IR 05000528/2018004
| ML19031B145 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 01/29/2019 |
| From: | O'Keefe N NRC/RGN-IV/DRP/RPB-D |
| To: | Bement R Arizona Public Service Co |
| References | |
| IR 2018004 | |
| Download: ML19031B145 (41) | |
Text
January 29, 2019
SUBJECT:
PALO VERDE NUCLEAR GENERATING STATION - NRC INTEGRATED INSPECTION REPORT 05000528/2018004, 05000529/2018004, AND 05000530/2018004
Dear Mr. Bement:
On December 31, 2018, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Palo Verde Nuclear Generating Station Units 1, 2, and 3. On January 2, 2019, the NRC inspectors discussed the results of this inspection with Mr. Bruce Rash and other members of your staff. The results of this inspection are documented in the enclosed report.
NRC inspectors documented two findings of very low safety significance (Green) in this report.
Both of 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. Further, inspectors documented a licensee-identified violation which was determined to be Severity Level IV in this report. The NRC is treating this violation as an NCV consistent with Section 2.3.2 of the 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; and the NRC resident inspector at the Palo Verde Nuclear Generating Station.
If you disagree with a cross-cutting aspect assignment or a finding not associated with a regulatory requirement 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 the 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/
Neil F. OKeefe, Chief Project Branch D Division of Reactor Projects
Docket Nos. 50-528, 50-529, and 50-530 License Nos. NPF-41, NPF-51, and NPF-74
Enclosure:
Inspection Report 05000528/2018004, 05000529/2018004, 05000530/2018004 w/Attachments:
1. Documents Reviewed 2. Request for Information
Enclosure U.S. NUCLEAR REGULATORY COMMISSION
Inspection Report
Docket Numbers:
05000528, 05000529, 05000530
License Numbers:
Report Numbers:
05000528/2018004, 05000529/2018004, and 05000530/2018004
Enterprise Identifier: I-2018-004-0013
Licensee:
Arizona Public Service Company
Facility:
Palo Verde Nuclear Generating Station
Location:
5801 South Wintersburg Road, Tonopah, AZ 85354
Inspection Dates:
October 1, 2018 to December 31, 2018
Inspectors:
C. Peabody, Senior Resident Inspector
D. Reinert, PhD, Resident Inspector
D. You, Resident Inspector
W. Sifre, Senior Reactor Inspector
J. Drake, Senior Reactor Inspector
C. Osterholtz, Senior Operations Engineer
C. Smith, Reactor Inspector
C. Stott, Reactor Inspector
S. Hedger, Emergency Preparedness Inspector
N. Hernandez, Operations Engineer
Approved By:
N. OKeefe, Chief
Project Branch D,
Division of Reactor Projects
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting baseline inspections 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.
NRC-identified and self-revealed findings, violations, and additional items are summarized in the table below. Licensee-identified non-cited violations are documented in the Inspection Results at the end of this report.
List of Findings and Violations
Failure to Properly Preplan Corrective Maintenance Cornerstone Significance Cross-cutting Aspect Inspection Procedure Initiating Event Green NCV 05000529/2018004-001 Closed
[H.5] -
Human Performance,
Work Management 71111.13
Maintenance Risk and Emergent Work Control The inspectors reviewed a Green, self-revealing, non-cited violation of Technical Specification 5.4.1.a for the licensees failure to properly preplan maintenance that can affect the performance of safety-related equipment. Specifically, while planning replacement of an automatic control element drive mechanism timer module (ACTM) card, the licensee failed to provide instructions that would prevent unseating the adjacent ACTM card and losing electrical continuity or else plan the work for a plant mode that would not result in a plant transient. This caused a control element assembly to drop to the bottom of the core resulting in an automatic reactor trip.
Failure to Establish an Adequate Procedure for Control of Potential Tornado Borne Missiles Cornerstone Significance Cross-cutting Aspect Inspection Procedure Mitigating Systems
Green NCV 05000529/2018004-002 Closed
[H.3] -
Human Performance,
Change Management 71111.20 Refueling and Other Outage Activities The inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B,
Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to prescribe instructions appropriate to the circumstances for the control and monitoring of transient missile hazards that have the potential to affect the operability of the essential spray ponds.
PLANT STATUS
Units 1 and 3 operated at or near full power for the entire inspection period.
Unit 2 operated at full power with the exception of a planned refueling outage which lasted from October 6, 2018 - December 3,
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 (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. [Include for integrated report: The inspectors performed plant status activities described in IMC 2515 Appendix D, Plant Status and conducted routine reviews using IP 71152, Problem Identification and Resolution.] 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
71111.04Equipment Alignment Partial Walkdown
The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:
- (1) Unit 2 shutdown cooling system A, on October 12, 2018
- (2) Unit 3 high pressure injection system A, on November 14, 2018
- (3) Unit 1 Class 1E 480V motor control centers system A, on December 20, 2018
71111.05AQFire Protection Annual/Quarterly Quarterly Inspection
The inspectors evaluated fire protection program implementation in the following selected areas:
- (1) Unit 3 electrical penetration rooms, Fire Zones 42 A and 42B, on October 1, 2018
- (2) Unit 2 main steam support structure 140 elevation, Fire Zones 74A and 74B, on October 3, 2018
- (3) Unit 2 containment near steam generator #2 economizer line, Fire Zone 63A, on October 17, 2018
- (4) Unit 3 fuel building, Fire Zones 27, 28, and 29, on October 18, 2018
- (5) Unit 1 main control room, Fire Zone 17 on November 16, 2018
71111.07Heat Sink Performance Heat Sink (Triennial)
The inspectors evaluated exchanger/sink performance on the following components November 6, 2018, to November 9, 2018:
- (1) Unit 2 Essential Cooling Water Heat Exchanger B, Section 02.02b
- (2) Unit 1 Essential Chiller A, Section 02.02c
- (3) Unit 1 Fuel Pool Cooling Heat Exchanger B, Section 02.02c
- (4) Unit 3 Auxiliary Feedwater Pump Turbine Oil Cooler, Section 02.02c
71111.08Inservice Inspection Activities (Unit 2)
The inspectors evaluated non-destructive examination activities by observing the following activities:
- (1) Ultrasonic Examinations a) Data Sheet 2-053-013 on steam to auxiliary feedwater system b) Data Sheet 2-054-006A on Steam Generator 1 feedwater system c) Data Sheet 2-054-010A on Steam Generator 1 feedwater system
- (2) Dye Penetrant Examinations a) Data Sheet 2-054-005-NW-1-18-PT-2003 on Steam Generator 1 feedwater system b) Data Sheet 2-054-006A-18-PT-2004 on Steam Generator 1 feedwater system
- (3) Phased Array Ultrasonic Examinations Data Sheet 2-027-009-011 on Pressurizer Spray 1A
- (4) Radiography Examination Data Sheet 18-946 on Steam Generator 1 feedwater system
The inspector evaluated non-destructive examination activities by reviewing the following activities:
- (1) Ultrasonic Examinations a) Data Sheet 2-058-001-NW-1 on Steam Generator 1 auxiliary feedwater system b) Data Sheet 2-058-004-NW-1 on Steam Generator 1 auxiliary feedwater system c) Data Sheet 2-058-019 on Steam Generator 1 auxiliary feedwater system
- (2) Dye Penetrant Examinations a) Data Sheet 2-058-001-NW-1-18-PT-2007 on Steam Generator 1 auxiliary feedwater system b) Data Sheet 2-058-004-NW-1-18-PT-2006 on Steam Generator 1 auxiliary feedwater system c) Data Sheet 2-058-019-18-PT-2001 on Steam Generator 1 auxiliary feedwater system
- (3) Magnetic Particle Examinations a) Data Sheet 2-053-013-18-MT-2012 on steam to auxiliary feedwater system b) Data Sheet 2-054-010A-18-MT-2011 on Steam Generator 2 feedwater system
- (4) Phased Array Ultrasonic Examinations a) Data Sheet 2-032-008-018 on Reactor Coolant Loop 1A drain a) Data Sheet 2-033-010-018 on Reactor Coolant Loop 1B drain b) Data Sheet 2-028-011-011 on Pressurizer Spray 1B c) Data Sheet 2-037-013-011 on Charging 2A d) Data Sheet 2-034-012-018 on Reactor Coolant Loop 2A drain e) Data Sheet 2-036-014-018 on Letdown 2B
The inspector evaluated welding activities by reviewing the following records:
Gas Tungsten Arc Welding (GTAW)a) Weld Data Sheet 5024637 on Steam Generator 1 feedwater system b) Weld Data Sheet 5024646 on Steam Generator 2 feedwater system
The Inspector evaluated the licensees boric acid control program performance.
The licensee did not perform a bare metal head inspection on the Unit 2 reactor vessel upper head penetrations.
The licensee did not perform steam generator tube eddy current examinations on the Unit 2 steam generators.
The Inspector evaluated a sample of condition reports associated with inservice inspection activities.
71111.11Licensed Operator Requalification Program and Licensed Operator Performance Operator Requalification
The inspectors observed and evaluated licensed operator continuing training simulator scenario on November 28, 2018. The inspectors assessed the performance of the operators and control room simulator.
Operator Performance (1 Sample)
The inspectors observed and evaluated operators performing a planned reactor coolant system cooldown for a refueling outage on October 6, 2018, and synchronizing the main generator to the grid following a stator rewind on December 3, 2018.
Operator Exams (1 Sample)
The inspectors reviewed and evaluated requalification examination results on December 27, 2018.
Operator Requalification Program (1 Sample)
The inspectors evaluated the operator requalification program from September 24, 2018, to December 27, 2018.
71111.12Maintenance Effectiveness Quality Control
The inspectors evaluated maintenance and quality control activities associated with the following equipment performance issues:
- (1) Unit 2 polar crane replacement parts acceptance review, on October 30, 2018
71111.13Maintenance Risk Assessments and Emergent Work Control
The inspectors evaluated the risk assessments for the following planned and emergent work activities:
- (1) Unit 1 emergent risk assessment due to the failure of safety injection system valve 660, on October 4, 2018
- (2) Units 1, 2, and 3, emergent risk due to unplanned loss of the west switchyard bus, on October 4, 2018
- (3) Unit 2 outage risk assessment during reactor coolant system lowered inventory, on October 12, 2018
- (4) Unit 2 reactor trip due to dropped control element assembly, on May 23, 2018
71111.15Operability Determinations and Functionality Assessments
The inspectors evaluated the following operability determinations and functionality assessments:
- (1) Unit 2 high pressure safety injection pump B operability during refueling water tank recirculation, on October 16, 2018
- (2) Unit 2 reactor coolant system operability following the identification of reactor head O-ring groove indication, on November 15, 2018
- (3) Unit 2 control room envelope allowable open area during maintenance/modification on the control room envelope boundary, on November 16, 2018
- (4) Unit 2 evaluation of visual indications on fuel assemblies, on November 16, 2018
71111.18Plant Modifications
The inspectors evaluated the permanent modification DMWO 3282819 Polar Crane Upgrade on November 5, 2018
71111.19Post Maintenance Testing
The inspectors evaluated the following post-maintenance tests:
- (1) 73ST-9ZZ25, Unit 2 boration flow path check valve testing following disassembly and inspection, on October 29, 2018
- (2) 73ST-9XI33, Unit 2 high pressure safety injection pump B full flow test and vibration readings, on October 30, 2018
- (3) 40ST-9DG01, Unit 3 diesel generator A surveillance test following spray pond pump A failure to start and spray pond pump breaker replacement, on November 2, 2018
- (4) 36ST-9SB04, Unit 3 plant protection system test of the main steam isolation signal channel B following an erroneous half-leg trip, on December 11, 2018
71111.20Refueling and Other Outage Activities
The inspectors evaluated refueling outage 2R21 activities from October 6, 2018, to December 3, 2018.
===71111.22Surveillance Testing The inspectors evaluated the following surveillance tests:
Routine===
- (1) 73ST-9SI06, Unit 1 B train containment spray pump, on December 20, 2018
In-service (1 Sample)
- (1) Unit 2 low pressure safety injection system A check valve 134 inservice test, on October 9, 2018
Containment Isolation Valve (1 Sample)
- (1) Unit 2 local leak rate testing of electrical and mechanical penetrations, on October 23, 2018
71114.04Emergency Action Level and Emergency Plan Changes
The inspector evaluated Palo Verde Nuclear Generating Station Emergency Plan, Revision 62, submitted on September 25, 2018; and Revision 63, submitted on December 6, 2018. Associated 10 CFR 50.54(q) emergency plan change process documentation was reviewed as well. The evaluation was performed in-office from December 3, 2018, to December 14, 2018. This evaluation does not constitute NRC approval.
OTHER ACTIVITIES - BASELINE
71152Problem Identification and Resolution Semiannual Trend Review
The inspectors reviewed the licensees corrective action program for trends that might be indicative of a more significant safety issue.
Annual Follow-up of Selected Issues (2 Samples)
The inspectors reviewed the licensees implementation of its corrective action program related to the following issues:
- (1) Unit 2 containment liner punch marks
- (2) Units 1 and 2 excore nuclear instruments missing O-rings at preamplifier-filter connections
71153Follow-up of Events and Notices of Enforcement Discretion Licensee Event Reports
The inspectors evaluated the following licensee event reports which can be accessed at https://lersearch.inl.gov/LERSearchCriteria.aspx:
- (1) Licensee Event Report 05000528/2018-001-00, automatic actuation of the reactor protection system resulting from a loss of reactor coolant pumps, on February 15, 2018
- (2) Licensee Event Report 05000528/2018-003-00, condition prohibited by technical specification 3.7.4 due to inoperable atmospheric dump valve, on May 14, 2018
- (3) Licensee Event Report 05000528/2018-004-00 and 05000528/2018-004-01, engineered safety feature pump room exhaust air cleanup system failure resulting in a condition prohibited by technical specifications, on May 17, 2018
- (4) Licensee Event Report 05000529/2018-001-00, Unit 2 reactor trip due to dropped control element assembly, on May 23,
INSPECTION RESULTS
Observation 71152 Problem Identification and Resolution
The inspectors reviewed the licensees response to a negative trend noted on their reactor vessel O-ring, specifically, the inner O-ring. The reactor vessel closure head interfaces with the reactor vessel by a raised face flanged design containing two metallic O-ring seals.
Between the inner and outer O-rings on the reactor vessel mating surface is a port that is piped to the reactor drain tank. This small pipe is also connected to a pressure instrument which is used to determine if a leak past the inner O-ring exists.
On November 5, 2017, Unit 1 had indication of rising pressure in the area between the inner and outer reactor vessel O-rings (CR 17-16139). On May 4, 2018, Unit 3 had a similar indication of rising pressure (CR 18-07701). Both of these cases indicate that there is leakage past the inner O-ring. However, there has yet to be indication of any leakage past the outer O-ring. It should be noted that in both cases, the leakage was observed shortly after coming out of a planned refueling outage.
Noting these two adverse trends, the licensee elected to perform an evaluation on the handling and installation of the metallic O-rings during the Unit 2 refueling outage which took place from October 6, 2018, to December 3, 2018. The licensee had the O-ring vendor Technetics onsite to perform the evaluation. Although the vendor did not observe any gross errors with the handling and installation, they did recommend the following actions:
- Placing the O-ring in a transportation crate from the warehouse to the head stand
- Use of tygon tubing for protecting the seals
- Maintain protective covering on the new O-ring seals until the old O-ring has been removed and the new O-ring is ready to install
- Consider repairing any defects greater than 0.0001 on the sealing surface
- A significant factor in sealing performance is percent of compression of the seal.
Currently, Palo Verde is seeing about 9 percent compression. The vendor is recommending 16 percent compression. This would entail using a thicker O-rings (from 0.5 cross section O-ring to 0.525)
The licensees engineering group is using the vendor report as an input to a final engineering evaluation of all the actions they plan on implementing to enhance their handling and installation of the reactor vessel O-rings. The licensee is aiming for these recommendations to be put into action prior to their next refueling outage (Unit 1, April 2019). These enhancement actions will be documented under CR 18-07701.
Observation 71111.18 Plant Modifications
The inspector discussed with licensee staff discrepancies noted in the vendor quality assurance paperwork that the licensee accepted deviations of the as-delivered components from NUREG-0554, ASME NOG-1, and Palo Verde Nuclear Generating Station design specification 13-CN-0390, Technical Requirements for Upgrading the Containment Building Polar Crane to Single Failure Proof, that was provided to Konecranes. The inspector noted the following exceptions listed below:
Licensee Design Document: 13-CN-0390, Technical Requirements for Upgrading the Containment Building Polar Crane to Single Failure Proof Requirement Observation Section 7.5.1, stated in part, The girders, end trucks, and trolleys shall be of basically welded construction. All bolted connections shall be in accordance with American Not all weldments and castings that could collect moisture have drainage holes.
Institute of Steel Construction specifications for structural joints using American Society for Testing and Materials specifications A325 or A490 bolts. All castings and weldments that may collect moisture shall have drain holes 1/2-inch minimum...
Section 7.5.2.1, stated in part, All welding and welding criteria shall be in accordance with the following codes and standards: All structural steel welding shall be performed in accordance with American Welding Society (AWS) specification D1.1...
Some of the vendors used AWS Specification D14.1, Specification for Welding of Industrial and Mill Cranes and other Material Handling Equipment versus AWS specification D1.1, Structural Welding Code - Steel.
Section 7.9.1, required the crane to be able to operate at a minimum temperature of 60ºF.
Konecranes NUREG 0612, Appendix C, Compliance Matrix, stated in part, Impact testing was not performed on the bridge when initially built, so Palo Verde Nuclear Generating Station has elected to perform critical lifts only when Containment Building temperatures are >70ºF.
Section 7.11.3.4, stated in part, All threaded fasteners which have the potential to fall into the fuel or equipment pools shall be installed utilizing thread locking fluid, lock wire, or staking to prevent loosening during crane operation. Lock washers shall not be utilized...
A number of fasteners used lock washers while the majority of the fasteners inspected had no lock washer, lock wire, staking, or apparent Locking fluid.
Section 7.11.4.1, did not permit aluminum components to be used inside containment, unless contained within a sealed terminal box or waterproof housing.
The terminal blocks in the new cabinets appear to be aluminum. Several of these cabinets are open to the containment atmosphere via cooling vents or conduits.
Vendor Document: CN390-A00359, Polar Crane Unit 2 Factory Acceptance Test Requirement Observation Step 7.46 a. auxiliary hoist micro speed test specification of 3 ft/min (+/- 10 percent)
Recorded values of 2.6 ft/min raise and 2.4 ft/min lower, outside the required range of 2.7 to 3.3 ft/min Step 7.46 b. hoist micro speed specification of 1.5 ft/min (+/- 10 percent)
Recorded values of 34.4 ft/min raise and 34.2 ft/min lower, outside the required range of 1.35 to 1.65 ft/min Step 7.46 c. auxiliary hoist electrical current specification limit not to exceed 66 amps No value recorded Step 8.48 calculated main hoist speed specification of 4.0 ft/min (+/- 10 percent)
Recorded main hoist drum speed not converted to main hoist speed Step 8.50 calculated main hoist micro speed specification of 3 in/min (+/-10 percent)
Recorded values of 23.6 raise and 22.9 lower with no units Step 8.50 b. main hoist electrical current specification limit not to exceed 84 amps No value recorded NUREG 0554, Single-Failure-Proof Cranes For Nuclear Power Plants Requirement Observation Section 2.3, Operating Environment, stated in part, The operating environment, including maximum and minimum pressure, maximum rate of pressure increase, temperature, humidity, and emergency corrosive or hazardous conditions, should be specified for the crane and lifting fixtures...
Vendor Document 13-CN-0390, Technical Requirements for Upgrading the Containment Building Polar Crane to Single Failure Proof, does not provide values for minimum pressure, maximum rate of pressure increase, and emergency corrosive or hazardous conditions. The licensee was unable to provide these values to the inspector.
Section 2.3, Operating Environment, stated in part, For cranes inside the containment structure, the closed box sections of the crane structure should be vented to avoid collapse during containment pressurization.
Drainage should be provided to avoid standing water in the crane structure.
There were a number of closed box sections of crane supports without the recommended vent and/or drain valves.
Section 2.8, Welding Procedures, stated in part,... Welds described in the recommendations of Section 2.6 should be post-weld heat treated in accordance with Subarticle 3.9 of AWS D1.1...
Some of the vendors used American Welding Society (AWS) specification D14.1, Specification for Welding of Industrial and Mill Cranes and Other Material Handling Equipment, versus AWS specification D1.1.
ASME NOG1, 2004, Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder)
Section 5427.1, specifically requires steel sheaves.
Konecranes August 23, 2017, letter stated, in part,... Since NOG-1-2004, Section 5427.1, specifically requires steel sheaves, we have taken an exception to NOG-1. Sections 5.3.5 of references 3 and 4 provide the technical justification on why the sheaves are acceptable.
Other Observations The proposed update to the updated final safety analysis report (UFSAR) stated, The containment polar cranes have been upgraded to meet the guidelines for single-failure-proof cranes contained in NUREG 0554.
The proposed update was not accurate for all of the as received polar crane components, as described in the above table for NUREG 0554.
Guidelines state that the crane meets ASME NOG-1 requirements.
Konecranes August 23, 2017, letter, states:
Since NOG-1-2004, Section 5427.1, specifically requires steel sheaves, we have taken an exception to NOG-1. Sections 5.3.5 of references 3 and 4 provide the technical justification on why the sheaves are acceptable.
The inspector identified additional discrepancies and verified that the licensee documented all discrepancies in Condition Report CR-18-17888.
Licensee-Identified Non-Cited Violation 71111.11 Licensed Operator Requalification Program This violation of very low safety-significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Violation: Title 10 CFR Part 50.74 requires licensees to notify the NRC within 30 days of a change in medical condition of a licensed operator.
Contrary to the above, from May 18, 2018, to October 30, 2018, the licensee failed to notify the NRC of a change in medical condition of a licensed operator.
Significance/Severity Level: Severity Level IV. The performance deficiency was evaluated in accordance with the reactor oversight process and was determined to be minor because a licensing decision was not made due to the absence of this medical information, the individual was not on shift during the period of May 18, 2018, to December 12, 2018, and has used this therapeutic device and taken the medications as prescribed. The ROPs significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it is necessary to address this violation which impedes the NRCs ability to regulate using traditional enforcement to deter noncompliance. The inspectors determined the violation to be a Severity Level IV violation similar to Example 6.4.d.1(b) in the NRC Enforcement Policy and because it impacted the ability of the NRC to perform its regulatory oversight function. Specifically, by failing to inform the NRC of a change in an operators medical condition, the licensee prevented the NRC from performing the regulatory reviews associated with that process.
Corrective Action Reference: Condition Report 18-17047
Failure to Properly Preplan Corrective Maintenance Cornerstone Significance Cross-cutting Aspect Inspection Procedure Initiating Event
Green NCV 05000529/2018004-001 Closed
[H.5] -
Human Performance, Work Management 71111.13 Maintenance Risk and Emergent Work Control
The inspectors reviewed a Green, self-revealing, non-cited violation of Technical Specification 5.4.1.a for the licensees failure to properly preplan maintenance that can affect the performance of safety-related equipment. Specifically, while planning replacement of an automatic control element drive mechanism (CEDM) timer module (ACTM) card, the licensee failed to provide instructions that would prevent unseating the adjacent ACTM card and losing electrical continuity or else plan the work for a plant mode that would not result in a plant transient. This caused a control element assembly to drop to the bottom of the core resulting in an automatic reactor trip.
Description:
On May 23, 2018, I&C technicians were performing planned maintenance on Unit 2. The activity was intended to replace control element drive mechanism control system (CEDMCS) logic power fuses and fuse holder caps in subgroups 15 and 16 with components of a newer design. With Unit 2 at 100 percent power, the licensee began the activity by placing subgroup 15 control element assemblies (CEAs) on the hold bus to allow the technicians to begin the fuse replacements without causing a loss of power to the subgroup 15 CEAs. As part of the maintenance activity, the technicians were to remove subgroup 15 ACTM cards one at a time. Once each set of fuses were replaced, the associated ACTM card would then be reseated. Three of the ACTM cards for subgroup 15 were reinstalled with no issues. However when attempting to reseat the ACTM card for CEA 62, the technicians encountered alignment issues. During their attempts to reseat the card, Unit 2 reactor tripped due to low departure from nucleate boiling ratio (DNBR) trip signals. There were no issues with the standard post trip actions.
The licensees post trip investigation determined that the automatic trip was the result of CEA 64 dropping to the bottom of the core causing the core protection calculators to generate a low DNBR and a high local power density trip signal. The technicians also determined that the ACTM card for CEA 64, located next to the CEA 62 card the technicians were trying to reseat and powered from a different subgroup, had become unseated and lost power. The licensee entered the event into their corrective action program as CR 18-08748 and performed a root cause evaluation.
The root cause evaluation revealed the following items:
- The alignment issue when trying to reseat an ACTM card caused the lower logic card, located behind and at right angles to the ACTM cards, to flex away from the row of ACTM cards causing adjacent ACTM cards to slowly unseat.
- The card alignment issue was a known equipment deficiency in Unit 2. Specifically, CR 17-09075, written on June 20, 2017, identified the issue that when manipulating a card, an adjacent card could be unseated.
- The card alignment issue was discussed in work planning meetings for this activity, but action to prevent the possibility affecting adjacent cards (especially in another subgroup) was not considered.
As a result the licensee did not establish any additional controls to mitigate the possibility that the known card alignment deficiency could cause unseating of adjacent cards. Alternatively, the inspectors noted that the licensee could have considered performing the fuse replacement activity during a plant mode when the card alignment deficiency would not have caused a plant transient, such as during a refueling outage.
Corrective Action: Unit 2 corrected the alignment issues during refueling outage 2R21. An extent of condition found that the alignment issue was only present in Unit 2.
Corrective Action Reference: CR 18-08748
Performance Assessment:
Performance Deficiency: The station failed to properly preplan maintenance such that the manipulation of one ACTM card would not affect the seating of another ACTM card due to the misalignment of the cards during work planning meetings. The misalignment was a known deficiency that was documented in their corrective action program.
Screening: The inspectors determined the performance deficiency was more than minor because it adversely affected the procedure quality attribute of the Initiating Events Cornerstone and affected the associated cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the station failed to consider the adverse consequences due to the ACTM card misalignments when generating work instructions to manipulate those cards. This ultimately caused a control rod to drop to the bottom of the core resulting in an automatic reactor trip.
Significance: The inspectors assessed the significance of the finding using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012. The inspectors determined that the finding had a very low safety significance (Green) because the finding resulted in a reactor trip, but did not involve a loss of mitigating equipment relied upon to transition the plant to a stable shutdown condition.
Cross-cutting Aspect: The finding has a cross-cutting aspect in the area of human performance associated with the work management component because the licensee failed to implement a process of planning, controlling, and executing work activities such that nuclear safety was the overriding priority. Specifically, the work was not effectively planned and executed by incorporating risk insights and controlling job site conditions, and the work process did not incorporate contingency plans, compensatory actions, or abort criteria as needed to prevent potential adverse consequences of known alignment problems with ACTM cards. [H.5]
Enforcement:
Violation: Technical Specification 5.4.1.a requires, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures listed in Appendix A to Regulatory Guide 1.33, Quality Assurance Program Requirements, Revision 2, February 1978. Regulatory Guide 1.33, Appendix A, Section 9.a, states, in part, that maintenance that can affect the performance of safety-related equipment should be properly planned and performed in accordance with written procedures appropriate to the circumstances.
Contrary to the above, on May 23, 2018, the licensee performed maintenance that can affect the performance of safety-related equipment which was not properly planned and performed in accordance with written work instructions that were appropriate to the circumstances.
Specifically, work instructions in Work Order 5006012 were not appropriate because they failed to address how to prevent unseating an ACTM card while seating another card, causing a loss of electrical continuity to an unaffected control element assembly.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy because of its very low safety significance (Green)and was entered in to the licensees corrective action program as CR 18-08748.
Failure to Establish an Adequate Procedure for Control of Potential Tornado Borne Missiles Cornerstone Significance Cross-cutting Aspect Inspection Procedure Mitigating Systems
Green NCV 05000529/2018004-002 Closed
[H.3] -
Human Performance, Change Management 71111.20 Refueling and Other Outage Activities The inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to prescribe instructions appropriate to the circumstances for the control and monitoring of transient missile hazards that have the potential to affect the operability of the essential spray ponds.
Description:
On September 28, 2018, the inspectors performed a walkdown of the yard area around Unit 2 and identified a large number of unsecured potential tornado borne missile (PTBM) hazards within 400 feet of the Unit 2 essential spray ponds. A PTBM is defined as an object that could become airborne during a tornado and be transported to the essential spray ponds. In addition, the object could damage the spray pond nozzles when it impacts the exposed spray nozzles or nozzle piping. The unsecured PTBMs included piping spool pieces, rubber hoses, large toolboxes, wire material storage baskets, wooden pallets, piping storage racks, and metal stanchions. The materials had been staged in preparation for draining a portion of the essential spray pond system to facilitate piping inspections during the October 2018 refueling outage. The inspectors noted that many of the PTBMs were not labelled with PTBM permit tags as required. The inspectors notified the Unit 2 shift manager of the potentially nonconforming condition.
On October 12, 2018, the inspectors were performing walkdowns near the Unit 2 essential spray ponds and noted several hundred scaffolding poles, wooden boards, and miscellaneous scaffolding components staged near the essential spray pond pump house.
The inspectors observed that many of the scaffolding materials were bound together with a lightweight nylon rope that would be insufficient to restrain the materials from becoming airborne during a design basis tornado event. The inspectors informed the licensees outage control center of this potentially nonconforming condition.
The essential spray ponds function as the ultimate heat sink. Spray nozzles, located above the surface of the essential spray ponds, are used to maintain post-accident design temperatures within safety analysis assumptions. The spray nozzles have no protective features to prevent damage from airborne missiles and, as a result, they are vulnerable to airborne missiles generated during a high wind event. During original plant licensing, the licensee used a probabilistic analysis to justify not providing tornado missile protection for the essential spray pond nozzles. This probabilistic approach is described in UFSAR, Section 3.5.1.4. The key input into this analysis is the maximum missile density near each of the stations essential spray ponds. Higher missile densities lead to an increase in the probability that the essential spray pond will not be able to perform its function to provide cooling following a tornado. The probabilistic analysis concluded that a tornado event will not cause the loss of the heat removal function of the essential spray ponds so long as the licensee manages the storage of loose materials within the transient missile density limit. If the PTBM density limit is exceeded, however, the presumption of essential spray pond operability following a tornado event is lost.
The licensee uses Procedure 01DP-0XX01, Control and Monitoring of Potential Tornado Borne Missiles, Revision 4, to maintain the license basis density limits for PTBM hazards.
This procedure contains guidance for licensee personnel to be able to identify and control PTBM hazards. Transient items that meet the definition of a PTBM must be identified and tagged with a permit and entered into the licensees database for tracking PTBMs to ensure the operability limits are not exceeded.
For the piping spool pieces and other components identified on September 28, 2018, licensee maintenance personnel had not accurately accounted for all PTBMs. PTBM permit tags were found on only a few of the items. The number of items that had been entered into the licensees tracking database was 12. However, after questioning from the inspectors, the licensee evaluated the staged material and identified that at least 64 items met the definition of a PTBM. The operability limit defined in Procedure 01DP-0XX01 is 30 PTBMs within any 10,000 square foot area. The licensee determined that maintenance personnel had been unclear of how to identify PTBMs and establish accurate quantities when creating a PTBM permit and had obtained permits for groups of material rather than individual items that could become separate missile hazards.
For the scaffolding materials identified on October 12, 2018, a PTBM permit had been generated for the materials, but included only seven racks and five equipment bins. The many scaffolding poles, boards, and other equipment were not recognized as separate PTBMs and were not included in quantity of materials entered into the PTBM database.
Thus, for both examples identified by the inspectors, the licensee had unknowingly exceeded the analyzed PTBM density limits for which the essential spray ponds could be presumed to remain operable following a tornado event.
Operations personnel use the PTBM database to monitor the margin to the PTBM density limits and ensure the essential spray ponds remain operable. Without an accurate number of PTBM items entered into the database, the licensee had not questioned the operability of the essential spray ponds until challenged by the inspectors.
The licensee evaluated these conditions under Condition Reports 18-14363, 18-15171, and 18-16311. The licensee concluded that Procedure 01DP-0XX01, Control and Monitoring of PTBMs, is poorly written and confusing to follow. The procedure was first issued as an engineering procedure in 2010 and the procedure formatting and structure has remained technical in nature rather than providing explicit instructions for how to implement a defined process. The licensee also concluded that despite its title of Control and Monitoring of PTBMs, the procedure is not structured to facilitate control or monitoring. The procedure did not contain directions for controlling PTBMs to minimize high density clusters and did not facilitate monitoring localized densities within the PTBM tracking database. The inspectors verified that the licensee generated action item 18-14363-015 to revise Procedure 01DP-0XX01 to address these deficiencies.
Corrective Action: The licensee took immediate corrective action by removing the transient PTBMs from the vicinity of the Unit 2 essential spray ponds.
Corrective Action References: Condition Reports 18-14363, 18-15171, and 18-16311
Performance Assessment:
Performance Deficiency: The licensees failure to establish an adequate procedure for the control and monitoring of potential tornado borne missiles is a performance deficiency.
Screening: The performance deficiency is more-than-minor and a finding because it is associated with the protection against external factors attribute of the Mitigating Systems cornerstone, and adversely affected the cornerstone objective of ensuring the availability, reliability and capability of the ultimate heat sink to respond to initiating events to prevent undesirable consequences. Specifically, as a result of unclear instructions, the licensee failed to control the storage of loose material that constituted potential tornado borne missile hazards in the vicinity of the Unit 2 spray ponds, and as a result exceeded the analyzed limits for which the essential spray ponds would be presumed to remain operable following a tornado event.
Significance: The inspectors performed the initial significance determination using NRC Inspection Manual Chapter 0609, Appendix A, Significance Determination Process for Findings at Power, Exhibit 4, External Events Screening Questions. Step 1.a. required a senior reactor analyst to perform a detailed risk evaluation because if the equipment or safety function is assumed to be completely failed or unavailable, it would degrade one or more trains of a system that supports a risk significant system or function.
A regional senior reactor analyst performed a detailed risk evaluation and determined that the finding was of very low safety significance (Green). The analyst used the tornado missile frequency of 1.74E-6 tornados per year for the Palo Verde site developed from tornado data from January 1, 1950, to December 31, 2006, within 100 kilometers of the plant and reviewed per Review of Methods for Estimation of High Wind and Tornado Hazard Frequencies, dated December 2012.
A postulated tornado at this frequency was assumed to cause a switchyard-centered loss of offsite power which could not be recovered within a 24-hour probabilistic risk assessment mission time. The analyst then set the failure to start basic events for both essential spray pond motor driven pumps to TRUE, thereby totaling failing both essential service water spray ponds, in the Palo Verde SPAR model, Version 8.50, run on SAPHIRE, Version 8.1.8. This failure resulted in an increase in the conditional core damage probability of 7.1E-2 during the postulated tornado event.
The analyst then applied the assumed exposure time of six days to the tornado frequency and the increase in the conditional core damage probability to obtain a bounding estimate of the increase in core damage frequency from the performance deficiency of 2.1E-9 per year.
This estimate made the significance Green or of very low safety significance for core damage frequency. Because not all of the spray nozzles would be damaged and some heat exchange capability would likely remain, the actual increase in core damage frequency would be expected to be less than this estimated value. Since the increase in core damage frequency was less than 1.0E-7, the increase in large early release frequency was not analyzed.
Losses of offsite power initiated by tornados were the dominant core damage sequences which were mitigated by the remaining auxiliary feedwater system and emergency diesel generators.
Cross-cutting Aspect: The finding has a cross-cutting aspect in the area of human performance associated with the change management component because licensee leaders failed to use a systematic process for implementing change so that nuclear safety remains the overriding priority. Specifically, when Procedure 01DP-0XX01 was initially issued in 2010, it was accompanied by several other program elements designed to raise awareness of the potential tornado borne missile program. These other program elements included issuance of a communications plan, site-wide news releases, periodic mandatory employee training, designated area owners, and operations department PTBM performance indicators. Since 2010, the licensee has terminated most of these other elements of the PTBM program. The licensee failed to re-evaluate and reinforce Procedure 01DP-0XX01 to compensate for removing the other program elements that were designed to ensure the PTBM program standards were met. [H.3]
Enforcement:
Violation: Title 10 CFR Part 50, Appendix B, Criterion V, Instruction, Procedures, and Drawings, requires that activities affecting quality shall be prescribed by instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with those instructions, procedures, or drawings. Updated Final Safety Analysis Report, Section 3.5.1.4, Missiles Generated by Natural Phenomena (Tornados), provided probabilistic criteria to ensure essential spray pond operability.
Study 13-NS-A106, Probabilistic Risk Assessment of Potential Tornado Missile Damage to the Station Ultimate Heat Sink, Revision 0, provided the missile density requirements to ensure the probabilistic criteria in the UFSAR, Section 3.5.1.4, are met.
Procedure 01DP-0XX01, Control and Monitoring of Potential Tornado Borne Missiles, Revision 4, implemented the control of transient missile hazards, an activity affecting quality, to ensure the missile density requirements of Calculation 13-NS-A106 and, therefore, the operability requirements for spray pond nozzles are met.
Contrary to the above, since 2010, Procedure 01DP-0XX01, intended to implement the control and monitoring of transient missiles that could have an effect on the operability of the essential spray ponds, an activity affecting quality, was not appropriate to the circumstances.
Specifically, licensee personnel failed to establish instructions to ensure that potential tornado missile density criteria are understood and controlled within limits.
Enforcement Action: This violation is being treated as a non-cited violation consistent with Section 2.3.2.a of the Enforcement Policy because of its very low safety significance (Green)and was entered in to the licensees corrective action program as CR 18-14363.
EXIT MEETINGS AND DEBRIEFS
On October 19, 2018, the inspectors presented the inservice inspection activities inspection results to Mr. B. Rash, Vice President, Engineering, and other members of the licensee staff.
The inspector confirmed that proprietary information was controlled to protect from public disclosure.
On November 5, 2018, the inspector presented the Unit 2 Polar Crane inspection results to Mr. M. McLaughlin, Vice President, Operations Support, and other members of the licensee staff. The inspector verified no proprietary information was retained or documented in this report.
On November 9, 2018, the inspectors presented the triennial heat sink performance inspection results to Ms. M. Lacal, Senior Vice President, and other members of the licensee staff. The inspectors verified no proprietary information was retained or documented in this report.
On December 14, 2018, the inspector communicated the emergency action level and emergency plan changes inspection results telephonically to Ms. C. Shields, Manager, Emergency Preparedness, and other members of the licensee staff. Unless otherwise noted, no proprietary information was retained by the inspector or documented in this report.
On December 27, 2018, the inspector presented the licensed operator requalification inspection results to Mr. D. Elkinton, Section Leader Compliance, and other members of the licensee staff.
The inspectors verified no proprietary information was retained or documented in this report.
On January 2, 2019, the inspectors presented the quarterly resident inspector inspection results to Mr. B. Rash, Vice President, Engineering, and other members of the licensee staff. The inspectors confirmed that proprietary information was controlled to protect from public disclosure.
THIRD PARTY REVIEWS
The inspectors reviewed a World Association of Nuclear Operator reports that was issued during the inspection period.
DOCUMENTS REVIEWED
71111.04Equipment Alignment
Procedures
Number
Title
Revision
Shutdown Cooling Initiation
Shutdown Cooling Flow Verification
High Pressure Safety Injection System Alignment
Verification
Drawings
Number
Title
Revision
01-M-SIP-001
P&I Diagram Safety Injection and Shutdown Cooling
System
01-M-SIP-001
P&I Diagram Safety Injection & Shutdown Cooling
System
01-E-PHA-001
Single Line Diagram 480V Class 1E Power System
Motor Control Center M31
01-E-PHA-003
Single Line Diagram 480V Class 1E Power System
Motor Control Center M33
01-E-PHA-005
Single Line Diagram 480V Class 1E Power System
Motor Control Center M35
71111.05AQFire Protection Annual/Quarterly
Work Orders
24637
5033525
4850349
Miscellaneous
Documents
Number
Title
Revision
PVNGS Pre-Fire Strategies Manual
71111.07Heat Sink Performance
Condition Reports
16-03487
16-02624
17-04346
16-18445
16-02413
Condition Reports Generated During the Inspection
18-18022
18-18023
18-18087
18-18088
18-18142
Work Orders
4752012
264546
4414212
210815
210816
4403647
4396131
Procedures
Number
Title
Revision
Heat Exchanger Program
Operations Condition Reporting Process and Operability
Determination/Functional Assessment
Systems Chemistry Specifications
Essential Auxiliary Feedwater System
Fuel Pool Cooling
Drawings
Number
Title
Revision
M021-00207
Oil Cooler Whitlock Type 1-R-4
F
13-C-SPS-375
Nuclear Service Spray Ponds Plan
Miscellaneous
Documents
Number
Title
Revision
or Date
1PCB Heat Exchanger Thermal Performance Analysis
for 1R16
October 29,
2011
1PCB Heat Exchanger Thermal Performance Analysis
for 1R15
April 13, 2010
1PCB Heat Exchanger Thermal Performance Analysis
for 1R14
October 16,
2008
1PCB Heat Exchanger Thermal Performance Analysis
for 1R13
June 8, 2007
VTD-S445-0002
Installation, Operation & Maintenance For Struthers
Wells Corp. Fuel Pool Heat Exchangers
VTD-S445-0006
Specifications, Spare Parts, Special Tools, and Torque
Requirements for Struthers Wells Corp. Fuel Pool Heat
Exchangers
VTD-S445-0007
Specifications, Spare Parts, Special Tools, and
Torque Requirements for Struthers Sells Corp. Essential
Cooling Water Heat Exchangers
Miscellaneous
Documents
Number
Title
Revision
or Date
VTD-C150-00014
Carrier Application Data for the Condenser and
Economizer (Essential Chiller)
Revised Response to NRC Generic Letter (GL) 89-13
Recommended Actions
October 1, 1993
Annual 10 CFR 50.59, 10 CFR 72.48 and Commitment
Change Report (January - December 2005)
December 22,
2006
1MPCAE01
Eddy Current Inspection Summary - Fuel Pool Cooler
PCAE01
February 2016
1MPCBE01
Eddy Current Inspection Summary - Fuel Pool Cooler
PCBE01
February 2016
2MPCAE01
Eddy Current Inspection Summary - Fuel Pool Cooler
PCAE01
February 2015
2MPCBE01
Eddy Current Inspection Summary - Fuel Pool Cooler
PCBE01
February 2015
3MPCAE01
Eddy Current Inspection Summary - Fuel Pool Cooler
PCAE01
January 2015
3MPCBE01
Eddy Current Inspection Summary - Fuel Pool Cooler
PCBE01
February 2015
Essential Cooling Water Heat Exchanger Thermal
Performance Test Report
October 31,
2012
Essential Cooling Water Heat Exchanger Thermal
Performance Test Report
April 23, 2014
Essential Cooling Water Heat Exchanger Thermal
Performance Test Report
May 1, 2017
FAI 13-0471
Evaluation of Acceptance Criteria Based on
Waterhammer Phenomena in the AF Piping for APS
Nuclear Generating Station, Units 1, 2 and 3
FAI 14-0364
Evaluation of Acceptance Criteria Based on
Waterhammer Phenomena in the PC Piping for APS
Nuclear Generating Station, Units 1, 2 and 3
13-MM-0021
Material Requisition for Auxiliary Feedwater Pumps
DMWO 4011490
Engineering Disposition for Spray Pond Wall
Refurbishment
13-CN-0365
Specification for Forming, Placing, Finishing, and Curing
of Grout and Concrete
Calculations
Number
Title
Revision
RE-01-C19-2014
U1R18 As-Left Spent Fuel Pool Decay Heat Projections
13-NC-PC-0203
Loss of Spent Fuel Pool Cooling Evaluations
13-MC-PC-0217
Spent Fuel Cooling System - Shutdown Cooling and
Pool Cooling Heat Transfer Evaluation
13-MC-SP-0306
MINET Hydraulic Analysis of SP System
13-MC-SP-0307
SP/EW System Thermal Performance Design Bases
Analysis
SDOC 02-MN950-
A00003
EW Heat Exchanger Replacement - B&W Essential
Cooling Water Heat Exchanger Thermal-Hydraulic
Performance & Flow-Induced Vibration Report
71111.08Inservice Inspection Activities
Condition Reports
17-04807
17-06872
17-07462
17-08423
17-08521
17-08756
17-09718
17-1114
17-12329
17-13114
17-15464
18-00733
18-00942
18-02177
18-02487
Work Orders
4704651
4791185
4595676
4718617
24637
Procedures
Number
Title
Revision
Welding and Brazing Control
Weld Filler Material Control
Radiographic Examination
Dry Magnetic Particle Examination
Liquid Penetrant Examination
Ultrasonic Examination of Pipe and Vessel Welds
Welding of Carbon and Low Alloy Steels to Stainless
and Nickel Alloys
Welding of Ferritic and Martensitic Steels
Visual Inspection of Code Welds
Radiation Protection Oversight of Radiography
Procedures
Number
Title
Revision
Boric Acid Walkdown Leak Detection
Boric Acid Corrosion Control Program
EPRI-PIPE-MPA-1 Procedure for Manual Phased Array Examination of
Austenitic and Ferritic Pipe Welds
PDI-UT-5
PDI Generic Procedure for the Straight Beam Ultrasonic
Examination of Bolts and Studs,
PDI-UT-8
PDI Generic Procedure for the Ultrasonic Examination of
Weld Overlaid Similar and Dissimilar Metal Welds
PDI-UT-11
PDI Generic Procedure for the Ultrasonic Examination of
Reactor Pressure Vessel Nozzle-to-Shell Welds and the
Nozzle Inner Corner Radius
PDI-UT-10
PDI Generic Procedure for the Ultrasonic Examination of
PDI-UT-1
PDI Generic Procedure for the Ultrasonic Examination of
Ferritic Pipe Welds
PDI-UT-2
PDI Generic Procedure for the Ultrasonic Examination of
Austenitic Pipe Welds
PDI-UT-3
PDI Generic Procedure for the Ultrasonic Through-Wall
Sizing of Planar Flaws in Similar Metal Piping Welds
Miscellaneous
Documents
Number
Title
Date
Audit 2017-003
Engineering Programs
December 6, 2017
2-07657-
MLL/TNW
Palo Verde Nuclear Generating Station Units 1, 2, and
3, Docket Nos. STN 50-528/529/530
Supplement to Relief Request 56 - Unit 2, Third
10-Year Inservice Inspection (ISI) Interval Extension
March 2, 2018
SWMS No.
4362861
Formal Self-Assessment of the Inservice Inspection
Program
August 16, 2013
SWIMS No.17-
09564
Welding Program Self-Assessment
June 30, 2017
71111.11 Licensed Operator Requalification Program and Licensed Operator Performance
Condition Reports
18-19310
18-19315
18-05351
18-06620
18-17047
17-01597
17-00651
17-02134
16-20182
16-19366
Condition Reports
17-00284
17-06366
17-03722
17-02671
17-08110
17-10495
17-10434
17-00092
17-00521
17-01060
17-01076
17-01439
17-01470
17-02201
17-02378
17-02541
18-00450
18-02315
18-02693
18-04454
Work Orders
28401
Procedures
Number
Title
Revision
Main Turbine Operations to Support Validation Testing
of the EX2100E Main Generator Excitation System Post
Generator Stator Rewind
RCS and Pressurizer Heat-up and Cooldown Rates
LOCT Annual and Biennial Operating Exam Sample
Plan Development
LOCT Biennial Written Exam Development and Sample
Plan
LOCT Annual and Biennial Exam Administration
NRC Examination Security
LOCT Scenario and JPM Development
LOCT Training Program Description
Simulator Operator Feedback
Simulator Design Control
Simulator Loadout Control
Simulator Performance Testing
Simulator Configuration
PV-1281
Exam Security Briefing Checklist
PV-1282
Pre-Exam Activities Security Checklist
PV-1283
Simulator Pre-Exam Activities Security Checklist
PV-1286
Initial License Exam Security Checklist for Admin JPMs
PV-1822
Initial License Exam Security Checklist for Written
Examinations
PV-1824
Initial License Exam Security for Simulator Scenarios
Procedures
Number
Title
Revision
Licensed Operator Medical Examinations
Operator Licensing and Requalification Process
Miscellaneous
Documents
Number
Title
Revision
or Date
2018 LOCT Operations Test Sample Plan
Operations Training Critical Task List
Written Exam Remediation December 2017
Written Exam Remediation #1 December 2018
Written Exam Remediation #2 December 2018
71111.12Maintenance Effectiveness
Condition Reports
18-01796
VC-MMH1-18-001
Quality Records
214832
214833
214834
214835
214837
214840
214864
214865
215151
215457
215458
215467
215477
215481
215489
215490
215713
215715
215716
215717
215718
215720
216789
216795
216801
216803
216804
216837
216849
216850
216855
216856
217047
217048
217049
217050
217051
217052
217251
217256
217777
217820
217821
217845
217849
217850
217898
218089
218220
218226
218227
218228
218229
218230
218232
218234
218236
218237
218238
218240
218242
218243
218244
218245
218246
218541
218548
218551
218552
218553
218560
218565
218566
218567
218568
219237
219250
219506
219568
219603
220093
220096
220099
220128
220180
220221
220340
220963
221321
221403
222066
222273
222274
227399
232514
232515
238523
238524
241230
241706
241707
273380
54300927
54357112
2596991
2596992
2596993
2596989
2596994
2596996
2410653
Work Orders
4151870-24
71111.13Maintenance Risk Assessments and Emergent Work Control
Condition Reports
18-15424
18-05353
16-12956
10-00150
15-08614
18-08748
17-09075
15-11063
18-05303
18-06515
Work Orders
22983
5006012
Procedures
Number
Title
Revision
Protected Equipment
Online Integrated Risk
Conduct of Maintenance
Miscellaneous
Documents
Number
Title
Date
Unit 2 Shutdown Safety Function Assessment
October 12, 2018
Schedulers Evaluation for PV Units 1, 2, and 3
October 4, 2018
71111.15Operability Determinations and Functionality Assessments
Condition Reports
16-15545
14-00406
17-003961
18-15424
18-18407
17-16139
18-07701
17-16139
18-18549
18-15398
18-18548
18-18551
Work Orders
49005003
4911770
Procedures
Number
Title
Revision
Refueling Water Tank (RWT) Operations
Procedures
Number
Title
Revision
HPSI Pumps Miniflow - Inservice Test
Locked Valve, Breaker, and Component Tracking
139
Drawings
Number
Title
Revision
01-M-SIP-001
P & I Diagram Safety Injection & Shutdown Cooling
System
71111.18Plant Modifications
Condition Reports
17-14312
18-08540
18-17888
Procedures
Number
Title
Revision
CN390-A00193
APS 225/35 Ton Single Failure Proof Trolley and
Controls Upgrade Installation Procedure (Vendor)
Reactor Vessel Head Removal and Installation
Drawings
Number
Title
Revision
or Date
219237
Main Hoist Gearcase - Structural Welds
217521
Main Hoist Equalizer - Structural Welds
215467
Main Hoist Bottom Block - Structural Welds
220963
Main Hoist Drum - Structural Welds
219237
Main/Auxiliary Hoist Brake Base - Structural Welds
214833
Trolley Truck Right - Structural Welds
241707
Trolley Seismic Restraints - Structural Welds
222274
Auxiliary Hoist Drum - Structural Welds
216795
Auxiliary Hoist Upper Block - Structural Welds
CN390-A00055
SRI-Trolley Frame, Trolley Frame Weldment
Miscellaneous
Documents
Number
Title
Revision
13-CN-390
Technical Requirements for Upgrading the Containment
Building Polar Crane to Single Failure Proof
Engineering Disposition for ENG-DMWO 3282819
(ZC-1311) Polar Crane Upgrade Modification
Miscellaneous
Documents
Number
Title
Revision
484-09275-MM/ac APS Palo Verde Nuclear Generating Station Acceptance
of the use of Nylatron for Sheaves on the Replacement
Trolley Assemblies
Various NDE Records for Crane System Critical Welds
S-16-0014
CFR 50.59 Screening / Evaluation
17-F026
Final LDCR for DMWO 3282819
Vendor
Documents
Number
Title
Revision
KNES 36676-80
Tripod Bail Load Test Procedure
CN-36676-21
NUREG 0612, Appendix C, Compliance Matrix for
Konecranes/Konecranes Supersafe' Single Failure
Proof Trolley
CN-3667622
Konecranes Nuclear Equipment and Services
Supersafe' Single Failure Proof Upgrade for Palo
Verde Nuclear Generating Station Polar Crane
3667623
Palo Verde Nuclear Generating Station Polar Crane
Compliance Matrix
13-CN390-A00183 Safety Analysis Report for Konecranes Supersafe'
Single Failure Proof Trolley Palo Verde Nuclear
Generating Station Polar Crane
2018-00899
Polar Crane QA Document folder
CN390-A00359
Polar Crane Unit 2 Factory Acceptance Test
13-CN-0390
Technical Requirements for Upgrading the Containment
Building Polar Crane to Single Failure Proof (Vendor
document with same document number and title as
licensee document above)
Modification
Number
Title
Revision
DMWO 3282819
(ZC-1311)
Polar Crane Upgrade Modification
Vendor
Calculation
Number
Title
Revision
CN390-A00181
Palo Verde Nuclear Generating Station - APS Main
Vendor
Calculation
Number
Title
Revision
Hoist Reeving Calculation
71111.19Post-Maintenance Testing
Condition Reports
18-17331
3157405
18-17372
18-17577
18-17740
Work Orders
22458
4909877
4911663
5065857
5079029
22458
Procedures
Number
Title
Revision
Check Valve Predictive Maintenance and Monitoring
Program
Check Valve Disassembly, Inspection, and Manual
Exercise
HPSI Pump and Check Valve Full Flow Test
Diesel Generator A Test
PPS Function Test - RPS/ESFAS Logic
Miscellaneous
Documents
Number
Title
Date
Last Measurement Deviation Report
October 30, 2018
71111.20Refueling and Other Outage Activities
Condition Reports
18-18582
18-18066
18-16233
18-16164
18-14363
18-15171
18-16311
Procedures
Number
Title
Revision
Miscellaneous Containment Building Heavy Loads
Operations Maintenance Activities
Fuel Transfer Machine
26A
Outage GOP
Control and Monitoring of PTBMs
Drawings
Number
Title
Revision
241707
SRI-Bracket, Seismic Restraint Weldment
Miscellaneous
Documents
Number
Title
Revision
282819
Engineering Disposition for Polar Crane Upgrade
Modification
18-18582-009
Engineering Evaluation
53702
Event Notification Worksheet: Personnel Medical
Condition
13-NS-A106
Missiles Generated by Natural Phenomena (Tornados)
Engineering
Reports Number
Title
Revision
2016-00631
Engineering Document Change
71111.22Surveillance Testing
Work Orders
4909840
28971
4932057
Procedures
Number
Title
Revision
Check Valve Non-intrusive Testing or Examination -
Inservice Test
Containment Spray Pumps and Check Valves -
Inservice Test
Containment Leakage Type B and C Testing
71114.04Emergency Action Level and Emergency Plan Changes
Procedures
Number
Title
Revision
Emergency Response Organization (ERO) Position
Checklists
Miscellaneous
Documents
Number
Title
Date
2-7799-CS/WP
Palo Verde Nuclear Generating Station (PVNGS), Units
1, 2, and 3 and Independent Spent Fuel Storage
Installation; Docket Nos. 50-528, 50-529, 50-530, and
2-44; License Nos. NPF-41, NPF-51, and NPF-74;
PVNGS Emergency Plan, Revision 62
September 25,
2018
Evaluation Tracking
Number 2018-002E
Effectiveness Evaluation Form, Revision 62, Palo
Verde Nuclear Generating Station Emergency Plan
August 29,
2018
2-07839-CS/MA
Palo Verde Nuclear Generating Station (PVNGS), Units
1, 2, and 3 and Independent Spent Fuel Storage
Installation; Docket Nos. 50-528, 50-529, 50-530, and
2-44; License Nos. NPF-41, NPF-51, and NPF-74;
PVNGS Emergency Plan, Revision 63
December 6,
2018
71152Problem Identification and Resolution
Condition Reports
18-18408
18-16091
18-07701
17-16139
18-18420
18-02569
18-16928
18-12217
Work Orders
2891465
26510
3490456
250774
3374818
3374819
Procedures
Number
Title
Revision
Reactor Vessel O-Ring Replacement
Operations Condition Reporting Process and Operability
Determination/Functional Assessment
Miscellaneous
Documents
Number
Title
Revision
or Date
ODMI: Unit 1 Reactor Vessel Flange Indication of
Leakage from Inner O-ring
ODMI: Unit 3 Reactor Vessel Flange Indication of
Leakage from Inner O-ring
May 22, 2018
17-16139-004
Engineering Evaluation
November 17,
2017
Vendor
Documents
Number
Title
Date
Technetics: Customer Contact Report
November 20,
Vendor
Documents
Number
Title
Date
2018
71153Follow-up of Events and Notices of Enforcement Discretion
Condition Reports
18-02605
18-08466
18-00202
Miscellaneous
Documents
Number
Title
Revision
or Date
18-02605-009
Unit 1 Main Turbine Excitation Trip Root Cause
Evaluation
18-00202-027
Engineering Evaluation
July 3, 2018
CRDR 4531542
Inability of ADV 2JSGAHV0184 to Stroke Open from the
Control Room and Remote Shutdown Panel
July 31, 2014
Information Request August 21, 2018
Notification of Inspection and Request for Information
Palo Verde Nuclear Generating Station, Unit 2
NRC Inspection Report 05000529/2018004
INSERVICE INSPECTION DOCUMENT REQUEST
Inspection Dates: October 11 - 18, 2018
Inspector: Wayne Sifre, Senior Reactor Inspector
A. Information Requested for the In-Office Preparation Week
The following information should be sent to the Region IV office in hard copy or electronic
format (ims.certrec.com preferred), in care of Wayne Sifre, by October 3, 2018, to facilitate
the selection of specific items that will be reviewed during the onsite inspection weeks. The
inspector will select specific items from the information requested below and then request
from your staff additional documents needed during the onsite inspection week (Section B
of this enclosure). We ask that the specific items selected from the lists be available and
ready for review on the first day of inspection. Please provide requested documentation
electronically if possible. If requested documents are large and only hard copy formats are
available, please inform the inspector, and provide subject documentation during the first
day of the onsite inspection.
If you have any questions regarding this information request, please call the inspector as
soon as possible.
PAPERWORK REDUCTION ACT STATEMENT
This letter does not contain new or amended information collection requirements
subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). Existing
information collection requirements were approved by the Office of Management and
Budget, Control Number 31500011. 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.
In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this
letter and its enclosure will be available electronically for public inspection in the NRC
Public Document Room or from the Publicly Available Records (PARS) component of
NRCs document system (ADAMS). ADAMS is accessible from the NRC web site at
http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
On October 11, 2018, a reactor inspector from the Nuclear Regulatory Commissions
Region IV office will perform the baseline inservice inspection at Palo Verde Nuclear
Generating Station, Unit 2, using NRC Inspection Procedure 71111.08, "Inservice
Inspection Activities. Experience has shown that this inspection is a resource intensive
inspection both for the NRC inspectors and your staff. The date of this inspection may
change dependent on the outage schedule you provide. In order to minimize the impact to
your onsite resources and to ensure a productive inspection, we have enclosed a request
for documents needed for this inspection. These documents have been divided into two
groups. The first group (Section A of the enclosure) identifies information to be provided
prior to the inspection to ensure that the inspector is adequately prepared. The second
group (Section B of the enclosure) identifies the information the inspector will need upon
arrival at the site. It is important that all of these documents are up to date and complete in
order to minimize the number of additional documents requested during the preparation
and/or the onsite portions of the inspection. We have discussed the schedule for these
inspection activities with your staff and understand that our regulatory contact for this
inspection will be Mr. Del Elkinton of your licensing organization. The tentative inspection
schedule is as follows:
Preparation week: October 3 - 10, 2018
Onsite weeks: October 11 - 18, 2018
Our inspection dates are subject to change based on your updated schedule of outage
activities. If there are any questions about this inspection or the material requested, please
contact the lead inspector Wayne Sifre at 817-200-1193.
(mail to: Wayne.Sifre@nrc.gov ).
A.1
ISI/Welding Programs and Schedule Information
a) A detailed schedule (including preliminary dates) of:
i. Nondestructive examinations planned for ASME Code Class Components
including containment, performed as part of your ASME Section XI, risk informed
(if applicable), and augmented inservice inspection programs during the
upcoming outage.
ii. Examinations planned for Alloy 82/182/600 components that are not included in
the Section XI scope (if applicable).
iii. Examinations planned as part of your boric acid corrosion control program
(mode 3 walk downs, bolted connection walk downs, etc.)
iv. Welding activities that are scheduled to be completed during the upcoming
outage (ASME Class 1, 2, or 3 structures, systems, or components). Include the
weld identification number, description of weld, category, class, type of exam and
procedure number, and date of examination.
b) A copy of ASME Section XI, Code Relief Requests and associated NRC safety
evaluations applicable to the examinations identified above. A list of ASME Code
Cases currently being used to include the system and/or component the Code Case
is being applied to.
c) A list of nondestructive examination reports which have identified recordable or
rejectable indications on any ASME Code Class components since the beginning of
the last refueling outage. This should include the previousSection XI pressure
test(s) conducted during start up and any evaluations associated with the results of
the pressure tests.
d) A list including a brief description (e.g., system, code class, weld category,
nondestructive examination performed) associated with the repair/replacement
activities of any ASME Code Class component since the beginning of the last
outage and/or planned this refueling outage.
e) If reactor vessel weld examinations required by the ASME Code are scheduled to
occur during the upcoming outage, provide a detailed description of the welds to be
examined and the extent of the planned examination. Please also provide
applicable procedures that will be used to conduct these examinations.
f) Copy of any 10 CFR Part 21 reports applicable to structures, systems, or
components within the scope of Section XI of the ASME Code that have been
identified since the beginning of the last refueling outage.
g) A list of any temporary non-code repairs in service (e.g., pinhole leaks).
h) Please provide copies of the most recent self-assessments for the inservice
inspection, welding, and Alloy 600 programs.
i)
Copy of the procedures for NDE and welding techniques that will be used during the
outage.
A.2
Boric Acid Corrosion Control Program
a) Copy of the procedures that govern the scope, equipment, and implementation of the
inspections required to identify boric acid leakage and the procedures for boric acid
leakage/corrosion evaluation.
b) Please provide a list of leaks (including code class of the components) that have been
identified since the last refueling outage and associated corrective action
documentation. If during the last cycle, the unit was shut down, please provide
documentation of containment walk down inspections performed as part of the boric
acid corrosion control program.
A.3
Additional Information Related to all Inservice Inspection Activities
a) A list with a brief description of inservice inspection, and boric acid corrosion control
program (e.g., condition reports) entered into your corrective action program since the
beginning of the last refueling outage. For example, a list based upon data base
searches using key words related to piping such as: inservice inspection, ASME
Code,Section XI, NDE, cracks, wear, thinning, leakage, rust, corrosion, boric acid, or
errors in piping examinations.
b) Provide training (e.g., Scaffolding, Fall Protection, FME, Confined Space) if they are
required for the activities described in A.1 through A.4.
c) Please provide names and phone numbers for the following program leads:
Inservice inspection (examination, planning)
Containment examinations
Snubbers and supports
Site welding engineer
Boric acid corrosion control program
B.
Information to be Provided Onsite to the Inspector at the Entrance Meeting
(October 11, 2018):
B.1
Inservice Inspection / Welding Programs and Schedule Information
a) Updated schedules for inservice inspection/nondestructive examination activities,
including steam generator tube inspections, planned welding activities, and schedule
showing contingency repair plans, if available.
b) For ASME Code Class welds selected by the inspector from the lists provided from
section A of this enclosure, please provide copies of the following documentation for
each subject weld:
i. Weld data sheet (traveler).
ii. Weld configuration and system location.
iii. Applicable Code Edition and Addenda for weldment.
iv. Applicable Code Edition and Addenda for welding procedures.
v. Applicable welding procedures used to fabricate the welds.
vi. Copies of procedure qualification records (PQRs) supporting the weld procedures
from B.1.b.v.
vii. Copies of welders performance qualification records (WPQ).
viii. Copies of the nonconformance reports for the selected welds (If applicable)
ix. Radiographs of the selected welds and access to equipment to allow viewing
radiographs (if radiographic testing was performed).
x. Copies of the preservice examination records for the selected welds.
xi. Readily accessible copies of nondestructive examination personnel qualifications
records for reviewing.
c) For the inservice inspection related corrective action issues selected by the inspectors
from section A of this enclosure, provide a copy of the corrective actions and supporting
documentation.
d) For the nondestructive examination reports with relevant conditions on ASME Code
Class components selected by the inspectors from Section A above, provide a copy of
the examination records, examiner qualification records, and associated corrective
action documents.
e) A copy of (or ready access to) most current revision of the inservice inspection program
manual and plan for the current interval.
f) For the nondestructive examinations selected by the inspectors from Section A of this
enclosure, provide a copy of the nondestructive examination procedures used to
perform the examinations (including calibration and flaw characterization/sizing
procedures). For ultrasonic examination procedures qualified in accordance with
ASME Code,Section XI, Appendix VIII, provide documentation supporting the
procedure qualification (e.g., the EPRI performance demonstration qualification
summary sheets). Also, include qualification documentation of the specific equipment
to be used (e.g., ultrasonic unit, cables, and transducers including serial numbers) and
nondestructive examination personnel qualification records.
B.2
Boric Acid Corrosion Control Program
a) Please provide boric acid walk down inspection results, an updated list of
boric acid leaks identified so far this outage, associated corrective action
documentation, and overall status of planned boric acid inspections.
b) Please provide any engineering evaluations completed for boric acid leaks
identified since the end of the last refueling outage. Please include a status
of corrective actions to repair and/or clean these boric acid leaks. Please
identify specifically which known leaks, if any, have remained in service or
will remain in service as active leaks.
B.3
Codes and Standards
a) Ready access to (i.e., copies provided to the inspector(s) for use during the
inspection at the onsite inspection location, or room number and location
where available) Applicable Editions of the ASME Code (Sections V, IX,
and XI) for the inservice inspection program and the repair/replacement
program.
b) Copy of the performance demonstration initiative (PDI) generic procedures
with the latest applicable revisions that support site qualified ultrasonic
examinations of piping welds and components (e.g., PDI-UT-1, PDI-UT-2,
PDI-UT-3, PDI-UT-10, etc.).
c) Boric Acid Corrosion Guidebook Revision 1 - EPRI Technical Report 1000975.
SUNSI Review:
ADAMS:
Non-Publicly Available
Non-Sensitive
Keyword:
By: JDixon
Yes No
Publicly Available
Sensitive
OFFICE
DRP/SRI
DRP/RI
DRP/RI
C:DRS/EB1
C:DRS/EB2
C:DRS/OB
NAME
CPeabody
DReinert
DYou
VGaddy
FRamirez
GWerner
SIGNATURE
DRR
DDY
vgg
FCR
GEW
DATE
1/14/2019
1/11/2019
1/15/2019
1/24/19
1/25/2019
01/28/2019
OFFICE
C:DRS/PS2
TL:IPAT
C:DRP/D
NAME
HGepford
RKellar
NOKeefe
SIGNATURE
hjg
RLK
DATE
1/24/2019
1/23/2019
1/29/19