ML24141A101
| ML24141A101 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 05/20/2024 |
| From: | John Dixon, Alfred Sanchez NRC/RGN-IV/DORS/PBD |
| To: | Sullivan J Entergy Operations |
| References | |
| IR 2023004 | |
| Download: ML24141A101 (31) | |
See also: IR 05000382/2023004
Text
Joseph Sullivan, Site Vice President
Entergy Operations, Inc.
17265 River Road
Killona, LA 70057
SUBJECT:
WATERFORD STEAM ELECTRIC STATION, UNIT 3 - AMENDED
INTEGRATED INSPECTION REPORT 05000382/2023004
Dear Joseph Sullivan:
On December 31, 2023, the U.S. Nuclear Regulatory Commission (NRC) completed an
inspection at Waterford Steam Electric Station, Unit 3. On January 24, 2024, 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. This letter and enclosure
amend Inspection Report 05000382/2023004, issued on February 12, 2004, ADAMS accession
number ML24039A199.
This amendment clarifies the completed samples identified in Inspection Procedure 71111.08 to
reflect that samples were completed in Sections 03.02 and 03.04
Two findings of very low safety significance (Green) are documented in this report. Two of these
findings involved violations of NRC requirements. We are treating these violations as non-cited
violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.
If you contest the violations or the significance or severity of the violations documented in this
inspection report, you should provide a response within 30 days of the date of this inspection
report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN:
Document Control Desk, Washington, DC 20555-0001; with copies to the Regional
Administrator, Region IV; the Director, Office of Enforcement; and the NRC Resident Inspector
at 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.
May 20, 2024
J. Sullivan
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 Title 10 of the Code of Federal Regulations 2.390, Public
Inspections, Exemptions, Requests for Withholding.
Sincerely,
John L. Dixon, Jr., Chief
Reactor Projects Branch D
Division of Operating Reactor Safety
Docket No. 05000382
License No. NPF-38
Enclosure:
As stated
cc w/ encl: Distribution via LISTSERV
Signed by Dixon, John
on 05/20/24
J. Sullivan
3
WATERFORD STEAM ELECTRIC STATION, UNIT 3 - AMENDED INTEGRATED
INSPECTION REPORT 05000382/2023004 - DATED MAY 20, 2024
DISTRIBUTION:
JMonninger, ORA
JLara, ORA
GMiller, DORS
MHay, DORS
DCylkowski, RC
FGaskins, RIV/OEDO
VDricks, ORA
LWilkins, OCA
JDrake, NRR
AMoreno, RIV/OCA
RAlexander, RSLO
JDixon, DORS
ASanchez, DORS
APatz, DORS
DChilds, DORS
LReyna, DORS
R4-DORS-IPAT
R4Enforcement
DOCUMENT NAME: WATERFORD STEAM ELECTRIC STATION, UNIT 3 - AMEMDED INTEGRATED
INSPECTION REPORT 05000382/2023004
ADAMS Accession Number: ML24141A101
SUNSI Review
ADAMS:
Non-Publicly Available
Non-Sensitive
Keyword:
By: AAS
Yes No
Publicly Available
Sensitive
OFFICE
BC:DORS/EB2
BC:DORS/D
NAME
NTaylor
JDixon
SIGNATURE
/RA/
/RA/
DATE
05/20/2024
05/20/2024
J. Sullivan
4
OFFICIAL RECORD COPY
U.S. NUCLEAR REGULATORY COMMISSION
Inspection Report
Docket Number:
05000382
License Number:
Report Number:
Enterprise Identifier:
I-2023-004-0009
Licensee:
Entergy Operations, Inc.
Facility:
Waterford Steam Electric Station, Unit 3
Location:
Killona, LA 70057
Inspection Dates:
October 1, 2023, to December 31, 2023
Inspectors:
D. Childs, Resident Inspector
J. Drake, Senior Reactor Inspector
N. Greene, Senior Health Physicist
R. Kopriva, Senior Project Engineer
J. O'Donnell, Senior Health Physicist
A. Patz, Senior Resident Inspector
B. Tharakan, Technical Assistant
Approved By:
John L. Dixon, Jr., Chief
Reactor Projects Branch D
Division of Operating Reactor Safety
2
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees
performance by conducting an integrated inspection at Waterford Steam Electric Station, 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.
List of Findings and Violations
Failure to Follow ALARA Planning and Control Procedures Resulting in Unplanned Dose
Cornerstone
Significance
Cross-Cutting
Aspect
Report
Section
Occupational
Radiation Safety
Green
Open/Closed
[H.4] -
Teamwork
The inspectors identified a Green finding and associated non-cited violation (NCV) of
Technical Specification 6.8.1.a for a failure to follow as low as reasonably achievable
(ALARA) planning and control procedures during the 2024 Unit 1 refueling outage.
Specifically, the licensee's planning or radiological controls did not prevent unplanned dose for
two separate work activities conducted during the 2024 refueling outage.
Failure to Maintain FLEX Equipment Starting Batteries
Cornerstone
Significance
Cross-Cutting
Aspect
Report
Section
Mitigating
Systems
Green
Open/Closed
[H.12] - Avoid
Complacency
The inspectors reviewed a self-revealed Green finding and associated NCV of 10 CFR
50.155(b)(1), which states, in part, strategies and guidelines to mitigate beyond-design-basis
events from natural phenomena must be capable of being implemented site-wide and must
include maintaining or restoring core cooling capabilities. Specifically, from approximately
February 14 to May 16, 2023, the licensee failed to ensure the starting batteries for the
FLEX N and N+1 diesel generators had sufficient capacity to perform their required functions.
Additional Tracking Items
Type
Issue Number
Title
Report Section
Status
Steam Generator 1 In-Situ
Tube Pressure Testing
Failures.
Open
3
PLANT STATUS
Unit 3 began the inspection period at rated thermal power. On October 14, 2023, the unit was
shut down for refueling outage 25 and remained shut down 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 activities described in IMC 2515,
Appendix D, Plant Status, observed risk significant activities, and completed on-site portions of
IPs. The inspectors reviewed selected procedures and records, observed activities, and
interviewed personnel to assess licensee performance and compliance with Commission rules
and regulations, license conditions, site procedures, and standards.
REACTOR SAFETY
71111.01 - Adverse Weather Protection
Impending Severe Weather Sample (IP Section 03.02) (1 Sample)
(1)
The inspectors evaluated the adequacy of the overall preparations to protect
risk-significant systems against external flooding from heavy rains and high winds on
November 20, 2023.
71111.04 - Equipment Alignment
Partial Walkdown Sample (IP Section 03.01) (1 Sample)
The inspectors evaluated system configurations during partial walkdowns of the following
systems/trains:
(1)
train B 7KV, 4KV and 480V safety-related electrical distribution systems while train A
was out for planned maintenance on November 2, 2023
Complete Walkdown Sample (IP Section 03.02) (1 Sample)
(1)
The inspectors evaluated system configurations during a complete walkdown of the
containment fan cooler system on October 31, 2023.
4
71111.05 - Fire Protection
Fire Area Walkdown and Inspection Sample (IP Section 03.01) (8 Samples)
The inspectors evaluated the implementation of the fire protection program by conducting a
walkdown and performing a review to verify program compliance, equipment functionality,
material condition, and operational readiness of the following fire areas:
(1)
fire area RAB 5-001, elevation +35.00' reactor auxiliaries building electrical
penetration room B on October 17, 2023
(2)
fire area RAB 6-001, elevation +35.00' reactor auxiliaries building electrical
penetration room A on October 18, 2023
(3)
fire area RCB-001, elevations -4.00' and +21.00' reactor containment building on
October 20, 2023
(4)
fire area RAB 16-001, elevation +21.00' emergency diesel generator 3A room on
October 23, 2023
(5)
fire area RCB-001, elevation +46.00' reactor containment building on October 24,
2023
(6)
fire area RAB 8C-001, elevation +21' switchgear room AB on October 30, 2023
(7)
fire area RAB 9-001, elevation +21.00' remote shutdown room on October 30, 2023
(8)
fire area RAB 1E-001, elevation +35.00' cable vault on November 8, 2023
71111.07A - Heat Exchanger/Sink Performance
Annual Review (IP Section 03.01) (1 Sample)
The inspectors evaluated readiness and performance of:
(1)
component cooling water heat exchanger A on November 3, 2023
71111.08P - Inservice Inspection Activities (PWR)
The inspectors verified that the reactor coolant system boundary, 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 October 23 to November 30, 2023.
PWR Inservice Inspection Activities Sample - Nondestructive Examination and Welding
Activities (IP Section 03.01) (1 Sample)
The inspectors verified that the following nondestructive examination and welding activities
were performed appropriately:
5
(1)
Dye Penetrant Examination
Reactor Coolant System, Component ID # RCI TE0112 CD1, 1B Cold Leg
Thermowell, Report No. BOP-PT-23-069
Magnetic Particle Examination
Main Steam, Component ID # 04-071, S/G #2 Upper Key Support Lug Weld
@ 0 Degree Axis, Report No W-ISI-MT-23-001
Visual Examination
Component Cooling Water, Component ID # CCRR-00322, Rigid Restraint,
Report No. W-ISI-VT-23-009
Primary Containment (PC), Component ID # DS-5, Containment Dome Outer
Surface, Report No. W-CISI-VT23-001
Primary Containment (PC), Component ID # WS-13, Containment Liner Outer
Surface 352.8 degrees - 138 degrees Azimuth, Report No. W-CISI-VT23-003
Primary Containment (PC), Component ID # WS-01, Containment Liner Inner
Surface 0 degrees- 90 degrees Azimuth at - 4-foot Elevation, Report
No. W-CISI-VT23-006
Primary Containment (PC), Component ID # WS-10, Containment Liner Inner
Surface 90 degrees-180 degrees Azimuth at + 46-foot Elevation, Report
No. W-CISI-VT23-014
Ultrasonic Examination
Charging (CH), Component ID # 30-002, 2-inch Pipe to Elbow Weld, Report
No. W-ISI-UT-23-011
Charging (CH), Component ID # 30-018, Elbow to 2-inch Pipe Weld, Report
No. W-ISI-UT-23-015
Charging (CH), Component ID # 30-009, 2-inch Pipe to Tee Weld, Report
No. W-ISI-UT-23-013
Charging (CH), Component ID # 30-010, 2-inch Pipe to Tee Weld, Report
No. W-ISI-UT-23-014
Charging (CH), Component ID # 30-008, 2-inch Pipe to Pipe Weld, Report
No. W-ISI-UT-23-012
Welding Activities
o
Reactor Coolant System, ID # RC ITE0112 DC1, Thermowell
Cap - Fillet Weld FW-1
o
Safety Injection System, ID # SI MVAA303 A, Valve, Socket
Welds - FW-1 and SW-6
6
PWR Inservice Inspection Activities Sample - Vessel Upper Head Penetration Inspection
Activities (IP Section 03.02 (1 Sample)
The inspectors verified that the license conducted the following vessel upper head
penetration inspections and addressed any identified defects appropriately:
(1)
Visual examination, bare metal visual, reactor vessel closure head.
PWR Inservice Inspection Activities Sample - Boric Acid Corrosion Control Inspection Activities
(IP Section 03.03) (1 Sample)
The inspectors verified the licensee is managing the boric acid corrosion control program
(1)
Evaluation # 22-WF3-0029, Component ID # BAMMVAAA118B,
Evaluation # 22-WF3-0030, Component ID # SI MPMP0002A,
Evaluation # 22-WF3-0031, Component ID # SI MVAAA2031A,
Evaluation # 22-WF3-0032, Component ID # CVCIDPI0203,
Evaluation # 22-WF3-0033, Component ID # BAMMVAAA118B,
Evaluation # 22-WF3-0034, Component ID # CS MPMP0001B,
Evaluation # 22-WF3-0035, Component ID # FS MPMP0001B,
Evaluation # 22-WF3-0036, Component ID # CS MPMP0001A,
Evaluation # 22-WF3-0037, Component ID # SI MPMP0002A,
Evaluation # 22-WF3-0038, Component ID # BAMMVAAA141,
Evaluation # 22-WF3-0039, Component ID # FS MVAAA426,
CR-WF3- 22-6910
Evaluation # 23-WF3-0001, Component ID # FS MVAAA512,
Evaluation # 23-WF3-0002, Component ID # CS MPMP0002B,
Evaluation # 23-WF3-0003, Component ID # BM MPMP0009,
Evaluation # 23-WF3-0004, Component ID # BM MPMP0001,
Evaluation # 23-WF3-0005, Component ID # SI MVAAA119B,
Evaluation # 23-WF3-0006, Component ID # SI MVAAA205A,
Evaluation # 23-WF3-0007, Component ID # SI MPMP0001A.
CR-WF3-3-0180
Evaluation # 23-WF3-0008, Component ID # SI MVAAA2351,
7
CR-WF3- 23-1215
Evaluation # 23-WF3-0009, Component ID # CVCMVAAA189A,
Independent Boric Acid Walkdown, October 27, 2023
Boric Acid Walkdown with Boric Acid Engineer, October 28, 2023
PWR Inservice Inspection Activities Sample - Steam Generator Tube Inspection Activities
(Section 03.04) (1 Sample)
The inspectors verified that the licensee is monitoring the steam generator tube integrity
appropriately through a review of the results of the 100 percent full length eddy current
inspection of all tubes with bobbin coil probe. Four tubes in replacement steam generator 1
exhibited wear that exceeded the tube integrity criteria provided in the degradation
assessment (DA).
1.
There were four tubes that required in situ pressure testing to support the condition
monitoring assessment based on the DA and Electric Power Research Institute in situ
pressure test guidelines. Additional discussion of these activities is included in an
unresolved item in the results section of this report.
Two tubes from steam generator 1 (R1 C112 and R1 C138) were tested over the
range of prescribed test pressures and successfully reached and maintained the
structural limit pressure test of 5500 psi. No tube leakage was measured at any test
pressure for these two tubes.
Two tubes from steam generator 1 (R1 C4 and R2 C35) were tested over the range
of prescribed test pressures. Tube R1 C4 was unable to reach the structural limit test
pressure as it experienced pop-through at 5243 psi. No leakage was measured in
this tube at lower test pressures prior to the pop-through. Tube location R2 C35 was
able to temporarily achieve the structural limit test pressure point at 5500 psi, but lost
leak tight integrity via pop-through after a combined 131 seconds above the target
pressure of 5500 psi. The combined 131 seconds at pressure was achieved by a
period of 41 seconds above the test target, then briefly dropping below 5500 psi
before being re-established above 5500 psi for 90 seconds prior to the pop through.
No tube leakage was observed at any test pressure below the structural limit test.
2.
No tube leakage was reported during this operating interval. The inspectors verified that
the licensee is monitoring the steam generator tube integrity appropriately through
a review of the examinations.
There were a total of 48 tubes plugged, including 27 tubes in steam generator 1 and
21 tubes in steam generator 2.
Problem Identification and Resolution. Review of in-service inspection items. (Inspection
Procedure 71152 - Problem Identification and Resolution). The inspector evaluated a
sample of 16 condition reports associated with in-service inspection activities. No
findings or violations of more than minor significance were identified.
8
71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance
Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01)
(1 Sample)
(1)
The inspectors observed and evaluated licensed operator performance in the control
room during unit shutdown for refueling outage on October 13-14, 2023.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
(1)
The inspectors observed and evaluated a licensed operator exam in the simulator on
December 12, 2023.
71111.12 - Maintenance Effectiveness
Maintenance Effectiveness (IP Section 03.01) (5 Samples)
The inspectors evaluated the effectiveness of maintenance to ensure the following
structures, systems, and components remain capable of performing their intended function:
(1)
containment spray pump A following breaker failure on September 22, 2023
(2)
permanent temporary emergency diesel generator following failure of heating,
ventilation, and air conditioning system on November 27, 2023
(3)
shield building ventilation train B failures on December 13, 2023
(4)
controlled ventilation area system following identification of incorrect open and close
times in design basis calculations on December 14, 2023
(5)
essential services chilled water chiller AB following trip while in service for train A on
December 26, 2023
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (5 Samples)
The inspectors evaluated the licensee's justifications and actions associated with the
following operability determinations and functionality assessments:
(1)
containment particulate-iodine-gas radiation monitor operability following restoration
of particulate channel only on October 2, 2023
(2)
shutdown cooling trains A and B following instrument air transients on October 15,
2023
(3)
low pressure safety injection train B following identification of condensation inside
minimum flow recirculation valve actuator on November 28, 2023
(4)
plant stack radiation monitoring following failures and maintenance of plant stack
particulate-iodine-gas and plant stack wide range gas monitor on November 30, 2023
(5)
engineered safety features actuation system trains A and B following identification of
no fire seals on December 13, 2023
71111.20 - Refueling and Other Outage Activities
Refueling/Other Outage Sample (IP Section 03.01) (1 Sample)
9
(1)
(Partial)
The inspectors evaluated refueling outage 25 activities from October 14, 2023, to the
end of the inspection period, December 31, 2023. The sample will be closed in a
future inspection report.
71111.24 - Testing and Maintenance of Equipment Important to Risk
The inspectors evaluated the following testing and maintenance activities to verify system
operability and/or functionality:
Post-Maintenance Testing (PMT) (IP Section 03.01) (3 Samples)
(1)
startup transformer B following breaker repair on October 12, 2023
(2)
low pressure safety injection pump B following identification of condensation in
minimum flow valve on December 4, 2023
(3)
auxiliary component cooling water train B following modification implementation for
flow control valve on December 19, 2023
Surveillance Testing (IP Section 03.01) (4 Samples)
(1)
emergency diesel generator A safety injection actuation test with concurrent loss of
offsite power on October 18, 2023
(2)
N+1 FLEX diesel generator on November 14, 2023
(3)
auxiliary component cooling water train B on December 7, 2023
(4)
charging pump A for boron flowrate verification on December 14, 2023
Inservice Testing (IP Section 03.01) (2 Samples)
(1)
safety injection valve 307A, safety injection tank 1A fill/drain valve testing on
November 6, 2023
(2)
controlled ventilation area system train B on December 18, 2023
Containment Isolation Valve Testing (IP Section 03.01) (1 Sample)
(1)
leak rate test on containment isolation valve SI-407A, loop 2 shutdown cooling
suction outside containment isolation, on October 23, 2023
RADIATION SAFETY
71124.01 - Radiological Hazard Assessment and Exposure Controls
Radiological Hazard Assessment (IP Section 03.01) (1 Sample)
(1)
The inspectors evaluated how the licensee identifies the magnitude and extent of
radiation levels and the concentrations and quantities of radioactive materials and
how the licensee assesses radiological hazards.
Instructions to Workers (IP Section 03.02) (1 Sample)
10
(1)
The inspectors evaluated how the licensee instructs workers on plant-related
radiological hazards and the radiation protection requirements intended to protect
workers from those hazards.
Contamination and Radioactive Material Control (IP Section 03.03) (2 Samples)
The inspectors observed/evaluated the following licensee processes for monitoring and
controlling contamination and radioactive material:
(1)
surveys of potentially contaminated material leaving the radiologically controlled area
exit
(2)
workers exiting the reactor containment building during a refueling outage
Radiological Hazards Control and Work Coverage (IP Section 03.04) (3 Samples)
The inspectors evaluated the licensee's control of radiological hazards for the following
radiological work:
(1)
Move of the upper guide structure from the reactor vessel to the lower cavity using
radiation work permit (RWP) 2023-702.
(2)
Chemical sampling and engineering inspection on the reactor vessel head using
RWP 2023-0714.
(3)
Breach and disassembly of gaseous waste valve (NG MVAAA 230A) using
RWP 2023-0404.
High Radiation Area and Very High Radiation Area Controls (IP Section 03.05) (5 Samples)
The inspectors evaluated licensee controls of the following high radiation areas (HRAs) and
very high radiation areas (VHRAs):
(1)
(HRA) top of containment sump (+7' elevation in the reactor containment building)
(2)
(HRA) pre-concentrator filter cubicle A/B (-35' elevation in the reactor auxiliary
building [RAB])
(3)
(HRA) purification ion exchange (IX) room A/B (-4' elevation in the RAB)
(4)
(HRA) pre-concentrator IX room A/B (-4' elevation in the RAB)
(5)
(HRA) fuel pool and chemical volume control filter cubicles and their respective hoist
pendants (-4' elevation in the RAB)
Radiation Worker Performance and Radiation Protection Technician Proficiency
(IP Section 03.06) (1 Sample)
(1)
The inspectors evaluated radiation worker and radiation protection technician
performance as it pertains to radiation protection requirements.
71124.04 - Occupational Dose Assessment
Source Term Characterization (IP Section 03.01) (1 Sample)
(1)
The inspectors evaluated licensee performance as it pertains to radioactive source
term characterization.
11
External Dosimetry (IP Section 03.02) (1 Sample)
(1)
The inspectors evaluated how the licensee processes, stores, and uses external
dosimetry.
Internal Dosimetry (IP Section 03.03) (2 Samples)
The inspectors evaluated the following internal dose assessments:
(1)
NRC Form 5 and dose assessment information for four workers, dated
October 2, 2020
(2)
NRC Form 5 and dose assessment information for one worker, dated April 18, 2022
Special Dosimetric Situations (IP Section 03.04) (2 Samples)
The inspectors evaluated the following special dosimetric situations:
(1)
NRC Form 5 and dose information for four declared pregnant workers
(2)
NRC Form 5 and assessments for four workers using effective dose equivalent
monitoring for non-uniform radiation fields
71124.08 - Radioactive Solid Waste Processing & Radioactive Material Handling, Storage, &
Transportation
Shipment Preparation (IP Section 03.04) (1 Sample)
(1)
The inspectors observed the preparation of radioactive shipment 23-1009 consisting
of two intermodal containers (ESUU200865 and ESUU200404) of dry active waste on
October 26, 2023.
OTHER ACTIVITIES - BASELINE
71151 - Performance Indicator Verification
The inspectors verified licensee performance indicators submittals listed below:
BI01: Reactor Coolant System (RCS) Specific Activity Sample (IP Section 02.10) (1 Sample)
(1)
October 1, 2022, through September 30, 2023
BI02: RCS Leak Rate Sample (IP Section 02.11) (1 Sample)
(1)
October 1, 2022, through September 30, 2023
OR01: Occupational Exposure Control Effectiveness Sample (IP Section 02.15) (1 Sample)
(1)
April 1, 2021, through June 30, 2023
PR01: Radiological Effluent Technical Specifications/Offsite Dose Calculation Manual
Radiological Effluent Occurrences Radiological Effluent Occurrences Sample (IP Section 02.16)
(1 Sample)
12
(1)
April 1, 2021, through June 30, 2023
71152A - Annual Follow-up Problem Identification and Resolution
Annual Follow-up of Selected Issues (Section 03.03) (3 Samples)
The inspectors reviewed the licensees implementation of its corrective action program
related to the following issues:
(1)
containment fan cooler train B failures on November 21, 2023
(2)
component cooling water flow deviations to high pressure safety injection pump, low
pressure safety injection pumps, and containment spray pumps on December 1, 2023
(3)
FLEX N and N+1 diesel generator starting battery failures on December 21, 2023
71152S - Semiannual Trend Problem Identification and Resolution
Semiannual Trend Review (Section 03.02) (1 Sample)
(1)
The inspectors reviewed the licensees corrective action program for potential
adverse trends in lead-acid battery performance that might be indicative of a more
significant safety issue. The inspectors observed a negative trend in performance and
longevity of flooded lead-acid battery performance. This observation is further
detailed in the results section of this report.
INSPECTION RESULTS
Unresolved Item
(Open)
Steam Generator 1 In-Situ Tube Pressure Testing Failures.
Description: The inspectors identified an unresolved item (URI) associated with the licensees
failure to meet the steam generator tube integrity performance criterion in technical
specification (TS) 6.5.9.b.1, Steam Generator Program. Specifically, Waterford 3s Steam
Generator Program structural integrity performance criterion includes retaining a safety factor
of 3.0 against burst under normal steady state full power operation primary to secondary
pressure differential and a safety factor of 1.4 against burst applied to the design basis
accident primary to secondary pressure differentials. The licensee extended the inspection
interval for the tube inspections from three cycles to four based on NRC approval of
TSTF-577 (Technical Specification Task Force), Revised Frequencies for Steam Generator
Tube Inspections and reevaluation of the refueling outage 21 (2017) operational
assessment. During the Unit 3 refueling outage 25 four tubes failed to meet the condition
monitoring criteria.
Technical specification 6.5.9, Steam Generator Program, requires that a program be
established and implemented to ensure that steam generator tube integrity is maintained.
Pursuant to TS 6.5.9, tube integrity is maintained when the steam generator performance
criteria are met. There are three steam generator performance criteria: structural integrity,
accident induced leakage, and operational leakage. Meeting the steam generator
performance criteria provides reasonable assurance of maintaining tube integrity at normal
and accident condition. TS 6.5.9 also states that the Steam Generator Program shall include
provisions for steam generator tube plugging criteria. Tubes found by in-service inspection to
contain flaws with a depth equal to or exceeding 40 percent on the nominal tube wall
13
thickness shall be plugged.
In steam generator 1, there were four tubes identified as having flaws that exceeded the
condition monitoring structural limit at the tube support plates. Eddy current testing and sizing
was performed, and the structural equivalent flaw parameters were calculated. The structural
equivalent parameters were compared to the condition monitoring limit curve and determined
that deficiencies existed. Since the tube performance criteria were not met analytically, in-situ
pressure testing of the four tubes was required. Other than the four tubes in-situ pressure
tested, all other tubes satisfied performance criteria analytically. In steam generator 2, the
tube performance criteria were satisfied analytically.
Two of the four tubes in-situ pressure tested in steam generator 1 failed to meet Structural
Integrity Performance Criterion. The examination results were also used, together with outage
repairs (i.e., tube plugging), to demonstrate that the performance criteria would be met for
upcoming cycles 26 through 27.
Upon completion of the tube examinations of pre and post pressure testing, +Point and Array
probe data confirmed that flaws in steam generator 1, tube Row 1 (R1) - Column 4 (C4) and
in tube R2-C35 had failed and burst. The inspectors reviewed condition report
CR-WF3-2023-17005 which provides additional information and a causal evaluation.
The event was reported as an 8-hour, non-emergency notification per 10CFR72(b)(3)(ii)9A)
as a degraded condition for not meeting the performance criteria for steam generator
structural integrity in accordance with TS 6.5.9.b.1, Steam Generator Program, due to two
tube failures in steam generator 1. Event notification56834 was reported to NRC
operations center on November 5, 2023.
The licensees apparent cause analysis and EN 56834 identified that the vendor used
non-conservation assumptions in the revised operational assessment to extend the
inspection interval. Additional inspection is required to determine if there is a performance
deficiency associated with this issue.
Planned Closure Actions: The NRC staff will review the available information, including a
pending vendor causal evaluation, to determine if any performance deficiencies exist and
identify any possible regulatory outcomes.
Licensee Actions: The licensee has placed the information into their corrective action
program and will have the document reviews and corrective actions developed in
January 2024.
Corrective Action References: Condition Reports CR-WF3-2023-17220 and
Failure to Follow ALARA Planning and Control Procedures Resulting in Unplanned Dose
Cornerstone
Significance
Cross-Cutting
Aspect
Report
Section
Occupational
Radiation Safety
Green
Open/Closed
[H.4] -
Teamwork
14
The inspectors identified a Green finding associated with a non-cited violation (NCV) of
Technical Specification (TS) 6.8.1.a for a failure to follow as low as reasonably achievable
(ALARA) planning and control procedures during the refueling outage 24 (2022). Specifically,
the licensee's planning or radiological controls did not prevent unplanned dose for two
separate work activities conducted during the refueling outage.
Description: During refueling outage 24 (2022), the licensee performed work activities under
RWP 2022-0512, "1RE24 Steam Generator 1 and 2 Feedring Mod," and RWP 2022-0615,
"1RE24 Remove/Replace Pressurizer Heater." The accrued dose for each of these activities
exceeded the planned dose estimate by 64 percent due to issues that NRC deemed were
preventable or reasonably foreseeable. This presented two examples for a failure to follow
ALARA planning and control procedures. These issues involved changing radiological
conditions, delays in staging materials needed for work, inaccurate person-hours from various
teams, and uncoordinated resources.
The first example is relative to RWP 2022-0512, "1RE24 Steam Generator 1 and 2 Feedring
Mod," revision 2, which addressed the radiological work with the steam generators and the
feedring. While conducting this work, multiple issues occurred. The steam generator design
for installation was a new design to the site. The new design had higher u-tubes relative to
the feedwater injection area than the previous steam generators installed. The licensee
determined that the new design required more shielding for the foreign object search and
retrieval activities, so they added more magnetic tungsten shielding. However, this additional
shielding was not as effective as planned relative to the body positioning of the workers in
that the licensee did not account for the larger plane source of the steam generators. The
workers were exposed to higher than planned levels of radiation resulting in the additional
dose. The actual dose for this task significantly increased the planned dose due to various
issues identified during the review of the work activity.
Some of these issues were:
There was difficulty with the torquing of bolts, in which multiple bolts were
over-torqued and had to be addressed. The NRC deemed this as a human
performance error and therefore preventable. NRC gave no credit for this additional
dose.
There were delays in staging material due to improper planning for the needed
resources. For instance, the polar crane hook was unavailable when needed to stage
materials. NRC deemed this as a human performance error and preventable. NRC
gave no credit for this additional dose.
Teams involved with the work activity underestimated activities and resources
needed. For instance, the project team underestimated resources needed to assist
the containment coordinator. NRC deemed this as a human performance error and
preventable. NRC gave no credit for this additional dose.
The licensee used surveys from mockup activities during the pre-outage phase and
subsequently, the radiological levels increased. However, the licensee failed to
confirm the new radiological conditions and properly address the changing
radiological conditions in their planning phase prior to work. NRC deemed this as a
human performance error and preventable. NRC gave no credit for this additional
dose.
During the job, the licensee experienced retrieval of foreign material on the secondary
side of the steam generators. NRC determined this was an emergent issue that was
not preventable or foreseeable. NRC gave additional dose credit in the amount of
208 millirem.
15
Based on the above information reviewed, the NRC determined that an additional
208 millirem may be added to the licensee's initial dose estimate of 3.656 rem, resulting in a
new NRC revised dose estimate of 3.864 rem. When comparing this to the actual accrued
dose of the RWP (6.352 rem), NRC determined that the actual collective dose exceeded the
revised dose estimate by approximately 64 percent.
The second example is relative to RWP 2022-0615, "1RE24 Remove/Replace Pressurizer
Heater," revision 6, which addressed the radiological work to remove and replace the
pressurizer heaters. While conducting this work, the licensee had trouble in various aspects
of the activity. The three primary issues involved: (1) higher dose rates on the instrument
lines requiring more shielding, (2) removing the packaging of the new heater equipment, and
(3) more time needed to remove the heaters due to issues with the type of respiratory
equipment used.
The details of these three issues included:
The licensee surveyed the instrument lines prior to work and identified additional
shielding was needed to protect workers from unintended dose. NRC gave additional
dose credit in the amount of 132 millirem for adding shielding for this activity.
The licensee had difficulty removing the type of packaging used on the new heater
equipment, which seemed to have crystallized, and there was also wire meshing that
proved difficult to remove. NRC gave additional dose credit in the amount of 372
millirem.
The licensee chose to use a Pureflo respirator hood, described as a loose-fitting,
all-in-one powered air purifying respirator (PAPR). Workers experienced fogging of
these PAPRs that slowed down work significantly. However, NRC determined that the
time estimate used for removal of each heater was inadequate and underestimated.
The licensee estimated the removal of ten heaters at approximately 36 minutes per
heater but needed about 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> per heater. Partial credit for additional dose was
given due to unforeseen conditions of the PAPRs fogging resulting in slower work
performance, but not for the general underestimation of man-hours needed for each
heater removal. As a result, NRC gave additional dose credit in the amount of 724
millirem. NRC added this additional dose to the initial dose estimate, which generally
allowed about one additional man-hour for the removal of each heater.
Based on the above information reviewed, the NRC determined that a total of 1.228 rem
(132 millirem + 372 millirem + 724 millirem) in additional dose may be credited to the
licensee's initial dose estimate of 3.262 rem, resulting in a new NRC revised dose estimate of
4.490 rem. When comparing this to the actual accrued dose of the RWP (7.383 rem), NRC
determined that the actual collective dose exceeded the revised dose estimate by
approximately 64 percent.
As the inspectors reviewed the ALARA procedure, EN-RP-110, Step 4.0[8], the following
steps were not consistently followed in RWPs 2022-0512 and 2022-0615:
[8] Planning and Scheduling / Outage Groups: Responsibilities include the following:
Providing accurate worksite person-hours and accurate work locations for ALARA
Planning purposes.
16
o
In NRC's review, the worksite person hours for removal and replacement of
the pressurizer heaters and the steam generator activities were not accurate
for planning purposes to maintain doses ALARA
Providing detailed work plans to allow for ALARA Planning to designate adequate
radiological controls.
o
During NRC's review, in RWP 2022-0615, there were no written plans for
sequence and steps of the pressurizer heater removal. Poor planning resulted
in not maintaining doses ALARA. In RWP 2022-0512, the ALARA planning
phase did not account for the larger plane source of the new steam generator
design resulting in challenges with radiological exposures. Also, in the
planning of this RWP, the surveys used were from the mockup during the pre-
outage phase. When the radiological conditions changed, the licensee failed to
adjust the planned dose estimate to account for the higher dose rates during
the outage.
Coordinating scheduling of work with radiation protection (RP) personnel to assure
work is completed in a manner and sequence that supports the ALARA Program.
o
In NRC's review, there were examples of licensee groups not coordinating
activities, such as delays in staging material needed to conduct the work and
informal work plans. Activities and resources needed for work within the RCA
were not coordinated and accounted for appropriately. In the post-outage
review of RWP 2022-0512, the licensee deemed the delays in
staging/de-staging as the largest percentage of unproductive RWP person-
hours. In the post-outage review of RWP 2022-0615, the licensee stated that
the RP technicians supporting the activity did not have good firsthand
knowledge of the project scope and equipment being used which challenged
effective team building. Also, informal discussions between the project team
and RP staff for removal of the pressurizer heaters, in RWP 2022-0615,
resulted in uncertainty regarding the sequence and steps of execution.
Therefore, NRC determined that multiple procedural steps were missed during the planning
of these two work activities, RWP 2022-0512 and RWP 2022-0615, which resulted in
unplanned dose to workers and challenging ALARA principles.
Corrective Actions: The licensee addressed the deficiencies identified during the work activity
in their ALARA package post-job reviews. They also documented the failure to maintain
doses ALARA for these work activities in a new condition report for assessment of applicable
corrective actions.
Corrective Action References: CR-WF3-2023-16870
Performance Assessment:
Performance Deficiency: The licensee failed to follow ALARA planning procedures and did
not properly plan the scope of work activities.
Screening: The inspectors determined the performance deficiency was more than minor
because it was associated with the Program & Process attribute of the Occupational
Radiation Safety cornerstone and adversely affected the cornerstone objective to ensure the
adequate protection of the worker health and safety from exposure to radiation from
17
radioactive material during routine civilian nuclear reactor operation. Additionally, the finding
was similar to Example 6(i) in Appendix E to Inspection Manual Chapter 0612, Power
Reactor Inspection Reports - Examples of Minor Issues. This example states that an issue is
more than minor if it results in a collective dose greater than 5 person-Rem, and the actual
dose accrued exceeds the estimated dose by greater than 50 percent. Specifically, the actual
dose accrued for each work activity exceeded 5 rem and both exceeded the revised dose
estimate, as determined by the NRC, by 64 percent.
Significance: The inspectors assessed the significance of the finding using IMC 0609
Appendix C, Occupational Radiation Safety SDP. The inspectors determined the finding had
very low safety significance (Green) because: (1) it was associated with ALARA planning and
work controls; and (2) the licensees latest 3-year rolling average collective dose was less
than 135 person-Rem.
Cross-Cutting Aspect: H.4 - Teamwork: Individuals and work groups communicate and
coordinate their activities within and across organizational boundaries to ensure nuclear
safety is maintained. Specifically, the licensee failed to implement the process of planning
work activities with proper communication and coordination from each workgroup involved to
include person-hour estimates, resources, and formal work steps needed for the job activities.
This resulted in delays in staging material needed, inaccurate person-hours needed to
perform work activities, and uncoordinated resources needed for work activities.
Enforcement:
Violation: Technical Specification 6.8.1.a requires, in part, that written procedures shall be
established, implemented, and maintained covering the procedures recommended in
Regulatory Guide 1.33, Appendix A, Revision 2, dated February 1978. Section 7(e) of
Appendix A requires radiation protection procedures. Licensee Procedure EN-RP-110,
ALARA Program, revision 14, described the planning and scheduling responsibilities for
outage groups, which included providing accurate work site person-hours, providing detailed
work plans to allow ALARA planning to designate adequate radiological controls, and
coordinating scheduling of work with Radiation Protection personnel to support ALARA.
Contrary to the above, during refueling outage 24 in the spring of 2022, the licensee failed to
implement their ALARA program procedures for planning and controlling two work activities.
Specifically, for two RWPs-2022-0512 and -2022-0615, the licensee failed to provide
accurate work site person-hours, failed to provide detailed work plans for the pressurizer
heater removals or account for the larger plane source of the new steam generator design,
and failed to coordinate work activities and resources resulting in delays in staging materials
and unavailable resources. This resulted in not maintaining doses ALARA for workers during
these activities.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with
Section 2.3.2 of the Enforcement Policy.
Failure to Maintain FLEX Equipment Starting Batteries
Cornerstone
Significance
Cross-Cutting
Aspect
Report
Section
Mitigating
Systems
Green
Open/Closed
[H.12] - Avoid
Complacency
18
The inspectors reviewed a self-revealed Green finding and associated NCV of 10 CFR
50.155(b)(1), which states, in part, strategies and guidelines to mitigate beyond -design-basis
events from natural phenomena must be capable of being implemented site-wide and must
include maintaining or restoring core cooling capabilities. Specifically, from approximately
February 14 to May 16, 2023, the licensee failed to ensure the starting batteries for the
FLEX N and N+1 diesel generators had sufficient capacity to perform their required
functions.
Description: As part of the licensees phase 2 strategies as required by NRC
Order EA-12-049, Order Modifying Licenses with Regard to Requirements for Mitigation
Strategies for Beyond-Design-Basis External Events, the licensee committed to the guidance
described in NEI 12-06, Diverse and Flexible Coping Strategies (FLEX) Implementation
Guide, revision 0. NRC Order EA-12-049 has since been codified by 10 CFR 50.155,
Mitigation of beyond-design-basis events.
Specifically for FLEX AC power supply, the licensee developed mitigating strategies that
utilize a FLEX N diesel generator as a 480V power supply that can be hooked up into a safety
bus. A FLEX N+1 diesel generator was stored outside the protected area as a backup that
can be brought into the protected area and connected into a safety bus. These two diesel
generators are the only dedicated means of providing 480V power for a beyond-design-basis
station blackout event. The diesel generators are started by a set of two commercial 8D
batteries for each generator.
On May 6, 2023, power was lost for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> to the FLEX N+1 building which maintains the
FLEX N+1 diesel generator starting battery charge. On May 13, 2023, the licensee was
performing weekly rounds when it was identified the control panel of the FLEX N+1 diesel
generator had no power. The capacity of the batteries was too low to restart the battery
charger to provide the float charge. The batteries would not have had the capacity to start the
FLEX N+1 diesel generator if needed. On May 14, 2023, the degraded starting batteries were
replaced with the charged and ready set of spare FLEX starting batteries.
On May 16, 2023, the licensee removed power to the FLEX N diesel generator for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for
maintenance. The power was reconnected 24 minutes later, and the licensee attempted to
start the FLEX N diesel. However, the generator failed to start due to degraded capacity of
the starting battery. In both failures, the cause was a starting battery that had degraded
capacity. Because there was a set of ready spare batteries that would be able to be changed
out in an actual event, the function of the FLEX AC power supply was not considered fully
lost. All FLEX functions could still be completed within the time allotted.
The licensee makes plans to replace the starting batteries on a 4-year frequency. No tests
are performed specifically on the batteries to ensure their capacity is adequate beyond
performing a start of the FLEX N and FLEX N+1 diesel generators every six months. Both
sets of starting batteries were purchased and installed in May 2020. There is no expected
lifetime of the battery provided by the manufacturer. The warranty on the batteries is for 6
months with a pro-rated replacement that extends until 30 months of life. As evidenced by the
failure to start of the diesel generators, the capacity of these starting batteries was degraded
beyond the ability to start the FLEX diesel generators.
The date on which the starting batteries had degraded to no longer be functional is unable to
be determined with accuracy. The degradation mechanism is not able to be identified on the
licensee weekly or monthly checks of the equipment. The previous successful surveillances
that started the FLEX N and N+1 diesel generators were on November 15, 2022. The
19
inspectors assume the degradation occurred halfway from the last successful surveillance to
when both FLEX diesel generators were repaired. This date was determined to be
February 14, 2023.
Corrective Actions: The licensee replaced the starting batteries for both FLEX N and
FLEX N+1 diesel generators. After the initial replacement, the licensee performed another
replacement with longer-lasting absorbed glass-mat batteries. Additionally, the licensee
implemented preventive maintenance to perform monthly battery load testing for the FLEX N
and FLEX N+1 diesel generator starting batteries.
Corrective Action References: CR-WF3-2023-13265, CR-WF3-2023-13293
Performance Assessment:
Performance Deficiency: The licensee failed to maintain mitigation strategies for beyond-
design basis external events.
Screening: The inspectors determined the performance deficiency was more than minor
because it was associated with the Equipment Performance attribute of the Mitigating
Systems Cornerstone and adversely affected the cornerstone objective to ensure the
availability, reliability, and capability of systems that respond to initiating events to prevent
undesirable consequences. Specifically, the licensee failed to maintain the FLEX N and
FLEX N+1 diesel generator batteries so their respective generators could start and provide
power in accordance with the licensee mitigating strategies.
Significance: The inspectors assessed the significance of the finding using IMC 0609
Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using
Exhibit 2, Mitigating Systems Screening Questions, Section E, the inspectors determined
the finding to be of very low safety significance (Green), because the performance deficiency
was associated with equipment not solely purposed for spent fuel pool instrumentation or for
containment venting, but it was associated with equipment credited in a Phase 2 FLEX
strategy such that all FLEX functions could still be completed in accordance with existing
plant procedures within the time allotted.
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 licensee failed to
recognize and plan for the possibility of starting batteries to degrade faster than their service
life and result a loss of ability of the FLEX diesels to start.
Enforcement:
Violation: 10 CFR 50.155(b)(1), states, in part, strategies and guidelines to mitigate
beyond-design-basis events from natural phenomena must be capable of being implemented
site-wide and must include maintaining or restoring core cooling capabilities.
Contrary to the above, from approximately February 14 to May 16, 2023, the licensee failed
to maintain mitigation strategies for beyond-design basis external events. Specifically, the
licensee failed to maintain the FLEX N and FLEX N+1 diesel generator batteries so their
respective generators could start and provide power in accordance with the licensee
mitigating strategies.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with
Section 2.3.2 of the Enforcement Policy.
20
Observation: Flooded Lead-Acid Battery Performance
The inspectors reviewed the licensees corrective action program for potential adverse trends
in lead-acid battery performance that might be indicative of a more significant safety issue.
The inspectors observed a negative trend in performance and longevity of flooded lead-acid
battery performance. In addition to the FLEX N and FLEX N+1 starting battery issues detailed
in the IP 71152A section, the inspectors identified five other battery failures in 2023:
CR-WF3-2023-01793: The starting batteries for the non-safety permanently-installed
temporary emergency diesel generator were degraded and unable to perform their
function.
CR-WF3-2023-14593: The starting batteries for the security diesel generator were
degraded and unable to perform their function.
CR-WF3-2023-15322: The starting battery for the portable ultimate heat sink
replenishment pump were below the required voltage.
CR-WF3-2023-15407: The starting battery for the diesel-driven dry cooling tower
sump pump was degraded and unable to perform its function.
CR-WF3-2023-15858: The starting batteries for diesel-driven fire pump A were
degraded and unable to perform their function.
These five diesel generators are considered non-safety but perform important functions
for the site. The licensee documented the NRC concern about a negative trend in
performance in CR-WF3-2023-15830 and performed an analysis of the issue. The
corrective actions included replacement of the batteries and a reconsideration of the
preventive maintenance strategies. No findings of significance were identified.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
On October 27, 2023, the inspectors presented the occupational radiation safety
inspection results to Joseph Sullivan, Site Vice President, and other members of the
licensee staff.
On November 2, 2023, the inspectors presented the radiation inspection results to
Joseph Sullivan, Site Vice President, and other members of the licensee staff.
On November 30, 2023, the inspectors presented the inservice inspection results to
Joseph Sullivan, Site Vice President, and other members of the licensee staff.
On January 24, 2024, the inspectors presented the integrated inspection results to
Joseph Sullivan, Site Vice President, and other members of the licensee staff.
21
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Engineering
Evaluations
Flood Hazard Reevaluation Report
07/21/2015
Procedures
OP-901-521
Severe Weather and Flooding
342
Corrective Action
Documents
CR-WF3-YYYY-
NNNN
2022-04265, 2022-06268, 2023-16951
Miscellaneous
W3-DBD-010
Containment Cooling HVAC and Related Systems
301
Miscellaneous
W3-DBD-011
Electrical Distribution (AC portion)
302
Procedures
OP-006-001
Plant Distribution System (7KV, 4KV, and SSD) System
346
Procedures
OP-008-003
Containment Cooling System
303
Work Orders
00580779, 00580781
Fire Plans
RAB 16-001
Emergency Diesel Generator Room 3A
12
Fire Plans
Cable Vault
11
Fire Plans
RAB 5-001
Electrical Penetration Room B
10
Fire Plans
RAB 6-001
Electrical Penetration Room A
10
Fire Plans
Switchgear Room AB
12
Fire Plans
RAB 9-001
Auxiliary Control Panel Room
9
Fire Plans
RCB-001
RCB General Area
12
Miscellaneous
W3-DBD-4
Component Cooling Water Auxiliary Component Cooling
Water Design Basis Document
307
Work Orders
52586237, 53000031
Corrective Action
Documents
CR-WF3-YYYY-
NNNNN
2022-01969, 2022-02400, 2022-02472, 2022-02656,
2022-02665, 2022-02915, 2022-03207, 2022-03855,
2022-04131, 2022-04850, 2022-05025, 2022-05227,
2022-05244, 2022-05355, 2022-08116, 2023-01326,
2023-01346, 2023-01565, 2023-16490, 2023-16753,
2023-16755, 2023-16971, 2023-91568, 2023-01568,
2023-01632
Corrective Action
Documents
Resulting from
Inspection
CR-WF3-YYYY-
NNNNN
2023-16714, 2023-16720, 2023-16883, 2023-16938,
2023-16966, 2023-16971, 2023-16985, 2023-16990,
2023-17005, 2023-17042, 2023-17043, 2023-17044,
2023-17058, 2023-17070, 2023-17219, 2023-17220,
2023-17259, 2023-17278, 2023-376
22
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Drawings
S/G 32 Hot Primary Face, Hardware Repair Status Pre-
3R25-10/23 - S/G 32 Hot Leg
08/11/2023
Drawings
02-9367763-E-
000
S/G 31 Cold Primary Face, Hardware Repair Status Pre-
3R25-10/23 - S/G 31 Cold Leg
08/11/2023
Drawings
02-9367764-E-
000
S/G 31 Hot Primary Face, Hardware Repair Status Pre-
3R25-10/23 - S/G 31 Hot Leg
08/11/2023
Drawings
02-9367765-E-
000
S/G 32 Cold Primary Face, Hardware Repair Status Pre-
3R25-10/23 - S/G 32 Cold Leg
08/11/2023
Drawings
6660E03
Replacement Steam Generator Waterford 3 Water Level Vs.
Span
2
Drawings
H33760-1201,
Sheet 1 of 4
Rosemount Engineering Company, Certified Configuration
Drawing - Sensor, Temperature Platinum Resistance Type
C
Engineering
Changes
Waterford 3 - Steam Generator Strategic Plan Document
Plan Per EN-DC-317, Para 7.13
000
Engineering
Changes
ASME Section XI VT-3 examination of rigid strut support
FWRR-0017 under WO-554302
Miscellaneous
Certificate of
Calibration No.
20846-502
Parker Research Corporation, TB-10 Magnetic Weight Lift
Test Bar
04/12/2007
Miscellaneous
LA191736
SOCOTEC WF3 Feedwater Piping Monitoring for RSG Flow
Diverter Modification
001
Miscellaneous
PQR 344
Procedure Qualification Record, Manual Gas tungsten Arc
Welding (GTAW)
1
Miscellaneous
PQR 456
Procedure Qualification Record, Manual Gas Tungsten &
Shielded Metal Arc Welding (GTAW and SMAW)
0
Miscellaneous
WPS-NI-43/43-B
Manual Gas Tungsten Arc Welding (GTWA) of P-No.43
nickel alloys, in all positions, for all joint types, fillets and
repairs using F-No. 43 filler metal, without Postweld Heat
Treatment (PWHT).
0
NDE Reports
BOP-PT-23-069
1B Cold Leg Thermowell, Component ID: RCI TE0112 CD1
11/04/2023
NDE Reports
PT-VT-22-031
Bolted Connection RC MRCT0001 (RV Studs)
04/15/2022
NDE Reports
PT-VT-22-039
S/G System - RCB/Outside D-Rings
06/24/2022
NDE Reports
W-CISI-VT 22-
002
Inner Moisture Between Col. 19 and Col. 21 (Approx.)
04/27/2022
23
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
NDE Reports
W-CISI-VT22-007
Inner Moisture Barrier Between Col. 11 and Col. 13
(Approx.)
04/27/2022
NDE Reports
W-CISI-VT22-013
Moisture Barrier Inside Annulus 0 degrees to 103 degrees
azimuth.
04/27/2022
NDE Reports
W-CISI-VT22-014
Moisture Barrier Inside Annulus 103 degrees to 256 degrees
Azimuth
04/27/2022
NDE Reports
W-CISI-VT22-015
Moisture Barrier Inside Annulus 256 degrees to 360 degrees
Azimuth
04/27/2022
NDE Reports
W-ISI-VT-22-009
ASME Section XI VT-3 examination of rigid strut support
FWRR-0017 under WO-554302. A loose lock nut was not in
the proper location according to design drawing FWRR-117
SH 1 of 3 and the Bergen-Paterson Pipe Support Corp.
10/30/2023
Procedures
CEP-BAC-001
Boric Acid Corrosion Control (BACC) Program Plan
2
Procedures
CEP-NDE-0400
Ultrasonic Examination
9
Procedures
CEP-NDE-0404
Manual Ultrasonic Examination of Ferritic Piping Welds
(ASME XI)
9
Procedures
CEP-NDE-0407
Straight Beam Ultrasonic Examination of Bolts and Studs
(ASME XI)
6
Procedures
CEP-NDE-0423
Manual Ultrasonic Examination of Austenitic Piping Welds
(ASME XI)
9
Procedures
CEP-NDE-0641
Liquid Penetrant Examination (PT) for ASME Section XI
10
Procedures
CEP-NDE-0731
Magnetic Particle Examination (MT) for ASME Section XI
7
Procedures
CEP-NDE-0901
VT-1 Examination
6
Procedures
CEP-NDE-0902
VT-2 Examination
10
Procedures
CEP-NDE-0903
VT-3 Examination
8
Procedures
CEP-NDE-0965
Visual Welding Inspection ASME, ANSI B31-1
7
Procedures
CEP-PT-0001
ASME Section XI Pressure Test (PT) Program
313
Procedures
CEP-RR-001
ASME Section XI Repair/Replacement Program
320
Procedures
CEP-SG-002
Steam Generator Secondary Side Examinations and
Maintenance
5
Procedures
CEP-WP-GWS-1
General Welding Standard ASME/ANSI
8
Procedures
Boric Acid Corrosion Control Program (BACCP)
13
Procedures
Entergy Nuclear Welding Program
008
Procedures
Entergy Repair/Replacement Program
004
24
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Procedures
Inservice Inspection Program Duties and Responsibilities
007
Procedures
Corrective action Program
049
Procedures
SEP-BAC-WF3-
001
Waterford 3 Boric Acid Corrosion Control Program (BACCP)
Program Section
4
Procedures
SEP-ISI-104
Program Section For ASME Section XI, Division 1 WF3
Inservice Inspection Program
14
Procedures
SEP-ISI-104
Program Section for ASME Section X, Division 1 WF3
Inservice Inspection Program
14
Procedures
SEP-PT-WF3-
001
Waterford 3 Inservice Inspection Pressure Testing (PT)
Program Section
001
Procedures
SEP-SG-WF3-
001
Waterford -3 (W3/WF3) Steam Generator Program
4
Self-Assessments LO-HQNLO-
2021-19
2022 Welding Program Assessment
02/17/2022
Self-Assessments LO-WLO-2022-
0060-CA
Pre-NRC RF25 ISI Activities Self -Assessment Report
08/08/2023
Self-Assessments LO-WLO-2022-
0060-CA-3
Pre-NRC RF25 ISI Activities Self-Assessment Report
08/08/2023
Work Orders
WO No.
572188-24, 589604-15
Procedures
Conduct of Operations
31
Procedures
OP-010-005
Plant Shutdown
345
Procedures
OP-901-311
Loss of Train B Safety Bus
313
Procedures
OP-901-521
Severe Weather and Flooding
343
Procedures
OP-902-001
Reactor Trip Recovery
21
Procedures
OP-902-003
Loss of Offsite Power / Loss of Forced Circulation Recovery
11
Corrective Action
Documents
CR-WF3-YYYY-
NNNN
2022-06818, 2023-01910, 2023-01944, 2023-13294,
2023-13313, 2023-13331, 2023-14317, 2023-14967,
2023-16596, 2023-13943, 2023-14310, 2023-14314
Corrective Action
Documents
Resulting from
Inspection
CR-WF3-YYYY-
NNNN
2024-00169
01/10/2024
Engineering
Changes
Engineering Change
09/14/2023
25
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Miscellaneous
TD G080.0095
General Electric Switchgear Magne Blast Breakers
6
Procedures
Maintenance Rule Monitoring
9
Procedures
ME-004-115
4.16/6.9 kV G.E. Magne-Blast Breaker Overhaul
6, 9
Procedures
OP-903-094
ESFAS Subgroup Relay Test - Operating
35
Work Orders
00517244, 00586519, 52790255, 52805142, 54034973,
54038818
Corrective Action
Documents
CR-WF3-YYYY-
NNNN
2023-16283, 2023-16372, 2023-16376, 2023-17876,
2023-15594
Engineering
Changes
Engineering Change
0
Procedures
OP-009-005
45
Procedures
OP-901-511
Instrument Air Malfunction
20
Corrective Action
Documents
CR-WF3-YYYY-
NNNN
2023-17399
Corrective Action
Documents
CR-WF3-YYYY-
NNNNN
2019-01293, 2023-18027, 2017-03359, 2017-04081,
2018-00948,
Engineering
Changes
ACC-127B Input to Operability CR-23-18244/18245
12/21/2023
Engineering
Changes
Use of instrumentation for ACCW System Flow Balance
PE-004-024
05/12/2017
Procedures
FSG-005
Initial Assessment and FLEX Equipment Staging
15
Procedures
OP-903-003
Charging Pump Operability Check
315
Procedures
OP-903-052
Controlled Ventilation Area System Operability Check
15
Procedures
OP-903-096
Boron Flowrate Verification
11
Procedures
OP-903-115
Train A Integrated Emergency Diesel
59
Procedures
OP-903-121
Safety Systems Quarterly IST Valve Tests
36
Procedures
PE-004-024
ACCW & CCW System Flow Balance
310
Procedures
STA-001-004
Local Leak Rate Test (LLRT)
320
Work Orders
53013043, 54002710, 00586332, 53017375, 54067505,
54085552, 00474102, 00495521, 00502714, 00517264,
00518612
ALARA Plans
RWP 2022-0512
1RE24 Steam Generator 1 and 2 Feedring Mod
2
ALARA Plans
RWP 2022-0615
1RE24 Remove/Replace Pressurizer Heater
6
26
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Corrective Action
Documents
CR-WF3-YYYY-
XXXXX
2022-01953, 2023-00421, 2022-03390, 2022-06963,
2023-00518, 2023-01766, 2023-01234, 2022-02542,
2022-07912, 2023-00114, 2023-16348, 2023-16474,
Corrective Action
Documents
Resulting from
Inspection
CR-WF3-YYYY-
XXXXX
2023-16870, 2023-16872, 2023-16893
Procedures
Radiation Worker Expectations
14
Procedures
Access Control for Radiologically Controlled Areas
17
Procedures
Radiological Control
008
Procedures
ALARA Program
14
Procedures
Radioactive Material Control
19
Procedures
EN-RP-141-01
Job Coverage Using Remote Monitoring Technology
8
Procedures
Conduct of Radiation Protection
008
Procedures
HPI-001-123
Plant Conditions and Radiological Concerns
010
Radiation
Surveys
RAB -4 Purification Ion Exchangers
01/24/2023
Radiation
Surveys
RAB -35 Spent Resin Tank Pump Room / Waste
Condensate IX
08/22/2023
Radiation
Surveys
RAB -35 Boric Acid Pre-Concentrator Filters
09/14/2023
Radiation
Surveys
FHB +46 Fuel Handling Area
09/18/2023
Radiation
Surveys
RAB -4 Center Wing
09/23/2023
Radiation
Surveys
Radwaste Solidification Building
09/26/2023
Radiation
Surveys
RAB -4 Flash Tank / Purification Filter Area
10/05/2023
Radiation Work
Permits (RWPs)
2022-0623
REFUEL 24 - Perform miscellaneous contaminated system
valve work in the Regen Hx Room including all support
activities, troubleshooting, walkdowns, tagouts, tours and
inspections.
01
Radiation Work
2022-0641
REFUEL 24 - Emergent Dose added Inside the Reactor
00
27
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Permits (RWPs)
Containment Building.
Radiation Work
Permits (RWPs)
2023-0404
REFUEL 25 - Plant Maintenance Valve Work on
Contaminated and Clean System Valves outside the Reactor
Containment Building.
00
Radiation Work
Permits (RWPs)
2023-0702
REFUEL 25 - Disassembly of Reactor Head and All
Associated Work Activities.
02
Radiation Work
Permits (RWPs)
2023-0714
REFUEL 25 - Cleaning of the Reactor Head Surface.
Includes all supporting activities and Bare Metal Inspections.
00
Self-Assessments LO-WLO-2022-
0051 CA-00004
Radiological Hazard Assessment and Exposure Controls
08/21/2023
Corrective Action
Documents
CR-WF3-YYYY-
NNNN
2020-01981, 2020-02198, 2020-03232, 2020-07014,
2021-00302, 2021-02028, 2022-01780, 2022-01921,
2022-03253, 2022-07004, 2023-01604, 2023-15043,
2023-16119
Miscellaneous
Evaluation of DLR/SRD Discrepancies and DLRs Not
Returned for Processing
06/30/2020
Miscellaneous
Evaluation of DLR/SRD Discrepancies and DLRs Not
Returned for Processing
07/10/2022
Miscellaneous
NRC Annual Dose Report (REIRS)
2022
Miscellaneous
15403
Dose Assessment from PCE
10/02/2020
Miscellaneous
56286
Dose Assessment from PCE
10/02/2020
Miscellaneous
57700
Dose Assessment from PCE
10/02/2020
Miscellaneous
64832
Dose Assessment from PCE
10/02/2020
Miscellaneous
92905
Dose Assessment from PCE
04/18/2022
Procedures
Alpha Monitoring
10
Procedures
Dosimetry Administration
5
Procedures
Dose Assessment
10
Procedures
Special Monitoring Requirements
11
Procedures
EN-RP-204-01
Effective Dose Equivalent (EDEX) Monitoring
3
Procedures
Prenatal Monitoring
5
Procedures
Dosimeter of Legal Record Quality Assurance
7
Procedures
Whole Body Counting/In-Vitro Bioassay
7
Self-Assessments LO-WLO-2022-
00051
Occupational Dose Assessment
10/05/2023
28
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Self-Assessments QA-14/15-2021-
W3-01
Quality Assurance Audit: Combined Radiation Protection
and Radwaste
10/25/2021
Shipping Records
RSN#: 23-1009
Shipment consisting of two 20-foot intermodal containers
(ESUU200404 and ESUU200865) of dry active waste,
UN2912, radioactive material, low specific activity (LSA-I)
10/26/2023
Corrective Action
Documents
CR-WF3-YYYY-
NNNN
2022-01874, 2022-03111, 2022-06393, 2022-06647,
2022-06852, 2023-15179, 2023-15245, 2023-16237
Corrective Action
Documents
Resulting from
Inspection
CR-WF3-YYYY-
NNNN
2023-14746, 2023-14747, 2023-14895, 2023-15933,
2023-15424
Work Orders
53005507, 53022055, 53022177, 53005391, 54003998
Corrective Action
Documents
CR-WF3-YYYY-
NNNN
2023-01793, 2023-01911, 2023-14593, 2023-15322,
2023-15407, 2023-15858, 2023-16043
Corrective Action
Documents
Resulting from
Inspection
CR-WF3-YYYY-
NNNN
2023-15830