IR 05000275/2018001

NRC Inspection Report 05000275/2018001, 05000323/2018001, and 07200026/2018001

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

April 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

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

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 Aspect Report Section Mitigating Systems Green NCV 05000323/2018001-001 Closed H.5 - Human Performance -

Work Management 71111.20 - Refueling and Other Outage Activities 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 Aspect Report Section Mitigating Systems Green FIN 05000323/2018001-02 Closed None 71152 - Problem 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.

Additional Tracking Items

Type Issue number Title Report Section Status URI 07200026/2016001-01 Applicability of required NDE inspections on the Lift Yoke in accordance with ANSI N14.6 60855.1 - Operation of an ISFSI at Operating Plants Closed

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.

71111.01 - Adverse Weather Protection

Impending Severe Weather (1 Sample)

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 (5 Samples)

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 (5 Samples)

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 (1 Sample)

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 (1 Sample)

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 (1 Sample)

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 (2 Samples)

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,

71124.01 - Radiological Hazard Assessment and Exposure Controls

Radiological Hazard Assessment (1 Sample)

The inspectors evaluated radiological hazards assessments and controls.

Instructions to Workers (1 Sample)

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 (1 Sample)

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

(2) RWP 17-2050, 2R20 RCP Maintenance

(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 (2 Samples)

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 Aspect Report Section Mitigating Systems

Green NCV 05000323/2018001-01 Closed H.5 - Human Performance - Work Management

71111.20 - Refueling and

Other Outage 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 Aspect Report Section Mitigating Systems

Green FIN 05000323/2018001-02 Closed None

71152 - Problem

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 (Closed)

Applicability of required NDE inspections on the Lift Yoke in accordance with ANSI N14.6 URI 07200026/2016001-01 60855.1 - Operation of an ISFSI at Operating Plants

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

Work Order 64030635

Procedures

Number

Title

Revision

AD4.ID4

Temporary Storage Process

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

Startup

23B

OP B-7:I

SFP-Make Available and Place in Service

24A

OP K-10D

Auxiliary Feedwater System Alignment Checklist

OP L-0

Mode 4 to 3 Transition Checklist

STP I-1D

Routine Monthly Checks Required by Licenses

STP P-CSP-A21 Comprehensive Testing of Containment Spray Pump 2-1

13A

Drawings

Number

Description

Revision

106716-17

Unit 2, Condensate Water Supply

180

106716-11

Unit 1 and 2 Raw Water System

181

106709

106718-6

Firewater System One Line Diagram

176

106720

107703-3

Unit 2, Auxiliary Feedwater System One Line Diagram

107709

Safety Injection System, Sheet 4

107710

Containment Spray System OVID

107710

Residual Heat Removal System, Sheet 2

107713

Spent Fuel Pool System

Other

Number

Description

Revision

DCM S-9

Safety Injection System

DCM S-10

Residual Heat Removal System

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

RA-1

Fire Drawing: Radiological Control Area 54 & 64

RA-31

Fire Drawing: Containment Building Elev. 91

RA-32

Fire Drawing: Containment Building Elev. 117

RA-33

Containment Building Elev. 91 & 117

Drawings

Number

Description

Revision

RA-33

Containment Building Elev. 91 & 117

RA-34

Fire Drawing: Containment Building Elev. 140

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

Other

Number

Description

Revision

M-988

ASW flows, temperatures and pressure

71111.07: Heat Sink Performance

Procedures

Number

Title

Revision

OP F-2

CCW System

8A

OP F-2: I

CCW Make Available

Drawings

Number

Description

Revision

106714

CCW System

663212

CCW Mechanical Heat Exchanger Tube Plugging Map,

Sheet 1

Other

Number

Description

Date

26139

Component Cooling Water Heat Exchangers CCW 2-1 and

CCW 2-2 Eddy Current Report

2/2018

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

266373-TR-002

RHR Structural Weld Overlay Repair

AD4.ID2

Plant Leakage Evaluation

NDE PDI-UT-1

Ultrasonic Examination of Ferritic Piping

NDE PDI-UT-2

Ultrasonic Examination of Austenitic Piping

NDE PDI-UT-3

Ultrasonic Through-Wall Sizing in Pipe Welds

NDE PT-1

Liquid Dye Penetrant Examination Procedure

NDE UT-

BMSPRD

Ultrasonic Beam Spread Determination

NDE VT 3-L

VT-3 Visual Examination of the Containment Liner

WPS 11

Welding of P8 Materials with GTAW and/or SMAW

WPS 5

Welding of P1 Materials with GTAW and/or SMAW

Miscellaneous Documents

Number

Title

Date

50601807

Quick HIT Self-Assessment for Reactor Coolant System

Materials Degradation Management Program (RCS MDMP)

and Steam Generator Management Program (SGMP)

11/6/2013

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

Regarding "Request for Approval of Alternative for Application

of Full Structural Weld Overlay, REP-RHRSWOL, Units 1 and

2"

11/20/2017

DCL-17-083

Request for Approval of Alternative for Application of Full

Structural Weld Overlay, REP-RHR-SWOL. Units 1 and 2

9/26/2017

71111.11: Licensed Operator Requalification Program and Licensed Operator

Performance

Notification

50959163

Procedures

Number

Title

Revision

OP1.DC10

Conduct of Operations

STP M-15

Integrated Test of Engineered Safeguards and Diesel

Generators

STP M-26A

FCV-601, ASW Unit 1 and 2 Cross-tie Dividing Valve Flow

Test

STP M-9D2

Diesel Generator Partial Load Rejection Test

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

MA1.ID17

Maintenance Rule Monitoring Program

PEP V-7B

Test of ECCS Valve Interlocks

Drawings

Number

Description

Revision

2648

Sheet 1

2650

Sheet 2

2675

Sheet 34

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

Supplies to Vital Buses

AD7.DC6

On-Line Maintenance Risk Management

AD7.ID14

Assessment of Integrated Risk

AD7.ID17

PRA Significant Component Management

0A

AD8.DC54

Containment Closure

AD8.ID4

Outage Fire Risk Mitigation

MP M-45.1

Containment Hatch Door Closure

OM10.ID6

Equipment Important to Emergency Response

OP O-36

Protected Equipment Postings

Drawings

Number

Description

Revision /

Date

Unit 2 Refueling Outage Safety Schedule -2R20

2/14/17

495888

Functional Diagram Seismic Trip

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

OM7.ID12

Operability Determination

OM7.ID7

Emerging Issues and Event Investigations

OP C-2:III

Main Steam and Steam Dump Systems

OP O-2

Operation of Process Hand Controllers

RCP D-230

Radiological Control for Containment Entry

STP M-15

Integrated Test of Engineering Safeguards

STP V-2U4D

Exercising SG 1-4 AFW Supply Valves LCV-113

STP V-3P6B

Exercising Valves LCV-113

Drawings

Number

Description

Revision

106703

Auxiliary Feedwater, Sheet 3

107704

Unit 2-One line Diagram of Main Steam System

109803

Functional Loop Diagram LC-89

Drawings

Number

Description

Revision

441305

Reactor Coolant Pumps Electrical, Sheet 1

663049-1

Mechanical Assembly of MSIV

663165

Wiring Diagram NH91 Electric Actuator

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

CF4.ID10

Design Change Development

CF4.ID7

Temporary Alteration

PEP V-7B

Test of ECCS Valve Interlocks

Drawings

Number

Description

Revision

2648

Sheet 1

2650

Sheet 2

2675

Sheet 34

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

OPE-5:1

Alignment ASW Pump 1

STP M-39D

Routine Surveillance Test of Carbon Dioxide Hose Reels

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

STP P-SIP-A21

Comprehensive Test of Safety Injection Pump 21

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

Supplies to Vital Buses

AD8.DC54

Containment Closure

ER1.ID2

Boric Acid Corrosion Control Program

Procedures

Number

Title

Revision

MP I-2.28

Activation and Deactivation of the Rx Vsl Refueling Lvl

Indication System (RVRLIS)

MP M-45.1

Containment Equipment Hatch Door Opening and Closing

OM14.ID1

Fatigue Management Rule Program

OM6.ID7

Activities Near High Voltage Equipment

OP A-2:II

Reactor Vessel - Draining the RCS to the Vessel Flange -

With Fuel in Vessel

OP A-2:IX

Reactor Vessel - Vacuum Refill of the RCS

OP A-2:X

RVRLIS Alignments for Refueling Outages

8A

OP B-2:V

RHR - Place in Service

OP B-8DS2

Core Loading

OP L-0

Mode Transition Checklists

OP L-4

Normal Operation at Power

OP L-5

Plant Cooldown From Minimum Load to Cold Shutdown

(unit 2)

OP2.ID1

Clearances

STP M-45A

Containment Inspection Prior to Establishing Containment

Integrity

STP R-30

Reload Cycle Initial Criticality

19A

STP R-6

Low Power Reload Physics Tests

STP R-8C

Containment Walk down for Evidence of Boric Acid Leakage 10

Other

Number

Description

Revision /

Date

2C20 R-17-002B Circulating Water Pump 2-1 Clearance

2/13/2018

2C20 R-17-

23A,

Auxiliary Saltwater Pump 2-1 Clearance

2/14/2018

AD8.DC55

(Section 5.2)

2R20 Outage Safety Plan

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

Trevitest System

STP M-15

Integrated Test of Engineered Safeguards and Diesel

Generators

STP M-26

ASW System Flow Monitoring

STP M-26A

FCV-601, ASW Unit 1 and 2 Cross-Tie Dividing Valve, Flow

Test

STP M-7

Integrated Leak Rate Test (ILRT) Type A

STP M-77

Safety and Relief Valve Testing

STP M-7E

Containment Penetration Valve Lineup for the Integrated

Leakage Rate Test (ILRT)

STP M-7W

Containment Structural Integrity Inspection

STP R-1B

Rod Drop Measurement

37A

STP V-651B

Penetration 51B Containment Isolation Valve Leak Testing

Calculations

Number

Description

Revision

663199-122-3

Digital Rod Position Indication system vendor manual

C-M-26-1

ASW Flow Sensor Uncertainty

M-988

Evaluation of ASW Bypass Piping

Other

Number

Description

Date

ANSI/ANS-56.8-

2002Property "ANSI code" (as page type) with input value "ANSI/ANS-56.8-</br></br>2002" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process.

Containment System Leakage Testing Requirements

Other

Number

Description

Date

MRP 54 Cal

Data

Leak Rate Monitor Calibration Data

1/12/2018

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

Radiation Areas

RCP D-240

Radiological Posting

25, 26

RCP D-310

RCA Access Control

RCP D-330

Personnel Dosimetry Evaluations

RCP D-335

Radiation Exposure Reporting

RCP D-500

Routine and Job Coverage Surveys

RCP D-620

Radioactive Source Control Program

RP1

Radiation Protection (Program Directive)

RP1.DC6

RP Code of Conduct

RP1.ID10

Embryo-Fetus Protection Program

RP1.ID16

Radiation Worker Expectations

RP1.ID6

Personnel Dose Limits and Monitoring Requirements

Audits and Self-Assessments

Number

Description

Date

50943634

Rad Risk Assessment Benchmark

10/16/2017

50958040

NRC Pre-Inspection on Radiological Hazard Assessment

and Exposure Controls

1/19/2018

50960956

ANI Nuclear Liability Insurance Inspection

1/29/2018

173330010

Quality Performance Assessment Report

2/19/2017

Radiation Work Permits

Number

Description

Revision

2002

2R20 Scaffold Work

2031

2R20 Regen HX Room Work

2050

2R20 Reactor Coolant Pump Work

2061

2R20 Containment Valves and Breaches

2070

2R20 RHR Weld Overlay (WIB 245)

Radiation Surveys

Number

Description

Date

59137

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.

Post Forced O2 Procedure

2/12/2018

59612

R 8 C Walkdowns Install Barrier in Letdown Orifice Room

and Downpost to HRA

2/13/2018

59789

2R20 91 Scaffold

2/15/2018

59811

Unit 2 140 Containment Remove Drive Shafts from Internals

to DSSR

2/16/2018

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

Air Sample Surveys

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

RCP D-202

RWP Work Instructions

RP1.ID1

ALARA Program

RP1.ID9

Radiation Work Permits

Audits and Self-Assessments

Number

Description

Date

50905151

Formal Self-Assessment Effectiveness Review: Unit 1

Radiation Dose Control

7/11/2017

50958659

Quick Hit Self-Assessment Report: 71124.02 Occupational

ALARA Planning and Controls

1/22/2018

ALARA Planning, In-Progress Reviews, and Post-Job Reviews

Number

Description

Date

16-0032B

2R20 ISFSI: Spent Fuel Outage Campaign SFP Purification

Required for Work in the SFP

2/14/2018

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

MOR

NEI 99-02

Regulatory Assessment Performance Indicator Guideline

XI1.ID2

Regulatory Reporting Requirements

2B

XI1.ID5

Collection and Submittal of NRC Performance Indicators

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

MP E-8.1

Centrifugal Charging Pump Motor Overhaul

OM4.ID14

Notification Review Team

30A

OM4.ID3

Operating Experience Program

OP O-35

Bumped Component Protection Program

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

663055

Feedwater Pump Outline, Sheet 41

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)

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.

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

By: MSH2/dll

Non-Sensitive

Sensitive

Publicly Available

Non-Publicly Available

Keyword:

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