IR 05000373/2018004
| ML19214A183 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 07/31/2019 |
| From: | Kenneth Riemer NRC/RGN-III/DRP/B1 |
| To: | Bryan Hanson Exelon Generation Co, Exelon Nuclear |
| References | |
| IR 2018004 | |
| Download: ML19214A183 (25) | |
Text
First initial L UNITED STATES NUCLEAR REGULATORY COMMISSION uly 31, 2019
SUBJECT:
ERRATALASALLE COUNTY STATION, UNITS 1 AND 2NRC INTEGRATED INSPECTION REPORT 05000373/2018004 AND 05000374/2018004
Dear Mr. Hanson:
On February 14, 2019, the NRC issued NRC Integrated Inspection Report 05000373/2018004 and 2018004 (ADAMS Accession Number ML19045A403). An internal NRC review concluded that the 10 CFR Part 50, Appendix B, Criterion V violation associated with the Technical Specification Surveillance Procedure for drywall oxygen concentration was more appropriately characterized as a violation of Technical Specification 5.4.1.a and the applicable procedures recommended in NRC Regulatory Guide 1.33, Revision 2, Appendix A, February 1978.
Accordingly, this report is being reissued in its entirety to more appropriately characterize the violation. The significance of the violation and the NRCs understanding of your corrective actions remains unchanged. Therefore, no action is required on your part unless you contest the new characterization of the finding and associated violation.
On December 31, 2018, the U.S. Nuclear Regulatory Commission (NRC) completed an integrated inspection at your LaSalle County Station, Units 1 and 2. On January 9, 2019, the NRC inspectors discussed the results of this inspection with Mr. W. Trafton, and other members of your staff. The results of this inspection are documented in the enclosed report.
Based on the results of this inspection, the NRC has identified two issues that were evaluated under the risk significance determination process as having very-low safety significance (Green). The NRC has also determined that two violations are associated with these issues.
Because the licensee initiated condition reports to address these issues, these violations are being treated as Non-Cited Violations (NCVs), consistent with Section 2.3.2 of the Enforcement Policy. These NCVs are described in the subject inspection report. Further, inspectors documented a licensee-identified violation which was determined to be of very low safety significance. The NRC is treating this violation as a non-cited violation consistent with Section 2.3.2.a of the Enforcement Policy.
If you contest the violations or significance of these NCVs, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region III; the Director, Office of Enforcement; and the NRC Resident Inspector at the LaSalle County Station. If you disagree with a cross-cutting aspect assignment or a finding not associated with a regulatory requirement in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region III; and the NRC resident inspector at the LaSalle County Station.
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 10 CFR 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely,
/RA/
Kenneth Riemer, Chief Branch 1 Division of Reactor Projects Docket Nos. 50-373; 50-374;72-070 License Nos. NPF-11; NPF-18 Enclosure:
IR 05000373/2018004; 05000374/2018004 cc: Distribution via LISTSERV
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring licensees performance by conducting an integrated quarterly inspection at LaSalle County Station, Units 1 and 2, 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. Findings and violations being considered in the NRCs assessment are summarized in the table below. A Licensee identified Non-Cited violation is documented in report section: 71153 List of Findings and Violations Failure to Implement Scaffolding Program Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green [H.13] 71111.05 Systems NCV 05000373/2018004-01 Consistent Opened/Closed Process The inspectors identified a finding of very low safety significance (Green) and an associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (CFR) Part 50,
Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to follow station procedure MA-AA-716-025, Scaffold Installation, Modification, and Removal Request Process, Revision 15. Specifically, the licensee erected two scaffolds that were in close proximity to safety-related equipment without engineering approval for less than minimum clearances. Additionally, these scaffolds were not being tracked administratively, and as a result were installed in the plant for greater than 90 days with neither a 10 CFR 50.59 review nor engineering approval to make the installation a permanent scaffold.
Control of Instrument Used in an Activity Affecting Quality Not Appropriate to the Circumstance Cornerstone Significance Cross-Cutting Report Aspect Section Barrier Integrity Green None 71152 NCV 05000373/2018004-02 Opened/Closed The inspectors identified a finding of very low safety significance (Green) and an associated NCV of Technical Specification 5.4.1.a Instructions, Procedures, and Drawings, for the licensees failure to establish procedures of a type appropriate to the circumstances that ensured the drywell oxygen monitor was properly controlled, calibrated, and adjusted at specified periods to maintain accuracy. Specifically, procedures for calibration and operation of the continuous oxygen monitor did not calibrate the instrument over the expected range of use or account for all uncertainties or instrument drift during operation to maintain accuracy of the instrument.
Additional Tracking Items Type Issue Number Title Report Status Section LER 05000373/2018-003-00 Two Main Steam Safety 71153 Closed and 05000373/2018- Relief Valves Failed 003-01 Inservice Lift Inspection Pressure Test
TABLE OF CONTENTS
PLANT STATUS
INSPECTION SCOPES
................................................................................................................
REACTOR SAFETY
.....................................................................................................................
RADIATION SAFETY
...................................................................................................................
OTHER ACTIVITIES - BASELINE
..............................................................................................
INSPECTION RESULTS
............................................................................................................
EXIT MEETINGS AND DEBRIEFS
............................................................................................ 17
DOCUMENTS REVIEWED
......................................................................................................... 17
PLANT STATUS
Unit 1 began the inspection period at rated thermal power. On October 17, 2018, the unit was
down powered to approximately 92 percent power due to a feedwater heater (15A) emergency
drain valve failure and power excursion. The affected drain valve was repaired and the unit was
returned to full power on October 18, 2018. On December 14, 2018, the unit was down
powered to approximately 77 percent for a rod pattern adjustment and to complete turbine valve
testing. The unit was returned to full power the following day. The unit remained at or near
rated thermal power for the remainder of the inspection period.
Unit 2 began the inspection period at rated thermal power. On November 2, 2018, the unit
was down powered to approximately 80 percent power to perform a rod sequence exchange
and control rod testing. The unit was returned to full power the following day. On
November 20, 2018, the unit was down powered to approximately 85 percent power to
perform a rod sequence exchange, control rod testing and turbine valve testing. On
December 20, 2018, the unit was down powered to approximately 85 percent power to perform
a rod sequence exchange, control rod testing and turbine valve testing. The unit was returned
to full power the same day. The unit remained at or near rated thermal power for the remainder
of the inspection period.
INSPECTION SCOPES
Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in
effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with
their attached revision histories are located on the public website at http://www.nrc.gov/reading-
rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared
complete when the IP requirements most appropriate to the inspection activity were met
consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection
Program - Operations Phase. The inspectors performed plant status activities described in
IMC 2515 Appendix D, Plant Status and conducted routine reviews using IP 71152, Problem
Identification and Resolution. 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
71111.01Adverse Weather Protection
Seasonal Extreme Weather (1 Sample)
The inspectors evaluated readiness for seasonal extreme weather conditions prior to the
onset of seasonal cold temperatures on November 14, 2018.
External Flooding (1 Sample)
The inspectors evaluated readiness to cope with external flooding on October 23, 2018.
71111.04Equipment Alignment
Partial Walkdown (2 Samples)
The inspectors evaluated system configurations during partial walkdowns of the following
systems/trains:
(1) Unit 2 standby gas treatment system during maintenance on the Unit 1 standby gas
treatment system on October 1, 2018; and
(2) Unit 2 B low pressure coolant injection mode of residual heat removal (RHR) on
October 18, 2018.
Complete Walkdown (1 Sample)
The inspectors evaluated system configurations during a complete walkdown of the Unit 1,
Division II electrical distribution system during Division I battery replacement on
November 14, 2018.
71111.05AQFire Protection Annual/Quarterly
Quarterly Inspection (4 Samples)
The inspectors evaluated fire protection program implementation in the following selected
areas:
(1) Fire zone 3H3, Unit 2 B RHR, 694 elevation;
(2) Fire zone 3I3, Unit 2 B RHR pump, 673 elevation;
(3) Fire zone 3I4, Unit 2 low pressure core spray system, 6734 elevation; and
(4) Fire zone 3H4, Reactor Core Isolation Cooling (RCIC) pump cubicle, 694 elevation.
Annual Inspection (1 Sample)
The inspectors evaluated fire brigade performance on Unit 1, condensate pump aisle, on
November 1, 2018.
71111.06Flood Protection Measures
Internal Flooding (2 Samples)
The inspectors evaluated internal flooding mitigation protections in the Unit 2, Division II
corner room and Unit 2 raceway, both at 673 elevation, on October 23, 2018.
71111.07Heat Sink Performance
Heat Sink (1 Sample)
The inspectors evaluated the Unit 1, Division III emergency diesel generator cooling
performance on November 20, 2018.
71111.11Licensed Operator Requalification Program and Licensed Operator Performance
Operator Requalification (1 Sample)
The inspectors observed and evaluated the out-of-the-box (OBE) ESG-80 on
October 10, 2018.
Operator Performance (1 Sample)
The inspectors observed and evaluated operators in the control room during the Unit 2 down
power to support control valve testing, scram time testing and post maintenance testing on
motor-driven reactor feed pump on December 8, 2018.
Operator Exams (1 Sample)
The inspectors reviewed and evaluated requalification examination results on
December 18, 2018.
71111.12Maintenance Effectiveness
Routine Maintenance Effectiveness (3 Samples)
The inspectors evaluated the effectiveness of routine maintenance activities associated
with the following equipment and/or safety significant functions:
(1) Reactor building flood seals on October 23, 2018;
(2) Licensee a(3) evaluation; and
(3) Reactor building ventilation check dampers.
71111.13Maintenance Risk Assessments and Emergent Work Control (3 Samples)
The inspectors evaluated the risk assessments for the following planned and emergent
work activities:
(1) Mobile crane exterior to Unit 1 and Unit 2 reactor buildings;
(2) Unit 1 and Unit 2 online risk yellow due to 0 emergency diesel generator maintenance;
(3) Unit 1, Division I and II protected equipment during yellow online risk for Division III
maintenance; and
(4) Unit 1 and Unit 2 yellow online risk during blizzard.
71111.15Operability Determinations and Functionality Assessments (5 Samples)
The inspectors evaluated the following operability determinations and functionality
assessments:
(1) Unit 1 A standby liquid control pump gearbox vibration in Alert range;
(2) Unit 1 drywell high temperature annunciator during B containment chiller maintenance
window;
(3) Unit 1, Division I 125 volts direct current (VDC) battery online replacement;
(4) Secondary containment operability during forced helium dehydration modification
installation; and
(5) Unit 2 RCIC turbine inboard bearing oil level low.
71111.18Plant Modifications (1 Sample)
The inspectors evaluated the following temporary or permanent modification:
(1) Category II over I qualification for abandoned high pressure core spray suction line after
a temporary modification on November 20, 2018.
71111.19Post Maintenance Testing (4 Samples)
The inspectors evaluated the following post maintenance tests:
(1) Unit 2 diesel generator on November 15, 2018;
(2) Unit 1 standby gas treatment system on October 18, 2018;
(3) Unit 2 B RHR heat exchanger outlet valve breaker replacement on October 16, 2018;
and
(4) Unit 1, A control room heating, ventilation and cooling, D radiation monitor
post-maintenance testing on October 29, 2018.
71111.22Surveillance Testing
The inspectors evaluated the following surveillance tests:
Routine (1 Sample)
(1) Diesel fire pump 0A (0FP01KA) operational check on December 18, 2018.
In-Service (1 Sample)
(1) Unit 2 B RHR system operability and inservice test on December 19, 2018.
71114.04Emergency Action Level and Emergency Plan Changes (1 Sample)
The inspector completed the evaluation of submitted Emergency Action Level and
Emergency Plan changes on November 30, 2018. This evaluation does not constitute
NRC approval.
RADIATION SAFETY
71124.02Occupational As Low As Reasonably Achievable Planning and Controls
Radiological Work Planning (1 Sample)
The inspectors evaluated the licensees radiological work planning by reviewing the
following activities:
(1) LA-02-17-00502; L2R16 drywell RP Department Activities;
(2) LA-02-17-00506; L2R16 drywell Scaffold;
(3) LA-02-17-00513; L2R16 drywell Control Rod Drive (CRD) Exchange;
(4) LA-02-17-00547; L2R16 drywell RR Motor Replacement; and
(5) LA-01-18-00510; L1R17 drywell Steam Safety Relief Valve Activities.
Verification of Dose Estimates and Exposure Tracking Systems (1 Sample)
The inspectors evaluated dose estimates and exposure tracking.
Implementation of As Low As Reasonably Achievable and Radiological Work Controls
(Partial Sample)
The inspectors reviewed as low as reasonable achievable practices and radiological work
controls by reviewing the following activities:
(1) LA-02-17-00502; L2R16 drywell RP Department Activities;
(2) LA-02-17-00506; L2R16 drywell Scaffold;
(3) LA-02-17-00513; L2R16 drywell Control Rod Drive (CRD) Exchange;
(4) LA-02-17-00547; L2R16 drywell RR Motor Replacement; and
(5) LA-01-18-00510; L1R17 drywell Steam Safety Relief Valve Activities.
OTHER ACTIVITIES - BASELINE
71151Performance Indicator Verification (7 Samples)
The inspectors verified licensee performance indicators submittals listed below:
(1) MS05: Safety System Functional Failures2 Samples; October 1, 2017 -
September 31, 2018;
(2) MS08: Heat Removal Systems2 Samples; October 1, 2017 - September 31, 2018;
(3) MS10: Cooling Water Support Systems2 Samples; October 1, 2017 -
September 31, 2018; and
(4) OR01: Occupational Exposure Control Effectiveness1 Sample; October 2017 -
September 2018.
71152Problem Identification and Resolution
Semiannual Trend Review (1 Sample)
The inspectors reviewed the licensees corrective action program for trends that might be
indicative of a more significant safety issue.
Annual Follow-Up of Selected Issues (1 Samples)
The inspectors reviewed the licensees implementation of its corrective action program
related to the following issues:
(1) Drywell continuous oxygen monitor.
71153Follow-Up of Events and Notices of Enforcement Discretion
Licensee Event Reports (1 Sample)
The inspectors evaluated the following licensee event reports which can be accessed at
https://lersearch.inl.gov/LERSearchCriteria.aspx:
(1) Licensee Event Reports 05000373/2018-003-00 and 05000373/2018-003-01,
Two Main Safety Relief Valves Failed Inservice Lift Inspection Pressure Test, on
July 25, 2018.
INSPECTION RESULTS
71111.05AQFire Protection Annual/Quarterly
Failure to Adhere to Scaffolding Program
Cornerstone Significance Cross-Cutting Report Section
Aspect
Mitigating Systems Green [H.13] - 71111.05
NCV 05000373/2018004-01 Consistent
Open/Closed Process
Introduction:
The inspectors identified a finding of very low safety significance (Green) and an associated
NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings,
for the licensees failure follow station procedure MA-AA-716-025, Scaffold Installation,
Modification, and Removal Request Process, Revision 15. Specifically, the licensee erected
two scaffolds that were in close proximity to safety-related equipment without engineering
approval for less than minimum clearances. Additionally, these scaffolds were not being
tracked administratively, and as a result were installed in the plant for greater than 90 days
with neither a 10 CFR 50.59 review nor engineering approval to make the installation a
permanent scaffold.
Description:
On October 17, 2018, during a fire protection walk down, NRC inspectors identified a scaffold
in contact with safety-related piping on the 687 feet elevation in the Unit 2, Division II core
standby cooling system (CSCS) pump room. The inspectors noted that there was no
engineering approval of the scaffold installation as required by Step 4.1.5 of MA-AA-716-025
for scaffolds in close proximity of safety-related equipment. The inspectors discussed the
issue with the licensee, as well as questioned the acceptability of a scaffold contacting safety-
related equipment. The inspectors concerns were documented in the corrective action
program by the licensee as AR 04185035.
On October 18, 2018, NRC inspectors identified a scaffold in contact with safety-related
piping on the 694 feet elevation in the Unit 2, Division II RHR corner room. Again, the
inspectors noted there was no engineering approval of the scaffold installation as required by
Step 4.1.5 of MA-AA-716-025. Additionally, the inspectors reviewed the Non-Permanent
Scaffold Request Form posted on the scaffold and noted that Section C, Pre-Erection
Operations Review, of the form documented that Operations did their jobsite review via
teleconference.
Procedure MA-AA-716-025, Steps 4.1.4.2 and 4.1.2.3, required operations staff to identify
bump sensitive equipment in the area of the scaffold and to determine if an operations
inspection of the installation was required (i.e. whether the scaffold was installed near safety-
related equipment). The inspectors questioned the ability of operations department personnel
to make that type of determination without physically walking down the area of the scaffold
installation. The inspectors discussed their concerns with the licensee, who documented
them in the corrective action program as AR 04185256.
In response to the inspectors questions, the licensee evaluated the scaffolds for the potential
impact on the equipment during a postulated seismic event. In each case it was determined
that there would be no adverse impact. To determine whether engineering approval was
required by step 4.1.5 of MA-AA-716-025, the licensee referred to Table 2 of engineering
standard NES-MS-04.1, Seismic Prequalified Scaffolds, Revision 7. Note 3 of this
standard states in part, that scaffolds to be installed within 4 inches of safety-related
equipment in the Auxiliary Building below 815 feet elevation and within 3.5 inches of
safety-related equipment in the Reactor Building below 843 feet elevation be approved by
engineering. Therefore, engineering approval was required for the scaffolds installed in the
Division II CSCS pump room and RHR corner room.
The inspectors also discussed the administrative tracking requirements in Step 3.6 of
MA-AA-716-025 with the licensee. Step 3.6 of MA-AA-716-025 required the following:
Scaffold Coordinator/Designee- Is responsible for the coordination of erection and
removal of all scaffolds on site. Maintaining a log or electronic equivalent of the status
of all scaffolds, and reviewing the log to ensure that any scaffolds approaching their
day limit are removed or converted to a permanent scaffold or requesting that an
individual 10 CFR 50.59 review be performed for the individual scaffold required to be
left in place beyond 90 days.
The inspectors noted that since the previously discussed scaffolds were installed in March
and April of 2017, greater than 90 days, that Step 3.6 would require that the licensee either
perform a 10 CFR 50.59 review of the temporary installed scaffolding or get engineering
approval to convert the temporary installed scaffolding to permanent scaffolding. The
licensee acknowledged that they did not perform either of these actions, contrary to Step 3.6
of MA-AA-716-025. Further, the licensee determined that the previously discussed scaffolds
were not being tracked in the scaffolding log, contrary to Step 3.6 of MA-AA-716-025.
Corrective Actions: The licensee removed the scaffolds in the Division II CSCS pump room
and RHR corner room to comply with MA-AA-716-025. Additionally, the licensee completed
a Corrective Action Program Evaluation. During this review the licensee discovered two
additional scaffolds in close proximity to safety-related equipment without the required
engineering approval. The licensee documented the issues in the corrective action program
as AR 04186864 and 04186868. The licensee evaluated the scaffolds for the potential impact
on the equipment during a postulated seismic event and determined there would be no impact
on equipment operability.
Corrective Action References: ARs 04185035, 04185256, 04186864, and 04186868
Performance Assessment:
Performance Deficiency: The inspectors identified that multiple examples of the failure to
follow procedure, MA-AA-716-025, as related to control of temporary scaffolding was
contrary to 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings,
and was a performance deficiency. Specifically, the licensee built scaffolds in close proximity
to safety-related equipment without engineering approval; and did not follow administrative
requirements to ensure temporary scaffolds installed in the plant greater than 90 days were
either reviewed under 10 CFR 50.59 or made a permanent scaffold.
Screening: The inspectors determined the performance deficiency was more than minor in
accordance with IMC 0612, Appendix E, Examples of Minor Issues. Specifically, the
inspectors concluded that this issue was similar to the more than minor criteria established in
Example 4.a, Insignificant Procedural Issues, since the licensee routinely failed to perform
the required engineering evaluations on seismically qualified scaffolds. Therefore, this
performance deficiency also impacted the Mitigating Systems Cornerstone objective of
protection against external events (i.e. seismic events).
Significance: The inspectors assessed the significance of the finding using SDP Appendix A,
The Significance Determination Process for Findings At-Power. The finding screened as
very low safety significance because it did not result in the loss of operability or functionality of
a Mitigating System.
Cross-Cutting Aspect: The inspectors determined this finding affected the cross-cutting area
of human performance in the aspect of consistent process, where Individuals use a
consistent, systematic approach to make decisions. Risk insights are incorporated as
appropriate. The inspectors found several scaffolds in the plant that incorporated all of the
necessary elements of MA-AA-716-025 related to control of temporary scaffolding.
However the scaffolds installed in the Division II CSCS pump room and RHR corner room did
not. [H.13]
Enforcement:
Violation: Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and
Drawings, requires, in part, that activities affecting quality be prescribed by instructions,
procedures, or drawings, of a type appropriate to the circumstance and shall be accomplished
in accordance with these instructions, procedures, or drawings.
Step 3.6 of procedure MA-AA-716-025, Scaffold Installation, Modification, and Removal
Request Process, Revision 15, required:
Scaffold Coordinator/Designee- Is responsible for the coordination of erection and
removal of all scaffolds on site. Maintaining a log or electronic equivalent of the status
of all scaffolds, and reviewing the log to ensure that any scaffolds approaching their
day limit are removed or converted to a permanent scaffold or requesting that an
individual 10 CFR 50.59 review be performed for the individual scaffold required to be
left in place beyond 90 days.
Step 4.1.5 of procedure MA-AA-716-025, Scaffold Installation, Modification, and Removal
Request Process, Revision 15, required:
Engineering APPROVE post erection inspections required by Engineering.
Step 4.1.5 of procedure MA-AA-716-025 is implemented by Table 2, Note 3, of engineering
standard NES-MS-04.1, Seismic Prequalified Scaffolds, Revision 7. NES-MS-04.01, Table
2, Note 3, required:
Movement of in-place systems/components are not included in the above clearances
and should be increased accordingly. For Byron/Braidwood/Clinton, a 3 clearance
shall be provided to account for movement of in-place systems/components, unless
otherwise approved by engineering. This clearance is measured from anywhere on
the outside surface of the in-place item to the closest point of the scaffolding. For the
other stations, Engineering shall be contacted to provide clearance requirements, as
required.
Contrary to the above:
From April 18, 2017, until October 25, 2018, scaffold 1867810 was installed in close proximity
to safety-related piping in on the 687 feet elevation, and from March 21, 2017, until October
25, 2018, scaffold 1929990 was installed in close proximity to safety-related piping in on the
694 feet elevation in the Unit 2, Division II core standby cooling system (CSCS) pump room,
for a period in excess of 90 days, without engineering approval, was not recorded in the
scaffolding log, and was not removed or converted to a permanent scaffold or requested that
an individual 10 CFR 50.59 review be performed.
From March 21, 2017, until October 25, 2018, scaffold 1929990 was installed in close
proximity to safety-related piping in on the 694 feet elevation in the Unit 2, Division II RHR
corner room, for a period in excess of 90 days, without engineering approval, was not
recorded in the scaffolding log, and was not removed or converted to a permanent scaffold or
requested that an individual 10 CFR 50.59 review be performed.
Disposition: This violation is being treated as an NCV, consistent with Section 2.3.2 of the
71152Problem Identification and Resolution
Control of instrument used in an activity affecting quality not appropriate to the circumstance
Cornerstone Significance Cross-Cutting Report Section
Aspect
Barrier Integrity Green None 71152
Open/Closed
Introduction:
The inspectors identified a finding of very low safety significance (Green) and an associated
NCV of Technical Specification 5.4.1.a Instructions, Procedures, and Drawings, for the
licensees failure to establish procedures of a type appropriate to the circumstances that
ensured the drywell oxygen monitor was properly controlled, calibrated, and adjusted at
specified periods to maintain accuracy. Specifically, procedures for calibration and operation
of the continuous oxygen monitor did not calibrate the instrument over the expected range of
use or account for all uncertainties or instrument drift during operation to maintain accuracy of
the instrument.
Description:
Technical Specification (TS) 3.6.3.2, Primary Containment Oxygen Concentration, requires,
in part, that primary containment oxygen concentration remain less than 4 percent oxygen by
volume during operation in Mode 1. In order to ensure that an event that produces any
amount of hydrogen (e.g. a loss of coolant accident) does not result in a combustible mixture
inside primary containment, which affects the safety-related function of primary containment,
TS surveillance requirement (SR) 3.6.3.2.1 requires the licensee to verify primary
containment hydrogen remain within the specified limits on a weekly basis. The licensee
implemented this surveillance requirement with procedure LOS-AA-W1, Technical
Specifications Weekly Surveillances, Revision 83. This procedure contained instructions to
re-inert the drywell at 3.5 percent oxygen to prevent exceeding the 4 percent oxygen limit of
On May 29, 2018, the licensee documented in AR 04141949 that the Unit 1 drywell
continuous oxygen monitor (1PL78J) had failed. This was determined while performing a
monthly channel check of the post-LOCA containment monitoring system, which found
1PL78J indicating 1 percent oxygen and the post-LOCA oxygen detection system indicating
4.2 percent oxygen in the drywell. In response to the high oxygen condition in the drywell, the
licensee entered TS 3.6.3.2 and re-inerted the drywell to less than the TS oxygen limit.
Additionally, the licensee replaced the 1PL78J oxygen detector. The licensee determined
that the issue was not reportable under 10 CFR 50.73(a)(2)(i)(B) for a condition prohibited by
TS.
The drywell continuous oxygen monitor is the instrument the licensee takes credit for in
procedure LOS-AA-W1 to accomplish the requirements of SR 3.6.3.2.1. The inspectors
reviewed the calibration data for 1PL78J and noted that the instrument calibration procedure,
LIP-CM-510, Unit 1 Primary Containment Continuous Oxygen Monitor Sensor Maintenance
and Standardization, Revision 7, did not include provisions for control and calibration of the
monitor to maintain its accuracy. Specifically, they noted that it was not calibrated, within its
range of use, to a reference standard, which is a requirement for instruments used for
activities affecting quality such as TS surveillances. It was also noted that the instrument loop
uncertainty was not quantified and accounted for in the procedural acceptance criteria used to
meet SR 3.6.3.2.1, like it is for other TS surveillance-related instruments.
Upon review of corrective action program documentation, the inspectors found several
instances of significant instrument drift for 1PL78J. This was revealed by abnormally low
oxygen readings upon de-inerting the drywell for refueling outages. In some instances,
drywell oxygen read as low as 14 percent when normal atmospheric conditions,
approximately 20.9 percent oxygen, were expected in the drywell. The inspectors discussed
their concerns with the licensee. Following a discussion with the instrument vendor, the
licensee determined that the oxygen monitor drift was likely due to the instrumentation
remaining energized and that this practice led to the detector reading erratically after 75-120
days. The licensee documented this issue in the corrective action program as AR 4203425.
The inspectors determined that since the calibration periodicity for 1PL78J was 120 days, and
the vendor concluded that the detector could behave erratically after being energized
continuously for 75 days, the practice of leaving the detector energized when not in use was
not appropriate to the circumstances. Specifically, this practice could lead to an erroneous
reading during the weekly TS surveillance test, which is a condition adverse to quality. The
licensee documented the inspectors concern in the corrective action program as AR
209125.
Corrective Actions: The licensee will revise the operating and calibration procedures for
1PL78J to de-energize the instrument when not in use to preclude instrument drift. The
licensee was also evaluating the use of a 4 percent oxygen calibration gas during the
calibration procedure and the need to determine 1PL78J instrument loop uncertainty.
Corrective Action Reference: AR 4203425 and 4209125
Performance Assessment:
Performance Deficiency: The inspectors determined that the licensees failure to establish
procedures of a type appropriate to the circumstances that ensured the drywell oxygen
monitor was properly controlled, calibrated, and adjusted at specified periods to maintain
accuracy was a performance deficiency. Specifically, procedures for calibration and
operation of the continuous oxygen monitor did not calibrate the instrument over the expected
range of use or account for all uncertainties or instrument drift during operation to maintain
accuracy of the instrument.
Screening: The inspectors determined the performance deficiency was more than minor
because it adversely affected the procedure quality attribute of the barrier integrity
cornerstone objective to provide reasonable assurance that physical design barriers (fuel
cladding, reactor coolant system, and containment) protect the public from radionuclide
releases caused by accidents or events. Specifically, instrument uncertainty due to drift
caused primary containment oxygen levels to exceed levels permitted by T
- S.
Significance: The inspectors assessed the significance of the finding using SDP Appendix A,
The Significance Determination Process (SDP) for Findings At-Power. The inspectors
determined that the finding was of very low safety significance (Green) because it did not
represent an actual open pathway in the physical integrity of the reactor containment,
containment isolation system, or heat removal components.
Cross-cutting Aspect: The inspectors determined that there was no cross-cutting issue
because the issue was not indicative of current plant performance.
Enforcement:
Violation: Technical Specification Section 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.
NRC Regulatory Guide 1.33, Revision 2, Appendix A, Section 8, provides recommendations
for Procedures for Control of Measuring and Test Equipment and for Surveillance Tests,
Procedures, and Calibrations. Part a of Section 8 says procedures of a type appropriate to
the circumstances shall be provided to ensure that tools, gauges, instruments, controls, and
other measuring and testing devices are properly controlled, calibrated, and adjusted at
specified periods to maintain accuracy.
Contrary to the above, on February 4, 2015, the licensee failed to establish procedures of a
type appropriate to the circumstances to ensure that the Unit 1 drywell continuous oxygen
monitor was properly controlled, calibrated, and adjusted at specified periods to maintain
instrument accuracy. Specifically, procedures for the calibration and operation of the
continuous oxygen monitor failed to include provisions to ensure the instrument was properly
controlled and calibrated to maintain accuracy.
Disposition: This violation is being treated as an NCV, consistent with Section 2.3.2 of the
71153Follow-Up of Events and Notices of Enforcement Discretion
Licensee Identified Non-Cited Violation 71153Follow-Up of Events and Notices of
This violation of very low safety significant was identified by the licensee and has been
entered into the licensee corrective action program and is being treated as an NCV,
consistent with Section 2.3.2 of the Enforcement Policy.
Violation: Technical Specification 3.4.4 limited condition for operation (LCO) is (applicable for
Modes 1, 2 and 3) states: The safety function of 12 safety relief valves (S/RVs) shall be
OPERABLE; and action statement A states One or more required S/RVs inoperable - A.1 be
in mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and A.2 be in Mode 4 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. In addition, TS SR 3.4.4.1 stated
Verify the safety function lift setpoints of the required S/RVs are as follows:
Number of S/RVs Setpoint (psig)
1205 +/- 36.1
1195 +/- 35.8
1185 +/- 35.5
1175 +/- 35.2
1150 +/- 34.5
Contrary to the above, during a portion of the previous two Unit 1 operating cycles from
February 7, 2014 through February of 2018, two main steam S/RVs did not meet these lift
pressure setpoint requirements. Specifically S/RV 1B21 - F013R lifted at 1167 psig instead
of 1168.9 - 1241.1 psig and S/RV 1B21 - F013U lifted at 1109 psig instead of 1115.5 -
1184.5 psig (reference; Licensee Event Report 05000373/2018-003-01, Two Main Safety
Relief Valves Failed Inservice Lift Inspection Pressure Test). The licensee replaced the two
affected valves and submitted a license amendment request on February 27, 2018, to revise
TS 3.4.4.1 and lower the setpoint tolerance (minus five percent) to account for S/RV minor
setpoint drift in the conservative direction.
Significance/Severity: This licensee identified finding affected the Initiating Events
Cornerstone and was screened in accordance with IMC 0609.04, Initial Characterization of
Findings, and Exhibit 1 of IMC 0609, Appendix A, The Significance Determination Process
for Findings At Power. The two affected SRVs lifted low outside of their setpoint band, which
was conservative with respect to maintaining the reactor coolant system overpressure
protection safety function of these valves. Therefore, the inspectors determined that this
finding is of very low safety significance (Green) because after a reasonable assessment of
degradation, the finding would not have resulted in exceeding the reactor coolant system leak
rate for a small LOCA and did not affect other systems used to mitigate a loss-of-coolant
accident.
Corrective Action References: AR 04110929 and AR 04110933
EXIT MEETINGS AND DEBRIEFS
The inspectors confirmed that proprietary information was controlled to protect from public
disclosure. No proprietary information was documented in this report.
- On November 9, 2018, the inspector presented the radiation protection program inspection
results to Mr.
- J. Washko, Plant Manager, and other members of the licensee staff.
- On December 7, 2018, the inspector presented the emergency preparedness program
inspection results to Mr.
- M. Hayworth, Site Emergency Preparedness Manager.
- On December 19, 2018, the inspector discussed the completed 2018 Licensed Operator
Requalification Program annual operating test inspection results with Ms.
Operations Training Manager.
- On January 9, 2019, the inspector presented the quarterly integrated inspection results to
Mr.
- W. Trafton, and other members of the licensee staff.
DOCUMENTS REVIEWED
71111.01Adverse Weather Protection
- AR 4172878; EST, TCCP, and Config Control Review for LOS-ZZ-A2
- AR 4181689; EED Switchyard Winterization Inspection
- LOS-ZZ-A2; Preparation for Winter/Summer Operation; Revision 56
- WC-AA-107; Seasonal Readiness; Revision 20
- WO 4717497-01; LOS-ZZ-A2 Winterize Station; 12/6/2018
71111.04Equipment Alignment
- 4185035; NRC Identified Potential Scaffold Issue in U2 CSCS Pump Room
- MA-AA-716-025; Scaffold Installation, Modification, and Removal Request Process;
Revision 15
- LAP-100-65; Equipment/Parts Storage in Plant Areas Containing Safety-Related Equipment;
Revision 9
- MA-AA-796-024; Scaffold Installation, Inspection, and Removal; Revision 11
- AR 4179046; SBGT 1FS-VG009 Indication Out of Calibration
- LaSalle NPS 90 Day Scaffold Report; 10/24/2018
- NES-MS-04.1; Seismic Prequalified Scaffolds; Revision 7
71111.05AQFire Protection Annual/Quarterly
- Pre-Fire Plan FZ 3H4; RX Bldg. 694-6 Elevation U2 RCIC/LPCS Cubicle
- Pre-Fire Plan FZ 3I4; RX Bldg. 673-4 Elevation U2 LPCS/RCIC Pump Cubicle
71111.06Flood Protection Measures
- AR 1413252; FUK: Fukushima Flooding Elevation Surveys Doors 20 and 164
- C467110014-9939; ISOLRB3, Terminate Flood Before HPCS, RHR A & LPCS/RCIC Rooms
Reach Critical Height; 2013
- EC 388864; Evaluate Leakage by MS Tunnel Dampers During Flood; Revision 000
- EC 399280; Beyond Design Basis Flooding Analysis for NRC Fukushima NTTF
Recommendation 2.1Plant LIP Ingress; Revision 004
- LOR-1PM13J-A304; RB NE/NW Equip DRN Sump Trouble; Revision 2
- LSCS-UFSAR; 3.4; Water Level (Flood) Design; Revision 20
- PMRQ 64856-01; Inspection of Magenetrol for the U-1 RCIC Pump Room; 10/25/2018
- WO 1112731-01; Perform Inspection of Magentrol; 1/19/2010
71111.07Heat Sink Performance
- EC 347674; Loop Accuracy for the RHR A & B Heat Exchanger Service Water Inlet Flow;
3/2/2004
- EC 626435; Evaluation of Unit 2B RHR Heat Exchanger Thermal Performance Data Using
Alternate (EPRI) Methodology; Revision 000
- WO 1816960-01; 2E12-B001 BRHR HX Heat Xfer Test per LTS-200-17; 12/7/2018
71111.11Licensed Operator Requalification Program and Licensed Operator Performance
- L2C17-15; L2C17 December 2018 Rod Pattern Adjustment and Quarterly Surveillances;
2/1/2018
71111.12Maintenance Effectiveness
- NRC Question on Unused Penetrations
- 1E-0-3073; Electrical Installation Fire-Stop & Fire-Barrier Details; Revision H
- LS-PSA-012; LaSalle PRA Internal Flood Analysis; Revision 1
- LTS-1000-29; Water Tight Penetration Inspection; Revision 15
- AR 4193160; NRC IddInadequate Detail in VR MR Evaluation
- AR 3972328; Left Blade on Damper 2VR08Y is Sticking
- WO 1814934-01; Inspect Steam Tunnel, Check Dampers 2VR08Y thru 2VR14Y; 2/9/2017
- WO 821498-01; Inspect Steam Tunnel Check Dampers 2VR08Y thru 2VR14Y; 2/18/2011
- WO 1909964-01; IVR90Y Press Relief Damper Open Torque Verification; 2/24/2018
- WO 1911148-01; Inspect Steam Tunnel Check Dampers 1VR08Y thru 1VR14Y; Recor (sic);
2/24/2018
- ER-LA-450-1006; LaSalle Structures Monitoring Instructions; Revision 3
- ER-LA-450; LaSalle Structures Monitoring Program; Revision 002
- ER-AA-450; Structures Monitoring; Revision 7
- ER-AA-310-1003; Maintenance RulePerformance Criteria Selection; Revision 5
- 5423 (Drawing); Steam Tunnel Check Dampers; Revision 2
- PMRQ 76697-01; Inspect Steam Tunnel Check Dampers 1VR08Y thru 1VR14Y; 10/10/2018
- PMRQ 75772-01; Inspect Steam Tunnel Check Dampers 1VR08Y thru 1VR14Y Record In;
10/10/2018
- AR 912656; Apply Lubricant to Chain Hold Down Bolt on VR Check Damper
- AR 891020; NOS ID: OPS CPA PRA Key Operator Actions
- M-1438; High Pressure Core Spray Switchgear Room and High Radiation Sampling
Ventilation System Elevation 687-0; Revision J
- M-3460; HVAC/C&I Diagram Turbine Building Ventilation System; Revision C
- DWG 5559; General Arrangement Steam Tunnel Check Damper; Revision 2
- DWG 5561; Electrical and Control Schematic Steam Tunnel Check Dampers Unit 2 Turbine
Building Air Return Risers; Revision 1
- M-1460, Sheets 1 & 2; P&ID Turbine Building Ventilation System; Revision J
- AR 4186903; NRC MR Questions 10/23/2018
- Current Installed Protected Pathway List;12/11/2018
- IR 4302231, NRC Question On Div 3 Protected Paths; 12/11/18
- S-237; Reactor Building Framing Section 8-8 Lower Area; Revision X
- WO 1916774-01; Watertight Penetration Inspection, Unit 1; 2/24/2018
- WO 1807623-01; Watertight Penetration Inspection, Unit 2; 4/27/2017
- IE-2-4085AM; Schematic Diagram Turbine Building Ventilation Syst. VT Pt. 12; Revision E
71111.13Maintenance Risk Assessments and Emergent Work Control
- ECR 436522; Request Approval of Crane and Pad to Support LaSalle FHD Installation,
Reactor Building; 7/24/2018
- ECR 436569; Request Approval of Scaffold Plans in Support of LaSalle FHD Installation;
7/30/2018
- L-004116; HCVS Steel Tower Load Drop Analysis, EC 392353-02; Revision 000
- RP-01; Plan on Crane Setup for Scaffold and Pipe Installation; Revision 4
- RP-02; Elevation on Crane Setup for Scaffold and Pipe Installation; Revision 2
- RP-03; Crane Pad Layout Plans, Sections & Details; Revision 3
- WO 4776605-09; Install/Remove Crane Pad for FHD Cooling Project, Reference EC 622967;
7/11/2018
- WO 4776605-11; Erect Outside Tower Scaffold for FHD Cooling Modification
- WO 4776605-27; Install Supports per EC 622967 AWA #3
71111.15Operability Determinations and Functionality Assessments
- 1E-1-4000CU; Key Diagram 480V MCC 135X-3 (1AP73E); Revision P
- 1E-1-4000CU; Key Diagram 480V MCC 135X-3 (1AP73E); Revision P
- 1E-1-4000FB; Key Diagram 125V DC Distribution Essential Div. 1; Revision T
- 67062E; Turbine File 36687 Drawing; 12/31/1969
- A-261; Reactor Building Wall Sections; Revision D
- AR 4195336; NRC IDdRB Penetration Tracking
- AR 4195744; 4.0 Crew Critique and LL of 1DC07E Battery Replacement; 11/16/2018
- CC-AA-204; Control of Vendor Equipment Manuals; Revision 12
- EC 625223; Technical Evaluation of Procedure LEP-DC-116; Revision 0
- EPRI; Technical Report, Nuclear Maintenance Applications Center, RCIC; 2017
- HLA Brief for 1DC07E Div 1 125V Battery Replacement Unit 1 710 Aux Building; Undated
- LEP-DC-116; Division 1 and 2 Switchgear Room 125 Volt Battery Cell Replacement for Units
and 2; Revision 1
- LMP-RI-01; Replacement of Outboard Mechanical Seal Assembly; Revision 7
- LOS-DC-Q2; Battery Readings for Safety-Related 250 VDC and Div 1,2,3 125 VDC Batteries;
Revision 36
- LOS-RI-Q4; RCIC; Revision 22
- M-1776, Sheets 2 & 4; ASME Weld Map; 11/13/2018
- Money/Johnson Risk Management Communication; Paragon and PRA Model Update;
5/22/2018
- NRC ID: Low Oil Level in U2 RCIC Turbine Sight Glass
- RCIC GS-1, GS-2; Lubrication System; Undated
- WO 4758919-01; Install New Battery Cells for 1DC07E; Undated
- WO 4776605-59; Install Pipe Penetration A Through the Reactor Building Metal Siding Wall
on 843; 11/13/2018
71111.18Plant Modifications
- 4196929; NRC IdentifiedQuestions Regarding CSCS Project Work
- CC-AA-402; Maintenance Specification: Installation of Temporary Rigging; Revision 5
- EC 622658; HP Pipe Loading Impact; Revision 000
- M-74; P&ID Cycled Condensate Storage; Revision AD
- M-766; Outdoor Piping; Revision AD
- M-938; High Pressure Core Spray Piping; Revision F
- NES-MS-04.2; BWR Stations Temporary Rigging Load Criteria; Revision 2
- SK-M-622658-1/EC 622658; Temporary Support for 10: Hot Tap/Plagging Machine;
11/9/2018
- WO 4777021-10; CM Remove Section of 2HP01C-24 / EC 622658; 11/19/2018
- WO 4777021-10; Remove A Section of Line 2HP01C-24 To Provide Access for the Line Stop
Machine as Per EC 622658
71111.19Post Maintenance Testing
- WO 1564237-03; Replace Hydramotor for 2VD19Y; 9/25/2018
- WO 1665024-02; Replace Hydramotor for 2VD03YA/YB; 9/26/2018
- WO 1853112-02; EM Cubicle Insp. LES-GM-108 @ 243-1 7B-2VD07C (DG Inop);
9/25/2018
- WO 1853113-02; Perform LES-GM-108 for Dist Trans @ MCC 243-1 CUB 6D (2AP79E);
9/25/2018
- WO 1853115-02; EM Cubicle Insp. LES-GM-108 @ 243-1 6A-2VD05C (DG Inop);
9/25/2018
- WO 1853409-02; EM Cubicle Insp. LES-GM-108 @ 243-1 4E-2VD01C (DG Inop);
9/26/2018
- WO 1853412-02; EM Cubicle Insp. LES-GM-108 @ 243-1 5B-2D002P (DG Inop);
9/25/2018
- WO 1882382-05; 2E12-F003B Klockner Moeller MCC 2AP82E-A6 Cubicle Replacement;
10/15/2018
- WO 1882382-08; 2E12-F003B Klockner Moeller MCC 2AP82E-A6 Cubicle Replacement;
10/15/2018
- WO 1923264-02; Perform LES-DG-202 Att A, B & C if Applicable on the U2 B D; 9/26/2018
- WO 1937824-03; MM Replace 2E22-F316 With SS Valve per IT-7000-M-PP-16; 9/25/2018
- WO 1964571-02; Inspect 2B Diesel Generator Start Air Moisture Separate; 9/26/2018
- WO 4791067-02; Replace Jacket Water Outlet Gasket on 2B DG; 9/26/2018
- WO 4801283-01; LRA LOS-DG-M3 2B DG Fast Start ATT 2B-Fast; 9/26/2018
- WO 484154-01; IM-EWP-1D-K751D 1D VD Rad Monitor Downscale; 10/23/2018
71111.22Surveillance Testing
- LOS-FP-M6; Diesel Fire Pump 0A (0FP01KA) Operational Check; 12/19/2018
- LOS-RH-Q1; Unit 2 B RHR System Operability and Inservice Test; 12/18/2018
71114.04Emergency Action Level and Emergency Plan Changes
- 10 CFR 50.54(q) Evaluator Qualification Spreadsheet; Dated May 30, 2018
- 50.54(q) Evaluation No.17-105; EP-AA-1005, Addendum 3, Emergency Action Levels for
LaSalle Station (Revision 3) Evaluation and Effectiveness Review; Dated October 13, 2017
- 50.54(q) Evaluation No. 17-72; EP-AA-1005, Exelon Nuclear Radiological Emergency Plan
Annex for LaSalle Station (Revision 40) Evaluation and Effectiveness Review; Dated
August 28, 2017
- 50.54(q) Evaluation No. 17-89; EP-AA-1005, Addendum 3, Emergency Action Levels for
LaSalle Station (Revision 3) Evaluation and Effectiveness Review; Dated July 24, 2017
- 50.54(q) Evaluation No. 18-13; EP-AA-1005, Addendum 3, Emergency Action Levels for
LaSalle Station (Revision 4) Evaluation and Effectiveness Review; Dated February 21, 2018
- AR 04096437; Typographical Error Identified in EAL Matrix
- AR 04109555; Enhancement Opportunity to Improve Knowledge for the C6 EAL
- EP-AA-1000; Exelon Nuclear Standardized Radiological Emergency Plan: Revision 29
- EP-AA-1005 Addendum 1; LaSalle Station On-Shift Staffing Technical Basis; Revision 1
- EP-AA-1005, Addendum 3; Emergency Action Levels for LaSalle Station; Revisions 2, 3,
and 4
- EP-AA-1005; Exelon Nuclear Radiological Emergency Plan Annex for LaSalle Station;
Revisions 39 and 40
- EP-AA-120; Emergency Plan Administration; Revision 21
- EP-AA-120-1001; 10 CFR 50.54(q) Change Evaluation; Revision 9
71124.02Occupational As Low As Reasonably Achievable Planning and Controls
- AR 03949782; NRC Rad Protection Baseline Inspection Self-Assessment; 12/11/2017
- AR 03977765; Increased CRD Dose Rates in L2R16; 02/23/2017
- AR 03983436; Radiation Protection Baseline (71124.02) Self-Assessment; 08/30/2018
- Issue Report 4005125; ALARA Post-Job Review - L2R16 RB/DW Chemical Decontamination
Project; 04/17/2017
- LaSalle Station RP / ALARA Refuel Outage Report; L1R17; 2018
- LaSalle Station RP / ALARA Refuel Outage Report; L2R16; 2017
- Radiation Work Permit and Associated ALARA File; LA-01-18-00510; L1R17 DW Steam
Safety Relief Valve Activities
- Radiation Work Permit and Associated ALARA File; LA-02-17-00502; L2R16 DW RP
Department Activities
- Radiation Work Permit and Associated ALARA File; LA-02-17-00506; L2R16 DW Scaffold
- Radiation Work Permit and Associated ALARA File; LA-02-17-00513; L2R16 DW Control Rod
Drive (CRD) Exchange
- Radiation Work Permit and Associated ALARA File; LA-02-17-00547; L2R16 DW RR Motor
Replacement
- RP-AA-400; ALARA Program; Revision 15
- RP-AA-400-1001; Establishing Collective Radiation Exposure Annual Business Plan Goals;
Revision 5
- RP-AA-400-1004; Emergent Dose Control and Authorization; Revision 9
- RP-AA-400-1006; Outage Exposure Estimating and Tracking; Revision 8
- RP-AA-400-1008; Exposure Goal Recovery Plans; Revision 3
- RP-AA-401; Operational ALARA Planning and Controls; Revision 24
71151Performance Indicator Verification
- LS-AA-2140; Attachment 1; Monthly Data Elements for NRC Occupational Control
Effectiveness; October 2017 through September 2018
- Periodic Assessment of Maintenance Rule Program; LaSalle Station, Units 1 & 2; July 2016
through June 2018
- Periodic Assessment of Maintenance Rule Program; LaSalle Station, Units 0, 1 & 2; July 2014
through June 2016
71152Problem Identification and Resolution
- AR 00002269; Document Actions Being Taken to Correct Water That is Leaking; 03/29/1999
- AR 2420888; Unit 2 Reactor Cavity Skirt Plate to Drain Line Leakage; 12/04/2014
- AR 4193011; NRC ID: Potential NCV for LaSalle Unit 2 Cavity Leakage; 11/08/2018
- ATI 1470953-18-47; Actions Supporting Issues Identified by LS-AA-2001; 07/29/2014
- Drawing S-326; Sections and Details, Reactor Containment Liner Plate, Sheet 1; Revision AE
- PI-AA-125; Corrective Action Program (CAP) Procedure; Revision 6
71153Follow-Up of Events and Notices of Enforcement Discretion
- AR 04110929; 1B21-F013R Fails Set-Pressure Test; 02/27/2018
- AR 04110933; 1B21-F013U Fails Set-Pressure Test; 02/27/2018
- Licensee Event Report, 05000373/2013-003-01, Two Main Safety Relief Valves Failed
Inservice Lift Inspection Pressure Test, 07/25/2018
- NWS Technology, Letter; LaSalle SRV s/n N63790-05-0016- As found Failure; Revision 0
- NWS Technology, Letter; LaSalle SRV s/n N63790-05-0076- As found Failure; Revision
2