3A Panel B03A Alarm Responses 41
40AO-9ZZ02 Excessive Reactor Coolant System Leak Rate 19
40DP-9ZZ04 Time Critical Action (TCA) Program 13
40EP-9EO01 Standard Post Trip Actions 23
40EP-9EO010 Standard Appendices 106
40EP-9EO03 Loss of Coolant Accident 44
40EP-9EO04 Steam Generator Tube Rupture 11, 12, 33
40EP-9EO09 Functional Recovery 10, 11, 12, 62
40EP-9EO10- Appendix 43: Restarting MFPs 0
043
40OP-9OP01 Manual Operation of Air Operated Valves 2
70DP-0EE03 Reactor Trip Breaker Performance Monitoring 4
70DP-0MR01 Maintenance Rule 44
73DP-9XI01 Pump and Valve Inservice Testing Program 39
73DP-9XI01 Pump and Valve Inservice Testing Program 39A
73DP-9ZZ12 Motor-Operated Valve (MOV) Program 11
73ST-9SG01 MSIVs - Inservice Test 44
73ST-9SI10 HPSI Pumps Miniflow - Inservice Test 53
73ST-9XI06 CH and SS Valves - Inservice Test 32
73ST-9XI14 Train B HPSI Injection and Miscellaneous SI Valves 37
Quarterly - Inservice Test
73ST-9XI33 HPSI Pump and Check Valve Full Flow Test 64
73ST-9XI53 Train A HPSI Injection and Miscellaneous SI Valves - Cycle 12
- Inservice Test
Procedures
Revision
Number Title Date
73ST-9XI54 Train B HPSI Injection and Miscellaneous SI Valves - Cycle 10
- Inservice Test
73TI-9XI01 Vibration Data Collection for Surveillance Tests 22
73WP-0ZZ07 Welding of Stainless and Nickel Alloys 18
77ST-9SB21 CPCS Response Time Test 15
77ST-9SB22 CEA Drop Time Test 13
81DP-0CC28 Classification of Structures, Systems, and Components 15
81DP-0DC17 Temporary Configuration Changes 41
81DP-0EE10 Design Change Process 39
81DP-0EE10 Design Change Process 31
81-DP-0EE10 Design Change Process 42
81TD-0DC88 Design Guide for Instrument Uncertainty and Setpoint 1
Determination
93DP-0LC07 10 CFR 50.59 and 72.48 Screenings and Evaluations 28
ERM Event Reporting Manual 46
IP-ENG-001 Standard Design Process 0
MI 244102 Inspection, Cleaning, and Maintenance of Calvert Bus and N/A
Manual Transfer Switch
QAPD Palo Verde Nuclear Generating Station (PVNGS) 0
Operations Quality Assurance Program Description
Design Basis Documents
Number Title Revision
DBM CL Containment Integrity (Leakage and Isolation) Topical 14
DBM PB Class 1E 4.1 KV Power System 13
SG Main Steam 33
SI Safety Injection System 39
Vendor Manuals and Documents
Number Title Revision
13-VTD-I075- INGERSOLL-RAND Install Operation and Maintenance 5
00007 Instructions for R 4X11 CA-8 HPSI Pumps
13-VTD-I075- Ingersoll-Rand Sectional Assembly Parts List for 4X11CA-8 2
00037 HPSI Pumps
13-VTD-L200- Instructions and Maintenance Manual for Limitorque Type 2
0003-3 SMB Valve Operators
13-VTD-R165- Reliance Electric Co. Instruction Manual for Fractional 0
0010-1 Horsepower Duty Master A-C Motors
MSDS 7031 Fyrquel 220MLT Safety Data Sheet 3
VTD-A109-0001 Agastat Nuclear Qualified Time Delay Relays 2
VTD-A109-0045 Agastat EGP/EML/ETR Series, Nuclear Qualified Control 0
Relays
VTD-E270- Engineers and Fabricators Co. (EFCO) Technical Manual 3
00001 for SCHEs
VTD-L200-0039 Limitorque Technical Update 92-02, Recommended Spring 0
Pack Replacement for SMB Actuators
Engineering Changes
Number Title Revision
PB-1655 Degraded Voltage Relay Modification 0
DEC-00149 Replacement Spring Packs for Limitorque Actuators 2
DEC-00557 Fuse Addition to the PS1, PS2, and PS3 power supplies 1
internal to MSFIS Logic cabinets 1/2/3J-SGA/B-C01 to
enhance the existing over current protection
DEC-01035 Replacement of Obsolete TEC Molded Case Circuit 0
Breakers in Q Class application
Engineering Evaluations
Revision
Number Title Date
13-MS-B089 Cavitation in the Safety Injection System 4
Engineering Evaluations
Revision
Number Title Date
13-MS-C003 ESFAS Valve Stroke Time Adjustment Evaluation 1
EEQ-S124-001 Determination of Thermal Life for Skinner Solenoid Valve 14
EER-85-SI-105 Tube Plugging on Unit 2 SDCHX Room A 08/11/1985
EER-85-SI-108 Tube Plugging on Unit 2 SDCHX Room BA 08/12/1985
EER-91-SG-009 Instrument Flex conduit on MSIV's and FWIV's 01/16/1991
ENG-DMWO Improve Motor Actuated Valve Capability Margin 0
3197093
ERET 2827053 Operator Hydraulic SG 1 Economizer Feedwater Upstream 14
ISOL Valve
Quality Assurance Audits
Number Title Date
NAD Audit Plan Engineering Programs 07/14/2017
and Report
2017-003
NAD Audit Plan Design and Configuration Control 11/30/2017
and Report
2017-007
NAD Audit Plan Corrective Action 07/20/2018
and Report
2018-003
Other
Revision
Number Title Date
CR 19-03334-001 Evaluation 1
Level 3 Evaluation Report 19-03075-002 0
Time Critical Action Validation Package, Validation 03/03/2019
Description: TCA 70
01-J-ZZI-0004 Controlled Motor Operator Data Base 33
Other
Revision
Number Title Date
2-MN725-A PVNGS, Unit 2 Replacement Steam Generators and Power 2, 3
00185-2 Uprate - Steam Generator Tube Rupture Events
A-PV2-FE-0166
2-06244- (PVNGS, Units 1, 2, and 3; Docket Nos. STN 50-528, 08/27/2010
DCM/RAB/DFS 50-529, and 50-530; Request for Operating License
Amendment - Revision of Feedwater Line Break with Loss
of Offsite Power and Single Failure Analysis
2-06321- PVNGS, Units 1, 2, and 3; Docket Nos. STN 50-528, 02/11/2011
DCM/DFS 50-529, and 50-530; Response to Request for Additional
Information Regarding License Amendment Request to
Revise the Feedwater Line Break with Loss of Offsite
Power and Single Failure Analysis (TAC No. ME4596,
ME4597, and ME4598)
2-06364- PVNGS, Units 1, 2, and 3; Docket Nos. STN 50-528, 05/25/2011
DCM/DFS 50-529, and 50-530; Response to the Second Request for
Additional Information Regarding License Amendment
Request to Revise the Feedwater Line Break with Loss of
Offsite Power and Single Failure Analysis
(TAC No. ME4596, ME4597, and ME4598)
2-06948 Letter, Degraded Voltage License Amendment Request 09/26/2014
Commitment
2-07219 Letter, License Amendment Request 04/01/2016
13-NS-B062 At-Power PRA Study for Human Reliability Analysis 13
86-001-000 MSIV Stroke Time Requirements - Static Conditions 03/06/86
C2 PVNGS Design Basis Manual, Title: Hazards Topical 14
LDCR Log No. UFSAR Section 15.6.3 06/02/2011
2011-F009
S-07-0353 DMWO 2974904 Rev. 0, Trico Oil Level Installation, HPSI 1
Pumps, 10 CFR 50.59 Screening
S-10-0051 Generic Screening for Design Equivalent Changes 1
SABD-6.03 Time Critical Actions for LOCA, Non-LOCA, and Fire 1
Protection Accident Analysis
Sample 52163 1R19 MSIV Fyrquel Sample Results 01/22/2016
Other
Revision
Number Title Date
Sample 57291 Test Rig Fyrquel Sample Results 02/26/2019
SG System Health Report - Main Steam Q3-2018
SG Maintenance Rule Performance Criteria Basis 4
SP-PEC-157 Main Steam and Feedwater System Valve Design 04/28/1978
Requirements - System 80
TA-13-C00- Emergency Operating Procedure (EOP) Setpoint Document 0, 1, 2, 11
2000-001
TA-13-C00- Feedwater Line Break with LOP and Single Failure for Long 6
2002-002 Term Cooling
V-AFS-86-027 Palo Verde Nuclear Generating Station HPSI Hot Leg 12/19/1986
Injection Gate Valves SI-604, (-609) Relaxed Open/Close
Differential Pressure Requirements (525506)
V-CE-32369 CE Letter, PVNGS Safety Analysis Assumptions 04/30/1981
Condition Reports (CRs)
2-00608-001 12-00608-004 15-02686 15-06019 15-09485
15-11837 16-08132 16-08153 16-16517 16-16540
16-19076 16-19182 17-03158 17-05248 17-05253
17-05525 17-06480-001 17-06480-002 17-06480-004 17-06480-009
17-06911 17-08852-001 17-09268 17-09661 17-09665-001
17-09665-002 17-15064 17-16121 17-16133 18-02605
18-02676 18-10272 18-12153 18-16235-004 18-16683
18-16683-001 18-17372 18-17372-003 18-17463-002 18-17740
2-15-10350 2-16-07510 2303499 2941496 2970675
2973682 3354252 3355082 CRAI 4465947 CRDR 2856266
CRDR 4194803 PVAR 2967140
Condition Reports (CRs) Generated during the Inspection
19-02219 19-02289 19-02334 19-02371 19-02281
19-02285 19-02304 19-02305 19-02401 19-02402
19-02674 19-02679 19-02724 19-02779 19-02896
19-02902 19-02985 19-02987 19-03005 19-03020
19-03024 19-03044 19-03075 19-03334 19-04087
Work Orders (WOs)
19-03024-002 19-02305-004 2750567 2893355 2974904
3011661 3197093 3321368 3772667 3798574
3969170 4194252 4284559 4316495 4344033
4431114 4530516 4538379 4601740 4624445
Work Orders (WOs)
4640402 4645712 4645718 4693502 4699032
4699035 4699035 4704605 4723071 4731640
4731896 4736941 4739208 4745966 4745966
4755588 4755588 4777631 4777968 4784379
4793504 4793838 4795732 4796044 4796099
4885553 4885786 4885888 4886937 4889402
4893176 4899614 4907976 4912100 5014881
ADDITIONAL DOCUMENT AND INSPECTION REQUESTS
Second Document Request for Palo Verde DBA Inspection
Inspection Report 2019011
February 05, 2019
This letter does not contain new or amended information collection requirements subject to the Paperwork Reduction Act of 1995
(44 U.
control number 3150-0011. The NRC may not conduct or sponsor, and a person is not required to respond to, a request for
information or an information collection requirement unless the requesting document displays a currently valid Office of
Management and Budget control number.
Below is the second document request for this inspection. Please add each item to the
electronic document database corresponding to the inspector requesting the information.
Please upload this information as soon as achievable. If you have any questions, please
contact the team leader, Gerond George.
The team requests meetings with the system engineers for their affected components
and modifications to occur on Tuesday, February 12.
Main Steam Isolation Valve XJSGE-UV171
1. CRs 15-02686, 16-19076 (with attachments)
2. CRDR 4403348, 3338918
3. CRAIs 4564954, 4564948, 4564947
4. ASME OM Code Inservice Testing Program and bases
5. List of preventive maintenance tasks, frequency, and bases
6. List of relevant operating experience
7. List of relevant CRs for past 3 years (with descriptions, all MSIVs)
8. Most recent testing results (i.e., completed work orders) including:
a. Surveillance testing
i. MSIV/IA Check Valves: Accumulator Drop Test
ii. Stroke testing
b. Leak rate testing
9. 13-MM-0234A
10. 01,02,03-M-SGP-0001
11. 13-J-082-0218
2. 13-MC-SG-0807
13. 13-JC-SG-0200 (EER 91-SG-125)
14. SP-PEC-157, Rev. 0
15. 13-MC-SG-0316
16. 13-NC-SG-0001
17. M234A-00107
18. M234A-00045
19. EER 91-SG-125
20. Any trend data
Susan Gardner
Containment Isolation
Documentation that supports the response times (calc, or SER, or standard) from signal
initiation to valve closures, including CE letter V-CE-32369, 5/14/85
HPSI motor replacement, APN 135508
1. Complete equivalency eval and PV engineering procedure for completing it. Please
verify that the procedure revision matches the date of eval.
2. Package that installed motor (maybe DM 17-01007-002?) including PMTs
3. Vendor documentation for motor, including motor curves
4. Calculation supporting motor protection
5. Does PV issue a component health report for motors? If so, please provide.
For RPS:
1. Root cause evaluations for trips of unknown cause, 18-02605-009 on unit 1 and
18-08748-005 on unit 2
2. Evaluation supporting RPS response times (13-JC-SB-0202?)
Train A ESF Service Transformer S03
1. Functional description of Transformer S03
2. System health report
3. Single line diagram
4. Transformer loading Calc
5. Transformer Protection system procedure and drawings
6. PM procedure for Transformer
7. Copies of four latest PM performed on Transformer
8. Vendor drawings for Transformer
9. Vendor PM and surveillance requirement for Transformer
10. Transformer Primary Cable sizing and protection calculation
11. 5 years previous CR associated with Service Transformer
4160 VAC Bus PBA-S03 and Control Relaying
1. Functional description of 4160 VAC bus PBA-S03
2. Logic and schematic bus and feeder breakers
3. System health report
4. Procedure for bus and FDR breaker maintenance, overhaul and surveillance
5. Vendor manual for bus and feeder breakers
6. Vendor PM requirements for SWGR and bus feeder breakers
7. Hard copy last five PMs
8. Coordination study and sizing of breakers
9. Refurbishment requirement for breaker
10. 5 years previous CRs and failures associated with SWGR and breaker
11. FDR breaker control and wiring drawing
2. Coordination and protection studies for the control relays
DEC-01035 - Replacement of Obsolete TEC Molded Case Breakers in Q Class Application
1. Modification package
2. Cable and circuit changes associated with mod package
3. Calculations changes from mod package
4. Schematic and control wiring diagram from mod changes
5. PM revisions from modification
6. Protection and coordination calculation changes
7. PM procedure and test breaker
8. Power feeder and control circuit voltage drop calculations
9. Vendor manual breaker
10. Vendor data sheet for new breakers
11. Breaker sizing calculation
MOV HV609 - Safety Injection Valve Loop 2 Hot Leg
1. One-line diagram for valve motor
2. Cable schedule and routing for power to valve motor
3. Cable sizing for valve motor
4. Control wiring diagram for valve motor
5. Control Transformer sizing for valve motor control
6. Motor brake HP calculation
7. Protection and coordination calculation for motor
8. Hard copy last surveillance test for valve motor
9. Power feeder and control circuit voltage drop calculations
10. Vendor manual for valve motor
11. Vendor data sheet for valve motor
2. Last 5 years CR for valve motor
13. Cable ampacity calculation for valve motor power cable
14. Electrical protection calculation.
15. Motor starting time calculation
NRC IN 2006-31 - Inadequate Interrupting Rating of Breakers
1. Licensee response to the IN
2. Calculation review and revisions from the IN
John Zudans
HPSI Pumps
Provide the following references:
I075-0001, 14273-1257, 14473-732, EER 90-SI-135, 9071884-14273, 9500068-91,
SP-PCE-047, SP-PEC-733, SP-PCE-745, SP-PCE-746, SP-PCE-782, SP-PSD-093,
SP-PSD-110, SP-PSD-111, SP-PSD-191, SP-PSD-675, V-CE-21827, V-CE-30391,
V-CE-31127, V-CE-31392, V-CE-31515, V-CE-32754, V-MPS-89-080, V-MPS-89-105,
V-PCE-1224, V-PCE-1536, V-PCE-4069, V-PSD-838, V-PSD-2422, V-SF-1262,
CDBR-0031, 13-MC-SI-0215, 13-MC-SI-0250, 13-MS-B089, 13-MC-SI-0316, 6
249-MPS-5CALC-001, SP-PEC-006, SP-PEC-007, V-PEC-254, DMWO 2974904
Third Document Request for Palo Verde DBA Inspection
Inspection Report 2019011
February 07, 2019
This letter does not contain new or amended information collection requirements subject to the Paperwork Reduction Act of 1995
(44 U.
control number 3150-0011. The NRC may not conduct or sponsor, and a person is not required to respond to, a request for
information or an information collection requirement unless the requesting document displays a currently valid Office of
Management and Budget control number.
Below is the third document request for this inspection. Please add each item to the electronic
document database corresponding to the inspector requesting the information. Please upload
this information as soon as achievable. If you have any questions, please contact the team
leader, Gerond George.
DEC-00149
1. 13-MC-ZZ-0217
2. VTD L200-00039
3. Procedure for making a Design Equivalent Change
4. Test new spring pack per Teledyne Technical Manual (i.e., completed work order)
5. Post-installation test documentation (i.e., completed work order)
6. Installation documentation (i.e., completed work order)
Safety Injection Valve Loop 2 Hot Leg MOV HV609
1. Vendor manuals (valve, actuator, motor) for installation, operation, and maintenance
2. IST program/bases document
3. 13-JC-ZZ-0307
4. 13-MC-SI-0222
5. 01-MC-SI-0324
6. DMWO 3197093
7. V-FS-C-047
8. V-AFS-86-027
9. List of CRs (past 3 years)
10. List of PM activities
11. List of surveillances, tests, etc.
2. Results from most recent static/dynamic testing (i.e., completed work order)
13. Trend data on MOV performance, if any
14. Calculation for voltage drop at MOV
15. Calculation of environmental conditions
16. Methods for selecting, setting, and adjusting MOV switch settings
17. MOV functional requirements under normal, abnormal, and accident conditions
Susan Gardner
Containment Isolation
1. System health report refers to "sticking issue" of containment isolation solenoid valves.
Please provide any CRs regarding this issue for last 5 years.
2. Provide drawings of how solenoids are used in isolation valve actuators
3. CR 16-09987
HPSI Motor Replacement, APN 135508
1. VTD-H987-00001 and M021-00048-motor data and M01-00062-Time vs current
2. Calculation supporting motor protection 01-EC-PB-0200
3. 13-MC-DG-0401 DG brake HP loads
4. 13-MS-B016 DG frequency and voltage variation Impact to mechanical systems
5. 01-EC-MA-0221 AC distribution calculation
6. PM procedure for new motor, 4736953
RPS:
1. RPS system health report
2. Many figures in FSAR 7.2 are illegible. Specifically, block and logic diagrams 7.2-5
through 7.2-11. Please provide improved figures.
From: Hedger, Sean
Sent: Sunday, March 03, 2019 5:54 PM
To: Sean.McCormack@aps.com
Cc: George, Gerond <Gerond.George@nrc.gov>
Subject: Request - EOPs and Setpoint Basis Document Revisions
Sean,
With the discussions this past Thursday, I wanted to understand in what revision of the EOP
setpoint basis document, as well as the pertinent EOPs, did the SG band of for covering the
tubes in a ruptured steam generator change from 40%NR to 45%NR. The documents that I
think will show this will be:
TA-13-C00-2000-01, "Emergency Operating Procedure (EOP) Setpoint Document"
40EP-9EO09, "Functional Recovery"
40EP-9EO04, "Steam Generator Tube Rupture"
If you could provide copies that reflect when the change was made, as well as their effective
dates, I would appreciate it. Thank you.
From: George, Gerond
To: "Lorraine.Weaver@aps.com"
Cc: "sean.dornseif@aps.com"; Hedger, Sean; Deese, Rick
Subject: Palo Verde - TCA 70 phone call and questions
Date: Tuesday, March 19, 2019 1:12:00 PM
Attachments: image001.png
Lorraine,
I would like to set up a phone call for tomorrow at either 9am or 1pm (Arizona Time) to
discuss a few additional questions on TCA 70. I would like Tom Hook and Jeff Fearn
available on the call as well. During the call we will have a few questions and would like to
discuss the differences between the DBA scenario and the PRA analysis. To be specific,
we will be asking the following questions:
1. Assuming the time available for recovery for 1AF-FWIV-2HR is 10 minutes, what
would be the resulting:
a. HEP
b. CCDP/CDP/RAW
2. If time to complete the action was 57 minutes or greater, how does this change the
failure probability (HEP) of 1AF-FWIV-2HR?
3. Could you provide us the Top 50 cutsets and CCDP/CDP/RAW results that include
1AF-FWIV-2HR when changing the HEP to the following:
a. 0.5
b. 1.0
c. Resultant HEP from Question 1.
4. How would the results of question 3 be affected if the nonsafety AFW pump (AFNP01)
were out of service for maintenance?
Thanks
USNRC Region IV, Division of Reactor Safety
8172001562
Gerond.George@nrc.gov
SUNSI Review: ADAMS: Non-Publicly Available Non-Sensitive Keyword: NRC-002
By: GAG Yes No Publicly Available Sensitive
OFFICE RI:EB1 RI:EB2 EPI:RCB SRA:EB2 SRI:EB1 AC:PBD C:EB1
NAME JBraisted NOkonkwo SHedger RDeese GGeorge DDodson VGaddy
SIGNATURE /RA/ /RA/ /RA/ /RA/ /RA/ /RA/ /RA/
DATE 4/29/2019 4/30/2019 5/1/2019 5/3/2019 4/30/2019 5/7/2019 5/10/2019