IR 05000275/2018001
| ML18114A835 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 04/24/2018 |
| From: | Mark Haire NRC/RGN-IV/DRP/RPB-A |
| To: | Welsch J Pacific Gas & Electric Co |
| Mark Haire | |
| References | |
| IR 2018001 | |
| Download: ML18114A835 (43) | |
Text
ril 24, 2018
SUBJECT:
DIABLO CANYON POWER PLANT - NRC INSPECTION REPORT 05000275/2018001, 05000323/2018001, and 07200026/2018001
Dear Mr. Welsch:
On March 31, 2018, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Diablo Canyon Power Plant Units 1 and 2. On April 18, 2018, the NRC inspectors discussed the results of this inspection with you 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.
One of these findings involved a violation of NRC requirements. The NRC is treating this violation as a non-cited violation (NCV) consistent with Section 2.3.2.a of the Enforcement Policy.
If you contest the violation or significance of this 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 Diablo Canyon Power Plant.
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 Diablo Canyon Power Plant. 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/
Mark S. Haire, Chief Project Branch A Division of Reactor Projects Docket Nos. 50-275, 50-323, and 72-026 License Nos. DPR-80, DPR-82, and SNM-2511
Enclosure:
Inspection Report 05000275/2018001, 05000323/2018001, and 07200026/2018001 w/ Attachments:
1. Documents Reviewed 2. RFI for Resident Inspection - 1st Quarter 3. RFI for O
Inspection Report
Docket Numbers: 05000275, 05000323, and 07200026 License Numbers: DPR-80, DPR-82, and SNM-2511 Report Numbers: 05000275/2018001, 05000323/2018001, and 07200026/2018001 Enterprise Identifier: I-2018-001-0001 Licensee: Pacific Gas and Electric Company Facility: Diablo Canyon Power Plant, Units 1 and 2, and associated Independent Spent Fuel Storage Installation (ISFSI)
Location: Avila Beach, California Inspection Dates: January 1, 2018 to March 31, 2018.
Inspectors: C. Newport, Senior Resident Inspector J. Reynoso, Resident Inspector W. Sifre, Senior Reactor Inspector (ISI)
L. Brookhart, Senior ISFSI inspector, FCDB (ISFSI)
N. Greene, PhD, Senior Health Physicist (Radiation Safety)
J. ODonnell, CHP, Health Physicist (Radiation Safety)
Approved By: Mark S. Haire, Chief, Project Branch A, Division of Reactor Projects 1 Enclosure
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Diablo Canyon Power Plant Units 1 and 2 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 Improper Troubleshooting Results in Reactor Trip Signal and Loss of Source Range Nuclear Instrument Power Cornerstone Significance Cross-cutting Report Section Aspect Mitigating Green H.5 - Human 71111.20 - Refueling and Systems NCV 05000323/2018001-001 Performance - Other Outage Activities Closed Work Management The inspectors reviewed a Green, self-revealed non-cited violation of Technical Specification 5.4.1.a Procedures, because PG&E personnel failed to follow the requirements of MA1.DC54, Conduct of Maintenance, Revision 15. Specifically, on March 20, 2018, with the reactor in Mode 3 during informal troubleshooting of high background count rate on source range nuclear instrument (NI) NI-32, PG&E personnel caused a short in NI cabinet B resulting in a blown fuse and the loss of power to the cabinet. This resulted in the loss of power to power range NI-42, intermediate range NI-36, source range NI-32, a reactor trip signal, a turbine trip signal, and all associated reactor protection interlocks. Power was automatically removed from the remaining source range NI due to reactor protection interlock P-10, resulting in no safety-related source range NI indication being available for control room operators.
Failure to Follow Operating Experience Procedures Results in Inadequate Screen of Operating Experience Report Cornerstone Significance Cross-cutting Report Section Aspect Mitigating Green None 71152 - Problem Systems FIN 05000323/2018001-02 Identification and Closed Resolution The inspectors identified a finding of very low safety significance (Green) because PG&E personnel failed to follow the requirements of OM4.ID3, Operating Experience Program,
Revision 20. Specifically, PG&E personnel failed to screen relevant operating experience relating to a safety-related centrifugal charging pump (CCP) journal bearing failure due to non-metallic anti-rotation pin shear failure. This operating experience notice was received by PG&E September 2011 and was not screened per OM4.ID3, Operating Experience Program, preventing actions from being identified and implemented that could have eliminated vulnerabilities and prevented a similar event from occurring at DCPP. On November 11, 2017,
CCP 2-1 was declared inoperable and determined to be non-functional due to a damaged journal bearing caused by non-metallic, anti-rotation pin shear failure.
Additional Tracking Items Type Issue number Title Report Section Status URI 07200026/2016001-01 Applicability of required 60855.1 - Operation of Closed NDE inspections on the an ISFSI at Operating Lift Yoke in accordance Plants with ANSI N14.6
PLANT STATUS
Units 1 and 2 began the inspection period at full power.
On February 11, 2018, Unit 2 was shut down for a planned refueling outage. On March 20, 2018, Unit 2 returned to operation and began a controlled power ascension; it returned to full power on March 26, 2018.
Units 1 and 2 operated at or near full power for the remainder of the inspection period.
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. 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
71111.01 - Adverse Weather Protection Impending Severe Weather
The inspectors evaluated readiness for impending adverse weather conditions following heavy rains and high winds on January 9, 2018.
71111.04 - Equipment Alignment Partial Walkdown
The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:
- (1) Unit 2, Mode 6 boration flowpath on March 2, 2018
- (2) Unit 2, containment spray on March 7, 2018
- (3) Unit 2, auxiliary feedwater on March 18, 2018
- (4) Unit 2, safety injection on March 23, 2018
- (5) Unit 2, spent fuel cooling on March 29, 2018
Complete Walkdown (1 Sample)
The inspectors evaluated system configurations during a complete walkdown of the Unit 2, residual heat removal system on March 22, 2018.
71111.05AQ - Fire Protection Annual/Quarterly Quarterly Inspection
The inspectors evaluated fire protection program implementation in the following selected areas:
- (1) Unit 1, auxiliary building 64 foot elevation on January 4, 2018
- (2) Unit 2, containment 140 foot elevation on February 13, 2018
- (3) Unit 2, containment 117 foot elevation on February 15, 2018
- (4) Unit 2, containment 91 foot elevation on February 22, 2018
- (5) Units 1 and 2, saltwater intake on March 2, 2018
71111.06 - Flood Protection Measures Internal Flooding
The inspectors evaluated internal flooding mitigation protections associated with the Units 1 and 2, circulating water and saltwater systems on March 20, 2018.
Cables (1 Sample)
The inspectors evaluated cable submergence protection in the following areas:
- (1) auxiliary saltwater pump vaults on February 17, 2018
71111.07 - Heat Sink Performance Heat Sink
The inspectors evaluated Unit 2, component cooling water heat exchanger 2-1 performance on March 1, 2018.
71111.08 - Inservice Inspection Activities
The inspectors evaluated pressurized water reactor non-destructive testing by reviewing the following examinations from February 22 to March 1, 2018:
- (1) Ultrasonic Examinations a) residual heat removal line 45 (Weld RB-45-3-04)b) residual heat removal line 45 (Weld RB-45-2)c) residual heat removal line 45 (Weld RB-45-6)d) residual heat removal line 45 (Weld RB-45-7)e) residual heat removal line 46 (Weld RB-46-4)f) residual heat removal line 46 (Weld RB-46-7)g) residual heat removal line 46 (Weld RB-46-8)
- (2) Dye Penetrant Examinations a) centrifugal charging pump discharge header line S6-1456-6 (welded lug attachments23-36R)b) main steam valve FCV-41 replacement weld c) main steam MS-2-1018 socket weld d) main steam MS-2-2016 socket weld
- (3) Phased Array Ultrasonic Examination a) residual heat removal line 46 (266373-TR-002 structural weld overlay)
- (4) Visual (VT-3)a) containment liner penetration 67, equipment hatch bolting
- (5) Gas Tungsten Arc Weld - Machine a) residual heat removal line 46 (266373-TR-002 structural weld overlay)
The Inspectors evaluated the licensees boric acid control program performance.
71111.11 - Licensed Operator Requalification Program and Licensed Operator Performance Operator Requalification
The inspectors observed and evaluated a crew of licensed operators in the plants simulator during training on refueling outage testing activities on February 5, 2018.
Operator Performance (1 Sample)
The inspectors observed and evaluated operator performance during the following activities:
- (1) Units 1 and 2, auxiliary saltwater cross flow test procedure to determine the capability to provide auxiliary saltwater flow to opposite units on January 22, 2018
- (2) Unit 2, entering coast down procedure, including the pre-job briefing, on January 22, 2018
- (3) Unit 1, test procedure to stroke and verify position of emergency core cooling system valve 8700B on January 25, 2018
- (4) Unit 2, reactor coolant system drain down for reactor head removal on February 14, 2018
71111.12 - Maintenance Effectiveness Routine Maintenance Effectiveness
The inspectors evaluated the effectiveness of routine maintenance activities associated with the following equipment and/or safety significant functions:
- (1) Unit 2, 2R20 motor operated valve maintenance
- (2) Units 1 and 2, auxiliary building ventilation fans and dampers
71111.13 - Maintenance Risk Assessments and Emergent Work Control
The inspectors evaluated the risk assessments for the following planned and emergent work activities:
- (1) Unit 2, vital battery charger 2-1, inspection and maintenance outage on January 17, 2018
- (2) Unit 2, reactor coolant system, reduced reactor coolant inventory window of 2R20 refueling outage on February 13-14, 2018
- (3) Unit 2, reactor coolant system, yellow risk during drain down for reactor head installation on March 6, 2018
- (4) Unit 2, reactor protection system, seismic trigger protection set 2, maintenance and calibration on March 30, 2018
71111.15 - Operability Determinations and Functionality Assessments
The inspectors evaluated the following operability determinations and functionality assessments:
- (1) Unit 1, auxiliary feedwater level control valve position indicator LCV-113 output failure on January 16-17, 2018
- (2) Unit 2, containment nitrogen system on January 24-25, 2018
- (3) Unit 2, auxiliary building ventilation system back damper M-20 stuck open on February 12, 2018
- (4) Unit 2, reactor coolant pump 2-1 failure to stop on demand from control room on February 12, 2018
- (5) Unit 2, auxiliary feedwater valve LCV-115 output failure on February 26, 2018
- (6) Unit 2, main steam insolation valve FCV-44 excessive wear on check valve disc on March 12, 2018
71111.18 - Plant Modifications
The inspectors evaluated the following temporary or permanent modifications:
- (1) Short duration modification on Unit 2, fuel transfer up-ender position switch inside containment on February 16, 2018
- (2) Emergency core cooling system valve external NAMCO limit switch removal on March 6, 2018
71111.19 - Post Maintenance Testing
The inspectors evaluated the following post maintenance tests:
- (1) Unit 2, containment spray pump 2-1, outboard bearing anti-rotation pin replacement on January 5, 2018
- (2) Units 1 and 2, replacement of fire suppression solenoid valve SV-103 on January 10, 2018
- (3) Unit 2, auxiliary saltwater pump 2-1, motor and pump replacement on February 21, 2018
- (4) Unit 2, safety injection pump 2-1, shaft coupling inspection and high vibration on March 7-8, 2018
- (5) Unit 1, centrifugal charging pump 1-2, testing following maintenance of 4 kV breaker and relays on March 27, 2018
71111.20 - Refueling and Other Outage Activities
The inspectors evaluated refueling outage 2R20 activities from February 11, 2018 to March 21, 2018.
71111.22 - Surveillance Testing The inspectors evaluated the following surveillance tests: Routine
- (1) Units 1 and 2, flow control valve FCV-601, auxiliary saltwater cross-tie valve flow test per STP M-26A on January 22, 2018
- (2) Unit 2, integrated test of engineered safeguards and diesel generators per procedure STP M-15 on February 13, 2018
- (3) Unit 2, rod drop measurement testing per procedure STP R-1B on March 19, 2018
In-service (1 Sample)
- (1) Unit 2, main steam lead 1, safety and relief valve testing per procedure STP M-77 on
February 6, 2018 Containment Isolation Valve (2 Samples)
- (1) Leakage testing on containment isolation valves associated with Unit 2, penetration 51 per STP V-651B on February 28, 2018
- (2) Unit 2, containment integrated leak rate testing per procedure STP M-7 on March 12,
RADIATION SAFETY
71124.01 - Radiological Hazard Assessment and Exposure Controls Radiological Hazard Assessment
===The inspectors evaluated radiological hazards assessments and controls.
Instructions to Workers===
The inspectors evaluated worker instructions.
Contamination and Radioactive Material Control (1 Sample)
The inspectors evaluated contamination and radioactive material controls.
Radiological Hazards Control and Work Coverage (1 Sample)
The inspectors evaluated radiological hazards control and work coverage.
High Radiation Area and Very High Radiation Area Controls (1 Sample)
The inspectors evaluated risk-significant high radiation area and very high radiation area controls.
Radiation Worker Performance and Radiation Protection Technician Proficiency (1 Sample)
The inspectors evaluated radiation worker performance and radiation protection technician proficiency.
71124.02 - Occupational As Low As Reasonably Achievable (ALARA) Planning and Controls Implementation of ALARA and Radiological Work Controls
The inspectors reviewed ALARA practices and radiological work controls by reviewing the following activities and/or their ALARA plans:
- (1) RWP 17-2024, 2R20 Guide Cards
- (3) RWP 17-2070, 2R20 RHR Weld Overlay
Radiation Worker Performance (1 Sample)
The inspectors evaluated radiation worker and radiation protection technician performance.
OTHER ACTIVITIES - BASELINE
71151 - Performance Indicator Verification
The inspectors verified licensee performance indicators submittals listed below:
- (1) IE01: Unplanned Scrams per 7000 Critical Hours Sample (01/01/2017-12/31/2017)
[1 sample per unit]
- (2) IE03: Unplanned Power Changes per 7000 Critical Hours Sample (01/01/2017-12/31/2017) [1 sample per unit]
- (3) IE04: Unplanned Scrams with Complications (USwC) Sample (01/01/2017-12/31/2017) [1 sample per unit]
- (4) OR01: Occupational Exposure Control Effectiveness Sample (05/01/2017-12/31/2017)
- (5) PR01: Radiological Effluent Technical Specifications/Offsite Dose Calculation Manual Occurrences (RETS/ODCM) Sample (05/01/2017-12/31/2017)
71152 - Problem Identification and Resolution Annual Follow-up of Selected Issues
The inspectors reviewed the licensees implementation of its corrective action program related to the following issues:
- (1) Diablo Canyon Power Plant (DCPP) safety related 4 kV motor journal bearing degradation, including centrifugal charging pump 2-1 outboard journal bearing damage caused by bearing rotation and 2-1 containment spray pump outboard journal bearing minor rotation, on February 13, 2018.
- (2) The inspectors performed an in-depth review of the licensees evaluation and corrective actions related to personnel not adhering to procedures or standards. The inspectors primary focus centered on Unit 2 refueling activities beginning February 1 through March 21, 2018, during which the licensee relied on vendor or temporary additional workers to help with scheduled maintenance activities. The basis for this inspection focused on a review of corrective action program issues which were documented associated with the Unit 2 refueling outage.
71153 - Follow-up of Events and Notices of Enforcement Discretion Events
- (1) The inspectors evaluated a steam leak at the suction line of main feedwater pump 2-2 and the licensees response on January 16, 2018.
- (2) The inspectors evaluated a screen wash pump motor fire and the licensees response on January 25,
OTHER ACTIVITIES
- TEMPORARY INSTRUCTIONS, INFREQUENT AND ABNORMAL 60855.1 - Operation of an Independent Spent Fuel Storage Installation (ISFSI)
The inspectors closed an Unresolved Item (URI) from a routine ISFSI inspection conducted in September of 2016 (see Inspection Results).
INSPECTION RESULTS
Improper Troubleshooting Results in Reactor Trip Signal and Loss of Source Range Nuclear Instrument Power Cornerstone Significance Cross-cutting Report Aspect Section Mitigating Systems Green H.5 - Human
71111.20 - NCV 05000323/2018001-01 Performance - Refueling and
Closed Work Other Outage Management Activities The inspectors reviewed a Green, self-revealed non-cited violation of Technical Specification (TS) 5.4.1.a Procedures, because PG&E personnel failed to follow the requirements of MA1.DC54, Conduct of Maintenance, Revision 15. Specifically, on March 20, 2018, with the reactor in Mode 3 during informal troubleshooting of high background count rate on source range nuclear instrument (NI) NI-32, PG&E personnel caused a short in NI cabinet B resulting in a blown fuse and the loss of power to the cabinet. This resulted in the loss of power to power range NI-42, intermediate range NI-36, source range NI-32, a reactor trip signal, a turbine trip signal, and all associated reactor protection interlocks. Power was automatically removed from the remaining source range NI due to reactor protection interlock P-10, resulting in no safety-related source range NI indication being available for control room operators.
Description:
On March 20, 2018, with the plant in Mode 3 at normal operating temperature and pressure, source range detector high voltage alignment of the NI-32 source range instrument could not be achieved during performance of STP I-4B4, Determination of Source Range Detector Characteristic Curves. This alignment is performed to establish appropriate baseline NI channel settings prior to reactor startup. An informal troubleshooting plan was subsequently developed to diagnose the source of potential electronic noise within the electronic circuitry that could be impacting the alignment. The informal troubleshooting plan consisted of installing an insulating metallic Faraday blanket around NI-32 power cables in the vicinity of the filter. The plan did not meet the requirements of MA1.DC54, Conduct of Maintenance, which requires, in part, that all work on plant systems, structures, or components (SSCs)should be performed using appropriate documentation such as work orders, notifications, procedures, or design drawings. No work order or work package was developed prior to the installation of the Faraday blanket. During installation of the Faraday blanket, the blanket inadvertently made contact with an exposed terminal causing a short in NI cabinet B and the main power fuse to fail, removing power from the cabinet. The loss of power to the cabinet resulted in the loss of power to power range NI-42, intermediate range NI-36, and source range NI-32. A second power range NI (NI-44) was already out of service at the time of the event. The loss of power to two of four power range NIs resulted in the two out of four coincidence logic being met for the reactor protection system (RPS) power range high flux trip, hi power trip, and low power trip. As a result, the reactor trip breakers opened, and the main turbine tripped. At the time of the event, all control rods were unlatched and fully inserted.
Additionally, the coincidence logic was made up for interlocks P-10, P-9, and P-8 resulting in power being removed from both source range NIs. Source range NIs are safety-related instruments relied upon in Mode 3 to protect against rod withdrawal accidents (source range hi flux RPS trip), inadvertent criticality from unanticipated reactor coolant system (RCS) dilution, and to provide indication to plant operators. As a result of the event and in accordance with applicable TS action statements, PG&E personnel stopped all activities that could add positive reactivity to the RCS, verified shutdown margin, and verified that the reactor trip breakers were open. For the duration of loss of power to the source range NIs, PG&E personnel were able to rely upon non-safety-related, post-accident monitoring NIs to provide indication of neutron activity with the core. Power was restored to one of the two safety-related source range NIs after approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Additional troubleshooting and engineering evaluations allowed PG&E personnel to restore power to NI cabinet B and complete STP I-4B4.
Corrective Actions: After the issue occurred, PG&E initiated corrective actions including replacement of the blown fuse and additional troubleshooting in accordance with applicable procedures.
Corrective Action References: Notification 50971776
Performance Assessment:
Performance Deficiency: The inspectors determined that PG&Es failure to follow MA1.DC54, Conduct of Maintenance, Revision 15, when diagnosing high background count rate on source range NI-32, was a performance deficiency within PG&Es ability to foresee and correct.
Screening: The performance deficiency was considered to be more than minor because it impacted the equipment performance attribute of the Mitigating Systems cornerstone and its objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, improper troubleshooting of source range NI-32 resulted in the loss of power to NI cabinet B and a subsequent reactor trip signal, associated interlocks, and a loss of power to both source range NIs which were required to be in service by TS at the time of the event.
Significance: Because the plant was in Mode 3 at normal operating temperature and pressure with decay heat removal secured, the finding was evaluated in accordance with Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings-At-Power, dated June 19, 2012. The inspectors determined that the finding screened to Green because it did not result in an actual loss of safety function of at least a single train for greater than its TS allowed outage time.
Cross-cutting Aspect: This finding is related to the cross-cutting area of Human Performance
- Work Management [H.5] because PG&E personnel did not implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority.
Specifically, during troubleshooting, PG&E personnel failed to develop an appropriate work order or instruction that would protect against inadvertently causing electrical transients within the NI cabinet.
Enforcement:
Violation: TS 5.4.1.a, Procedures, requires, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide (RG) 1.33, Revision 2, Appendix A. RG 1.33, Appendix A requires, in part, that maintenance that can affect the performance of safety-related equipment should be properly pre-planned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances. Procedure MA1.DC54, Conduct of Maintenance, requires, in part, that all work on plant systems, structures, or components (SSCs) should be performed using appropriate documentation such as work orders, notifications, procedures, or design drawings.
Contrary to the above, on March 20, 2018, no work order or work package was developed prior to the installation of a Faraday blanket in NI cabinet B. During installation of the Faraday blanket, the blanket inadvertently made contact with an exposed terminal causing a short in NI cabinet B and the main power fuse to actuate, removing power from the cabinet and causing an RPS reactor trip actuation, turbine trip, and loss of power to both safety-related source range NI channels. After the issue occurred, PG&E entered it into the corrective action program as Notification 50971776 and initiated corrective actions including replacement of the blown fuse and additional troubleshooting.
Disposition: This violation is being treated as a non-cited violation consistent with Section 2.3.2 of the Enforcement Policy.
Failure to Follow Operating Experience Procedures Results in Inadequate Screen of Operating Experience Report Cornerstone Significance Cross-cutting Report Aspect Section Mitigating Systems Green None
71152 - FIN 05000323/2018001-02 Problem Closed
Identification and Resolution The inspectors identified a finding of very low safety significance (Green) because PG&E personnel failed to follow the requirements of OM4.ID3, Operating Experience Program, Revision 20. Specifically, PG&E personnel failed to screen relevant operating experience relating to a safety-related centrifugal charging pump (CCP) journal bearing failure due to non-metallic anti-rotation pin shear failure. This operating experience notice was received by PG&E September 2011 and was not screened per OM4.ID3, Operating Experience Program, preventing actions from being identified and implemented that could have eliminated vulnerabilities and prevented a similar event from occurring at DCPP. On November 11, 2017, CCP 2-1 was declared inoperable and determined to be non-functional due to a damaged journal bearing caused by non-metallic, anti-rotation pin shear failure.
Description:
On September 7, 2011, PG&E received an external operating experience report describing the failure of CCP outboard and inboard motor bearings caused by a non-metallic anti-rotation pin age-related failure. The failure of the anti-rotation pin allowed the journal style bearings to rotate resulting in the oil slinger ring being prevented from adequately lubricating the bearing.
This lack of bearing lubrication subsequently caused bearing damage and failure. The applicable PG&E operating experience procedure in place at the time, OM4.ID3, Operating Experience Program, Revision 20, required that operating experience reports received by the site be screened per applicability per Section 5.3, Screening and Disseminating OPEX Documents. Section 5.3 required that relevant operating experience is screened by subject matter experts and appropriate corrective actions assigned to eliminate vulnerabilities and prevent a similar event from occurring at DCPP. When questioned, PG&E personnel were unable to identify any record of operating experience screening or corrective action documents generated as a result of the receipt of external operating experience report.
On November 7, 2017, a CCP 2-1 outboard motor bearing alarm was received in the DCPP Unit 2 control room. Further investigation revealed erratic bearing temperature indications and oil slinger ring movement, and a lack of oil present in the outboard bearing sight glass or tube while the motor was running. CCP 2-1 was subsequently secured and declared inoperable and an investigation was commenced. Upon disassembly of the outboard motor bearing for CCP 2-1, the bearing non-metallic anti-rotation pin was found broken in two pieces and the bearing was found shifted approximately 25 degrees. The bearing babbitt lining surface was damaged due to rotor-to-bearing contact. A PG&E cause evaluation determined that shear failure of the non-metallic, anti-rotation pin allowed the outboard motor bearing to rotate in the bearing housing, impinging the bearing on the oil slinger ring and preventing it from lubricating the pump shaft adequately. This resulted in damage to the bearing and CCP 2-1 being declared inoperable and unavailable for use for a period of approximately 4 days. Prior to the November 7, 2011, CCP 2-1 start and subsequent bearing failure, CCP 2-1 was last operated on September 18, 2017.
Corrective Actions: After the issue was identified by the inspectors, PG&E entered the issue into the corrective action program and initiated corrective actions including replacement of the CCP 2-1 outboard bearing, periodic inspection of all safety-related pump motors with susceptible bearings to verify that bearings have not rotated, and scheduled replacement of all bearing anti-rotation pins susceptible to a similar failure mechanism. The licensee determined that the Operating Experience coordinator position had been vacant between approximately December 15, 2011, and February 9, 2012, and that over 300 operating experience reports were received by the site which were not appropriately screened during this time period. Each of these reports will be properly screened for relevance to the site.
Corrective Action References: Notifications 50949662 and 50958899
Performance Assessment:
Performance Deficiency: The inspectors determined that PG&Es failure to screen relevant operating experience relating to a safety-related CCP journal bearing failure in accordance with OM4.ID3, Operating Experience Program, Revision 20, was a performance deficiency within PG&Es ability to foresee and correct.
Screening: This performance deficiency was considered to be more than minor because it impacted the equipment performance attribute of the Mitigating Systems cornerstone and its objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to appropriately screen operating experience prevented the implementation of actions that could have prevented a subsequent outboard journal bearing failure of safety-related CCP 2-1 and resulted in the pump being declared inoperable and non-functional.
Significance: The finding was evaluated in accordance with Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings-At-Power, dated June 19, 2012. The inspectors determined that a detailed risk evaluation by an NRC senior reactor analyst was required since the finding was associated with a loss of function of at least a single train of safety-related equipment for greater than its TS allowed outage time.
The regional senior reactor analyst performed a Phase 3 SDP analysis for the finding. The analyst used the DCPP 1 & 2 SPAR model, Version 8.54, to evaluate the risk of the finding.
The analyst noted that there was a difference between the success criteria in the Plant Information e-Book and the Fault Tree FAB, Feed and Bleed, regarding the success of the charging system during feed and bleed operations. The e-Book suggested that any single High Pressure Safety Injection or Charging Pump could be used to successfully complete the function. However, the Fault Tree FAB failed the feed and bleed function if the charging pumps failed, regardless of the availability of the safety-injection pumps. Personnel from Idaho National Laboratory informed the analyst that the change to the model was made at all PWRs based on the NUREG/CR-2187, Confirmatory Thermal-Hydraulic Analysis to Support Specific Success Criteria in the Standardized Plant Analysis Risk Models-Byron Unit 1, evaluation of Byron. In this study, the MELCOR runs suggested that feed and bleed would fail for scenarios with a single safety-injection pump and a single pressurizer pilot-operated relief valve. The analyst noted that DCPP had three pressurizer PORVs and the licensee had MAPP runs showing successful feed and bleed with a single safety-injection pump. Therefore, for this evaluation, the analyst used the plant-specific information to modify the SPAR model to show feed and bleed success for a single safety-injection pump.
Using the modified SPAR model, the analyst quantified a new baseline core damage frequency of 6.15 x 10-6/year and a case conditional core damage probability of 6.43 x 10-6 over a 1-year period. As a bounding assumption, the analyst used an exposure time of 133 days, which represented the time over which the pump had been tested and accumulated a run time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The resulting incremental conditional core damage probability was 1.02 x 10-7. The analyst identified two items that, if fully evaluated, would reduce this probability further. Specifically, early in the exposure time, the charging pump would have functioned for many hours providing time for recovery of other components and reducing the decay heat load. Also, at the time of discovery, the pump was still functioning and would have continued to run for some unknown period of time had it been the last defense against core damage. As a result, the analyst determined, qualitatively, that the incremental conditional core damage probability was less than 1.0 x 10-7 (Green).
Cross-cutting Aspect: A cross-cutting aspect was not assigned to the finding since the finding did not represent current licensee performance. The performance deficiency occurred when operating experience was received and improperly dispositioned by PG&E in September 2011.
Enforcement:
Enforcement action does not apply because the performance deficiency did not involve a violation of a regulatory requirement.
Unresolved Item Applicability of required NDE 60855.1 - Operation of an (Closed) inspections on the Lift Yoke in ISFSI at Operating Plants accordance with ANSI N14.6 URI 07200026/2016001-01
Description:
During the routine ISFSI inspection conducted on September 19-24, 2016, a URI was identified and documented Inspection Report 05000275/2016011, 05000323/2016011, and 07200026/2016001 (ADAMS Accession No ML16323A110).
After spent fuel is loaded into a Multi-Purpose Canister (MPC), DCPP utilizes a lift yoke to lift the Holtec HI-TRAC transfer cask and MPC from the spent fuel pool to a cask wash-down pit.
The MPC is then welded and dried before moving the MPC to the ISFSI for storage.
DCPPs lift yoke is designated as a special lifting device and is classified as Important to Safety (ITS). The DCPP ISFSI Final Safety Analysis Report (FSAR) Section 4.4.1.3.1 states:
The transfer-cask-lifting trunnions and the lift yoke are designed, fabricated, inspected, maintained, and tested in accordance with NUREG-0612, Control of Heavy Loads, to ensure that structural failures of these items are not credible.
NUREG-0612 Section 5.1(4) states: Special lifting devices should satisfy the guidelines of American National Standards Institute (ANSI) N14.6, Radioactive Materials - Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds or More.
ANSI N.14.6 (1993 version), Section 6.3.1 stated that each special lifting device shall be subjected annually to load testing and a visual inspection; or if the load testing is omitted, to dimensional testing, visual inspection, and nondestructive examination (NDE) of major load-carrying welds and critical areas.
During the routine ISFSI inspection conducted in September 2016, it was identified that DCPP had been performing dimensional and visual inspections on its lift yoke in accordance with the ANSI standard during annual inspections, but had not been performing the NDE inspections on critical areas of the lift yoke. The inspectors found that DCPP had been following Holtec (the cask vendor) guidance that stated the lift yoke did not have any critical areas and that NDE inspections of the lift yoke were not required. Holtec provided DCPP with Response to Request for Technical Information 2655-2 which documented the vendors position of why no critical areas existed and as such the NDE was not required. This position to not perform NDE on the lift yoke as part of the annual maintenance was identified by NRC Region IV inspectors as differing from other Region IV sites that utilize the Holtec systems. Other sites in Region IV do perform NDE inspections on their lift yokes in accordance with the ANSI standard on an annual basis. Since this appeared to be a non-conservative approach and it did not match the maintenance activities associated with other Region IV Holtec system users, the inspectors forwarded the information in a Technical Assistance Request (TAR) to NRC Headquarters Division of Spent Fuel Management (DSFM) for their assistance in this review and opened a URI to document the resolution of the issue.
The DSFM staff noted in the TAR response, dated November 7, 2017, that the term critical area was not explicitly defined in the ANSI N14.6 (1993) standard. For guidance, DSFM re-examined the definition of a critical load in Section 3.4 of ANSI N14.6 which states: Any lifted load whose uncontrolled movement or release could adversely affect any safety-related system when such system is required for unit safety or could result in potential off-site exposures.
Therefore, to avoid any uncontrolled movement of the critical load, those components of the lift yoke which could induce an uncontrolled movement if they fail (facture critical members where redundancy is not provided) were viewed by the DSFM staff as critical areas. The components that were determined to contain critical areas came from the drawing titled, Diablo Canyon HI-TRAC Lift Yoke Ancillary #702. The drawing, on page 11 of 13, tabulates 27 components that make up the lift yoke. Of these, components 1, 5, 6, 16, and 17 are fracture critical components. Holtec designated these as being ITS Category A. The DSFM determined that these items shall be maintained in accordance with the ANSI N14.6 standard Section 6.3.1 to perform NDE inspections on the items on an annual basis.
Federal regulation 10 CFR 72.146 states, in part, that licensees shall establish measures to ensure that applicable regulatory requirements and the design basis, as specified in the license for those structures, systems, and components to which this section applies, are correctly translated into specifications, drawings, procedures, and instructions.
Since the lift yoke is classified as an Important-to-Safety component, the licensees FSAR Section 4.4.1.3.1 required the lift yoke be maintained in accordance with the NUREG 0612 which invoked the ANSI standard, and the DFSM staff has interpreted that the lift yoke does contain critical areas, the licensee is required to perform annual NDE inspections on the lift yoke. Contrary to the above, DCPPs Work Order 64125530 Lift Yoke Annual Inspection, only required dimensional and visual inspection on the lift yoke and not NDE testing of critical areas as required by the ANSI N14.6 standard.
Closure Basis:
Since the DCPP had been performing the visual and dimensional testing in accordance with the ANSI N14.6 standard, entered the issue into their corrective action program (Notifications 50872940 and 50882291), and subsequently performed the NDE examination on the lift yoke and did not find any indications nor discrepancies on the critical areas of the lift yoke, the NRC has determined this failure to perform annual NDE inspections on the critical areas of the lift yoke in accordance with 10 CFR 72.146 requirements constitutes a minor violation that is not subject to enforcement action in accordance with Section 2 of the NRCs Enforcement Policy.
Inspectors confirmed that the licensee had initiated the appropriate change of the lift yokes Preventative Maintenance Plan MP 25358 per Notification 50953427 to include NDE on the critical areas of the lift yoke for future annual maintenance activities.
Corrective Action References: Notifications 50872940, 50882291, and
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
On February 27, 2018, the inspector presented the URI results from the 2016 routine ISFSI inspection to Ms. Paula Gerfen, Station Director, and other members of the licensee staff.
On March 1, 2018, the inspectors presented the occupational radiation protection inspection results to Mr. J. Welsch, Vice President Nuclear Generation and Chief Nuclear Officer, and other members of the licensee staff.
On March 1, 2018, the inspector presented the Inservice inspection results to Mr. J. Welsch, and other members of the licensee staff.
On April 18, 2018, the resident inspectors presented the quarterly resident inspector inspection results to Mr. J. Welsch and other members of the licensee staff.
DOCUMENTS REVIEWED
71111.01: Adverse Weather Protection
Notifications
50957352 50958613
Procedures
Number Title Revision
AD4.ID4 Temporary Storage Process 3
OP K-2C Fire Protection - Network Operations 45A
71111.04: Equipment Alignment
Notifications
50965347 50965742 50932504 50887271
50842872 50842873 50849878 50887203
50849758 50449163
Work Orders
60081257 60097453 60096938
Procedures
Number Title Revision
OP B-3A:II Safety Injection System Alignment Verification for Plant 23B
Startup
OP B-7:I SFP- Make Available and Place in Service 24A
OP K-10D Auxiliary Feedwater System Alignment Checklist 19
OP L-0 Mode 4 to 3 Transition Checklist 82
STP I-1D Routine Monthly Checks Required by Licenses 76
STP P-CSP-A21 Comprehensive Testing of Containment Spray Pump 2-1 13A
Attachment 1
Drawings
Number Description Revision
106716-17 Unit 2, Condensate Water Supply 180
106716-11 Unit 1 and 2 Raw Water System 181
106709 60
106718-6 Firewater System One Line Diagram 176
106720 36
107703-3 Unit 2, Auxiliary Feedwater System One Line Diagram 69
107709 Safety Injection System, Sheet 4 52
107710 Containment Spray System OVID 36
107710 Residual Heat Removal System, Sheet 2 29
107713 Spent Fuel Pool System 39
Other
Number Description Revision
DCM S-9 Safety Injection System 37
DCM S-10 Residual Heat Removal System 8
71111.05: Fire Protection
Notifications
50964085 50964087 50964089
Procedures
Number Title Revision
OM8.ID4 Control of Flammable and Combustible Materials 27A
Drawings
Number Description Revision
PA-2 Fire Drawing: Intake Structure 5
RA-1 Fire Drawing: Radiological Control Area 54 & 64 5
RA-31 Fire Drawing: Containment Building Elev. 91 4
RA-32 Fire Drawing: Containment Building Elev. 117 5
RA-33 Containment Building Elev. 91 & 117 3
Drawings
Number Description Revision
RA-33 Containment Building Elev. 91 & 117 3
RA-34 Fire Drawing: Containment Building Elev. 140 4
71111.06: Flood Protection Measures
Notifications
50872132 55774132 50836858 50888596
50916439 50934115
Work Orders
64101202 64103254
Procedures
Number Title Revision
M-17.9 Auxiliary Saltwater Maintenance 32
Other
Number Description Revision
M-988 ASW flows, temperatures and pressure 7
71111.07: Heat Sink Performance
Procedures
Number Title Revision
OP F-2 CCW System 8A
OP F-2: I CCW Make Available 44
Drawings
Number Description Revision
106714 CCW System 59
663212 CCW Mechanical Heat Exchanger Tube Plugging Map, 67
Sheet 1
Other
Number Description Date
26139 Component Cooling Water Heat Exchangers CCW 2-1 and 2/2018
CCW 2-2 Eddy Current Report
71111.08: Inservice Inspection
Notifications
50965970 50966007 50963737 50963849
50688241 50965369 50852155 50915871
50852586 50884077 50919783 50864574
50890899 50861150 50889754 50545275
50546153 50853443 50909260 50683435
Procedures
Number Title Revision
266373-TR-001 RHR SWOL Repair, Common Prerequisites and Closure 0
266373-TR-002 RHR Structural Weld Overlay Repair 1
AD4.ID2 Plant Leakage Evaluation 12
NDE PDI-UT-1 Ultrasonic Examination of Ferritic Piping 7
NDE PDI-UT-2 Ultrasonic Examination of Austenitic Piping 12
NDE PDI-UT-3 Ultrasonic Through-Wall Sizing in Pipe Welds 4
NDE PT-1 Liquid Dye Penetrant Examination Procedure 6
NDE UT- Ultrasonic Beam Spread Determination 0
BMSPRD
NDE VT 3-L VT-3 Visual Examination of the Containment Liner 1
WPS 11 Welding of P8 Materials with GTAW and/or SMAW 8
WPS 5 Welding of P1 Materials with GTAW and/or SMAW 8
Miscellaneous Documents
Number Title Date
50601807 Quick HIT Self-Assessment for Reactor Coolant System 11/6/2013
Materials Degradation Management Program (RCS MDMP)
and Steam Generator Management Program (SGMP)
50688241 DCPP Inservice Inspection Program Self-Assessment 2015 5/20/2015
Miscellaneous Documents
Number Title Date
DCL-17 -095 Response to NRC Request for Additional Information 11/20/2017
Regarding "Request for Approval of Alternative for Application
of Full Structural Weld Overlay, REP-RHRSWOL, Units 1 and
2"
DCL-17-083 Request for Approval of Alternative for Application of Full 9/26/2017
Structural Weld Overlay, REP-RHR-SWOL. Units 1 and 2
71111.11: Licensed Operator Requalification Program and Licensed Operator
Performance
Notification
50959163
Procedures
Number Title Revision
OP1.DC10 Conduct of Operations 47
STP M-15 Integrated Test of Engineered Safeguards and Diesel 68
Generators
STP M-26A FCV-601, ASW Unit 1 and 2 Cross-tie Dividing Valve Flow 13
Test
STP M-9D2 Diesel Generator Partial Load Rejection Test 26
STP V-2D2B Exercising and Position Verification of ECCS Valve 8700B 3A
71111.12: Maintenance Effectiveness
Notifications
50967198 50963220 50963637 50966575
50852067 50860985 50967447 50966553
50966889 50966552 50967030 50948544
50966584 50966402
Work Orders
60090369 68048760 60108699
Procedures
Number Title Revision
CF3.ID9 Design Change Notice 0
MA1.ID17 Maintenance Rule Monitoring Program 31
PEP V-7B Test of ECCS Valve Interlocks 11
Drawings
Number Description Revision
2648 Sheet 1 30
2650 Sheet 2 19
2675 Sheet 34 22
71111.13: Maintenance Risk Assessments and Emergent Work Control
Notification
50973413
Work Orders
50968047 50968076 50968640
Procedures
Number Title Revision
AD7.DC51 Outage Safety Management Control of Off-Site Power 18
Supplies to Vital Buses
AD7.DC6 On-Line Maintenance Risk Management 25
AD7.ID14 Assessment of Integrated Risk 15
AD7.ID17 PRA Significant Component Management 0A
AD8.DC54 Containment Closure 15
AD8.ID4 Outage Fire Risk Mitigation 0
MP M-45.1 Containment Hatch Door Closure 14
OM10.ID6 Equipment Important to Emergency Response 6
OP O-36 Protected Equipment Postings 16
Drawings
Revision /
Number Description Date
Unit 2 Refueling Outage Safety Schedule -2R20 12/14/17
495888 Functional Diagram Seismic Trip 2
71111.15: Operability Determinations and Functionality Assessments
Notifications
50662149 50870033 50952407 50958526
50958614 50962232 50959948 50959757
50959829 50963116 50963109 50967066
50966370 50966501 50857028 50963639
50963548 50683856 50857018
Work Orders
60093787 60076383
Procedures
Number Title Revision
OM7.ID1 Corrective Action Program 51
OM7.ID12 Operability Determination 36
OM7.ID7 Emerging Issues and Event Investigations 19
OP C-2:III Main Steam and Steam Dump Systems 13
OP O-2 Operation of Process Hand Controllers 23
RCP D-230 Radiological Control for Containment Entry 22
STP M-15 Integrated Test of Engineering Safeguards 68
STP V-2U4D Exercising SG 1-4 AFW Supply Valves LCV-113 12
STP V-3P6B Exercising Valves LCV-113 29
Drawings
Number Description Revision
106703 Auxiliary Feedwater, Sheet 3 71
107704 Unit 2-One line Diagram of Main Steam System 83
109803 Functional Loop Diagram LC-89 2
Drawings
Number Description Revision
441305 Reactor Coolant Pumps Electrical, Sheet 1 19
663049-1 Mechanical Assembly of MSIV 47
663165 Wiring Diagram NH91 Electric Actuator 2
71111.18: Plant Modifications
Notifications
50963670 50971609 50967447 50966553
50966889 50966552 50948544 50966584
50966402
Work Orders
60108120 60108122 68048760 60108699
Procedures
Number Title Revision
CF3.ID9 Design Change Notice 0
CF4.ID10 Design Change Development 53
CF4.ID7 Temporary Alteration 30
PEP V-7B Test of ECCS Valve Interlocks 11
Drawings
Number Description Revision
2648 Sheet 1 30
2650 Sheet 2 19
2675 Sheet 34 22
71111.19: Post-Maintenance Testing
Notifications
50956906 50946587 50920677 50968083
50968647 50968145 50973096 50963712
Work Orders
64030635 64103254 64136640 64077703
64161423
Procedures
Number Title Revision
M-17.9 Auxiliary Saltwater Pump Maintenance 32
OPE-5:1 Alignment ASW Pump 1 0
STP M-39D Routine Surveillance Test of Carbon Dioxide Hose Reels 17
STP P-CCP-12 Routine Surveillance Test of Centrifugal Charging Pump 1-2 31
STP P-CSP-21 Routine Surveillance Test of Containment Spray Pump 2-1 17
STP P-SIP-A21 Comprehensive Test of Safety Injection Pump 21 9
71111.20: Refueling and Other Outage Activities
Notifications
50857018 50890593 50957240 50958945
50959693 50960202 50960259 50960914
50962645 50964023 50968609 50968632
50968640 50968705 50968707 50968708
50968721 50970403 50971046 50971833
50971885 50972015 50972054
Work Orders
60101827 64136443 64175744 64136008
64136218
Procedures
Number Title Revision
AD8.DC51 Outage Safety Management Control of Off-Site Power 18
Supplies to Vital Buses
AD8.DC54 Containment Closure 15
ER1.ID2 Boric Acid Corrosion Control Program 7
Procedures
Number Title Revision
MP I-2.28 Activation and Deactivation of the Rx Vsl Refueling Lvl 28
Indication System (RVRLIS)
MP M-45.1 Containment Equipment Hatch Door Opening and Closing 14
OM14.ID1 Fatigue Management Rule Program 27
OM6.ID7 Activities Near High Voltage Equipment 10
OP A-2:II Reactor Vessel - Draining the RCS to the Vessel Flange - 53
With Fuel in Vessel
OP A-2:IX Reactor Vessel - Vacuum Refill of the RCS 30
OP A-2:X RVRLIS Alignments for Refueling Outages 8A
OP B-2:V RHR - Place in Service 36
OP B-8DS2 Core Loading 61
OP L-0 Mode Transition Checklists 82
OP L-4 Normal Operation at Power 77
OP L-5 Plant Cooldown From Minimum Load to Cold Shutdown 86
(unit 2)
OP2.ID1 Clearances 40
STP M-45A Containment Inspection Prior to Establishing Containment 34
Integrity
STP R-30 Reload Cycle Initial Criticality 19A
STP R-6 Low Power Reload Physics Tests 17
STP R-8C Containment Walk down for Evidence of Boric Acid Leakage 10
Other
Revision /
Number Description Date
2C20 R-17-002B Circulating Water Pump 2-1 Clearance 2/13/2018
2C20 R-17- Auxiliary Saltwater Pump 2-1 Clearance 2/14/2018
23A,
AD8.DC55 2R20 Outage Safety Plan 0
(Section 5.2)
EMPCenter 2018 Manager Time Entry Data Base 4/2018 and
3/2018
71111.22: Surveillance Testing
Notifications
50962900 50963118 50963045 50963034
50970198 50964023 50963720 50965673
Work Orders
64134469 64130743 64135924 64136578
Procedures
Number Title Revision
MP M-4.18A Check of Main Steam Safety Valve Lift Point With the 14
Trevitest System
STP M-15 Integrated Test of Engineered Safeguards and Diesel 68
Generators
STP M-26 ASW System Flow Monitoring 32
STP M-26A FCV-601, ASW Unit 1 and 2 Cross-Tie Dividing Valve, Flow 13
Test
STP M-7 Integrated Leak Rate Test (ILRT) Type A 14
STP M-77 Safety and Relief Valve Testing 39
STP M-7E Containment Penetration Valve Lineup for the Integrated 7
Leakage Rate Test (ILRT)
STP M-7W Containment Structural Integrity Inspection 5
STP R-1B Rod Drop Measurement 37A
STP V-651B Penetration 51B Containment Isolation Valve Leak Testing 19
Calculations
Number Description Revision
663199-122-3 Digital Rod Position Indication system vendor manual
C-M-26-1 ASW Flow Sensor Uncertainty 1
M-988 Evaluation of ASW Bypass Piping 7
Other
Number Description Date
ANSI/ANS-56.8- Containment System Leakage Testing Requirements
2002
Other
Number Description Date
MRP 54 Cal Leak Rate Monitor Calibration Data 1/12/2018
Data
71124.01: Radiological Hazard Assessment and Exposure Controls
Notifications
50919030 50921983 50923635 50924825
50924826 50926292 50934918 50944790
50954298 50958712 50960955 50961004
Procedures
Number Title Revision
RCP D-220 Control of Access to High, Locked High, and Very High 52
Radiation Areas
RCP D-240 Radiological Posting 25, 26
RCP D-310 RCA Access Control 26
RCP D-330 Personnel Dosimetry Evaluations 11
RCP D-335 Radiation Exposure Reporting 5
RCP D-500 Routine and Job Coverage Surveys 43
RCP D-620 Radioactive Source Control Program 13
RP1 Radiation Protection (Program Directive) 8
RP1.DC6 RP Code of Conduct 4
RP1.ID10 Embryo-Fetus Protection Program 9
RP1.ID16 Radiation Worker Expectations 10
RP1.ID6 Personnel Dose Limits and Monitoring Requirements 15
Audits and Self-Assessments
Number Description Date
50943634 Rad Risk Assessment Benchmark 10/16/2017
50958040 NRC Pre-Inspection on Radiological Hazard Assessment 1/19/2018
and Exposure Controls
50960956 ANI Nuclear Liability Insurance Inspection 1/29/2018
173330010 Quality Performance Assessment Report 12/19/2017
Radiation Work Permits
Number Description Revision
2002 2R20 Scaffold Work 0
2031 2R20 Regen HX Room Work 0
2050 2R20 Reactor Coolant Pump Work 0
2061 2R20 Containment Valves and Breaches 0
2070 2R20 RHR Weld Overlay (WIB 245) 0
Radiation Surveys
Number Description Date
59137 64 Quarterly 1/29/2018
59511 Unit 2 140 Open QOTTC and Cart Inspection 2/12/2018
59529 Containment 91 IS Bio-Shield at RCP 2-3 & 2-4. 2/12/2018
Post Forced O2 Procedure
59612 R 8 C Walkdowns Install Barrier in Letdown Orifice Room 2/13/2018
and Downpost to HRA
59789 2R20 91 Scaffold 2/15/2018
59811 Unit 2 140 Containment Remove Drive Shafts from Internals 2/16/2018
to DSSR
59960 2R20 Move SLD2 from upper internals to lower internals 2/19/2018
60003 Regen Hx Pre-Job Scaffold Survey 2/19/2018
60110 Remove CVCS-2-5505 2/20/2018
60186 Scaffold 2R20 115 Ctmt 2/21/2018
60446 2R20 Tri-Nuke Replacement 2/27/2018
Number Description Date
59697 2R20 Breach PZR Relief Valve 8010B 2/14/2018
59905 2R20 LWS-2-33 Valve Cutout RCDT Room 2/18/2018
59969 U2 CTMT RHR Weld Overlay 91 2-4 Area 2/19/2018
60110 Remove CVCS-2-5505 2/20/2018
60446 2R20 Tri-Nuke Replacement 2/27/2018
Miscellaneous Documents
Number Description Date
NSTS Annual Inventory Reconciliation Report 1/10/2018
Unit 1 Spent Fuel Pool Inventory - Non-Fuel Items 2/1/2018
Unit 2 Spent Fuel Pool Inventory - Non-Fuel Items 2/1/2018
2944 Sealed Source Leak Test - 1st Quarter 2017 2/6/2017
57108 Sealed Source Leak Test - 2nd Quarter 2017 8/9/2017
59154 Sealed Source Leak Test - 1st Quarter 2018 1/29/2018
71124.02: Occupational ALARA Planning and Controls
Notifications
50617408 50839148 50839149 50846171
50848759 50849592 50850918 50855741
50864083 50877280 50914066 50924222
Procedures
Number Title Revision
RCP D-200 Writing RWPs and ALARA Processes 57
RCP D-202 RWP Work Instructions 15
RP1.ID1 ALARA Program 10
RP1.ID9 Radiation Work Permits 13
Audits and Self-Assessments
Number Description Date
50905151 Formal Self-Assessment Effectiveness Review: Unit 1 7/11/2017
Radiation Dose Control
50958659 Quick Hit Self-Assessment Report: 71124.02 Occupational 1/22/2018
ALARA Planning and Controls
ALARA Planning, In-Progress Reviews, and Post-Job Reviews
Number Description Date
16-0032B 2R20 ISFSI: Spent Fuel Outage Campaign SFP Purification 2/14/2018
Required for Work in the SFP
1066 1R20 Emergent Work 7/7/2017
ALARA Planning, In-Progress Reviews, and Post-Job Reviews
Number Description Date
1090 1R20 Baffle Bolt Inspection 7/12/2017
24 2R20 Guide Cards 2/22/2018
2050 2R20 RCP Maintenance 2/8/2018
2070 2R20 RHR Weld Overlay (WIB 245) 2/8/2018
Miscellaneous Documents
Number Description Date
1R20 Post Outage ALARA Report
2R19 Outage ALARA Report
DCPP Dose Reduction Strategy 2018
Radiation Protection Section Annual Review - 2015 9/7/2016
Radiation Protection Section Annual Review - 2016 5/5/2017
Unit 2 LD Filter Dose & D/P, RCS Co-58 & Co-60 1/25/2018
71151: Performance Indicator Verification
Procedures
Number Title Revision
AWP L-001 NRC Performance Indicators Initiating Events, SSFFs, and 9
NEI 99-02 Regulatory Assessment Performance Indicator Guideline 7
XI1.ID2 Regulatory Reporting Requirements 42B
XI1.ID5 Collection and Submittal of NRC Performance Indicators 1
71152: Problem Identification and Resolution
Notifications
50954496 50958899 50949938 50949531
50961430 50949662 50958984 50970395
50966193 50970650 50958547 50960208
50960047 50964614 50966876 50920007
50922188 50961014 40970426 5094717
50946885 50948215 50951450 50951534
50951986 50951539 50951891 50951894
50963245 50964323 50964036 50966150
50965966 50969293 50966459
Procedures
Number Title Revision
AD2.ID1 Procedure and Work Plan Use Adherence 26
MP E-8.1 Centrifugal Charging Pump Motor Overhaul 9
OM4.ID14 Notification Review Team 30A
OM4.ID3 Operating Experience Program 20
OP O-35 Bumped Component Protection Program 13
71153: Follow-up of Events and Notices of Enforcement Discretion
Notifications
50960279 50960218 50958544
Drawings
Number Description Revision
047237 Mechanical Design Standard Piping Specification K12 16
663055 Feedwater Pump Outline, Sheet 41 2
Other
Number Description Date
EN# 53184 Diablo Canyon Power Plant Unit 1 and 2 Event Notification 1/25/2018
60855.1: Operation of an Independent Spent Fuel Storage Installation (ISFSI)
Notifications
50882891 50872940 50953469 50953427
Request for Information for
Resident Inspection at
Diablo Canyon Power Plant -
1st Quarter 2018
Please provide the following documents (use date range of document request unless
otherwise noted):
1. Summary listing of all root and apparent cause evaluations performed on this system
2. Summary listing of all condition reports written on this system, sorted by category type
3. Summary listing of OE screening and dispositioning reviews
4. Summary listing of all corrective maintenance work orders, with description of work,
performed on this system
5. List of all work orders, with description of work, planned within the next 6 months (from
start of applicable quarter).
6. List of maintenance rule functional failure assessments - regardless of the result -
performed on this system for previous two years. Include condition report numbers for
each item on the list
7. Summary list of system design calculations
8. Summary list of fire impairments associated with this system
9. List of completed engineering changes and planned engineering changes (within the
next six months of start of applicable quarter) associated with this system.
10. Schedule of system surveillance and work activities for the planned quarter of interest.
11. Summary listing of temporary and permanent modifications associated with the system
completed within the last two years
2. Summary listing of 50.59 screenings and evaluations related to the system
13. Maintenance rule scoping and basis documents
14. Maintenance rule expert panel meeting minutes for meetings related to system
15. Current UA hours and UR data tracking documents (system health report)
Attachment 2
1st Quarter 2018
System: Residual Heat Removal
IMS Upload Date Requested By: December 25, 2017
Date Range of Document Request: January 1, 2016 - Current
The following items are requested for the
Occupational Radiation Safety Inspection
Diablo Canyon
February 26 - March 2, 2018
Integrated Report 2018001
Inspection areas are listed in the attachments below.
Please provide the requested information on or before February 9, 2018.
Please submit this information using the same lettering system as below. For example, all
contacts and phone numbers for Inspection Procedure 71124.01 should be in a file/folder titled
1- A, applicable organization charts in file/folder 1- B, etc.
If information is placed on ims.certrec.com, please ensure the inspection exit date entered is at
least 30 days later than the onsite inspection dates, so the inspectors will have access to the
information while writing the report.
In addition to the corrective action document lists provided for each inspection procedure listed
below, please provide updated lists of corrective action documents at the entrance meeting.
The dates for these lists should range from the end dates of the original lists to the day of the
entrance meeting.
If more than one inspection procedure is to be conducted and the information requests appear
to be redundant, there is no need to provide duplicate copies. Enter a note explaining in which
file the information can be found.
If you have any questions or comments, please contact John ODonnell at (817) 200-1441 or
john.odonnell@nrc.gov.
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 3150-0011.
Attachment 3
1. Radiological Hazard Assessment and Exposure Controls (71124.01) and
Performance Indicator Verification (71151)
Date of Last Inspection: May 8, 2017
A. List of contacts and telephone numbers for the Radiation Protection Organization Staff
and Technicians
B. Applicable organization charts
- C. Audits, self-assessments, and LERs written since date of last inspection, related to this
inspection area
D. Procedure indexes for the radiation protection procedures
E. Please provide specific procedures related to the following areas noted below.
Additional Specific Procedures may be requested by number after the inspector reviews
the procedure indexes.
1. Radiation Protection Program Description
2. Radiation Protection Conduct of Operations
3. Personnel Dosimetry Program
4. Posting of Radiological Areas
5. High Radiation Area Controls
6. RCA Access Controls and Radiation Worker Instructions
7. Conduct of Radiological Surveys
8. Radioactive Source Inventory and Control
9. Declared Pregnant Worker Program
F. List of corrective action documents (including corporate and sub-tiered systems) since
date of last inspection
a. Initiated by the radiation protection organization
b. Assigned to the radiation protection organization
NOTE: The lists should indicate the significance level of each issue and the search
criteria used. Please provide in document formats which are searchable so that
the inspector can perform word searches.
If not covered above, a summary of corrective action documents since date of last
inspection involving unmonitored releases, unplanned releases, or releases in which any
dose limit or administrative dose limit was exceeded (for Public Radiation Safety
Performance Indicator verification in accordance with IP 71151)
G. List of radiologically significant work activities scheduled to be conducted during the
inspection period (If the inspection is scheduled during an outage, please also include a
list of work activities greater than 1 rem, scheduled during the outage with the dose
estimate for the work activity.)
H. List of active radiation work permits
I. Radioactive source inventory list
a. All radioactive sources that are required to be leak tested
b. All radioactive sources that meet the 10 CFR Part 20, Appendix E, Category 2
and above threshold. Please indicate the radioisotope, initial and current activity
(w/assay date), and storage location for each applicable source.
J. The last two leak test results for the radioactive sources inventoried and required to be
leak tested. If applicable, specifically provide a list of all radioactive source(s) that have
failed its leak test within the last two years
- K. A current listing of any non-fuel items stored within your pools, and if available, their
appropriate dose rates (Contact / @ 30cm)
L. Computer printout of radiological controlled area entries greater than 100 millirem since
the previous inspection to the current inspection entrance date. The printout should
include the date of entry, some form of worker identification, the radiation work permit
used by the worker, dose accrued by the worker, and the electronic dosimeter dose
alarm set-point used during the entry (for Occupational Radiation Safety Performance
Indicator verification in accordance with IP 71151).
2. Occupational ALARA Planning and Controls (71124.02)
Date of Last Inspection: July 31, 2017
A. List of contacts and telephone numbers for ALARA program personnel
B. Applicable organization charts
- C. Copies of audits, self-assessments, and LERs, written since date of last inspection,
focusing on ALARA
D. Procedure index for ALARA Program
E. Please provide specific procedures related to the following areas noted below.
Additional Specific Procedures may be requested by number after the inspector reviews
the procedure indexes.
1. ALARA Program
2. ALARA Committee
3. Radiation Work Permit Preparation
F. A summary list of corrective action documents (including corporate and sub-tiered
systems) written since date of last inspection, related to the ALARA program. In addition
to ALARA, the summary should also address Radiation Work Permit violations,
Electronic Dosimeter Alarms, and RWP Dose Estimates
NOTE: The lists should indicate the significance level of each issue and the search
criteria used. Please provide in document formats which are searchable so that
the inspector can perform word searches.
- G. List of work activities greater than 1 rem, since date of last inspection,
Include original dose estimate and actual dose.
H. Site dose totals and 3-year rolling averages for the past 3 years (based on dose of
record)
I. Outline of source term reduction strategy
- J. If available, provide a copy of the ALARA outage report for the most recently completed
outages for each unit
K. Please provide your most recent Annual ALARA Report.
SUNSI Review Non-Sensitive Publicly Available Keyword:
By: MSH2/dll Sensitive Non-Publicly Available NRC-002
OFFICE SRI:DRP/A RI:DRP/A C:DRS/EB1 C:DRS/EB2 C:DRS/OB C:DNMS/FCDB
NAME CNewport JReynoso TFarnholtz JDrake VGaddy RKellar
SIGNATURE /RA/ /RA/ /RA/ /RA/ /RA/ /RA/
DATE 04/19/18 04/19/18 04/23/18 04/24/18 04/20/18 04/19/18
OFFICE C:DRS/PSB2 C:DRS/IPAT SPE:DRP/A BC:DRP/A
NAME HGepford GGeorge RAlexander MHaire
SIGNATURE /RA/ /RA/ /RA/ /RA/
DATE 04/23/18 04/24/18 04/19/2018 4/24/18