IR 05000382/2019001
| ML19122A121 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 04/30/2019 |
| From: | O'Keefe N NRC/RGN-IV/DRP/RPB-D |
| To: | Dinelli J Entergy Operations |
| References | |
| IR 2019001 | |
| Download: ML19122A121 (52) | |
Text
ril 30, 2019
SUBJECT:
WATERFORD STEAM ELECTRIC STATION, UNIT 3 - NRC INTEGRATED INSPECTION REPORT 05000382/2019001
Dear Mr. Dinelli:
On March 31, 2019, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Waterford Steam Electric Station, Unit 3. On April 11, 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 two findings of very low safety significance (Green) in this report.
Both of these findings involved violations of NRC requirements. The NRC is treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.
If you contest the violations or significance 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 Waterford Steam Electric Station, Unit 3.
If you disagree with a cross-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region IV; and the NRC resident inspector at Waterford Steam Electric Station, Unit 3.
J. Denilli 2 This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with 10 CFR 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely,
/RA/
Neil F. O'Keefe, Chief Reactor Projects Branch D Docket No. 05000382 License No. NPF-38
Enclosure:
Inspection Report 05000382/2019001 w/attachments:
1. Request for Information (Inservice Inspection)
2. Request for Information (O
Inspection Report
Docket Number(s): 05000382 License Number(s): NPF-38 Report Number(s): 05000382/2019001 Enterprise Identifier: I-2019-001-0002 Licensee: Entergy Operations, Inc.
Facility: Waterford Steam Electric Station, Unit 3 Location: 17265 River Road Killona, LA 70057 Inspection Dates: January 1, 2019 to March 31, 2019 Inspectors: B. Baca, Health Physicist N. Greene, PhD, Senior Health Physicist R. Kopriva, Senior Reactor Inspector J. Melfi, Acting Resident Inspector F. Ramirez, Senior Resident Inspector C. Speer, Acting Senior Resident Inspector Approved By: Neil F. O'Keefe, Chief Reactor Projects Branch D Division of Reactor Projects
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting a Quarterly inspection at Waterford Unit 3 in accordance with the Reactor Oversight Process. The Reactor Oversight Process is the NRCs program for overseeing the safe operation of commercial nuclear power reactors. Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information. Findings and violations being considered in the NRCs assessment are summarized in the table below.
List of Findings and Violations Failure to Lock a Reactor Coolant System Drain Valve Results in Loss of Inventory during Lowered Inventory Conditions Cornerstone Significance Cross-cutting Report Aspect Section Initiating Events Green [H.12] - Avoid 71111.13 NCV 05000382/2019001-01 Complacency Closed The inspectors identified Green non-cited violation of Technical Specification 6.8.1.a associated with the licensees failure to implement procedures for equipment control.
Specifically, on January 7, 2019, the licensee failed to lock the boron management/reactor coolant system (RCS) level monitoring system isolation to reactor drain tank valve, BM-1071, when placing it into service as an RCS drain valve contrary to licensee Procedure OP-100-009, Control of Valves and Breakers, Revision 43. As a result, on January 10, 2019, personnel working in the area inadvertently opened BM-1071, resulting in a loss of RCS inventory during lowered inventory operations.
Inadequate Procedure Contributed to Damage to a Fuel Assembly Cornerstone Significance Cross-cutting Report Aspect Section Barrier Integrity Green [P.3] - Resolution 71111.20 NCV 05000382/2019001-02 Closed The inspectors reviewed a self-revealed, Green non-cited violation of 10 CFR Part 50,
Appendix B, Criterion V, Instructions, Procedures, and Drawings, which occurred because procedures for handling spent fuel in the spent fuel pool (SFP) were not appropriate to the circumstances. Specifically, licensees procedure for placing fuel assemblies into cells in the SFP, RF-005-002, Refueling Equipment Operation, did not adequately verify that spent fuel would not be damaged during fuel movements. As a result, on January 21, 2019, when placing spent fuel assembly LAGA06 into the SFP cell LL30, it made contact with a deformed portion of the SFP cell and damaged the fuel assembly such that it was no longer acceptable for use in the core.
Additional Tracking Items Type Issue number Title Report Section Status LER 05000382/2019-003-00 Relevant Indications 71153 Closed Identified in Reactor Coolant Pump 1A and 2A Suction Drain Nozzle Dissimilar Metal Welds Resulting in the Condition of the Nuclear Power Plant, Including its Principal Safety Barriers,
Being Seriously Degraded
PLANT STATUS
The Waterford Steam Electric Station, Unit 3, began the inspection period at 100 percent power.
The plant was shut down on January 5, 2019, to begin Refueling Outage 22. On March 16, 2019, operators restarted the reactor startup, and the unit achieved full power on March 22, 2019. The unit remained at full power for the remainder of the inspection period.
INSPECTION SCOPES
Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors performed plant status activities described in IMC 2515 Appendix D, Plant Status and conducted routine reviews using IP 71152, Problem Identification and Resolution. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.
REACTOR SAFETY
71111.04 - Equipment Alignment Partial Walkdown (IP Section 02.01)
The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:
- (1) Shutdown cooling train A with train B out of service for maintenance on January 18, 2019
- (2) Spent fuel pool cooling train A during a period of heightened risk due to core offload on January 24, 2019
- (3) Emergency diesel generator train B with train A out of service for maintenance on February 7, 2019
- (4) Low pressure safety injection train B following system re-alignment on March 18, 2019
71111.05Q - Fire Protection Quarterly Inspection (IP Section 03.01)
The inspectors evaluated fire protection program implementation in the following selected areas:
- (1) Fire area RCB, reactor containment building general area, on January 9, 2019
- (2) Fire area RAB 2, heating and ventilation mechanical room, on January 14, 2019
- (3) Fire area RAB 39, general area, on January 14, 2019
- (4) Fire area NS-TB-001, +15 turbine generator building switchgear/east area, on February 27, 2019
- (5) Fire area NS-TB-002, +15 turbine generator building west area, on February 27, 2019
71111.08P - Inservice Inspection Activities (PWR) PWR Inservice Inspection Activities Sample (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 January 3, 2019, to March 22, 2019:
03.01.a - Nondestructive Examination and Welding Activities.
The Inspector directly observed or reviewed records of the following non-destructive examination activities:
- (1) Magnetic Particle Examinations a) Reactor Coolant System. Support Skirt to Pressurizer Weld.
Report No. ISI-MT-19-001.
- (2) Ultrasonic Examinations a) Reactor Pressurizer. Pressurizer Safety Nozzle to Top Head Weld at 450.
Report No. ISI-UT-19-006.
b) Reactor Pressurizer. Pressurizer Safety Nozzle to Top Head Weld at 1350.
Report No. ISI-UT-19-007.
c) Reactor Coolant System. Reactor coolant system Loop 1A Lower Cold Leg, nozzle to safe-end circumferential weld. Fully Encoded Phased Array Ultrasonic Examination of Dissimilar Metal Welds. One, unacceptable, axial planar flaw indication was identified. Report No. ISI-VE-19-009. Summary No. W3.89.21.001.
d) Reactor Coolant System. Reactor coolant system Loop 2A Lower Cold Leg, nozzle to safe-end circumferential weld. Fully Encoded Phased Array Ultrasonic Examination of Dissimilar Metal Welds. One, unacceptable, axial planar flaw indication was identified. Report No. ISI-VE-19-013. Summary No. W3.89.21.006.
e) Reactor Coolant System. Reactor coolant system Loop 1A Lower Cold Leg, nozzle to safe end circumferential weld. Component ID: 07-009-WOL.
Report No. BOP-VE-19-002.
f) Reactor Coolant System. Reactor coolant system Loop 2A Lower Cold Leg, nozzle to safe end circumferential weld. Component ID: 11-007-WOL.
Report No. BOP-VE-19-003.
g) Reactor Coolant System. Reactor coolant system Loop 1A Lower Cold Leg, safe end to 2 inch pipe weld. Component ID: 33-001-WOL.
Report No. BOP-VE-19-004.
h) Reactor Coolant System. Reactor coolant system Loop 2A Lower Cold Leg, safe end to 2 inch pipe weld. Component ID: 35-001-WOL.
Report No. BOP-VE-19-005.
- (3) Visual Examinations a) Moisture Barrier Inspection Inside Containment. Inner Moisture Barrier between Columns 1 and Columns 3. Report No. ISI-VT-19-003.
b) Moisture Barrier Inspection Inside Containment. Inner Moisture Barrier between Columns 3 and Columns 5. Report No. ISI-VT-19-004.
c) Moisture Barrier Inspection Inside Containment. Inner Moisture Barrier between Columns 5 and Columns 7. Report No. ISI-VT-19-005.
d) Moisture Barrier Inspection Inside Containment. Inner Moisture Barrier between Columns 7 and Columns 9. Report No. ISI-VT-19-006.
e) Moisture Barrier Inspection Inside Containment. Inner Moisture Barrier between Columns 11 and Columns 13. Report No. ISI-VT-19-008.
f) Moisture Barrier Inspection Inside Containment. Inner Moisture Barrier between Columns 13 and Columns 15. Report No. ISI-VT-19-009.
g) Moisture Barrier Inspection Inside Containment. Inner Moisture Barrier between Columns 15 and Columns 17. Report No. ISI-VT-19-010.
h) Moisture Barrier Inspection Inside Containment. Inner Moisture Barrier between Columns 17 and Columns 19. Report No. ISI-VT-19-011.
i) Moisture Barrier Inspection Inside Containment. Inner Moisture Barrier between Columns 19 and Columns 21. Report No. ISI-VT-19-012.
j) Moisture Barrier Inspection Inside Containment. Inner Moisture Barrier between Columns 23 and Fuel Transfer Tunnel Shield Boundary. Report No. ISI-VT-19-014.
k) Moisture Barrier Inspection Outside of Containment in Annulus. Moisture Barrier Inside Annulus 1030 to 2560 azimuth. Report No. ISI-VT-19-016.
l) Moisture Barrier Inspection Outside of Containment in Annulus. Moisture Barrier Inside Annulus 2070 to 3600 azimuth. Report No. ISI-VT-19-017.
m) Reactor Vessel Head Bolts. Bolted Connection Reactor Coolant MRCT0001 (Reactor Vessel Studs). Report No. ISI-VT-19-018.
n) Main Steam. Moisture Separator Reheater Component ID: MSRR-0264A. Rigid Restraint Attachment Weld. Report No. ISI-VT-19-024.
.
o) Main Steam. Moisture Separator Reheater Component ID: NSRR-0264. RAB Roof, Rigid Restraint. Report No. ISI-VT-19-025.
p) Reactor Vessel Closure Head. Control Element Drive Mechanism Nozzles.
Report No. BOP-VT-19-005.
The inspector directly observed or reviewed records of the following welding activities:
- (1) Gas Tungsten Arc Weld - Manual a) Chemical Volume and Control System Valve Replacement. Component No. CVCMVAAA205 Valve.
- (2) Gas Tungsten Arc Weld - Machine a) Reactor Coolant System. Weld Overlay. Reactor coolant system Loop 1A Lower Cold Leg, nozzle to safe-end circumferential weld. Component ID: 07-009-WOL.
b) Reactor Coolant System. Weld Overlay. Reactor coolant system Loop 2A Lower Cold Leg, nozzle to safe-end circumferential weld. Component ID: 11-007-WOL.
03.01.b - Pressurized-Water Reactor Vessel Upper Head Penetration Examination Activities The bare metal visual inspection of the reactor vessel upper head penetrations was not required to be performed during this outage as it had just been completed in Refueling Outage RF21. However, after performing a visual inspection on the reactor head for the initial boric acid walkdown and the required visual examination, VT-2, of the reactor head bolts, the inspector and licensee noted significant amounts of a dry, white, substance covering a significant area of the reactor head, flange, bolts, and control element drive mechanisms and nozzles. As a follow-up assessment of the condition identified in CR-WF3-2017-02567, a visual examination was performed on selected reactor head penetrations. This was based on a white substance on the reactor head attributed to a component cooling water leakage in RF21. This examination was performed in accordance with Entergy Procedure CEP-NDE-0955. General Electric Inspection Technologies performed a bare metal visual exam per Work Order No. 516245-04 and the examination results were documented in Report No. ISR-190112JS248. The examination coverage area included 360 degrees around the annulus of each penetration examined.
Chemical samples were taken to aid in identifying the source of leakage. Site chemical analysis results concluded the source of leakage was component cooling water, and the white residue was molybdates in the component cooling water system. Additional samples have been sent out for independent analysis. No evidence of nozzle leakage or active boron was identified during the examination. There was no discoloration, corrosion, wastage, or staining that would represent active nozzle leakage. A light residue remains on the head, in the area of head vent piping due to access restrictions.
The inspection results were acceptable.
03.01.c - Pressurized-Water Reactor Boric Acid Corrosion Control Activities 17-WF3-0119 Condition Report # 17-4996 June 8, 2017 17-WF3-0120 Condition Report # 17-5233 June 19, 2017 17-WF3-0121 Condition Report # 17-5904 July 18, 2017 17-WF3-0122 Condition Report # 17-5906 July 18, 2017 17-WF3-0123 Condition Report # 17-6604 August 7, 2017 17-WF3-0124 Condition Report # 17-6643 August 8, 2017 17-WF3-0125 Condition Report # 17-6642 August 8, 2017 17-WF3-0126 Condition Report # 17-7115 August 28, 2017 17-WF3-0127 Condition Report # 17-07504 September 14, 2017 17-WF3-0128 Condition Report # 17-10007 December 12, 2017 18-WF3-0001 Condition Report # 18-0284 January 15, 2018 18-WF3-0002 Condition Report # 18-0376 January 20, 2018 18-WF3-0003 Condition Report # 18-0698 February 3, 2018 18-WF3-0004 Condition Report # 18-1219 March 1, 2018 18-WF3-0005 Condition Report # 18-1830 March 31, 2018 18-WF3-0006 Condition Report # 18-1490 March 14, 2018 18-WF3-0007 Condition Report # 18-1829 March 31, 2018 18-WF3-0008 Condition Report # 18-3372 June 14, 2018 18-WF3-0009 Condition Report # 18-3372 June 14, 2018 18-WF3-0010 Condition Report # 18-4272 July 31, 2018 18-WF3-0011 Condition Report # 18-4552 August 10, 2018 18-WF3-0012 Condition Report # 18-6352 November 11, 2018 18-WF3-0013 Condition Report # 18-6408 December 12, 2018 03.01.d - Steam Generator Tube Examination Activities 1. The steam generator tube inspections were not required to be performed during this outage.
Identification and Resolution of Problems The inspectors reviewed 60 condition reports which dealt with inservice inspection activities and found the corrective actions for inservice inspection issues were appropriate.
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 plant shutdown and cooldown on January 5, 2019.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
The inspectors observed and evaluated licensed operator performance in the simulator during just in time startup training on March 8 and 10, 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) Containment building on February 20, 2019
- (2) Auxiliary component cooling water system on March 5, 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) Elevated risk due to reactor coolant system draining down to the reactor vessel flange for reactor head removal on January 9, 2019
- (2) Emergent work review due to unexpected lowering of reactor coolant system level during reduced inventory conditions on January 10, 2019
- (3) Elevated risk associated with lifting the reactor vessel head on January 11, 2019
- (4) Emergent work review due to discovery of a weld flaw associated with reactor coolant system loop 2A on January 22, 2019
- (5) Emergent work review due to the failure of the plant main computer on February 13, 2019
- (6) Emergent work review due to inadvertent lifting of a fuel assembly on February 7, 2019
- (7) Elevated risk due to reactor coolant draindown to the reactor vessel flange for reactor head installation on February 14, 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) Operability determination associated with a weld flaw in reactor coolant system Loop 1A on January 18, 2019
- (2) Operability determination associated with startup transformer B unexpected test indications on January 28, 2019
- (3) Operability determination associated with failed containment coatings on January 29, 2019
- (4) Functionality assessment associated with excessive wear on fuel assembly upper end fitting guide posts and coil hold-down springs on February 12, 2019
- (5) Operability determination associated with cracks found in the shield building on February 26, 2019
- (6) Operability determination associated with slow open stroke times of the steam generator No. 2 atmospheric dump valve, on March 8, 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) Temporary modification to connect loads equivalent to safety-related loads for testing of emergency diesel generator A on January 23, 2019
- (2) Permanent modification associated with ultimate heat sink motor control centers on February 14, 2019
- (3) Permanent modification associated with letdown inside containment isolation valve, CVC-103, on March 28, 2019
71111.19 - Post Maintenance Testing Post Maintenance Test Sample (IP Section 03.01)
The inspectors evaluated the following post maintenance tests:
- (1) Component cooling water makeup pump B discharge check valve, CMU-510B, following disassembly for inspection on February 5, 2019
- (2) Containment sump pump inside containment isolation valve, SP-105, after limit switch replacement on February 6, 2019
- (3) Auxiliary feedwater pump flow control valves, EFW-223B and 229B, following system maintenance on February 7, 2019
- (4) Auxiliary component cooling water pump A following system maintenance on February 9, 2019
- (5) Low pressure safety injection pump A following system maintenance on February 12, 2019
- (6) Main steam isolation valve 2, MS-124B, following system maintenance on March 11, 2019
71111.20 - Refueling and Other Outage Activities Refueling/Other Outage Sample (IP Section 03.01)
The inspectors evaluated the licensee's performance during Refueling Outage 22 from January 5, 2019, to March 18, 2019.
71111.22 - Surveillance Testing The inspectors evaluated the following surveillance tests: Containment Isolation Valve (ISO) (IP Section 03.01)
Stroke time testing of Reactor Coolant Loop 2 shutdown cooling inside containment isolation valve, SI-405A, on January 5, 2019
In Service Testing (IST) (IP Section 03.01) (2 Samples)
- (1) Stroke time testing of component cooling water header B return from essential chiller isolation valve, CC-322B, on January 16, 2019
- (2) Stroke test of emergency feedwater header B check valve, EFW-2191B, on February 27, 2019
Routine Surveillance Testing (IP Section 03.01) (2 Samples)
- (1) Integrated test of emergency safety features and emergency diesel generator B on January 26, 2019
- (2) Containment integrated leak rate test on February 20,
RADIATION SAFETY
71124.01 - Radiological Hazard Assessment and Exposure Controls Contamination and Radioactive Material Control (IP Section 02.03)
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:
- The inspectors performed walk downs of the licensee's sealed sources, per the inventory provided, and had no regulatory concerns with the storage, labeling, or physical condition of the observed sources.
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.
Instructions to Workers (IP Section 02.02) (1 Sample)
The inspectors evaluated instructions to workers including radiation work permits used to access high radiation areas:
Radiation work packages:
- RWP 20190702, Refuel 22: Disassembly of Reactor Head and All Associated Work Activities, Revision 0
- RWP 20190705, Refuel 22: Reassembly of Reactor Head and Associated Work Activities including Staging/Destaging of Equipment, Revision 0
- RWP 20190708, Refuel 22: ICI Removal/Installation/Cut Up of ICIs, Work on ICI Equipment and Replacement Swageloc Bodies, Revision 0
- RWP 20190724, Refuel 22: Waterford 3 Fuel Transfer System Maintenance and Repairs including Lower Cavity Decon Activities, Support Activities, HP Job Coverage and Surveys, Revision 0 Electronic alarming dosimeter alarms:
- No alarms of significance occurred during this period.
Labeling of containers:
- All containers observed had proper labeling.
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.
Radiological Hazard Assessment (IP Section 02.01) (1 Sample)
The inspectors evaluated radiological hazards assessments and controls. The inspectors reviewed the following:
Radiological surveys:
- WF3-1901-0037, -35 RAB Hallway, January 2, 2019
- WF3-1901-0120, Radwaste Solidification Building, January 4, 2019
- WF3-1901-0121, FHB +46 Fuel Handling Area, January 4, 2019
- WF3-1901-0361, -11 Reactor Containment Building, January 8, 2019
- WF3-1901-0368, +46 Reactor Containment Building, January 8, 2019
- WF3-1901-0474, -4 Reactor Containment Building, January 9, 2019
- WF3-1901-0765, RAB +21 Drumming Station, January 14, 2019 Risk significant radiological work activities:
- RWP 20190702, Refuel 22: Disassembly of Reactor Head and All Associated Work Activities, Revision 0
- RWP 20190705, Refuel 22: Reassembly of Reactor Head and Associated Work Activities including Staging/Destaging of Equipment, Revision 0
- RWP 20190708, Refuel 22: ICI Removal/Installation/Cut Up of ICIs, Work on ICI Equipment and Replacement Swageloc Bodies, Revision 0 Air sample survey records:
- OL-050218-015, Air Sample: Fuel Pool Purification Pump Check Valve (FS-426)
Inspection, May 2, 2018
- OL-110118-061, Air Sample: Welding Purification Ion Exchanger B Plug, November 1, 2018
- OL-121818-076, Air Sample: Chemistry Sampling of SDC B, December 18, 2018
- OL-010119-001, Air Sample: Safeguards 'A' Pump Room, January 1, 2019 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 package for areas with airborne radioactivity were reviewed via the following air sample surveys:
- OL-050218-015, Fuel Pool Purification Pump Check Valve, FS-426, Inspection, May 2, 2018
- OL-110118-061, Welding Purification Ion Exchanger B Plug, November 1, 2018
- OL-121818-076, Chemistry Sampling of shutdown cooling train B, December 18, 2018
- OL-010119-001, Safeguards 'A' Pump Room, January 1, 2019
71124.02 - Occupational ALARA Planning and Controls Implementation of ALARA and Radiological Work Controls (IP Section 02.03)
The inspectors reviewed ALARA practices and radiological work controls by reviewing the following activities:
- RWP 20190613, Refuel 22: Alloy 600 Inspections in Containment and Support Activities including Decontamination and Preparation Support, Revision 0
- RWP 20190617, Refuel 22: Various Decontamination Activities in the Upper Reactor Cavity, Staging/Destaging including Change out of Tri-Nuc Filters and Underwater Vacuuming, Revision 0
- RWP 20190702, Refuel 22: Disassembly of Reactor Head and All Associated Work Activities, Revision 0
- RWP 20190708, Refuel 22: In Core Instrumentation (ICI) Removal/Installation/Cut up of ICIs, Work on ICI Equipment and Replacement Swageloc Bodies, Revision 0
Radiation Worker Performance (IP Section 02.04) (1 Sample)
The inspectors evaluated radiation worker and radiation protection technician performance during the reactor head decontamination and disassembly, movement and cutting of in-core instrumentation, movement of the reactor upper guide structure, and Alloy 600 component inspection and testing.
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) ===
January 1, 2018, through December 31, 2018 IE03: Unplanned Power Changes per 7000 Critical Hours Sample (IP Section 02.02)
(1 Sample)
January 1, 2018, through December 31, 2018 IE04: Unplanned Scrams with Complications (USwC) Sample (IP Section 02.03) (1 Sample)
January 1, 2018, through December 31, 2018 OR01: Occupational Exposure Control Effectiveness Sample (IP Section 02.15) (1 Sample)
January 1, 2018, through December 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)
January 1, 2018, through December 31, 2018
71152 - Problem Identification and Resolution Semiannual Trend Review (IP Section 02.02)
The inspectors reviewed the licensees corrective action program for potential adverse trends that might be indicative of a more significant safety issue. The inspectors identified an observation related to a programmatic weakness in performing Maintenance Rule functional failure evaluations that is documented in the Inspections Results section below.
71153 - Follow-up of Events and Notices of Enforcement Discretion Event Report (IP Section 03.02)
The inspectors evaluated the following licensee event reports (LERs):
- LER 05000382/2019-003-00, Relevant Indications Identified in Reactor Coolant Pump 1A and 2A Suction Drain Nozzle Dissimilar Metal Welds Resulting in the Condition of the Nuclear Power Plant, Including its Principal Safety Barriers, Being Seriously Degraded The inspectors concluded that no violation of NRC requirements occurred.
INSPECTION RESULTS
Failure to Lock an RCS Drain Valve Results in Loss of Inventory during Lowered Inventory Conditions Cornerstone Significance Cross-cutting Report Aspect Section Initiating Events Green [H.12] - Avoid 71111.13 NCV 05000382/2019001-01 Complacency Closed The inspectors identified Green non-cited violation of Technical Specification 6.8.1.a associated with the licensees failure to implement procedures for equipment control.
Specifically, on January 7, 2019, the licensee failed to lock the boron management/reactor coolant system (RCS) level monitoring system isolation to reactor drain tank valve, BM-1071, when placing it into service as an RCS drain valve contrary to licensee Procedure OP-100-009, Control of Valves and Breakers, Revision 43. As a result, on January 10, 2019, personnel working in the area inadvertently opened BM-1071, resulting in a loss of RCS inventory during lowered inventory operations.
Description:
On January 7, 2019, the licensee placed the reactor coolant system level monitoring system (RCSLMS) into service to support work during refueling outage 22. The RCSLMS provides redundant RCS level indication to operations personnel during outages when the RCS is not full. To place the RCSLMS into service, the normally-locked RCS hot leg 1 drain and drain isolation valves, RC-106 and RC-107, were opened and the boron management/RCSLMS isolation to reactor drain tank valve, BM-1071, was closed. Although BM-1071 was locked closed to prevent inadvertent mispositioning during the installation of the RCSLMS equipment, the locking mechanism was removed when the RCSLMS was placed into service.
On January 10, 2019, the licensee partially drained the RCS below the reactor vessel flange to remove the reactor vessel head. Control room operators established an RCS level control band of 18.88 to 19.5. At 11:19 a.m., control room operators noticed RCS level dropping and a corresponding rise in reactor drain tank level. At 11:24 a.m., operators started charging pump A to restore RCS level. Indicated RCS level dropped from approximately 19 MSL to as low as 18.5, falling below the pre-prescribed band for RCS level control.
At 11:45 a.m., the licensee evacuated unnecessary personnel from the reactor containment building. At 11:51 a.m., the licensee entered OP-901-111, Reactor Coolant System Leak, and OP-901-131, Shutdown Cooling System Malfunction, to address the loss of inventory.
At 12:16 p.m., the licensee began additional injection with high pressure safety injection pump A to restore RCS level more quickly. At 12:51 p.m., personnel dispatched to the reactor containment building to investigate the loss of level found BM-1071 open, which allowed RCS inventory to drain to the reactor drain tank. Operators closed BM-1071 to stop the loss of RCS inventory.
The licensee determined that a scaffolding crew working in the vicinity of BM-1071 partially opened the valve inadvertently, which led to the loss of RCS inventory. The licensee further determined that the pre-job brief and job site review for the scaffolding work did not identify that the risk-significant valve was in the area and was subject to potential inadvertent operation since it was located near the path to the job site, not at the job site. Additionally, the licensee found that licensee Procedure EN-OP-115-05, Control of Components, was revised in December 2018 to require measures be used to prevent inadvertent operation of drain valves like BM-1071 in high traffic areas, but the revision to this corporate procedure was inadvertently not issued.
In their review of the event, the inspectors identified that licensee Procedure OP-100-009, Control of Valves and Breakers, Revision 43, gave the licensee guidance for locking valves important to plant operations. Step 5.7.1.7 of OP-100-009 specifically lists Reactor Coolant System vent and drain valves as valves that should be locked. Procedure OP-100-009 was available to licensee personnel in their electronic system for accessing procedures, but it was not implemented to consider locking BM-1071 or similar valves in drain paths for the RCS to prevent inadvertent operation when the RCSLMS was placed into service.
In discussing the issue with licensee personnel, the inspectors found that the licensee only applied the OP-100-009 guidance to lock RCS vent and drain valves in the normal alignment.
The licensee did not consider locking valves like BM-1071 that provide potential RCS vent and drain paths during temporary or off-normal alignments. However, the guidance contained in OP-100-009 is not limited to normal system alignments.
Corrective Action(s): The licensees immediate corrective action was to place locking devices on or otherwise secure BM-1071 and other valves that provided an RCS drain path while the RCSLMS was aligned. The licensees planned corrective actions include identifying valves in the Reactor Containment Building that warrant additional controls such as locking devices to prevent inadvertent operation.
Corrective Action Reference(s): CR-WF3-2019-00513, CR-HQN-2019-00230, CR-WF3-2019-03986
Performance Assessment:
Performance Deficiency: The failure to follow procedure OP-100-009 and to lock drain Valve BM-1071 when placing RCSLMS into service was a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Configuration Control attribute of the Initiating Events cornerstone. Specifically, the licensee did not lock BM-1071 to prevent inadvertent operation after aligning it as an RCS drain valve when placing RCSLMS into service. As a result, the valve was not protected from inadvertent opening when a scaffolding crew was working in the area. When the valve was inadvertently opened, it resulted in a loss of RCS inventory when the system was already in a reduced inventory condition.
Significance: The inspectors assessed the significance of the finding using Appendix G, Shutdown Safety SDP. Using Appendix G, Attachment 1, Exhibit 2, Initiating Events Screening Questions, the inspectors determined that the finding required a Phase 2 estimation because it resulted in leakage such that, if the leakage were undetected and/or unmitigated in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or less, it would cause the operating decay heat removal method to fail (e.g., level would drop below the hot leg suction of the low pressure safety injection pumps used for shutdown cooling).
Utilizing Inspection Manual Chapter 0609, Appendix G, Attachment 2, Phase 2 Significance Determination Process Template for PWR During Shutdown, the senior reactor analyst conducted a Phase 2 approximation of the risk significance of the subject finding. The analyst made the following conclusions:
1. The finding was a precursor finding because it increased the likelihood that the operating RHR train would be lost; 2. The initiating event that best models the impact of the subject finding was a Loss of Reactor Inventory; 3. The plant operating state during the time that the finding impacted the plant was Plant Operating State 2, because water level was at reduced inventory (below the flange)and the pressurizer manway was removed; and 4. The outage was in the early time window, indicating that Plant Operating State 3 had not yet been entered for the outage.
The analyst performed the following steps from Appendix G, Attachment 2:
Step 4.3.1: The analyst determined that the finding only impacted plant risk in Plant Operating State 2 and the Early Time Window.
Step 4.3.2: The analyst utilized Table 3, Initiating Event Likelihood (IELs) for Loss of Inventory Precursors, to determine the estimated IEL for a loss of reactor inventory initiator. The applicable questions were answered as follows:
Time to RHR loss: Using the information provided, the analyst calculated that the drain rate during the event was approximately 1.6 feet per hour. This would have required approximately 9.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to drain to midloop. Therefore, the time to loss of RHR pump suction was greater than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
Was RCS level indication a reasonable reflection of RCS level? The analyst noted that while level indication was inaccurate during the drain, it continued to indicate a decreasing reactor water level that was greater than the actual loss of inventory. Therefore, in consultation with the resident inspector, the analyst answered this question, YES.
Can leak path be readily identified within 1/2 time to loss of RHR? The analyst noted that under the actual conditions it took operators 92 minutes to identify and isolate the source of the leak. Therefore, the analyst answered this question, YES.
Can drain path be isolated by at least one functional valve such that a train of RHR can be restarted? Because operators were able to isolate the drain path and maintain the RHR system running during the actual drain, the analyst answered this question, YES.
Based on these four answers, Table 3 estimated the IEL as being in the magnitude of 10-4, or 4.
Step 4.3.3: In accordance with this step, the analyst used Worksheet 6, SDP for a PWR Plant - Loss of Inventory in Plant Operational State 2 (RCS Vented) to perform the risk estimation.
The analyst determined the function credit for each top event in Worksheet 6.
Mitigation Credit for Loss of Reactor Inventory Initiator Top Event Basis for Success Credit Credit Injection There were multiple, diverse trains of injection available 4 before Core during the subject draining event. Also, level indication Damage and core exit thermocouple output were available. The time to core damage was greater than 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> without injection. The analyst determined that this function was limited by operator error.
Terminate At least one valve was available to isolate the leak and 3 Leak Path this path would not prevent operation of decay heat before RWST removal. This provided an equipment credit of 3. The Depletion time to refueling water storage tank (RWST) depletion was significantly longer than 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. Therefore, operator credit was 3.
Decay Heat All equipment necessary to vent an RHR pump and 3 Removal restart it before RWST depletion was available, Recovery including inlet and outlet temperature indication and RHR flow indication. The time to RWST depletion was greater than 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. The analyst determined that this function was limited by operator error.
Borated Water There was sufficient equipment available for operators 2 Makeup to to makeup to the RWST before RWST depletion and RWST core damage, including RWST level indication and a low level alarm. The time to RWST depletion and subsequent core damage was greater than 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />.
The analyst determined that this function was limited by operator error.
Using the initiating event likelihood and the mitigation credit documented above, the analyst quantified the results of the core damage sequences recorded on Worksheet 6. The following results were documented:
Worksheet 6 Results Core Damage Sequence IEL Mitigation Credit Result Loss of Reactor Inventory with Failure 4 3+2 9 to Recover RHR and Failure to Makeup to the RWST before Core Damage Loss of Reactor Inventory with Failure 4 3+2 9 to Stop the Leak Before RWST Depletion and Failure to Makeup to the RSWT before Core Damage Loss of Reactor Inventory with a Failure 4 4 8 of RCS Injection Before Core Damage Based on these results, the analyst determined that the change in core damage frequency associated with the finding was less than 1 x 10-7. Therefore, the finding is of very low safety significance (Green).
Cross-cutting Aspect: H.12 - Avoid Complacency: Individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Individuals implement appropriate error reduction tools. The finding had a cross-cutting aspect in the area of human performance associated with avoiding complacency because individuals did not recognize and plan for the possibility of mistakes, latent issues, and inherent risk even while expecting successful outcomes. Specifically, the licensee did not recognize the inherent risk of potentially mispositioning BM-1071 or other valves that could drain the RCS when changing from the normal RCS lineup and only applied guidance for locking valves to normal rather than off-normal equipment alignments.
Enforcement:
Violation: Technical Specification 6.8, Procedures and Programs, Section 1.a, requires, in part, that procedures shall be established, implemented and maintained covering, the applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2.
Regulatory Guide 1.33, Quality Assurance Program Requirements, Revision 2, Appendix A, Section 1.c, requires that instructions be established for Equipment control (e.g., locking and tagging). The licensee established licensee procedure OP-100-009, Control of Valves and Breakers, Revision 43, to meet the Regulatory Guide 1.33 requirement. Step 5.7.1.7 of licensee Procedure OP-100-009 directs that Reactor Coolant System vent and drain valves should be locked.
Contrary to the above, on January 7, 2019, after installing temporary RCS level monitoring equipment as part of Refueling Outage 22, the licensee did not lock reactor coolant system drain valve BM-1071. As a result, on January 10, 2019, personnel working in the area inadvertently opened BM-1071, leading to an unexpected loss of RCS inventory during lowered inventory operations.
Enforcement Action: This violation is being treated as a Non-Cited Violation, consistent with Section 2.3.2 of the Enforcement Policy.
Inadequate Procedure Contributed to Damage to a Fuel Assembly Cornerstone Significance Cross-cutting Report Aspect Section Barrier Integrity Green [P.3] - 71111.20 NCV 05000382/2019001-02 Resolution Closed The inspectors reviewed a self-revealed, Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, which occurred because the licensee did not prescribe procedures for handling spent fuel in the spent fuel pool (SFP)that were appropriate to the circumstances. Specifically, licensees procedure for placing fuel assemblies into cells in the SFP, RF-005-002, Refueling Equipment Operation, was not appropriate to the circumstances because it did not ensure that spent fuel would not be damaged during fuel movements. As a result, on January 21, 2019, when placing spent fuel Assembly LAGA06 into the SFP cell LL30, it made contact with a deformed portion of the SFP cell and damaged the fuel assembly such that it was no longer acceptable for use in the core.
Description:
On January 21, 2019, during Refueling Outage 22, the licensee attempted to place spent fuel assembly LAGA06 into SFP cell LL30. When lowering the fuel assembly into the cell, the SFP handling machine hoist stopped and an underload condition occurred. An overload condition occurred when the licensee attempted to raise the fuel assembly. After initial troubleshooting, the licensee moved the assembly laterally and was able to raise it smoothly without issue. The licensee then placed the assembly in an alternate SFP cell.
Subsequent inspection of LAGA06 identified damage on one face of the assembly. The licensee found damage on the bottom grid strap, on two fuel pins, and on the top grid.
Although the assembly was intended to be reloaded into the core, the licensee determined LAGA06 was not suitable for continued operation. The licensee reconstituted the fuel pins into a new fuel assembly to address the damage prior to reloading the bundle in the core.
The licensee inspected SFP cell LL30 found damage to the top edge of one wall of the cell, which was bent inwards. The licensee determined that the deformation was the cause of the damage to LAGA06 when it was lowered into the fuel cell.
The licensees fuel movements in the SFP are governed by Procedure RF-005-002, Refueling Equipment Operation, Revision 341. Step 5.3.2.3.16 requires the licensee to verify location is acceptable for insertion or removal of fuel assembly (i.e., little or no damage to the fuel rack) when placing a fuel assembly into a SFP cell. The licensees fuel handling personnel indicated that the procedural step is achieved with visual examinations from the SFP handling machine without the use of cameras or other tools. Procedure RF-005-002 does not provide any specific criteria, method, or other means for performing this verification.
The licensee previously identified multiple instances regarding fuel assemblies interacting with portions of the SFP cells:
- CR-WF3-2015-01763 documented damage found to cell NN31 discovered while placing a fuel assembly
- CR-WF3-2014-02346 documented damage found to cell EE30 discovered while placing a fuel assembly
- CR-WF3-2014-02206 documented damage found to cell RR36 discovered after placing a fuel assembly
- CR-WF3-2014-00610 documented the inability to remove a fuel assembly from cell XX37 due to rack damage The inspectors reviewed the corrective actions from the previous instances and determined that in each case the licensee had not taken action to assure that the procedural requirements of RF-005-002 associated with moving fuel assemblies in the SFP were adequate prevent damage to the fuel assemblies.
Corrective Action(s): The licensees immediate corrective actions were to procedurally prohibit SFP cell LL30 for use and to reconstitute fuel bundle LAGA06. The licensees planned future corrective actions include revising RF-005-002 to require the use of a camera during movements in the SFP to aid in positioning of fuel assemblies prior to placing them into SFP cells to more easily identify conditions that could damage fuel assemblies. The licensee plans to also implement a task to perform a detailed inspection of SFP cells prior to each outage to identify pre-existing rack damage or other conditions that could damage fuel assemblies.
Corrective Action Reference(s): CR-WF3-2019-01244
Performance Assessment:
Performance Deficiency: The failure to prescribe procedures for placing spent fuel into SFP cells that were appropriate to the circumstances was a performance deficiency. Specifically, the procedure did not include adequate instructions to ensure that spent fuel rack would not damage the fuel prior to placing a bundle in the rack.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Procedure Quality attribute of the Barrier Integrity cornerstone. Specifically, the failure to prescribe procedures appropriate to the circumstances for handling spent fuel in the SFP resulted in damage to a fuel assembly.
Significance: The inspectors assessed the significance of the finding using Appendix A, Significance Determination of Reactor Inspection Findings for At - Power Situations. Using Exhibit 3, Barrier Integrity Screening Questions, Section D, Spent Fuel Pool, the inspectors determined the finding to be of very low safety significance (Green), because the performance deficiency did not affect fuel pool temperature or level, did not affect neutron absorber capability or result in a fuel assembly being misplaced, and did not cause mechanical damage to fuel clad with a detectible release of radionuclides.
The inspectors were initially directed by IMC 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, to IMC 0609 Appendix G, Shutdown Operations Significance Determination Process, to assess the significant of the finding. However, the inspectors did not find any applicable criteria to evaluate this issue in Appendix G. As allowed by IMC 0609, Appendix M, Significance Determination Process Using Qualitative Criteria, the inspectors used alternate significance determination tools to assess the finding.
Using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power.
Cross-cutting Aspect: P.3 - Resolution: The organization takes effective corrective actions to address issues in a timely manner commensurate with their safety significance. The finding had a cross-cutting aspect in the area of Problem Identification and Resolution associated with resolution because the licensee did not take effective corrective actions to address issues in a timely manner commensurate with their safety significance. Specifically, the licensee has documented the potential for damaged fuel pool rack locations to interact with fuel assemblies multiple times in the past but did not ensure that the corrective actions taken were effective to prevent subsequent interactions between fuel assemblies and the spent fuel pool rack that could lead to fuel damage.
Enforcement:
Violation: Title 10 of the Code of Federal Regulations, Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances.
Contrary to the above, as of January 21, 2019, the licensees procedure for placing spent fuel assemblies into spent fuel pool rack locations, an activity affecting quality, was not appropriate to the circumstances. Specifically, licensee Procedure RF-005-002, Refueling Equipment Operation, Revision 341, did not contain adequate instructions to ensure that fuel rack locations were acceptable for insertion of spent fuel assemblies. Step 5.3.2.3.16 required refueling personnel to verify that spent fuel rack locations were acceptable for insertion or removal of assembly. (i.e. little or no damage to the fuel rack), but did not prescribe any means or criteria for making that determination. As a result, on January 21, 2019, when placing a spent fuel assembly LAGA06 into the spent fuel pool rack location LL30, the fuel assembly made contact with a deformed portion of the spent fuel pool rack and damaged the fuel assembly such that it was not in an acceptable condition for use in the core.
Enforcement Action: This violation is being treated as a Non-Cited Violation, consistent with Section 2.3.2 of the Enforcement Policy.
Observation 71152 The inspectors identified an adverse trend involving Maintenance Rule function failure evaluations. As part of the Maintenance Rule program, the licensee established functional failure criteria for the containment function that included any failure of a containment isolation valve to meet its operational leakage limit or any failure that could prevent a containment isolation valve from closing. The inspectors identified three instances where containment isolation valves failed local leak rate testing, but those failures were not evaluated against the applicable functional failure criteria:
- CR-WF3-2019-00982 documented a failed local leak rate test for the reactor coolant system loop one shutdown cooling system suction inside containment isolation valve and suction header pressure equalizing valve, SI-405B and SI-4052B
- CR-WF3-2019-01281 documented a failed local leak rate test for the component cooling water return inside containment isolation valve backup air supply valve, SA-9085
- CR-WF3-2019-01303 documented a failed local leak rate test for the containment annulus purge inside annulus and inside containment valves, CAP-103 and CAP-104 For each failure identified above, the valves were evaluated against the maintenance rule functional failure criteria for the system they were explicitly assigned to (i.e., SI-405B was evaluated against the safety injection system criteria), but were not evaluated against the applicable containment building system criteria. After discussing the issue with the licensee, the inspectors found that the licensee did not establish any formal system to assure that equipment failures are evaluated against all applicable Maintenance Rule. Instead, condition reports are reviewed by system engineering personnel to perform functional failure reviews. It is left up to the engineers to determine which Maintenance Rule criteria apply through knowledge, experience, and judgement. In the case of the above condition reports, the system engineers repeatedly failed to recognize that the containment building criteria applied to the subject valves. This condition represented a programmatic weakness in the licensees Maintenance Rule program. However, the inspectors noted that accounting for these functional failures would not result in the containment building system being reclassified from Maintenance Rule (a)(2) to (a)(1).
The licensee captured the inspectors observation in condition report CR-WF3-2019-03069.
The licensees immediate corrective action was to perform functional failure determinations for the identified condition reports. The licensee also planned to take action to assess the extent of condition for potentially missed functional failure determinations and to address the potential programmatic weakness in their Maintenance Rule program.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
- On January 11, 2019, the inspector presented the ISI inspection to Mr. J. Dinelli and other members of the licensee staff.
- On January 18, 2019, the inspector presented the baseline radiation safety inspection results to Mr. J. Dinelli and other members of the licensee staff.
- On April 2, 2019, the inspector presented the ISI re-exited to Mr. W. Steelman and other members of the licensee staff.
- On April 11, 2019, the inspector presented the quarterly resident inspector inspection results to Mr. J. Dinelli and other members of the licensee staff.
DOCUMENTS REVIEWED
71111.04 - Equipment Alignment
Procedures
Number Title Revision
OP-002-006 Fuel Pool Cooling and Purification 320
OP-009-002 Emergency Diesel Generator 348
OP-009-005 Shutdown Cooling 40
OP-009-008 Safety Injection System 41
71111.05 - Fire Protection
Procedures
Number Title Revision
RCB-001 Waterford S.E.S Prefire Strategy RCB General Area 11
RAB 2-001 Waterford S.E.S Prefire Strategy Elev. +46.00 RAB(RCA) 13
H&V Mechanical Room
RAB 39-001 Waterford S.E.S Prefire Strategy Elev. -35.00 RAB(RCA) 13
General Area
NS-TB-001 Waterford S.E.S Prefire Strategy TB +15.00 East Including 4
Turbine Building Switchgear, Feedwater Pump B, Instrument and
Station Air Compressors
NS-TB-002 Waterford S.E.S Prefire Strategy TB +15.00 West Including 3
Feedwater Pump A, Auxiliary Feedwater Pump, Main Lube Oil
Storage Tank and Condensate Pump
71111.08 - Inservice Inspection Activities
Condition Reports
CR-WF3-2017-02868 CR-WF3-2017-03650 CR-WF3-2017-00466 CR-WF3-2017-03503
CR-WF3-2017-03329 CR-WF3-2017-04039 CR-WF3-2017-03409 CR-WF3-2017-05007
CR-WF3-2017-06550 CR-WF3-2017-06956 CR-WF3-2017-07012 CR-WF3-2017-07366
CR-WF3-2017-09316 CR-WF3-2017-08962 CR-WF3-2017-07387 CR-WF3-2017-09286
CR-WF3-2017-07436 CR-WF3-2017-02767 CR-WF3-2017-08141 CR-WF3-2017-02768
CR-WF3-2017-01710 CR-WF3-2017-01674 CR-WF3-2017-00233 CR-WF3-2017-01790
CR-WF3-2017-00314 CR-WF3-2017-01791 CR-WF3-2017-00602 CR-WF3-2017-01964
CR-WF3-2017-00849 CR-WF3-2017-01971 CR-WF3-2017-00888 CR-WF3-2017-02323
CR-WF3-2017-01147 CR-WF3-2017-02544 CR-WF3-2017-01176 CR-WF3-2017-02596
CR-WF3-2017-01244 CR-WF3-2017-02696 CR-WF3-2017-01605 CR-WF3-2017-02698
CR-WF3-2018-01005 CR-WF3-2018-02367 CR-WF3-2018-03627 CR-WF3-2018-03946
CR-WF3-2018-05076 CR-WF3-2018-05157 CR-WF3-2018-05866 CR-WF3-2018-01858
CR-WF3-2018-04649 CR-WF3-2018-04663 CR-WF3-2018-04854 CR-WF3-2018-04908
CR-WF3-2018-05569 CR-WF3-2018-00290 CR-WF3-2018-00340 CR-WF3-2018-00459
CR-WF3-2018-00527 CR-WF3-2018-00558 CR-WF3-2018-00580 CR-WF3-2018-00808
Work Orders
475217-01 480840 499227 480436 481611 52794849 480836
495866
Evaluations
Number Title Date
17-WF3-0119 Condition Report #17-4996 06/08/2017
17-WF3-0120 Condition Report #17-5233 06/19/2017
17-WF3-0121 Condition Report #17-5904 07/18/2017
17-WF3-0122 Condition Report #17-5906 07/18/2017
17-WF3-0123 Condition Report #17-6604 08/07/2017
17-WF3-0124 Condition Report #17-6643 08/08/2017
17-WF3-0125 Condition Report #17-6642 08/08/2017
17-WF3-0126 Condition Report #17-7115 08/28/2017
17-WF3-0127 Condition Report #17-07504 09/14/2017
17-WF3-0128 Condition Report #17-10007 12/12/2017
Evaluations
Number Title Date
18-WF3-0001 Condition Report #18-0284 01/15/2018
18-WF3-0002 Condition Report #18-0376 01/20/2018
18-WF3-0003 Condition Report #18-0698 02/03/2018
18-WF3-0004 Condition Report #18-1219 03/01/2018
18-WF3-0005 Condition Report #18-1830 03/31/2018
18-WF3-0006 Condition Report #18-1490 03/14/2018
18-WF3-0007 Condition Report #18-1829 03/31/2018
18-WF3-0008 Condition Report #18-3372 06/14/2018
18-WF3-0009 Condition Report #18-3372 06/14/2018
18-WF3-0010 Condition Report #18-4272 07/31/2018
18-WF3-0011 Condition Report #18-4552 08/10/2018
18-WF3-0012 Condition Report #18-6352 11/11/2018
18-WF3-0013 Condition Report #18-6408 12/12/2018
Calculations
Number Title Revision
ECM-19-001 Waterford 3 Cold Leg Drain Nozzle Weld Overlay Design 3
Calculation
Drawings
Number Title Revision
1900073.510 Waterford Steam Electric Station, Unit 3, Cold Leg Drain 0
Nozzle Weld Overlay Design
7-DH-22A Sht. 1 Large Pipe Isometric Decay Heat Removal to Reactor 14
7-DH-23 Sht. 1 Large Pipe Isometric Decay Heat Removal to Reactor 21
E-74470-761-001 Material Identification 5
E-74470-771-014 Primary Pipe Assembly 6
Engineering
Change
Number Title Revision
EC-81233 Emergent RF22 Structural Weld Overlay (SWOL) repairs for two Cold 2
Leg drain welds. (Ref: CR-WF3-2019-00967 and
Procedures
Number Title Revision
CEP-NDE-0404 Manual Ultrasonic Examination of Ferritic Piping Welds 8
(ASME XI)
CEP-NDE-0423 Manual Ultrasonic Examination of Austenitic Piping Welds 8
(ASME XI)
CEP-NDE-0497 Manual Ultrasonic Examination of Welds in Vessels 7
(Non-App. VIII)
CEP-NDE-0504 Ultrasonic Examination of Small Bore Diameter Piping for 5
Thermal Fatigue Damage
CEP-NDE-0641 Liquid Penetrant Examination (PT) for ASME Section XI 8
CEP-NDE-0731 Magnetic Particle Examination (MT) for ASME Section XI 6
CEP-NDE-0901 VT-1 Examination 6
CEP-NDE-0902 VT-2 Examination 8
CEP-NDE-0903 VT-3 Examination 6
CEP-NDE-0955 Visual Examination (VE) of Bare-Metal Surfaces 303
CEP-PT-001 ASME Section XI Pressure Testing Program 308
EN-DC-319 Boric Acid Corrosion Control Program (BACCP) 11
EN-DC-328 Entergy Nuclear Welding Program 4
EPRI-DMW-PA-1 Procedure for Manual Phased Array Ultrasonic 6
Examination of Dissimilar Metal Welds
EPRI-WOL-PA-1 Procedure for Manual Phased Array Ultrasonic 4
Examination of Weld Overlaid Similar and Dissimilar
Metal Welds
LMT-07-PAUT-005 Performance of Phased Array Instrument Screen Height 1
and amplitude Control Linearity Checks
LMT-08 EPRI-DMW-1 Procedure for Manual Phased Array Ultrasonic 0
Examination of Dissimilar Metal Welds
LMT-08-EPRI-WOL-1 Procedure for Manual Phased Array Ultrasonic 0
Examination of Weld Overlaid Similar and Dissimilar
Metal Welds
LMT-08-PAUT-005 Performance of Phased Array Instrument Screen Height 0
and Amplitude Control Linearity Checks
LMT-10-PAUT-002 Manual Phased Array Ultrasonic Examination and 1
Austenitic and Ferritic Piping Welds
LMT-10-PAUT-007 Fully Encoded Phased Array Ultrasonic Examination of 3
Dissimilar Metal Piping Welds
LMT-10-PAUT-011 Fully Encoded Phased Array Ultrasonic Examination of 1
Weld Overlaid Similar and Dissimilar Metal Piping Welds
LMT-10-PAUT-019 Manual Phased Array Ultrasonic Examination of 0
Procedures
Number Title Revision
Dissimilar Metal Piping Welds
LMT-08-PAUT-005 Performance of Phased Array Instrument Screen Height 0
and Amplitude Control Linearity Checks
SEP-BAC-WF3-001 Waterford 3 Boric Acid Corrosion Control Program 2
(BACCP) Program Section
SEP-ISI-104 Program Section for ASME Section XI, Division 1 WF3 7
Inservice Inspection Program
SEP-CISI-104 Program Section for ASME Section XI, Division 1 WF3 2
Containment Inservice Inspection Program
SEP-PT-WF3-001 Waterford 3 Inservice Inspection Pressure Testing (PT) 0
Program Section
Miscellaneous
Documents Number
Title Date
LO-WLO-2012-00028 Self-Assessment. Pressure Testing Program 07/12/2012
LO-WLO-2012-00046 Self-Assessment. Snapshot Assessment of the 10/05/2012
Waterford 3 Welding and Section XI Repair/Replacement
Programs.
LO-WLO-2012-00078 Self-Assessment. RF22 WF3 Inservice Inspection 09/28/2018
Pre-NRC Inspection Assessment
71111.11 - Licensed Operator Requalification Program and Licensed Operator Performance
Condition Reports
CR-WF3-2019-00367 CR-WF3-2019-00178
Procedures
Number Title Revision
CE-002-002 Maintaining Condensate and Feedwater Chemistry 306
EN-DC-202 NEI 03-08 Materials Initiative Process 7
EN-DC-317 Steam Generator Program 10
NE-002-003 Post-Refueling Startup Testing Control 24
OP-002-005 Chemical and Volume Control 62
OP-002-005 Chemical and Volume Control 63
OP-003-003 Condensate 308
OP-003-035 Auxiliary Feedwater 310
OP-010-003 Plant Startup 347
Procedures
Number Title Revision
OP-010-005 Plant Shutdown 334
OP-010-006 Outage Operations 332
71111.12 - Maintenance Effectiveness
Condition Reports
CR-WF3-2019-01814 CR-WF3-2019-01718 CR-WF3-2019-00982 CR-WF3-2019-00168
CR-WF3-2018-01993 CR-WF3-2017-03493 CR-WF3-2017-08203 CR-WF3-2019-03069
CR-WF3-2019-01303 CR-WF3-2019-01281
Procedures
Number Title Revision
EN-DC-203 Maintenance Rule Program 4
EN-DC-204 Maintenance Rule Scope and Basis 4
EN-DC-205 Maintenance Rule Monitoring 6
EN-DC-206 Maintenance Rule (A)(1) Process 3
71111.13 - Maintenance Risk and Emergent Work
Condition Reports
CR-WF3-2019-00336 CR-WF3-2019-00338 CR-WF3-2019-00520 CR-WF3-2019-01041
CR-WF3-2019-00518 CR-WF3-2019-00513 CR-HQN-2019-00230
Procedures
Number Title Revision
EN-HU-106 Procedure and Work Instruction Use and Adherence 7
EN-MA-119 Material Handling Program 33
EN-OP-115-05 Operation of Components 1
EN-OU-108 Shutdown Safety Management Program (SSMP) 9
MM-008-001 Inside Maintenance Access Hatch and Outside Maintenance 13
Access Hatch Shield Door Opening, Inspection, and Closing
OI-037-000 Operations Risk Assessment Guideline 313
OP-001-003 Reactor Coolant System Drain Down 319
OP-010-006 Outage Operations 32
OP-100-009 Control of Valves and Breakers 43
RF-001-009 Reactor Head 314
Procedures
Number Title Revision
UNT-007-008 Control of Heavy and Critical Loads 321
UNT-007-008 Control of Heavy and Critical Loads 322
Miscellaneous
Documents
Number Title Date
RF-22 Outage Report 01/05/2019
Calculations
Number Title Revision
EC 23453 Waterford 3 RCS Time-to-Boil due to LOSDC at Various Levels 0
and Temperature
ECS00-007 PSA-Study Calc - Basis for Qualitative Level 2, Eternal Events, 2
and Non-PSA SSC Guidance
71111.15 - Operability Determination and Functionality Assessments
Condition Reports
CR-WF3-2019-00967 CR-WF3-2019-01304 CR-WF3-2019-01580 CR-WF3-2019-01826
CR-WF3-2019-01711 CR-WF3-2019-02638 CR-WF3-2019-02640 CR-WF3-2019-03025
CR-WF3-2019-03026
Procedures
Number Title Revision
EN-OP-104 Operability Determination Process 16
Miscellaneous
Documents
Number Title Revision
WF3-ME-09-00007 Flaw Evaluation of CE Design RCP Suction and Discharge, 0
and Safety Injection Nozzle Dissimilar-Metal Welds
Vendor
Documents
Number Title Date
PE-19-14 Functionality Assessment of CE16NFG Upper End Fitting Guide 02/09/2019
Post and Coil Holddown Spring Wear As-Found Condition at
Waterford Unit 3, End of Cycle 22
71111.18 - Plant Modifications
Condition Reports
CR-WF3-2018-01471 CR-WF3-2019-03302
Work Orders
20785-15
Procedures
Number Title Revision
EN-DC-136 Temporary Modifications 18
EN-DC-149 Acceptance of Vendor Documents 14
Miscellaneous
Documents Revision
Number Title or Date
EC 08996 Load Temporary 125KVA SUPS and Disable Shunt Trip on ID 0
EBKR2572 for Surveillance Testing During OP-903-115 and
OP-903-116
EC 74123 Computer/Security SUPS Upgrade (Parent) 0
4158549 Engineering Order 4158549 03/16/2019
EC-82101 Leak Repair CVC-103 Body to Bonnet 0
71111.19 - Post Maintenance Testing
Condition Reports
CR-WF3-2019-00225 CR-WF3-2019-00234 CR-WF3-2019-03139 CR-WF3-2019-03140
Work Orders
2780171 52496259 52776777 52813702
Procedures
Number Title Revision
OP-903-014 Emergency Feedwater Flow Verification 314
OP-903-033 Cold Shutdown IST Valve Tests 52
OP-903-050 Component Cooling Water and Auxiliary Component Cooling 41
Water Pump and Valve Operability Test
OP-903-119 Secondary Auxiliaries Quarterly ISI Valve Test 34
OP-903-121 Safety System Quarterly IST Operability Tests 19
71111.20 - Refueling and Outage Activities
Condition Reports
CR-WF3-2019-03370 CR-WF3-2019-00367 CR-WF3-2019-00178 CR-WF3-2015-01763
CR-WF3-2019-01244 CR-WF3-2015-08122 CR-WF3-2019-01341 CR-WF3-2019-01421
CR-WF3-2019-01459 CR-WF3-2019-01477
Procedures
Number Title Revision
EN-FAP-OU-001 Outage Position Fundamentals 2
EN-FAP-OU-105 Outage Execution 7
EN-OU-108 Shutdown Safety Management Program (SSMP) 9
EN-RE-302 PWR Reactivity Maneuver 5
NE-002-030 Initial Criticality 307
OI-041-000 Operations Outage Guide 17
OP-001-003 Reactor Coolant System Drain Down 319
OP-010-003 Plant Startup 347
OP-010-005 Plant Shutdown 334
OP-010-006 Outage Operations 331
PLG-009-014 Conduct of Planned Outages 315
RF-002-001 New Fuel Receipt 326
RF-005-001 Fuel Movement 321
RF-005-002 Refueling Equipment Operation 341
Drawings
Number Title Revision
1564-8353 Refueling Machine Hoist Assembly 3
5817-14348, Hoist Fuel Hoist Box Assembly 0
sheet 7
71111.22 - Surveillance Testing
Condition Reports
CR-WF3-2019-00165 CR-WF3-2019-00880 CR-WF3-2019-01534 CR-WF3-2019-01535
CR-WF3-2019-01572 CR-WF3-2019-01524 CR-WF3-2019-01514 CR-WF3-2019-02208
Work Orders
2550737
Procedures
Number Title Revision
OP-903-027 Containment Closeout Inspection 305
OP-903-033 Cold Shutdown IST Valve Tests 52
OP-903-116 Train B Integrated Emergency Diesel Generator/Engineering 47
Safety Features Test
OP-903-116 Train B Integrated Emergency Diesel Generator/Engineering 48
Safety Features Test
PE-005-001 Containment Integrated Leak Rate Test 302
PE-005-001 Containment Integrated Leak Rate Test 303
PE-005-001 Containment Integrated Leak Rate Test 304
Miscellaneous
Documents
Number Title Date
Waterford 3 2019 ILRT Test Report 03/2019
Letter W3F1-92-0473 Review of NRC Information Notice 92-20, Inadequate 12/22/1992
71124.01 - Radiological Hazard Assessment and Exposure Controls
Condition Reports
CR-WF3-2017-03340 CR-WF3-2018-00506 CR-WF3-2018-02332 CR-WF3-2018-02915
CR-WF3-2018-04095 CR-WF3-2018-04512 CR-WF3-2018-06188 CR-WF3-2018-06202
CR-WF3-2018-06296 CR-WF3-2019-00037 CR-WF3-2019-00513 CR-WF3-2019-00737
CR-WF3-2019-00869
Procedures
Number Title Revision
EN-LI-114 Regulatory Performance Indicator Process 12
EN-RE-220 PWR Control of Miscellaneous Material in the Spent Fuel Pool 03
EN-RP-100 Radiation Worker Expectations 12
EN-RP-101 Access Control for Radiologically Controlled Areas 15
EN-RP-108 Radiation Protection Posting 21
EN-RP-121 Radioactive Material Control 15
EN-RP-143 Source Control 13
Audits and Self-Assessments
Number Title Date
Benchmark Report: Facility Visit, Tour, Observations at 05/24/2018
Catawba
LO-WLO-2017-00070 Pre-NRC Self-Assessment: Radiation Safety - Public and 12/06/2018
Occupational - IP 71124.01 Access Control to
Radiologically Significant Areas
Radiation Work Permits
Number Title Revision
20190702 Refuel 22 - Disassembly of Reactor Head and All Associated Work 00
Activities
20190705 Refuel 22 - Reassembly of Reactor Head and Associated Work 00
Activities including Staging/Destaging of Equipment
20190708 Refuel 22 - ICI Removal/Installation/Cut Up of ICIs, Work on ICI 00
Equipment and Replacement Swageloc Bodies
20190724 Refuel 22 - Waterford 3 Fuel Transfer System Maintenance and 00
Repairs including Lower Cavity Decon Activities, Support Activities,
HP Job Coverage and Surveys
Radiological Surveys
Number Title Date
WF3-1901-0037 -35 RAB Hallway 01/02/2019
WF3-1901-0120 Radwaste Solidification Building 01/04/2019
WF3-1901-0121 FHB +46 Fuel Handling Area 01/04/2019
WF3-1901-0361 -11 Reactor Containment Building 01/08/2019
WF3-1901-0368 +46 Reactor Containment Building 01/08/2019
WF3-1901-0474 -4 Reactor Containment Building 01/09/2019
WF3-1901-0765 RAB +21 Drumming Station 01/14/2019
Air Sample Surveys
Number Title Date
OL-050218-015 Air Sample: Fuel Pool Purification Pump Check 05/02/2018
Valve (FS-426) Inspection
OL-110118-061 Air Sample: Welding Purification Ion Exchanger B Plug 11/01/2018
OL-121818-076 Air Sample: Chemistry Sampling of SDC B 12/18/2018
OL-010119-001 Air Sample: Safeguards A Pump Room 01/01/2019
Miscellaneous
Documents
Number Title Date
2016-2000 Waterford 3 STM Alarm Setpoint Evaluation 03/24/2016
Comparison of Detection Limits Using Argos, Gem5, TEM 07/11/2017
and Pancake Probe
WO 52798853-01 Semi-Annual Sealed Source Leak Test 06/12/2018
WO 52827432-01 Semi-Annual Sealed Source Leak Test 12/06/2018
NRC Form 748 NSTS Annual Inventory Reconciliation Form 01/01/2019
Ready for Issue High, Locked High, and Very High 01/15-17/2019
Radiation Area Key Inventory
RF22 Rapid Trending Report 01/15-18/2019
71124.02 - Occupational ALARA Planning and Controls
Condition Reports
CR-WF3-2018-5439 CR-WF3-2018-5568 CR-WF3-2018-5720 CR-WF3-2018-5780
CR-WF3-2018-6299 CR-WF3-2018-0772
Procedures
Number Title Revision
EN-RP-102 Radiological Control 6
EN-RP-105 Radiological Work Permits 18
EN-RP-110-02 Elemental Cobalt Sampling 0
EN-RP-110-03 Collective Radiation Exposure (CRE) Reduction Guidelines 4
EN-RP-110-04 Radiation Protection Risk Assessment Process 7
EN-RP-110-06 Outage Dose Estimating and Tracking 1
HP-001-114 Control of Temporary Shielding 17
Radiation Work Permits, ALARA Planning, In-Progress Reviews
Number Title Revision
20190401 REFUEL 22 - Air Operated Valve (AOV)/Motor Operated Valve(MOV) 0
Valve Group Contaminated and Clean System Valve Work Outside the
Reactor Containment Building
20190613 REFUEL 22 - Alloy 600 Inspections in Containment and support 0
activities including Decontamination and Preparation Support
20190617 REFUEL 22 - Various decon activities in the Upper Reactor Cavity, 0
Staging/Destaging including Change out of Tri-Nuc Filters and
Radiation Work Permits, ALARA Planning, In-Progress Reviews
Number Title Revision
Underwater Vacuuming. NO Entry into any posted Locked High
Radiation Areas
20190702 REFUEL 22 - Disassembly of Reactor Head and All Associated Work 0
Activities
20190705 REFUEL 22 - Reassembly of Reactor Head and Associated 0
Work Activities including Staging/Destaging of Equipment
20190724 REFUEL 22 - Waterford 3 Fuel Transfer System 0
Maintenance and Repairs including Lower Cavity Decon Activities,
Support Activities, HP Job Coverage and Surveys
Radiation Surveys
Number Title Date
WF3-1705-0127 RAB-35 Center Wing Area: Valve BM-110 05/03/2017
WF3-1705-0285 RAB-35 Center Wing Area: SI MVAA602B 05/06/2017
WF3-1901-0833 RAB-4 Letdown Heat Exchanger: CVC-113B 01/15/2019
WF3-1901-0885 RAB-4 Letdown Heat Exchanger: CVC-113A 01/17/2019
Miscellaneous
Documents
Number Title Date
RF21 RP Critique/Lessons Learned 09/20/2017
WF3 2019 Radiation Work Permit Totals (millirem) 01/13/2019
WF3 Site Dose Totals Online/Outage for the past 3 years 01/09/2019
(dose in Person-Rem)
EN-RW-104 Scaling Factors - 10 CFR Part 61 Waste Stream Sample 02/02/2108
Screening and Evaluation
71151 - Performance Indicator Verification
Condition Reports
CR-WF3-2018-6202 CR-WF3-2018-6296
Procedures
Number Title Revision
EN-LI-114 Regulatory Performance Indicator Process 12
EN-RP-101 Access Control for Radiologically Controlled Areas 15
Miscellaneous
Documents
Number Title Date
W3F1-2018-0024 NRC Performance Indicator (PI) Data - 1st Quarter 2018 04/12/2018
ROP Data
W3F1-2018-0042 NRC Performance Indicator (PI) Data - 2nd Quarter 2018 07/13/2018
ROP Data
W3F1-2018-0066 NRC Performance Indicator (PI) Data - 3rd Quarter 2018 10/21/2018
ROP Data
W3F1-2019-0003 NRC Performance Indicator (PI) Data - 4th Quarter 2018 01/10/2018
ROP Data
EN-LI-114 Regulatory Performance Indicator Process - NRC 04/03 and
Performance Indicator Technique/Data Sheet: WF3 2018 12/2018
First Quarter
EN-LI-114 Regulatory Performance Indicator Process - NRC 07/02 and
Performance Indicator Technique/Data Sheet: WF3 2018 03/2018
Second Quarter
EN-LI-114 Regulatory Performance Indicator Process - NRC 10/03 and
Performance Indicator Technique/Data Sheet: WF3 2018 08/2018
Third Quarter
EN-LI-114 Regulatory Performance Indicator Process - NRC 01/01/2019
Performance Indicator Technique/Data Sheet: WF3 2018
Fourth Quarter
71152 - Identification & Resolution of Problems
Condition Reports
CR-WF3-2019-00982 CR-WF3-2019-01281 CR-WF3-2019-01303 CR-WF3-2019-03069
Procedures
Number Title Revision
EN-DC-203 Maintenance Rule Program 4
EN-DC-205 Maintenance Rule Monitoring 6
EN-DC-206 Maintenance Rule (A)(1) Process 3
STA-001-002 Containment Purge Valve Leakage Test 303
STA-001-004 Local Leak Rate Test (LLRT) 316
STA-001-005 Leakage Testing of Air and Nitrogen Accumulators for Safety 320
Related Valves
71153 - Event Notification and Follow-up
Condition Reports
Vendor
Documents
Number Title Revision
1900073.405 Acceptability Evaluation for RCS Operation for the Cold Leg Drain A
Nozzle-to-Safe End Dissimilar Metal Welds (07-009 and 11-007)
Axial Indication at Waterford Steam Electric Plant, Unit 3 to Support
Past Operability Determination
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, Requests for Withholding.
Information Request
November 13, 2018
Notification of Inspection and Request for Information
Waterford Steam Electric Station, Unit 3
NRC Inspection Report 05000382/2019001
INSERVICE INSPECTION DOCUMENT REQUEST
Inspection Dates: January 4 - 12, 2019
Inspector: Ronald A. Kopriva
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 Ron Kopriva, by
December 28, 2018, 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 January 4, 2019, a reactor inspector from the Nuclear Regulatory Commissions (NRC)
Region IV office will perform the baseline inservice inspection at Waterford Steam Electric
Station, Unit 3, 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. These documents have been divided into two groups. The first group
(Section A of the enclosure) identified information to be provided prior to the inspection to
ensure that the inspector is adequately prepared. The second group (Section B of the
enclosure) identifies the information the inspector will need upon arrival at the site. It is
important that all of these documents are up to date and complete in order to minimize the
number of additional documents requested during the preparation and/or the onsite portions
of the inspection.
We have discussed the schedule for these inspection activities with your staff and
understand that our regulatory contact for this inspection will be Ms. Maria Zamber of your
licensing organization. The tentative inspection schedule is as follows:
Preparation week: December 28, 2018 - January 3, 2019
Onsite dates: January 4 - 12, 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 the lead inspector Ronald A. Kopriva at (817) 200-1104. (email to:
ron.kopriva@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 previous Section 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
There are no Reactor Pressure Vessel Head inspections planned during this scheduled
outage.
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 Steam Generator Inspections
There are no Steam Generator Tube inspections planned during this scheduled
outage. It is our understanding that you will be opening the secondary side of the
Steam Generators for inspection and measurements for the Feedwater Ring
modification. We would like to obtain information as to your inspection scope and
planned activities.
A.5 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., Condition Reports) entered into your corrective action
program since the beginning of the last refueling outage. For example, a list based
upon data base searches using key words related to piping such as: inservice
inspection, ASME Code, Section XI, NDE, cracks, wear, thinning, leakage, rust,
corrosion, boric acid, or errors in piping examinations.
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
Updated schedules for inservice inspection/nondestructive examination activities, including
planned welding activities, and schedule showing contingency repair plans, if available.
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.
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.
There are no Reactor Pressure Vessel Head inspections planned during this scheduled
outage.
Boric Acid Corrosion Control Program
1. Please provide boric acid walk down inspection results, an updated list of boric acid
leaks identified so far this outage, associated corrective action documentation, and
overall status of planned boric acid inspections.
2. Please provide any engineering evaluations completed for boric acid leaks identified
since the end of the last refueling outage. Please include a status of corrective actions
to repair and/or clean these boric acid leaks. Please identify specifically which known
leaks, if any, have remained in service or will remain in service as active leaks.
Steam Generator Inspections
There are no Steam Generator Tube inspections planned during this scheduled
outage. It is our understanding that you will be opening the secondary side of the
Steam Generators for inspection and measurements for the Feedwater Ring
modification. We would like to obtain information as to your inspection scope and
planned activities.
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
at Waterford-3
Dates of Inspection: 01/14/2019 to 01/18/2019
Integrated Report 2019001
Inspection areas are listed in the attachments below.
Please provide the requested information on or before Thursday, January 03, 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.
The information should be provided in electronic format or a secure document management
service. 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 Natasha Greene at 817-200-1154 or via
e-mail at Natasha.Greene@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: January 22, 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.
I. Provide a summary of any changes to plant operation that have resulted or could result
in a significant new radiological hazard. For each change, please provide the
assessment conducted on the potential impact and any monitoring done to evaluate it.
J. List of active radiation work permits and those specifically planned for the on-site
inspection week.
K. Please provide a list of air samples taken to verify engineering controls and a separate
list for breathing air samples in airborne radiation areas or high contamination work
areas. Please include the RWP the breathing air sampling supports.
- L. Please provide the current radioactive source inventory, listing all radioactive sources
that are required to be leak tested. Indicate which sources are deemed 10 CFR Part 20,
Appendix E, Category 1 or Category 2. Please indicate the radioisotope, initial and
current activity (w/assay date), and storage location for each applicable source.
M. The last two leak test results for all required/applicable radioactive sources that have
failed its leak test within the last two years. Provide any applicable condition reports.
- N. A list of all non-fuel items stored in the spent fuel pools, and if available, their appropriate
dose rates (Contact / @ 30cm)
inspection date. The list 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).
P. A list describing VHRAs and TS HRAs (> 1 rem/hour) that are current and historical.
Include their current status, locations, and control measures.
Q. Temporary effluent monitor locations and calibrations (AMS-4) used to monitor normally
closed doors or off-normal release points (e.g., equipment hatch or turbine heater bay
doors). Include any CRs associated with this monitoring or instrumentation.
2. Occupational ALARA Planning and Controls (71124.02)
Date of Last Inspection: September 10, 2018
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 date of last
inspection, focusing on ALARA
D. Procedure index for ALARA Program procedures and other related disciplines.
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 Planning
3. ALARA Reviews
4. ALARA Committee
5. Radiation Work Permit Preparation
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) written since the date of last inspection, related to the ALARA
program, including exceeding RWP Dose Estimates.
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 work activities (RWPs) greater than 1 rem, since date of last inspection,
including the original dose estimates and actual doses accrued. (Excel format
preferred). Please provide all revisions/changes, as well as any related RWPs that
support the work activity.
dose and dose rate settings, and if available, the planned dose. Include planning
documents and surveys. Include radiological risk assessments and proposed control
measures.
J. Site dose totals for the past 3 years (based on dose of record). Also provide the current
year-to-date (YTD) collective radiation exposure (CRE). In addition, please provide
another document that separates the online and outage doses for the past 3 years.
- K. Most recent assessment of your isotopic mix, including the hard-to-detect radionuclides
and alpha hazards. Include a list of new and historical exposure issues (radiological
source term or high exposure areas/activities).
- L. If available, provide a copy of the lessons learned from the most recently completed
outage for each unit. Include a summary list of any associated corrective action
documents and the current status of any corrective actions assigned.
M. Please provide the methods/reports that are in your process to meet the requirements of
CFR 20.1101(c) for periodic review of your RP program.
N. Current exposure trends (BRAC dose rates and/or source term information).
SUNSI Review ADAMS: Non-Publicly Available Non-Sensitive Keyword:
By: Yes No Publicly Available Sensitive NRC-002
OFFICE SRI:DRP/D SRI:DRP/D RI:DRP/D TL:DRS/IPAT BC:DRS/EB1 BC:DRS/EB2
NAME FRamirez CSpeer JMelfi RKellar VGaddy GPick
SIGNATURE FCR CAS JFM RLK vgg GAP
DATE 4/26/19 4/25/19 4/26/19 4/29/19 4/29/19 4/26/19
OFFICE BC:DRS/RCB BC:DNMS/RIB BC:DRS/OB BC:DRP/D
NAME NMakris GWarnick GWerner NOKeefe
SIGNATURE NM JFK for GEW NFO
DATE 4/29/19 4/30/19 04/29/19 4/30/19