ML23310A032
| ML23310A032 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 11/13/2023 |
| From: | William Schaup NRC/RGN-IV/DORS/PBC |
| To: | Hansett P Entergy Operations |
| Schaup W | |
| References | |
| EA-23-065 IR 2023003 | |
| Download: ML23310A032 (44) | |
See also: IR 05000458/2023003
Text
November 13, 2023
Phil Hansett, Site Vice President
Entergy Operations, Inc.
5485 U.S. Highway 61N
St. Francisville, LA 70775
SUBJECT: RIVER BEND STATION - INTEGRATED INSPECTION
REPORT 05000458/2023003 AND NOTICE OF VIOLATION
Dear Phil Hansett:
On September 30, 2023, the U.S. Nuclear Regulatory Commission (NRC) completed an
inspection at River Bend Station. On November 13, 2023, 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.
The enclosed report documents a cited Severity Level IV violation for the failure to obtain a
license amendment prior to making a change to the licensing basis for the ultimate heat sink for
the River Bend Station. The NRC evaluated this violation in accordance with section 2.3.2 of the
NRC Enforcement Policy, which can be found at http://www.nrc.gov/about-
nrc/regulatory/enforcement/enforce-pol.html. The violation is cited in Enclosure 1, Notice of
Violation (Notice). We determined that this violation did not meet the criteria to be treated as a
non-cited violation (NCV), consistent with section 2.3.2 of the Enforcement Policy, because the
licensee failed to restore compliance within a reasonable period of time after the violation was
identified.
You are required to respond to this letter and should follow the instructions specified in the
enclosed Notice of Violation when preparing your response. If you have additional information
that you believe the NRC should consider, you may provide it in your response. The NRCs
review of your response will also determine whether further enforcement action is necessary to
ensure your compliance with regulatory requirements.
Additionally, six findings of very low safety significance (Green) are documented in this report.
All of these findings involved violations of NRC requirements, one of which was determined to
be severity level IV. One severity level IV violation without an associated finding is documented
in this report. We are treating these violations as NCVs consistent with section 2.3.2 of the
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
P. Hansett 2
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 River Bend Station.
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 River Bend Station.
This letter, its enclosures, and your response 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,
Signed by Schaup, William
on 11/13/23
William T. Schaup, Jr., Acting Chief
Reactor Projects Branch C
Division of Operating Reactor Safety
Docket No. 05000458
License No. NPF-47
Enclosures:
1. Notice of Violation
2. Inspection Report 05000458/2023003
cc w/ encl: Distribution via LISTSERV
SUNSI Review ADAMS: Non-Publicly Available Non-Sensitive Keyword:
By: Yes No Publicly Available Sensitive
OFFICE RI:DORS/C SRI:DORS/C SPE:DORS/C TL:ACES AD:DORS
NAME EPowell CWynar MBloodgood JGroom TClark
SIGNATURE ELP1 CJW5 MRB4 JRG2 TXV
DATE 11/06/23 11/06/23 11/06/23 11/09/23 11/13/23
OFFICE ABC:DORS/C
NAME WSchaup
SIGNATURE WTS1
DATE 11/13/23
NOTICE OF VIOLATION
Entergy Operations, Inc. Docket No. 05000458
River Bend Station License No. NPF-47
During an NRC inspection conducted from July 1 through September 30, 2023, a violation of
NRC requirements was identified. In accordance with the NRC Enforcement Policy, the violation
is listed below:
10 CFR 50.59(c)(2)(ii) requires, in part, that a licensee shall obtain a license amendment
pursuant to 10 CFR 50.90 prior to implementing a proposed change if the change would
result in more than a minimal increase in the likelihood of occurrence of a malfunction of
a structure, system, or component (SSC) important-to-safety previously evaluated in the
final safety analysis report (as updated).
Contrary to the above, from October 27, 2011, to September 30, 2023, the licensee
failed to obtain a license amendment pursuant to 10 CFR 50.90 prior to implementing a
proposed change, test, or experiment if this activity would result in more than a minimal
increase in the likelihood of occurrence of a malfunction of an SSC important-to-safety
previously evaluated in the final safety analysis report (as updated). Specifically, the
licensee changed the design basis of the ultimate heat sink inventory requirements from
providing a 30-day cooling water supply without the need for makeup to providing a less
than 30-day cooling water supply with makeup capability, without obtaining a license
amendment. The licensee implemented temporary compensatory measures that would
ensure a 30-day ultimate heat sink cooling water supply.
This is a Severity Level IV violation (Enforcement Policy Section 6.1.d.2).
Pursuant to 10 CFR 2.201, Entergy Operations, Inc. is hereby required to submit a written
statement or explanation to the U.S. Nuclear Regulatory Commission, ATTN: Document Control
Desk, Washington, DC 20555-0001, with a copy to the Regional Administrator, U.S. Nuclear
Regulatory Commission, Region IV, 1600 East Lamar Blvd., Arlington, Texas 76011-4511, and
the NRC Resident Inspector at River Bend Station, and email it to R4Enforcement@nrc.gov
within 30 days of the date of the letter transmitting this Notice of Violation. This reply should be
clearly marked as a Reply to a Notice of Violation, NRC Inspection Report 05000458/2023003
and should include for the violation: (1) the reason for the violation, or, if contested, the basis for
disputing the violation or severity level; (2) the corrective steps that have been taken and the
results achieved; (3) the corrective steps that will be taken; and (4) the date when full
compliance will be achieved.
Your response may reference or include previous docketed correspondence if the
correspondence adequately addresses the required response. If an adequate reply is not
received within the time specified in this Notice of Violation, the NRC may issue an order or a
demand for information requiring you to explain why your license should not be modified,
suspended, or revoked, or why such other action as may be proper should not be taken. Where
good cause is shown, consideration will be given to extending the response time.
If you contest this enforcement action, please provide an additional copy of your response, with
your basis for denial, to the Director, Office of Enforcement, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001.
Enclosure 1
Because your response will be made available electronically for public inspection in the NRC
Public Document Room and from the NRCs ADAMS, accessible from the NRC website at
http://www.nrc.gov/reading-rm/adams.html, to the extent possible, it should not include any
personal privacy or proprietary information so that it can be made available to the public without
redaction.
If personal privacy or proprietary information is necessary to provide an acceptable response,
then please provide a bracketed copy of your response that identifies the information that
should be protected and a redacted copy of your response that deletes such information. If you
request that such material is withheld from public disclosure, you must specifically identify the
portions of your response that you seek to have withheld and provide in detail the bases for your
claim (e.g., explain why the disclosure of information will create an unwarranted invasion of
personal privacy or provide the information required by 10 CFR 2.390(b) to support a request for
withholding confidential commercial or financial information).
Dated this 13th day of November 2023
1-2
U.S. NUCLEAR REGULATORY COMMISSION
Inspection Report
Docket Number: 05000458
License Number: NPF-47
Report Number: 05000458/2023003
Enterprise Identifier: I-2023-003-0012
Licensee: Entergy Operations, Inc.
Facility: River Bend Station
Location: St. Francisville, Louisiana
Inspection Dates: July 1, 2023, to September 30, 2023
Inspectors: D. Antonangeli, Resident Inspector
B. Baca, Health Physicist
E. Powell, Resident Inspector
H. Strittmatter, Emergency Preparedness Inspector
C. Wynar, Senior Resident Inspector
Approved By: William T. Schaup, Jr., Acting Chief
Reactor Projects Branch C
Division of Operating Reactor Safety
Enclosure 2
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees
performance by conducting an integrated inspection at River Bend Station, 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 Take Risk Mitigating Actions for Unit Cooler A Work
Cornerstone Significance Cross-Cutting Report
Aspect Section
Mitigating Green H.5 - Work 71111.13
Systems NCV 05000458/2023003-01 Management
Open/Closed
The inspectors identified a Green finding and associated non-cited violation of 10 CFR
50.65(a)(4) for failing to manage an increase in risk from maintenance activities. Specifically,
on June 27, 2023, the licensee failed to manage the increase in risk from planned
maintenance on the division 1 containment unit cooler.
Failure to Obtain an Amendment for Technical Specification Bases Change
Cornerstone Significance/Severity Cross-Cutting Report
Aspect Section
Mitigating Green None (NPP) 71111.15
Systems Severity Level IV
Open/Closed
The inspectors identified a Green finding and associated severity level IV non-cited violation
of 10 CFR Part 50.59, Changes, Tests, and Experiments, when the licensee failed to
perform an adequate 10 CFR 50.59 evaluation and obtain a license amendment prior to
implementing change package 2000-11 on September 5, 2000. This change added a note to
the technical specification bases that changed the intent of the associated surveillance
requirement.
Failure to Perform a Radiation Survey in Accordance with 10 CFR 20.1501(a) to Ensure
Occupational Doses were Controlled Within Regulatory Limits.
Cornerstone Significance Cross-Cutting Report
Aspect Section
Occupational Green [H.14] - 71124.05
Radiation Safety NCV 05000458/2023003-03 Conservative
Open/Closed Bias
The inspectors identified a Green non-cited violation for the licensees failure to perform a
radiation survey (10 CFR 20.1501(a)) to ensure occupational doses were controlled within
regulatory limits (10 CFR 20.1201). Specifically, between July 10 and July 14, 2023, the
inspectors identified two examples of area radiation monitors in alert and alarm conditions
with dose rate readings that were not reflective of the current area radiological conditions.
Each monitor had reportedly been in this condition at or greater than 29 days. These monitors
were installed to provide constant dose rate information locally and/or in the control room and
provided local audible and visual alarms upon reaching a preset dose rate.
2-2
Notice of Violation of 10 CFR 50.59, Changes, tests and experiments, for the failure to
obtain a license amendment for a change to licensing basis for the River Bend ultimate heat
sink.
Cornerstone Severity Cross-Cutting Report
Aspect Section
Not Applicable Severity Level IV Not Applicable 71152A
Open
The inspectors identified a Severity Level IV Notice of Violation of 10 CFR 50.59, Changes,
tests and experiments, for the licensees failure to obtain a license amendment for a change
to the licensing basis for the ultimate heat sink that required prior NRC approval. Specifically,
the licensee failed to seek a license amendment for a change to the facility that allowed the
use of makeup to meet the 30-day ultimate heat sink inventory requirements.
Failure to Maintain Accurate Information in Updated Safety Analysis Report
Cornerstone Severity Cross-Cutting Report
Aspect Section
Not Applicable Severity Level IV Not Applicable 71152A
Open/Closed
The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.71(e) when the
licensee failed to update the updated safety analysis report to assure that the information
included in the report contains the latest information developed. Specifically, the licensee
failed to ensure that the design capacity of the diesel generator air start system was updated
to reflect a change to the diesel engine air start capacity and start time requirements.
Failure to Follow Procedures for Time Critical Actions
Cornerstone Significance Cross-Cutting Report
Aspect Section
Mitigating Green H.14 - 71152S
Systems NCV 05000458/2023003-06 Conservative
Open/Closed Bias
The inspectors identified a Green finding and associated non-cited violation of
10 CFR Part 50, appendix B, criterion V, Instructions, Procedures, and Drawings, for the
licensees failure to complete required actions in accordance with Procedure EN-OP-123,
Time Critical Action Program Standard, revision 007. Specifically, the licensee failed to
scope in actions associated with Procedure AOP-0004, Loss of Offsite Power, as time
critical and failed to take all time critical actions for shutting SSW-MOV96, Normal Service
Water Isolation Valve.
Failure to Properly Categorize Standby Service Water Valves in the Inservice Testing
Program
Cornerstone Significance Cross-Cutting Report
Aspect Section
Mitigating Green None (NPP) 71153
Systems NCV 05000458/2023003-07
Open/Closed
2-3
The inspectors identified a Green finding and associated non-cited violation of
10 CFR 50.55a(f)(4), In-service Testing Requirements, for the licensees failure to categorize
standby service water isolation valves as Category A valves. Specifically, the licensees
inservice testing program did not test safety-related valves (SWP-MOV-57A and
SWP-MOV-57B) in accordance with ASME OM code Subsection ISTC-1300, Valve
Categories, to ensure they could meet seat leakage requirements. This caused the licensee
to be in violation of Technical Specification 3.7.1 for the operability of the ultimate heat sink.
Engineering Changes Failed to Evaluate Effects on Technical Specification Surveillance
Requirements
Cornerstone Significance Cross-Cutting Report
Aspect Section
Mitigating Green [P.1] - 71153
Systems NCV 05000458/2023003-08 Identification
Open/Closed
The inspectors identified a non-cited violation of 10 CFR Part 50, appendix B, criterion III and
associated technical specification violation of Technical Specification 3.8.1, AC Sources
Operating, when the licensee failed to incorporate original circuitry design into a new digital
control unit upgrade for the control building air conditioning system. This was identified when
the licensee performed STP-309-0602, Division II ECCS Test, revision 056, and the chillers
failed to meet the acceptance criteria and load onto the emergency diesel generator in the
required time in the final safety analysis report.
Additional Tracking Items
Type Issue Number Title Report Section Status
LER 05000458/2023-001-00 Ultimate Heat Sink 71153 Closed
Inoperable due to Boundary
Valve Leakage
LER 05000458/2023-001-01 Ultimate Heat Sink 71153 Closed
Inoperable due to Boundary
Valve Leakage (Supplement)
LER 05000458/2023-002-00 Division I and II Diesel 71153 Closed
Generators Inoperable due
to Exceeding Load
Sequence Times
LER 05000458/2023-002-01 Division I and II Diesel 71153 Closed
Generators Inoperable due
to Exceeding Load
Sequence Times
(Supplement)
2-4
PLANT STATUS
The plant began the quarter at 100 percent rated thermal power. On July 11, 2023, the unit was
down powered to 85 percent for a planned rod pattern adjustment and returned to 100 percent
on July 12, 2023. On August 18, 2023, the unit was down powered to 65 percent for a planned
rod pattern adjustment and returned to 100 percent power on August 20, 2023, where it
remained for the remainder of the third quarter.
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
71111.04 - Equipment Alignment
Partial Walkdown Sample (IP Section 03.01) (4 Samples)
The inspectors evaluated system configurations during partial walkdowns of the following
systems/trains:
(1) standby liquid control system on August 30, 2023
(2) high pressure core spray system on August 30, 2023
(3) residual heat removal (RHR) system on September 19, 2023
(4) 125 Vdc system on September 19, 2023
71111.05 - Fire Protection
Fire Area Walkdown and Inspection Sample (IP Section 03.01) (5 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) cable chase III, fire area C-9, on August 7, 2023
(2) high pressure core spray battery room and battery charger room, fire area C-21, on
August 10, 2023
(3) division 1 emergency diesel generator (EDG), fire area DG-6/Z-1, on August 16, 2023
2-5
(4) RHR pump C room, fire area AB-4/Z-1 and Z-2, on September 7, 2023
(5) reactor core isolation cooling (RCIC) pump room, fire areas AB-1/Z-1 and AB-15/Z-1,
on September 7, 2023
Fire Brigade Drill Performance Sample (IP Section 03.02) (1 Sample)
(1) The inspectors evaluated the onsite fire brigade training and performance during an
unannounced fire drill on September 21, 2023.
71111.06 - Flood Protection Measures
Flooding Sample (IP Section 03.01) (1 Sample)
(1) The inspectors evaluated internal flooding mitigation protections in the RHR A
cubicle.
71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance
Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01)
(2 Samples)
(1) The inspectors observed and evaluated licensed operator performance in the control
room during downpower to 85 percent to repair feed pump B oil leak on July 11,
2023.
(2) The inspectors observed and evaluated licensed operator performance in the control
room during downpower to 60 percent for rod pattern adjustment on August 21, 2023.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (2 Samples)
(1) The inspectors observed and evaluated an assistant operations manager drill and
exercise performance scenario for shift manager proficiency on August 29, 2023.
(2) The inspectors observed and evaluated a graded crew exam on September 19, 2023.
71111.12 - Maintenance Effectiveness
Maintenance Effectiveness (IP Section 03.01) (2 Samples)
The inspectors evaluated the effectiveness of maintenance to ensure the following
structures, systems, and components (SSCs) remain capable of performing their intended
function:
(1) division 1 EDG turbo charger bolt failure on August 14, 2023
(2) maintenance preventable functional failures for fire dampers on September 18, 2023
2-6
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management Sample (IP Section 03.01) (7 Samples)
The inspectors evaluated the accuracy and completeness of risk assessments for the
following planned and emergent work activities to ensure configuration changes and
appropriate work controls were addressed:
(1) elevated risk during RCIC maintenance on July 11, 2023
(2) elevated risk from unplanned limiting condition for operation (LCO) entry for control
room fresh air due to division 2 inoperability on July 20, 2023
(3) main turbine rotor crane lift high integrated risk and trip risk on August 2, 2023
(4) elevated risk due to RHR A STP-203-1300, revision 24, new revision to probabilistic
risk assessment on August 3, 2023
(5) elevated risk during work on containment A unit cooler on August 7, 2023
(6) elevated risk during RHR A breaker current injection testing on August 7, 2023
(7) elevated risk for containment unit cooler B during standby service water (SSW)
STP-256-6604 on August 22, 2023
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (5 Samples)
The inspectors evaluated the licensees justifications and actions associated with the
following operability determinations and functionality assessments:
(1) time critical operator actions for SSW-MOV-96B effect on system operability on
August 10, 2023
(2) emergency core cooling system drift calculations effect on system operability on
August 10, 2023 (CR-RBS-2023-01183, CR-RBS-2023-01184, and
(3) fire suppression system piping rupture effect on fire suppression operability on
August 10, 2023
(4) operability of RHR C shutdown cooling suction line pipe support bend on August 11,
2023 (CR-RBS-2023-05492)
(5) division 2 EDG stator temperatures on August 22, 2023 (CR-RBS-2023-06501)
71111.18 - Plant Modifications
Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02)
(2 Samples)
The inspectors evaluated the following temporary or permanent modifications:
(1) Engineering Change (EC) 93741, upgrade EDG turbo mounting bolts, on August 16,
2023
(2) EC 94390, main steam positive leakage control system MOVF028 replacement, on
September 18, 2023
71111.24 - Testing and Maintenance of Equipment Important to Risk
2-7
The inspectors evaluated the following testing and maintenance activities to verify system
operability and/or functionality:
Post-Maintenance Testing (PMT) (IP Section 03.01) (7 Samples)
(1) work order (WO) 54013857, RCIC limit switch replacement, on August 11, 2023
(2) WO 00523755, E12-VF063B valve repair, on August 11, 2023
(3) WO 54038407, SWP-MOV507B failed to open, on August 24, 2023
(4) WO 53018691, division 2 control building chilled water system HVC-FS62B,
1HVC-ACU1B flow switch repair, on August 24, 2023
(5) WO 00501472/WO 00574687/WO 00584865, component cooling water secondary
pump CCS-P1A repair, on September 8, 2023
(6) WO 52973338, standby liquid control pump 1A discharge header pressure relief valve
replacement, on September 8, 2023
(7) WO 54047693, reactor water cleanup pump B tubing replacement, on September 14,
2023
Surveillance Testing (IP Section 03.01) (4 Samples)
(1) STP-403-1301, HVR-UC1B System B Timer Channel Calibration, revision 09, on
July 1, 2023
(2) STP-207-4241, RCIC Equipment Room Ambient Temperature High Channel
Calibration and LSFT, revision 15, on July 7, 2023
(3) STP-207-4538, RCIC Isolation - RCIC Steam Supply Pressure - Low Channel
Functional Test, revision 305, on September 6, 2023
(4) STP-309-0202, DIV 2 EDG Operability Test, revision 339, on September 11, 2023
Inservice Testing (IST) (IP Section 03.01) (2 Samples)
(1) STP-208-6302, Div II MSIV Leakage Control Quarterly Valve Operability Test,
revision 12, on September 7, 2023
(2) STP-201-6310, SLC Pump and Valve Operability Test, revision 316, on
September 12, 2023
71114.02 - Alert and Notification System Testing
Inspection Review (IP Section 02.01-02.04) (1 Sample)
(1) The inspectors evaluated the licensees maintenance, testing, and audit program for
the alert and notification system between July 1, 2021, and August 4, 2023.
71114.03 - Emergency Response Organization Staffing and Augmentation System
Inspection Review (IP Section 02.01-02.02) (1 Sample)
(1) The inspectors evaluated the readiness of the Emergency Preparedness
Organization between July 1, 2021, and August 4, 2023. Inspectors also evaluated
the licensees ability to staff their emergency response facilities in accordance with
emergency plan commitments.
71114.04 - Emergency Action Level and Emergency Plan Changes
2-8
Inspection Review (IP Section 02.01-02.03) (1 Sample)
(1) The inspectors evaluated the 10 CFR 50.54(q) emergency plan change process and
practices between July 1, 2021, and August 4, 2023. The evaluation reviewed
screenings and evaluations documenting implementation of the process. The reviews
of the change process documentation do not constitute NRC approval.
71114.05 - Maintenance of Emergency Preparedness
Inspection Review (IP Section 02.01 - 02.11) (1 Sample)
(1) The inspectors evaluated the maintenance of the emergency preparedness program
between July 1, 2021, and August 4, 2023. The evaluation reviewed evidence of
completing various emergency plan commitments, the conduct of drills and exercises,
licensee audits and assessments, and the maintenance of equipment important to
71114.06 - Drill Evaluation
Select Emergency Preparedness Drills and/or Training for Observation (IP Section 03.01)
(1 Sample)
(1) full site emergency preparedness drill with external agency participation on
August 10, 2023
Drill/Training Evolution Observation (IP Section 03.02) (1 Sample)
The inspectors evaluated:
(1) graded crew exam with drill and exercise performance opportunity on August 29,
2023
RADIATION SAFETY
71124.05 - Radiation Monitoring Instrumentation
Walkdowns and Observations (IP Section 03.01) (8 Samples)
The inspectors evaluated the following radiation detection instrumentation during plant
walkdowns:
(1) air samplers (AMS4 and low volume air sampler) in the fuel handling and turbine
buildings
(2) area radiation monitors (ARMs) in auxiliary, fuel handling, radwaste, reactor, and
turbine buildings
(3) radiation detection and measurement equipment (friskers, high purity germanium
detectors, liquid scintillator, proportional counter, portable air sampler) staged for use
in the chemistry counting laboratory
(4) portable radiation instrumentation (friskers, iSolo alpha beta counter) staged for use
at radiologically controlled area egress
2-9
(5) portable radiation instrumentation (friskers, ion chambers, beta counter, telepoles)
staged for use in the instrument storage room
(6) stationary radiation instrumentation (tool and equipment monitors, personnel
contamination monitors, and portal monitors) located at the radiologically controlled
area egress
(7) stationary radiation instrumentation (portal monitors) at the protected area egress
(8) whole body counter and portal monitor in the dosimetry office
Calibration and Testing Program (IP Section 03.02) (14 Samples)
The inspectors evaluated the calibration and testing of the following radiation detection
instruments:
(1) ARGOS-5AB: #140-184 dated April 27, 2023; #140-186 dated October 20, 2022;
- 1812-369 dated June 23, 2023; and #2206-142 dated January 13, 2023
(2) ASP-1 frisker: CHP-MF-040 dated February 28, 2022, and CHP-MF-124 dated
February 28, 2022
(3) CRONOS: #1011-060 dated September 13, 2022; #1011-061 dated November 20,
2022; #1412-364 dated January 25, 2023; and #2205-102 dated February 17, 2023
(4) DMC-3000: #940981 dated February 2, 2023, and A04D33 dated August 17, 2022
(5) Eberline AMS-4: RBS06 dated February 6, 2022, and RBS07 dated September 1,
2022
(6) Eberline BC-4: CHP-BC-001 dated November 8, 2022
(7) Eberline PM7: #419-08 dated August 11, 2022
(8) GEM-5: #1410-188 dated April 21, 2023; #1410-189 dated September 27, 2022;
- 1902-041 dated November 19, 2022; #2112-320 dated 2023; #2112-321 dated
March 23, 2023; and #2206-141 dated February 17, 2023
(9) high purity germanium detectors: Det1 dated March 28, 2023, and Det4 dated March
28, 2023
(10) iSolo: CH-C-012 dated February 6, 2023, and CH-C-066 dated January 1, 2023
(11) Ludlum 144: CHP-CR-075 dated August 4, 2022; CHP-CR-092 dated October 6,
2022; CHP-CR-110 dated August 4, 2022; CHP-CR-118 dated August 4, 2022; CHP-
CR-137 dated September 6, 2022; CHP-CR-155 dated October 6, 2022; and CHP-
CR-237 dated October 6, 2022
(12) Ludlum 9-3 ion chamber: CHP-DR-487 dated September 13, 2022; CHP-DR-519
dated February 28, 2023; CHP-DR-529 dated February 28, 2023; and CHP-DR-852
dated January 10, 2023
(13) Mirion RDS-31(iTx): CHP-DR-541 dated November 16, 2022; CHP-DR-563 dated
December 12, 2022; and CHP-DR-594 dated October 12, 2022
(14) telepole (Tele-Pole and Tele-Pole II): CHP-TEL030 dated November 7, 2022, and
CHP-TEL-097 dated August 4, 2022
Effluent Monitoring Calibration and Testing Program Sample (IP Section 03.03) (2 Samples)
The inspectors evaluated the calibration and maintenance of the following radioactive
effluent monitoring and measurement instrumentation:
(1) radwaste exhaust (RE-06A/B) and plant stack (RE-126) ventilation systems
(2) liquid effluent (RE-107) monitoring system
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71124.08 - Radioactive Solid Waste Processing & Radioactive Material Handling, Storage, &
Transportation
Radioactive Material Storage (IP Section 03.01) (3 Samples)
The inspectors evaluated the licensees performance in controlling, labeling, and securing
the following radioactive materials:
(1) storage of radioactive material located within the low level radioactive material
building
(2) storage of radioactive material located within the Stone & Webster building
(3) storage of radioactive material located within in the radwaste building
Radioactive Waste System Walkdown (IP Section 03.02) (1 Sample)
The inspectors walked down the following accessible portions of the solid radioactive waste
systems and evaluated system configuration and functionality:
(1) radwaste resin transfer system
Waste Characterization and Classification (IP Section 03.03) (2 Samples)
The inspectors evaluated the following characterization and classification of radioactive
waste:
(1) reactor water cleanup system, waste stream analysis dated 08/09/2022
(2) dry activated waste, waste stream analysis dated 04/17/2023
Shipment Preparation (IP Section 03.04)
There was no sample available when the inspectors were on site.
Shipping Records (IP Section 03.05) (5 Samples)
The inspectors evaluated the following non-excepted radioactive material shipments through
a record review:
(1) mixed bed ion exchange media, waste class A unstable, LSA-II, shipment number
(2) mixed bed ion exchange media, waste class A unstable, LSA-II, Shipment number
(3) filter/hardware liner, waste class B, Type B(U), Shipment number RBS-2022-031
(4) mixed bed ion exchange media, waste class A unstable, LSA-II, Shipment number
(5) mixed bed ion exchange media, waste class A unstable, LSA-II, Shipment number
OTHER ACTIVITIES - BASELINE
71151 - Performance Indicator Verification
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The inspectors verified licensee performance indicators submittals listed below:
EP01: Drill/Exercise Performance (DEP) Sample (IP Section 02.12) (1 Sample)
(1) October 1, 2022, through June 30, 2023
EP02: Emergency Response Organization (ERO) Drill Participation (IP Section 02.13)
(1 Sample)
(1) October 1, 2022, through June 30, 2023
EP03: Alert and Notification System (ANS) Reliability Sample (IP Section 02.14) (1 Sample)
(1) October 1, 2022, through June 30, 2023
71152A - Annual Follow-up Problem Identification and Resolution
Annual Follow-up of Selected Issues (Section 03.03) (1 Sample)
The inspectors reviewed the licensees implementation of its corrective action program
related to the following issue:
(1) heat exchanger inspection on ultimate heat sink (Inspection Report
Number 2017-003) on September 20, 2023
71152S - Semiannual Trend Problem Identification and Resolution
Semiannual Trend Review (Section 03.02) (1 Sample)
The inspectors reviewed the licensees corrective action program and other licensee
trending programs for potential adverse trends that might be indicative of a more significant
safety issue. The inspectors provided an observation in the results section with further
details of their review.
71153 - Follow Up of Events and Notices of Enforcement Discretion
Event Report (IP Section 03.02) (2 Samples)
The inspectors evaluated the following licensee event reports (LERs):
(1) LER 05000458/2023-001-00, Ultimate Heat Sink Inoperable due to Boundary Valve
Leakage (ADAMS Accession No. ML23143A276) and LER 05000458/2023-001-01,
Ultimate Heat Sink Inoperable due to Boundary Valve Leakage Supplement
The inspection conclusions associated with this LER and an associated non-cited
violation are documented in this report under IP 71153 in the Inspection Results
section. This LER is closed.
(2) LER 05000458/2023-002-00, Division I and II Diesel Generators Inoperable due to
Exceeding Load Sequence Times (ML23180A213) and LER 05000458/2023-002-01,
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Division I and II Diesel Generators Inoperable due to Exceeding Load Sequence
Times Supplement (ML23256A200)
The inspection conclusions associated with this LER and an associated non-cited
violation are documented in this report under IP 71153 in the Inspection Results
section. This LER is closed.
INSPECTION RESULTS
Failure to Take Risk Mitigating Actions for Unit Cooler A Work
Cornerstone Significance Cross-Cutting Report
Aspect Section
Mitigating Green H.5 - Work 71111.13
Systems NCV 05000458/2023003-01 Management
Open/Closed
The inspectors identified a Green finding and associated non-cited violation of 10 CFR
50.65(a)(4) for failing to manage an increase in risk from maintenance activities. Specifically,
on June 27, 2023, the licensee failed to manage the increase in risk from planned
maintenance on the division 1 containment unit cooler.
Description: The inspectors identified one example of maintenance activities that were
assessed as placing the plant in an elevated risk category but were not appropriately
managed in accordance with industry guidance and approved licensee procedures.
On June 27, 2023, the licensee was performing Procedure STP-051-4279, Containment Unit
Cooler System Instrumentation, Unit Cooler A - Containment to Annulus Differential Pressure
High Channel Calibration and Logic System Functional Test, revision 011. This makes
containment unit cooler 1A unavailable and places the plant in a yellow risk condition. During
a control room tour, the inspectors noticed that containment unit cooler 1B was not protected.
The inspectors reviewed Procedure EN-OP-119, Protected Equipment Postings,
revision 16, and noted that section 5.2.4 states, When the loss of the redundant component
or system would result in a technical specification (TS) action statement that requires an
immediate plant shutdown (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or less) and not performing a surveillance that meets the
requirements of step 5.2.9 then protect the component or system.
Procedure EN-OP-119, states, the shift manager is responsible for ensuring the hanging,
tracking and prompt removal of protected equipment postings based on the Risk Assessment
Process or requirements of Section 5.2, (Conditions that Require Posting of Protected
Equipment).
The inspectors determined that the licensee met the conditions in this section of procedure
EN-OP-119 that would require them to protect the division 2 containment unit cooler 1B,
since loss of the unit cooler 1B would result in a shutdown action statement of 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />. The
licensee failed to take the prescribed risk mitigating actions (RMAs) until questioned by the
inspectors.
Corrective Actions: The licensee entered this issue into their corrective action program.
Corrective Action References: CR-RBS-2023-05478
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Performance Assessment:
Performance Deficiency: Title 10 CFR 50.65(a)(4) requires the licensee to assess and
manage the increase in risk that may result from maintenance activities before performing
them. The inspectors determined that during the maintenance on containment unit cooler 1A
system, the licensee failed to assess and manage the increase in risk for maintenance
activities by taking required RMAs and was therefore a performance deficiency.
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 failure to protect the redundant train of
containment unit cooler adversely affected the Mitigating Systems cornerstone because the
failure to take RMAs for the division 2 containment unit cooler 1B increased the likelihood that
other mitigating systems could be affected.
Significance: The inspectors assessed the significance of the finding using IMC 0609
appendix K, Maintenance Risk Assessment and Risk Management SDP. The inspectors
requested that the licensee perform a risk assessment of the specific configurations of both
conditions. In their assessments, the licensee estimated the risk deficits and incremental core
damage probabilities for each condition were less than 1.0E-6. A regional senior reactor
analyst independently reviewed the licensees assessments and confirmed the licensees risk
estimates. The inspectors applied this information to the flowcharts in appendix K to
determine this finding had very low safety significance (Green).
Cross-Cutting Aspect: H.5 - Work Management: The organization implements a process of
planning, controlling, and executing work activities such that nuclear safety is the overriding
priority. The work process includes the identification and management of risk commensurate
to the work and the need for coordination with different groups or job activities. Specifically,
the shift manager failed to implement the RMAs required by EN-OP-119.
Enforcement:
Violation: Title 10 CFR 50.65(a)(4) requires, in part, that the licensee to assess and manage
the increase in risk that may result from maintenance activities.
Contrary to the above, on June 27, 2023, the licensee failed to assess and manage the
increase in risk before performing maintenance activities. Specifically, the licensee failed to
implement appropriate RMAs to manage the increase in risk that resulted from maintenance
on the containment unit cooler 1A system.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with
section 2.3.2 of the Enforcement Policy.
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Failure to Obtain an Amendment for Technical Specification Bases Change
Cornerstone Significance/Severity Cross-Cutting Report
Aspect Section
Mitigating Green None (NPP) 71111.15
Systems Severity Level IV
Open/Closed
The inspectors identified a Green finding and associated Severity Level IV non-cited violation
of 10 CFR 50.59, Changes, Tests, and Experiments, when the licensee failed to perform an
adequate 10 CFR 50.59 evaluation and obtain a license amendment prior to implementing
change package 2000-11 on September 5, 2000. This change added a note to the technical
specification bases that changed the intent of the associated surveillance requirement.
Description: The control building air conditioning (AC) system consists of two independent,
redundant subsystems that provide cooling and heating of the control building air. The
associated control building chilled water subsystem (HVK) supplies chilled water to both
subsystem AC units, as well as to the main control room AC units.
The design of the AC electrical power system provides independence and redundancy to
ensure an available source of power to the engineered safety feature (ESF) systems. Each
ESF bus has a dedicated onsite diesel generator (DG). A DG starts automatically on loss of
coolant accident (LOCA) or an ESF bus degraded voltage, undervoltage/loss of power (LOP)
signal. Certain required plant loads in divisions 1 and 2 (including HVK) are returned to
service in a predetermined sequence to prevent overloading the DG.
The licensee completed modifications to the control building chiller controllers for
HVK-CHL 1C and 1D in 2014 and then completed the same modifications to HVK-CHL 1A
and 1B in 2020. Historical data recorded shows that since the control building chiller control
modifications, the start of the control building chillers may be delayed by approximately
40 seconds past the design start time.
On February 12, 2023, testing of the division 2 emergency core cooling system (ECCS)
revealed that both division 2 control building chillers (HVK-CHL 1B and 1D) failed to
sequence onto the emergency bus within the TS required time. Investigation of the issue
revealed that the same condition was present for the division 1 control building chillers (HVK-
CHL 1A and 1C). The licensee determined the condition was caused by the modification that
replaced the control building chiller controllers.
The licensee documented this issue in CR-RBS-2023-04966 and performed an operability
evaluation for the condition.
In the operability evaluation the licensee referenced a note in the TS bases for surveillance
requirement (SR) 3.8.1.18 that states this surveillance requirement pertains only to the load
sequence timer itself, and not to the interposing logic which comprises the remainder of the
circuit. The evaluation further states that in the case of HVK-CHL 1B, the load sequence
timer is relay EJS-SWG1B-62 and is set for a 150 second time delay and that the correct
timing of the load sequence timers for the chiller is demonstrated by the start of the air
handling units within the 135-165 second band. The note in the TS bases was added by the
licensee in change package 2000-11 as part of the 10 CFR 50.59 process after a similar
ECCS test failure of the low-pressure core spray pump in 1999.
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During their review of the issue, the inspectors questioned the licensee on the intent of the TS
bases note, as well as the 50.59 evaluation performed in 2000 as part of the change
package. After receiving the licensees response that the note was added to clarify the SR
intent, the inspectors requested assistance from the technical specification branch of the
Office of Nuclear Reactor Regulation to provide clarification on the intent of SR 3.8.1.18.
River Bend Station TS SR 3.8.1.18 states Verify sequence time is within +/- 10% of design for
each load sequencer timer. The bases states that the intent of SR 3.8.1.18 is that the
10 percent load sequence time tolerance ensures that sufficient time exists for the bus power
supply to restore frequency and voltage prior to applying the next load and that safety
analysis assumptions regarding ESF equipment time delays are not violated. River Bend
Station Updated Safety Analysis Report (USAR) Tables 8.3-2a and 8.3-2b (automatic and
manual loading of ESF buses for divisions 1 and 2) include the start time for each ESF load
calculated from the time of the LOOP/LOCA signal to start the EDGs.
River Bend Station SR 3.8.1.18 acceptance criteria for the timing sequence intervals should
include the start of the associated loads. The function of load sequencing is based on 1) the
accident analysis load assumptions and 2) ensuring the EDG has sufficient voltage and
frequency to start each load.
The inspectors determined that the added note contradicts the preceding sentence that
describes the intent of the surveillance and appears to contradict the licensees current
surveillance procedure acceptance criteria. Additionally, the note is also not specific to any
load. If it were applied to a load timed earlier than or larger than the control building chiller
(which has a relatively long period before the next load would start), it could negatively affect
subsequent loads or the EDG itself.
The note is technically incorrect, as it defeats the purpose of meeting the USAR Table 8.3-2b
load starting time requirements. Although, this TS bases note did not adversely impact the
safety analyses for the chillers because they are the final load sequenced, it can adversely
affect the starting of other ESF loads. Any change to SR bases which defeats the real
purpose of the SR is incorrect.
Changes to the licensees TS bases are made in accordance with TS 5.5.11, Technical
Specifications (TS) Bases Control Program, that provides a means for processing changes
to the bases of these TS. The TS states the following:
a. Changes to the bases of the TS shall be made under appropriate administrative
controls and reviews.
b. Licensees may make changes to bases without prior NRC approval provided the
changes do not involve either of the following:
1. A change in the TS incorporated in the license; or
2. A change to the USAR or bases that involves an unreviewed safety question as
2-16
defined in 10 CFR 50.59.
c. The bases control program shall contain provisions to ensure that the bases are
maintained consistent with the USAR.
d. Proposed changes that do not meet the criteria of either TS 5.5.11.b.1 or
TS 5.5.11.b.2 above shall be reviewed and approved by the NRC prior to
implementation. Changes to the bases implemented without prior NRC approval
shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).
The licensee did perform a 50.59 evaluation as part of the change package in July 2000 and
as part of that evaluation 10 CFR 50.59(c)(1)(i) required licensees evaluate whether a change
to the TS is required prior to making changes to the facility.
The 10 CFR 50.59 evaluation performed by the licensee concluded that no change was
required to the TS because the SR 3.8.1.18 timing requirement pertains only to the load
sequencer itself and not to the interposing logic which comprises the remainder of the circuit.
This does not compromise the effectiveness of sequencing loads onto the standby DG,
therefore, there is no increase in the probability of an accident or a malfunction of equipment
important to safety previously evaluated in the River Bend Station safety analysis report.
The inspectors concluded that this change constituted a change to the intent of associated
TS and defeated the purpose of meeting the USAR Table 8.3-2b load starting time
requirements and can adversely affect the starting of other ESF loads. The added note
changes the intent of the testing requirement and defeats the purpose of SR 3.8.1.18, and the
licensee incorrectly concluded that the change did not impact the TS and that a license
amendment was not required.
Corrective Actions: The preliminary root cause identified that when the controllers were
replaced, a delay was introduced in the start logic. This condition resulted in the inability to
comply with a SR of TS 3.8.1 and the issuance of licensee event report LER-2023-002. At the
time of discovery, River Bend Station was in a mode of operation for which TS 3.8.1 did not
apply. Corrective actions to restore TS compliance were completed prior to the next mode of
applicability. Additionally, the licensee has a corrective action in place to gather more
information to determine a solution that will provide a resolution to address the TS bases note
in question.
Corrective Action References: CR-RBS-2023-04060 and CR-RBS-2023-04966
Performance Assessment:
Performance Deficiency: The inspectors determined that the failure of the licensee to obtain a
license amendment that impacted the TS prior to making changes to the facility as required
by 10 CFR 50.59(c)(2)(ii) was a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor
because it was associated with the Design Control 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 obtain a license amendment prior to making
a change to the TS bases for SR 3.8.1.18 that changed the intent of the SR and had the
potential to adversely affect the starting of ESF loads by the DG.
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Significance: The inspectors assessed the significance of the finding using IMC 0609
appendix A, The Significance Determination Process (SDP) for Findings At-Power.
Specifically, using exhibit 2, Mitigating Systems Screening Questions, the inspectors
determined that this finding is of very low safety significance (Green) because the impacted
SSCs maintained their operability and probabilistic risk assessment (PRA) functionality.
Cross-Cutting Aspect: Not Present Performance. No cross-cutting aspect was assigned to
this finding because the inspectors determined the finding did not reflect present licensee
performance.
Enforcement: The ROPs significance determination process does not specifically consider
the regulatory process impact in its assessment of licensee performance. Therefore, it is
necessary to address this violation which impedes the NRCs ability to regulate using
traditional enforcement to adequately deter non-compliance.
Severity: Using the NRC Enforcement Policy, dated January 13, 2023, the inspectors
determined the violation was a severity level IV violation in accordance with section 6.1.d.2
because the violation of 10 CFR 50.59 resulted in conditions evaluated as having very low
safety significance (Green) by the SDP.
Violation: Title 10 CFR 50.59(c)(2)(ii) requires, in part, that a licensee shall obtain a license
amendment pursuant to 10 CFR 50.90 prior to implementing a proposed change that would
result in a more than minimal increase in the likelihood of occurrence of a malfunction of a
SSC important-to-safety previously evaluated in the USAR.
Contrary to the above, on September 5, 2000, the licensee failed to obtain a license
amendment pursuant to 10 CFR 50.90 prior to implementing a proposed change that would
result in a more than minimal increase in the likelihood of occurrence of a malfunction of a
SSC important-to-safety previously evaluated in the USAR. Specifically, the licensee made a
change to the TS bases without obtaining a license amendment that changed the intent of the
associated USAR and TS SRs. The licensee failed to recognize that the note added to the TS
bases for SR 3.8.1.18, in effect, constituted a change to the intent of associated TS and
defeated the purpose of meeting the USAR Table 8.3-2b load starting time requirements.
This type of change would, therefore, require a license amendment pursuant to
Enforcement Action: This violation is being treated as a non-cited violation, consistent with
section 2.3.2 of the Enforcement Policy.
Failure to Perform a Radiation Survey in Accordance with 10 CFR 20.1501(a) to Ensure
Occupational Doses were Controlled Within Regulatory Limits.
Cornerstone Significance Cross-Cutting Report
Aspect Section
Occupational Green [H.14] - 71124.05
Radiation Safety NCV 05000458/2023003-03 Conservative
Open/Closed Bias
The inspectors identified a Green non-cited violation for the licensees failure to perform a
radiation survey (10 CFR 20.1501(a)) to ensure occupational doses were controlled within
regulatory limits (10 CFR 20.1201). Specifically, between July 10 and July 14, 2023, the
inspectors identified two examples of area radiation monitors in alert and alarm conditions
with dose rate readings that were not reflective of the current area radiological conditions.
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Each monitor had reportedly been in this condition at or greater than 29 days. These monitors
were installed to provide constant dose rate information locally and/or in the control room and
provided local audible and visual alarms upon reaching a preset dose rate.
Description: 10 CFR 20.1003, defines a survey as an evaluation of the radiological conditions
and potential hazards incident to the production, use, transfer, release, disposal, or presence
of radioactive material or other sources of radiation.
River Bend Stations USAR, revision 27, section 12.5.2.2.5, established the area radiation
monitor (ARM) system in areas where it was desirable to have continuous dose rate
information. These monitors indicated dose rate locally and/or in the control room and
provided local audible and visual alarms upon reaching a preset dose rate. Therefore, the
intent of the ARM system was to provide a continuous survey of radiological conditions of the
work areas and inform workers of changing radiological conditions.
During an onsite walkdown of installed plant radiation monitors from July 10-14, 2023, the
inspectors identified two ARMs in alert and alarm conditions with dose rate readings not
representative of the alert and alarm conditions. The first ARM, RMS-RE-204, Condensate
Demineralizer Sink Area, was found to be in an alert condition, i.e., yellow beacon light
illuminated. The ARMs displayed dose rate radiation reading was inconsistent with the
inspectors independent measurement for activities and materials stored within the local area
and was not reflective of the most recent area survey (RBS-2306-00116 dated June 6, 2023).
This area was not an area frequently surveyed by radiation protection (RP) based on the low-
level radiological conditions during normal operations. The turbine building, where the ARM
was located, was routinely surveyed by RP on a quarterly schedule.
As noted in EIP-2-001, Classification of Emergencies, revision 30, attachment 2 RBS EAL
Basis Document, RMS-RE-204 dose rate information was utilized to inform responders
(operators, chemistry and radiation protection staff, and fire brigade) of sustained general
radiation levels which may preclude access to areas requiring continuous occupancy. In
corrective action CR-RBS-2023-05855, dated July 17, 2023, the control room noted a loss of
RMS-RE-204 communication and dispatched an operator to review the local conditions. The
operator noted the alert beacon was illuminated and radiation dose rate readings were
normal. This identification was a week after the inspectors identified RMS-RE-204 was not
providing a correct radiation dose rate reading, meaning the operator may not have had an
accurate radiation reading on the ARMs radiation dose rate display. The inspectors
interviewed an RP technician regarding the alert condition of RMS-RE-204 and was
informed the monitor had been intermittently going into alert status for some time and workers
were told to disregard the warning indicator. This guidance had the effect of desensitizing
workers to radiation alarms.
Based on a review of corrective action documents, the inspectors determined RMS-RE-204
had been having intermittent issues, i.e., reaching an alert/alarm condition without valid
conditions present, having erroneous radiological readings, and loosing communication with
the control room, since 2011 (CR-RBS-2011-2573). The inspectors reviewed corrective
action documents linked to the current alert condition and identified CR-RBS-2021-03740
dated May 19, 2021, as the most recent corrective action document for the monitors issues.
The licensee confirmed no additional corrective actions were generated for RMS-RE-204
since 2021. The inspectors requested confirmation of compensatory continuous dose rate
monitoring for the local area. The licensee demonstrated compensatory continuous dose rate
monitoring was established on November 15, 2021. This means the area was without
continuous dose rate monitoring for approximately 180 days. However, the area was
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surveyed for dose rates during the quarterly surveys dated July 14, 2021 (RBS-2107-00140)
and October 27, 2021 (RBS-2110-00229). These surveys provided an assessment of the
area 56 days after the ARM was identified deficient and 19 days prior to the compensatory
monitoring being established with 105 days in between the quarterly surveys the ARM was
deficient. Although quarterly surveys were performed and no abnormal conditions were
identified, RMS-RE-204 was not performing its intended function to provide accurate
continuous dose rate monitoring and informing workers of elevated radiation dose rates
through alert and alarm warning indications.
The compensatory monitoring was provided through the installation of an assigned DMC-
3000 teledosimeter (dosimeter number 912967) co-located at RMS-RE-204s detector. The
teledosimeter transmitted data via the ATOMS telemetry server to monitoring stations at the
radiologically controlled area access point. In addition, the teledosimeter readings could be
remotely accessed through the licensees network, as needed. The licensee performed
WO 52986825-01 dated December 12, 2021, to correct the monitor issue. However, the
intermittent conditions returned in the subsequent months based off an interview with an RP
technician and the inspectors observation during another facility walkdown during March 13-
16, 2023.
The second ARM observed during the onsite inspection and facility walkdown, RMS-RE-162,
67-foot Elevation Offgas Regeneration Area, was found in an alarm condition (audio alarm
engaged with no warning indication lights). The inspectors were unable to validate the dose
rate readings of the ARM due a low light environment, i.e., confirm the ARM was functional.
However, the inspectors independent measurement was not consistent with the ARMs
warning indication as the dose rates measured by the inspectors were near background
levels. The inspectors interviewed an RP technician regarding the length of time this monitor
was in the alarm condition. The RP technician believed the monitor had been in this condition
for approximately a month. The licensee provided CR-RBS-2023-04658 dated May 28, 2023,
for RMS-RE-162s current alarm condition with no indicator (alert or alarm) lights observed.
This means the ARM was in an alarm condition for approximately 42 days when the
inspectors found the monitor in alarm. The licensee generated work request WR 514828
dated May 28, 2023, to resolve the issue. The offgas regeneration area was routinely
surveyed by RP on a quarterly basis due to the low levels of radiation during normal
operations. The most current radiological survey was performed on June 26, 2023 (RBS-
2306-00400) and radiation dose rates were as anticipated, i.e., close to background levels.
However, this condition provided 29 days the area was not continuously monitored
for changes to radiation dose rates nor provided a valid local audible and visual alarm upon
reaching a preset dose rate, i.e., RMS-RE-162 was not performing its intended function.
Further, the continued audio alarm desensitized workers to valid warning indications.
The inspectors reviewed applicable corrective action documents associated with RMS-RE-
162 and determined this monitor was experiencing false elevated dose rate readings which
triggered alerts and alarms since 2005 (CR-RBS-2005-1546). The latest troubleshooting and
resolution to the intermittent issues with RMS-RE-162 was performed under WO 52989527-
01 dated July 17, 2022. The intermittent alarms returned as the radiation monitor was found
in alarm by the inspectors.
Corrective Actions: The licensee initiated the troubleshooting WOs to address the intermittent
alert and alarm conditions of RMS-RE-162 and RMS-RE-204, as well as the communication
losses to the control room. Compensatory monitoring was only initiated for RMS-RE-204 after
six months.
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Corrective Action References: The conditions were entered into the corrective action program
as CR-RBS-2021-03740, CR-RBS-2023-04658, and CR-RBS-2023-05855.
Performance Assessment:
Performance Deficiency: The failure to perform a radiation survey in accordance with 10 CFR 20.1501(a) to ensure occupational doses were controlled within regulatory limits was a
performance deficiency.
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. Specifically, the failure to
perform radiological surveys increased the workers potential exposure to unknown and/or
elevated radiation levels in the work area and desensitized the workers to radiation alarms.
Significance: The inspectors assessed the significance of the finding using IMC 0609
appendix C, Occupational Radiation Safety SDP. The inspectors determined the finding to
be of very low safety significance (Green) because (1) it was not associated with as low as is
reasonably achievable (ALARA) planning or work controls, (2) there was no overexposure,
(3) there was no substantial potential for an overexposure, and (4) the ability to assess dose
was not compromised.
Cross-Cutting Aspect: H.14 - Conservative Bias: Individuals use decision making-practices
that emphasize prudent choices over those that are simply allowable. A proposed action is
determined to be safe in order to proceed, rather than unsafe in order to stop. Specifically,
the licensee relied on operator and RP knowledge and recent quarterly surveys of the
affected areas to understand the current and present radiological conditions of these areas
until corrective actions to resolve the alerts and alarms were completed. This led workers to
rely on their electronic alarming dosimetry for unsuspected radiological changes when in
these areas rather than the continuous dose rate monitoring and warning indications of the
ARMs.
Enforcement:
Violation: Title 10 CFR 20.1003 defines a Survey as an evaluation of the radiological
conditions and potential hazards incident to the production, use, transfer, release, disposal, or
presence of radioactive material or other sources of radiation.
Title 10 CFR 20.1501(a) requires in part, that each licensee make or cause to be made
surveys of areas that may be necessary for the licensee to comply with the regulations in
10 CFR Part 20, and that surveys are reasonable under the circumstances to evaluate the
magnitude and extent of radiation levels, concentrations or quantities of radioactive materials,
and the potential radiological hazards that could be present.
Title 10 CFR 20.1201(a) requires, in part, that the licensee shall control occupational doses to
individual adults less than the annual limit of 5 rem total effective dose equivalent.
River Bend Stations USAR, revision 27, section 12.5.2.2.5, established the ARM system in
areas where it was desirable to have continuous dose rate information. These monitors
indicated dose rate locally and/or in the control room and provided local audible and visual
2-21
alarms upon reaching a preset dose rate. Therefore, the intent of the ARM system was to
provide a continuous survey of radiological conditions of the work areas and inform workers
of changing radiological conditions.
Contrary to the above, from approximately May 19 through November 15, 2021, and May 28
through June 26, 2023, the licensee failed to make or cause to be made, with two examples,
surveys of areas that may be necessary for the licensee to comply with the regulations in
10 CFR Part 20; and are reasonable under the circumstances to evaluate the magnitude and
extent of radiation levels. Specifically, the licensee did not provide compensatory continuous
dose rate surveys when the ARM system, in areas where it was desirable to have continuous
dose rate information, was unable to provide indications of dose rates locally and/or in the
control room and provide local audible and visual alarms upon reaching a preset dose rate.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with
section 2.3.2 of the Enforcement Policy.
Notice of Violation of 10 CFR 50.59, Changes, tests and experiments, for the failure to
obtain a license amendment for a change to licensing basis for the River Bend ultimate heat
sink.
Cornerstone Severity Level Cross-Cutting Report
Aspect Section
Mitigating Severity Level IV Not Applicable 71152A
Systems NOV 05000458/2023003-04
Open
The inspectors identified a Severity Level IV Notice of Violation of 10 CFR 50.59, Changes,
tests and experiments, for the licensees failure to obtain a license amendment for a change
to the licensing basis for the River Bend ultimate heat sink that required prior NRC approval.
Specifically, the licensee failed to seek a license amendment for a change to the facility that
allowed the use of makeup to meet the 30-day ultimate heat sink inventory requirements.
Description: The inspectors reviewed corrective actions associated with non-cited violation
(NCV)05000458/2011008-06 of 10 CFR 50.59, Changes, Tests, and Experiments, and
identified that the licensee did not restore compliance with the requirements of 10 CFR 50.59.
Specifically, the licensee failed to correct the NCV and restore compliance when they closed
corrective actions for the violation without obtaining a license amendment.
In October 2011, inspectors reviewed River Bend Stations USAR, section 9.2.5, Ultimate
Heat Sink, which describes the standby cooling tower and water storage basin that functions
as the ultimate heat sink (UHS) for River Bend Station during accident conditions. USAR
section 9.2.5.1, Design Bases, describes the criteria to which the UHS is designed in
accordance with General Design Criteria 44, Cooling Water. In particular, USAR section
9.2.5.1, criterion 2 states: The capacity of the [Ultimate Heat Sink] water storage basin is
designed to provide necessary cooling for the period of time (30 days) needed to evaluate the
situation, to take corrective action to mitigate the consequences of an accident, and if
required to take any necessary measures to permit water replenishment. In addition, alternate
methods are available for ensuring the continued capability of the sink beyond 30 days
(section 9.2.5.2).
Criterion 6 states: The UHS is designed to perform its intended safety function assuming any
single active or passive failure coincident with a loss of offsite power.
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When River Bend Station was originally licensed in November 1985, the licensee submitted
the design of the UHS that supplied a 30-day supply of water from the standby cooling tower
basin for decay heat removal, without a makeup water source. The NRC reviewed and
approved this design using Regulatory Guide 1.27, Ultimate Heat Sink for Nuclear Power
Plants. The NRCs position was reflected in the River Bend Station Safety Evaluation Report,
NUREG-0989, section 9.2.5, which states, the UHS contains more than a 30-day supply of
water for decay heat removal without makeup, in accordance with General Design Criteria 44.
Makeup water required after 30 days of UHS operation can be provided from the non-safety
related makeup water system if this system is available.
The original submitted Final Safety Analysis Report, revision 0 had the same criterion for 2
and 6 as it is presently stated in the current USAR revision.
Additionally, the UHS design of a standby cooling tower basin containing a 30-day supply
without makeup is confirmed in the River Bend Station Technical Specification Bases, section B 3.7.1, Standby Service Water System and Ultimate Heat Sink. Page B 3.7-1 states, the
basin is sized such that sufficient water inventory is available to provide heat removal
capability to safely shut down the plant and to maintain it in a cold shutdown condition for a
30-day period with no external makeup water source available (Regulatory Guide 1.27, Ref.
1).
The licensee intended to correct NCV 05000458/2011008-06 by submitting license
amendment request RBS LAR-2013-0018 that would credit makeup to the ultimate heat sink
in less than 30 days to account for system leakage and operation with more than one division
of standby service water in operation; however, the request was later withdrawn on July 7,
2014, due to NRC concerns. This license amendment was intended to correct the 2011 50.59
NCV and to restore compliance.
The following is an excerpt from the NRC response:
The NRC staff has concluded that the licensee did not provide sufficient information to
support its position that UHS/SSW leakage is beyond the design basis. In its letter dated May
29, 2014, the licensee cited NUREG-0800, Standard Review Plan [SRP] for the Review of
Safety Analysis Reports for Nuclear Power Plants: LWR Edition, section 9.2.1, Station
Service Water System, as part of the basis for its position that leakage is not a criterion for
design of the UHS. However, SRP section 9.2.1 also states that the SWS shall have the
capability to isolate components, subsystems, or piping if required so that the system safety
function will not be compromised. As such, the NRC staff believes that the licensees position
appears contrary to the guidance in SRP section 9.2.1. Therefore, if Entergy decides to re
submit the request, it must include the following information:
1) Provide the basis for why accounting for leakage, which compromises a safety function, is
not part of RBSs design basis for UHS inventory.
2) If leakage is determined to be part of the licensing basis, specify a leakage limit from the
3) Provide the basis for why the operation of one division of ECCS and SSW is the bounding
single failure when determining UHS inventory requirements.
4) From the submittals provided, the NRC staff could not verify that the alternative sources
2-23
of makeup water to the UHS provide acceptable methods to replenish the UHS. Provide
more detail on the three alternate sources of makeup water with regards to capacity,
implementation, and design.
A 2014 component design basis inspection reviewed the licensees corrective actions to
address NCV 05000458/2011008-06 and issued a NCV against 10 CFR 50, Appendix B,
Criterion XVI, Corrective Actions, because the licensee failed to take prompt corrective
actions to restore compliance for the 2011 violation.
As part of corrective actions from the 2011 and 2014 NCVs, the licensee revised the Final
Safety Analysis Report (FSAR), Section 9.2.5, to remove the following sentence describing
the licensing basis of the UHS:
Additional makeup is required for system leakage under licensing basis condition and when
operating two divisions with system leakage.
While removal of this sentence addressed a portion of the FSAR changes that caused the 10 CFR 50.59 NCV issued in 2011, the inspectors determined that additional portions of the
RBS licensing basis still relied on the use of make-up to meet the 30-day requirement for the
UHS and that the licensee had still not received NRC approval for this change. Specifically,
FSAR Section 9.2.5.1.2 states that, the capacity of the UHS water storage basin is designed
to provide necessary cooling for the period of time (30 days) needed to evaluate the situation,
to take corrective action to mitigate the consequences of an accident, and if required to take
any necessary measures to permit water replenishment. This is different than the approved
section from the SER that stated the UHS had more than a 30-day supply with no makeup
required. The statement above has a section to allow makeup if required i.e., cant make the
30-day time as well as a beyond 30-day makeup statement. Further, the inspectors noted
that AOP-004, Loss of Offsite Power, still has section 5.13.3 and a note to provide makeup
at day 10 of an accident.
The inspectors reviewed the current calculation of record, G13.18.13.2*086, Effects of
Maximum Safeguards Operation on the Ultimate Heat Sink (Standby Cooling Tower),
revision 3 issued in September 2018 and noted that for all design-basis scenarios, the station
can maintain SSW inlet temperature under 95 degrees Fahrenheit provided they take the
procedural actions in the USAR and AOP-004, Loss of Offsite Power. These actions are
based on the previous 10 CFR 50.59 changes that were the subject of NCV 05000458/2011008-06. The calculation states that, if both divisions of standby cooling tower
fans start on initiation of the event and no EDG failure occurs, even if the actions in AOP-004
are taken, the inventory will only last 21.4 days. This duration also assumes no SSW
inventory loss through SWP-MOV 96A/B from a delayed closure. When a bounding loss of
SSW is assumed through SWP-MOV 96A/B, where one of the normal service water (NSW) to
SSW isolation valves fails to close remotely and requires local operation (the worst case
single active failure), UHS inventory lasts for 20.4 days.
In summation, the inspectors determined that the licensee failed to restore compliance to
NCV 05000458/2011008-06 because they failed to obtain a license amendment for a change
to licensing basis for the River Bend ultimate heat sink that required prior NRC approval. In
particular, the original RBS license basis stated that the UHS will last 30 days assuming
single active failure with no makeup required. However, the current design basis calculation
G13.18.13.2*086, revision 3 shows inventory will not last 30 days and the current USAR and
AOP-004 allow the licensee to makeup to the UHS to meet the 30-day requirement. The
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allowance for makeup to the UHS was done so without prior NRC approval, contrary to 10 CFR 50.59.
Corrective Actions: The licensee entered this issue into their corrective action program.
Corrective Action References: CR-RBS-2023-07894 and CR-RBS-2023-02086
Performance Assessment: The inspectors determined this violation was associated with a
minor performance deficiency.
Enforcement: The Reactor Oversight Process SDP does not specifically consider the
regulatory process impact in its assessment of licensee performance. Therefore, it is
necessary to address this violation which impedes the NRCs ability to regulate using
traditional enforcement to adequately deter noncompliance.
Severity: Because this performance deficiency had the potential to impact the NRCs ability to
perform its regulatory function, it is necessary to address this violation using traditional
enforcement to adequately deter noncompliance. Using the NRC Enforcement Policy, dated
January 13, 2023, the violation was determined to be Severity Level IV in accordance with
section 6.1.d.2 because it involved a violation of 10 CFR 50.59 where the conditions
associated with the inappropriate change were evaluated as having very low safety
significance (i.e., green) by the SDP.
Violation: Title 10 CFR 50.59(c)(2)(ii) requires, in part, that a licensee shall obtain a license
amendment pursuant to 10 CFR 50.90 prior to implementing a proposed change if the
change would result in more than a minimal increase in the likelihood of occurrence of a
malfunction of a structure, system, or component (SSC) important-to-safety previously
evaluated in the final safety analysis report (as updated).
Contrary to the above, from October 27, 2011, to September 30, 2023, the licensee failed to
obtain a license amendment pursuant to 10 CFR 50.90 prior to implementing a proposed
change, test, or experiment if this activity would result in more than a minimal increase in the
likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated
in the final safety analysis report (as updated). Specifically, the licensee changed the design
basis of the ultimate heat sink inventory requirements from providing a 30-day cooling water
supply without the need for makeup to providing a less than 30-day cooling water supply with
makeup capability, without obtaining a license amendment. The licensee implemented
temporary compensatory measures that would ensure a 30-day ultimate heat sink cooling
water supply.
Enforcement Action: This violation is being cited because the licensee failed to restore
compliance within a reasonable period of time after the violation was identified, consistent
with section 2.3.2 of the Enforcement Policy.
Failure to Maintain Accurate Information in the Updated Safety Analysis Report
Cornerstone Severity Cross-Cutting Report
Aspect Section
Not Severity Level IV Not 71152A
Applicable NCV 05000458/2023003-05 Applicable
Open/Closed
2-25
The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.71(e) when the
licensee failed to update the updated safety analysis report to assure that the information
included in the report contains the latest information developed. Specifically, the licensee
failed to ensure that the design capacity of the diesel generator air start system was updated
to reflect a change to the diesel engine air start capacity and start time requirements.
Description: During a review of the DG air start system, the inspectors noted that the design
capacity of the starting air banks for the standby DG was described in the USAR as capable
of starting the standby DG eight times with five of the starts each within 10 seconds.
Specifically, section 9.5.6 of the USAR revision 27 states:
Each redundant DGSS train is capable of providing the standby diesel generator with eight
starts (five of them are 10 sec starts) from two air receivers without recharging the associated
air receivers.
While the inspectors were researching the licensing basis for the River Bend Station DGs, the
licensee issued EC 93806 to modify USAR section 9.5.6.1.5 to replace the original design
basis of the DG air start system with a revised design basis.
The revised design basis (USAR revision 27) stated:
Each redundant DGSS train is capable of providing the standby diesel generator with five
consecutive starts from two air receivers without recharging the associated air receivers.
It initially appeared that this change was made without prior NRC approval. During follow-up
inspection, the inspectors identified that the NRC approved a change to the diesel air starting
system that corresponded to the change that was implemented under EC 93806 with
Amendment No. 91 to Facility Operating License No. NPF-47, on January 16, 1997.
However, the licensee did not update the USAR until the time of the inspection in 2022.
Corrective Actions: The licensee entered this issue into their corrective action program.
Corrective Action References: CR-RBS-2023-07903
Performance Assessment: The inspectors determined this violation was associated with a
minor performance deficiency.
Enforcement: The ROPs significance determination process does not specifically consider
the regulatory process impact in its assessment of licensee performance. Therefore, it is
necessary to address this violation which impedes the NRCs ability to regulate using
traditional enforcement to adequately deter non-compliance.
Severity: Because this performance deficiency had the potential to impact the NRCs ability to
perform its regulatory function, it is necessary to address this violation using traditional
enforcement to adequately deter noncompliance. In accordance with NRC Enforcement
Policy, section 6.1.d.3, the violation was determined to be severity level IV because the
licensee failed to update the USAR as required by 10 CFR 50.71(e) and the lack of up-to-
date information had a material impact on safety or licensed activities.
Violation: Title 10 CFR 50.71(e) requires, in part, licensees shall update periodically, as
provided in paragraphs (e)(3) and (4) of 10 CFR 50.71, the USAR to assure that the
information included in the report contains the latest information developed.
2-26
Contrary to the above, from January 16, 1997, to September 30, 2023, the licensee failed to
update the USAR to assure that the information included in the report contains the latest
information developed. Specifically, the licensee failed to update the FSAR to reflect the
NRCs approval of River Bend Station, Amendment No. 91 to Facility Operating License No.
NPF-47, issued on January 16, 1997.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with
section 2.3.2 of the Enforcement Policy.
Failure to Follow Procedures for Time Critical Actions
Cornerstone Significance Cross-Cutting Report
Aspect Section
Mitigating Green H.14 - 71152S
Systems NCV 05000458/2023003-06 Conservative
Open/Closed Bias
The inspectors identified a Green finding and associated non-cited violation of
10 CFR Part 50, appendix B, criterion V, Instructions, Procedures, and Drawings, for the
licensees failure to complete required action in accordance with procedure EN-OP-123,
Time Critical Action Program Standard, revision 007. Specifically, the licensee failed to
scope in actions associated with Procedure AOP-0004, Loss of Offsite Power, as time
critical and failed to take all time critical actions for shutting SSW-MOV96, Normal Service
Water Isolation Valve.
Description: While investigating the UHS and SSW systems, the inspectors discovered two
specific cases where the licensee had not followed all of the requirements of EN-OP-123. The
procedure defines time critical actions (TCAs) as a manual action or series of actions, the
performance of which within specified time constraints has been credited in the plant
deterministic design analysis or licensing basis. Section 5.1 requires a list of all sources of
plant-specific TCAs including the Updated Final Safety Analysis Report (USAR).
The first scenario is related to the SSW to NSW isolation valves failing to shut during an
event. Section 9.2.5 of the USAR is for the UHS and 9.2.5.2, system description, describes
the importance of shutting either SSW-MOV96 A or B, the NSW to SSW isolation valves, in
20 minutes to mitigate the loss of SSW inventory to the non-safety related NSW system.
Attachment 6 of EN-OP-123 lists verifying the closure of these valves as a TCA; however,
during the review of the associated TCA conducted on October 17, 2020, the inspector the
inspector identified that the licensee was not validating the local operation of SSW-MOV96 A.
The licensee, in response to questions from the inspector performed a validation, on May 17,
2023, for the local operation of SSW-MOV96 A. During the validation the inspector identified
that multiple different steps of the procedure were not completed for this TCA. Specifically,
EN-OP-123, section 5.6, Validation Requirements, step 3, states to ensure any special
equipment, devices, or supplies required to support the TCA are readily available to operate
SSW-MOV96A.
Specifically, the inspector identified that a step stool or ladder would be required for an
operator to operate the valve. Not having the required equipment staged nearby to complete
the task results in the 20-minute requirement not being met by the time the proper equipment
was retrieved. Additionally, the licensee did not calculate the time required for an operator to
arrive on station or operate the valve under actual system conditions as required per sections
2-27
4.3.b.5, 5.6.3, 5.7.3.b, 5.7.5, and 5.8.4. Additionally, section 5.2 step 9 was also not
completed which requires a rim pull calculation to be done when stroking the valve under
system pressure is not practical.
The second scenario is related to the actions the licensee must take to conserve UHS
inventory to complete the 30-day mission time and prevent the UHS water basin from
exceeding the design limit of 95 degrees Fahrenheit. Section 9.2.5.3 is the safety evaluation
section in the USAR for the UHS. This section required various actions to be completed in 1-2
hours if both divisions of diesels and SSW operate on a LOOP combined with a LOCA. For
this case, the USAR states that various actions need to be taken such as throttling flow to the
RHR heat exchangers. Additionally, abnormal operating procedure AOP-004, Loss of Offsite
Power, revision 73, has section 5.13, Standby Service Water Operations, that has multiple
actions to be completed to conserve UHS inventory when both diesels and SSW systems
start. Examples include step 5.13.2 requiring operators to secure the division 2 RHR heat
exchanger and containment unit cooler as well as throttling flow to the division 1 RHR heat
exchanger. None of the actions associated to both safety trains starting at an event were
scoped into the TCAs and governed by EN-OP-123.
Corrective Actions: The licensee entered this issue into their corrective action program.
Corrective Action References: CR-RBS-2023-00543, EC 54034873
Performance Assessment:
Performance Deficiency: The licensee failed to scope actions in the USAR and AOP-0004 for
preserving UHS water inventory for its 30-day mission time and keeping the basin
temperature under 95 degrees Fahrenheit as TCAs. Additionally, the licensee failed to take
all of the actions quality procedure EN-OP-123 required for the TCA of closing
SWP-MOV96A(B) within 20 minutes of failing to close. Not applying TCA standards to
securing division 2 loads and not completing all required steps for TCAs is within the
licensees ability to foresee and correct and therefore a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor
because it was associated with the Procedure Quality 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.
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, the inspectors determined that the finding was of very low
safety significance (Green) because the finding was not a deficiency affecting the design or
qualification of the system, did not represent a loss of the PRA function of a single train TS
system for greater than its TS-allowed outage time, did not represent a loss of PRA function
of two separate TS systems for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, did not represent a loss of a PRA
system and/or function as defined in the PRIB or licensee's PRA for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,
and did not represent a loss of the PRA function of one or more non-TS trains of equipment
designated as risk-significant in accordance with the licensee's maintenance rule program for
more than 3 days.
Cross-Cutting Aspect: H.14 - Conservative Bias: Individuals use decision making-practices
that emphasize prudent choices over those that are simply allowable. A proposed action is
2-28
determined to be safe in order to proceed, rather than unsafe in order to stop. Specifically,
the licensee failed to identify that the required actions of EN-OP-123 could be performed.
Enforcement:
Violation: Title 10 CFR 50, appendix B, criterion V, Instructions, Procedures, and Drawings,
requires, in part, that activities affecting quality shall be described by documented
instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be
accomplished in accordance with these instructions, procedures, or drawings. The licensee
established Procedure EN-OP-123, Time Critical Action Program Standard, as the
implementing procedure for establishing TCAs that are assumed to be completed within
specified time limits and periodically validates and documents the plant staff capability to
perform such actions within specified time limits and tracks actual performance times, an
activity affecting quality.
Contrary to the above, from September 27, 2022, to September 30, 2023, for TCAs, an
activity affecting quality, the licensee failed to describe by a documented procedure, and
failed to accomplish the activity affecting quality in accordance with a procedure. Specifically,
for procedure EN-OP-123, the TCA implementing procedure, the licensee failed to validate
the proper equipment was staged as directed by step 3 of section 5.6, failed to calculate rim
pull as directed by step 9 of section 5.2, and failed to scope in TCAs directed by the USAR
and AOP-004 into the TCA program as directed by section 5.1.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with
section 2.3.2 of the Enforcement Policy.
Observation: Improper Storage of Equipment in Safety-Related Buildings 71152S
While conducting routine plant status walkdowns, the inspectors made frequent observations
of loose tools, equipment, hoses, etc. being left in various locations in safety-related
buildings. The licensee utilizes work in progress (WIP) signs to identify temporary services
and equipment (TSE) in the plant supporting active work activities. These materials often had
expired WIP signs associated with them or incomplete or inaccurate information on the WIP
signs. When brought to the attention of the control room, condition reports (CRs) were
generated. Occasionally, the equipment would be picked up and stored in its proper location,
but the usual fix was to revise the removal dates on the WIP signs to a new date 90 days out.
Ninety days is the maximum amount of time equipment can be left out without further
evaluation. By continually updating the WIP with new removal dates, multiple staged
equipment locations had material in place that far exceeded the 90-day time without any
evaluation. The residents identified at least ten CRs describing locations where there were
either no WIPs, TSEs were improperly restrained to safety-related piping, or the TSE there
had exceeded 90 days.
The residents determined two specific examples were of note. In the control building AC
rooms, there are two safety-related valves, one in each division, SWP-PVY32A/B. These
valves have associated CRs describing leakage. The valves have a catch installed
underneath them with hoses routed to collection barrels. On multiple occasions the barrels
have been identified to either not be restrained or restrained to safety-related piping. The WO
numbers referenced on the WIP signs were for different valves. Additionally, the catch, hose,
and drum meet the definition of TSE and have been installed in the plant in excess of two
years. When the WIP sign expired, the licensee updated the sign with a new 90-day interval.
ADM-0073 requires an evaluation as a temporary modification or permanent change.
2-29
NUMARC 93-01 requires a 50.59 evaluation for alterations related to a maintenance activity
that exceeds 90 days.
None of the issues rose to more than minor and the resident staff plans to continue to
observe and document any changes to this issue.
Failure to Properly Categorize Standby Service Water Valves in the Inservice Testing
Program
Cornerstone Significance Cross-Cutting Report
Aspect Section
Mitigating Green None (NPP) 71153
Systems NCV 05000458/2023003-07
Open/Closed
The inspectors identified a Green finding and associated non-cited violation of
10 CFR 50.55a(f)(4), In-service Testing Requirements, for the licensees failure to
categorize standby service water isolation valves as Category A valves. Specifically, the
licensees inservice testing program did not test safety-related valves (SWP-MOV-57A and
SWP-MOV-57B) in accordance with ASME OM code Subsection ISTC-1300, Valve
Categories, to ensure they could meet seat leakage requirements. This caused the licensee
to be in violation of Technical Specification 3.7.1 for ultimate heat sink inoperable.
Description: As documented in NRC inspection report 05000458/2022-001 (ML22112A213),
the licensee received a non-cited violation for the failure to correct a condition adverse to
quality. Specifically, the NRC reviewed CR-RBS-2019-06296 which documented an instance
where the SSW C discharge check valve SWP-V148 did not fully seat after a periodic test of
the C SSW pump. The condition was noticed by licensee staff after they observed NSW
system leaking by to the SSW system. Check valve SWP-V148 should fully seat in all
conditions to prevent leak-by that could divert system flow from SSW. The NRC issued the
non-cited violation for the licensees failure to correct the leaking SWP-V148 valve for nearly
three years.
The NRC inspectors questioned why other isolation valves within the SSW system were not
tested for leakage given that the design basis calculation PM-194, Standby Cooling Tower
Performance and Evaporation Losses Without Drywell Coolers, revision 11, identifies that
the maximum allowable leakage for the entire SSW system is 6.9 gallons per minute. After
researching the issue, the inspectors found that since the year 2000, River Bend Station
documented (CR-RBS-2000-01977) the potential for SSW inventory loss through boundary
valves that encroached on the 10 gallon per minute limit specified in PM-194, revision 5
(which was the revision of record at the time). Further, to date, the licensee does not have a
clearly defined population of SSW boundary valves or what the actual leakage rates are (only
the limit of 6.9 gallons per minute).
As stated above, calculation PM-194 assumed that the worst-case leakage from all SSW
boundary valves was 6.9 gallons per minute; however, not all the SSW isolation valves have
been tested for seat leakage. The licensee incorrectly assumed that inservice testing
requirements did not apply to valves SWP-MOV-57A and SWP-MOV-57B, because there
were check valves in the flow path. In other words, the license took credit for the redundant
check valves, though it is possible the check valves themselves could leak, to exclude valves
SWP-MOV-57A and SWP-MOV-57B from the inservice testing program.
2-30
Corrective Actions: In response to this issue, the licensee performed maintenance on
SWP-MOV-57A and SWP-MOV-58B to correct the seat leakage. Additionally, the licensee
re-categorized valves SWP-MOV-57A and SWP-MOV-58B as Category A valves, updated
the inservice testing requirements, and established preventative maintenance tasks for the
valves. Finally, the licensee is developing an engineering assessment to ensure that all the
SSW boundary valves are correctly categorized in the inservice testing program and tested
for leakage. The current worst-case leakage rates of SSW valves do not challenge the
available water inventory for over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Corrective Action References: CR-RBS-2023-01249
Performance Assessment:
Performance Deficiency: The failure to correctly test and classify safety-related valves
SWP-MOV-57A and SWP-MOV-58B as Category A in accordance with 10 CFR 50.55(a)(f)(4)
was a performance deficiency.
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 failure to test for valve leakage on valves
SWP-MOV-57A and SWP-MOV-58B adversely affected the reliability of the SSW system.
When tested, SSW valves SWP-MOV-57A and SWP-MOV-58B had leakage in excess of
40 gallons per minute which challenged the available water inventory in the UHS. This
resulted in the inability for it to meet its 30-day mission time and TS requirements.
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, the inspectors determined that the finding was of very low
safety significance (Green) because the finding was not a deficiency affecting the design or
qualification of the system, did not represent a loss of the PRA function of a single train TS
system for greater than its TS-allowed outage time, did not represent a loss of PRA function
of two separate TS systems for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, did not represent a loss of a PRA
system and/or function as defined in the PRIB or licensee's PRA for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,
and did not represent a loss of the PRA function of one or more non-TS trains of equipment
designated as risk-significant in accordance with the licensee's maintenance rule program for
more than 3 days.
Cross-Cutting Aspect: Not Present Performance. No cross-cutting aspect was assigned to
this finding because the inspectors determined the finding did not reflect present licensee
performance.
Enforcement:
Violation: Title 10 CFR 50.55a(f), Inservice Testing Requirements, subsection (4) requires,
in part, that pumps and valves which are classified as ASME Class 1, Class 2, and Class 3
must meet the inservice test requirements set forth in the ASME OM Code and addenda that
become effective subsequent to editions and addenda specified in paragraphs (f)(2) of this
section and that are incorporated by reference in paragraph (a)(1)(iv) of this section.
Furthermore, subsection (f)(4)(ii) requires inservice tests to verify operational readiness of
pumps and valves, whose function is required for safety, conducted during successive
120-month intervals must comply with the requirements of the latest edition and addenda of
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the ASME OM Code incorporated by reference in paragraph (a)(1)(iv) of this section
18 months before the start of the 120-month interval. ASME Code subsection ISTC-1300,
Valve Categories, requires, in part, that valves within this subsection shall be placed in one
or more of the following categories. Category A is for valves for which the seat leakage is
limited to a specific maximum amount in the closed position for fulfillment of their required
function as specified in ISTA-1100.
TS 3.7.1, Standby Service Water System (SSW) and Ultimate Heat Sink (UHS), applicable
in modes 1, 2, & 3, condition D requires that the UHS shall be operable.
Contrary to the above, from November 2000, to September 30, 2023, the licensee failed to
categorize SSW valves SWP-MOV-57A and SWP-MOV-58B as Category A valves in
accordance with the ASME code requirements. Specifically, by categorizing these valves as
B their leakage was not tested and accounted for to ensure UHS operability. After NRC
questioning, it was discovered the valves leaked more than 40 gallons per minute. This
resulted in the UHS being inoperable for a time longer than permitted by TS, which required
the plant to be in mode 3 within 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br />.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with
section 2.3.2 of the Enforcement Policy.
Engineering Changes Failed to Evaluate Effects on Technical Specification Surveillance
Requirements
Cornerstone Significance Cross-Cutting Report
Aspect Section
Mitigating Green [P.1] - 71153
Systems NCV 05000458/2023003-08 Identification
Open/Closed
The inspectors identified a non-cited violation of 10 CFR Part 50, appendix B, criterion III and
associated technical specification violation of Technical Specification 3.8.1, AC Sources
Operating, when the licensee failed to incorporate original circuitry design into a new digital
control unit upgrade for the control building air conditioning system. This was identified when
the licensee performed STP-309-0602, Division II ECCS Test, revision 056, and the chillers
failed to meet the acceptance criteria and load onto the emergency diesel generator in the
required time in the final safety analysis report.
Description: On February 4, 2023, while the licensee was in mode 4 during refueling
outage 22, they conducted STP-309-0602, Division II ECCS Test, revision 056, to meet TS
SR 3.1.18. The purpose of the test is to simulate a LOOP in conjunction with an ECCS
actuation signal and verify the de-energization of the emergency busses and load shedding of
the emergency busses. Additionally, it is to verify the EDG starts on the auto start signal and
energizes the emergency buses within 10 seconds and then energize the auto connected
loads through the sequencing logic.
Both control building division 2 chillers, HVK-CHL1B and HVK-CHL1D, failed to sequence
onto the emergency bus within the TS required time. Chiller HVK-CHL1B sequenced on the
EDG in 237.1 seconds when the acceptance criteria were to sequence between 180.9 and
221.1 seconds, and HVKl-CHL1D sequenced on in 241 seconds with the same acceptance
criteria. Being outside of the acceptance criteria caused the EDG to fail the surveillance and
therefore be inoperable and enter the LCO, TS 3.8.1. Further investigation revealed that the
division 1 control building chillers, HVK-CHL1A and HVK-CHL1C, had also failed to meet the
load sequence acceptance criteria when they were tested the previous outage.
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The cause of all four chillers from both divisions failing to load onto the emergency bus in the
required time was the result of a modification to upgrade the chiller controllers from analog to
digital. During the upgrade under EC 00031803 completed in July 2012, a delay was
introduced into the start logic that prevented them from meeting the required times. The
changes made from the approved EC were implemented in 2014 and 2020 for division 1 and
2 respectively.
Corrective Actions: The licensee entered this issue into their corrective action program. The
chillers sequence time was corrected under changes to the EC and new WOs.
Corrective Action References: CR-RBS-2023-01153, WO 594333(4)(5)(6), EC 95152
Performance Assessment:
Performance Deficiency: Title 10 CFR Part 50, appendix B, criterion III, Design Control,
requires, that design changes, including field changes, be subject to design control measures
commensurate with those applied to the original design. When implementing
EC 0000311803, a design change, the licensee failed to incorporate the original design
requirements for the HVK system into the EC and was therefore a performance deficiency.
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, by not loading onto the emergency busses within the
required times, the HVK chillers could cause a trip of the EDGs.
Significance: The inspectors assessed the significance of the finding using IMC 0609
appendix A, The Significance Determination Process (SDP) for Findings At-Power.
Specifically, using exhibit 2, Mitigating Systems Screening Questions, the inspectors
determined that this finding is of very low safety significance (Green) because the impacted
SSCs maintained their operability and PRA functionality.
Cross-Cutting Aspect: P.1 - Identification: The organization implements a corrective action
program with a low threshold for identifying issues. Individuals identify issues completely,
accurately, and in a timely manner in accordance with the program. During previous
executions of the ECCS surveillance, the HVK chillers failed to meet acceptance criteria and
loaded onto the emergency busses outside of the required times. The licensee
inappropriately used a TS bases to call the EDG operable in each instance. This was a
missed opportunity to investigate why the acceptance criteria was not met and ultimately
identify the error in the digital control system EC.
Enforcement:
Title 10 CFR Part 50, appendix B, criterion III, Design Control, requires, in part, that design
changes, including field changes, be subject to design control measures commensurate with
those applied to the original design.
SR 3.0.3 requires that if a SR is not met then the LCO is not met. Since the SR acceptance
criteria was not met, the SR was failed and therefore TS 3.8.1, AC Sources Operating, was
not met due to Condition D not having actions completed.
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Contrary to the above, from July 10, 2012, to April 24, 2023, the licensee performed a design
change and failed to subject it to design control measures commensurate with those applied
to the original design. Specifically, when implementing EC 000031803, the licensee failed to
incorporate the original design requirements for the HVK system into section 3.1.25 specific
circuit logic that would have temporarily bypassed the start signal for the HVK chillers until
SSW flow was adequate thus preventing unnecessary trips that prevented the system from
loading onto the EDG per its prescribed sequence time. This resulted in a condition that
prevented a SR for TS 3.8.1 from being met and therefore inoperable for longer than time
permitted by TS.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with
section 2.3.2 of the Enforcement Policy.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
- On July 18, 2023, the inspectors presented the occupational and public radiation safety
inspection results to Phil Hansett, Site Vice President, and other members of the
licensee staff.
- On August 17, 2023, the inspectors presented the emergency preparedness program
inspection results to Phil Hansett, Site Vice President, and other members of the
licensee staff.
- On October 18, 2023, the inspectors presented the integrated inspection results to Phil
Hansett, Site Vice President, and other members of the licensee staff.
- On November 13, 2023, the inspectors presented the integrated inspection results to
Phil Hansett, Site Vice President, and other member of the licensee staff.
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DOCUMENTS REVIEWED
Inspection Type Designation Description or Title Revision or
Procedure Date
71111.04 Procedures EN-DC-355 Implementation of the Technical Specification Surveillance 005
Frequency Control Program
SOP-0031 Residual Heat Removal 345
SOP-0049 125VDC System 048
71111.06 Calculations PN-314 Moderate Energy Line Crack Flow Rates 0
PN-315 MELC Maximum Discharge and Suction Line Pressures for 0
RCIC, HPCS, LPCS, and RHR Fill Pumps
PN-316 Maximum Leak Rate for a Moderate Energy Line Crack 0
MELC During RHR Shutdown Cooling
PN-317 MELC-Max Flood Elevations for Moderate Energy Line 002
Cracks in Cat 1 Structures
71111.12 Corrective Action CR-RBS- 2022-05212, 2023-06233
Documents
Engineering EC 93741
Changes
Procedures EN-DC-204 Maintenance Rule Scope and Basis 008
EN-DC-205 Maintenance Rule Monitoring 009
71111.13 Corrective Action CR-RBS- 2023-05428
Documents
Procedures ADM-0096 Risk Management Program Implementation and On-Line 338
Maintenance Risk Assessment
EN-OP-119 Protected Equipment Postings 016
STP-051-4279 Containment Unit Cooler System Instrumentation Unit Cooler 11
A Functional Test
STP-051-4522 ECCS RCIC Reactor Vessel Water Level Channel Functional 14
Test (B21-692A, B21-691A)
STP-051-4524 ECCS RCIC Reactor Vessel Water Level Channel Functional 8
Test (B21-N692E, B21-691E)
71111.15 Procedures AOP-0004 Loss of Offsite Power 073
EN-OP-104 Operability Determination Process 018
EN-OP-123 Time Critical Action Program Standard 007
71111.24 Corrective Action CR-RBS- 2023-04358, 2023-06044, 2023-06570
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Inspection Type Designation Description or Title Revision or
Procedure Date
Documents
71114.02 Miscellaneous River Bend Station Alert and Notification System Design 01/04/2013
Report
Procedures EPP-2-701 Prompt Notification System Maintenance and Testing 34
71114.03 Miscellaneous ERO Notification System Test Results 03/23/2022
ERO Notification System Test Results 09/20/2022
ERO Notification System Test Results 06/21/2023
ERO Notification System Test Results 12/17/2021
River Bend Station On-Shift Staffing Analysis Final Report 1
Procedures EIP-2-006 Notifications 49
EN-EP-310 Emergency Response Organization Notification System 11
EPP-2-502 Emergency Communications Equipment Testing 29
71114.04 Miscellaneous 10CFR50.54(Q)(2) Review for EN-EP-202 Revision 3 07/26/2021
10CFR50.54(Q)(2) Review for EIP-2-024 Revision 27 01/12/2023
10CFR50.54(Q)(3) Screening for PR-C-422 09/12/2022
10CFR50.54(Q)(2) Review for EIP-2-016 Revision 32 11/01/2021
Procedures EN-EP-305 Emergency Planning 10CFR50.54(q) Review Program 8
71114.05 Corrective Action CR-RBS- 2021-04918, 2021-05407, 2021-05963, 2021-06899, 2022-
Documents 02007, 2022-02599, 2022-03308, 2022-03409, 2022-03410,
2022-03629, 2022-05741, 2022-05992, 2022-06138, 2022-
02420, 2022-06185, 2023-00407, 2023-02200, 2023-02207,
2021-03346, 2021-03355, 2021-03512, 2021-03515
Corrective Action CR-RBS- 2023-06592, 2023-06596
Documents
Resulting from
Inspection
Miscellaneous ERO Team B Site Drill Report 10/13/2021 11/12/2021
ERO Team B Site Drill Report 11/10/2021 03/01/2022
ERO Team C Site Drill Report 02/23/2022 03/24/2022
2022 NRC IPX Exercise Drill Report 11/14/2022
Our Lady of the Lake MS-1 Drill Report 03/16/2022
Medical Emergency Drill On-Site 12/13/2022
Team B EOF Focused Drill Report 10/05/2022 10/06/2022
2-36
Inspection Type Designation Description or Title Revision or
Procedure Date
Team D EOF Focused Drill Report 10/12/2022 10/24/2022
2022 EMPE Drill Report 06/15/2022 07/14/2022
2022 Accountability Drill Report 12/22/2022 01/10/2023
HP Drill Report 06/22/2023 07/18/2023
2nd Quarter Team C Focused Drill Report 06/28/2023 07/12/2023
QA-7-2021-RBS- Emergency Preparedness Quality Assurance Audit Report 06/28/2020
1
QA-7-2023-RBS- Emergency Preparedness Quality Assurance Audit Report 04/28/2022
1
QS-2022-RBS- Emergency Preparedness Quality Assurance Surveillance 04/12/2022
002 Report
Procedures ADM-0060 First Aid Team Emergencies 16
EIP-2-001 Classification of Emergencies 30
EIP-2-002 Classification Actions 37
EIP-2-007 Protective Action Recommendation Guidelines 28
EN-EP-306 Drills and Exercises 11
EN-EP-609 Emergency Operations Facility (EOF) Operations 7
EN-EP-610 Technical Support Center (TSC) Operations 8
EN-EP-611 Operations Support Center Operations 8
EN-TQ-110 Emergency Response Organization Training 15
EPP-2-501 Emergency Facilities and Equipment Readiness 18
Work Orders WO-545130
71124.05 Corrective Action CR-RBS- 2021-03740, 2021-04317, 2021-05817, 2021-06840, 2022-
Documents 02036, 2022-04484, 2022-4618, 2022-4860, 2022-06323,
2022-06714, 2022-0621, 2023-04854
Corrective Action CR-RBS- 2023-05769, 2023-05771
Documents
Resulting from
Inspection
Miscellaneous Table of Routine Source Check Sources, Assay Date, and 06/12/2023
Current Activity
Procedures EN-RP-302 Operation of Radiation Protection Instrumentation 6
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Inspection Type Designation Description or Title Revision or
Procedure Date
EN-RP-306 Calibration and Operation of the Eberline PM-7 3
EN-RP-307 Operation and Calibration of the Eberline Personal 2
Contamination Monitors
EN-RP-308 Operation and Calibration of Gamma Scintillation Tool 9
Monitors
EN-RP-312 Operation and Calibration of the Canberra GEM-5 4
EN-RP-313 Operation and Calibration of the ARGOS-5AB Personnel 3
Contamination Monitor
EN-RP-315 Operation and Calibration of the CRONOS Contamination 3
Monitor
EN-RP-317-05 Calibration of Extendable Dose Rate Instruments 1
EN-RP-317-09 Calibration of Dosimeters 5
EN-RP-317-10 Calibration of Portable Dose Rate Instruments 3
RHP-0106 Calibration of the Canberra Fastscan and Accuscan II Whole 4
Body Counters
STP-511-4237 Main Plant Exhaust Duct Monitoring System Flow Rate 305
Monitor Channel Calibration RMS-FEX125, RMS-FEY125
Radiation RBS-2306-00116 3119 TB 95 Chemistry Sample Area 06/07/2023
Surveys RBS-2306-00400 3000 Turbine Building 67-foot Elevation 06/26/2023
Work Orders Work Order (WO) 52769324-01, 52948224-01, 00565523-01, 00585136-01,
00596070-01,
71124.08 Corrective Action CR-RBS- 2022-00723, 2022-04574, 2023-01499, 2023-02069, 2023-
Documents 02738
Corrective Action CR-RBS- 2023-05914, 2023-05940, 2023-05729, 2023-05730, 2023-
Documents 05732, 2023-05735, 2023-05930, 2023-05931, 2023-05933
Resulting from
Inspection
Miscellaneous Dry active waste, waste stream analysis 04/17/2023
Reactor water cleanup system waste stream analysis 08/09/2022
Radioactive source inventory list 06/27/2023
FCBT-MPC- Shipping training for radwaste workers 1
HAZSEC
FCBT-MPC-TSP Shipping security training for radwaste workers 3
Procedures EN-RP-121 Radioactive Material Control 18
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Inspection Type Designation Description or Title Revision or
Procedure Date
EN-RW-102 Radioactive Shipping Procedure 20
EN-RW-104 Scaling Factors 14
EN-RW-105 Process Control Program 5
EN-RW-106 Integrated Transportation Security Plan 7
SOP-0112 Solid Radwaste Processing 22
SOP-012 Solid Radwaste Collection 8
Self-Assessments LO-RLO-2020- RP Pre-NRC focused self-assessment plan 02/16/2021
00018
QA-14/15-2021- Combined Radiation Protection (RP) and Radwaste (RW) 09/16/2021
RBS-01 audit
Shipping Records RBS-2022-031 Shipping package for shipment number RBS-2022-031
RBS-2022-034 Shipping package for shipment number RBS-2022-034
RBS-2022-059 Shipping package for shipment number RBS-2022-059
RBS-2023-040 Shipping package for shipment number RBS-2023-040
RBS-2023-041 Shipping package for shipment number RBS-2023-041
71151 Miscellaneous Alert and Notification System Test Data 4Q2022 - 2Q2023
Drill and Exercise Performance Data 4Q2022 - 2Q2023
Emergency Response Organization Drill Participation Data
4Q2022 - 2Q2023
71152S Procedures ADM-0073 Temporary Services and Equipment 307
EDS-ME-002 Control of Loose Items 002
EN-OP-115 Conduct of Operations 031
EN-OP-115-09 Maintaining the Station Narrative Log 004
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