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Latest revision as of 03:11, 2 February 2022
| ML20219A414 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 08/06/2020 |
| From: | David Proulx NRC/RGN-IV/DRP/RPB-C |
| To: | Emily Larson Entergy Operations |
| References | |
| IR 2020002 | |
| Download: ML20219A414 (25) | |
See also: IR 05000416/2020002
Text
August 6, 2020
Mr. Eric Larson, Site Vice President
Entergy Operations, Inc.
Grand Gulf Nuclear Station
P.O. Box 756
Port Gibson, MS 39150
SUBJECT: GRAND GULF NUCLEAR STATION - INTEGRATED INSPECTION
REPORT 05000416/2020002
Dear Mr. Larson:
On June 30, 2020, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at
Grand Gulf Nuclear Station. On July 16, 2020, the NRC inspectors discussed the results of this
inspection with you and other members of your staff. The results of this inspection are
documented in the enclosed report.
Two findings of very low safety significance (Green) are documented in this report. Both
findings involved violations of NRC requirements. Two Severity Level IV violations without an
associated finding are also documented in this report. 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.
E. Larson 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,
David L. Digitally signed by
David L. Proulx
Proulx Date: 2020.08.06
10:38:19 -05'00'
David L. Proulx, Acting Chief
Reactor Projects Branch C
Division of Reactor Projects
Docket No. 05000416
License No. NPF-29
Enclosure:
As stated
cc w/ encl: Distribution via LISTSERV
Non-Sensitive Publicly Available
SUNSI Review
Sensitive Non-Publicly Available
OFFICE SRI:DRP/C RI:DRP/C BC:DRS/EB1 BC:DRS/EB2 BC:DRS/OB
NAME TSteadham MThomas VGaddy NTaylor GWerner
DATE 07/29/2020 07/28/2020 07/28/2020 07/28/2020 07/28/2020
OFFICE BC:DRS/RCB TL:DRS/IPAT BC:DNMS/RxIB SPE:DRP/C ABC:DRP/C
NAME MHaire AAgrawal GWarnick CYoung DProulx
DATE 07/28/2020 08/05/2020 7/28/2020 07/27/2020 08/06/2020
U.S. NUCLEAR REGULATORY COMMISSION
Inspection Report
Docket Number: 05000416
License Number: NPF-29
Report Number: 05000416/2020002
Enterprise Identifier: I-2020-002-0004
Licensee: Entergy Operations, Inc.
Facility: Grand Gulf Nuclear Station
Location: Port Gibson, MS
Inspection Dates: April 1, 2020 to June 30, 2020
Inspectors: T. Steadham, Senior Resident Inspector
S. Hedger, Emergency Preparedness Inspector
M. Thomas, Resident Inspector
Approved By: David L. Proulx, Acting Chief
Reactor Projects Branch C
Division of Reactor Projects
Enclosure
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
Failure to Prevent Storage-Related Damage and Deterioration of Safety Related Tachometer
Cornerstone Significance Cross-Cutting Report
Aspect Section
Mitigating Green [P.2] - 71111.12
Systems NCV 05000416/2020002-01 Evaluation
Open/Closed
An NRC-identified Green finding and associated non-cited violation of Title 10 of the Code of
Federal Regulations, Part 50 (10 CFR Part 50), Appendix B, Criterion XIII, Handling, Storage,
and Shipping, was identified for the licensees failure to store an emergency diesel generator
tachometer in a manner to prevent damage and deterioration. Specifically, the licensee
installed a tachometer for the Division 3 emergency diesel generator which failed 7 days later
due to damage and deterioration as a result of improper storage conditions in the warehouse.
Failure to Provide Complete and Accurate Information in a License Amendment Request to
Change Emergency Action Level Requirements
Cornerstone Severity Cross-Cutting Report
Aspect Section
Not Applicable Severity Level IV Not Applicable 71114.04
Open/Closed
An NRC-identified Severity Level IV non-cited violation of 10 CFR 50.9(a) was identified for
the licensees providing of inaccurate information to the NRC in a license amendment request
for an emergency action level scheme change. Specifically, the licensee provided emergency
action level bases information and reactor pressure vessel thresholds used in emergency
action level statements that were not accurate.
2
Failure to Report a Safety System Functional Failure for the Standby Gas Treatment System
Cornerstone Severity Cross-Cutting Report
Aspect Section
Not Applicable Severity Level IV Not Applicable 71151
Open/Closed
An NRC-identified Severity Level IV non-cited violation of 10 CFR 50.73 was identified when
the licensee failed to report an event or condition that could have prevented the fulfillment of
the safety function of structures or systems that are needed to control the release of
radioactive material. Specifically, Licensee Event Report 05000416/2019-006-00 (ADAMS
Accession No. ML19322C805) documented a condition where the B train of the standby gas
treatment system was inoperable for a period of approximately 19 days. The licensee event
report reported this event in accordance with 10 CFR 50.73(a)(2)(i)(B) as a condition which
was prohibited by the plants technical specifications. However, the licensee event report did
not identify a reportable condition in accordance with 10 CFR 50.73(a)(2)(v), on the basis that
the A train of the standby gas treatment system was concurrently inoperable during a period
of time that the B train was inoperable.
Failure to Follow Surveillance Procedure Causes Inadvertent Emergency Core Cooling
System Actuation
Cornerstone Significance Cross-Cutting Report
Aspect Section
Initiating Events Green None 71152
Open/Closed
A self-revealed Green finding and associated non-cited violation of Technical
Specification 5.4.1 was identified when the licensee failed to follow Surveillance
Procedure 06-IC-1B21-R-0015, Revision 103, Attachment IV, step 5.4.1. The licensees
failure to isolate an instrument from a common reference leg resulted in an inadvertent
actuation of the Division 2 emergency core cooling system residual heat removal system and
its subsequent injection into the reactor vessel.
Additional Tracking Items
None.
3
PLANT STATUS
Grand Gulf Nuclear Station, Unit 1, began this inspection period shut down for Refueling
Outage 22. On May 20, 2020, the unit was restarted following the refueling outage. On
May 25, 2020, a reactor scram occurred from 66 percent power due to a malfunction in the main
turbine speed sensor protective logic. Following repairs to the speed sensors, the unit was
restarted on May 30, 2020, and reached full rated power on June 9, 2020. On June 20, 2020,
the unit was downpowered to 50 percent power due to high vibrations associated with main
feedwater pump B. Following repairs, power was returned to full rated power on June 28, 2020.
The unit was downpowered to 71 percent power for a rod pattern adjustment on June 29, 2020.
Following the rod pattern adjustment, the unit returned to full rated power on June 30, 2020.
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 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.
Starting on March 20, 2020, in response to the National Emergency declared by the President
of the United States on the public health risks of the coronavirus (COVID-19), resident
inspectors were directed to begin telework and to remotely access licensee information using
available technology. During this time the resident inspectors performed periodic site visits
each week and during that time conducted plant status activities as described in IMC 2515,
Appendix D; observed risk-significant activities; and completed on site portions of IPs. In
addition, resident and regional baseline inspections were evaluated to determine if all or a
portion of the objectives and requirements stated in the IP could be performed remotely. If the
inspections could be performed remotely, they were conducted per the applicable IP. In some
cases, portions of an IP were completed remotely and on site. The inspections documented
below met the objectives and requirements for completion of the IP.
REACTOR SAFETY
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) Refueling Outage 22 post core alteration verification on April 20, 2020
(2) Division 1 standby diesel generator partial alignment on June 12, 2020
4
71111.05 - Fire Protection
Fire Area Walkdown and Inspection Sample (IP Section 03.01) (5 Samples)
The inspectors evaluated the implementation of the fire protection program by conducting a
walkdown and performing a review to verify program compliance, equipment functionality,
material condition, and operational readiness of the following fire areas:
(1) Division 1 switchgear and battery room, Elevation 111 feet, Area 25A, on
June 19, 2020
(2) Control building, Elevation 148 feet, lower cable spreading room, Fire Zone OC402,
on June 19, 2020
(3) Containment, Elevation 208 feet, Area 11, on June 22, 2020
(4) Division 2 switchgear and battery room, Elevation 111 feet, Area 25A, on
June 22, 2020
(5) Auxiliary building, Elevation 139 feet, Plant Chiller Area 1A322, on June 22, 2020
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 the startup from Refueling Outage 22 and subsequent startup from
Forced Outage 23-01 on May 31, 2020.
71111.12 - Maintenance Effectiveness
Maintenance Effectiveness (IP Section 03.01) (4 Samples)
The inspectors evaluated the effectiveness of maintenance to ensure the following
structures, systems, and components (SSCs) remain capable of performing their intended
function:
(1) Valve 1B21F022A, main steam isolation valve A inboard failed local leak rate test on
May 6, 2020
(2) Valve 1E12F027A, residual heat removal A outboard containment isolation valve
failed during emergency core cooling system remote shutdown room testing on
May 6, 2020
(3) Valve 1B33F067A, recirculation pump A discharge maintenance isolation valve failure
due to flow induced vibration on May 6, 2020
(4) Division 3 emergency diesel generator tachometer storage on June 26, 2020
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:
5
(1) Risk assessment during reduced inventory operations during the week of
April 27, 2020
(2) Water level inventory control and risk management during reactor recirc pump A
discharge valve repair on April 30, 2020
(3) Reactor water cleanup system being credited as available for decay heat removal
through analysis on May 13, 2020
(4) Risk management during standby service water subsystem A maintenance on
June 19, 2020
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (6 Samples)
The inspectors evaluated the licensee's justifications and actions associated with the
following operability determinations and functionality assessments:
(1) Engineering Change EC-85565, seismic qualification of Division 3 standby diesel
generator tachometer on April 8, 2020
(2) 1E12F027A operability following motor actuator winding failure on May 6, 2020
(3) 1B21F022A inboard main steam isolation valve A indication on valve body at gasket
location on May 15, 2020
(4) Condition Report CR-GGN-2020-06338, Division 1 diesel generator jacket water heat
exchanger leak on May 15, 2020
(5) Condition Report CR-GGN-2020-07188, Division 2 diesel generator lube oil full flow
filter oil leak on June 18, 2020
(6) Condition Report CR-GGN-2020-05268, standby gas treatment system B failed
vacuum drawdown test on June 24, 2020
71111.18 - Plant Modifications
Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02)
(1 Sample)
The inspectors evaluated the following temporary or permanent modifications:
(1) Engineering Change 86810, drywell floor drain sump pump replacement on
May 15, 2020
71111.19 - Post-Maintenance Testing
Post-Maintenance Test Sample (IP Section 03.01) (6 Samples)
The inspectors evaluated the following post maintenance test activities to verify system
operability and functionality:
(1) Work Order 52859292, control rod stroke time testing on April 30, 2020
(2) Work Order 540076, reactor recirculation pump A discharge valve repairs on
April 30, 2020
(3) Work Order 52838593, post maintenance (as-left) local leak-rate test following repair
of main steam isolation valve 1B21F022A on May 13, 2020
6
(4) Work Order 544968, turbine speed probe replacement post maintenance test after
being damaged and causing SCRAM on May 30, 2020
(5) Engineering Change 72780, 72779, 76310, turbine control system upgrades on
June 12, 2020
(6) Work Order 541874, post maintenance test following actuator repair for residual heat
removal subsystem A shutoff valve E12F027A on June 19, 2020
71111.20 - Refueling and Other Outage Activities
Refueling/Other Outage Sample (IP Section 03.01) (1 Sample)
(1) The inspectors evaluated outage-related activities related to Refueling Outage 22
from April 1, 2020, through May 25, 2020.
(2) The inspectors evaluated outage-related activities related to Forced Outage 23-01
from May 25, 2020, to June 3, 2020.
71111.22 - Surveillance Testing
The inspectors evaluated the following surveillance tests:
Surveillance Tests (other) (IP Section 03.01) (2 Samples)
(1) Work Order 52840019-1, Procedure 06IC-1B21-R-0015, Attachment IV, "ATWS
Reactor Vessel Level Calibration Channel D," on April 20, 2020
(2) Work Order 52882122, Procedure 06-OP-1T48-R-0003, Revision 122, standby gas
treatment B logic and vacuum test on May 10, 2020
Inservice Testing (IP Section 03.01) (3 Samples)
(1) Work Order 542396-06, Procedure 06-OP-1B21-V-0001, Revision 122, main steam
isolation valve 1B21F022A stroke time testing on April 25, 2020
(2) Work Order 52853178, Procedure 06-OP-1E51-C-0005, Revision 111, reactor core
isolation cooling pump low pressure flow verification test on May 20, 2020
(3) Work Order 52912690-01, Procedure 06-OP-1E51-Q-0003, Revision 145, reactor
core isolation cooling system quarterly pump operability verification on May 22, 2020
Containment Isolation Valve Testing (IP Section 03.01) (1 Sample)
(1) Work Order 52838593, inboard A main steam isolation valve as-found local leak rate
test on April 24, 2020
71114.04 - Emergency Action Level and Emergency Plan Changes
Inspection Review (IP Section 02.01-02.03) (1 Sample)
(1) The licensee submitted a summary of an emergency plan change (Revision 80) to the
NRC on March 11, 2020. In addition, the licensee made emergency action level
changes (Procedure 10-S-01-1, Revisions 131 and 132) effective on
February 28, 2020. The inspectors conducted an in-office review of the changes from
April 1 to June 30, 2020. This evaluation does not constitute NRC approval.
7
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, 2019, through March 31, 2020
IE04: Unplanned Scrams with Complications (USwC) Sample (IP Section 02.03) (1 Sample)
(1) April 1, 2019, through March 31, 2020
MS05: Safety System Functional Failures (SSFFs) Sample (IP Section 02.04) (1 Sample)
(1) October 1, 2019, through March 31, 2020
71152 - Problem Identification and Resolution
Semiannual Trend Review (IP Section 02.02) (1 Sample)
(1) The inspectors reviewed the licensees corrective action program for potential
adverse trends associated with monitoring and trending of site programs and systems
that might be indicative of a more significant safety issue.
Annual Follow-up of Selected Issues (IP Section 02.03) (3 Samples)
The inspectors reviewed the licensees implementation of its corrective action program
related to the following issues:
(1) Condition Report CR-GGN-2020-05225, Part 21 for Schulz electric motor vertical end
bell bearing oil reservoirs recoating on June 24, 2020
(2) Condition Report CR-GGN-2020-03778, unsecured gas bottles in containment on
May 29, 2020
(3) Condition Report CR-GGN-2020-05246, inadvertent actuation and injection of
Division 2 residual heat removal following an anticipated transient without scram
instrument calibration on June 30, 2020
(4) Condition Report CR-GGN-2020-05763, uncontrolled pressure transient in
condensate full flow filter inlet line on June 26, 2020
71153 - Follow-up of Events and Notices of Enforcement Discretion
Event Follow-up (IP Section 03.01) (1 Sample)
(1) The inspectors evaluated Forced Outage 23-01 and the licensees response on
May 25, 2020.
8
INSPECTION RESULTS
Failure to Prevent Storage-Related Damage and Deterioration of Safety Related Tachometer
Cornerstone Significance Cross-Cutting Report
Aspect Section
Mitigating Green [P.2] - 71111.12
Systems NCV 05000416/2020002-01 Evaluation
Open/Closed
An NRC-identified Green finding and associated non-cited violation of 10 CFR Part 50,
Appendix B, Criterion XIII, Handling, Storage, and Shipping, was identified for the licensees
failure to store an emergency diesel generator tachometer in a manner to prevent damage
and deterioration. Specifically, the licensee installed a tachometer for the Division 3
emergency diesel generator which failed 7 days later due to damage and deterioration as a
result of improper storage conditions in the warehouse.
Description: On January 10, 2020, the licensee completed a maintenance outage window for
the Division 3 emergency diesel generator (EDG) and returned the generator to service.
During that maintenance window, the licensee replaced the tachometer under Work
Order 52852015. On January 17, 2020, operators attempted to start the EDG to troubleshoot
a Division 3 EDG trouble control room annunciator; however, the EDG immediately tripped.
Investigation determined that the cause of the trip was a failed tachometer. The licensee
replaced the tachometer and returned the EDG to service.
The licensee opened the cover of the failed tachometer and discovered evidence of moisture
and corrosion on some of the circuit boards. The licensee sent the failed tachometer to the
vendor for evaluation, who determined that the tachometer failed due to moisture and
corrosion causing an internal short and significant overheating. The tachometer was
originally received under Material Receiving Request 2-3500 on February 24, 1983. Through
discussions with the licensee, the inspectors learned that the tachometer remained in storage
with no assigned shelf life and no interim storage maintenance while the device was in
storage. Because the tachometer contained aluminum electrolytic capacitors, the inspectors
determined that a shelf life of 12 years should have been assigned as per licensee
Procedure EN-MP-112, Shelf Life Program, Revision 7, dated April 25, 2018.
The inspectors reviewed recent corrective action history and identified three work orders
written since 2017 documenting discrepancies with either shelf life or in-service maintenance
activities associated with EDG electrical components containing electrolytic capacitors.
These condition reports (CRs) were CR-GGN-2017-03974, CR-GGN-2018-02854, and
CR-GGN-2018-03716. Corrective actions related to these CRs failed to discover that the
Division 3 EDG tachometer, which also contained electrolytic capacitors, likewise had neither
any in-service maintenance, as required by Procedure EN-MP-140, In-Storage Maintenance
Program, Revision 2, dated February 19, 2016, nor an assigned shelf life as required by
Procedure EN-MP-112.
The inspectors expanded their CR history search and identified a total of 14 previous CRs
detailing general warehouse storage deficiencies with electronic components containing
electrolytic capacitors that also could have identified the storage discrepancies with the failed
tachometer. Of these 14 CRs, notable examples that could have identified the tachometer
storage discrepancy included CR-GGN-1999-01539, CR-GGN-2002-00621,
CR-GGN-2007-00873, and CR-GGN-2008-05950.
As required by Procedure EN-MP-112, items that exceed their shelf life shall be evaluated for
9
suitability for installation in the plant, including performing any maintenance that is deemed
necessary by such evaluation. If the evaluation for either of these three CRs had been
expanded to include all EDG components with electrolytic capacitors, the licensee would
have likely discovered the storage discrepancies with the tachometer. This discovery would
have prompted an evaluation which likely would have discovered the tachometer degradation
and allowed for the licensee to take appropriate corrective action prior to the defective
tachometer being installed in the plant. However, when the tachometer was installed, it had
exceeded what should have been its shelf life by almost 25 years with no review for suitability
for installation because this shelf life expiration was not recognized.
The issue was NRC-identified because the licensee failed to identify the missed opportunities
for a more thorough extent of condition evaluation from previous CRs related to improper
storage of components with electrolytic capacitors prior to the NRC identifying this to the
licensee.
Corrective Actions: The licensee replaced the tachometer with a suitable replacement and
performed an equipment failure evaluation. Planned corrective actions included an extent of
condition review of other items in storage.
Corrective Action References: Condition Report CR-GGN-2020-00516
Performance Assessment:
Performance Deficiency: The failure to store a Division 3 EDG tachometer replacement part
in a manner that prevented damage and deterioration in accordance with
Procedure EN-MP-112 was a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor
because it was associated with the equipment performance attribute of the Mitigating
Systems Cornerstone and adversely affected the cornerstone objective to ensure the
availability, reliability, and capability of systems that respond to initiating events to prevent
undesirable consequences. Specifically, the failed tachometer would have prevented the
Division 3 EDG from performing its intended safety function during a loss of power event.
Significance: The inspectors assessed the significance of the finding using Appendix A, The
Significance Determination Process (SDP) for Findings At-Power. The finding screened as
Green because the inspectors answered No to all six questions. Notably, the maximum
exposure time with a degraded tachometer was 7 days, which was less than the technical
specification allowed outage time of 14 days.
Cross-Cutting Aspect: P.2 - Evaluation: The organization thoroughly evaluates issues to
ensure that resolutions address causes and extent of conditions commensurate with their
safety significance. The finding had a cross-cutting aspect in the Problem Identification and
Resolution area associated with Evaluation because recent CRs detailed storage
discrepancies with EDG electrical components containing electrolytic capacitors, but failed to
identify the same deficiencies with the Division 3 EDG tachometer.
Enforcement:
Violation: Title 10 CFR 50, Appendix B, Criterion XIII, Handling, Storage, and Shipping,
requires, in part, that measures shall be established to control the handling, storage, and
preservation of equipment in accordance with work and inspection instructions to prevent
damage or deterioration.
10
Contrary to the above, from approximately February 24, 1983, until January 8, 2020, the
licensee failed to establish measures to control the handling, storage, and preservation of
equipment in accordance with work and inspection instructions to prevent damage or
deterioration. Specifically, the licensee failed to control the storage of the replacement
Division 3 emergency diesel generator tachometer, a safety-related component, in
accordance with Procedure EN-MP-112, Revision 7, and Procedure EN-MP-140, Revision 2,
to prevent damage or deterioration. The licensee failed to assign a shelf life and in-storage
maintenance and failed to keep the tachometer internals dry while in storage and during
handling.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with
Section 2.3.2 of the Enforcement Policy.
Failure to Provide Complete and Accurate Information in a License Amendment Request to
Change Emergency Action Level Requirements
Cornerstone Severity Cross-Cutting Report
Aspect Section
Not Severity Level IV Not 71114.04
Applicable NCV 05000416/2020002-02 Applicable
Open/Closed
An NRC-identified Severity Level IV non-cited violation of 10 CFR 50.9(a) was identified for
the licensees providing of inaccurate information to the NRC in a license amendment request
for an emergency action level scheme change. Specifically, the licensee provided
emergency action level bases information and reactor pressure vessel thresholds used in
emergency action level statements that were not accurate.
Description: The NRC approved an emergency action level (EAL) scheme change on
March 12, 2019 (ADAMS Accession No. ML19025A023), to allow Grand Gulf Nuclear Station
to adopt the Nuclear Energy Institute (NEI) 99-01, Revision 6, scheme. Subsequently, the
licensee identified that one of their current EAL thresholds could not be implemented in
accordance with their emergency classification procedure. A cross discipline review of draft
Procedure 10-S-01-1, Activation of the Emergency Plan, Revision 132, revealed that the
indicated reactor pressure vessel Level 2 was incorrectly stated as -42 inches, when it was
supposed to be -41.6 inches (Condition Report CR-GGN-2020-00631). This affected EAL
CA1.1, which is declared with loss of RPV [sic., Reactor Pressure Vessel] inventory as
indicated by RPV water level < -42 inches (Level 2). According to the EAL CA1.1 basis text,
the RPV level is associated with the Level 2 actuation setpoint for high pressure core spray
and reactor core isolation cooling, which occurs at -41.6 inches. The licensee evaluated the
error and made a change to the EAL to correct the EAL so that it reflected the correct Level 2
actuation set point.
The inspectors reviewed the licensees license amendment request, dated April 27, 2018
(ADAMS Accession No. ML18117A514), License Amendment Request, Adoption of
Emergency Action Level Schemes Pursuant to NEI 99-01," Revision 6, and the licensees
response to a request for additional information dated October 10, 2018 (ADAMS Accession
No. ML18284A041), to determine whether the request correctly described the Level 2
actuation set point. The inspectors identified:
Both the April 27, 2018, submittal, and the response to the NRCs request for
additional information (RAI), dated October 10, 2018, incorrectly indicated that the
Level 2 actuation set point was -42 inches associated with EAL CA1.1.
11
Both the April 27, 2018, and the October 10, 2018, submittals incorrectly indicated
that the Level 1 actuation set point was -150 inches associated with EAL CS1.1. The
correct Level 1 actuation set point is -150.3 inches.
In the response to RAI 5 on October 10, 2018, the licensee indicated that the phrase
due to the loss of decay removal capability... would be removed from the basis
text for EAL CU3.1. NRC staff requested that it be removed because of its potential to
lead to misclassification. Revised EAL documentation provided in the submittal
showed the text removed from the previously proposed text. However, this text was
included in Procedure 10-S-01-1, Revisions 131 and 132.
In response to RAI 6 on October 10, 2018, the licensee indicated that the phrase
...due to the loss of decay removal capability... would be removed from the basis text
for EAL CA3.1. NRC staff requested that it be removed because of its potential to
lead to misclassification. Revised EAL documentation provided in the submittal
showed the text removed from the previously proposed text. However, this text was
included in Procedure 10-S-01-1, Revisions 131 and 132.
The issue was NRC-identified because, although the licensee identified the Level 2 set point
error in CA1.1 and took action to correct the error before EAL implementation, the licensee
failed to address the failure to ensure that technical information provided to the NRC in
support of the license amendment request was complete and accurate in all material aspects.
In addition, the licensee failed to identify the three other examples of erroneous information
provided to the NRC identified above.
Corrective Actions: The licensee entered these issues into the corrective action program. In
addition, the licensee issued communications and just-in-time training to their emergency
response organization to inform them of the condition in lieu of documentation changes which
are in progress.
Corrective Action References: Condition Reports CR-GGN-2020-05947,
CR-GGN-2020-06914, and CR-GGN-2020-07514.
Performance Assessment: The inspectors determined this violation was associated with a
minor performance deficiency. Specifically, it was determined that the licensee failed to
maintain the effectiveness of the emergency plan; however, the performance deficiencies
were minor due to the continued effectiveness of the EALs despite the errors.
Enforcement: The ROPs significance determination process does not specifically consider
the regulatory process impact in its assessment of licensee performance. Therefore, it is
necessary to address this violation which impedes the NRCs ability to regulate using
traditional enforcement to adequately deter noncompliance.
Severity: This issue was determined to be a Severity Level IV violation using the NRC
Enforcement Policy, dated January 15, 2020, Section 2.3.11, Inaccurate and Incomplete
Information, and Section 6.9, Inaccurate and Incomplete Information or Failure to Make a
Required Report. The Enforcement Policy, Section 6.9.c.1, provides that a violation is
characterized as Severity Level III if the accurate information would have caused the NRC to
reconsider a regulatory position or undertake further inquiry. There are no corresponding
Severity Level IV examples. Through discussion with the NRCs Office of Nuclear Security
and Incident Response (NSIR), it was determined that had accurate information been
provided (or had the NRC known the information was inaccurate), the NRC license reviewer
would have used the request for additional information process to address these problems
with the license amendment request. Specifically, the licensee would have been required to
12
revise their proposed EALs so they could be implemented before the emergency action
scheme change was approved. Because the request for additional information is a routine
NRC process, it was concluded that the failure to provide accurate information to the NRC
would not have caused the NRC to undertake substantial further inquiry (a threshold for
Severity Level III) and, therefore, the violation was appropriately characterized as Severity
Level IV.
Violation: Title 10 CFR 50.9(a) states, in part, that information provided to the Commission
by a licensee shall be complete and accurate in all material respects.
Contrary to the above, on April 27, 2018, and August 10, 2018, information was provided to
the Commission by the licensee that was not complete and accurate in all material respects.
Specifically, the licensees EAL scheme change submittal documents contained EAL
declaration threshold values that did not correctly correlate to Level 1 and 2 set points. In
addition, the licensee submitted EAL basis text for approval that reflected edits requested by
the NRC during review, but after approval, the licensee implemented the version of the EAL
basis text that did not reflect the edits approved. The inaccurate information was material to
the NRC because the errors affected the thresholds for implementing the EALs as required
by NRC regulations.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with
Section 2.3.2 of the Enforcement Policy.
Failure to Report a Safety System Functional Failure for the Standby Gas Treatment System
Cornerstone Severity Cross-Cutting Report
Aspect Section
Not Severity Level IV Not 71151
Applicable NCV 05000416/2020002-03 Applicable
Open/Closed
An NRC-identified Severity Level IV non-cited violation of 10 CFR 50.73 was identified when
the licensee failed to report an event or condition that could have prevented the fulfillment of
the safety function of structures or systems that are needed to control the release of
radioactive material. Specifically, Grand Gulf Nuclear Station Licensee Event
Report 05000416/2019-006-00 (ADAMS Accession No. ML19322C805), documented a
condition where the B train of the standby gas treatment system was inoperable for a period
of approximately 19 days. The licensee event report reported this event in accordance with
10 CFR 50.73(a)(2)(i)(B), for a condition which was prohibited by the plants technical
specification. However, the licensee event report did not identify this condition as a
10 CFR 50.73(a)(2)(v) condition when the A train of the standby gas treatment system was
concurrently inoperable during a period of time that the B train was inoperable.
Description: Licensee Event Report (LER) 05000416/2019-006-00 documented a condition
where the B train of standby gas treatment system (SGTS) was inoperable for
approximately 19 days from September 4, 2019, until September 23, 2019. The LER was
submitted to the NRC pursuant to 10 CFR 50.73(a)(2)(i)(B) for a condition which was
prohibited by the plants technical specifications.
The LER stated that the A train of SGTS was operable and available during the event.
However, while reviewing LER 05000416/2019-006-00, the inspectors identified that on
September 19, 2019, the A train of SGTS was declared inoperable during surveillance
testing per Procedure 06-OP-1T48-Q-0002, Standby Gas Treatment System A Valve Test,
13
Revision 111. Surveillance Procedure 06-OP-1T48-Q-0002, Revision 111, required kill
switches to be installed as well as the lifting of wires. As a result, operators declared the A
train of SGTS inoperable, as it would no longer provide its function while in this condition
absent additional operator action.
Following the guidance in NUREG-1022, Revision 3, Event Report Guidelines 10 CFR 50.72
and 50.73, the inspectors determined that the concurrent inoperability of the A and B
trains of SGTS should have been reported to the NRC as a condition that could have
prevented the fulfillment of a safety function, pursuant to 10 CFR 50.73(a)(2)(v). Specifically,
both trains of SGTS were inoperable at the same time due to one or more equipment
problems associated with degraded charcoal in the B SGTS.
On approximately May 29, 2020, the inspectors informed the licensee of the information they
discovered regarding a potential condition that could have prevented the fulfillment of the
SGTS safety function. The licensee entered the concern into its corrective action program on
July 7, 2020.
Corrective Actions: The licensee entered the NRC concern into its corrective action program
for resolution.
Corrective Action References: Condition Report CR-GGN-2020-07879
Performance Assessment: The inspectors determined this violation was associated with a
minor performance deficiency. Specifically, the failure to report a condition that could have
prevented the fulfillment of a safety function is contrary to licensee Procedure EN-LI-108,
Event Notification and Reporting, Revision 18.
Enforcement: The ROPs significance determination process does not specifically consider
the regulatory process impact in its assessment of licensee performance. Therefore, it is
necessary to address this violation which impedes the NRCs ability to regulate using
traditional enforcement to adequately deter noncompliance.
Severity: In accordance with the NRC Enforcement Policy, dated January 15, 2020,
Section 6.9.d.9, this violation is being dispositioned as a Severity Level IV violation because
the licensee failed to make a report required by 10 CFR 50.73.
Violation: Title 10 CFR 50.73(a)(1), requires, in part, that a licensee shall submit a LER for
any event of the type described in this paragraph within 60 days after the discovery of the
event. Specifically, 10 CFR 50.73(a)(2)(v) requires that the licensee shall report any event or
condition that could have prevented the fulfillment of the safety function of structures or
systems that are needed to (1) shut down the reactor and maintain it in a safe shutdown
condition; (2) remove residual heat; (3) control the release of radioactive material; or
(4) mitigate the consequences of an accident.
Contrary to the above, on November 18, 2019, the licensee failed to submit a LER for an
event of the type described in 10 CFR 50.73(a) within 60 days after the discovery of the
event. Specifically, when the licensee declared the A train of SGTS inoperable for a
surveillance test with the B train of SGTS concurrently inoperable due to a failed charcoal
test, a condition existed that could have prevented the fulfillment of a safety function of a
system that is needed, in part, to control the release of radioactive material.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with
Section 2.3.2 of the Enforcement Policy.
14
Failure to Follow Surveillance Procedure Causes Inadvertent Emergency Core Cooling
System Actuation
Cornerstone Significance Cross-Cutting Report
Aspect Section
Initiating Events Green None 71152
Open/Closed
A self-revealed, Green finding and associated non-cited violation of Technical Specification 5.4.1.a was identified when the licensee failed to follow Surveillance
Procedure 06-IC-1B21-R-0015, ATWS Reactor Vessel Level Calibration, Revision 103,
Attachment IV, step 5.4.1, by not adequately isolating the instrument from the common
reference leg which resulted in the inadvertent actuation of the Division 2 emergency core
cooling system, residual heat removal system, and its subsequent injection into the reactor
vessel.
Description: On April 16, 2020, instrumentation and controls (I&C) technicians were
performing Surveillance Procedure 06-IC-1B21-R-0015, ATWS Reactor Vessel Level
Calibration, Revision 103. The procedure required the level instrument that was being
calibrated to be isolated from its normal sensing lines so a calibration instrument could be
installed on the level instrument test ports for calibration. Specifically,
Procedure 06-IC-1B21-R-0015, Revision 103, Attachment IV, step 5.4.1, required, in part, the
level instrument isolation valves to be closed. This specific step was a concurrent verification
step which required an additional instrumentation and control technician to verify the
instrument isolation valves were closed. Early in the morning on April 16, 2020, the
instrumentation and control technicians conducting this surveillance had initialed these steps
as complete on the basis that they believed they had adequately performed the steps and
continued with the surveillance for calibrating the anticipated transient without scram (ATWS)
reactor vessel level instrument.
At approximately 1:58 a.m. on April 16, 2020, the control room received a spurious Division 2,
Level 1, initiation signal, which caused the residual heat removal (RHR) B train to realign to
its low pressure coolant injection (LPCI) mode. A couple of seconds later, operators in the
main control room secured the RHR B pump and terminated the inadvertent injection event,
thereby preventing any significant event from occurring as a result of the inadvertent LPCI
injection to the reactor vessel.
The licensee determined that the inadequate isolation of the level instrument, in accordance
with Procedure 06-IC-1B21-R-0015, Revision 103, Attachment IV, step 5.4.1, from its
common reference leg allowed the calibration instrument to pressurize not only the level
instrument but the downstream common reference leg as well. The pressurization of this
common reference leg, which fed various other pressure and level transmitters, including an
emergency core cooling system (ECCS) Channel B reactor vessel level transmitter, caused
the RHR B LPCI initiation signal.
The licensees investigation concluded that the isolation valves were not fully closed as part
of the required steps in the surveillance procedure.
Corrective Actions: The licensees corrective actions included submitting an LER to the NRC
(LER 05000416/2020-001-00, Residual Heat Removal System Inadvertent Actuation Due To
Human Error, ADAMS Accession No. ML20167A150), conducting an Adverse Condition
15
Analysis, performing a failure investigation on the level transmitter manifold, instituting
additional supervisor oversight for Refueling Outage 22 during manipulation of critical
transmitters with shared sensing lines, and ensuring that when calibrating other engineered
safeguards features (ESF) instruments on shared sensing lines, those other instruments are
appropriately bypassed to ensure another inadvertent actuation could not happen.
Corrective Action References: Condition Report CR-GGN-2020-05246
Performance Assessment:
Performance Deficiency: The licensees failure to isolate the level instrument from the
common sensing line in accordance with Procedure 06-IC-1B21-R-0015, Revision 103,
Attachment IV, step 5.4.1, 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 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 failure to isolate the level instrument from the common
sensing line resulted in an inadvertent RHR injection event that upset plant stability and
challenged critical safety functions such as LPCI.
Significance: The inspectors assessed the significance of the finding using Appendix G,
Shutdown Safety SDP. By answering no to the relevant questions in Exhibit 2, Initiating
Events Screening Questions, the inspectors screened this finding as having very low safety
significance (Green). The finding did not increase the likelihood of an internal flood that could
cause a shutdown initiating event. Specifically, the inspectors believed that when considering
the time it would take for the shutdown initiating event to occur as a result of overfilling the
cavity, and also considering the resultant complex sequence of events that would precede a
shutdown initiating event, the likelihood of an internal flood causing a shutdown initiating
event was not impacted.
Cross-Cutting Aspect: None
Enforcement:
Violation: Technical Specification 5.4.1.a requires, in part, that written procedures shall be
established, implemented, and maintained covering the applicable procedures recommended
in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Regulatory Guide 1.33,
Revision 2, Appendix A, Section 8, lists procedures for control of measuring and test
equipment and for surveillance tests, procedures, and calibrations. Section 8.a provides
examples of equipment that should be calibrated and tested in accordance with those
procedures specified by Section 8. The licensee established Procedure 06-IC-1B21-R-0015,
ATWS Reactor Vessel Level Calibration, Revision 103, to meet the Regulatory Guide 1.33
requirement. Step 5.4.1 of Procedure 06-IC-1B21-R-0015 required the level instrument
isolation valves to be closed.
Contrary to the above, on April 16, 2020, the licensee failed to ensure that level instrument
valves were closed during reactor vessel level instrument calibration, which caused an
inadvertent ECCS actuation event.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with
Section 2.3.2 of the Enforcement Policy.
16
Observation: Semiannual Trend Review 71152
In the 4 quarter 2019 Integrated Inspection Report 05000416/2019004 (ADAMS Accession
th
No. ML20044F733), the inspectors documented a semiannual trend observation that
discussed a concern with the lack of rigorous monitoring and trending of site programs and
systems and how that could be a common theme with many documented findings.
Specifically, the inspectors documented that if system/program monitoring was not improved,
then failures to identify deficiencies could continue to occur.
In the current quarter, the inspectors remained concerned with some site trending/monitoring
aspects specifically related to the examples discussed below, which were also subjects of the
4th quarter 2019 integrated inspection report semiannual trend review.
1. Division 1 standby diesel generator (SDG) jacket water (JW) heat exchanger (HX)
leak: Condition Report CR-GGN-2020-06338 reported an NRC identified leak, or
evidence thereof, on May 15, 2020, following an NRC walkdown of the Division 1
SDG. This component leak was the subject of the semiannual trend observation
reported in the 4th quarter 2019 Integrated Inspection Report, and a Green non-cited
violation (NCV) documented in the 4th quarter 2019 Integrated Inspection Report as
well. The NRCs identification of a subsequent leak from the same location on the
same component demonstrates that monitoring and trending in this area could
continue to be improved. Following the NRC identification of this subsequent leak on
the Division 1 SDG JW HX, the licensee added a specific operator shiftly round check
to the operator logs to assure this leak continues to be monitored and trended. The
residents believed this additional action was reasonable to mitigate immediate
concerns regarding monitoring and trending of leakage in this area.
2. Division 3 SDG Failed Tachometer due to Inadequate Storage Maintenance:
Condition Report CR-GGN-2020-00516 reported the failure of the Division 3 SDG to
run following the installation of a faulty tachometer. The NRC determined that
inadequate storage maintenance was a cause for the failed tachometer that led to the
unplanned SDG inoperability. Level A storage ambient monitoring was the subject of
the semiannual trend observation reported in the 4th quarter 2019 Integrated
Inspection Report, and the inspectors noted that inadequate monitoring and trending
of storage conditions could have contributed to the faulty tachometer that was
installed in the Division 3 SDG.
3. Degraded Fire Wall: Condition Report CR-GGN-2020-07286 reported an NRC-
identified deficiency with a 3-hour rated fire wall. The monitoring aspect associated
with this deficiency was also the subject of the semiannual trend observation reported
in the 4th quarter 2019 Integrated Inspection Report. Specifically, the inspectors
believe that a lack of monitoring of fire program equipment allowed for a 3-hour rated
fire wall to degrade and go unnoticed until identified by the inspectors.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
On June 29, 2020, the inspectors presented the Emergency Plan Revision In-Office
Review inspection results to Mr. M. Lewis, Emergency Preparedness Manager, and
other members of the licensee staff.
On July 16, 2020, the inspectors presented the integrated inspection results to
Mr. E. Larson, Site Vice President, and other members of the licensee staff.
17
DOCUMENTS REVIEWED
Inspection Type Designation Description or Title Revision or
Procedure Date
71111.04 Drawings M-1070A Standby Diesel Generator System 46
71111.04 Drawings M-1070B Standby Diesel Generator System 39
71111.04 Drawings M-1070C Standby Diesel Generator System 22
71111.04 Drawings M-1070D Standby Diesel Generator System 19
71111.04 Procedures 04-1-01-P75-1 Standby Diesel Generator System 114
71111.05 Corrective Action CR-GGN- 2020-07286
Documents
71111.05 Fire Plans Fire Pre-Plan A-27 Centrifugal Chiller Area 1A322 1
71111.05 Fire Plans Fire Pre-Plan A-47 Containment Fuel Pool, Steam Separator Storage Area, 1
Reactor Containment Area
71111.05 Fire Plans Fire Pre-Plan C-03 Division I SWGR Area and Battery Room 4
71111.05 Fire Plans Fire Pre-Plan C-07-1 Division II Switchgear Room and Battery Room 4
71111.05 Procedures EN-DC-330 Fire Protection Program 5
71111.12 Corrective Action CR-GGN- 2020-02838, 2020-03667, 2020-04079, 2020-04319, 2020-
Documents 04323, 2020-04409, 2020-04412, 2020-04430, 2020-
04522, 2020-04810, 2020-05009, 2020-05109, 2020-
05129, 2020-05154, 2020-05234, 2020-05291, 2020-
05351, 2020-05361, 2020-05447
71111.12 Drawings M-1078A P & I Diagram Reactor Recirculation System 38
71111.12 Drawings M-1085B P & I Diagram Residual Heat Removal System 64
71111.12 Procedures 04-1-05-B21-1 Main Steam and Drywell Pressure Transmitters 2
71111.12 Procedures 06-OP-1M61-V-0002 Local Leak Rate Test - Air Using Graftel Model 9623-7 13
Leak Rate Monitor
71111.12 Work Orders WO 00541874, 00542396
71111.13 Miscellaneous Operator Logs
71111.13 Miscellaneous Section 5.4.8 UFSAR
71111.13 Procedures EN-OP-119 Protected Equipment Postings 12
71111.15 Corrective Action CR-GGN- 2020-04319, 2020-04412, 2020-04430, 2020-04522, 2020-
Documents 05268, 2020-05270, 2020-05404, 2020-05405, 2020-
05427, 2020-05447, 2020-06046, 2020-06070, 2020-
06338, 2020-07188
18
Inspection Type Designation Description or Title Revision or
Procedure Date
71111.15 Drawings M-1085B P & I Diagram Residual Heat Removal System 64
71111.15 Engineering EC-85565 Replace 1P81K001 Div III EDG Speedswitch (SSA-1) 0
Changes
71111.15 Miscellaneous 00200225 Procurement Engineering Evaluation
71111.15 Miscellaneous BOP-VE-20-179 Ultrasonic examination of heat exchanger during RF22
71111.15 Miscellaneous E100.0 Technical Specification for Environmental Safety Related 8
Parameter
71111.15 Miscellaneous IEE-344-2013
71111.15 Procedures 06-OP-1C61-R-0002 Remote Shutdown Panel Control Check 117
71111.15 Procedures 13561-01-H-005 28" Main Steam Isolation Valve 1
71111.15 Procedures EN-DC-115 Engineering Change Process 27
71111.15 Procedures MC-Q1P75-90194 Lube Oil Requirements for the Division I and II Diesel 1
Generators
71111.15 Procedures Standard/Specification Purchase Specification for Main Steam Isolation Valves 5
Change Notice
00/0001 21A9506
71111.15 Work Orders WO 00537942
71111.18 Engineering EC-86810 Equivalent 1P45C001A Drwl Flr Drn Sump Pump 0
Changes
71111.19 Corrective Action CR-GGN- 2020-04810, 2020-05009, 2020-05109, 2020-05129, 2020-
Documents 05154, 2020-05234, 2020-05291, 2020-05351, 2020-
05361, 2020-05447, 2020-05905
71111.19 Engineering EC-72779 Turbine Control Protection System - Safety 0
Changes
71111.19 Engineering EC-72780 Turbine Control Protection System - Non-Safety 0
Changes
71111.19 Engineering EC-76310 Turbine RPS Instrument Setpoint and Uncertainty 0
Changes Evaluations
71111.19 Engineering EC-87061 Modification to the Speed Probe Bracket for TCU Upgrade 0
Changes
71111.19 Procedures 04-1-05-B21-1 Local Leak Rate Alignment Instructions Main Steam and 2
Drywell Pressure Transmitters Penetrations
71111.19 Procedures 06-OP-1M61-V-0002 Local Leak Rate Test - AIR (Using Graftel Model 9623-7 13
Leak Rate Monitor)
19
Inspection Type Designation Description or Title Revision or
Procedure Date
71111.19 Work Orders WO 540076, 541874, 542396, 544968, 52838593, 52859292
71111.22 Corrective Action CR-GGN- 2018-04934, 2020-05246, 2020-05268, 2020-05270, 2020-
Documents 05276, 2020-05404, 2020-05405, 2020-05413, 2020-
05427, 2020-05438, 2020-05640, 2020-05712, 2020-
06046, 2020-06070, 2020-06217, 2020-06552
71111.22 Drawings M-1077B P&I Diagram Nuclear Boiler System 35
71111.22 Miscellaneous DCP 87/3001 ATWS ARI/RPT Fabrication/Installation Package 0
71111.22 Procedures 06-IC-1B21-R-0015 ATWS Reactor Vessel Level Calibration 103
71111.22 Procedures 06-OP-1B21-V-0001 MSIV Operability Test 122
71111.22 Procedures 06-OP-1E51-C-0005 RCIC Pump Low Pressure Flow Verification Test 111
71111.22 Procedures 06-OP-1E51-Q-0003 RCIC System Quarterly Pump Operability Verification 145
71111.22 Procedures 06-OP-1T48-R-0003 Standby Gas Treatment B Logic and Vacuum Test 122
71111.22 Procedures JC-Q1B21-N682-1 Level 2 Setpoint Calculation 1
71111.22 Work Orders WO 542396-06, 52838593, 52840019-01, 52853178, 52882122
71114.04 Corrective Action CR-GGN- 2020-00631
Documents
71114.04 Corrective Action CR-GGN- 2020-05947, 2020-06708, 2020-06914, 2020-07514
Documents
Resulting from
Inspection
71114.04 Corrective Action Work Tracking (WT- 2020-00182
Documents GGN-)
Resulting from
Inspection
71114.04 Miscellaneous 10 CFR 50.54(q)(3) Screening; Procedure/Document 02/19/2020
Number: 10-S-01-1, Revision: 132
71114.04 Miscellaneous 10 CFR 50.54(q)(3) Screening; Procedure/Document 01/22/2020
Number: Emergency Plan, Revision: 80
71114.04 Miscellaneous 10 CFR 50.54(q)(3) Evaluation: Procedure/Document 01/22/2020
Number: Emergency Plan, Revision: 80
71114.04 Miscellaneous 10 CFR 50.54(q)(3) Screening; Procedure/Document 02/18/2020
Number: Emergency Plan, Revision: 80
71114.04 Miscellaneous 10 CFR 50.54(q)(3) Evaluation; Procedure/Document 02/19/2020
Number: 10-S-01-1, Revision: 132
20
Inspection Type Designation Description or Title Revision or
Procedure Date
71114.04 Miscellaneous 10 CFR 50.54(q)(3) Screening; Procedure/Document 01/22/2020
Number: Activation of the Emergency Plan/10-S-01-1,
Revision: 131
71114.04 Miscellaneous GLP-OPS-E2201 High Pressure Core Spray System (HPCS) - E22-1 13
71114.04 Miscellaneous GNRO-2018/00008 License Amendment Request, Adoption of Emergency 04/27/2018
Action Level Schemes Pursuant to NEI 99-01, Revision 6;
Grand Gulf Nuclear Station (GGNS), Docket No. 50-416;
License No. NPF-29
71114.04 Miscellaneous GNRO-2018/00048 Response to Request for Additional Information, License 10/10/2018
Amendment Request; Adoption of Emergency Action Level
Scheme Pursuant to NEI 99-01, Revision 6; Grand Gulf
Nuclear Station, Unit 1; Docket No. 50-416; License No.
71114.04 Miscellaneous GNRO-2020/00011 Emergency Plan, Revision 80; Grand Gulf Nuclear Station, 03/11/2020
Unit 1; NRC Docket No. 50-416; Renewed Facility
Operating License No. NPF-29
71114.04 Procedures 01-S-02-9 Procedure Change Process 5
71114.04 Procedures 01-S-10-3 Emergency Planning Department Responsibilities 22
71114.04 Procedures 02-S-01-40 EP Technical Bases 8
71114.04 Procedures 04-1-02-1H13-P601 Alarm Response Instruction, Panel No.: 1H13-P601 170
71114.04 Procedures 05-S-01-EP-4 Emergency Procedure, Auxiliary Building Control 29
71114.04 Procedures 10-S-01-1 Activation of the Emergency Plan 131, 132
71114.04 Procedures 10-S-01-12 Radiological Assessment and Protective Action 47
Recommendations
71151 Corrective Action CR-GGN- 2019-03822, 2019-07766, 2020-07879
Documents
71151 Engineering EC-82963 Evaluate Loss of HVAC in the Drywell for May 12, 2019 0
Changes SCRAM
71151 Procedures 04-1-01-T48-1 Standby Gas Treatment 40
71151 Procedures 06-ME-1000-R-0007 Charcoal Adsorber Chemical Analysis 106
71151 Procedures 06-OP-1P75-M-0001 Standby Diesel Generator 11 Functional Test 147
71151 Procedures 06-OP-1T48-Q-0002 Standby Gas Treatment System A Valve Test 111
71151 Procedures EN-LI-102 Corrective Action Program 40
71151 Procedures EN-LI-108 Event Notification and Reporting 18
21
Inspection Type Designation Description or Title Revision or
Procedure Date
71151 Work Orders WO 52887447
71152 Corrective Action CR-GGN- 2018-04934, 2020-03778, 2020-05225, 2020-05246, 2020-
Documents 05276, 2020-05413, 2020-05438, 2020-05640, 2020-
05697, 2020-05712, 2020-05763, 2020-06217
71152 Procedures 06-IC-1B21-R-0015 ATWS Reactor Vessel Level Calibration 103
71152 Work Orders WO 543966, 544219, 544220, 544221, 544222, 52840019-01
22