ML24131A151
| ML24131A151 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 05/13/2024 |
| From: | Robert Ruiz NRC/RGN-III/DORS/RPB1 |
| To: | Rhoades D Constellation Energy Generation, Constellation Nuclear |
| References | |
| IR 2024001 | |
| Download: ML24131A151 (1) | |
See also: IR 05000373/2024001
Text
David P. Rhoades
Senior Vice President
Constellation Energy Generation, LLC
President and Chief Nuclear Officer (CNO)
Constellation Nuclear
4300 Winfield Road
Warrenville, IL 60555
SUBJECT:
LASALLE COUNTY STATION - INTEGRATED INSPECTION REPORT
05000373/2024001 AND 05000374/2024001
Dear David P. Rhoades:
On March 31, 2024, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection
at LaSalle County Station. On April 17, 2024, the NRC inspectors discussed the results of this
inspection with Christopher Smith, Plant Manager, and other members of your staff. The results
of this inspection are documented in the enclosed report.
Four findings of very low safety significance (Green) are documented in this report. Four of
these findings involved violations of NRC requirements. We are treating these violations as
non-cited violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.
If you contest the violations or the significance or severity of the violations documented in this
inspection report, you should provide a response within 30 days of the date of this inspection
report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission,
ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional
Administrator, Region III; the Director, Office of Enforcement; and the NRC Resident Inspector
at LaSalle County 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 III; and the
NRC Resident Inspector at LaSalle County Station.
May 13, 2024
D. Rhoades
2
This letter, its enclosure, and your response (if any) will be made available for public inspection
and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document
Room in accordance with Title 10 of the Code of Federal Regulations 2.390, Public
Inspections, Exemptions, Requests for Withholding.
Sincerely,
Robert Ruiz, Chief
Reactor Projects Branch 1
Division of Operating Reactor Safety
Docket Nos. 05000373 and 05000374
License Nos. NPF-11 and NPF-18
Enclosure:
As stated
cc w/ encl: Distribution via LISTSERV
Signed by Ruiz, Robert
on 05/13/24
SUNSI Review
Non-Sensitive
Sensitive
Publicly Available
Non-Publicly Available
OFFICE
RIII
EICS
RIII
NAME
NShaw:sw
GEdwards via email
RRuiz
DATE
05/13/2024
05/07/2024
05/13/2024
Enclosure
U.S. NUCLEAR REGULATORY COMMISSION
Inspection Report
Docket Numbers:
05000373 and 05000374
License Numbers:
Report Numbers:
05000373/2024001 and 05000374/2024001
Enterprise Identifier:
I-2024-001-0074
Licensee:
Constellation Nuclear
Facility:
LaSalle County Station
Location:
Marseilles, IL
Inspection Dates:
January 01, 2024 to March 31, 2024
Inspectors:
J. Benjamin, Senior Resident Inspector
J. Cassidy, Senior Health Physicist
T. Hooker, Health Physicist
J. Meszaros, Resident Inspector
E. Rosario, Reactor Inspector
N. Shah, Senior Project Engineer
A. Shaikh, Senior Reactor Inspector
M. Siddiqui, Allegations/Enforcement Specialist
C. St. Peters, Resident Inspector
Approved By:
Robert Ruiz, Chief
Reactor Projects Branch 1
Division of Operating Reactor Safety
2
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees
performance by conducting an integrated inspection at LaSalle County 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 Follow Seismic Storage Requirements for a Temporary Battery Charger Stacked on
an Unrestrained Cart Configuration Adjacent to the Safety-Related Unit 1 125Vdc Division 2
Battery Rack and Exposed Terminals
Cornerstone
Significance
Cross-Cutting
Aspect
Report
Section
Mitigating
Systems
Green
Open/Closed
[H.8] -
Procedure
Adherence
The inspectors identified a Green finding and an associated non-cited violation (NCV) of
10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, when the
licensee failed to follow the requirements of station procedure LAP-100-56, Equipment Parts
Storage in Plant Areas Containing Safety-Related Equipment, Revision 10. Specifically, the
licensee failed to ensure that a portable battery charger stacked on the top of a mobile cart
with a combined height to width ratio of greater than 2.0 was either stored in an approved
seismic storage area, seismically restrained, approved by engineering, or stored at least a
distance of 2 feet plus the height of the stacked configuration away from the safety-related
batteries with exposed terminals in accordance with the procedure requirements.
Failure to Test Motor-Operated Valve in Accordance with the Inservice Test Program
Cornerstone
Significance
Cross-Cutting
Aspect
Report
Section
Green
Open/Closed
[H.11] -
Challenge the
Unknown
The inspectors identified a finding of very low safety significance (Green) and a non-cited
violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50.55a(f)(4)(ii)
for the licensees failure to meet the in-service testing requirements set forth in the American
Society of Mechanical Engineers (ASME) Operations and Maintenance Code and Addenda
Code Case OMN-1 after performing maintenance that could affect motor-operated valve
(MOV) performance. Specifically, the licensee failed to perform testing on primary
containment isolation MOV, 2B21-F016, prior to returning the valve to service after electrically
backseating the valve, which was maintenance that could affect the valves performance.
3
Failure to Use Calibrated Measuring and Test Equipment to Electrically Backseat a Safety
Related Motor-Operated Valve
Cornerstone
Significance
Cross-Cutting
Aspect
Report
Section
Green
Open/Closed
[H.11] -
Challenge the
Unknown
The inspectors identified a finding of very low safety significance (Green) and a NCV of
Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion XII,
Control of Measuring and Test Equipment, for the licensees failure to assure a tool used in
activities affecting quality was properly controlled, calibrated, and adjusted at specified
periods to maintain accuracy within necessary limits. Specifically, the licensee failed to apply
quality assurance requirements to the MOV BSRT when using the tool to electrically backseat
a Unit 2 safety-related valve.
Failure to Promptly Correct Degraded Pressure Switches in the Unit 1 and Unit 2 Main Steam
Line High Flow Isolation Logic System
Cornerstone
Significance
Cross-Cutting
Aspect
Report
Section
Green
NCV 05000373,05000374/2024001-04
Open/Closed
None (NPP)
The inspectors identified a Green finding and associated non-cited violation of
10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to
promptly correct a condition adverse to quality associated with the Unit 1 and Unit 2 main
steam line (MSL) isolation logic system. Specifically, the licensee identified in a 2008
equipment apparent cause evaluation that pressure switches installed in the Unit 1 and Unit 2
MSL high-flow isolation logic trip systems were susceptible to multiple failure modes and had
been exposed to peak inductive currents during frequent calibration activities that may have
damaged switch contacts. A subsequent corrective action replaced half of the impacted
switches while the other half continued to be replaced on an as-needed basis.
Additional Tracking Items
Type
Issue Number
Title
Report Section
Status
05000373,05000374/20
23002-01
Clarification of the Dew Point
Specification for the MSA
Firehawk M7 SCBA System
Closed
4
PLANT STATUS
Unit 1 began the inspection period at rated thermal power. On February 19, 2024, the unit
coasted down to approximately 86 percent power and commenced refueling outage L1R20.
On March 8, 2024, the unit started up and reached rated thermal power on March 13, 2024.
On March 17, 2024, the unit was down powered to approximately 60 percent to perform power
suppression testing for a suspected fuel cladding failure. The suspected fuel failure was
suppressed and later that day, the unit returned to rated thermal power and remained at or near
rated thermal power for the duration of the inspection period.
Unit 2 began the inspection period at rated thermal power. On January 12, 2024, the unit was
down powered to approximately 80 percent for a control rod sequence exchange. The unit was
returned to rated thermal power on the same day and remained at rated thermal power for the
remainder of the inspection period.
INSPECTION SCOPES
Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in
effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with
their attached revision histories are located on the public website at http://www.nrc.gov/reading-
rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared
complete when the IP requirements most appropriate to the inspection activity were met
consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection
Program - Operations Phase. The inspectors performed activities described in IMC 2515,
Appendix D, Plant Status, observed risk significant activities, and completed on-site portions of
IPs. The inspectors reviewed selected procedures and records, observed activities, and
interviewed personnel to assess licensee performance and compliance with Commission rules
and regulations, license conditions, site procedures, and standards.
REACTOR SAFETY
71111.01 - Adverse Weather Protection
Seasonal Extreme Weather Sample (IP Section 03.01) (1 Sample)
(1)
The inspectors evaluated the licensees readiness for extreme cold weather condition
preparation for the following systems:
1.
unit common A train firewater pump
2.
cooling water intake systems
3.
FLEX buildings
71111.04 - Equipment Alignment
Partial Walkdown Sample (IP Section 03.01) (2 Samples)
The inspectors evaluated system configurations during partial walkdowns of the following
systems/trains:
(1)
1A residual heat removal system on January 4, 2024, after quarterly pump run
(2)
Unit 1 and Unit 2 station air following the trip of a station air compressor on
March 26, 2024
5
71111.05 - Fire Protection
Fire Area Walkdown and Inspection Sample (IP Section 03.01) (9 Samples)
The inspectors evaluated the implementation of the fire protection program by conducting a
walkdown and performing a review to verify program compliance, equipment functionality,
material condition, and operational readiness of the following fire areas:
(1)
Fire Zone 312, reactor building, elevation 673'-4", Unit 2 high-pressure core spray
cubicle on January 17, 2024
(2)
Fire Zone 3H2, reactor building, elevation 694'-6", Unit 2 high-pressure core spray
cubicle on January 19, 2024
(3)
Fire Zone 2K, reactor building, elevation 687'-0" to 768'-0", Unit 1 steam tunnel on
February 22, 2024
(4)
Unit 2 auxiliary electrical equipment room on March 13, 2024
(5)
Fire Zone 3B1, reactor building, elevation 820'-6", Unit 2 general area and standby
gas treatment area on March 13, 2024
(6)
Fire Zone 2B1, reactor building, elevation 820'-6", Unit 1 general area and standby
gas treatment area on March 13, 2024
(7)
Fire Zone 5D3, elevation 687'-0", Unit 1 high-pressure core spray switchgear area on
March 14, 2024
(8)
Fire Zone 5D2, elevation 687'-0", Unit 2 high-pressure core spray switchgear area on
March 14, 2024
(9)
Fire Zone 3I4, reactor building, Unit 2 low-pressure coolant system (LPCS)/reactor
core isolation cooling (RCIC) pump cubicle room on March 29, 2024
71111.06 - Flood Protection Measures
Flooding Sample (IP Section 03.01) (1 Sample)
(1)
The inspectors evaluated the temporary floor drain plugging configuration in the
turbine building to support the movement of 250Vdc batteries (RE: Engineering
Change 640690).
71111.08G - Inservice Inspection Activities (BWR)
BWR Inservice Inspection Activities Sample - Nondestructive Examination and Welding
Activities (IP Section 03.01) (1 Sample)
The inspectors evaluated boiling-water reactor nondestructive testing by reviewing the
following examinations from February 26-March 1, 2024:
(1)
1.
visual examination of residual heat removal (RHR) constant support
RH40-1004C
2.
visual examination of MS restraint MS01-1352X
3.
ultrasonic examination of reactor pressure vessel (RPV) recirculation outlet
nozzle inner radius 1-NIR-10
4.
ultrasonic examination of RPV core spray nozzle inner radius 1-NIR-16
5.
ultrasonic examination of RPV bottom head meridional welds GEL-1006 DB,
6.
weld repair on Unit 1 RHR HX weld 1RH-HX1B-9A
6
71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance
Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01)
(1 Sample)
(1)
The inspectors observed and evaluated licensed operator performance in the control
room during reactor startup from outage L1R20 on March 8, 2024.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
(1)
The inspectors observed and evaluated license operator requalification simulator
out-of-box evaluation on March 19, 2024.
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management Sample (IP Section 03.01) (4 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)
Unit 2 elevated action Green risk due to main steam line high-flow channel spurious
trips on January 17, 2024
(2)
Unit 2 L1R20 shutdown safety plan
(3)
Unit 2 planned elevated action Green risk for behind-the-meter related activities on
February 1, 2024
(4)
Unit 1 planned reduced inventory to support head removal on February 20, 2024
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (8 Samples)
The inspectors evaluated the licensees justifications and actions associated with the
following operability determinations and functionality assessments:
(1)
Unit 2 channel A1 main steam isolation valve high-flow isolation operability between
spurious signals on January 9, 2024, and January 12, 2024
(2)
operability assessment of the Unit 2B emergency diesel generator after
small amounts of metal debris were observed under the top deck cover on
January 23, 2024
(3)
Unit 2 RCIC issues with manual control
(4)
Unit 2 main stem line valve drain inboard isolation valve 2B21-F016 back seating
activity to reduce suspected packing leakage
(5)
Unit 1 1B reactor recirculation seal pressure temperature and pressure fluctuations
(6)
Unit 1 1A reactor recirculation seal pressure temperature and pressure fluctuations
(7)
Unit 1 L1R20 water rod inspection identification of three fuel spacers that shifted up
near the upper core plate
(8)
2CM023B operability as a primary containment isolation valve and associated
containment monitoring channel operability given continued indication and
manufacturing issues
7
71111.18 - Plant Modifications
Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02)
(1 Sample)
(1)
Unit 2 main steam inboard drain primary containment isolation valve backseat to
reduce suspected packing leakage
71111.20 - Refueling and Other Outage Activities
Refueling/Other Outage Sample (IP Section 03.01) (1 Sample)
(1)
The inspectors evaluated Unit 1 refueling outage L1R20 related activities from
February 19, 2024, to March 8, 2024.
71111.24 - Testing and Maintenance of Equipment Important to Risk
The inspectors evaluated the following testing and maintenance activities to verify system
operability and/or functionality:
Post-Maintenance Testing (PMT) (IP Section 03.01) (4 Samples)
(1)
Unit 1 division 2 safety-related battery cell 56 equalizing charge on January 31, 2024
(2)
Unit 2 main stem line valve drain inboard isolation valve 2B21-F016 back seating
(3)
Unit 1 reactor vessel leakage test on March 5, 2024
(4)
Unit 1 1A emergency diesel generator fast start on March 29, 2024
Surveillance Testing (IP Section 03.01) (5 Samples)
(1)
Unit 2 main steam line high-flow main steam isolation valve (MSIV) isolation
calibration per LIS-MS-202 on January 9, 2024
(2)
Unit 1 MSIV stroke time testing on February 19, 2024
(3)
Unit 1 reactor recirculation flow control valve lockup testing on February 19, 2024
(4)
Unit 1 integrated division 1 response time surveillance per LOS-DG-109 on
February 21, 2024
(5)
Unit 2 2A emergency diesel generator idle start testing on March 15, 2024
Inservice Testing (IST) (IP Section 03.01) (2 Samples)
(1)
Unit 2 2A standby liquid control pump on March 15, 2024
(2)
Unit 2 2B emergency diesel generator cooling water pump comprehensive inservice
test on March 18, 2024
8
RADIATION SAFETY
71124.01 - Radiological Hazard Assessment and Exposure Controls
Radiological Hazard Assessment (IP Section 03.01) (1 Sample)
(1)
The inspectors evaluated how the licensee identifies the magnitude and extent of
radiation levels, the concentrations and quantities of radioactive materials, and how
the licensee assesses radiological hazards.
Instructions to Workers (IP Section 03.02) (1 Sample)
(1)
The inspectors evaluated how the licensee instructs workers on plant-related
radiological hazards and the radiation protection requirements intended to protect
workers from those hazards.
Contamination and Radioactive Material Control (IP Section 03.03) (2 Samples)
The inspectors observed/evaluated the following licensee processes for monitoring and
controlling contamination and radioactive material:
(1)
licensee surveys and decontamination of potentially contaminated material leaving
the main radiologically controlled area
(2)
licensee surveys and decontamination of potentially contaminated material leaving
the radiologically controlled area at the north service building
Radiological Hazards Control and Work Coverage (IP Section 03.04) (4 Samples)
The inspectors evaluated the licensees control of radiological hazards for the following
radiological work:
(1)
Unit 1 undervessel installation
(2)
Unit 1 high-pressure turbine diaphragm inspection
(3)
refueling floor cavity platform activities
(4)
Unit 1 F004 valve room activities
High Radiation Area and Very High Radiation Area Controls (IP Section 03.05) (3 Samples)
The inspectors evaluated licensee controls of the following high radiation areas and very
(1)
Unit 1 reactor building personnel access
(2)
Unit 1 turbine deck center court
(3)
Unit 1 drywell entry
Radiation Worker Performance and Radiation Protection Technician Proficiency (IP Section
03.06) (1 Sample)
(1)
The inspectors evaluated radiation worker and radiation protection technician
performance as it pertains to radiation protection requirements.
9
71124.03 - In-Plant Airborne Radioactivity Control and Mitigation
Temporary Ventilation Systems (IP Section 03.02) (1 Sample)
The inspectors evaluated the configuration of the following temporary ventilation systems:
(1)
LAS-416 HEPA unit
Use of Respiratory Protection Devices (IP Section 03.03) (1 Sample)
(1)
The inspectors evaluated the licensees use of respiratory protection devices.
71124.08 - Radioactive Solid Waste Processing & Radioactive Material Handling,
Storage, & Transportation
Shipment Preparation (IP Section 03.04) (1 Sample)
(1)
The inspectors observed the preparation of radioactive shipment LM24-003 of control
rod drive boxes on February 29, 2024.
OTHER ACTIVITIES - BASELINE
71151 - Performance Indicator Verification
The inspectors verified licensee performance indicators submittals listed below:
IE01: Unplanned Scrams per 7000 Critical Hours Sample (IP Section 02.01) (2 Samples)
(1)
Unit 1 (January 1, 2023, through December 31, 2023)
(2)
Unit 2 (January 1, 2023, through December 31, 2023)
IE03: Unplanned Power Changes per 7000 Critical Hours Sample (IP Section 02.02)
(2 Samples)
(1)
Unit 1 (January 1, 2023, through December 31, 2023)
(2)
Unit 2 (January 1, 2023, through December 31, 2023)
IE04: Unplanned Scrams with Complications (USwC) Sample (IP Section 02.03) (2 Samples)
(1)
Unit 1 (January 1, 2023, through December 31, 2023)
(2)
Unit 2 (January 1, 2023, through December 31, 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 issues:
10
(1)
The inspectors evaluated spurious Unit 2 channel A1 main steam line isolation valve
isolation signals that occurred on several dates as documented in Action Requests
(ARs) 4730751 and 4727857. A finding of very low safety significance (Green) and
non-cited violation is associated with the inspectors review and are documented in
the Inspection Results section of this report.
71153 - Follow Up of Events and Notices of Enforcement Discretion
Event Report (IP Section 03.02) (1 Sample)
The inspectors evaluated the following licensee event reports (LERs):
(1)
LER 05000374/2023-003-00, LaSalle County Station, Unit 2, Automatic Actuation of
Reactor Protection System (RPS) During Restoration from Hydrostatic Test
Conditions (ADAMS Accession No. ML23122A016). The inspectors determined that
the cause of the condition described in the LER does not represent a finding because
the unit was already shut down at the time of actuation and the RPS signal was not
applicable. This LER is closed.
INSPECTION RESULTS
Failure to Follow Seismic Storage Requirements for a Temporary Battery Charge Stacked on
an Unrestrained Cart Configuration Adjacent to the Safety-Related Unit 1 125Vdc Division 2
Battery Rack and Exposed Terminals
Cornerstone
Significance
Cross-Cutting
Aspect
Report
Section
Mitigating
Systems
Green
Open/Closed
[H.8] -
Procedure
Adherence
The inspectors identified a Green finding and an associated non-cited violation (NCV) of
10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, when the
licensee failed to follow the requirements of station procedure LAP-100-56, Equipment Parts
Storage in Plant Areas Containing Safety-Related Equipment, Revision 10. Specifically, the
licensee failed to ensure that a portable battery charger stacked on the top of a mobile cart
with a combined height to width ratio of greater than 2.0 was either stored in an approved
seismic storage area, seismically restrained, approved by engineering, or stored at least a
distance of 2 feet plus the height of the stacked configuration away from the safety-related
batteries with exposed terminals in accordance with the procedure requirements.
Description:
During the walkdown portion of this inspection, the inspectors identified that a portable battery
charger stacked on a mobile cart was within in very close proximity to the Unit 1 division 2
125Vdc safety-related batteries. The licensee was using the temporary battery charger for an
extended single charge of cell #56 per licensee procedure LEP-DC-01. The inspectors
identified a concern that if a seismic event occurred, or if bumped, that the stacked metal
frame battery charger unit could fall off the plastic cart and onto the exposed battery terminals
resulting in an arc event and loss of 125Vdc battery charger or batteries themselves. The
inspectors immediately notified the Unit 1 senior reactor operator of their concerns. The
licensee promptly secured the single cell charge, removed the cart and battery charger from
the 125Vdc division 2 battery room. The licensee discussed the station storage requirements
11
and expectations with the appropriate site personnel and entered this issue into their
corrective action program (AR 4739747). The licensee concluded that this configuration did
not meet station requirements and expectations, however, the battery and associated
permanent battery charger remained operable in this configuration based upon their
judgement.
The inspectors reviewed the licensee corrective action document and procedure that
maintains structure, system, and component quality with respect to equipment storage,
LAP-100-56, revision 10, Equipment/Parts Storage in Plant Areas Containing Safety-Related
Equipment. A specific purpose listed in the procedure was to define the LaSalle station
requirements for equipment/parts storage in plant areas containing safety-related
equipment. The inspectors approximated the height (H) to weight (W) ratio to be 6 based
upon the smallest width of the battery charger combined with the total height of the stacked
configuration to be approximately 48 inches. The inspectors approximated the distance from
the stacked metal frame battery charge to be approximately 1 foot higher and 10 inches from
the open battery terminals. The inspectors identified the specific procedural requirements not
met included:
Specifically (Ref: LAP-100-56, revision 10):
B.4.2.1 Stored items with a height (H) to width (W) ration (H/W) greater than 2.0 may
require a seismic restraint.
B.4.2.2 If H/W is greater than 2.0, proceed to B.4.4 . . .
B.4.3 Stored items with H/W < 2.0. . . .
B.4.4 Stored items with H/W > 2.0.
B.4.4.1 Items stored inside or outside one of the approved storage areas.
B.4.4.1.1 If the edge of the stored item is located at least 24-inches plus the height (H) of
the stored item away from the nearest piece of safety-related equipment/component. . .
otherwise proceed to B.4.6 (seismic restraint required).
B.4.6 Seismic Restraint
B.4.6.1 All stored items not meeting the requirements of B.4.3 and B.4.4 shall require
seismic restraints unless otherwise approved by Engineering. . .
Procedure LAP-100-56, revision 10, attachment, provides an activity flow chart which restates
the procedural requirements in a different format.
Corrective Actions: The licensee entered this condition into the stations corrective action
program AR 4739747. Corrective actions include promptly securing the single cell charger
and relocating the cart and stacked battery charger out of the Unit 1 125Vdc division 2 battery
room, individual coaching, and an assignment to determine if the procedure could be
enhanced.
Corrective Action References: AR 4739747
12
Performance Assessment:
Performance Deficiency: The licensees failure to follow station procedure LAP-100-56,
Equipment Parts Storage in Plant Areas Containing Safety-Related Equipment, Revision 10,
was a performance deficiency. Specifically, the procedure accurately captured the licensee
requirements in both the procedure body and associated attachment A and therefore was
reasonable for the licensee to foresee and prevent.
Screening: The inspectors determined the performance deficiency was more than minor
because it was associated with the Configuration 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, one of the purposes of licensee procedure LAP-100-56 is to
ensure safety-related components such as the Unit 1 division 2 125Vdc battery remain
available, reliable, and capable, during design-basis events. The cart and stacked battery
charger were in an unrestrained configuration such that if a seismic event occurred, the
battery charger could have tipped forward onto and shorted out 125Vdc battery terminal(s)
resulting in the loss of Unit 1 division 2 125Vdc safety-related batteries and/or charger.
Absent a seismic event, the charger could have also been knocked over onto the
safety-related battery terminals during normal day-to-day plant operations. In addition,
the inspectors reviewed the more-than-minor examples in NRC Inspection Manual
Chapter 0612, Appendix E, Examples of Minor Issues. The inspectors informed their use of
the more-than-minor questions by comparing this finding to the more-than-minor example 3a.
This example illustrates a calculational error with the potential to adversely affect the
mitigating system cornerstone objective. Similar to example 3a, the inspectors determined
that the finding was more than minor, regardless of the licensees operability assessment,
based upon the inspectors reasonable doubt that this stacked configuration could have
adversely effected Unit 1 division 2 125Vdc battery availability, reliability, and capability had
the battery charger fallen off the mobile cart during a seismic event.
Significance: The inspectors assessed the significance of the finding using IMC 0609
Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The
finding is a deficiency affecting the qualification of a mitigating SSC that maintained its
operability.
Cross-Cutting Aspect: H.8 - Procedure Adherence: Individuals follow processes, procedures,
and work instructions. Specifically, the inspectors determined that a primary cause for the
performance deficiency was because the licensee did not follow procedure LAP-100-56,
Equipment Parts Storage in Plant Areas Containing Safety-Related Equipment, Revision 10,
as part of this work activity.
Enforcement:
Violation: Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and
Drawings requires, in part, that activities affecting quality be prescribed by documented
procedures of a type appropriate to the circumstances and be accomplished in accordance
with these procedures. The licensee established LAP-100-56, Equipment Parts Storage in
Plant Areas Containing Safety-Related Equipment, Revision 10, as the implementing
procedure for storing equipment near safety-related equipment, as an activity affecting
quality.
13
Procedure LAP-100-56, Revision 10, states:
B.4.2.1 Stored items with a height (H) to width (W) ration (H/W) greater than 2.0 may
require a seismic restraint.
B.4.2.2 If H/W is greater than 2.0, proceed to B.4.4 . . .
B.4.3 Stored items with H/W < 2.0. . ..
B.4.4 Stored items with H/W > 2.0.
B.4.4.1 Items stored inside or outside one of the approved storage areas.
B.4.4.1.1 If the edge of the stored item is located at least 24-inches plus the height (H) of
the stored item away from the nearest piece of safety-related equipment/component. . .
otherwise proceed to B.4.6 (seismic restraint required).
B.4.6 Seismic Restraint
B.4.6.1 All stored items not meeting the requirements of B.4.3 and B.4.4 shall require
seismic restraints unless otherwise approved by Engineering. . .
Contrary to the above, from January 29, 2024, to January 31, 2024, the licensee failed to
follow step B.4.6.1 of procedure LAP-100-56, Revision 10. Specifically, the licensee failed to
use required seismic restraints or have the storage configuration approved by Engineering
when a temporary battery charger was stored in a stacked configuration on top of a mobile
cart with an overall height to width ratio of approximately 6 within approximately 10 inches of
the safety-related Unit 1 division 2 125Vdc batteries.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with
Section 2.3.2 of the Enforcement Policy.
Failure to Test Motor-Operated Valve in Accordance with the Inservice Test Program
Cornerstone
Significance
Cross-Cutting
Aspect
Report
Section
Green
Open/Closed
[H.11] -
Challenge the
Unknown
The inspectors identified a finding of very low safety significance (Green) and a non-cited
violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50.55a(f)(4)(ii)
for the licensees failure to meet the in-service testing requirements set forth in the American
Society of Mechanical Engineers (ASME) Operations and Maintenance Code and Addenda
Code Case OMN-1 after performing maintenance that could affect motor-operated valve
(MOV) performance. Specifically, the licensee failed to perform testing on primary
containment isolation MOV, 2B21-F016, prior to returning the valve to service after electrically
backseating the valve, which was maintenance that could affect the valves performance.
Description:
In December 2023, the licensee noticed an increasing trend in Unit 2 reactor coolant system
(RCS) unidentified leakage inside the drywell. The licensee performed troubleshooting to
14
identify the possible leak sources and identified a potential packing leak of the main steam
isolation valve (MSIV) drain header inboard isolation valve, 2B21-F016. This was based, in
part, on the as-left condition of this valve during the last RCS hydrostatic test. Due to the
exponentially increasing leak rate, the licensee chose to electrically backseat the MOV. On
January 25, 2024, the licensee was successful in limiting the RCS unidentified leak rate back
to baseline by electrically backseating the MOV using a new tool to operate the actuator
remotely from motor control center (MCC).
Backseating a valve is a maintenance activity allowing the stem to contact the backseat,
which can either reduce or stop packing leakage. Additionally, backseating a valve can affect
the valves performance (e.g., cause damage to the valve or bind it into its backseat). Since
this activity can affect the valve, NUREG-1482, revision 3, Guidelines for Inservice Testing at
Nuclear Power Plants, Section 4.4.2, Post-Maintenance Testing After Stem Packing
Adjustments and Backseating of Valves to Prevent Packing Leakage, provides guidance for
stroking the valve stem away from the backseat after the initial backseating operation to
demonstrate the valve did not become bound in the backseat. Although the licensee reviewed
this guidance describing an NRC approved testing method for validating a valves
performance, the licensee chose not to implement this guidance. Consequently, the licensee
did not stroke the valve away from the backseat after the initial backseating operation of the
valve.
The licensee performed an engineering evaluation in support of backseating the valve under
engineering change request (ECR) 461530. The inspectors reviewed the evaluation and
noted there were two reasons for performing this evaluation. One aspect of the evaluation
was to evaluate the backseating evolution, which was to be performed for the first time with a
new MOV backseat relay tool (BSRT), under Work Order (WO) 5443427-01. This tool is
installed at the MCC and bypasses the open limit switch to allow the stem to contact the
backseat. The second purpose of the evaluation was to perform a formal technical evaluation
to assess the limitation of the backseating, the potential effects on the MOV structural
capability, and the valve requirements. The licensee evaluated the following three criteria:
increased stroke time, thrust and torque loads applied during backseating of the valve
compared to valve/actuator structural capability, and post-maintenance
requirements/testing/evaluations. From the review, the inspectors noted the structural
capabilities of the valve and actuator were calculated to be acceptable within design limits.
Regarding the licensees evaluation on PMT requirements, MA-AA-716-012, Post
Maintenance Testing, (PMT) revision 28, contains an MOV PMT test matrix which lists
common maintenance activities, such as valve replacement, and assigns pre-established
PMT verification(s) to be performed. Although neither manual nor electrical backseating
were included in the test matrix, the licensee considered the listed maintenance activity of
control circuit disconnect and reconnect to be applicable since this would occur during the
MOV BSRT installation. Therefore, the listed test verifications included rotation and logic
check, control room functional stroke, and an IST operability stroke time test. However,
attachment 2 of the PMT procedure, Waiver Requirement Guidance, allowed engineering to
waive a PMT provided a written justification was completed. The licensee also provided
guidance for engineering to waive the PMT under procedure ER-AA-302-1006, revision 21,
Motor-Operated Valve Maintenance Testing Guidelines, under step 4.2, Guidelines for
Exempting Certain Motor-Operated Valve (MOV) Post-Maintenance Diagnostic (PMDT)
Recommendations from MA-AA-716-012. The written justification for waiving the referenced
test verification was documented in the engineering evaluation. Therefore, no testing was
performed prior to returning the valve to service after the valve was electrically backseated.
15
LaSalle County Station (LSCS) Inservice Testing (IST) Program Plan - 4th Interval,
Revision 0, states the Code of Record for the Fourth 10-Year IST Program interval is the
ASME Code for Operation and Maintenance of Nuclear Power Plants (OM) Code, 2004
Edition through 2006 Addenda. The IST requirements apply, in part, to valves required to
perform a specific function in shutting down the reactor to the safe shutdown condition, in
maintaining the safe shutdown condition, or in mitigating the consequences of an accident.
The IST Program Plan - 4th Interval incorporates ASME Code Case OMN-1, 2006 Addenda,
through the alternative granted by the Nuclear Regulatory Commission (NRC) in valve relief
request - RV-01, Utilization of ASME Code Case OMN-1. With this alternative granted by
the NRC for this IST Program interval, the inspectors noted primary containment isolation
MOV 2B21-F016 was subject to ASME Code Case OMN-1, Alternative Rules for Preservice
and Inservice Testing of Active Electric Motor-Operated Valve Assemblies in Light-Water
Reactor Power Plants.
The IST Program Plan - 4th Interval states 2B21-F016 is an ASME Class 1, Category A,
normally open, motor-operated, active valve with a safety function in the closed position. The
ASME OM-2006 Code Case OMN-1, paragraph 3.4, states, in part, When an MOV or its
control system is replaced, repaired, or undergoes maintenance that could affect the valves
performance, new inservice test values shall be determined, or the previously established
inservice test values shall be confirmed before the MOV is returned to service... This testing
is intended to demonstrate that performance parameters, which could have been affected by
the replacement, repair, or maintenance, are within acceptable limits. The Owners program
shall define the level of testing required after replacement, repair, or maintenance...
Since both the valve and its control system underwent maintenance that could affect the
MOVs performance, the inspectors determined the valve was required to be tested in
accordance with Paragraph 3.4 of Code Case OMN-1. The Code Case does not provide the
allowance to perform an evaluation in lieu of the required inservice testing. Although this does
not conform to the requirements, the licensees evaluation and work results provided
reasonable assurance the structural integrity of the MOV was not exceeded. However, not
performing any testing after maintenance prior to returning the valve to service did not
maintain the requisite level of assurance for the valve. Considering operating experiences
with backseating, the number of assumptions embedded in the evaluation, and the use of the
new MOV BSRT, the inspectors noted there were several uncertainties associated with the
engineering evaluation and its use to restore operability. Based on the inspectors review of
the IST plan and procedures for the valve, the inspectors determined the licensee failed to
ensure the testing required after maintenance under WO 5443427-01 was performed in
accordance with Code Case OMN-1. In addition, the licensee did not request relief from the
code via an ASME Code relief request to the NRC which, if approved, would have allowed
the valve to be returned to service without performing the required testing.
Corrective Actions: The licensee entered this issue into their corrective action program. The
licensee is evaluating the technical and regulatory requirements for resolution and alignment.
Corrective Action References: AR 4754319, NRC ID ASME OM Code Potential
Finding/Violation; and AR 4753350, NRC IS Questions on Valve Backseating Activity
Performance Assessment:
Performance Deficiency: The licensees failure to perform required testing after
maintenance for the primary containment isolation MOV 2B21-F016 in accordance with
ASME OM Code-2004, 2006 Addenda, Code Case OMN-1, Paragraph 3.4, was a violation of
16
10 CFR 50.55a(f)(4)(ii) and a performance deficiency. Specifically, the licensee failed to
perform required testing on the MOV prior to returning the valve to service after electrically
backseating the valve, which was maintenance that could affect the valves performance.
Screening: The inspectors determined the performance deficiency was more than minor
because it was associated with the SSC and Barrier Performance attribute of the Barrier
Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable
assurance that physical design barriers protect the public from radionuclide releases caused
by accidents or events. The inspectors determined the finding was associated with the SSC
and Barrier Performance objective of the Barrier Integrity cornerstone and adversely affected
the cornerstone objective to provide reasonable assurance physical design barriers (fuel
cladding, reactor coolant system, and containment) protect the public from radionuclide
releases caused by accidents or events. Specifically, by not performing the required testing,
the licensee did not maintain the requisite level of assurance of the valves reliability of
performing its intended function after performing maintenance that could affect the valves
performance.
Significance: The inspectors assessed the significance of the finding using IMC 0609
Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The
inspectors determined the finding was of very low safety significance (Green) because they
answered No to all exhibit 3, Barrier Integrity Screening Questions, section C, Reactor
Containment, screening questions.
Cross-Cutting Aspect: H.11 - Challenge the Unknown: Individuals stop when faced with
uncertain conditions. Risks are evaluated and managed before proceeding. Specifically,
when presented with an emergent and exponentially increasing RCS unidentified leakage,
the licensee determined testing after maintenance was not required prior to returning the
valve to service through an engineering evaluation. This was the licensees first time
performing this evolution and using the new MOV BSRT, and the licensee did not adequately
challenge, understand, and manage the activity affecting quality to ensure regulatory
requirements were met.
Enforcement:
Violation: Title 10 CFR 50.55a(f)(4)(ii), requires, in part, 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 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 (or the optional ASME Code
Cases listed in NRC Regulatory Guide 1.192 as incorporated by reference in paragraph
(a)(3)(iii) of this section).
LaSalle County Station IST Program Plan - 4th Interval, Revision 0, establishes the Code of
Record for the Fourth 10-Year IST Program Interval (October 12, 2017 - October 11, 2027)
as the ASME OM Code, 2004 Edition through 2006 Addenda, as incorporated by reference in
10 CFR 50.55a. LSCS submitted a Valve Relief Request RV-01 to implement the optional
ASME Code Case OMN-1 in their Fourth 10-Year IST Program Plan.
ASME OM Code-2006, Code Case OMN-1, Paragraph 3.4, Effect of MOV Replacement,
Repair, or Maintenance, states, in part, When an MOV or its control system is replaced,
repaired, or undergoes maintenance that could affect the valves performance, new inservice
test values shall be determined, or the previously established inservice test values shall be
17
confirmed before the MOV is returned to service.
Contrary to the above, on January 25, 2024, the licensees inservice tests to verify
operational readiness of pumps and valves, whose function is required for safety,
did not comply with the requirements of the 2004 Edition through the 2006 Addenda of the
ASME OM Code as incorporated by reference in 10 CFR 50.55a for the current 10-Year IST
program interval at LaSalle County Station effective October 17, 2017. Specifically, the
licensee failed to perform any testing on primary containment isolation MOV 2B21-F016, a
valve within the scope of the ASME OM Code and Addenda, before returning the valve to
service after electrically backseating the valve, which was maintenance that could affect the
valves performance.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with
Section 2.3.2 of the Enforcement Policy.
Failure to Use Calibrated Measuring and Test Equipment to Electrically Backseat a Safety
Related Motor-Operated Valve
Cornerstone
Significance
Cross-Cutting
Aspect
Report
Section
Green
Open/Closed
[H.11] -
Challenge the
Unknown
The inspectors identified a finding of very low safety significance (Green) and a NCV of
Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion XII,
Control of Measuring and Test Equipment, for the licensees failure to assure a tool used in
activities affecting quality was properly controlled, calibrated, and adjusted at specified
periods to maintain accuracy within necessary limits. Specifically, the licensee failed to apply
quality assurance requirements to the MOV BSRT when using the tool to electrically backseat
a Unit 2 safety-related valve.
Description:
On January 25, 2024, the licensee used a tool, the MOV BSRT, to electrically backseat the
B MSIV drain header inboard isolation valve 2B21-F016, a safety-related containment
isolation MOV. The valve was backseated to address a suspected packing leak, which was
contributing to an increase of Unit 2 RCS unidentified leakage. Although backseating is a
known method available to mitigate valve packing leaks, backseating is not commonly
performed. Furthermore, electrically backseating is an even more uncommon evolution. In
both manual and electrical backseating of valves, operating experience has shown
backseated valves can be either damaged or become bound into their backseats. The NRC
has provided guidance on backseating under section 4.4.2, Post-Maintenance Testing After
Stem Packing Adjustments and Backseating of Valves to Prevent Packing Leakage, of
NUREG-1482, revision 3, Guidelines for Inservice Testing at Nuclear Power Plants.
The inspectors performed a review of the MOV BSRT used to accomplish this
activity. Engineering evaluation documented under ECR 461530 was reviewed and
found to incorporate guidance of draft procedure MA-AA-723-304, Electrical Backseating
Motor-Operated Valves Remotely from a Motor-Control Center - Current Cut-Off Method, for
using this tool. From the evaluations conclusion, the inspectors noted this was the first use of
the MOV BSRT on a currently installed valve at the site. The guidance from the referenced
draft procedure was also incorporated in WO 5443427-01, which was implemented to
18
perform the electrical backseating of the valve with the MOV BSRT. The MOV BSRT was
installed remotely at the MOVs MCC on the open control circuit in parallel with the Limitorque
actuator open limit and torque switch. This allowed the tool to bypass the MOV open limit
settings to stroke the valve in the open direction. Normally, during an open stroke of the
valve, the valves actuator would stop the valve travel near the fully open position when the
open limit switch trips before the stem contacts the backseat. The MOV BSRT uses a
specified threshold percentage of minimum recorded running current to stop the valve travel
after the valve stem contacts the backseat.
The inspectors questioned the quality of the MOV BSRT used for this maintenance evolution
as this was an activity affecting quality. The licensee stated the MOV BSRT was not
considered measurement and testing equipment (M&TE) per procedure MA-AA-716-040,
revision 16, Control of Portable Measurement and Test Equipment Program. The licensee
also stated, The backseat relay tool monitors current and then performs its function in terms
of percent. Since it gives functions in terms of percentages and not absolute terms, no
calibration for current reading is necessary.
The inspectors reviewed the MOV BSRT users manual, TM201602, revisions 9 and 13. In
both revisions, section 9, Maintenance, states, in part, Calibration of the device may be
performed but is not required. Trip setpoints are specified as a percentage of current. As
such, the actual current does not matter to meet the intent. Current readings are provided for
information. Refer to the calibration procedure available at the link below. The licensee
determined the Quality Assurance (QA) required calibration and verification of the tool
measured current would not be performed based on the users manual. Through discussions
with the licensee, the inspectors noted that the users manual step 4.9 of revision 13 the
licensee later referenced was not found in the evaluation, which referenced revision 9.
Revision 13, step 4.9, states, Since all setpoints are expressed as percentages, accurate
calibration of the relay or probes is not relevant to trip functioning. The inspectors also
reviewed the MOV BSRT functional test and calibration procedure, TP201602-03, revision 1,
and found the stated purpose was to provide a means of bench verifying the relay responds
correctly to current inputs. The test also verifies that each phase current input is operational
as well as the displayed value of current. Additionally, the procedure states, Since the relay
operates on relative current readings to detect increased motor load as the valve reaches the
backseat, calibration of current reading is not required for proper functioning of the device.
Therefore, although the current reading displayed is not required for proper functioning of the
tool, it is still necessary to test and calibrate the MOV BSRT to ensure both the relays operate
correctly and the phase current inputs are operational.
When Constellation procured the MOV BSRT in 2019, the tool was classified as the following:
For General Use Only, Not for Qualitative or Quantitative Measurements, and Indication
Only. After purchasing from a commercial vendor, the MOV BSRT was entered into the
companys M&TE log for tool traceability. However, the licensee did not establish QA related
controls for the tool.
The inspectors reviewed the following quality assurance requirement. Title 10 CFR Part 50,
Appendix B, Criterion XII, states: Measures shall be established to assure that tools, gages,
instruments, and other measuring and testing devices used in activities affecting quality are
properly controlled, calibrated, and adjusted at specified periods to maintain accuracy within
necessary limits. Since the tool was used in activities affecting quality, the inspectors
determined the licensee failed to classify the MOV BSRT as a tool to be controlled and
calibrated in accordance with 10 CFR 50, Appendix B, Criterion XII, and their Quality
19
Assurance Topical Report (QATR) section 12, Control of Measuring and Test Equipment.
Therefore, the licensee failed to established measures to assure that the MOV BSRT, which
was used in an activity affecting quality, was properly controlled, calibrated, and adjusted at
specified periods to maintain accuracy within necessary limits.
Corrective Actions: The licensee has entered the inspectors concern into their corrective
action program. The licensee developed and performed just-in-time training for operating the
MOV BSRT prior to performing WO 5443427-01. Also, QA controlled MOV diagnostic testing
equipment was installed at the MCC to monitor current in parallel with the MOV BSRT current
monitoring probes. The licensee compared the data from the QA equipment to previous
diagnostic testing data and determined the tool operated as expected.
Corrective Action References: AR 4754320, NRC ID Use of Backseating Tool Potential
Finding/Violation
Performance Assessment:
Performance Deficiency: The licensees failure to establish quality assurance measures for
the MOV backseat relay tool used to electrically backseat a safety-related valve was contrary
to 10 CFR 50, Appendix B, Criterion XII, and was a performance deficiency. Specifically, the
licensee failed to assure the MOV BSRT was properly controlled, calibrated, and adjusted at
specified periods to maintain accuracy within necessary limits.
Screening: The inspectors determined the performance deficiency was more than minor
because if left uncorrected, it would have the potential to lead to a more significant safety
concern. Specifically, continued use of an uncontrolled and uncalibrated MOV BSRT has the
potential to cause structural damage to a valve.
Significance: The inspectors assessed the significance of the finding using IMC 0609
Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The
inspectors determined the finding was of very low safety significance (Green) because they
answered No to all exhibit 3, Barrier Integrity Screening Questions, section C, Reactor
Containment, screening questions.
Cross-Cutting Aspect: H.11 - Challenge the Unknown: Individuals stop when faced with
uncertain conditions. Risks are evaluated and managed before proceeding. Specifically,
when electrically backseating a safety-related valve, the licensee did not evaluate and
manage the risk of using an unqualified tool to perform a maintenance activity affecting
quality.
Enforcement:
Violation: Title 10 CFR Part 50, Appendix B, Criterion XII, Control of Measuring and Test
Equipment, requires, that measures shall be established to assure that tools, gauges,
instruments, and other measuring and testing devices used in activities affecting quality are
properly controlled, calibrated, and adjusted at specified periods to maintain accuracy within
necessary limits.
Contrary to the above, on January 25, 2024, the licensee failed to assure that a tool used in
activities affecting quality was properly controlled, calibrated, and adjusted to maintain
accuracy within necessary limits. Specifically, the licensee failed to establish quality
assurance measures for the MOV backseat relay tool used to electrically backseat a
safety-related valve.
20
Enforcement Action: This violation is being treated as a non-cited violation, consistent with
Section 2.3.2 of the Enforcement Policy.
Unresolved Item
(Closed)
Clarification of the Dew Point Specification for the MSA
Firehawk M7 SCBA System
URI 05000373,05000374/2023002-01
Description:
The inspectors reviewed a previously identified unresolved item for a condition where the
quality of breathing air used to fill self-contained breathing apparatus (SCBA) bottles did not
meet all of the parameters specified by the manufacturer as specified in the user instructions
manual. National Institute for Occupational Safety and Health (NIOSH) regulations and
guidance state that user instructions are included as part of the NIOSH approval. Nuclear
Regulatory Commission regulations require that NRC licensees use NIOSH approved
equipment in their respiratory protection programs, or that they obtain approval from the
USNRC to use equipment that has not been approved by NIOSH. However, it was not clear if
the NIOSH approval was contingent upon the dew point guidance that applied to Grade D air,
or a dew point of -65°F, two parameters of breathing air quality that conflicted with each
within the user instructions.
Review of this issue was discontinued in accordance with the Very Low Safety Significance
Issue Resolution (VLSSIR) process as documented in this report. No further evaluation is
required.
This item is closed.
Corrective Action Reference(s): AR 4686286
Very Low Safety Significance Issue Resolution Process: Very Low Safety
Significance Issue Resolution Process: Dew Point Specification for the MSA
M7XT SCBA System
This issue is a current licensing basis question and inspection effort is being discontinued in
accordance with the Very Low Safety Significance Issue Resolution (VLSSIR) process. No
further evaluation is required.
Description:
The NRC promulgated requirements for the use of respiratory protection and controls to
restrict internal exposure in Subpart H to 10 CFR 20 Standards for Protection Against
Radiation. Within this regulation are requirements when a licensee assigns or permits the
use of respiratory protection equipment to limit the intake of radioactive material. Some
anticipated uses of respiratory protective equipment to reduce the intake of radioactive
material include repair of highly contaminated equipment, decontamination of large surface
areas, and responding to an accident or a fire involving radioactive contamination. The
respiratory protection equipment with highest protection factor is the SCBA. These are used
for responding to an accident or a fire involving radioactive contamination. The SCBA unit is a
device that includes a face mask connected to a bottle of compressed air carried on the back
of the user. This is also known as an atmosphere-supplying respirator, as the only air
available to the user is from the bottle of compressed air.
21
Atmosphere-supplying respirators, such as SCBAs, must be supplied with respirable air of
Grade D quality or better as defined by the Compressed Gas Association in publication
G-7.1, Commodity Specification for Air, 1997 and included in the regulations of the
Occupational Safety and Health Administration (29 CFR 1910.134(i)(1)(ii)(A) through (E)).
Grade D quality air criteria include
(1) oxygen content (v/v) of 19.5-23.5%
(2) hydrocarbon (condensed) content of 5 milligrams per cubic meter of air or less
(3) carbon monoxide (CO) content of 10 ppm or less
(4) carbon dioxide content of 1,000 ppm or less
(5) lack of noticeable odor
Compressed Gas Association in publication G-7.1, Commodity Specification for Air, 1997
Table 1 - Directory of Limiting Characteristics, also includes maximum dew point or
moisture content for compressed air used as breathing air. The value listed in the table for
Grade D air is blank but covered by a footnote from the dew point parameter. Specifically, this
states The water content of compressed air required for any particular quality verification
level may vary with the intended use from saturated to very dry. For breathing air used in
conjunction with a self-contained breathing apparatus in extreme cold where moisture can
condense and freeze causing the breathing apparatus to malfunction, a dew point not to
exceed -65°F (24 ppm v/v) or 10 degrees Fahrenheit lower than the coldest temperature
expected in the area is required. If a specific water limit is required, it should be specified as a
limiting concentration in ppm (v/v) or dew point...
The inspectors have observed test results for breathing air quality consistently achieved
results with moisture content between 63 ppm (v/v) and 24 ppm (or dew point between -50°F
and -65°F).
The inspectors reviewed the operation and instructions manual published by the SCBA
manufacturer to identify limitations or evaluations that might assist with the evaluating the
apparent failure to ensure the SCBA will remain functional if used in extreme cold where
moisture can condense and freeze causing the breathing apparatus to malfunction. The
inspectors identified inconsistencies as it pertains to dew point specifications for SCBA.
Specifically, within the same user instruction documentation, in one section of the instructions
include a more conservative dew point might be prescribed when compared to the dew point
specified in another section. The least stringent dew point inspectors observed corresponds
to Grade D air (i.e., -50°F or 10°F) less than coldest expected ambient temp), whereas the
more conservative dew point corresponds to that of Grade L air (i.e., -65°F). Additionally, the
manual includes a special or critical user instruction that states the equipment is approved for
use at temperatures above -25°F.
The licensee has revised air quality testing procedure parameters. Since these changes were
implemented, the licensee has demonstrated the breathing air used to fill bottles used for
SCBA consistently satisfies the more stringent standard (-65°F).
Licensing Basis: The requirements for use of respiratory protection equipment to limit the
intake of radioactive material are established in 10 CFR 20.1703. Specifically, § 20.1703
states -
(a) The licensee shall use only respiratory protection equipment that is tested and certified
by the National Institute for Occupational Safety and Health (NIOSH) except as otherwise
noted in this part.
22
User instructions are part of the NIOSH certification; therefore, the inconsistency in the user
instructions introduces a situation where the equipment cannot be used per its instructions;
presumably leading to a violation of the NIOSH certification and our requirements.
Significance: The inspectors determined the issue was of very low safety significance
because the less stringent dew point specification of Grade D air (-50°F) was sufficient for the
environment (ambient temperatures above -25°F) in which the equipment was used or would
be used. Additionally, some of this equipment has been in service for many years without
issue.
For the purpose of the VLSSIR process, the inspectors screened the issue of concern
through IMC 0609, Appendix C and determined the issue of concern would likely be
Green had a performance deficiency been identified. Specifically, it would not have been an
as-low-as-reasonably-achievable planning issue, there would not have been overexposures,
nor substantial potential for overexposures, and the licensees ability to assess dose would
not be compromised. Therefore, the condition represents an issue of very low safety
significance that does not warrant additional review.
Technical Assistance Request: The inspectors did not enter either the TIA or technical
assistance request process. However, the inspectors contacted the cognizant branch in the
Office of Nuclear Reactor Regulation (NRR). Attempts to resolve whether the inconsistency in
the user instructions introduced a situation where the equipment cannot be used per its
instructions or invalidated the NIOSH certification were inconclusive. Consequently, the
inspectors could not determine whether the respiratory protection equipment was used as
certified by the NIOSH and required by 10 CFR 20.1703(a).
Corrective Action Reference: AR 4686286
Failure to Promptly Correct Degraded Pressure Switches in the Unit 1 and Unit 2 Main Steam
Line High Flow Isolation Logic System
Cornerstone
Significance
Cross-Cutting
Aspect
Report
Section
Green
NCV 05000373,05000374/2024001-04
Open/Closed
None (NPP)
The inspectors identified a Green finding and associated non-cited violation of
10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to
promptly correct a condition adverse to quality associated with the Unit 1 and Unit 2 main
steam line (MSL) isolation logic system. Specifically, the licensee identified in a 2008
equipment apparent cause evaluation that pressure switches installed in the Unit 1 and Unit 2
MSL high-flow isolation logic trip systems were susceptible to multiple failure modes and had
been exposed to peak inductive currents during frequent calibration activities that may have
damaged switch contacts. A subsequent corrective action replaced half of the impacted
switches while the other half continued to be replaced on an as-needed basis.
Description:
On January 9, 2024, LaSalle County Station (LSCS) Unit 2 received a main steam line
isolation valve (MSIV) half-isolation signal from the A trip system during a scheduled MSL
high-flow isolation calibration activity. Troubleshooting performed by the licensee
subsequently identified a failed pressure switch in the MSL high-flow isolation logic that they
23
believe caused the spurious half-isolation signal.
There are four differential pressure switches connected to each of the four MSLs in the MSL
high-flow isolation logic system, for a total of 16 differential pressure switches per unit. Each
of the four pressure switches connected to a steam line input to a trip channel and there are
two trip channels in each trip system. A half-isolation signal with no associated MSIV closure
occurs when the relays in a single trip system drop out. Currently, the licensee uses pressure
switches manufactured by Static-O-Ring (SOR). Two SOR models are installed in the trip
channels including the 102 series that was installed starting in the 1980s. The licensee began
to replace the 102 series with the 131 series on an as-needed basis in 2005 due to internal
diaphragm failures, contact quality issues, and setpoint drift experienced with the 102 series.
The resident inspectors evaluated the maintenance history of the failed pressure switch and
concluded that it was a 102-series SOR switch and was 22 years old at the time of the
spurious Unit 2 half isolation. Further, the resident inspectors reviewed an equipment
apparent cause evaluation (EACE) from 2008 for a similar spurious MSIV half-isolation signal
that occurred on Unit 1 in the B trip system. The EACE notes that the half-isolation signal
was caused by a failed pressure switch. The EACE also notes that the faulty pressure switch
removed after the event presented with contacts that were severely arc damaged. It suggests
that observed damage was caused by peak inductive currents generated during the
calibration activity. In 2008, LaSalle technical specifications only required this calibration
activity every 2 years, though the licensee was performing it every 92 days to address
setpoint drift exhibited by 102 series switches as noted above. Thus, the EACE suggests that
the observed damage to switch contacts was caused by the accelerated frequency of the
pressure switch calibration activity.
The cited apparent cause listed in the 2008 EACE was the design of the MSL High Flow
Switchis susceptible to multiple failure modes. The basis associated with a subsequent
causal factor further notes that shortly after the 2008 half-isolation, only 8 of the 32 (16 per
unit) MSL pressure switches had been upgraded to 131 series and reflects that if the
replace as they fail philosophy were not employed, this event could have been avoided.
Corrective actions listed in the EACE included replacement of the failed defective flow switch
and implementation of a replacement schedule so that at least one string in each trip system
will be replaced in approximately 6 months. An action item to develop a test box was
further listed in the EACE that could be installed during the calibration activity and suppress
the inductive current experienced by the switch contacts.
The residents reviewed maintenance records associated with the MSL isolation high-flow
pressure switches and determined that all Unit 1 and Unit 2 pressure switches in the B1 and
B2 trip channels were replaced by May 2009 in response to the corrective actions noted
above. They reviewed calibration procedures and noted that the test box identified by
another corrective action was developed and implemented in May 2011 via test procedure
LIS-MS-102/202, Unit 1/Unit 2 Main Steam Line High Flow MSIV Isolation Calibration.
The resident inspectors note however, that at the time of the January 2024 Unit 2 MSIV
half-isolation, only 8 of the 16 Unit 1 and Unit 2 pressure switches on the A1 and A2 trip
channels had been upgraded to the 131-series model. In fact, switches in these trip channels
have a mean in-service age of 28 years. The resident inspectors have not identified
a corrective action assigned to the pressure switches in these trip channels, even those the
licensee concluded in the 2008 EACE that the half-isolation was caused by the multiple
failure modes of the SOR model 102 pressure switches and frequent calibration activities
24
without surge protection most likely induced degradation across the pressure switch contacts.
Instead, these switches have been replaced as needed based on quarterly calibration testing.
The inspectors also reviewed several other maintenance issues associated with the MSL
high-flow pressure switches that occurred between 2008 and 2024. In 2014, the licensee
identified that both the 102 and 131-series SOR switches were exhibiting an increasing trend
in diaphragm failures. Another EACE documented in 2016 logged a series of calibration
issues seen in both 102 and 131-series SOR switches. An action item resulting from that
EACE produced a project to replace SOR-manufactured switches with Rosemount
Transmitters and trip units. Although this project was originally planned to be implemented in
approximately 2018, the upgrade has been delayed until 2028. Currently, the licensee
maintains a quarterly calibration frequency and is replacing pressure switches with the
SOR 131-series model on an as-needed basis.
Corrective Actions: In response to the January 2024 half-isolation signal, the licensee
replaced the bad SOR 102-series switch with a SOR 131-series switch. They also performed
a failure analysis on the bad switch and determined that the half-isolation signal occurring in
2024 was most likely caused by a poor solder connection at the micro-switch.
In response to the inspectors observations, the licensee wrote corrective action AR 4761815
to evaluate Unit 1 and Unit 2 MSIV isolation pressure switches that have not been replaced
since May 2011 and generate work requests as needed.
Corrective Action References: ARs 4727857, 4730751, 844283, 2607807, and 4761815
Performance Assessment:
Performance Deficiency: The inspectors determined that the licensees failure to promptly
replace SOR pressure switches in the A1 and A2 MSL high-flow trip channels that were
exposed to potential degradation across contacts is a performance deficiency. Specifically, a
2008 EACE evaluated a spurious MSIV half-isolation similar to the isolation that occurred in
January 2024 and cited the known multiple failure modes associated with the switches as an
apparent cause of the 2008 half-isolation. It also identified that the switch contacts had been
exposed on multiple occasions to peak inductive currents, potentially causing degradation
to those contacts. A corrective action to replace half the impacted switches was identified
and implemented at the time. The remainder of the switches have been replaced on an
as-needed schedule. To date, seven pressure switches on the Unit 1 and Unit 2 MSLs have
not been replaced.
Screening: The inspectors determined the performance deficiency was more than minor
because it was associated with the Equipment Performance attribute of the Initiating Events
cornerstone and adversely affected the cornerstone objective to limit the likelihood of events
that upset plant stability and challenge critical safety functions during shutdown as well as
power operations. Specifically, the licensees failure to replace SOR pressure switches in the
A1 and A2 MSL high-flow trip channels after identifying that the switches were susceptible
to multiple failure modes and had been exposed to peak inductive currents increased the
potential for a spurious MSL isolation and reactor scram. Unit 1 is currently operating with five
pressure switches which were subjected to the noted degradation mechanism and Unit 2 is
currently operating with two such switches.
Significance: The inspectors assessed the significance of the finding using IMC 0609
Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The
25
finding screened as Green, or very low safety significance, because the inspectors answered
No to Exhibit 1, Section B questions. To date, no full MSIV isolation/reactor trip has been
attributed to faulty pressure switches at LaSalle.
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: 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, states that
measures shall be established to assure that conditions adverse to quality, such as failures,
malfunctions, deficiencies, deviations, defective material and equipment, and
nonconformances are promptly identified and corrected.
Contrary to the above, as of April 17, 2024, the licensee failed to establish measures to
assure that conditions adverse to quality are promptly corrected. Specifically, the licensee
failed to correct pressure switches installed in the Unit 1 and Unit 2 MSL high-flow isolation
logic trip systems that are susceptible to multiple failure modes and have been exposed to
peak inductive currents during frequent calibration activities that may have damaged the
switch contacts. The licensee replaced all switches in the B1 and B2 trip channels while
switches in the A1 and A2 trip systems have been replaced on an as-needed basis.
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 April 17, 2024, the inspectors presented the integrated inspection results to
Christopher Smith, Plant Manager, and other members of the licensee staff.
On March 1, 2024, the inspectors presented the radiation protection inspection results to
John VanFleet, Site Vice President, and other members of the licensee staff.
On March 1, 2024, the inspectors presented the inservice inspection results to
John VanFleet, Site Vice President, and other members of the licensee staff.
26
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
M-96, Sheet 1
P&ID Residual Heat Removal System (RHRS)
Drawings
M-96, Sheet 4
P&ID Residual Heat Removal System (RHRS)
AH
Procedures
LOS-RH-Q1
RHR (LPCI) and RHR Service Water Pump and
Valve Inservice Test for Modes 1, 2, 3, 4, and 5
101
FZ 2K
Rx Bldg. 687'-0" to 768'-0" Elev. U1 Steam Tunnel
Fire Plans
FZ 3I4
Rx Bldg. 673'-4" Elev. U2 LPCS/RCIC Pump Cubicle
2
Engineering
Changes
Evaluate Plugging Floor Drains to Support Maintenance in the
Replacing of the U1 L1R20 250V Batteries (1DC01E)
02/08/2024
Ultrasonic Examination of RPV Recirculation Outlet Nozzle
IR 1-NIR-10
02/24/2024
Ultrasonic Examination of RPV Core Spray Nozzle
IR 1-NIR-16
02/25/2024
Ultrasonic Examination of RPV Bottom Head Meridional Weld
GEL-1006-DB
02/27/2024
Ultrasonic Examination of RPV Bottom Head Meridional Weld
GEL-1006-DC
02/27/2024
Ultrasonic Examination of RPV Bottom Head Meridional Weld
GEL-1006-DD
02/27/2024
Visual Examination of MS Restraint MS01-1352X
02/21/2024
NDE Reports
Visual Examination of RHR Constant Support RH40-1004C
02/23/2024
GEH-PDI-UT-1
PDI Generic Procedure for the Ultrasonic Examination of
Ferritic Welds
12.1
GEH-PDI-UT-2
PDI Generic Procedure for the Ultrasonic Examination of
Austenitic Pipe Welds
13
GEH-UT-300
Procedure for the Manual Ultrasonic Examination of Reactor
14
Procedures
GEH-UT-311
Procedure for Manual Ultrasonic Examination of Nozzle Inner
Radius, Bore, and Selected Nozzle to Vessel Regions
20
Work Orders
Unit 1 RHR HX Weld 1RH-HX1B-9A Repair
03/04/2022
LGP-1-1
Normal Unit Startup
134
NF-AB-720-F-1
L1C21 Startup Sequence
02
Procedures
L1C21 BOC Startup ReMA
20
27
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
1B RR Seal Pressure and Temperature Fluctuations
0
1A RR Seal Pressure and Temperature Fluctuations
0
LAS Named in 10CFR Part 21 Notification from
Valcor Eng Co.
12/27/2023
Metal Shavings Discovered during 2B DG Top Deck
Inspection
01/23/2024
Corrective Action
Documents
GGJ319 GNF Water Rod Inspection Not Complete
02/21/2004
Work Orders
WO 8003976-fa
Failure Analysis of Woodward EGR Actuator
0
Work Orders
L1R20 Water Rod Inspections
02/22/2024
Seismic Qualification of Velan 3" Class 900 MO Gate Valves
for Use in Applications 1(2)B21-F016, 19
1
Calculations
LAS-2B21-F016
MIDACALC MOV Datasheet
13
1AP04E-13 Will Not Close - 1A RR LFMG BKR 1A
02/21/2024
Corrective Action
Documents
HYDRO Test Issues
03/04/2024
NRC ID Questions on Valve Backseating Activity
02/26/2024
NRC ID ASME OM Code Potential Finding/Violation
02/29/2024
NRC ID Use of Backseating Tool Potential Finding/Violation
02/29/2024
NRC-Identified IR for WO 5403063
03/22/2024
Corrective Action
Documents
Resulting from
Inspection
Work Order 05443427-01 Amendment
03/27/2024
Engineering
Changes
Evaluate Electrically Backseating 2B21-F016 Remotely from
Motor Control Center Using MOV Backseat Relay Tool
2
IST-LAS-BDOC-
V-14
IST Basis Document for 2B21-F016
03/01/2019
IST-LAS-PLAN
Inservice Testing Program Plan Fourth Ten-Year Interval
0
Quality Assurance Topical Report
98
TM201602-AC
MOV Backseat Relay for AC Motors Users Manual
13
TM201602-AC
MOV Backseat Relay for AC Motors Users Manual
9
Miscellaneous
TP201602-03
MOV Backseat Relay Model 201602-AC Functional Test and
Calibration
1
LIS-MS-202
Unit 2 Main Steam Line High Flow MSIV Isolation Calibration
27
LOS-DG-109
Unit 1 Integrated Division 1 Response Time Surveillance
30
LOS-NB-R1
U1 Reactor Vessel Leakage Test
34
Procedures
Post Maintenance Testing
28
28
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Control of Portable Measurement and Test Equipment
Program
16
Electrical Backseating Motor-Operated Valves Remotely from
a Motor Control Center - Current Cut-Off Method
0
IST Comprehensive Pump Test for 2E22-C002
03/18/2024
Div 1 Integrated Divisional Response Time Test
02/22/2024
ASME XI ISI of Class I Components and Associated Piping
03/05/2024
IM LIS-MS-202 U2 MSL High Flow MSIV Isolation Cal
01/11/2024
WO Needed to Perform LEP-DC-01 for 1DC 14E Cell 56
01/29/2024
LOS-SC-Q1 U2 B SBLC Pump Quarterly
03/15/2024
U2 Drywell Leakage Has Increased
01/25/2024
Work Orders
2A Diesel Generator Idle Start
03/15/2024
Work Orders
1A DG Fast Start
03/29/2024
LA-01-24-00502
Refueling Outage: Drywell Radiation Protection Dept.
Activities
30
LA-01-24-00513
L1R20 Control Rod Drive (CRD)/Undervessel Activities
30
ALARA Plans
LA-01-24-00601
L2R20 RB RWCU System Maintenance Activities
30
Corrective Action
Documents
APS: Unexpected Dose (WO 5236584-04)
02/20/2024
NRC Observations of Radiological Activities
02/28/2024
Corrective Action
Documents
Resulting from
Inspection
NRC Observations
02/29/2024
Miscellaneous
LA-01-24-00601
TEDE ALARA Evaluation Screening Worksheet for
L1R20 RB RWCU System Maintenance Activities 500k G/A,
40 dpm alpha
01/31/2024
Determination of Alpha Levels and Monitoring
12
Refueling Outage DW Under Vessel CRD Preps/Exchange
Protective Clothing Matrix
Procedures
Controls for Very High Radiation Areas
8
LA-01-24-00903
L1R20 RFF Cavity Platform IVVI and Associated Activities
04/19/2023
Radiation Work
Permits (RWPs)
LA-010-24-00601
L1R20 RB Reactor Water Clean Up System Maintenance
00
Shipping Records
Shipment LM24-
Control Rod Drive Boxes
02/29/2024
29
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
003
RM - U1 Chemistry Sampling Indicates Potential Fuel Defect
12/28/2022
71151
Corrective Action
Documents
U1 Tertiary Oil Addition Troubleshooting
03/22/2023
EACE for Critical Component Failure of 2E31-N011D
01/05/2016
Spurious U2 A1 MSIV Isol Trip
01/09/2024
UMCRA: 2H13-P601-F504 CHAN A1/A2 MSIV ISOL Trip
01/12/2024
Corrective Action
Documents
AR 844283844283Unexpected Group 1 MSIV Half-Isolation Signal
11/12/2008
Corrective Action
Documents
Resulting from
Inspection
NRC ID - MSL Flow Switch Degradation Not Addressed
03/28/2024
Drawings
Schematic Diagram Primary Containment & Reactor Vessel
Isolation System PC (B21H) Part 2
Engineering
Changes
SOR Model Replacement/Alternative Solution
03/16/2010
Miscellaneous
Licensee Event
Report
05000374/2023-
003-00
Automatic Actuation of Reactor Protection System (RPS)
05/02/2023