ML24039A199
| ML24039A199 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 02/12/2024 |
| From: | John Dixon NRC/RGN-IV/DORS/PBD |
| To: | Sullivan J Entergy Operations |
| References | |
| IR 2023004 | |
| Download: ML24039A199 (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 - 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.
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.February 12, 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 02/12/24
ADAMS: ML24039A199
OFFICE SRI:DRP/D RI:DRP/D SPE:DORS/D BC:DORS/D
NAME APatz AChilds ASanchez JDixon
DATE 02/09/2024 02/09/2024 02/09/2024 2/12/2024
U.S. NUCLEAR REGULATORY COMMISSION
Inspection Report
Docket Number: 05000382
License Number: NPF-38
Report Number: 05000382/2023004
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
Enclosure
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 Report
Aspect Section
Occupational Green [H.4] - 71124.01
Radiation Safety NCV 05000382/2023004-02 Teamwork
Open/Closed
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 Report
Aspect Section
Mitigating Green [H.12] - Avoid 71152A
Systems NCV 05000382/2023004-03 Complacency
Open/Closed
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
URI 05000382/2023004-01 Steam Generator 1 In-Situ 71111.08P Open
Tube Pressure Testing
Failures.
2
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.
3
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:
4
(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
No. W-ISI-UT-23-011
No. W-ISI-UT-23-015
No. W-ISI-UT-23-013
No. W-ISI-UT-23-014
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
5
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
through a review of the following evaluations:
(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,
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
6
PWR Inservice Inspection Activities Sample - Steam Generator Tube Inspection Activities
(Section 03.04)
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.
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.
7
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)
(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.
8
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)
(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.
9
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.
External Dosimetry (IP Section 03.02) (1 Sample)
(1) The inspectors evaluated how the licensee processes, stores, and uses external
dosimetry.
10
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)
(1) April 1, 2021, through June 30, 2023
11
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 Steam Generator 1 In-Situ Tube Pressure Testing Failures. 71111.08P
(Open) URI 05000382/2023004-01
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
thickness shall be plugged.
12
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 notification (EN) 56834 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 Report
Aspect Section
Occupational Green [H.4] - 71124.01
Radiation Safety NCV 05000382/2023004-02 Teamwork
Open/Closed
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,
13
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.
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
14
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.
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
15
- 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
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
16
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 Report
Aspect Section
Mitigating Green [H.12] - Avoid 71152A
Systems NCV 05000382/2023004-03 Complacency
Open/Closed
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
17
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
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.
18
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.
19
Observation: Flooded Lead-Acid Battery Performance 71152S
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.
20
DOCUMENTS REVIEWED
Inspection Type Designation Description or Title Revision or
Procedure Date
71111.01 Engineering W3F1-2015-0042 Flood Hazard Reevaluation Report 07/21/2015
Evaluations
71111.01 Procedures OP-901-521 Severe Weather and Flooding 342
71111.04 Corrective Action CR-WF3-YYYY- 2022-04265, 2022-06268, 2023-16951
Documents NNNN
71111.04 Miscellaneous W3-DBD-010 Containment Cooling HVAC and Related Systems 301
71111.04 Miscellaneous W3-DBD-011 Electrical Distribution (AC portion) 302
71111.04 Procedures OP-006-001 Plant Distribution System (7KV, 4KV, and SSD) System 346
71111.04 Procedures OP-008-003 Containment Cooling System 303
71111.04 Work Orders 00580779, 00580781
71111.05 Fire Plans RAB 16-001 Emergency Diesel Generator Room 3A 12
71111.05 Fire Plans RAB 1E-001 Cable Vault 11
71111.05 Fire Plans RAB 5-001 Electrical Penetration Room B 10
71111.05 Fire Plans RAB 6-001 Electrical Penetration Room A 10
71111.05 Fire Plans RAB 8C-001 Switchgear Room AB 12
71111.05 Fire Plans RAB 9-001 Auxiliary Control Panel Room 9
71111.05 Fire Plans RCB-001 RCB General Area 12
71111.07A Miscellaneous W3-DBD-4 Component Cooling Water Auxiliary Component Cooling 307
Water Design Basis Document
71111.07A Work Orders 52586237, 53000031
71111.08P Corrective Action CR-WF3-YYYY- 2022-01969, 2022-02400, 2022-02472, 2022-02656,
Documents NNNNN 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
71111.08P Corrective Action CR-WF3-YYYY- 2023-16714, 2023-16720, 2023-16883, 2023-16938,
Documents NNNNN 2023-16966, 2023-16971, 2023-16985, 2023-16990,
Resulting from 2023-17005, 2023-17042, 2023-17043, 2023-17044,
Inspection 2023-17058, 2023-17070, 2023-17219, 2023-17220,
2023-17259, 2023-17278, 2023-376
21
Inspection Type Designation Description or Title Revision or
Procedure Date
71111.08P Drawings S/G 32 Hot Primary Face, Hardware Repair Status Pre- 08/11/2023
3R25-10/23 - S/G 32 Hot Leg
71111.08P Drawings 02-9367763-E- S/G 31 Cold Primary Face, Hardware Repair Status Pre- 08/11/2023
000 3R25-10/23 - S/G 31 Cold Leg
71111.08P Drawings 02-9367764-E- S/G 31 Hot Primary Face, Hardware Repair Status Pre- 08/11/2023
000 3R25-10/23 - S/G 31 Hot Leg
71111.08P Drawings 02-9367765-E- S/G 32 Cold Primary Face, Hardware Repair Status Pre- 08/11/2023
000 3R25-10/23 - S/G 32 Cold Leg
71111.08P Drawings 6660E03 Replacement Steam Generator Waterford 3 Water Level Vs. 2
Span
71111.08P Drawings H33760-1201, Rosemount Engineering Company, Certified Configuration C
Sheet 1 of 4 Drawing - Sensor, Temperature Platinum Resistance Type
71111.08P Engineering EC# 0000084109 Waterford 3 - Steam Generator Strategic Plan Document 000
Changes Plan Per EN-DC-317, Para 7.13
71111.08P Engineering EC-0054070627 ASME Section XI VT-3 examination of rigid strut support
Changes FWRR-0017 under WO-554302
71111.08P Miscellaneous Certificate of Parker Research Corporation, TB-10 Magnetic Weight Lift 04/12/2007
Calibration No. Test Bar
20846-502
71111.08P Miscellaneous LA191736 SOCOTEC WF3 Feedwater Piping Monitoring for RSG Flow 001
Diverter Modification
71111.08P Miscellaneous PQR 344 Procedure Qualification Record, Manual Gas tungsten Arc 1
Welding (GTAW)
71111.08P Miscellaneous PQR 456 Procedure Qualification Record, Manual Gas Tungsten & 0
Shielded Metal Arc Welding (GTAW and SMAW)
71111.08P Miscellaneous WPS-NI-43/43-B Manual Gas Tungsten Arc Welding (GTWA) of P-No.43 0
nickel alloys, in all positions, for all joint types, fillets and
repairs using F-No. 43 filler metal, without Postweld Heat
Treatment (PWHT).
71111.08P NDE Reports BOP-PT-23-069 1B Cold Leg Thermowell, Component ID: RCI TE0112 CD1 11/04/2023
71111.08P NDE Reports PT-VT-22-031 Bolted Connection RC MRCT0001 (RV Studs) 04/15/2022
71111.08P NDE Reports PT-VT-22-039 S/G System - RCB/Outside D-Rings 06/24/2022
71111.08P NDE Reports W-CISI-VT 22- Inner Moisture Between Col. 19 and Col. 21 (Approx.) 04/27/2022
002
22
Inspection Type Designation Description or Title Revision or
Procedure Date
71111.08P NDE Reports W-CISI-VT22-007 Inner Moisture Barrier Between Col. 11 and Col. 13 04/27/2022
(Approx.)
71111.08P NDE Reports W-CISI-VT22-013 Moisture Barrier Inside Annulus 0 degrees to 103 degrees 04/27/2022
azimuth.
71111.08P NDE Reports W-CISI-VT22-014 Moisture Barrier Inside Annulus 103 degrees to 256 degrees 04/27/2022
Azimuth
71111.08P NDE Reports W-CISI-VT22-015 Moisture Barrier Inside Annulus 256 degrees to 360 degrees 04/27/2022
Azimuth
71111.08P NDE Reports W-ISI-VT-22-009 ASME Section XI VT-3 examination of rigid strut support 10/30/2023
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.
71111.08P Procedures CEP-BAC-001 Boric Acid Corrosion Control (BACC) Program Plan 2
71111.08P Procedures CEP-NDE-0400 Ultrasonic Examination 9
71111.08P Procedures CEP-NDE-0404 Manual Ultrasonic Examination of Ferritic Piping Welds 9
(ASME XI)
71111.08P Procedures CEP-NDE-0407 Straight Beam Ultrasonic Examination of Bolts and Studs 6
(ASME XI)
71111.08P Procedures CEP-NDE-0423 Manual Ultrasonic Examination of Austenitic Piping Welds 9
(ASME XI)
71111.08P Procedures CEP-NDE-0641 Liquid Penetrant Examination (PT) for ASME Section XI 10
71111.08P Procedures CEP-NDE-0731 Magnetic Particle Examination (MT) for ASME Section XI 7
71111.08P Procedures CEP-NDE-0901 VT-1 Examination 6
71111.08P Procedures CEP-NDE-0902 VT-2 Examination 10
71111.08P Procedures CEP-NDE-0903 VT-3 Examination 8
71111.08P Procedures CEP-NDE-0965 Visual Welding Inspection ASME, ANSI B31-1 7
71111.08P Procedures CEP-PT-0001 ASME Section XI Pressure Test (PT) Program 313
71111.08P Procedures CEP-RR-001 ASME Section XI Repair/Replacement Program 320
71111.08P Procedures CEP-SG-002 Steam Generator Secondary Side Examinations and 5
Maintenance
71111.08P Procedures CEP-WP-GWS-1 General Welding Standard ASME/ANSI 8
71111.08P Procedures EN-DC-319 Boric Acid Corrosion Control Program (BACCP) 13
71111.08P Procedures EN-DC-328 Entergy Nuclear Welding Program 008
71111.08P Procedures EN-DC-342 Entergy Repair/Replacement Program 004
23
Inspection Type Designation Description or Title Revision or
Procedure Date
71111.08P Procedures EN-DC-351 Inservice Inspection Program Duties and Responsibilities007
71111.08P Procedures EN-LI-102 Corrective action Program 049
71111.08P Procedures SEP-BAC-WF3- Waterford 3 Boric Acid Corrosion Control Program (BACCP) 4
001 Program Section
71111.08P Procedures SEP-ISI-104 Program Section For ASME Section XI, Division 1 WF3 14
Inservice Inspection Program
71111.08P Procedures SEP-ISI-104 Program Section for ASME Section X, Division 1 WF3 14
Inservice Inspection Program
71111.08P Procedures SEP-PT-WF3- Waterford 3 Inservice Inspection Pressure Testing (PT) 001
001 Program Section
71111.08P Procedures SEP-SG-WF3- Waterford -3 (W3/WF3) Steam Generator Program 4
001
71111.08P Self-Assessments LO-HQNLO- 2022 Welding Program Assessment 02/17/2022
2021-19
71111.08P Self-Assessments LO-WLO-2022- Pre-NRC RF25 ISI Activities Self -Assessment Report 08/08/2023
0060-CA
71111.08P Self-Assessments LO-WLO-2022- Pre-NRC RF25 ISI Activities Self-Assessment Report 08/08/2023
0060-CA-3
71111.08P Work Orders WO No. 572188-24, 589604-15
71111.11Q Procedures EN-OP-115 Conduct of Operations 31
71111.11Q Procedures OP-010-005 Plant Shutdown 345
71111.11Q Procedures OP-901-311 Loss of Train B Safety Bus 313
71111.11Q Procedures OP-901-521 Severe Weather and Flooding 343
71111.11Q Procedures OP-902-001 Reactor Trip Recovery 21
71111.11Q Procedures OP-902-003 Loss of Offsite Power / Loss of Forced Circulation Recovery11
71111.12 Corrective Action CR-WF3-YYYY- 2022-06818, 2023-01910, 2023-01944, 2023-13294,
Documents NNNN 2023-13313, 2023-13331, 2023-14317, 2023-14967,
2023-16596, 2023-13943, 2023-14310, 2023-14314
71111.12 Corrective Action CR-WF3-YYYY- 2024-00169 01/10/2024
Documents NNNN
Resulting from
Inspection
71111.12 Engineering EC 54051011 Engineering Change 09/14/2023
Changes
24
Inspection Type Designation Description or Title Revision or
Procedure Date
71111.12 Miscellaneous TD G080.0095 General Electric Switchgear Magne Blast Breakers 6
71111.12 Procedures EN-DC-205 Maintenance Rule Monitoring 9
71111.12 Procedures ME-004-115 4.16/6.9 kV G.E. Magne-Blast Breaker Overhaul 6, 9
71111.12 Procedures OP-903-094 ESFAS Subgroup Relay Test - Operating 35
71111.12 Work Orders 00517244, 00586519, 52790255, 52805142, 54034973,
54038818
71111.15 Corrective Action CR-WF3-YYYY- 2023-16283, 2023-16372, 2023-16376, 2023-17876,
Documents NNNN 2023-15594
71111.15 Engineering EC 54056366 Engineering Change 0
Changes
71111.15 Procedures OP-009-005 Shutdown Cooling 45
71111.15 Procedures OP-901-511 Instrument Air Malfunction 20
71111.15 Corrective Action CR-WF3-YYYY- 2023-17399
Documents NNNN
71111.24 Corrective Action CR-WF3-YYYY- 2019-01293, 2023-18027, 2017-03359, 2017-04081,
Documents NNNNN 2018-00948,
71111.24 Engineering EC 54093486 ACC-127B Input to Operability CR-23-18244/18245 12/21/2023
Changes
71111.24 Engineering EC 72080 Use of instrumentation for ACCW System Flow Balance 05/12/2017
Changes PE-004-024
71111.24 Procedures FSG-005 Initial Assessment and FLEX Equipment Staging 15
71111.24 Procedures OP-903-003 Charging Pump Operability Check 315
71111.24 Procedures OP-903-052 Controlled Ventilation Area System Operability Check 15
71111.24 Procedures OP-903-096 Boron Flowrate Verification 11
71111.24 Procedures OP-903-115 Train A Integrated Emergency Diesel 59
71111.24 Procedures OP-903-121 Safety Systems Quarterly IST Valve Tests 36
71111.24 Procedures PE-004-024 ACCW & CCW System Flow Balance 310
71111.24 Procedures STA-001-004 Local Leak Rate Test (LLRT) 320
71111.24 Work Orders 53013043, 54002710, 00586332, 53017375, 54067505,
54085552, 00474102, 00495521, 00502714, 00517264,
00518612
71124.01 ALARA Plans RWP 2022-0512 1RE24 Steam Generator 1 and 2 Feedring Mod 2
71124.01 ALARA Plans RWP 2022-0615 1RE24 Remove/Replace Pressurizer Heater 6
25
Inspection Type Designation Description or Title Revision or
Procedure Date
71124.01 Corrective Action CR-WF3-YYYY- 2022-01953, 2023-00421, 2022-03390, 2022-06963,
Documents XXXXX 2023-00518, 2023-01766, 2023-01234, 2022-02542,
2022-07912, 2023-00114, 2023-16348, 2023-16474,
71124.01 Corrective Action CR-WF3-YYYY- 2023-16870, 2023-16872, 2023-16893
Documents XXXXX
Resulting from
Inspection
71124.01 Procedures EN-RP-100 Radiation Worker Expectations 14
71124.01 Procedures EN-RP-101 Access Control for Radiologically Controlled Areas 17
71124.01 Procedures EN-RP-102 Radiological Control 008
71124.01 Procedures EN-RP-110 ALARA Program 14
71124.01 Procedures EN-RP-121 Radioactive Material Control 19
71124.01 Procedures EN-RP-141-01 Job Coverage Using Remote Monitoring Technology 8
71124.01 Procedures EN-RP-152 Conduct of Radiation Protection 008
71124.01 Procedures HPI-001-123 Plant Conditions and Radiological Concerns 010
71124.01 Radiation WF3-2301-00269 RAB -4 Purification Ion Exchangers 01/24/2023
Surveys
71124.01 Radiation WF3-2308-00181 RAB -35 Spent Resin Tank Pump Room / Waste 08/22/2023
Surveys Condensate IX
71124.01 Radiation WF3-2309-00144 RAB -35 Boric Acid Pre-Concentrator Filters 09/14/2023
Surveys
71124.01 Radiation WF3-2309-00185 FHB +46 Fuel Handling Area 09/18/2023
Surveys
71124.01 Radiation WF3-2309-00225 RAB -4 Center Wing 09/23/2023
Surveys
71124.01 Radiation WF3-2309-00251 Radwaste Solidification Building 09/26/2023
Surveys
71124.01 Radiation WF3-2310-00066 RAB -4 Flash Tank / Purification Filter Area 10/05/2023
Surveys
71124.01 Radiation Work 2022-0623 REFUEL 24 - Perform miscellaneous contaminated system 01
Permits (RWPs) valve work in the Regen Hx Room including all support
activities, troubleshooting, walkdowns, tagouts, tours and
inspections.
71124.01 Radiation Work 2022-0641 REFUEL 24 - Emergent Dose added Inside the Reactor 00
26
Inspection Type Designation Description or Title Revision or
Procedure Date
Permits (RWPs) Containment Building.
71124.01 Radiation Work 2023-0404 REFUEL 25 - Plant Maintenance Valve Work on 00
Permits (RWPs) Contaminated and Clean System Valves outside the Reactor
Containment Building.
71124.01 Radiation Work 2023-0702 REFUEL 25 - Disassembly of Reactor Head and All 02
Permits (RWPs) Associated Work Activities.
71124.01 Radiation Work 2023-0714 REFUEL 25 - Cleaning of the Reactor Head Surface. 00
Permits (RWPs) Includes all supporting activities and Bare Metal Inspections.
71124.01 Self-Assessments LO-WLO-2022- Radiological Hazard Assessment and Exposure Controls 08/21/2023
0051 CA-00004
71124.04 Corrective Action CR-WF3-YYYY- 2020-01981, 2020-02198, 2020-03232, 2020-07014,
Documents NNNN 2021-00302, 2021-02028, 2022-01780, 2022-01921,
2022-03253, 2022-07004, 2023-01604, 2023-15043,
2023-16119
71124.04 Miscellaneous Evaluation of DLR/SRD Discrepancies and DLRs Not 06/30/2020
Returned for Processing
71124.04 Miscellaneous Evaluation of DLR/SRD Discrepancies and DLRs Not 07/10/2022
Returned for Processing
71124.04 Miscellaneous NRC Annual Dose Report (REIRS) 2022
71124.04 Miscellaneous 15403 Dose Assessment from PCE 10/02/2020
71124.04 Miscellaneous 56286 Dose Assessment from PCE 10/02/2020
71124.04 Miscellaneous 57700 Dose Assessment from PCE 10/02/2020
71124.04 Miscellaneous 64832 Dose Assessment from PCE 10/02/2020
71124.04 Miscellaneous 92905 Dose Assessment from PCE 04/18/2022
71124.04 Procedures EN-RP-122 Alpha Monitoring 10
71124.04 Procedures EN-RP-201 Dosimetry Administration 5
71124.04 Procedures EN-RP-203 Dose Assessment 10
71124.04 Procedures EN-RP-204 Special Monitoring Requirements 11
71124.04 Procedures EN-RP-204-01 Effective Dose Equivalent (EDEX) Monitoring 3
71124.04 Procedures EN-RP-205 Prenatal Monitoring 5
71124.04 Procedures EN-RP-206 Dosimeter of Legal Record Quality Assurance 7
71124.04 Procedures EN-RP-208 Whole Body Counting/In-Vitro Bioassay 7
71124.04 Self-Assessments LO-WLO-2022- Occupational Dose Assessment 10/05/2023
00051
27
Inspection Type Designation Description or Title Revision or
Procedure Date
71124.04 Self-Assessments QA-14/15-2021- Quality Assurance Audit: Combined Radiation Protection 10/25/2021
W3-01 and Radwaste
71124.08 Shipping Records RSN#: 23-1009 Shipment consisting of two 20-foot intermodal containers 10/26/2023
(ESUU200404 and ESUU200865) of dry active waste,
UN2912, radioactive material, low specific activity (LSA-I)
71152A Corrective Action CR-WF3-YYYY- 2022-01874, 2022-03111, 2022-06393, 2022-06647,
Documents NNNN 2022-06852, 2023-15179, 2023-15245, 2023-16237
71152A Corrective Action CR-WF3-YYYY- 2023-14746, 2023-14747, 2023-14895, 2023-15933,
Documents NNNN 2023-15424
Resulting from
Inspection
71152A Work Orders 53005507, 53022055, 53022177, 53005391, 54003998
71152S Corrective Action CR-WF3-YYYY- 2023-01793, 2023-01911, 2023-14593, 2023-15322,
Documents NNNN 2023-15407, 2023-15858, 2023-16043
71152S Corrective Action CR-WF3-YYYY- 2023-15830
Documents NNNN
Resulting from
Inspection
28