IR 05000275/2019001

NRC Integrated Inspection Report 05000275/2019001 and 05000323/2019001

ML19130A234
Person / Time
Site:	Diablo Canyon
Issue date:	05/10/2019
From:	Mark Haire NRC/RGN-IV/DRP
To:	Welsch J Pacific Gas & Electric Co
References
IR 2019001
Download: ML19130A234 (47)

Text

SUBJECT:

DIABLO CANYON POWER PLANT, UNITS 1 AND 2 - NRC INTEGRATED INSPECTION REPORT 05000275/2019001 AND 05000323/2019001

Dear Mr. Welsch:

On March 31, 2019, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Diablo Canyon Power Plant, Units 1 and 2. On April 4, 2019, 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 three findings of very low safety significance (Green) in this report.

Two of these findings involved violations of NRC requirements. The NRC is treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2.a of the Enforcement Policy.

If you contest the violations or significance or severity of the violations documented in this inspection report, 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 and 50-323 License Nos. DPR-80 and DPR-82

Enclosure:

Inspection Report 05000275/2019001 and 05000323/2019001 w/attachments:

1. RFI Inservice Inspection 2. RFI O

Inspection Report

Docket Numbers: 05000275 and 05000323 License Numbers: DPR-80 and DPR-82 Report Numbers: 05000275/2019001 and 05000323/2019001 Enterprise Identifier: I-2019-001-0001 Licensee: Pacific Gas & Electric Company Facility: Diablo Canyon Power Plant, Units 1 and 2 Location: Avila Beach, California Inspection Dates: January 1, 2019 to March 31, 2019 Inspectors: R. Alexander, Senior Project Engineer B. Baca, Health Physicist L. Carson, Senior Health Physicist J. Drake, Senior Reactor Inspector S. Hedger, Emergency Preparedness Inspector G. Kolcum, Senior Resident Inspector, Columbia C. Newport, Senior Resident Inspector J. Reynoso, Resident Inspector 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 a Quarterly 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 Failure to Promptly Correct an Adverse Condition Related to Reactor Coolant Pump Seal Damage Cornerstone Significance Cross-cutting Report Aspect Section Initiating Events Green None 71152 NCV 05000275/2019001-01 Closed The inspectors reviewed a Green, self-revealed non-cited violation of 10 CFR Part 50,

Appendix B, Criterion XVI, Corrective Action, for the licensees failure to promptly identify and correct a condition adverse to quality. Specifically, prior to February 4, 2019, corrective actions generated as a result of a root cause evaluation did not adequately correct a condition adversely impacting reactor coolant pump seals.

Inadequate Corrective Actions Leads to Increased Time at Yellow Risk and Lowered Inventory Cornerstone Significance Cross-cutting Report Aspect Section Initiating Events Green [P.3] - Resolution 71152 FIN 05000275/2019001-02 Closed The inspectors reviewed a Green, self-revealed finding in that PG&E personnel failed to promptly correct a condition in accordance with OM7.ID1, Problem Identification and Resolution. Specifically, personnel did not promptly correct a condition associated with the Unit 1, polar crane variable frequency drive motor used during the removal and replacement of the reactor vessel head from the reactor vessel during refueling outages that was first identified in 2015. This led to a recent event in which the reactor vessel head was suspended above the reactor vessel during a period of yellow outage safety risk with the reactor vessel water level in lowered inventory condition for approximately 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> longer than necessary.

Human Performance Event Leads to Sodium Hydroxide Leak and Loss of Level Indication for Containment Spray System Spray Additive Tank Cornerstone Significance Cross-cutting Report Aspect Section Mitigating Green [H.12] - Avoid 71153 Systems NCV 05000323/2019001-03 Complacency Closed The inspectors reviewed a Green, self-revealed non-cited violation of Technical Specification 5.4.1.a, Procedures, because personnel failed to follow the requirements of MP I-12-L931, Spray Additive Tank Level Channel 931 Calibration, Revision 2. Specifically, on March 4, 2019, during calibration of a Unit 2, containment spray additive tank level transmitter, PG&E technicians inadvertently removed a venting plug from the wrong level transmitter resulting in a pressurized leak of sodium hydroxide solution from the spray additive tank and the inability to determine the level of the spray additive.

PLANT STATUS

Units 1 and 2 began the inspection period at full power.

On February 10, 2019, Unit 1 was shut down for a planned refueling outage. On March 16, 2019, Unit 1 returned to operation and began a controlled power ascension, and returned to full power on March 22, 2019.

Units 1 and 2 operated at or near full power for the remainder of the inspection period.

INSPECTION SCOPE

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 Sample (IP Section 03.03)

The inspectors evaluated readiness for impending adverse weather conditions for high ocean swell and heavy debris loading from ocean storms on January 10-11, 2019.

71111.04 - Equipment Alignment Partial Walkdown (IP Section 02.01)

The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:

(1) Unit 2, residual heat removal train A on January 14, 2019

(2) Unit 2, emergency diesel generator 2-2 on January 29, 2019

(3) Unit 2, emergency diesel generator 2-3 on January 29, 2019

(4) Unit 1, shutdown cooling train B on February 11, 2019

(5) Unit 1, reactor vessel refueling level indication system on March 8, 2019

71111.05Q - Fire Protection Quarterly Inspection (IP Section 03.01)

The inspectors evaluated fire protection program implementation in the following selected areas:

(1) Unit 2, auxiliary building 100 foot elevation on January 4, 2019

(2) Units 1 and 2, intake structure on January 16, 2019

(3) Unit 1, containment 140 foot elevation on February 26, 2019

(4) Unit 1, containment 91 foot elevation on March 5, 2019

(5) Unit 1, containment 117 foot elevation on March 5, 2019

71111.07A - Heat Sink Performance Annual Review (IP Section 02.01)

The inspectors evaluated readiness and performance of the Unit 1, component cooling water heat exchangers 1-1 and 1-2 on March 13, 2019.

71111.08P - Inservice Inspection Activities (PWR) PWR Inservice Inspection Activities (IP Section 03.01)

The inspectors verified that the reactor coolant system boundary, steam generator tubes, reactor vessel internals, risk-significant piping system boundaries, and containment boundary are appropriately monitored for degradation and that repairs and replacements were appropriately fabricated, examined and accepted by reviewing the following activities from February 18 to February 26, 2019:

03.01.a - Nondestructive Examination and Welding Activities

(1) Dye Penetrant Examination a) charging and volume control system, CVCS-1-8435, line 1452 b) main feedwater system, main feed pump 1-1 impeller

(2) Visual Examination a) safety injection system, accumulator fill lines b) auxiliary feedwater system, K-561-8 line c) auxiliary feedwater system, K-6012-8 line d) auxiliary feedwater system, K-638-8 line e) pressurizer system, PZR-8010A studs and nuts

(3) Ultrasonic Examination a) pressurizer system, girth weld 5

(4) The inspectors directly observed or reviewed records of the following welding activities:

a) Gas Tungsten Arc Weld - Manual

• chemical volume and control system valve replacement, CVCS-1-8435, line 1452 b) Gas Tungsten Arc Weld - Machine

• Reactor Coolant System. Weld Overlay. Reactor coolant system loop 2A lower cold leg, nozzle to safe-end circumferential weld, WIB-228 03.01.b - Pressurized-Water Reactor Vessel Upper Head Penetration Inspection Activities Vessel upper head penetration inspection was not required to be performed in this outage.

03.01.c - Pressurized-Water Reactor Boric Acid Corrosion Control Activities Inspectors reviewed the following boric acid notifications and corrective actions:

===50956677 50956730 50988345 50927788 50315950 51008168 03.01.d - Steam Generator Tube Inspection Activities The steam generator tube inspections were not required to be performed in this outage.

03.01.e - Identification and Resolution of Problems The inspectors reviewed 19 notifications that dealt with inservice inspection issues and found that items were entered into the corrective action program at the appropriate level and addressed appropriately.

71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01)

The inspectors observed and evaluated licensed operator performance in the control room during the following activities:

(1) Unit 1, plant shutdown for 1R21 refueling outage on February 9-10, 2019

(2) Unit 2, initiation and alignment of shutdown cooling on February 10, 2019 Licensed Operator Requalification Training/Examinations (IP Section 03.02) ===

The inspectors observed and evaluated a crew of licensed operators in the plants simulator during reactor startup training on March 13, 2019.

71111.12 - Maintenance Effectiveness Routine Maintenance Effectiveness Inspection (IP Section 02.01)

The inspectors evaluated the effectiveness of routine maintenance activities associated with the following equipment and/or safety significant functions:

(1) Unit 1, hot work on control room level recorder resulting in blown fuse on January 16, 2019

(2) Units 1 and 2, safety-related switchgear room ventilation and cooling changes to equipment on March 26, 2019

71111.13 - Maintenance Risk Assessments and Emergent Work Control Risk Assessment and Management Sample (IP Section 03.01)

The inspectors evaluated the risk assessments for the following planned and emergent work activities:

(1) Unit 2, emergency diesel generator 2-1 maintenance outage on January 29-30, 2019

(2) Unit 1, yellow outage risk during drain down for reactor head removal on February 17, 2019

(3) Unit 1, yellow outage risk during drain down for reactor head installation on March 8, 2019

71111.15 - Operability Determinations and Functionality Assessments Sample Selection (IP Section 02.01)

The inspectors evaluated the following operability determinations and functionality assessments:

(1) Unit 1, auxiliary saltwater pump 1-1 vital bus 4 kV circuit breaker loose support bolt on January 9, 2019

(2) Unit 1, elevated reactor coolant system calculated leakage on February 1, 2019

(3) Unit 1, valve RHR-1-8726B reach rod failure on February 12, 2019

(4) Unit 1, valve SI-8982A seat leakage on February 15, 2019

(5) Unit 1, valve RHR-1-8724A bonnet stud material specification on March 5, 2019

(6) Unit 1, fuel assembly thimble screw missing on March 12, 2019

71111.18 - Plant Modifications Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02)

The inspectors evaluated the following temporary or permanent modifications:

(1) Unit 1, emergency diesel generator 1-2 governor controls upgrade on February 5, 2019

71111.19 - Post Maintenance Testing Post Maintenance Test Sample (IP Section 03.01)

The inspectors evaluated the following post maintenance tests:

(1) Unit 1, vital battery charger 1-2, performance testing and load testing on January 16, 2019

(2) Unit 1, valve RHR-1-8702, thermal overload device replacement on March 7, 2019

(3) Unit 1, turbine driven auxiliary feedwater pump 1-1, testing following turbine maintenance on March 16, 2019

(4) Unit 2, centrifugal charging pump 2-3, testing following maintenance on March 26, 2019

71111.20 - Refueling and Other Outage Activities Refueling/Other Outage Sample (IP Section 03.01)

The inspectors evaluated refueling outage 1R21 activities from February 9, 2019 to March 18, 2019.

71111.22 - Surveillance Testing The inspectors evaluated the following surveillance tests: Containment Isolation Valve (ISO) (IP Section 03.01)

(1) Unit 1, containment integrated leak rate testing, per procedure STP M-7, on February 13, 2019

(2) Unit 1, containment isolation valve testing of penetration 52B associated with the nitrogen system, per procedure STP V-652B, on February 26, 2019

In Service Testing (IST) (IP Section 03.01) (1 Sample)

Unit 1, comprehensive testing of turbine-driven auxiliary feedwater pump 1-1, per STP P-AFW-PS11, on March 19, 2019 Reactor Coolant System (RCS) Leak Detection (IP Section 03.01) (1 Sample)

Unit 1, reactor coolant leakage and water inventory balance evaluation, per STP I-1A, on January 10, 2019

Surveillance Testing (IP Section 03.01) (3 Samples)

(1) Unit 1, 12 kV bus reactor coolant pump under-frequency and under-voltage calibration and time response test, per procedure STP I-9, on January 15, 2019

(2) Unit 1, integrated test of engineered safeguards and diesel generators, per procedure STP M-15, on February 14, 2019

(3) Unit 1, emergency core cooling system check valve leak test, per procedure STP V-5A1, on March 11, 2019

71114.04 - Emergency Action Level and Emergency Plan Changes Inspection Review (IP Section 02.01-02.03)

The licensee submitted a copy of the "Emergency Action Level Technical Basis Manual,"

Appendix D, Revision 5 of the site emergency plan, to the NRC on October 4, 2018. The inspectors conducted an in-office review of the changes from January 25 to February 5, 2019. This evaluation does not constitute NRC approval.

71124.01 - Radiological Hazard Assessment and Exposure Controls Radiological Hazard Assessment (IP Section 02.01)

The inspectors evaluated radiological hazards assessments and controls. The inspectors reviewed the following documents:

Radiological surveys:

• 66011, Weld Overlay

• 65894, Containment

• 66031, Containment

• 66023, Containment

• 65787, Containment

• 65937, Containment Risk significant radiological work activities (Radiation Work Permit (RWP)):

• RWP 19-1004: Weld Overlay

• RWP 19-1006: Pressurizer Boric Acid Removal RCS 1-8054B, 9365B, 8033A

• RWP 19-1023: 1R21 Fuel Movement Dose Rate Verification Air sample survey records:

64476 12/10/18 Particulate Air Sample Filter Storage in Vault 3 61738 05/02/19 17-F-001 Filter Liner Solidification Preps 65063 01/29/19 Weekly D-230 Noble Gas 65176 02/05/19 Weekly D-230 Noble Gas Instructions to Workers (IP Section 02.02)

The inspectors evaluated instructions to workers including radiation work permits (RWPs)used to access high radiation areas:

Radiation work permits:

• RWP 19-1020: 64128865-0100 Remove In-core Thermocouples

• RWP 19-1050: 641358461-0100 Inspect RCS Flange for RCS Leakage

• RWP 19-1070: 601109330-0000 RHR Weld Overlay Mobilization

• RWP 19-1004: 64146800-0100 Survey Letdown Line

• RWP 19-1006: 64135120-0020 Unit-1 Core Plate Vacuum Electronic alarming dosimeter alarms:

• DN 50975426: 04/10/18

• DN 50968587: 03/05/18

• DA 50968669: 03/05/18

• DN 50968145: 03/08/18

• DA 51017812: 02/19/19 Labeling of containers:

• No, 1060, Cl-36

• No. 1748, Cs-137

• No. 1750, Sr-90

• No. 1777, Cs-137 Contamination and Radioactive Material Control (IP Section 02.03) (1 Sample)

The inspectors evaluated licensee processes for monitoring and controlling contamination and radioactive material. The inspectors verified the following sealed sources are accounted for and are intact:

• No. 656, Cs-137

• No. 710, Cs-137

• No. 825, Cs-137 Radiological Hazards Control and Work Coverage (IP Section 02.04) (1 Sample)

The inspectors evaluated in-plant radiological conditions during facility walkdowns and observation of radiological work activities.

Radiological work permits for areas with airborne radioactivity:

• RWP 19-1020: 64128865-0100 Remove In-core Thermocouples

• RWP 19-1050: 641358461-0100 Inspect RCS Flange for RCS Leakage

• RWP 19-1070: 601109330-0000 RHR Weld Overlay Mobilization

• RWP 19-1004: 64146800-0100 Survey Letdown Line

• RWP 19-1006: 64135120-0020 Unit-1 Core Plate Vacuum High Radiation Area and Very High Radiation Area Controls (IP Section 02.05) (1 Sample)

The inspectors evaluated risk-significant high radiation area and very high radiation area controls.

Radiation Worker Performance and Radiation Protection Technician Proficiency (IP Section 02.06) (1 Sample)

The inspectors evaluated radiation worker performance and radiation protection technician proficiency.

71124.03 - In-Plant Airborne Radioactivity Control and Mitigation Engineering Controls (IP Section 02.01)

The inspectors evaluated airborne controls and radioactive monitoring. The inspectors reviewed the following systems:

Installed ventilation systems

• Units 1 and 2, auxiliary building ventilation

• Units 1 and 2, control room ventilation

• Unit 2, fuel handling building ventilation Temporary ventilation system setups

• reactor head stand

• containment 91 foot inside bioshield 1-4

• pressurizer FCV-459 valve work

• reactor coolant pump 1-2 work Portable or installed monitoring systems

• containment AMS-4 (multiple) and portable grab samplers

• SPING #6 and associated back up AMS-4

• pressurizer FCV-459 valve work

• residual heat removal weld overlay Use of Respiratory Protection Devices (IP Section 02.02) (1 Sample)

The inspectors evaluated respiratory protection. The inspectors reviewed the following evaluations and tests:

Evaluations for the use of respiratory protection

• ALARA Plan 19-1026, "Unit 1 - 1R21 Lower Reactor Cavity Entries," Revision 0

• ALARA Plan 19-1061, "Unit1 - 1R21 Containment Valves and Breaches," Revision 0

• RWP 19-1020, "1R21 Reactor Disassembly and Reassembly," Revision 0, Respirator Use Evaluation

• RWP 19-1026, "1R21 Lower cavity and Transfer Canal Work, Lower Cavity Decontamination," Revision 0, Respirator Use Evaluation Respiratory protection use during work activities

• RWP 19-1020, "1R21 Reactor Disassembly and Reassembly," Revision 0

• RWP 19-1061, "1R21 Containment Valves and Breaches," Revision 0 Medical fitness for use of respiratory protection devices

• three contract radiation protection technicians providing work coverage

• three in-house radiation protection technicians providing work coverage

• three on-shift crew A, Unit 1 reactor operators

• two radiation protection supervisors Observation of donning, doffing and functional test

• three on-shift crew A Unit 1 reactor operators and two radiation protection technicians Respiratory protection device evaluation

• MaxAir: 082A, 174A, 183A, 186A

• MSA SCBA kits: EP-AAAL088346, EP-AAAL088348, EP-AAAL88354

• Ultra Elite: UEL2004A, UEM0214A, UEM0244A, UES1057A

• Ultra Twin: UTL0382A, UTM0304A, UTM0452A, UTS0362A Self-Contained Breathing Apparatus for Emergency Use (IP Section 02.03) (1 Sample)

The inspectors evaluated SCBA program implementation.

Status and surveillance records for SCBAs:

Second Stage Serial No.

First Stage Serial No.

EP-APA202352 EAB183440 EP-OAB274213 EAB183439 EP-OAB282747 EAB213170 EP-OAB301743 EAB246744 SCBA fit for on-shift operators

• Three on-shift crew A, Unit 1 reactor operators SCBA maintenance check:

Second Stage Serial No.

First Stage Serial No.

EP-APA202352 EAB183440 EP-OAB274213 EAB183439 EP-OAB282747 EAB213170 EP-OAB301743 EAB246744

OTHER ACTIVITIES - BASELINE

===71151 - Performance Indicator Verification The inspectors verified licensee performance indicators submittals listed below:

IE01: Unplanned Scrams per 7000 Critical Hours Sample (IP Section 02.01) ===

(1) Unit 1, date range: 01/01/2018-12/31/2018

(2) Unit 2, date range: 01/01/2018-12/31/2018 IE03: Unplanned Power Changes per 7000 Critical Hours Sample (IP Section 02.02)

(2 Samples)

(1) Unit 1, date range: 01/01/2018-12/31/2018

(2) Unit 2, date range: 01/01/2018-12/31/2018 IE04: Unplanned Scrams with Complications (USwC) Sample (IP Section 02.03)

(2 Samples)

(1) Unit 1, date range: 01/01/2018-12/31/2018

(2) Unit 2, date range: 01/01/2018-12/31/2018 OR01: Occupational Exposure Control Effectiveness Sample (IP Section 02.15) (1 Sample)

(1) date range: 01/01/2018 - 12/31/2018 PR01: Radiological Effluent Technical Specifications/Offsite Dose Calculation Manual Radiological Effluent Occurrences (RETS/ODCM) Radiological Effluent Occurrences

Sample. (IP Section 02.16) (1 Sample)

(1) date range: 01/01/2018 - 12/31/2018

71152 - Problem Identification and Resolution Annual Follow-up of Selected Issues (IP Section 02.03)

The inspectors reviewed the licensees implementation of its corrective action program related to the following issues:

(1) evaluation and corrective actions associated with trends in reactor coolant pump seal leakage on March 5, 2019

(2) evaluation and corrective actions related to a Unit 1, polar crane main hoist overspeed trip during the reactor vessel head lift on March 9, 2019

71153 - Follow-up of Events and Notices of Enforcement Discretion Event Follow-up (IP Section 03.01)

The inspectors evaluated a sodium hydroxide leak from the Unit 2 containment spray system spray additive tank and the licensees response on March 4,

OTHER ACTIVITIES

- TEMPORARY INSTRUCTION, INFREQUENT, AND ABNORMAL Evaluation of Diablo Canyon Safety Condition in Light of Financial Conditions Because Pacific Gas & Electric Corporation, and its subsidiary, Pacific Gas & Electric Company (the licensee) was under bankruptcy protection/reorganization during the inspection period, NRC Region IV conducted special reviews of processes at Diablo Canyon. Using the flexibilities in the baseline inspection program, the inspectors evaluated several aspects of the licensees operations to determine whether the financial condition of the licensee impacted plant safety. The factors reviewed during this period included:

(1) impact on staffing,

(2) corrective maintenance backlog,

(3) changes to the planned maintenance schedule,

(4) corrective action program implementation, and

(5) any reduction in outage scope, including risk-significant modifications. Considering the licensees financial difficulties, the inspectors verified that licensee personnel continued to identify problems at an appropriate threshold and enter these problems into the corrective action program for resolution. The inspectors also verified that the licensee continued to develop and implement corrective actions commensurate with the significance of the problems identified.

The special review of processes at Diablo Canyon included continuous reviews by the resident inspectors, as well as the specialist-led baseline inspections completed during the inspection period - emergency preparedness, inservice inspection, and radiation protection - which are documented in this report. During this period, the inspectors did not identify any indications that the licensees financial circumstances were adversely affecting plant performance and safety.

INSPECTION RESULTS

Failure to Promptly Correct an Adverse Condition Related to Reactor Coolant Pump Seal Damage Cornerstone Significance Cross-cutting Report Aspect Section Initiating Events Green None 71152 NCV 05000275/2019001-01 Closed The inspectors reviewed a Green, self-revealed non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to promptly identify and correct a condition adverse to quality. Specifically, prior to February 4, 2019, corrective actions generated as a result of a root cause evaluation did not adequately correct a condition adversely impacting reactor coolant pump (RCP) seals.

Description:

System interconnections between RCP seal drains, the pressurizer relief tank (PRT), and the radioactive drain collection tank (RCDT) comprise a closed system. RCP seal drains are hard piped to the RCDT and are designed to be free flowing. Evolutions such as draining the PRT into the RCDT can result in pressure imbalances which, if not properly controlled, can result in the backfilling of the RCP seal drains. This pressure imbalance can result in RCP seal instability and damage. Damage to the RCP seals can manifest itself through increased calculated RCS leakage.

In February 2013, prior to refueling outage 2R17, the 2-2 RCP seals were damaged and required replacement. As a result of this damage, PG&E performed a root cause evaluation and identified a root cause of backfilling of the RCP seals leading to seal damage. The root cause generated corrective actions that focused on correcting procedures that could lead to the backfill condition, as well as additional operator training on the RCDT/PRT system interconnections and draining procedures to prevent such damage from recurring.

On January 8, 2019, prior to the Unit 1 refueling outage 1R21, the inspectors noted increasing trends in reactor coolant system leakage and discussed with PG&E their actions to address this trend. PG&E initially believed that the observed changes to RCS leakage were due to leak rate calculational inaccuracies caused by the increasing frequency of dilutions required at end of core life. Additional review by PG&E prompted by a continued upward trend in identified RCS leakage rate determined the source of the RCS leakage to be from the 1-2 RCP #2 seal. Inspection of the 1-2 RCP seal package during the outage identified seal damage and resulted in the decision being made to replace the seal assembly.

PG&E subsequently determined that the draining of the PRT to the RCDT while carrying out the actions of alarm response procedure AR PK-05-25, Pressurizer Relief Tank (PRT)

Pressure/Level/Temperature, resulted in the backfilling of the system and damage to the seals. PG&E determined that revisions to this procedure should have been included as a corrective action from the 2013 root cause evaluation but were missed and not included.

Corrective Action(s): After the cause of the RCP seal leakage was identified, PG&E revised Annunciator Response Procedure AR PK-05-25, Pressurizer Relief Tank (PRT)

Pressure/Level/Temperature, to ensure proper control when draining the PRT to the RCDT during power operations and conducted additional operator training.

Corrective Action Reference(s): Notification 51014929

Performance Assessment:

Performance Deficiency: The inspectors determined that PG&Es failure to promptly identify and correct a degraded condition associated with the reactor coolant pumps seals during draining of the PRT to the RCDT was a performance deficiency within PG&Es ability to foresee and correct.

Screening: The performance deficiency was more than minor because it impacted the equipment performance attribute of the Initiating Events Cornerstone and its objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the licensees failure to ensure that appropriate corrective actions were taken in response to RCP seal damage caused by improper system operation in 2013 resulted in RCP seal damage to recur in 2019.

Significance: The inspectors assessed the significance of the finding using Appendix A, Significance Determination of Reactor Inspection Findings for At - Power Situations. The issue was evaluated in accordance with IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, and determined to be of very low safety significance (Green) since it did not result in exceeding the RCS leak rate for a small break LOCA and did not affect systems used to mitigate a LOCA.

Cross-cutting Aspect: The inspectors did not identify a cross-cutting aspect associated with this finding because it did not reflect current performance.

Enforcement:

Violation: Title 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requires, in part, that measures shall be established to assure that conditions adverse to quality are promptly identified and corrected. Contrary to the above, prior to February 4, 2019, the licensee failed to assure that a condition adverse to quality was promptly identified and corrected. Specifically, in conducting a root cause evaluation in 2013 on RCP seal vulnerability to damage when draining the PRT to the RCDT, the licensee failed to adequately identify and correct all pertinent operating procedures.

Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2.a of the Enforcement Policy.

Inadequate Corrective Actions Leads to Increased Time at Yellow Risk and Lowered Inventory Cornerstone Significance Cross-cutting Report Aspect Section Initiating Events Green [P.3] - 71152 FIN 05000275/2019001-02 Resolution Closed The inspectors reviewed a Green, self-revealed finding in that PG&E personnel failed to promptly correct a condition in accordance with OM7.ID1, Problem Identification and Resolution. Specifically, personnel did not promptly correct a condition associated with the Unit 1 polar crane variable frequency drive (VFD) motor used during the removal and replacement of the reactor vessel head from the reactor vessel during refueling outages that was first identified in 2015. This led to a recent event in which the reactor vessel head was suspended above the reactor vessel during a period of yellow outage safety risk with reactor vessel water level in lowered inventory condition for approximately 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> longer than necessary.

Description:

On March 8, 2019, the Diablo Canyon Power Plant, Unit 1, polar crane main hoist tripped during reactor head installation preventing the reactor vessel head from being raised or lowered. At the time of the trip, the reactor vessel head was suspended approximately 30 feet above the reactor vessel, reactor core, and upper internals. The station was in a yellow risk condition due, in part, to reactor vessel water level being in a lowered inventory condition. The reactor vessel head was in the credited heavy lift load path at all times, and all load drop safety devices on the polar crane functioned as designed.

After the polar crane main hoist trip, refueling outage day shift personnel commenced fact finding and troubleshooting activities. As a conservative measure, the containment equipment hatch was closed, and contingency preparations were made to flood up the refueling cavity. Approximately 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> later, shortly after turnover with day shift, outage nightshift subject matter experts recognized the fault was due to a VFD overspeed trip and promptly reset the condition, allowing the head to be lowered to its intended position.

Additional investigation by PG&E, including a review of the corrective action program database, revealed that the polar crane hoist mechanism for both units had suffered from similar VFD overspeed trip conditions on multiple occasions and that a work request had been generated in 2015 to replace a coupling device found in the VFD with a different type not susceptible to unnecessary overspeed trips. This work request was deleted from the scope of outage 1R19 and moved to 1R20. It was then deleted from the scope of 1R20 and moved to outage 1R21. During outage 1R21, the work was again deleted and deferred to a later outage. Additionally, no formal communication, operating experience, or procedure precautions or notes were generated as a result of the identified condition. These failures to promptly correct a condition associated with the polar crane VFD led to the reactor vessel head being unnecessarily suspended approximately 30 feet above the reactor vessel, core, and upper internals and the station remaining in a yellow risk condition with reactor vessel water level in a lowered inventory state for approximately 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> longer than required.

Corrective Action(s): After the event, PG&E performed an evaluation of the event, initiated changes to the applicable procedures, and scheduled work to upgrade the VFD such that similar events will not occur.

Corrective Action Reference: Notification 51017606

Performance Assessment:

Performance Deficiency: The inspectors determined that PG&Es failure to promptly correct a condition associated with the polar crane VFD motor used during the removal and replacement of the reactor vessel head from the reactor vessel during refueling outages was a performance deficiency within PG&Es ability to foresee and correct.

Screening: The performance deficiency was more than minor because it impacted the configuration control attribute of the Initiating Events cornerstone and its objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to promptly correct a condition associated with the polar crane VFD led to an event in which the reactor vessel head was suspended above the reactor vessel during a period of yellow outage safety risk with reactor vessel water level in lowered inventory condition for approximately 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> longer than necessary.

Significance: The inspectors assessed the significance of the finding using Appendix G, Shutdown Safety SDP. Because the plant was shutdown with at least one fuel bundle in the reactor and temperature and pressure within normal decay heat removal conditions, the finding was evaluated in accordance with Attachment 1 of IMC 0609, Appendix G, Shutdown Operations Significance Determination Process, dated May 9, 2014. The finding was determined to be of very low safety significance (Green) since all questions in Attachment 1, Exhibit 2 were answered NO.

Cross-cutting Aspect: This finding is related to the cross-cutting area of Problem Identification and Resolution - Resolution because PG&E did not take effective corrective actions to address issues in a timely manner commensurate with its safety significance

[P.3]. Specifically, PG&Es failure to promptly correct a condition associated with the polar crane VFD led to the reactor vessel head being unnecessarily suspended approximately 30 feet above the reactor vessel, core, and upper internals and the station remaining in a yellow risk condition with reactor vessel water level in a lowered inventory state for approximately 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> longer than required.

Enforcement:

Inspectors did not identify a violation of regulatory requirements associated with this finding.

Human Performance Event Leads to Sodium Hydroxide Leak and Loss of Level Indication for Containment Spray System Spray Additive Tank Cornerstone Significance Cross-cutting Report Aspect Section Mitigating Green [H.12] - Avoid 71153 Systems NCV 05000323/2019001-03 Complacency Closed The inspectors reviewed a Green, self-revealed non-cited violation of Technical Specification 5.4.1.a Procedures, because personnel failed to follow the requirements of MP I-12-L931, Spray Additive Tank Level Channel 931 Calibration, Revision 2. Specifically, on March 4, 2019, during performance of a calibration of a Unit 2 containment spray additive tank level transmitter, PG&E technicians inadvertently removed a venting plug from the wrong level transmitter resulting in a pressurized leak of sodium hydroxide solution from the spray additive tank, and the inability to determine the level of the spray additive tank.

Description:

The containment spray system is a safety-related system designed to actuate in the event of certain accident scenarios to minimize peak pressure in containment and to reduce iodine fission product inventory in the containment atmosphere. The spray additive system is a subsystem of the containment spray system that provides a solution of sodium hydroxide to the containment spray pump discharge via a storage tank and an eductor. Two redundant level indicators, LI-931 and LI-932, are provided to ensure that the storage tank maintains the technical specification (TS) required level of sodium hydroxide solution, and so that actions can be taken during emergency operation prior to draining the spray additive tank to the point that the containment spray pumps become susceptible to air entrainment and binding.

On March 4, 2019, procedure MP I-12-L931, Spray Additive Tank Level Channel 931 Calibration, was being performed on LT-931, one of the two, Unit 2, containment spray system spray additive tank level indicators to ensure its proper functionality. This procedure is used to satisfy a TS surveillance requirement. During performance of this procedure, maintenance technicians inadvertently removed the venting plug from the in-service level indicator, LT-932, instead of the procedurally directed level indicator LT-931. This resulted in a pressurized stream of caustic sodium hydroxide solution coming from the vent plug for approximately 45 minutes and evacuation of the area until emergency response personnel were able to halt the leak. During this time, approximately 300 gallons of sodium hydroxide solution was released from the tank and the Unit 2 containment spray system was declared inoperable until operations personnel could verify that the spray additive tank levels were within TS required limits. The system remained inoperable for approximately 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br />. Later evaluation determined that the tank level never decreased below the TS required limit. Additionally, had they been called upon during an emergency event, emergency operating procedures for the containment spray system require that the spray additive tank be secured prior to reaching 16 percent level to ensure that the containment spray pumps do not become susceptible to damage from air entrainment and binding. During the approximately 22-hour period prior to restoration of level indication, operations personnel would not have been able to carry out this requirement due to a lack of valid spray additive tank level indication.

Corrective Action(s): After the leak was stopped and the caustic sodium hydroxide solution was removed from the area, PG&E personnel restored the level transmitter to service and verified that spray additive tank levels were within TS required limits. Additionally, PG&E initiated a cause evaluation and implemented corrective actions including a department level clock reset, station wide communication, and emphasis on the importance of the use of human performance tools prior to any work on station equipment.

Corrective Action Reference(s): Notification 51020536

Performance Assessment:

Performance Deficiency: The inspectors determined that PG&Es failure to follow procedure MP I-12-L931, Spray Additive Tank Level Channel 931 Calibration, Revision 2, was a performance deficiency within PG&Es ability to foresee and correct.

Screening: The performance deficiency was 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 response to initiating events to prevent undesirable consequences. Specifically, the human performance event associated with the containment spray system spray additive tank level transmitter calibration resulted in a pressurized leak of sodium hydroxide solution from the spray additive tank, evacuation of the area, and the inability to detect spray additive tank level (as required by emergency procedures) for approximately 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br />.

Significance: The inspectors assessed the significance of the finding using Appendix A, Significance Determination of Reactor Inspection Findings for At - Power Situations. The issue was evaluated in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, and determined to be of very low safety significance (Green) since the sodium hydroxide leak and inoperability of the Unit 2 containment spray system did not result in a loss of system function.

Cross-cutting Aspect: This finding is related to the cross-cutting area of Human Performance

- Avoid Complacency because PG&E did not recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. This included the implementation of appropriate error reduction tools [H.12]. Specifically, PG&E technicians failed to perform appropriate activity and work site reviews prior to performing work, use appropriate error reduction tools, or consider potential undesired consequences when conducting a level transmitter calibration procedure on the Unit 2 containment spray system.

Enforcement:

Violation: Technical Specification 5.4.1.a, Procedures, 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. Contrary to the above, on March 4, 2019, PG&E personnel failed to follow the requirements of procedure MP I-12-L931, Spray Additive Tank Level Channel 931 Calibration, Revision 2. Specifically, on March 4, 2019, during performance of a calibration of a Unit 2 containment spray additive tank level transmitter, PG&E technicians inadvertently removed a venting plug from the wrong level transmitter resulting in a pressurized leak of sodium hydroxide solution from the spray additive tank, evacuation of the area, and the inability to detect spray additive tank level.

Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2.a of the Enforcement Policy.

EXIT MEETINGS AND DEBRIEFS

The inspectors verified no proprietary information was retained or documented in this report.

• On February 5, 2019, the inspectors presented the emergency preparedness inspection results to Mr. M. Ginn, Manager, Emergency Preparedness, and other members of the licensee staff.

• On February 26, 2019, the inspectors presented the inservice inspection results to Ms. P. Gerfen, Station Director, and other members of the licensee staff.

• On February 28, 2019, the inspectors presented the baseline radiation protection inspection results to Ms. P. Gerfen and other members of the licensee staff.

• On April 4, 2019, the inspectors presented the quarterly resident inspection results to Mr. J. Welsch, Vice President of Nuclear Generation and Chief Nuclear Officer, and other members of the licensee staff.

DOCUMENTS REVIEWED

71111.01: Adverse Weather Protection

Procedure

Number Title Revision

OP O-28 Operations Response to Storms or Biofouling 22

71111.04: Equipment Alignment

Notifications

51012721 51016118

Procedures

Number Title Revision

MP M-51.5 Testing and Maintenance of Safety/Relief Valves 32

OP A-2:X RVRLIS Alignment for Refueling Outage 9

OP B-2:1 Residual Heat Removal Alignment Verification 26

OP B-2:V RHR - Place in Service 37

OP B-2:VI RHR - Draining the Refueling Cavity 29

OP B-2:VII Residual Heat Removal System-Fill and Vent 16

OP J-6B:II-A Diesel Generator 2-3 - Alignment Checklist 0

OP J-6B:III-A Diesel Generator 2-2 - Alignment Checklist 0

Drawing

Number Description Revision

106710-2 RHR - System OVID 43

71111.05: Fire Protection

Notifications

50948204 50996387 51003929

Work Orders

60109333 64136730

Procedure

Number Title Revision

PEP 18-03 Containment Fire Water Jumper Installation and Removal 3

Drawings

Number Description Revision

PA1/2 Pre-Fire Plans Intake Structure 2

PA26/27 Pre-Fire Plans Containment Building 91 and 117, Unit 1 4

RA-29 Containment Building Elevation 140, Unit 1 4

RA-7 Radiological Control Area & H Block Elev. 100, Unit 2 8

Other

Number Description Date

Hot Work Permit 1769 03/09/2019

71111.07: Heat Sink Performance

Notifications

51021896 51018644 51022042

Work Orders

220069 64220070

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

27985 Component Cooling Water Heat Exchangers CCW 1-1 and 02/2019

1-2 Eddy Current Report

71111.08: Inservice Inspection Activities (PWR)

Notifications

50315950 50914808 50915722 50915871

50920366 50920367 50920368 50920369

50921421 50925730 50927788 50928799

50956677 50956730 50998093 50988345

51004997 51008168 51017806

Work Orders

64025645 60105066 68052869 60100625

60100791 60094591 64132540 60101082

Procedures

Number Title Revision

AD4.ID2 Plant Leakage Evaluation 12

AD7.ID11 Fluid Leak Management Program 5

ER1.ID2 Boric Acid Corrosion Control Program 7

ISI P-SI-18 SIS Test System Piping Pressurization for ISI Inspection 7

ISI X CRDM Reactor Vessel Top and Bottom Head Visual Inspections 6

ISI_P-AFW-13U1 ISI Pressure Test of AFW Suction Supply Piping Aligned to 2

the Raw Water Storage Reservoir Using a Hydro Pump

NDE VT 2-1 Visual Examination During Section XI System Pressure Test 4

NDE_PDI-UT-2 Ultrasonic Examination of Austenitic Piping 13

NDE_PT-1 Visible Dye Liquid Penetrant Examination Procedure 6

NDE_UT-4 Ultrasonic Examination of Pressure Vessel Welds Other 4

than Reactor Vessels

NDE_UT-I Ultrasonic Detection and Sizing of Vessel Nozzle Inside 0

Radius

NDE_VT_1-1 Visual Examination of Component Surfaces 2

Procedures

Number Title Revision

NDE_VT_2-1 Visual Examination During Section XI System Pressure Test 4

NDE_VT_3-1 Visual Examination of Component and Piping Supports 3

OM8.ID4 Control of Flammable and Combustible Materials 29

PD MA3 Special Processes 3

STP M-86 Leak Reduction of Systems Outside Containment Likely to 22

Contain Radioactive Materials following an Accident

(NUREG-0737)

STP M-87 Operational Leakage of ECCS Recirc Paths Outside 20

Containment

STP R 8C Containment Walkdown for Evidence of Boric Acid Leakage 10

Drawing

Number Description Revision

663056 Auxiliary Feedwater Terry Turbine Pump 2

71111.11: Licensed Operator Requalification Program and Licensed Operator

Performance

Notification

51016118

Procedures

Number Title Revision

EOP E-0 Reactor Trip or Safety Injection 45B

JSIM-507 Evaluator Skills 0

OP B-2:V RHR-Place in Service 37

OP L-4 Normal Operations at Power 97

OP L-5 Plant Cooldown from Minimum Load to Cold Shutdown 105

OP L-5 Plant Cooldown From Minimum Load to Cold Shutdown 105A

OP L-7 Plant Stabilization Following Reactor Trip 25A

TQ2.DC3 Licensed Operator Continuing Training Program 29

71111.12: Maintenance Effectiveness

Notifications

51012969 51021080 50605355 50288884

50636744 51017754

Other

Revision /

Number Description Date

DCN 24558 Damper Modifications 4 kV Rooms 11/28/2018

PC-23E1-01 Maintenance Rule Impacted Function 1

71111.13: Maintenance Risk Assessments and Emergent Work Control

Procedures

Number Title Revision

AD7.DC6 On-Line Maintenance Risk Management 26

OP A-2:X RVRLIS Alignment for Refueling Outages 9

OP O-36 Protected Equipment Postings 18

Other

Number Description Revision

1R21 Outage Safety Plan 0

71111.15: Operability Determinations and Functionality Assessments

Notifications

51011651 51011752 51011572 51010580

51016998 51016118 51016269 51022188

51022414 51020413 51020574 51020501

Work Orders

60116677 60117408

Procedures

Number Title Revision

OM7.ID12 Operability Determinations 38

Procedures

Number Title Revision

STP I-1B Routine Daily Checks Required by Licenses 129

STP R-10C Reactor Coolant System Water Inventory Balance 46

Drawing

Number Description Revision

106710 Residual Heat Removal, Sheet 2 36

Other

Number Description Date

Unit 1 Containment Pressure Trend Data 01/2019

Unit 1 Containment Temperature Trend Data 01/2019

Unit 1 Radiation Monitor RM-11 (Containment Air 01/2019

Particulate) Trend Data

Unit 1 Radiation Monitor RM-12 (Containment Radiological 01/2019

Gas) Trend Data

Unit 1 RCS Leak Rate Calculation Packages per STP I-1B 01/20/2019 -

01/30/2019

Unit 1 RCS Leak Rate Calculation Packages per STP R- 01/20/2019 -

10C 01/30/2019

Unit 1 Reactor Coolant Pump 1-1, 1-2, 1-3, and 1-4 Seal 01/2019

No. 1 Leakoff Trend Data

Unit 1 Reactor Coolant Pump 1-1, 1-2, 1-3, and 1-4 Seal 01/2019

Water Inlet Trend Data

51011572 Emerging Issue - Elevated Unit 1 RCS Calculated Leak 01/15/2019,

Rates 01/29/2019,

01/31/2019

71111.18: Plant Modifications

Notifications

50986428 51015255 51015257 51015176

51015256 51011572

Work Order 68053881

Procedures

Number Title Revision

CF3.ID9 Design Change Notice 56

MP M-21.8 Diesel Engine Governor Actuator Maintenance 25

TS3.ID2 Licensing Basis Impact Evaluations 44A

Other

Number Description Revision

DCP 25262 EDG Governor Controls Upgrade 0

71111.19: Post-Maintenance Testing

Notifications

51023045 51022989 51018293 51017982

51018598 51019542 51023045 51022989

Work Orders

64159084 64158929 60116970 60116968

64123070 60117148 60102262 64157306

Procedures

Number Title Revision

MP E-50.1 Thermal Overload Relay and Cubicle Maintenance 53

MP E-57.10C 4 kV Motor Preventative Maintenance Testing 8

MP E-67.3C Maintenance of Solidstate Controls 400A Vital Station 11

Battery Chargers

STP M-12B Battery Charger Performance Test 17

STP M-83A Penetration Overcurrent Protection 29A

STP P-ASW-11 Routine Surveillance Test of Turbine-Driven Auxiliary 11

Feedwater pump 1-1

STP P-CCP-23 Routine Surveillance Test of Centrifugal Charging Pump 2-3 8

71111.20: Refueling and Other Outage Activities

Notifications

51018584 51018358 51019722 51016530

51019158 51017079 51014929 51016562

51019171 51011572 51016362 51016543

51015960 51017642 51023590 51018335

51016804 51016680 51016126

Work Order 68053873

Procedures

Number Title Revision

AD8.DC51 Outage Safety Management Control of Off-Site Power 18

Supplies to Vital Buses

AD8.DC54 Containment Closure 15

AD8.DC55 Outage Safety Scheduling 42

EOP E-0 Reactor Trip of Safety Injection 45B

ER1.ID2 Boric Acid Corrosion Control Program 7

ISI X-2 Visual Inspection of Reactor Upper Internals 1

MP I-2.28 Activation and Deactivation of the Rx Vsl Refueling Lvl 28A

Indication System (RVRLIS)

MP M-45.1 Containment Equipment Hatch Door Opening and Closing 14

MP M-7.6A Reactor Vessel Upper Internals Removal 15

OM14.ID1 Fatigue Management Rule Program 29

OM14.ID1 Fatigue Management Rule Program 29

OP A-2:II Reactor Vessel - Draining the RCS to the Vessel Flange - 49

With Fuel in Vessel

OP A-2:IX Reactor Vessel - Vacuum Refill of the RCS 25

OP A-2:X RVRLIS Alignments for Refueling Outages 9

OP B-1A:VIII Makeup Control System Operation 60

OP B-2:V RHR - Place in Service 37

OP B-2:VI RHR-Draining the Refueling Cavity 29

OP B-8DS2 Core Loading 66

Procedures

Number Title Revision

OP L-0 Mode Transition Checklists 85

OP L-2 Hot Standby to Startup Mode 44

OP L-4 Normal Operation at Power 97

OP L-5 Plant Cooldown From Minimum Load to Cold Shutdown 105A

OP L-7 Plant Stabilization Following Reactor Trip 25A

OP2.ID1 Clearances 42

STP M-45A Containment Inspection Prior to Establishing Containment 34A

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

57725 Containment Equipment Drawing 91&100 foot 32

1R21 Outage Safety Plan 0

Calculation Mode Transition With CFCUs 1-4 and 1-4 Unavailable 0

PRA19-02

DCPP Form 69- Fatigue Management Rule Waiver Attachment 3 02/12/2019

21350

EmpCenter Time Hours Entry Data Sheets 02/18/2019

Entry

Security Door Door Reader Transactions 02/10/2019

History

71111.22: Surveillance Testing

Notifications

51016229 51016422 51016489 51016141

51910387 51011572 50910387 51010580

51012111 51012113 51012096 51022989

51023045 51023787

Work Orders

222140 64033435 64194016 64136202

201622 64136401

Procedures

Number Title Revision

ER1.ID2 Boric Acid Corrosion Control Program 7

STP I-1B Routine Daily Checks Required By Licenses 129

STP I-9 12 kV Bus RCP U/F and U/V Channels Calibration and Time 10

Response Test

STP M-15 Integrated Test of Engineered Safeguards and Diesel 71

Generators

STP M-7 Integrated Leak Rate Test (ILRT) Type A 27

STP M-7E Containment Penetration Valve Lineup for the Integrated 10

Leakage Rate Test (ILRT)

STP M-7W Containment Structural Integrity Inspection 5

STP P-AFW- Preservice Testing of Turbine-Driven Auxiliary Feedwater 7

PS11 Pump 1-1

STP R-10 Reactor Coolant System Leakage Evaluation 12

STP R-10C Reactor Coolant System Water Balance Inventory 46

STP V-5A1 Emergency Core Cooling System Check Valve Leak Test, 21

Post-Refueling/Post-Maintenance Valves 8819 A-D and

8956 A-D

STP V-600 General Containment Isolation Valve Leak Test 28

STP V-654B Penetration 52B Containment Isolation Valve Leak Testing 11

Drawing

Number Description Revision

106709 Safety Injection, Sheet 2 61

Other

Number Description Revision

ANSI/ANS-56.8- Containment System Leakage Testing Requirements 2

2002

71114.04: Emergency Action Level and Emergency Plan Changes

Miscellaneous Documents

Number Title Date

PG&E Letter Diablo Canyon Units 1 and 2; Docket No. 50-275, OL-DPR- 10/25/2016

DCL 16-099 80; Docket No. 50-323, OL-DPR-82; License Amendment

Request 16-04; Request to Adopt Emergency Action Level

Schemes Pursuant to NEI 99-01, Revision 6, Development

of Emergency Action Levels for Non-Passive Reactors

PG&E Letter Docket No. 50-275, OL-DPR-80; Docket No. 50-323, OL- 06/21/2017

DCL-17-055 DPR-82; Diablo Canyon Units 1 and 2; Response to NRC

Request for Additional Information Regarding License

Amendment Request 16-04, Request to Adopt Emergency

Action Level Schemes Pursuant to NEI 99-01, Revision 6,

Development of Emergency Action Levels for Non-Passive

Reactors

PG&E Letter Diablo Canyon Units 1 and 2; Docket No. 50-275, OL-DPR- 08/17/2017

DCL-17-073 80; Docket No. 50-323, OL-DPR-82; Supplement to

License Amendment Request 16-04; Request to Adopt

Emergency Action Level Schemes Pursuant to NEI 99-01,

Revision 6, Development of Emergency Action Levels for

Non-Passive Reactors

PG&E Letter Docket No. 50-275, OL-DPR-80; Docket No. 50-323, OL- 10/04/2018

DCL-18-077 DPR-82; Diablo Canyon Units 1 and 2; Docket No 72026,

PG&E Letter Materials License No. SNM-2511; Diablo Canyon

DIL-18-013 Independent Spent Fuel Storage Installation; Emergency

Plan Implementing Procedure Update

71124.01: Radiological Hazard Assessment and Exposure Controls

Notifications

51012602 51012603 51012604 51017804

50975426 50968020 50968587 50969660

50989145 51019595

Procedures

Number Title Revision

RCP NISP-RP.05 Access Controls for High Radiation Areas 1

RCP NISP-RP.02 Radiation & Contamination Surveys 0

RCP D-335 Radiation Exposure Reporting 6

RCP D-620 Radioactive Source Control Program 15

Procedures

Number Title Revision

RCP NISP-RP.04 Radiological Posting & Labelling 1

RP1-1D17 Control & Storage of Non-SNM 0

RP1-1D16 Radiation Work Expectations 12

RP1-1D9 Radiation Work Permits 13

RP1-1D14 Radioactive Material Controls 8

RP1-1D15 Radiological Risk Assessment 5

Audits and Assessments

Number Description Date

173180001 2018 Radiation Protection Programs Audit 02/25/2018

QV 180850014 Quality Verification 2R20 Integrated Outage Assessment 04/17/2018

Notification NRC Pre-Inspection on Radiological Hazard Assessment 01/14/2019

51012084 and Exposure Controls 71124.01

Radiation Work Permits

Number Description Revision

RWP 19-1004 WO 64146800-0100 Survey Letdown Line 0

RWP 19-1006 WO 64135120-0020 Unit-1 Core Plate Vacuum 0

RWP 19-1020 WO 64128865-0100 Remove In-core Thermocouples 0

RWP 19-1050: WO 641358461-0100 Inspect RCS Flange for RCS

Leakage 0

RWP 19-1070 WO 601109330-0000 RHR Weld Overlay Mobilization 0

Radiological Surveys

Number Description Date

65894 Containment 02/24/2019

65937 Containment 02/25/2019

66011 Weld Overlay 02/26/2019

66023 Containment 02/26/2019

66031 Containment 02/26/2019

66787 Containment 02/22/2019

Air Sample Surveys

Number Description Date

61738 Air Sample: 17-F-001 Filter Liner Solidification Preps 05/02/2018

64476 Air Sample: Particulate Air Sample Filter storage in Vault 3 12/10/2018

65063 Air Sample: D-230 Noble Gas 01/29/2019

65536 Air Sample: D-230 Noble Gas 02/05/2019

Miscellaneous Documents

Number Description Date

Ready for Issue High, Locked High, and Very High 02/26/2019

Radiation Area Key Inventory

GPAA102 DCPP General Employee Training 06/30/2016

NRC Form 748 NSTS Annual Inventory Reconciliation Form 01/08/2019

RPS17J DCPP Radiological Self-Monitoring 03/18/2014

WO 64178986 Semi-annual Sealed Source Inventory/Leak Test 02/05/2019

WO 64187031 Semi-Annual Sealed Source Inventory/Leak Test 08/18/2018

71124.03: In-Plant Airborne Radioactivity Control and Mitigation

Notifications

50930710 50931849 50934622 50944439

50946792 50947936 50950636 50954291

50960981 50961668 50965343 50966263

50975820 50976449 50978069 50999314

51012085 51013771

Procedures

Number Title Revision

OM6.ID10 Respiratory Protection Program 8

PEP M-92 Post LOCA Sampling Center Ventilation Systems DOP and 4

HALIDE Penetration Tests

RCP D-410 Issuing Respiratory Protective Equipment 20

RCP D-645 HEPA Integrity Testing 3

RCP D-700 Processing Respiratory Protection Equipment at DCPP 17

RCP D-707A MSA FireHawk (NIOSH) Self-Contained Breathing 1A

Procedures

Number Title Revision

Apparatus Inspection

RCP D-709 Posi-Chek3 Tester Operation 2

RCP D-712 MAXAIR Powered Air Purifying Respirator 1

RCP D-732 Respirator Fit Testing 26

RCP D-772 UNICUSIII Cylinder Recharging Station Operation 4

RCP NISP-RP.03 Radiological Air Sampling 1

RCP NISP-RP.08 Use and Control HEPA Filtration and Vacuum Equipment 1

STP M-53 Control Room Ventilation System DOP and Halide 21A

Penetration Tests

STP M-6A Routine Surveillance Testing of Control Room Ventilation 57

System

TQ1.DC20 Respirator Training 8

Audits and Self-Assessments

Number Description Date

DS-16-008 NUPIC Commercial Grade Survey - F&J Specialty Products 08/02/2016

229 NUPIC Audit -GEL Laboratories, LLC 11/03/2016

24351 NUPIC Audit - Mirion Technologies (Canberra), Inc. 11/06/2017

24371 NUPIC Audit - Reuter-Stokes LLC 12/21/2017

ASMP SAQH Quick Hit Self-Assessment (QHSA) - NRC Inspection

51012085 71124.03 01/29/2019

Respirator Testing, Inspection, and Inventory Records

Number Description Date

Weekly Respirator Locker Inventory 01/14/2019

EP-APA202352 Complete SCBA Test 01/09/2019

EP-APA202352 Monthly and Annual Inspection and Maintenance Checklist

for MSA Firehawk SCBA 2017-2018

EP-OAB274213 Complete SCBA Test 02/01/2018

EP-OAB274213 Complete SCBA Test 02/02/2017

EP-OAB274213 Monthly and Annual Inspection and Maintenance Checklist

for MSA Firehawk SCBA 2017-2018

Respirator Testing, Inspection, and Inventory Records

Number Description Date

EP-OAB282747 Monthly and Annual Inspection and Maintenance Checklist

for MSA Firehawk SCBA 2017-2018

EP-OAB301743 Complete SCBA Test 02/22/2017

EP-OAB301743 Monthly and Annual Inspection and Maintenance Checklist 2017-2018

for MSA Firehawk SCBA

Engineered System Filter Test Records

Number Description Date

60076341-5000 STP G-11, Procedure for Obtaining Charcoal Filer Media

for laboratory Testing (Methyl Iodine): Fuel Handling

Building Ventilation System, Revision 20 [old charcoal] 01/14/2019

60076341-5010 STP G-11, Procedure for Obtaining Charcoal Filer Media

for laboratory Testing (Methyl Iodine): Fuel Handling

Building Ventilation System, Revision 20 [new charcoal] 01/14/2019

60078958-5000 STP G-11, Procedure for Obtaining Charcoal Filer Media

for laboratory Testing (Methyl Iodine): Control Room

Ventilation System, Revision 20 04/25/2017

64114011-0100 STP M-53, Unit 2 Control Room Ventilation System DOP

and Halide Penetration Tests, Revision 13 04/19/2017

64114015-0200 STP G-9, Unit 2 General HEPA Filter Bank Penetration

Test: Control Room Ventilation System, Revision 9; and

STP G-10, Unit 1 General Charcoal Filter Bank Penetration

Test: Control Room Ventilation System, Revision 8 04/25/2017

64156652-0100 STP G-9, Unit 1 General HEPA Filter Bank Penetration

Test: Fuel Handling Building Ventilation System, Revision 9

[old charcoal] 01/14/2019

64156652-0200 STP G-9, Unit 1 General HEPA Filter Bank Penetration

Test: Fuel Handling Building Ventilation System, Revision 9

[new charcoal] 01/14/2019

64156654-0100 STP G-10, Unit 1 General Charcoal Filter Bank Penetration

Test: Fuel Handling Building Ventilation System, Revision 8

[old charcoal] 01/14/2019

64156654-0200 STP G-10, Unit 1 General Charcoal Filter Bank Penetration

Test: Fuel Handling Building Ventilation System, Revision 8

[new charcoal] 01/14/2019

64156655-0100 STP M-41, Unit 1 Fuel Handling Building Ventilation

System DOP and Halide Penetration Tests, Revision 20

[old charcoal] 01/14/2019

Engineered System Filter Test Records

Number Description Date

64156655-0300 STP M-41, Unit 1 Fuel Handling Building Ventilation

System DOP and Halide Penetration Tests, Revision 20

[new charcoal] 01/14/2019

64171179-0010 STP M-3A, Unit 1 DOP and Halide Penetration Test:

Auxiliary Building Ventilation System, Revision 23 02/06/2019

64171179-0040 STP G-9, Unit 1 General HEPA Filter Bank Penetration

Test: Auxiliary Building Ventilation System, Revision 9; STP

G-10, Unit 1 General Charcoal Filter Bank Penetration Test:

Auxiliary Building Ventilation System, Revision 8; and STP

G-11, Procedure for Obtaining Charcoal Filer Media for

laboratory Testing (Methyl Iodine): Auxiliary Building

Ventilation System, Revision 20 02/21/2019

Compressed Air System Testing Records

Number Description Date

Annual UNICUS Internal High-Pressure Storage Cylinders

Maintenance and Testing (DCPP Form 69-21535): ACPP1 2018 - 2019

Annual UNICUS Internal High-Pressure Storage Cylinders

Maintenance and Testing (DCPP Form 69-21535): ACPP2 2018 - 2019

Quarterly ACPP1 BAC 0-1 Trace Analytics CGA G-71. 03/13/2017 -

Grade L (SCBA) Air Quality Check 02/18/2019

Quarterly ACPP2 BAC 0-2 Trace Analytics CGA G-71. 03/13/2017 -

Grade L (SCBA) Air Quality Check 11/27/2018

Miscellaneous Documents

Revision /

Number Description Date

ALARA Plan 19-1026, "Unit 1, 1R21 Lower Reactor Cavity 01/29/2019

Entries"

ALARA Plan 19-1061, "Unit1, 1R21 Containment Valves 02/01/19

and Breaches

ALARA Plan 19-1071, Unit 1, 1R21 RHR Weld Overlay 01/16/2019

Respiratory Use Training, Fit Testing, and Medical 01/01/2017 -

Evaluation for Selected Staff 02/25/2019

RWP 19-1020, Unit 1, 1R21 Reactor Disassembly and 01/31/2019

Reassembly Respirator Use Evaluation

Miscellaneous Documents

Revision /

Number Description Date

RWP 19-1026, Unit 1, 1R21 Lower Cavity and Transfer 01/31/2019

Canal Work, Lower Cavity Decontamination Respirator

Use Evaluation

RWP 19-1061, "Unit 1, 1R21 Containment Valves and 0

Breaches

71151: Performance Indicator Verification

Procedures

Number Title Revision

AWP L-001 NRC Performance Indicators Initiating Events, SSFFs, and 9

MOR

NEI 99-02 Regulatory Assessment Performance Indicator Guideline 7

XI1.ID2 Regulatory Reporting Requirements 43

XI1.ID5 Collection and Submittal of NRC Performance Indicators 1

71152: Problem Identification and Resolution

Notifications

51014929 50538422 51014888 51023735

51013122 51011572 51012537 51018347

51018600 51022176 50268718 50403406

50544835 50663799 51017693 50351332

50535340 50662412 50815303 51017606

Procedures

Number Title Revision

AR PK05-25 Annunciator Response PRT PRESS/LVL TEMP 18

OM4.ID14 Notification Review Team (NRT) 31

OM7.ID1 Problem Identification and Resolution 53

OP1.DC40 Operations Equipment Deficiency and Adverse Condition 10

Monitoring

PAPERWORK REDUCTION ACT STATEMENT

This letter does not contain new or amended information collection requirements subject to the

Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). Existing information collection

requirements were approved by the Office of Management and Budget, Control

Number 31500011. The NRC may not conduct or sponsor, and a person is not required to

respond to, a request for information or an information collection requirement unless the

requesting document displays a currently valid Office of Management and Budget control

number.

This letter and its enclosure 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, and Requests for

Withholding.

Information Request

December 12, 2018

Notification of Inspection and Request for Information

Diablo Canyon Unit 1

NRC Inspection Report 05000275/2019001

INSERVICE INSPECTION DOCUMENT REQUEST

Inspection Dates: February 11 - 22, 2019

Inspector: Jim Drake

A. Information Requested for the In-Office Preparation Week

The following information should be sent to the Region IV office in hard copy or electronic

format or via a secure document management service, in care of Jim Drake, by January 22,

2019, to facilitate the selection of specific items that will be reviewed during the onsite

inspection week. The inspector will select specific items from the information requested

below and then request from your staff additional documents needed during the onsite

inspection week (Section B of this enclosure). We ask that the specific items selected from

the lists be available and ready for review on the first day of inspection. Please provide

requested documentation electronically if possible. If requested documents are large and

only hard copy formats are available, please inform the inspector(s), and provide subject

documentation during the first day of the onsite inspection.

If you have any questions regarding this information request, please call the inspector as

soon as possible.

On February 11, 2019, a reactor inspector from the Region IV office will perform the

baseline inservice inspection at Diablo Canyon Unit 1, using NRC Inspection

Procedure 71111.08, "Inservice Inspection Activities. Experience has shown that this

inspection is a resource intensive inspection both for the NRC inspector and your staff. The

date of this inspection may change dependent on the outage schedule you provide. In order

to minimize the impact to your onsite resources and to ensure a productive inspection, we

have enclosed a request for documents needed for this inspection. The information

identified on this request (Section A) is to be provided prior to the inspection to ensure that

the inspector(s) are adequately prepared. The section identified as Documents Upon

Request is intended to provide guidance to the type of information an inspector(s) will be

requesting to complete the inspection. It is important that all of these documents are up to

date and complete in order to minimize the number of additional documents requested

during the preparation and/or the onsite portions of the inspection (i.e., condition reports with

attachments).

We have discussed the schedule for these inspection activities with your staff and

understand that our regulatory contact for this inspection will be David Madsen of your

licensing organization. The tentative inspection schedule is as follows:

Preparation week: January 28 to February 1, 2019

Onsite weeks: February 11 - 22, 2019

Our inspection dates are subject to change based on your updated schedule of outage

activities. If there are any questions about this inspection or the material requested, please

contact Jim Drake at (817) 200-1558. (email to: James.Drake@nrc.gov)

A.1 ISI/Welding Programs and Schedule Information

1. A detailed schedule (including preliminary dates) of:

1.1. Nondestructive examinations planned for ASME Code Class Components

performed as part of your ASME Section XI, risk informed (if applicable), and

augmented inservice inspection programs during the upcoming outage.

1.2. Examinations planned for Alloy 82/182/600 components that are not included in

the Section XI scope (If applicable)

1.3. Examinations planned as part of your boric acid corrosion control program

(Mode 3 walkdowns, bolted connection walkdowns, etc.)

1.4. Welding activities that are scheduled to be completed during the upcoming

outage (ASME Class 1, 2, or 3 structures, systems, or components)

2. A copy of ASME Section XI Code Relief Requests and associated NRC safety

evaluations applicable to the examinations identified above.

2.1. A list of ASME Code Cases currently being used to include the system and/or

component the Code Case is being applied to.

3. A list of nondestructive examination reports which have identified recordable or

rejectable indications on any ASME Code Class components since the beginning of

the last refueling outage. This should include the previousSection XI pressure test(s)

conducted during start up and any evaluations associated with the results of the

pressure tests.

4. A list including a brief description (e.g., system, code class, weld category,

nondestructive examination performed) associated with the repair/replacement

activities of any ASME Code Class component since the beginning of the last outage

and/or planned this refueling outage.

5. If reactor vessel weld examinations required by the ASME Code are scheduled to

occur during the upcoming outage, provide a detailed description of the welds to be

examined and the extent of the planned examination. Please also provide reference

numbers for applicable procedures that will be used to conduct these examinations.

6. Copy of any 10 CFR Part 21 reports applicable to structures, systems, or

components within the scope of Section XI of the ASME Code that have been

identified since the beginning of the last refueling outage.

7. A list of any temporary non-code repairs in service (e.g., pinhole leaks).

8. Please provide copies of the most recent self-assessments for the inservice

inspection, welding, and Alloy 600 programs.

9. A copy of (or ready access to) most current revision of the inservice inspection

program manual and plan for the current interval.

10. Copy of NDE procedures for NDE that will be used during the outage.

11. Copy of overarching site procedure for welding.

A.2 Reactor Pressure Vessel Head

1. Provide a detailed scope of the planned bare metal visual examinations (e.g., volume

coverage, limitations, etc.) of the vessel upper head penetrations and/or any

nonvisual nondestructive examination of the reactor vessel head including the

examination procedures to be used.

1.1. Provide the records recording the extent of inspection for each penetration

nozzle including documents which resolved interference or masking issues

that confirm that the extent of examination meets 10 CFR 50.55a(g)(6)(ii)(D).

1.2. Provide records that demonstrate that a volumetric or surface leakage path

examination assessment was performed.

2. Copy of current calculations for EDY, and RIY as defined in Code Case N-729-1 that

establish the volumetric and visual inspection frequency for the reactor vessel head

and J-groove welds.

A.3 Boric Acid Corrosion Control Program

1. Copy of the procedures that govern the scope, equipment and implementation of the

inspections required to identify boric acid leakage and the procedures for boric acid

leakage/corrosion evaluation.

2. Please provide a list of leaks (including code class of the components) that have

been identified since the last refueling outage and associated corrective action

documentation. If during the last cycle, the unit was shutdown, please provide

documentation of containment walkdown inspections performed as part of the boric

acid corrosion control program.

A.4 Additional Information Related to all Inservice Inspection Activities

1. A list with a brief description of inservice inspection, and boric acid corrosion control

program related issues (e.g., PVAR) entered into your corrective action program

since the beginning of the last refueling outage. For example, a list based upon data

base searches using key words related to piping such as: inservice inspection,

ASME Code,Section XI, NDE, cracks, wear, thinning, leakage, rust, corrosion, boric

acid, or errors in piping examinations.

2. Provide training (e.g. Scaffolding, Fall Protection, FME, Confined Space) if they are

required for the activities described in A.1 through A.3.

3. Please provide names and phone numbers for the following program leads:

Inservice inspection (examination, planning)

Containment exams

Reactor pressure vessel head exams

Snubbers and supports

Repair and replacement program

Licensing

Site welding engineer

Boric acid corrosion control program

DOCUMENTS UPON REQUEST

Inservice Inspection / Welding Programs and Schedule Information

1. Updated schedules for inservice inspection/nondestructive examination activities,

including planned welding activities, and schedule showing contingency repair plans, if

available.

2. For ASME Code Class welds selected by the inspector please provide copies of the

following documentation (as applicable) for each subject weld:

• Weld data sheet (traveler).

• Weld configuration and system location.

• Applicable welding procedures used to fabricate the welds.

• Copies of procedure qualification records (PQRs).

• Welders performance qualification records (WPQ).

• Nonconformance reports for the selected welds (If applicable).

• Radiographs of the selected welds and access to equipment to allow

viewing radiographs (if radiographic testing was performed).

• Preservice and inservice examination records for the selected welds.

• Readily accessible copies of nondestructive examination personnel

qualifications records for reviewing.

3. For ultrasonic examination procedures qualified in accordance with ASME Code,

Section XI, Appendix VIII, provide documentation supporting the procedure

qualification (e.g. the EPRI performance demonstration qualification summary sheets).

Also, include qualification documentation of the specific equipment to be used (e.g.,

ultrasonic unit, cables, and transducers including serial numbers) and nondestructive

examination personnel qualification records.

Reactor Pressure Vessel Head

1. Nondestructive personnel qualification records for the examiners who will perform

examinations of the reactor pressure vessel head replacement. (If applicable)

2. Drawings showing the following:

• Reactor pressure vessel head and control rod drive mechanism nozzle

configurations

• Reactor pressure vessel head insulation configuration

Note: The drawings listed above should include fabrication drawings for the nozzle

attachment welds as applicable.

Boric Acid Corrosion Control Program

1. Boric acid walk down inspection results, an updated list of boric acid leaks identified

so far this outage, associated corrective action documentation, and overall status of

planned boric acid inspections.

2. List of boric acid evaluation and corrective action documents associated with the

leakage.

Codes and Standards

1. Ready access to (i.e., copies provided to the inspector(s) for use during the

inspection at the onsite inspection location, or room number and location where

available):

• Applicable Editions of the ASME Code (Sections V, IX, and XI) for the

inservice inspection program and the repair/replacement program.

2. Copy of the performance demonstration initiative (PDI) generic procedures with the

latest applicable revisions that support site qualified ultrasonic examinations of piping

welds and components (e.g., PDI-UT-1, PDI-UT-2, PDI-UT-3, PDI-UT-10, etc.).

3. Boric Acid Corrosion Guidebook Revision 1 - EPRI Technical Report 1000975.

The following items are requested for the

Occupational Radiation Safety Inspection

Diablo Canyon

Inspection Dates February 25 - March 1, 2019

Integrated Report 2019001

Inspection areas are listed in the attachments below.

Please provide the requested information on or before February 15, 2019

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 your normal data management system, please ensure the inspection

exit date entered is at least 45 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 Louis Carson at (817)200-1221,

Louis.Carson@nrc.gov or John ODonnell at (817)200-1441, 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: March 1, 2018

A. List of contacts and telephone numbers for the Radiation Protection Organization Staff

and Technicians, as well as the Licensing/Regulatory Affairs staff. Please include area

code and prefix. If work cell numbers are appropriate, then please include them as well.

B. Applicable organization charts including position or job titles. Please include as

appropriate for your site, Site Management, RP, Chemistry, Maintenance (I&C),

Engineering, and Emergency Protection. (Recent pictures are appreciated.)

C. Copies of audits, self-assessments, LARs, and LERs written since the last inspection

date, related to this inspection area

D. Procedure indexes for the radiation protection procedures and other related disciplines.

E. Please provide procedures related to the following areas noted below. Additional

procedures may be requested by number after the inspector reviews the procedure

indexes.

1. Radiation Protection Program

2. Radiation Protection Conduct of Operations, if not included in #1.

3. Personnel Dosimetry

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. Fuel Pool Inventory Access and Control

F. Please provide a list of NRC Regulatory Guides and NUREGs that you are currently

committed to relative to this program. Please include the revision and/or date for the

commitment and where this may be located in your current licensing basis documents.

G. Please provide a summary list of corrective action documents (including corporate and

sub-tiered systems) since the last inspection date.

1. Initiated by the radiation protection organization

2. Assigned to the radiation protection organization

NOTE: These lists should include a description of the condition that provides sufficient

detail that the inspectors can ascertain the regulatory impact, the significance

level assigned to the condition, the status of the action (e.g., open, working,

closed, etc.) and the search criteria used. Please provide in document formats

which are sortable and searchable so that inspectors can quickly and

efficiently determine appropriate sampling and perform word searches, as

needed. (Excel spreadsheets are the preferred format.) If codes are used,

please provide a legend for each column where a code is used.

H. 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.) Please include the radiological risk assigned to each activity.

ML19130A234

SUNSI Review Non-Sensitive Publicly Available Keyword:

By: RDA Sensitive Non-Publicly Available

OFFICE SRI:DRP/A RI:DRP/A C:DRS/EB1 AC:DRS/EB2 C:DRS/OB C:DNMS/RIB

NAME CNewport JReynoso VGaddy JDrake GWerner GWarnick

SIGNATURE /RA/ /RA/ /RA/ /RA/ /RA/ /RA/

DATE 05/02/19 05/02/19 05/06/19 05/02/19 05/05/19 05/07/19

OFFICE C:DRS/IPAT AC:DRS/RCB SPE:DRP/A BC:DRP/A

NAME RKellar NMakris RAlexander MHaire

SIGNATURE /RA/ /RA/ /RA/

DATE 05/06/19 05/07/19 05/02/19