ML19045A403

NRC Integrated Inspection Report 05000373/2018004 and 05000374/2018004

ML19045A403
Person / Time
Site:	LaSalle
Issue date:	02/14/2019
From:	Kenneth Riemer NRC/RGN-III/DRP/B1
To:	Bryan Hanson Exelon Generation Co, Exelon Nuclear
References
Ir 2018004
Download: ML19045A403 (24)

See also: IR 05000373/2018004

Text

First initial La

UNITED STATES

NUCLEAR REGULATORY COMMISSION

REGION III

2443 WARRENVILLE RD. SUITE 210

LISLE, ILLINOIS 60532-4352

February 14, 2019

Mr. Bryan C. Hanson

Senior VP, Exelon Generation Company, LLC

President and CNO, Exelon Nuclear

4300 Winfield Road

Warrenville, IL 60555

SUBJECT: LASALLE COUNTY STATION, UNITS 1 AND 2NRC INTEGRATED

INSPECTION REPORT 05000373/2018004 AND 05000374/2018004

Dear Mr. Hanson:

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

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.

B. Hanson -2-

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

B. Hanson -3-

Letter to Bryan Hanson from Kenneth Riemer dated February 14, 2019

SUBJECT: LASALLE COUNTY STATION, UNITS 1 AND 2NRC INTEGRATED

INSPECTION REPORT 05000373/2018004 AND 05000374/2018004

DISTRIBUTION:

Michael McCoppin

RidsNrrDorlLpl3

RidsNrrPMLaSalle

RidsNrrDirsIrib Resource

Darrell Roberts

John Giessner

Jamnes Cameron

Allan Barker

DRPIII

DRSIII

ROPreports.Resource@nrc.gov

ADAMS Accession Number: ML19045A403

OFFICE RIII RIII

NAME CPhillips:bw KRiemer

DATE 1/29/2019 2/14/2019

OFFICIAL RECORD COPY

U.S. NUCLEAR REGULATORY COMMISSION

REGION III

Docket Numbers: 50-373; 50-374;72-070

License Numbers: NPF-11; NPF-18

Report Numbers: 05000373/2018004; 05000374/2018004

Enterprise Identifier: I-2018-004-029

Licensee: Exelon Generation Company, LLC

Facility: LaSalle County Station, Units 1 and 2

Location: Marseilles, IL

Dates: October 1 through December 31, 2018

Inspectors: W. Schuap, Senior Resident Inspector, LaSalle

J. Havertape, Resident Inspector, LaSalle

C. Phillips, RIII Project Engineer

J. Cassidy, Sr. Health Physicist

D. Sargis, Resident Inspector, Clinton

M. Holmberg, Sr. Reactor Inspector

V. Meghani, Reactor Inspector

G. Hansen, Sr. Emergency Preparedness Inspector

C. Zoia, Sr. Operations Engineer

R. Zuffa, Illinois Emergency Management Agency

Approved by: K. Riemer, Chief

Branch 1

Division of Reactor Projects

Enclosure

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.

Technical Specification Surveillance Procedure Aceptance Criteria Did Not Consider

Instrument Uncertainty

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 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings,

for the licensees failure to include an appropriate acceptance criterion for drywell oxygen

concentration in surveillance procedures. Specifically, the acceptance criterion did not

account for instrument uncertainty.

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

2

TABLE OF CONTENTS

PLANT STATUS ........................................................................................................................... 4

INSPECTION SCOPES ................................................................................................................ 4

REACTOR SAFETY ..................................................................................................................... 4

RADIATION SAFETY ................................................................................................................... 7

OTHER ACTIVITIES - BASELINE .............................................................................................. 8

INSPECTION RESULTS .............................................................................................................. 9

EXIT MEETINGS AND DEBRIEFS ............................................................................................ 15

DOCUMENTS REVIEWED ......................................................................................................... 16

3

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 ruturn 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.

4

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.

5

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;

6

(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;

7

(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.

8

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.

9

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

90 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

10

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

90 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:

11

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

Enforcement Policy.

71152Problem Identification and Resolution

Acceptance Criteria for Technical Specification Surveillance Requirement Did Not Consider

Instrument Uncertainty

Cornerstone Significance Cross-Cutting Report Section

Aspect

Barrier Integrity Green None 71152

NCV 05000373/2018004-02

Open/Closed

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 to include an appropriate acceptance criterion for drywell oxygen

concentration in surveillance procedures. Specifically, the acceptance criterion did not

account for instrument uncertainty.

12

Description:

Technical Specification (TS) 3.6.3.2, Primary Containment Oxygen Concentration, required,

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 required 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

TS 3.6.3.2.

On 29 May, 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, during which the

1PL78J indicated 1 percent oxygen and the post-LOCA oxygen detection system indicated

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. The

licensee determined that the issue was not reportable under 10 CFR 50.73(a)(2)(i)(B) for a

condition prohibited by TS.

Since 1PL78J is an instrument permitted by procedure LOS-AA-W1 to accomplish

SR 3.6.3.2.1, the inspectors reviewed the calibration data for 1PL78J under work order 4718759 and 4789722. During their review, the inspectors noted that instrument calibration

procedure, LIP-CM-510, Unit 1 Primary Containment Continuous Oxygen Monitor Sensor

Maintenance and Standardization, Revision 7, Step E.2.10.5, required that oxygen

concentration stabilize at approximately 20 to 22 percent and greater than 19.5 percent

oxygen as read in air. The inspectors determined that Step E.2.10.5 of LIP-CM-510

contained a leave alone tolerance of approximately 1.4 percent oxygen since air normally

contains 20.9 percent oxygen.

The inspectors concluded that since LOS-AA-W1 contained approximately 0.5 percent of

margin, and LIP-CM-510 allowed for as much as 1.4 percent deviation in the instrument

response of 1PL78J, that the acceptance criteria of LOS-AA-W1 had the potential to allow a

condition prohibited by TSs. Further, the inspectors found several instances of instrument

drift in the continuous oxygen monitor. This was revealed by abnormally low oxygen readings

upon de-inerting the drywell for refueling outages documented in AR 1170271, 1325742,

1689590, 2626000, and 3970533. 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 who

documented them in the corrective action program as AR 4209125.

Corrective Actions: The licensee intends on performing a loop uncertainty determination for

the continuous oxygen monitor and to re-evaluate the acceptance criteria of LIP-CM-510 and

LOS-AA-W1 based on the results of the uncertainty determination.

Corrective Action Reference: AR 4209125

13

Performance Assessment:

Performance Deficiency: The inspectors determined that the failure to include appropriate

acceptance criteria for drywell oxygen concentration in TS surveillance procedures, an activity

affecting quality, was contrary to 10 CFR Part 50, Appendix B, Criterion V, Instructions,

Procedures, and Drawings, and was a performance deficiency.

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, the failure to include instrument

uncertainty as part of the acceptance criteria that performs the weekly TS SR had the

potential to allow a condition prohibited by TS.

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) 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: Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and

Drawings, requires, in part, that instructions, procedures, or drawings include appropriate

quantitative or qualitative acceptance criteria for determining that important activities have

been satisfactorily accomplished.

Contrary to the above, from January 16, 1999, to December 31, 2018, the licensee failed to

include appropriate acceptance criteria in procedures that implement TS surveillance

requirements when the continuous oxygen monitor was installed under engineering

change 55203. Specifically, the drywell oxygen acceptance criteria in procedure

LOS-AA-W1 did not account for instrument uncertainties to ensure compliance with

TS limits.

Disposition: This violation is being treated as an NCV, consistent with Section 2.3.2 of the

Enforcement Policy.

71153Follow-Up of Events and Notices of Enforcement Discretion

Licensee Identified Non-Cited Violation 71153Follow-Up of Events and Notices of

Enforcement Discretion

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.

14

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)

2 1205 +/- 36.1

3 1195 +/- 35.8

2 1185 +/- 35.5

4 1175 +/- 35.2

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. D. Fuson,

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.

15

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

16

71111.11Licensed Operator Requalification Program and Licensed Operator Performance

- L2C17-15; L2C17 December 2018 Rod Pattern Adjustment and Quarterly Surveillances;

12/1/2018

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; Sturctures 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 912656912656 Apply Lubricant to Chain Hold Down Bolt on VR Check Damper

- AR 891020891020 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 Secion 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

17

- 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

1 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

18

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

19

- 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

20

- 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

21