ML24207A079
| ML24207A079 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 07/29/2024 |
| From: | Jeffrey Josey NRC/RGN-IV/DORS/PBC |
| To: | Kapellas B Entergy Operations |
| References | |
| IR 2024002 | |
| Download: ML24207A079 (27) | |
See also: IR 05000416/2024002
Text
July 29, 2024
Brad Kapellas, Site Vice President
Entergy Operations, Inc.
P.O. Box 756
Port Gibson, MS 39150
SUBJECT:
GRAND GULF NUCLEAR STATION - INTEGRATED INSPECTION REPORT
Dear Brad Kapellas:
On June 30, 2024, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at
Grand Gulf Nuclear Station. On July 10, 2024, the NRC inspectors discussed the results of this
inspection with you and other members of your staff. The results of this inspection are
documented in the enclosed report.
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 IV; the Director, Office of Enforcement; and the NRC Resident Inspector
at Grand Gulf Nuclear 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 Grand Gulf Nuclear Station.
B. Kapellas
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,
Jeffrey E. Josey, Chief
Reactor Projects Branch C
Division of Operating Reactor Safety
Docket No. 05000416
License No. NPF-29
Enclosure:
As stated
cc w/ encl: Distribution via LISTSERV
Signed by Josey, Jeffrey
on 07/29/24
x
SUNSI Review
x
Non-Sensitive
Sensitive
x
Publicly Available
Non-Publicly Available
OFFICE
SRI:DORS/C
SPE:DORS/C
BC:DORS/C
NAME
ASmallwood
RAzua
JJosey
SIGNATURE
/RA/
RA/
/RA/
DATE
07/25/24
07/25/24
07/29/24
Enclosure
U.S. NUCLEAR REGULATORY COMMISSION
Inspection Report
Docket Number:
05000416
License Number:
Report Number:
Enterprise Identifier:
I-2024-002-0003
Licensee:
Entergy Operations, Inc.
Facility:
Grand Gulf Nuclear Station
Location:
Port Gibson
Inspection Dates:
April 1, 2024, to June 30, 2024
Inspectors:
R. Azua, Senior Reactor Inspector
S. Hedger, Sr Emergency Preparedness Inspector
J. Melfi, Project Engineer
E. Powell, Resident Inspector
A. Smallwood, Sr Resident Inspector
Approved By:
Jeffrey E. Josey, Chief
Reactor Projects Branch C
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 Grand Gulf Nuclear 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
Engineered Safety Feature Room Cooler 1A309 Elevated Temperature
Cornerstone
Significance
Cross-Cutting
Aspect
Report
Section
Mitigating
Systems
Green
Open/Closed
None (NPP)
An NRC Green finding and associated non-cited violation of 10 CFR Part 50, Appendix B,
Criterion III, Design Control, was identified when inspectors noted temperatures in the
1A309 engineered safety feature room in the auxiliary building exceeded 90F as described in
the Updated Final Safety Analysis Report, sections 3.11.4.3, 9.4.5.2.5 and 9.4.5.5.4. The
licensee failed to translate this design change associated with the power uprate into adequate
control measures.
Failure to Detect and Prevent Failure of A Safety Related Motor Operated Valve
Cornerstone
Significance
Cross-Cutting
Aspect
Report
Section
Mitigating
Systems
Green
Open/Closed
[H.11] -
Challenge the
Unknown
The inspectors are documenting an NRC identified Green finding and associated non-cited
violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, when the threads failed
in the stem nut of a motor operated valve in the residual heat removal system. Specifically,
when the stem nut failed in the E12-F064A, residual heat removal minimum flow to the
suppression pool valve, the system became inoperable. This inoperability resulted from
incorrect acceptance limits used during periodic valve diagnostic testing.
Failure to Take Timely Corrective Action for Error in Drywell Temperature Measurement
Cornerstone
Significance
Cross-Cutting
Aspect
Report
Section
Mitigating
Systems
Green
Open/Closed
None
The inspectors identified a finding and associated Green non-cited violation of 10 CFR 50,
Appendix B, Criterion XVI, Corrective Action, for the licensees failure to promptly correct a
condition adverse to quality. Specifically, from December 13, 2001, to June 30, 2024,
Technical Specifications 3.6.5.5 contained acceptance criteria for drywell temperature that
was not justified by analysis, and thus were nonconservative. This was contrary to Regulatory
Guide 1.239 guidance, which stated that licensees must take prompt action for technical
specifications that are not restrictive enough to ensure plant safety. The licensee identified
3
this issue in December of 2001, but failed to take corrective action to properly amend the
technical specifications as of June 30, 2024.
Failure to Have Environmental Qualification Evaluation for Safety-Related Rosemount
Cornerstone
Significance
Cross-Cutting
Aspect
Report
Section
Mitigating
Systems
Green
Open/Closed
None (NPP)
The inspectors identified a finding of very low safety significance (Green) and associated non-
cited violation of Title 10 of the Code of Federal Regulations 50.49, Environmental
Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants, for the
licensees failure to qualify an item of electrical equipment important to safety by acceptable
methods. Specifically, installed Rosemount transmitters were not in the configuration qualified
by the licensees environmental qualification reports. Since the configuration in the plant is not
in accordance with the environmental qualification report, the licensee is required to qualify
the Rosemount transmitters by one of the methods as described in Title 10 of the Code of
Federal Regulations 50.49 to qualify the configuration in the plant.
Additional Tracking Items
None.
4
PLANT STATUS
At the beginning of the inspection period Grand Gulf Nuclear Station, Unit 1, was operating at
74 percent rated thermal power (RTP) due to replacement of a condensate booster pump and
rod pattern adjustments. On April 5, 2024, the unit reached 100 percent RTP. On May 11, 2024,
the unit powered down to 10 percent of RTP to repair a recirculation pump hydraulic motor leak
in the drywell. After repairs to the recirculation pump motor were completed, rod pattern
adjustments were accomplished during the ensuing power ascension. The unit achieved full
RTP on May 15, 2024, where it remained at or near for the rest 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
71111.04 - Equipment Alignment
Partial Walkdown Sample (IP Section 03.01) (1 Sample)
The inspectors evaluated system configurations during partial walkdowns of the following
systems/trains:
(1)
low pressure core spray system on May 24, 2024.
71111.05 - Fire Protection
Fire Area Walkdown and Inspection Sample (IP Section 03.01) (3 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)
division 2 switchgear room control building 111-foot elevation on May 14, 2024
(2)
fire area 1A401 in the auxiliary building 166-foot elevation on May 16, 2024
(3)
control building licensed operator control room 166-foot elevation on May 29, 2024
5
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 brigade drill response to fire division 2 engineered safety feature
diesel generator day tank on May 31, 2024.
71111.07A - Heat Exchanger/Sink Performance
Annual Review (IP Section 03.01) (1 Sample)
The inspectors evaluated readiness and performance of:
(1)
division 1, room 1A309, switchgear engineered safety feature room cooler on the
139-foot elevation of the auxiliary building on April 30, 2024
71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
(1)
The inspectors observed and evaluated licensed operator training scenario in the
simulator on May 29, 2024.
71111.12 - Maintenance Effectiveness
Maintenance Effectiveness (IP Section 03.01) (1 Sample)
The inspectors evaluated the effectiveness of maintenance to ensure the following
structures, systems, and components (SSCs) remain capable of performing their intended
function:
(1)
residual heat removal E12-F064A stem nut failure and repair on June 17, 2024.
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management Sample (IP Section 03.01) (3 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)
fuse failure associated with CR-GGN-2024-02597 on April 19, 2024
(2)
drywell high temperature due to drywell chiller unplanned maintenance on May 3,
2024
(3)
protected system lineup during division 2 standby service water maintenance on
May 6, 2024
6
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (4 Samples)
The inspectors evaluated the licensee's justifications and actions associated with the
following operability determinations and functionality assessments:
(1)
CR-GGN-2024-02428 division 2 emergency diesel generator load sequencing
April 11, 2024
(2)
CR-GGN-2024-02568 main steam isolation valve B21-F022D functional test on
April 16, 2024
(3)
CR-GGN-2024-02102 Rosemount model 1153 transmitter environmental qualification
on May 6, 2024
(4)
CR-GGN-2024-03127 main steam isolation valve B21-F022C functional test on
May 12, 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) (1 Sample)
(1)
work order (WO) 54001061 standby service water semi-annual fan inspection and
repair on May 22, 204
Surveillance Testing (IP Section 03.01) (3 Samples)
(1)
WO 54099571 standby liquid control division 2 quarterly pump run on April 8, 2024
(2)
WO 54100117 main steam line flow isolation functional test on June 7, 2024
(3)
WO 54147901 division 1 residual heat removal, low pressure core injection
subsystem, motor operated valve functional test on June 18, 2024
Inservice Testing (IST) (IP Section 03.01) (1 Sample)
(1)
WO 54062068 main steam isolation valve functional testing on May 17, 2024
71114.02 - Alert and Notification System Testing
Inspection Review (IP Section 02.01-02.04) (1 Sample)
(1)
The inspectors evaluated the maintenance and testing of the alert and notification
system between June 18, 2022, and June 8, 2024.
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 June 18, 2022, and June 8, 2024. Inspectors also evaluated
7
the licensee's ability to staff their emergency response facilities in accordance with
emergency plan commitments.
71114.04 - Emergency Action Level and Emergency Plan Changes
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 June 18, 2022, and June 8, 2024. 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 June 18, 2022, and June 8, 2024. 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
OTHER ACTIVITIES - BASELINE
71151 - Performance Indicator Verification
The inspectors verified licensee performance indicators submittals listed below:
IE03: Unplanned Power Changes per 7000 Critical Hours Sample (IP Section 02.02) (1 Sample)
(1)
April 1, 2023, through March 31, 2024
IE04: Unplanned Scrams with Complications (USwC) Sample (IP Section 02.03) (1 Sample)
(1)
April 1, 2023, through March 31, 2024
MS05: Safety System Functional Failures (SSFFs) Sample (IP Section 02.04) (1 Sample)
(1)
April 1, 2023, through March 31, 2024
EP01: Drill/Exercise Performance (DEP) Sample (IP Section 02.12) (1 Sample)
(1)
January 1, 2023, through March 31, 2024
EP02: Emergency Response Organization (ERO) Drill Participation (IP Section 02.13)
(1 Sample)
(1)
January 1, 2023, through March 31, 2024
8
EP03: Alert And Notification System (ANS) Reliability Sample (IP Section 02.14) (1 Sample)
(1)
January 1, 2023, through March 31, 2024
INSPECTION RESULTS
Engineered Safety Feature Room Cooler 1A309 Elevated Temperature
Cornerstone
Significance
Cross-Cutting
Aspect
Report
Section
Mitigating
Systems
Green
Open/Closed
None (NPP)
An NRC Green finding and associated non-cited violation of 10 CFR Part 50, Appendix B,
Criterion III, Design Control, was identified when inspectors noted temperatures in the
1A309 engineered safety feature room in the auxiliary building exceeded 90oF as described
in the Updated Final Safety Analysis Report, sections 3.11.4.3, 9.4.5.2.5 and 9.4.5.5.4. The
licensee failed to translate this design change associated with the power uprate into adequate
control measures.
Description: The engineered safety feature (ESF) switchgear divisions 1 and 2 busses are in
the control building. The switchgear power is then routed from the control building to the
reactor building. There are 10 total ESF switchgear rooms in the reactor building on the 2nd,
3rd and 4th levels. Per the initial design calculations, the ESF switchgear rooms had a
maximum design temperature during normal operations of 104F, which is tracked by the
licensee in the control room logs and governed by the technical requirements manual (TRM)
specification 6.7.1. The basis for this temperature limit is the initial starting temperature of the
rooms during a loss of coolant accident (LOCA) as described in the updated final safety
analysis report (UFSAR) in sections 3.11.4.3, 9.4.5.2.5, 9.4.5.5.4 and shown in table 3.11-1.
The engineering analysis conducted for the extended power uprate as described in the safety
evaluation associated with license amendment 191 issued on July 18, 2012, stated in part
that the normal operating temperature for rooms 1A308 and 1A309 was to be 90F or less.
This change affected the LOCA calculations associated with these two ESF switchgear
rooms located in the reactor building.
The ESF switchgear room coolers in the reactor building may be cooled by two different
service water systems. During accident conditions, the standby service water (SSW) would
cool the rooms. The SSW system is the credited ESF cooling system for the plant. During
normal operations the system utilizes the plant service water (PSW) system. During warmer
months, during normal operations, the temperatures in the switchgear rooms can exceed
104F due to the high load on the PSW system.
On October 23, 2023, the licensee created condition report (CR) CR-GGN-2023-16422 to
report abnormal vibrations on the ESF switchgear room cooler in room 1A208. Engineering
evaluated the condition as part of CR-GGN-2023-16454 and determined the room cooler
could be repaired at power if room temperature did not exceed the 104oF limit of the TRM,
section 6.7.1. The residents took this information and began researching the ESF switchgear
room cooler design basis and requirements. During this investigation the resident staff
became aware of lower normal operating temperature for rooms 1A308 and 1A309 as
described in the UFSAR. The residents also knew from plant tours and daily plant status
updates that all the ESF switchgear rooms were being tracked during warmer months when
cooled from PSW to ensure they did not exceed the 104oF limit from TRM, section 6.7.1. The
9
residents confirmed on October 27, 2023, that rooms 1A308 and 1A309 had exceeded 90F
during the summer and challenged operations staff on operability of the rooms as
documented in CR-GGN-2023-16535. Room 1A309 was still over 90F on some other days in
October. Engineering gathered plant data from the PDS system and confirmed room 1A309
exceeded the maximum normal operating temperature of 90F more than 100 days, and the
1A308 room exceeded 90F on one day during a 2-year period. Neither room ever exceeded
the 104F TRM 6.7.1 limit. The TRM 6.7.1 specification did not account for the UFSAR 90oF
normal operating temperature for the 1A308 and 1A309 switchgear room coolers. Operations
was unaware of this change to the UFSAR and did not have measures in place to control the
normal operating temperature of rooms 1A308 or 1A309. The licensee created CR-GGN-
2023-17131 observing that the ESF switchgear rooms on division 1 are typically warmer than
the rooms on division 2, and that the rooms are warmer when cooled by PSW then when
cooled by SSW. When the SSW is aligned to these rooms, the temperatures drop rapidly to
the cooler temperatures of the SSW basins.
Licensee procedure EN-LI-113, titled Licensing Basis Document Change Process,
revision 23, states in part that the licensee identify all impacted license basis documents
associated with a change. The licensee failed to identify this update to the UFSAR lowering
the normal operating temperatures of the 1A308 and 1A309 switchgear rooms should be
controlled and monitored. 10 CFR 50 Appendix B, Criterion III, Design Control states in part
that measures shall be established to assure the design basis are correctly translated into
specifications, drawings, procedures, and instructions. In this case, measures have not been
established to procedures, instructions, or specifications to limit the normal operating
temperature of the 1A308 and 1A309 ESF switchgear rooms as described in the UFSAR.
Corrective Actions: The licensee entered this condition into their corrective action program.
Corrective Action References: The licensee entered this condition into their corrective action
program as CR-GGN-2023-17749 and CR-GGN-2023-17750.
Performance Assessment:
Performance Deficiency: The failure to verify all impacted license basis documents in
accordance with station procedure EN-LI-133, revision 22, section 7.2.1, step 2.a.2 was a
performance deficiency. Specifically, the licensee failed to capture the 90F temperatures limit
described in the FSAR for rooms 1A308 and 1A309 into an actionable task for the station and
translate it to other license basis documents. Engineered safety feature room cooler
temperatures are controlled by the TRM specification 6.7.3. The area temperature limit for
1A308 and 1A309 is 104F per the TRM specification 6.7.3. The current design basis
temperature limit for the 1A308 and 1A309 rooms is 90F.
Screening: The inspectors determined the performance deficiency was more than minor
because it was associated with the Human Performance attribute of the Mitigating Systems
cornerstone and adversely affected the cornerstone objective to ensure the availability,
reliability, and capability of systems that respond to initiating events to prevent undesirable
consequences. Specifically, the licensee did not establish measures to correctly translate the
normal operating design basis limit of 90F for the 1A308 and 1A309 ESF switchgear rooms.
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 that this finding is of very low safety significance (Green) significance
because the finding did not represent a deficiency affecting design or qualification of a
10
mitigating structure, system, or component; did not involve the loss of a single-train technical
specification (TS) system longer than its TS allowed outage time; did not represent the loss of
probabilistic risk assessment (PRA) function one train of a multi-train system for greater than
its TS allowed outage time; did not represent the 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 the loss of a PRA system and/or function
as defined in the PRIB or the licensees 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
the loss of the PRA function of one or more non-TS trains of equipment designated as risk
significant in accordance with the licensees maintenance rule program for greater than
3 days. Additionally, the finding did not involve external events mitigating systems, the reactor
protection system, fire brigade, or flexible coping strategies.
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. This violation is not indicative of current licensee performance; therefore, no
cross-cutting aspect is identified.
Enforcement:
Violation: As required, in part, by 10 CFR Part 50, appendix B, criterion III, Design Control,
requires in part measures shall be established to assure the design basis are correctly
translated into specifications, drawings, procedures, and instructions.
Contrary to the above, from July 18, 2012 to March 8, 2024, the licensee failed to establish
measures to assure the design basis is correctly translated into specifications, drawings,
procedures and instructions. Specifically, the licensee failed to ensure that the temperature
limitations during normal operation of 90F for the engineered safety feature room coolers
located in rooms 1A308 and 1A309 as described in the UFSAR were translated into
procedures or instructions to limit temperature in these rooms.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with
Section 2.3.2 of the Enforcement Policy.
Failure to Detect and Prevent Failure of A Safety Related Motor Operated Valve
Cornerstone
Significance
Cross-Cutting
Aspect
Report
Section
Mitigating
Systems
Green
Open/Closed
[H.11] -
Challenge the
Unknown
The inspectors are documenting an NRC identified Green finding and associated non-cited
violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, when the threads failed
in the stem nut of a motor operated valve in the residual heat removal system. Specifically,
when the stem nut failed in the E12-F064A, residual heat removal minimum flow to the
suppression pool valve, the system became inoperable. This inoperability resulted from
incorrect acceptance limits used during periodic valve diagnostic testing.
Description: On February 13, 2024, a license operator discovered the residual heat
removal (RHR) E12-F064A division 1 minimum flow to the suppression pool valve indicated
closed on the control panel. This valve is normally open when the RHR system is in standby.
An operator was sent to the valve location to determine the position. After approximately
three minutes, the position indication in the control room switched from closed to open. By
comparing the stem positions on the RHR divisions 1 and 2 valves, the local operator
determined the valve to be closed while having an open indication in the control room.
11
Further investigation found the motor operator to be running continuously and at
approximately 200F. The breaker supplying power to the motor operated valve was opened,
and the RHR division 1 system was declared inoperable.
The licensee opened work order 54113910 to troubleshoot the valve and it was found the
inner stem nut threads had worn and were not engaged with the threads on the valve stem.
This allowed the motor to turn and rotate controlling the valve limits switches and position
indications without engaging the torque switches. A new stem nut was threaded by the
licensee under the same work order, installed, tested for functionality, and the valve was
returned to service. After a post maintenance test was performed on the valve using
approved procedures, the licensee declared the division 1 RHR system to be operable and
exited the applicable technical specification limiting condition for operation.
When reviewing the trend for stem nut wear for tests conducted from February 8, 2011, to
January 18, 2023, the resident staff noted an abnormal trend. The final test in the series of
four periodic component monitoring tests measured less stem nut wear than the previous
three diagnostic tests. Stem nut wear was measured at 0.0782 inches on February 8, 2011,
0.0853 inches on April 30, 2013, 0.0846 inches on January 24, 2019, and 0.0668 inches on
January 18, 2023. This trend in stem nut wear does not match the expected trend of
increasing wear until the stem nut would be replaced before reaching wear limits. The
resident staff inquired if indications of wear debris from the stem nut were present during the
troubleshooting and repair. Initially, reports of debris were uncertain, but after questioning
from NRC residents and further inspection, debris from the failing stem nut was observed.
Further complicating the detection of wear and preventing failure of the E12-F064A stem nut,
the wrong administrative limits were in place to detect excess stem nut wear. For all the
measurements before the valve failure, the administrative limit in place was 0.0958 inches.
The correct administrative limit to detect wear and prevent failure for this valve was 0.0353
inches. The E12-F064A valve was more than this administrative wear limit from the first
diagnostic test performed on February 11, 2011. The licensee applied the correct
administrative limit to the post repair valve testing completed on February 15, 2024. The
correct administrative limits to detect and prevent stem nut wear leading to valve failure; are
documented in licensee procedure EN-DC-312, titled Motor Operated Valve Test Data
Review.
Corrective Actions: The licensee replaced the stem nut using WO 54113910.
Corrective Action References: This issue was entered into licensee's corrective action
program as CR-GGN-2024-00790 and CR-GGN-2024-00820.
Performance Assessment:
Performance Deficiency: Title 10 CFR Part 50, Appendix B, Criteria XI, Test Control,
requires, in part, that licensees establish a test program to demonstrate that structures,
systems and components will perform satisfactorily in service is identified and performed in
accordance with written test procedures which incorporate the requirements and acceptance
limits contained in acceptable design documents. The failure to use the correct acceptance
limits as outlined in station procedure EN-DC-312 for measuring stem nut wear for the
E12-F064A valve was a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor
because it was associated with the Human Performance attribute of the Mitigating Systems
12
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, failed stem nut prevented the minimum flow valve, E12-F064A,
from being operable preventing RHR system from being able to perform its safety functions.
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 that this finding is of very low safety significance (Green) because the
finding did not represent a deficiency affecting design or qualification of a mitigating structure,
system, or component; did not involve the loss of a single-train technical specification (TS)
system longer than its TS allowed outage time; did not represent the loss of probabilistic risk
assessment (PRA) function one train of a multi-train system for greater than its TS allowed
outage time; did not represent the 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 the loss of a PRA system and/or function as defined
in the PRIB or the licensees 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 the loss of
the PRA function of one or more non-TS trains of equipment designated as risk-significant in
accordance with the licensees maintenance rule program for greater than 3 days.
Additionally, the finding did not involve external events mitigating systems, the reactor
protection system, fire brigade, or flexible coping strategies.
Cross-Cutting Aspect: H.11 - Challenge the Unknown: Individuals stop when faced with
uncertain conditions. Risks are evaluated and managed before proceeding. Specifically, the
licensee did not challenge and investigate a reported decreasing trend in stem nut wear for
the E12-F064A valve.
Enforcement:
Violation: Title 10 CFR Part 50, Appendix B, Criterion XI, Test Control, requires, in part,
requires, in part, that licensees establish a test program to demonstrate that structures,
systems and components will perform satisfactorily in service is identified and performed in
accordance with written test procedures which incorporate the requirements and acceptance
limits contained in acceptable design documents. The licensee established procedure
EN-DC-312, Motor Operated Valve Test Data Review, revision 9, as implementing
procedure to test the E12-F064A motor operated valve for stem nut wear, an activity affecting
quality.
Contrary to the above from February 8, 2011, to February 15, 2024, the licensee failed to
ensure the correct acceptance limits for measuring stem nut wear on the E12-F064A valve.
Specifically, the use of the incorrect acceptance limit led to failure of the stem nut and
inoperability of the E12-F064A valve rendering the RHR system inoperable and unable to
perform its safety functions. Contrary to the above from February 8, 2011, to February 15,
2024, the licensee failed to use the correct acceptance criteria for the administrative wear
limit for the stem nut on the E12-F064A valve. Specifically, the licensee was using an
incorrect wear limit of 0.0958 inches when the correct acceptance limit for excess wear was
0.0353 inches for the E12-F064A valve.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with
Section 2.3.2 of the Enforcement Policy.
13
Failure to Take Timely Corrective Action for Error in Drywell Temperature Measurement
Cornerstone
Significance
Cross-Cutting
Aspect
Report
Section
Mitigating
Systems
Green
Open/Closed
None
The inspectors identified a finding and associated Green non-cited violation of 10 CFR 50,
Appendix B, Criterion XVI, Corrective Action, for the licensees failure to promptly correct a
condition adverse to quality. Specifically, from December 13, 2001, to June 30, 2024,
Technical Specifications 3.6.5.5 contained acceptance criteria for drywell temperature that
was not justified by analysis, and thus were nonconservative. This was contrary to Regulatory
Guide 1.239 guidance, which stated that licensees must take prompt action for technical
specifications that are not restrictive enough to ensure plant safety. The licensee identified
this issue in December of 2001, but failed to take corrective action to properly amend the
technical specifications as of June 30, 2024.
Description: On December 13, 2001, the licensee identified potential issues related to
temperature measurement error in the instrumentation in the drywell. The drywell contains
12 temperature detectors at different elevations and radial placement in the drywell. Technical
specifications (TS) surveillance requirement (SR) 3.6.5.5.1 reports drywell temperature as the
average of the twelve temperature sensors. The TS 3.6.5.5 limit for drywell temperature is
135F to prevent exceeding the drywell design temperature limit of 330F during a small loss
of coolant accident. The licensee-initiated condition report CR-GGN-2001-01951 to document
and correct instrumentation errors in drywell temperature measurements. The licensee noted
that a surveillance result could be deemed applicable when in fact it is not bounded by the
safety analysis. At the conclusion of the corrective actions, the licensee calculated a 6.5oF
instrument error for drywell temperature measurement.
On May 2, 2024, the licensee had a drywell chiller in maintenance when an additional drywell
chiller skid tripped due to high chill water temperature. While the drywell chillers are not safety
related, they directly impact the ability of the licensee to maintain drywell temperature within
design limits. At 9:12 p.m., on May 2, 2024, the licensee declared the drywell inoperable per
T.S. 3.6.5.5 at 128.5F and entered the limiting condition for operation (LCO) allowed outage
time of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The drywell chillers were recovered before the 8-hour allowed outage time
lapsed, and the system was declared operable, and the TS LCO was excited when
temperature stabilized below 128.5F. The licensee notified the resident office of the TS entry
and exit. These events were entered into the licensees corrective action program in condition
report CR-GGN-2024-02941.
The resident staff met with the licensee regarding the basis for the 128.5F which was used to
enter the TS 3.6.5.5 LCO. The licensee provided calculation JC-Q1M71-N605-1 titled Drywell
Air Temperature Monitoring Uncertainty Calculation. This calculation estimates the amount of
uncertainty in instrument measurement of drywell air temperature. The residents questioned if
the use of 128.5F was an administrative limit separate from the TS requirement of 135F.
The licensees initial response was to follow the guidance of Administrative Letter 98-10
Dispositioning of Technical Specifications that are Insufficient to Assure Plant Safety, which
was applicable at the time. This guidance stated that upon discovery that a TS was not
sufficient to assure plant safety, the licensee should immediately impose administrative limits
in plant procedures to assure safety as a short-term interim measure, then promptly submit a
license amendment request (LAR) to correct the condition in accordance with 10 CFR 50,
14
Appendix B, Criterion XVI. The administrative letter gave two examples where the licensees
had inadequate responses to nonconservative TS, in which the licensees took over an
operating cycle (18 months) to resolve the deficient TS.
The inspectors noted that Regulatory Guide 1.239, Licensee Actions to Address
Nonconservative Technical Specifications, states the following:
Occasionally, licensees have determined that the TS may be nonconservative.
Examples include, but are not limited to, an improper or inadequate TS value, required
action, or completion time. When this occurs, in the interim, licensees typically enter
the nonconservative TS into their corrective action program, conduct an evaluation,
and, if necessary, institute administrative controls that instruct the operators to
maintain a more restrictive value for a particular parameter or to take a more
conservative action. Following the implementation of such administrative controls,
most licensees have properly considered reporting under 10 CFR 50.72,
10 CFR 50.73, or both and have promptly submitted a license amendment request to
correct the TS. However, some licensees have failed to comply with NRC reporting
requirements, significantly delayed submitting a license amendment request to correct
the TS, or improperly concluded that a license amendment request was unnecessary
if administrative controls are implemented.
Regulatory Guide 1.239 also superseded Administrative Letter 98-10 and endorsed the
guidance of NEI 15-03 Licensee Actions to Address Nonconservative Technical
Specifications. The guidance of NEI 15-03 repeated the actions of NRC Administrative
Letter 98-10, that, upon identification of a nonconservative TS, implement administrative
controls on a short-term interim basis, and promptly submit a license amendment. It also
states that the generic nature of a nonconservative TS does not alleviate the licensees
responsibility to promptly correct their TS. The NEI guidance also contains an example of a
violation of 10 CFR, appendix B, criterion XVI, Corrective Action, in which a licensee waited
5 years to correct a nonconservative TS related to diesel generator frequency.
Corrective Actions: The licensee initiated a condition report to enter this item into the
corrective action program.
Corrective Action References: CR-GGN-2024-04113
Performance Assessment:
Performance Deficiency: The licensee's failure to promptly correct a condition adverse to
quality was a performance deficiency. Specifically, since December 13, 2001, the licensee
has not taken prompt action to address uncertainty errors in drywell temperature
measurement associated with TS 3.6.5.5.
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 performance deficiency was associated with failure to correct
nonconservative TS, that would allow drywell operation outside the bounds of the safety
analysis.
15
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 impacted mitigating systems and used exhibit 2 to evaluate
the condition. The finding was determined to be of very low safety significance (Green)
because it (1) was not a deficiency affecting design or qualification of a mitigating system,
(2) does not represent a loss of the probability risk analysis (PRA) function of a single train
technical specification system for greater than allowed outage time, (3) does not represent a
loss of PRA function of one train of a multi-train technical specification system for greater
than its allowed outage time, (4) does not represent a loss of the PRA function of two
separate technical specification systems for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, (5) does not represent a
loss of PRA system and/or function as defined by the plant risk information e-book or the
licensees 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 (6) does not result in the loss of a high
safety-significant, nontechnical specification train for greater than 3 days.
Cross-Cutting Aspect: None No cross-cutting aspect is proposed, since this issue is not
indicative of current licensee performance
Enforcement:
Violation: Title 10 of the Code of Federal Regulations Part 50, Appendix B, Criterion XVI,
Corrective Action, states, in part, that measures shall be established to assure that
conditions adverse to quality are promptly identified and corrected.
Contrary to the above, from December 13, 2001, to June 30, 2024, a condition adverse to
quality was not promptly corrected. Specifically, the licensee determined that TS 3.6.5.5 was
nonconservative because the acceptance criteria of 135F was not justified based on
instrument error and should be 128.5F. However, the licensee had not taken action to
correct the nonconservative technical specifications as of June 2024.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with
Section 2.3.2 of the Enforcement Policy.
Failure to Have Environmental Qualification Evaluation for Safety-Related Rosemount
Cornerstone
Significance
Cross-Cutting
Aspect
Report
Section
Mitigating
Systems
Green
Open/Closed
None (NPP)
The inspectors identified a finding of very low safety significance (Green) and associated
non-cited violation of Title 10 of the Code of Federal Regulations 50.49, Environmental
Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants, for the
licensees failure to qualify an item of electrical equipment important to safety by acceptable
methods. Specifically, installed Rosemount transmitters were not in the configuration qualified
by the licensees environmental qualification reports. Since the configuration in the plant is
not in accordance with the environmental qualification report, the licensee is required to
qualify the Rosemount transmitters by one of the methods as described in Title 10 of the
Code of Federal Regulations 50.49 to qualify the configuration in the plant.
Description: During a walkdown of the plant, the inspectors identified a difference in
configurations for various models of Rosemount pressure and level transmitters. The
inspectors noted different components (i.e. fittings, valves, and plugs) on the process
connections for Rosemount series 1153 transmitters. The licensee documented the issue in
16
the corrective action program as condition report CR-GGN-2024-02102.
The inspectors reviewed the environmental qualification assessment reports (EQARs) for
model 1153 transmitters and concluded that configurations installed in the plant were not
environmentally qualified based upon the following.
For the 1153 model transmitter, the vendor manual design specifications broke down the
serial numbers, and a note for certain models specifically stated that the customer assumes
responsibility for qualifying process interfaces on these options. The vendor provides a
specific model option that was environmentally qualified and other model options with the
above note. Rosemount report 108025, titled Qualification report for pressure transmitters
Rosemount model 1153 series B, revision E states in part that mechanical and electrical
interfaces must be considered in qualification.
The qualification reports referenced in each EQAR state that the only tested configuration
qualifying these transmitters were ones with welded compression fit Swagelok valves on the
process connection interface and the vent/drain interface. The configurations installed in the
plant are not ones with the compression fit Swagelok connections and therefore the note
about end user having qualification responsibility is applicable. The licensee had no
documentation to support qualification of the alternate configurations which used different
fittings (1/4 inch national pipe thread fittings and sealant) than what was specified by the
vendor.
The inspectors identified examples of installed transmitters that were of the model containing
the note about customer responsibility for qualifying the equipment. The installed transmitters
on site are not in the tested configuration qualified in the EQARs and test reports. Therefore,
the inspectors concluded that these transmitters had not been environmentally qualified by
the licensee and as a result, there was no assurance that they could be relied upon to
perform the required function for the specified mission time under all accident conditions.
Corrective Actions: The licensee entered this issue into their corrective action program.
Corrective Action References: CR-GGN-2024-02102 and CR-HQN-2024-00332
Performance Assessment:
Performance Deficiency: Failure to maintain equipment required to be environmentally
qualified is reasonably in the licensee 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 Equipment Performance attribute of the Mitigating
Systems cornerstone and adversely affected the cornerstone objective to ensure the
availability, reliability, and capability of systems that respond to initiating events to prevent
undesirable consequences. Specifically, the licensee failed to ensure the installed
configuration of transmitters were environmentally qualified to perform their safety function.
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 this finding to 16 be of very low safety significance (Green) because it
was a deficiency affecting the design or qualification of equipment, but the equipment
maintained its PRA functionality.
17
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 50.49, Environmental Qualification of Electrical Equipment Important
to Safety for Nuclear Power Plants, section (f) requires that each item of electric equipment
important to safety must be qualified by one of the following methods:
(1) testing an identical item of equipment under identical conditions or under similar
conditions with a supporting analysis to show that the equipment to be qualified is
acceptable
(2) testing a similar item of equipment with a supporting analysis to show that the
equipment to be qualified is acceptable
(3) experience with identical or similar equipment under similar conditions with a
supporting analysis to show that the equipment to be qualified is acceptable
(4) analysis in combination with partial type test data that supports the analytical
assumptions and conclusions.
Contrary to the above since 1985 to June 30, 2024, the licensee failed to qualify an item of
electrical equipment important to safety by one of the following methods:
(1) testing an identical item of equipment under identical conditions or under similar
conditions with a supporting analysis to show that the equipment to be qualified is
acceptable
(2) testing a similar item of equipment with a supporting analysis to show that the
equipment to be qualified is acceptable
(3) experience with identical or similar equipment under similar conditions with a
supporting analysis to show that the equipment to be qualified is acceptable
(4) analysis in combination with partial type test data that supports the analytical
assumptions and conclusions.
Specifically, the licensee has numerous Rosemount transmitters, important to safety, installed
in multiple systems in a configuration that the licensee has not qualified that may not perform
their intended function under the environmental conditions described by 10 CFR 50.49.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with
Section 2.3.2 of the Enforcement Policy.
Observation: Condition Reports Missing Noun Names or Component
Identification Information
The inspectors reviewed documents entered the corrective action program for the following:
Complete, accurate, and timely documentation of the issue identified in the corrective
action program
Evaluation and timely disposition of operability and reportability issues
Consideration of extent of condition and cause, generic implications, common cause,
and previous occurrences
18
Classification and prioritization of the problems resolution commensurate with the
safety significance
Identification of corrective actions that are appropriately focused to correct the
problem
Completion of corrective actions in a timely manner commensurate with the safety
significance of the issue
Identification of negative trends associated with human or equipment performance
that can potentially impact nuclear safety
Operating experience is adequately evaluated for applicability, and applicable lessons
learned are communicated to appropriate organizations and implemented
For the inspection period, the inspectors noted some condition reports were missing either
noun names of components or component identification numbering. Missing information in
condition reports impacts the inspectors ability to readily determine safety significance of the
system or component, and potentially impacts the licensee ability to determine operability or
screen condition report priority. NRC inspectors sometimes must interact with regulatory
affairs personnel to disposition and trace corrective actions, locate operability determinations,
and determine the final disposition of issues in the corrective action program. This places
additional burden on the licensee staff and complicates inspections. Four examples of
condition reports missing key information are provided as examples. Condition report
CR-GGN-2024-02938 describes a gouge in a valve stem with component identification
number SP64F013. This condition report does not use the noun name to identify which
specific valve in the fire protection system is affected or inform the reviewing personnel which
portions of the fire protection system would potentially be impacted. Condition report
CR-GGN-2024-02960 describes leakage past valve SP64-003 prevents isolation of another
SP64 system valve. No noun name descriptions are provided for either valve making it
difficult to assess impact to the entire fire protection system. Condition report
CR-GGN-2024-02970 concerns a stripped stud on a heat exchanger with component
identification number 1P72B001B. The condition report does not use the noun name for the
component informing readers this is a heat exchanger for a drywell chiller. Condition report
CR-GGN-2024-03160 documents an oil leak on the division 2 recirculation pump oil fed
bearing temperature thermocouple. The component identification number is not provided
which complicates tracking of both issue resolution and component history.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
On July 10, 2024, the inspectors presented the integrated inspection results to
Brad Kapellas, Site Vice President and other members of the licensee staff.
On July 11, 2024, the inspectors presented the emergency preparedness program
inspection results to Brad Kapellas, Site Vice President and other members of the
licensee staff.
19
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Corrective Action
Documents
CR-GGN-YYYY-NNNN
2023-14298, 2023-14686, 2023-15879, 2024-02623,
2024-03265
Corrective Action
Documents
CR-GGN-
2023-16535, 2023-17131, 2023-17749, 2023-17750
Work Orders
52691834, 52962074
Corrective Action
Documents
CR-GGN-YYYY-NNNN
2024-00790, 2024-00820
Procedures
Motor Operated Valve Test Data Review
9
Work Orders
54113910
Corrective Action
Documents
CR-GGN-YYYY-NNNN
2024-02597, 2024-02941, 2001-01951
Corrective Action
Documents
CR-GGN-YYYY-NNNN
2024-02428, 2024-02102, 2024-02568, 2024-03127
Procedures
06-OP-1C71-SA-0001
MSIV closure RPS functional test
MSIV = Main Steam Isolation Valve
RPS = Reactor Protection System
101
Work Orders
54001061, 54099571, 54147901, 54100117, 54062068
Corrective Action
Documents
CR-GGN-YYYY-NNNN
2023-00759, 2024-00725, 2024-02961
Corrective Action
Documents
Resulting from
Inspection
CR-GGN-YYYY-NNNN
2024-03768, 2024-03769, 2024-03779
Tone Activated Receiver Repair Records, Date Range
8/5/22 to 5/2/2024
Various
ANS Siren Bi-Annual Unit Maintenance/Testing
Records, 2022-2023
Various
Miscellaneous
ENTANSERGGN201217
ANS Evaluation Report, Grand Gulf Nuclear
0
Procedures
01-S-10-3
Emergency Planning Department Responsibilities
25
CR-GGN-YYYY-NNNN
2024-00445
Corrective Action
Documents
Document Change
Request (DRN-HQN-)
2024-00074
20
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
GIN2022-00116
Quarterly Off-Hours Unannounced Quarterly
Everbridge Test - 2nd Quarter 2022
06/11/2022
GIN2022-00168
Quarterly Off-Hours Unannounced Quarterly
Everbridge Test - 3rd Quarter 2022
09/29/2022
GIN2022-00191
Quarterly Off-Hours Unannounced Quarterly
Everbridge Test - 4th Quarter 2022
11/04/2022
GIN2023-00029
Quarterly Off-Hours Unannounced Quarterly
Everbridge Test - 1st Quarter 2023
03/15/2023
GIN2023-00093
Quarterly Off-Hours Unannounced Quarterly
Everbridge Test - 2nd Quarter 2023
07/06/2023
GIN2023-00139
Quarterly Off-Hours Unannounced Quarterly
Everbridge Test - 3rd Quarter 2023
10/03/2023
GIN2023-00160
Quarterly Off-Hours Unannounced Quarterly
Everbridge Test - 4th Quarter 2023
12/27/2023
Miscellaneous
GIN2024-00033
Quarterly Off-Hours Unannounced Quarterly
Everbridge Test - 1st Quarter 2024
04/24/2024
Emergency Response Organization Notification
System
6, 11
Procedures
Emergency Notification
1
10CFR50.54(q)(2) Review - EPP 12-02, Radiological
Assessment Guide, Revision 19
11/10/2022
10CFR50.54(q)(3) Screening - Standing Order 23.011,
Revision 1
11/02/2023
10CFR50.54(q)(2) Review - Replacement of GGNS
Operational Hotline (WO-2021-027_10638823-000,
Revision 0)
04/05/2022
10CFR50.54(q)(3) Evaluation - EP Compensatory
Action 21-001
01/08/2021
10CFR50.54(q)(3) Evaluation - Joint Information
Center (JIC) Operations, 10-S-01-34, Revision 24
11/13/2023
Miscellaneous
Standing Order 23-011
Amplifying Information for EAL SU5.1, RCS Leakage
for 15 Minutes or Longer
10/28/2023
10-S-01-34
Joint Information Center (JIC) Operations
24
Procedures
Emergency Planning 10CFR50.54(q) Review Program
8
21
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
CR-GGN-YYYY-NNNN
2022-06842, 2022-08537, 2022-08538, 2022-08566,
2022-08567, 2022-08568, 2022-11290, 2023-13295,
2023-13392, 2023-13633, 2023-13694, 2023-13695,
2023-14758, 2023-15892, 2023-16893, 2024-00443,
2024-03321, 2024-03326
Corrective Action
Documents
Fleet Condition Reports
(CR-HQN-)
2023-03065
CR-GGN-YYYY-NNNN
2024-03742, 2024-03774, 2024-03790, 2024-03794,
2024-03804
Corrective Action
Documents
Resulting from
Inspection
Fleet Condition Reports
(CR-HQN-)
2024-00699
2023/02/22 Dress Rehearsal
03/21/2023
Equipment Important to Emergency Response (EITER)
4
GIN 2023-00049
Emergency Preparedness Letter of Agreement (LOA)
Annual Review - 2022
05/08/2023
GIN 2024-00017
Emergency Preparedness Letter of Agreement (LOA)
Annual ORO Review - 2023
01/07/2024
GIN#: 2022-00119
GGNS 2022-004 EOF Mini Drill Report
06/24/2024
GIN#: 2022-00172
GGNS 2022-011 EOF Mini Drill Report
09/29/2022
GIN#: 2022-00178
GGNS 2022/09/21 Red Team Drill
10/19/2022
GIN#: 2022-00203
2022-016 EOF Mini Drill - Blue Team Report - 11-
16/2022
11/23/2022
GIN#: 2022-00204
2022-016 EOF Mini Drill - Red/Yellow/Green Team
Report
11/23/2022
GIN#: 2022-00205
October 26, 2022, Green Team ERO Drill
11/23/2022
GIN#: 2022-00220
GGNS Second Half Semi-Annual Health Physics Drill
Report
12/14/2022
GIN#: 2023-00054
GGNS March 22, 2023 - Yellow Team NRC Exercise
Report
04/21/2023
GIN#: 2023-00059
GGNS 2023-002 EOF Mini Drill Report
06/15/2023
GIN#: 2023-00087
GGNS 2023-003 EOF Mini Drill Report
06/15/2023
GIN#: 2023-00096
GGNS 2023 First Half Semi-Annual Health Physics
Drill Report
06/29/2023
Miscellaneous
GIN#: 2023-00115
GGNS July 26, 2023 - Blue Team Drill Report
08/25/2023
22
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
GIN#: 2023-00140
GGNS September 13, 2023 - Red Team Drill Report
10/13/2023
GIN#: 2023-00155
GGNS October 25, 2023 - Green Team Drill Report
11/15/2023
GIN#: 2024-00010
2023 GGNS On-Site Accountability Drill
01/08/2024
GIN#: 2024-00014
Second Half Semi-Annual RP Drill 2023
01/10/2024
GIN: 2022-00120
GGNS First Half Semi-Annual Health Physics Drill
Report
07/12/2022
GIN: 2022-00169
GGNS Medical Drill Report
09/28/2022
GIN: 2022-00219
2022 Annual Media Training
12/19/2022
GIN: 2022-0137
Quarterly ERDS Test, Third Quarter 2022
07/13/2022
GIN: 2022/00221
2022 Annual Fire Brigade Drill with Offsite Support
12/21/2022
GIN: 2023-00010
Quarterly Emergency Response Facilities Inventory
Report - Fourth Quarter 2022
01/18/2023
GIN: 2023-00104
Quarterly Emergency Response Facilities Inventory
Report - Second Quarter 2023
08/3/2023
GIN: 2024-00005
Quarterly ERDS Test, 2nd, 3rd, and 4th Quarters 2023
01/04/2024
GIN: 2024-00018
2023 Annual Media Training
1/22/2024
GIN: 2024-00049
Quarterly Emergency Response Facilities Inventory
Report - First Quarter 2024
04/04/2024
GIN: 2024/00022
2023 Annual Fire Brigade Drill with Offsite Support
01/25/2024
GIN: 2024/00057
State and Local Annual Review of GGNS Emergency
Action Levels - 2023
04/25/2024
KLD TR-1263
Grand Gulf Nuclear Station, Development of
Evacuation Time Estimates, August 16, 2022
0
KLD TR-1355
Grand Gulf Nuclear Station, 2023 Population Update
Analysis, August 15, 2023
0
2024 Pre-NRC Emergency Planning Program
Inspection Assessment, Grand Gulf Nuclear Station
(GGNS)
02/22/2024
QA-7-2023-GGN-01
Audit Area Title: Emergency Preparedness Program;
Audit Period: April 5, 2023, through June 15, 2023
06/15/2023
Surveillance Report
Number QS-2024-ECH-
001
The 2024 Fleet NIOS Surveillance for Implementing
the 24-Month Frequency for the Emergency
Preparedness Program, April 22-25, 2024
04/25/2024
Work Tracker
2023-00268, 2024-00066
23
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Documents (WT-HQN-)
EN-EP-202-02
0
Declared Emergency Recovery and Re-entry
1
Corrective Action Program
52
QA Surveillance Process
13
Audit Process
43, 44
EP-FAP-EP-005
Emergency Preparedness Performance Indicators
18
Procedures
GIN: 2024-00012
2023 GGNS Site Medical Drill
01/09/2024
Work Orders
Work Orders
562859, 582078, 5409996
Corrective Action
Documents
CR-GGN-YYYY-NNNN
2023-00092, 2023-15577
Corrective Action
Documents
Resulting from
Inspection
CR-GGN-YYYY-NNNN
2024-03771, 2024-03773
GIN#: 2023-00004
Alert Notification System Test - January 2023
01/05/2023
GIN#: 2023-00052
Alert Notification System Test - April 2023
04/19/2023
GIN#: 2023-00135
Alert Notification System Test - September 2023
09/29/2023
GIN#: 2023-00145
Alert Notification System Test - November 2023
11/06/2023
GSES-LOR-WEX04
Supp Pool Makeup Timer Actuation/RFPT B Trip on
Failed Suction Press/ESF 21 Lockout with 17AC Bus
Lockout/LOCA/LOP/Failure of Div. 1 D/G Trip
15
Miscellaneous
GSES-LOR-WEX09
APRM Upscale/Loss of Feedwater
Heating/ATWS/Suppression Pool Leak (EP-2A, EP-3,
EP-4)
22
71151
Procedures
Regulatory Performance Indicator Process
21