IR 05000373/2019001
| ML19128A260 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 05/08/2019 |
| From: | Kenneth Riemer NRC/RGN-III |
| To: | Bryan Hanson Exelon Generation Co, Exelon Nuclear |
| References | |
| IR 2019001 | |
| Download: ML19128A260 (25) | |
Text
SUBJECT:
LASALLE COUNTY STATION, UNITS 1 AND 2NRC INTEGRATED INSPECTION REPORT 05000373/2019001 AND 05000374/2019001
Dear Mr. Hanson:
On March 31, 2019, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your LaSalle County Station, Units 1 and 2. On April 10, 2019, the NRC inspectors discussed the results of this inspection with Mr. William Trafton and other members of your staff. The results of this inspection are documented in the enclosed report.
NRC inspectors documented two findings of very low safety significance (Green) in this report.
One of these findings involved a violation of NRC requirements. Because the licensee has initiated actions within their corrective action program to address this issue, the violation is being treated as Non-Cited Violation (NCV), consistent with Section 2.3.2 of the Enforcement Policy. The NCV is described in the subject inspection report.
If you contest the violation or significance or severity of the violation documented in this inspection report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN:
Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region III; the Director, Office of Enforcement; and the NRC resident inspector at LaSalle.
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 LaSalle. 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.: 05000373; 05000374 License Nos.: NPF-11; NPF-18
Enclosure:
IR 05000373/2019001; 05000374/2019001
Inspection Report
Docket Numbers: 05000373 and 05000374 License Numbers: NPF-11 and NPF-18 Report Numbers: 05000373/2019001 and 05000374/2019001 Enterprise Identifier: I-2019-001-0060 Licensee: Exelon Generation Company, LLC Facility: LaSalle County Station, Units 1 and 2 Location: Marseilles, IL Inspection Dates: January 01, 2019 to March 31, 2019 Inspectors: J. Bozga, Senior Reactor Inspector J. Cassidy, Senior Health Physicist G. Edwards, Health Physicist R. Elliott, Resident Inspector T. Go, Health Physicist J. Havertape, Resident Inspector C. Phillips, Project Engineer W. Schaup, Senior Resident Inspector L. Torres, Illinois Emergency Management Agency R. Zuffa, Illinois Emergency Management Agency Approved By: Kenneth Riemer, Chief Branch 1 Division of Reactor Projects Enclosure
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting a 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.
List of Findings and Violations Failure To Make an Individual Knowledgeable of Dose Rates Prior To Entry Into a High Radiation Area Cornerstone Significance Cross-Cutting Report Aspect Section Occupational Green [H.3] - Change 71124.01 Radiation Safety NCV 05000373,05000374/2019001- Management
Open/Closed An NRC identified Green finding and associated Non-cited Violation of Technical Specification 5.7.1. (e) High Radiation Areas, was identified when the licensee allowed an individual to make entry into a high radiation area where dose rates had been determined, but the individual was not made knowledgeable of the current dose rates. Specifically, the individual received a high radiation area briefing that used survey maps that did not reflect current conditions in the area.
Failure to Follow Station Erosion in Piping and Components Guide Results in Three Inch Hole in 2TEC5A-Header Cornerstone Significance Cross-Cutting Report Aspect Section Initiating Events Green [P.2] - Evaluation 71153 FIN 05000374/2019001-02 Open/Closed The inspectors documented a self-revealed finding of very low safety significance for the licensees failure to follow station procedure ER-AA-430-1004, Erosion in Piping and Components (EPC) Guide, Revision 1. Specifically, design engineering personnel failed to use a safety factor when determining the time (remaining service life) it would take for the 2TEC5A-header degradation to fall below the minimum wall limit. This evaluation resulted in no corrective actions being taken for cracks found on the header which eventually failed during the operating cycle. A temporary patch was installed at power that did not provide sealing under all operating conditions which later resulted in a loss of condenser vacuum causing the operators to manually scram the reactor.
Additional Tracking Items Type Issue Number Title Report Section Status URI 05000373,05000374/2018003- Potential Failure to 71152 Closed 06 Inspect Containment Post-Tensioned Tendons per Code Requirements and to Follow Corrective Action Program Process URI 05000374/2018003-07 Potential Failure to 71152 Closed Promptly Correct the Unit 2 Primary Containment Wall Cavity Leakage Condition and to Follow Corrective Action Program Process
PLANT STATUS
Unit 1 began the inspection period at rated thermal power. On March 23, 2019, the unit was down powered to 73 percent for a rod pattern adjustment, scram time testing, channel distortion testing and turbine valve testing. The unit was returned to thermal rated power the same day and remained at or near rated thermal power for the remainder of the inspection period.
Unit 2 began the inspection period at thermal rated power. On January 12, 2019, the unit was down powered to 85 percent for rod recovery and returned to thermal rated power the same day. On February 2, 2019, the unit was down powered to 80 percent to perform channel distortion testing and returned to thermal rated power the same day. On February 18, 2019, the unit was shutdown to commence refueling outage L2R17. On March 8, 2019, the unit was started up after completion of the refueling outage. On March 18, 2019 the unit reached rated thermal power and remained at or near thermal rated 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
71111.01 - Adverse Weather Protection Impending Severe Weather Sample (IP Section 03.03)
The inspectors evaluated readiness for impending adverse weather conditions for extreme cold temperatures during the polar vortex on January 29, 2019.
71111.04 - Equipment Alignment Partial Walkdown (IP Section 02.01)
The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:
- (1) Unit 2 Division I emergency diesel generator on February 20, 2019
- (2) Unit 1 standby gas treatment system on February 25, 2019
- (3) Unit 2 Division 2 diesel generator on March 1, 2019
- (4) Unit 2 reactor core isolation cooling system on March 9, 2019
71111.04S - Equipment Alignment Complete Walkdown (IP Section 02.02)
The inspectors evaluated system configurations during a complete walkdown of the Unit 2 B residual heat removal system after shutdown cooling restoration on March 20, 2019.
71111.05Q - Fire Protection Quarterly Inspection (IP Section 03.01)
The inspectors evaluated fire protection program implementation in the following selected areas:
- (1) Unit 2 reactor building, 673' elevation B/C residual heat removal corner room on January 29, 2019
- (2) Unit 2 reactor building, 694' elevation, B/C residual heat removal corner room, control of hot work on February 27, 2019
- (3) Unit 1 reactor building 694' elevation B/C residual heat removal corner room on March 27, 2019
- (4) Unit 2 auxiliary building 687' elevation, Division III switchgear room on March 27, 2019
- (5) Unit 2 auxiliary building 731' elevation, Division II switchgear room on March 19, 2019
71111.08G - In Service Inspection Activities BWR In Service Inspection Activities Sample - Nondestructive Examination and Welding Activities (IP Section 03.01)
The inspectors verified that the reactor coolant system boundary, reactor vessel internals, risk-significant piping system boundaries, and containment boundary are appropriately monitored for degradation and that repairs and replacements were appropriately fabricated, examined and accepted by reviewing the following activities from February 19, 2019 to March 8th 2019:
03.01.a - Nondestructive Examination and Welding Activities
- Ultrasonic (UT) examination of Reactor Pressure Vessel (RPV) Top Head Welds GEL-1060-DH, DJ and DK
- UT examination of Main Steam (MS) pipe to reducer weld IMS-2044-03
- UT examination of MS pipe-to elbow IMS-2044-11
- UT examination of MS pipe-to-valve weld IMS-2044-13
- UT examination of Residual Heat Removal (RHR) heat exchanger inlet nozzle-to-head weld IRH-HX2B-01
- Liquid Penetrant (PT) of 2B Reactor Recirculation Pump Lug 2B33-C001B
- Visual Examination (VT-3) of MS Pipe Support RR00-2061X
- RPV Shell Longitudinal Weld LCS-2-BA indication evaluation
- Jet pump 19/20 Riser Weld RS-1c indication evaluation
- Weld repair on 2C41-F304 valve
71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01)
The inspectors observed and evaluated licensed operator performance in the Control Room during Unit 2 shutdown for refueling outage L2R17 on February 17, 2019.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
The inspectors observed and evaluated OBE 19-2/ESG 61 on March 26, 2019, and just-in-time training for reactor startup on February 2, 2019.
71111.13 - Maintenance Risk Assessments and Emergent Work Control Risk Assessment and Management Sample (IP Section 03.01)
The inspectors evaluated the risk assessments for the following planned and emergent work activities:
- (1) Unit 1 Green online risk management during Division II work impacting credited offsite line for safety bus on February 8, 2019
- (2) Unit 2 Protected equipment posting during core standby cooling system license amendment period on February 20, 2019
- (3) Units 1 and 2 Yellow online risk for switchyard activities and Unit 2 Yellow risk for B residual heat removal and hardened vent maintenance on March 18, 2019
- (4) Unit 2 Yellow online risk during high pressure core spray/Division III diesel generator maintenance on March 26, 2019
- (5) Unit 1 Yellow online risk due to high wind event on February 24, 2019
71111.15 - Operability Determinations and Functionality Assessments Operability Determinations and Functionality Assessments (IP Section 02.01)
The inspectors evaluated the following operability determinations and functionality assessments:
- (1) Misaligned main steam strut found on 777' elevation of drywell
- (2) Foreign material found in Unit 2 jet pump diffuser nozzles during refueling outage L2R17
- (3) Eaton 480V motor control center m-contactor sticking issue
- (4) Missing bolt on Unit 1 Division III diesel generator turbo charger
- (5) Unit 2 traversing in-core probe channel compliance to Core Operating Limits Report requirements
- (6) Offsite power operability while using unit auxiliary transformer to power the Division II safety bus during ACB 1422 relay calibration
71111.19 - Post Maintenance Testing Post Maintenance Test Sample (IP Section 03.01)
The inspectors evaluated the following post maintenance tests:
- (1) Unit 2 reactor water cleanup outboard isolation valve stem and disc replacement on March 1, 2019
- (2) Unit 2 Division III Integrated response time testing on February 21, 2019
- (3) Unit 2 Division II Integrated response time testing on March 28, 2019
- (4) Unit 2 Division II room cooler replacement and piping modification on March 1, 2019
- (5) Unit 2 source range monitor 'D' dry tube replacement on March 29, 2019
- (6) Unit 2 source range monitor A, C and D testing on February 25, 2019
- (7) Unit 2 high pressure core spray pump, valve and diesel generator testing on March 2, 2019
- (8) Unit 2 Division III diesel generator start and acceptance testing on February 18, 2019
- (9) Unit 2 B standby liquid control pump and valve testing on March 8, 2019
- (10) Unit 2 traversing in-core probe C testing on March 14, 2019
- (11) Unit 2 B residual heat removal service water testing on March 4, 2019
71111.20 - Refueling and Other Outage Activities Refueling/Other Outage Sample (IP Section 03.01)
The inspectors evaluated refueling outage L2R17 activities from February 17 to March 18, 2019.
71111.22 - Surveillance Testing The inspectors evaluated the following surveillance tests: Containment Isolation Valve (ISO) (IP Section 03.01)
- (1) Unit 2 traversing in-core probe local leak rate test on March 14, 2019
- (2) Unit 2 main steam isolation valve local leak rate testing on February 18, 2019
In Service Testing (IST) (IP Section 03.01) (1 Sample)
Unit 2 residual heat removal A and B shutdown cooling return line check valve local leak rate testing on February 20, 2019.
Surveillance Testing (IP Section 03.01) (6 Samples)
- (1) Unit 1 quarterly reactor core isolation cooling operability run on January 14, 2019
- (2) Unit 2 reactor water cleanup outboard isolation valve as-left local leak rate testing on March 1, 2019
- (3) Unit 2 main steam line inboard drain header local leak rate testing on February 18, 2019
- (4) Unit 2 reactor vessel leakage test on March 7, 2019
- (5) Unit 2 control rod SCRAM time testing on March 7, 2019
- (6) Unit 1 Division III emergency diesel generator monthly idle start on February 1,
RADIATION SAFETY
71124.01 - Radiological Hazard Assessment and Exposure Controls Contamination and Radioactive Material Control (IP Section 02.03)
The inspectors evaluated licensee processes for monitoring and controlling contamination and radioactive material. The inspectors verified the following sealed sources are accounted for and are intact:
- CS-137, ID Number 81-565
- CS-137, ID Number 81-553 High Radiation Area and Very High Radiation Area Controls (IP Section 02.05) (1 Sample)
The inspectors evaluated risk-significant high radiation area and very high radiation area controls.
Instructions to Workers (IP Section 02.02) (1 Sample)
The inspectors evaluated instructions to workers including radiation work permits used to access high radiation areas:
Radiation work packages
- Radiation Work Permit (RWP) LA-0-18-00201, Regulator Activities, Revision 1
- No Alarms Occurred During This Period of Inspection Labeling of containers
- LaSalle Unit 2 Refuel Floor, Hot Trash Receptacle
- LaSalle Unit 2 Drywell, Entrance/Exit Hot Trash Receptacle
- LaSalle Unit 2 Radwaste, Trash Receptacle Radiation Worker Performance and Radiation Protection Technician Proficiency (IP Section 02.06) (1 Sample)
The inspectors evaluated radiation worker performance and radiation protection technician proficiency.
Radiological Hazard Assessment (IP Section 02.01) (1 Sample)
The inspectors evaluated radiological hazards assessments and controls. The inspectors reviewed the following:
Radiological surveys
- LaSalle Unit 2 Reactor Building A Residual Heat Removal Corner Room
- LaSalle Unit 2 Reactor Building Upper Raceway Map
- LaSalle Unit 2 Refuel Floor
- LaSalle Unit 2 Drywell; Elevations 740, 760 and 777 Risk significant radiological work activities
- LaSalle Unit 2 Control Rod Drive Exchange
- LaSalle Unit 2 Reactor Assembly and Disassembly
- LaSalle Unit 2 Drywell Scaffolding Assembly Air sample survey records
- LaSalle Unit 2 Elevation 740 General Area, <0.3 DAC
- LaSalle Unit 2 Turbine Sandblasting, <0.3 DAC
- LaSalle Unit 2 Control Rod Drive Exchange, <0.3 DAC Radiological Hazards Control and Work Coverage (IP Section 02.04) (1 Sample)
The inspectors evaluated in-plant radiological conditions during facility walkdowns and observation of radiological work activities.
Radiological work package for areas with airborne radioactivity
- RWP LA-02-19-00901, L2R17 Reactor Disassembly and Reassembly Activities, Revision 00
- RWP LA-02-19-00906, L2R17 Cavity Decontamination Activities, Revision 00
- RWP LA-02-19-00805, L2R17 Turbine Building Sandblasting Activities, Revision 00
==71124.02 - Occupational ALARA Planning and Controls Implementation of ALARA and Radiological Work Controls (IPSection 02.03) (1 Sample)*
The inspectors reviewed ALARA practices and radiological work controls by reviewing the following activities:
- RWP LA-02-19-00513, L2R17 DW Control Rod Drive (CRD) Exchange, Revision 00
- RWP LA-02-19-00701, L2R17 Suppression Pool Diving, Revision 00
- These inspection activities supplement samples that were completed in the
===05000373/2018004 and 05000374/2018004 quarterly report. This completes the 02.03 Implementation of ALARA and Radiological Work Controls sample.
Radiation Worker Performance (IP Section 02.04)==
The inspectors evaluated radiation worker and radiation protection technician performance during observation of work activities listed in
OTHER ACTIVITIES - BASELINE
71151 - Performance Indicator Verification The inspectors verified licensee performance indicators submittals listed below:
IE01: Unplanned Scrams per 7000 Critical Hours Sample (IP Section 02.01) ===
- (1) Unit 1, January 1, 2018 - December 31, 2018
- (2) Unit 2, January 1, 2018 - December 31, 2018 IE03: Unplanned Power Changes per 7000 Critical Hours Sample (IP Section 02.02)
(2 Samples)
- (1) Unit 1, January 1, 2018 - December 31, 2018
- (2) Unit 2, January 1, 2018 - December 31, 2018 IE04: Unplanned Scrams with Complications (USwC) Sample (IP Section 02.03)
(2 Samples)
- (1) Unit 1, January 1, 2018 - December 31, 2018
- (2) Unit 2, January 1, 2018 - December 31, 2018
==71152 - Problem Identification and Resolution Annual Follow-Up of Selected Issues (IP Section 02.03) (1 Partial)
The inspectors reviewed the licensees implementation of its corrective action program related to the following issues:
The inspectors completed review of Unresolved ltem (URl)05000373/2018003-06; 05000374/2018003-06 Potential Failure to Inspect Containment Post-Tensioned Tendons per Code Requirements and to Follow Corrective Action Program Process and URI 05000374/2018003-07 Potential Failure to Promptly Correct the Unit 2 Primary Containment Wall Cavity Leakage Condition and to Follow Corrective Action Program Process. The inspectors' review and closure of these URIs did not constitute an inspection sample.
71153 - Follow-Up of Events and Notices of Enforcement Discretion Event Report (IP Section 03.02)==
The inspectors evaluated the following licensee event reports which can be accessed at https://lersearch.inl.gov/LERSearchCriteria.aspx:
- (1) LER 05000373/2017-007-00, Low Pressure Core Spray System Inoperable Due to Loss of Cooling, ADAMS Accession: ML17230A288.
The inspectors determined that it was not reasonable to foresee or correct the cause discussed in LER 2017-007-00; therefore no performance deficiency was identified.
The inspectors also concluded that no violation of NRC requirements occurred.
- (2) LER 05000374/2018-001-00 and LER 05000374/2018-001-01, Manual Reactor Scram Due To Main Condenser Vacuum Degradation (ADAMS Accession:
The circumstances surrounding this LER are documented in the Results section.
INSPECTION RESULTS
Failure To Make an Individual Knowledgeable of Dose Rates Prior To Entry Into a High Radiation Area Cornerstone Significance Cross-Cutting Report Aspect Section Occupational Green [H.3] - Change 71124.01 Radiation Safety NCV 05000373,05000374/2019001-01 Management Open/Closed An NRC identified Green finding and associated Non-cited Violation of Technical Specification 5.7.1.
- (e) High Radiation Areas, was identified when the licensee allowed an individual to make entry into a high radiation area where dose rates had been determined, but the individual was not made knowledgeable of the current dose rates. Specifically, the individual received a high radiation area briefing that used survey maps that did not reflect current conditions in the area.
Description:
On December 27, 2018, an individual was granted access to the Unit 2 A Residual Heat Removal (RHR) Corner Room, an area where dose rates exceeded 100 millirem per hour (mrem/hr), after a high radiation area briefing was completed by radiation protection. The Radiation Protection Technician began the briefing by using the Plant Viewer Digital Survey System. The surveys that were used during this briefing were conducted on 07/25/2018.
However, plant operating parameters changed in late August 2018 causing increased dose rates in the Unit 2 A RHR Corner Room. The operating parameter changes in the plant caused several areas within the Unit 2 A RHR Corner Room to meet the criteria for a high radiation area. The highest dose rate in the area was 450 mrem/hr on contact/210 mrem/hr at 30 centimeters. This change was identified when the licensee conducted surveys on 09/04/2018 and 09/05/2018 when the area was designated as a High Radiation Area.
The individual receiving the briefing on December 27, 2018, had been in the Unit 2 A RHR Corner Room after the September 5, 2018 and questioned the Radiation Protection Technician about the radiological information presented. The Radiation Protection Technician affirmed that the information was current because the survey was the last record in the Plant Viewer Digital Survey System and was within the prescribed survey frequency. Specifically, the surveys used during this briefing were conducted on 07/25/2018, and the area was not due to be surveyed again until 01/21/2019.
The individual returned from the Unit 2 A RHR Corner Room and reported to Radiation Protection that the electronic dosimeter provided by the licensee indicated that dose rates were higher than briefed and similar to entries after September 5, 2018. Although the dose rates on the electronic dosimeter were lower than the 100 mrem/hour at 30 centimeters which defines a high radiation area, the individual and Radiation Protection Technician recognized that the individual had access to the entire Unit 2 A RHR Corner Room where radiation levels exceeding 100 mrem/hr were present in different localized areas. Had the individuals travel path been slightly altered, the individual could have entered into areas of higher radiation levels without being knowledgeable of the current dose rates. The Radiation Protection Technician then notified his supervisor of the event that had occurred.
Corrective Action: The licensee conducted an investigation and were able to conclude that only surveys that are marked as routine are included in the Plant Viewer Digital Survey System.
Surveys that are not marked routine, such as the surveys conducted to document changes in radiological conditions like the surveys performed 9/4/2018 and 9/5/2018 in the Unit 2 A RHR Corner Room, would not be available if an individual were to use the Plant Viewer Digital Survey System for radiological area briefings. The licensee suspended the use of all the Plant Viewer Digital Survey System for conducting high radiation area briefings.
Corrective Action Reference: AR: 042117520
Performance Assessment:
Performance Deficiency: The licensee allowed an individual to make entry into a high radiation area where dose rates had been determined, but the individual was not knowledgeable of the current dose rates. Specifically, the individual received a high radiation area briefing that used survey maps that did not reflect current conditions in the area.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Program & Process attribute of the Occupational Radiation Safety cornerstone. Specifically, worker entry into areas where workers are not made knowledgeable of current dose rates could lead to unintended dose.
Significance: The inspectors assessed the significance of the finding using Appendix C, Occupational Radiation Safety SDP. The inspectors determined that the finding was of very low safety significance (Green) because:
- (1) it did not involve as-low-as-reasonably-achievable planning or work controls,
- (2) there was no overexposure,
- (3) there was no substantial potential for an overexposure, and
- (4) the ability to assess dose was not compromised.
Cross-Cutting Aspect: H.3 - Change Management: Leaders use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority.
Specifically, the licensee implemented a new system to perform area radiation briefs without understanding limitations of the system, i.e. the Digital Plant Survey System only displayed a subset of all radiological surveys performed which may not include the most recent survey of each area.
Enforcement:
Violation: Technical Specification 5.7.1, requires in part that high radiation areas with dose rates not exceeding 1.0 rem/hour at 30 centimeters from the radiation source or from any surface penetrated by the radiation.
(e). Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such, individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas.
This dose rate determination, knowledge, and per-job briefing does not require documentation prior to initial entry.
Contrary to the above, on December 27, 2018, the licensee failed to make an individual knowledgeable of dose rates prior to allowing the individual entry into a high radiation area, in accordance with Technical Specification 5.7.1 (e). Specifically, the individual received a high radiation area briefing that used survey maps that did not reflect current conditions in the area.
Enforcement Action: This violation is being treated as an Non-Cited Violation, consistent with Section 2.3.2 of the Enforcement Policy.
Unresolved Item Potential Failure to Inspect Containment Post-Tensioned 71152 (Closed) Tendons per Code Requirements and to Follow Corrective Action Program Process 05000373,05000374/2018003-06
Description:
In 2003, the licensee identified and repaired degraded group A vertical tendons in Units 1 and 2 caused by moisture induced corrosion, and provided this information to the NRC in the Unit 1 and 2 post outage Inservice Inspection (ISI) summary reports as required by 10 CFR 50.55a. In these reports, the licensee assumed that the extent of corrosion induced degradation did not apply to the Group B vertical tendon locations because of a welded cover that precluded entry of water. In the LaSalle ISI Classification Basis Document, the licensee determined that both radiological and structural obstructions (welded covers) exist for the Group B vertical tendons on both Units. Therefore, the licensee considered these locations inaccessible and did not perform periodic visual examinations of the Group B vertical tendon anchorage locations as allowed by the American Society of Mechanical Engineers (ASME)
Code Section XI, Subsection IWL and 10 CFR 50.55a. However, the inspectors identified a previous licensee record of water found within a Group B inaccessible vertical tendon location and issued an Unresolved Item (URI)05000373/2018003-06; 05000374/2018003-06 associated with the licensees decision to delay followup inspections of Group B tendon anchorage locations to validate the assumption that the welded cover would preclude moisture induced corrosion. The inspectors determined that there was no immediate safety concern associated with this issue, because there was no known degraded conditions at the inaccessible Group B tendon locations. Additionally, based on review of the licensees procedure PI-AA-120 Issue Identification and Screening Process, the degraded Group A and the potential for degradation in the Group B tendons did not meet the licensees definition of a significant condition adverse to quality. In response to the inspectors review of this issue, the licensee initiated corrective action assignments to document evaluations of the tendon degradation in Units 1 and 2, and also coded work orders for inspection of a sample of Group B inaccessible tendon locations as Corrective Actions (CAs), (i.e. actions required to correct a condition adverse to quality) rather than Action Tracking Items (ACITs), (i.e. associated with actions not adverse to quality, to be completed only as improvement) as initially coded. Based on review of licensee corrective actions associated with this issue, no violations of NRC regulations were identified and URI 05000373/2018003-06; 05000374/2018003-06 is closed.
Unresolved Item Potential Failure to Promptly Correct the Unit 2 Primary 71152 (Closed) Containment Wall Cavity Leakage Condition and to Follow Corrective Action Program Process05000374/2018003-07
Description:
In 1998, the licensee identified leakage through the Unit 2 primary containment wall caused by a defective drain line to cavity skirt plate weld, but as of December 31, 2018, had not successfully corrected this condition. URI 05000374/2018003-07 was opened due to the inspectors concern that continued operation with this leakage could adversely affect the structural capability of the primary containment and to assess whether an enforcement action was applicable.
The licensee identified that the leakage was approximately 20-25 drops per minute at the primary locations and that the source of this water leakage through the primary containment wall was located at a 2 fuel pool cooling drain line weld connecting this line to the cavity skirt plate.
The licensee attempted to stop the leakage through installation of temporary plugs but this action was unsuccessful. The licensee also initiated a work order 98109950 to repair the defective weld, but the weld repair was never implemented and the licensee subsequently cancelled this work order. In 2010, the licensee again documented this leakage in AR 1086083, but closed this AR without correcting this condition.
In December of 2014, the licensee again documented this leakage through the Unit 2 primary containment wall in AR 2420888. This AR included a history of the actions associated with the cavity leakage since 1998 and identified planned corrective actions including ultrasonic examination of the liner every other refueling outage, a weld repair, and an evaluation of the structural impact of the leakage on the concrete, reinforcing steel, tendons, and liner. The licensee subsequently completed a technical evaluation (assignment ATI-1470953-18-47) in 2014 and considering the factors such as minor localized nature of leakage, short outage durations during which the leakage occurred, the alkaline environment of concrete surrounding steel liner / reinforcement, and the results of the ultrasonic thickness reading of the liner plate, concluded that there was no adverse impact on structural adequacy of the containment from this leakage and also referred to an existing work order 855785 for repair of the leakage in the subsequent outage. On March 10, 2017, the licensee closed the corrective action assignment (AR 2420888-04) for the weld repair to work order 855785 to be completed in 2019. Inspectors noted that the work order 855785 was originally scheduled for 2007 but had been repeatedly deferred. In response to the inspectors review, the licensee confirmed that the WO 855785 to perform weld repair was scheduled for implementation in the upcoming 2019 outage.
Closure Basis: The inspectors reviewed the licensees past activities, as described above, as well as the planned actions to stop the leakage by repair of the drain line to skirt plate weld at two locations during the 2019 refueling outage. The inspectors also assessed the licensee's evaluation of the leakage on the structural integrity of primary containment, including the concrete, reinforcing steel, tendons, and liner, and concluded there was no significant immediate impact. The inspectors subsequently followed up on the issue during the recent refueling outage and noted that repair work intended to stop the leakage was actually completed at one location (associated with the more significant leak) while the other location repair was deferred to the next outage due to difficulties encountered during performance of work. The inspectors discussed the leakage history and the associated actions with the NRC Region III Enforcement and Investigations Coordination Staff (EICS). Based on those discussions and the inspectors' assessment, no findings were identified and URI 05000374/2018003-07 is closed.
Failure to Follow Station Erosion in Piping and Components Guide Results in Three Inch Hole in 2TEC5A-Header Cornerstone Significance Cross-Cutting Report Aspect Section Initiating Events Green [P.2] - 71153 FIN 05000374/2019001-02 Evaluation Open/Closed The inspectors documented a self-revealed finding of very low safety significance for the licensees failure to follow station procedure ER-AA-430-1004, Erosion in Piping and Components (EPC) Guide, Revision 1. Specifically, design engineering personnel failed to use a safety factor when determining the time (remaining service life) it would take for the 2TEC5A-header degradation to fall below the minimum wall limit. This evaluation resulted in no corrective actions being taken for cracks found on the header which eventually failed during the operating cycle. A temporary patch was installed at power that did not provide sealing under all operating conditions which later resulted in a loss of condenser vacuum causing the operators to manually scram the reactor.
Description:
On August 31, 2018, while shutting down Unit 2 to perform repairs to the B reactor recirculation pump seals, the control room operators manually scrammed Unit 2 due to a loss of the main condenser vacuum. The loss of vacuum was caused by a leak in a temporary patch installed on the 2TEC5A-header, a return line to the main condenser, which allowed air in-leakage to the main condenser while shutting the unit down.
Four weeks earlier, operators noted a significant increase in Unit 2 off gas flow. During investigation of the off gas flow increase, the licensee discovered a steam leak from a three-inch hole in the 2TEC5A-header. To address the leak, the licensee installed a temporary patch that was sufficient to return off gas flow to normal readings, but did not completely seal the steam leak from the 2TEC5A-header. The licensee determined that during the plant shutdown, conditions in the 2TEC5A-header changed such that the incomplete seal formed by the temporary patch allowed air leakage into the main condenser, which led to the loss of vacuum and manual reactor scram.
During refueling outage L2R15 in 2015, an ultrasonic test inspection (UT) was performed on the 2TEC5A header that indicated wall thinning was occurring. Corrective actions were not required at the time and the licensee would examine the area again at the next available opportunity.
Additionally, the licensee ordered the required materials for possible replacement of the header in the next refueling outage.
During refueling outage L2R16 in 2017, as required by the station's erosion in piping and components program, an ultrasonic test inspection (UT) was performed on the 2TEC5A header based on the previously detected wall thinning. The UT inspection identified additional wall thinning, cracks in the piping and that the header was leaking in this area. This was documented in action request (AR) 03975569.
Design engineering personnel reviewed AR 03975569 and believed that the source of the leaks emanated from a weld defect and not axial and radial cracks in the bottom of the header.
Design engineering personnel did not review the non-destructive examination report that documented the wall thinning and contained a sketch and detailed description of cracks in the bottom of the header; however, programs engineering personnel were aware of the documented conditions and failed to ensure that design engineering personnel knew of the conditions.
Due to not receiving all the required parts to replace the header, design engineering was requested to perform a new allowable minimum wall thickness evaluation for the header with the intent to determine if the remaining wall thickness was sufficient to operate for another 2 years before replacement. Design engineering personnel calculated a new minimum wall thickness and determined that the remaining life, when the minimum wall thickness would be reached, for the header as 7 years with the assumption of a linear wear rate.
The group that normally performs remaining life calculations, programs engineering, was not informed of the new allowable minimum wall thickness. Design engineering personnel did not complete the remaining life calculation in accordance with station procedure ER-AA-430-1004, Erosion in Piping and Components (EPC) Guide. Specifically, design engineering personnel did not take into account that the header had been replaced in 2005 and that a correction factor of 1.5 should have been used to account for a non-linear wear rate. Had the calculation been performed in accordance with the station procedure the result would have been a remaining life of 1.9 years. The licensee stated that based on that result a repair would have been performed.
Corrective Action: As an interim corrective action, the licensee installed a more suitable patch while performing the B recirculation pump repairs prior to returning to full power and then replaced the header in the most recent unit 2 refueling outage in 2019, L2R17.
Corrective Action Reference: Action Request AR 04169146
Performance Assessment:
Performance Deficiency: The inspectors determined that the licensees failure to follow station procedure ER-AA-430-1004, Erosion in Piping and Components (EPC) Guide, was a performance deficiency. Specifically, design engineering's evaluation failed to use a safety factor as required by station procedure ER-AA-430-1004, step 4.7.2.2 when determining the time it would take for the 2TEC5A-header degradation to fall below the minimum wall limit. This evaluation resulted in no corrective actions being taken for cracks found on the header which eventually failed during the operating cycle. A temporary patch was installed at power that did not provide sealing under all operating conditions which later resulted in a loss of condenser vacuum causing the operators to manually scram the reactor.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Initiating Events cornerstone. Specifically, the failure to take corrective actions, in accordance with station procedure ER-AA-430-1004, after finding cracks in header 2TEC5A which resulted in a three inch hole in the header requiring a patch and which led to a loss of condenser vacuum causing the operators to manually scram the reactor.
Significance: The inspectors assessed the significance of the finding using Appendix A, Significance Determination of Reactor Inspection Findings for At - Power Situations. The finding was screened against the Initiating Events cornerstone and determined to be of very low safety significance (Green) because the reactor trip it did not result in the loss of mitigation equipment relied upon to transition the plant from the onset of a trip to a stable shutdown condition.
Cross-Cutting Aspect: P.2 - Evaluation: The organization thoroughly evaluates issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, programs engineering personnel should have verified design engineering was made aware of the existing conditions on the 2TEC5A header so that the evaluation was thorough and addressed the actual conditions identified in the non-destructive testing report.
Enforcement:
Inspectors did not identify a violation of regulatory requirements associated with this finding.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
- On April 10, 2019, the inspector presented the quarterly integrated inspection results to Mr. W. Trafton and other members of the licensee staff.
DOCUMENTS REVIEWED
71111.04Equipment Alignment
- 1E-2-4000CT; Key Diagram 480V MCCs 235X-1 (2AP71E) and 235X-2 (2AP72E);
Revision AN
- 1E-2-4220BD; Schematic Diagram Residual Heat Removal System RH (E12) Part 28;
Revision P
- 1E-2-4220CN; Schematic Diagram Residual Heat Removal System RH (E12) Part 61;
Revision K
- LOP-RH-04E; Unit 2 Residual Heat Removal System Electrical Checklist; Revision 16
- LOP-RH-04E; Unit 2 Residual Heat Removal System Electrical Checklist; Revision 15
- LOP-RH-2BM; Unit 2 Residual Heat Removal System Mechanical Checklist; Revision 4
- M-134; P&ID CSCS Equipment Cooling Water System; Revision BG
- M-142; P&ID Residual Heat Removal System (R.H.R.S.); Revisions BC, AZ, and AE
71111.05AQFire Protection Annual/Quarterly
- Doc 01883278; 00097170-01, TSTR, Fire Damper Visual Inspection, W: L00-LMS-PF-22-
Week11; 11/21/2017
- Doc 01944066; 00097161-01, TSTR, Fire Damper Visual Inspection, W: L00-LMS-PF-22-
Week03; 7/24/2018
- FZ 2H3; LaSalle County Generating Station Pre-Fire Plan; RHR Heat Exchanger B Cubicle
694'6"
- FZ 3I3; LaSalle County Generating Station Pre-Fire Plan; RX Bldg. 673--4 Elev. U2 RHR
Pump B & C Cubicle
- FZ 4E4; LaSalle County Generating Station Pre-Fire Plan; Unit 2Division 2 Essential
Switchgear Room 731'0"
- FZ 5D2; LaSalle County Generating Station Pre-Fire Plan; Unit 2HPCS Switchgear Zone
687'0"
- WO 1931093-01; EWP-MM-Fire Damper Visual Inspection LMS-FP-22 Att. P1; 9/24/2018
- WO 1944066-01; EWP-MM-Fire Damper Visual Inspection LMS-FP-22 Att. D; 7/24/2018
- WO 4575190-01; EM LES-FP-05 Ionization Fire Detector Test (Unit 2); 7/21/2018
71111.08Inservice Inspection Activities
- AR; Ultrasonic Indication in RPV Vert Weld LCS-2-BA; 02/16/17
- AR 4221690; ISI PT Inspection Indication on 2B Reactor Recirc Pump Lug; 02/19/19
- AR 4222405; NRC ID Calc L-004189 Unit 2 RPV Vertical Weld Flaw Eval; dated 2/19/19
- AR 4222601; VT Exam Results on RR Supports RR00-2060X and 2061X; 02/22/19
- AR 4224272; NRC ID: Question on Calc L-004189; 02/27/19
- AR 4224307; NRC ID: QUESTION ON RR00-2060X/2061X; 02/27/19
- Calculation No. L-002916; Jet Pump Assembly Weld Flaw Evaluation for LaSalle Unit 2;
Revision 2
- Calculation No. L-004189; LaSalle Unit 2 RPV Vertical Weld Flaw Evaluation; Revision 0
- Engineering Change (EC) 618338; Evaluate Indication In Reactor Vessel Vertical Weld
(IR 3975049) And In Jet Pump Weld RS-1c During L2R16; Revision 0
- Procedure No. ER-AA-335-003; Magnetic Particle (MT) Examination; Revision 8
- Procedure No. GEH-PDI-UT-1; PDI Generic Procedure for the Ultrasonic Examination of
Ferritic welds; Revision 12
- Procedure No. GEH-UT-300; Procedure for Manual Examination of Reactor Vessel Assembly
Welds in Accordance with PDI; Revision 12
- Work Order 01937934; Repair/ Replace 2C41-F304 as Required; 02/12/2017
71111.15Operability Determinations and Functionality Assessments
- 1E-1-4000DW; Key Diagram 480VAC MCC 136x-3 (1AP81E); Revision P
- 1E-1-4005CJ; Schematic Diagram Auxiliary Power System AP Part 57; Revision C
- 1E-1-4009AM; Schematic Diagram Diesel Generator 1A Generator/Engine Control System
DG Part 12; Revision L
- 1E-1-4009AQ; Schematic Diagram Diesel Generator System DG; Revision K
- 3.8; Electric Power Systems; Amendment No. 172/158
- 3.8.1-4; AC SourcesOperating; Amendment 194/181
- 3.8.1-4; AC Sources-Operating
- 3.8.1-1; Electrical Power Systems; Amendment No. 172/158
- 3.8.1-2; LCO 3.0.4.b is Not Applicable to DGs; Amendment No. 171/157
- AP-1; AC Distribution Training Diagram; Revision 0
- AR 1141298; CDBI, Fast Bus Transfer of 4KV Buses
- AR 4135381; 1VX05C Didnt Shut Off When Handswitch Taken to Stop
- AR 4136375; New Eaton Contactor Found Bound in the Energized State; 1/11/2019
- AR 1141298; CDBI, Fast Bus Transfer of 4KV Buses
- AR 4199873; 1A DG Oil Circ Pump Did Not Stop When C/S Taken to Off
- AR 4200198; 1A DG Soak Back Pump Did Not Stop When C/S Taken to Off
- AR 4216536; NRC Identified: Bolt Missing on 1B DG Exhaust Ducting
- AR 4217910; Replace Bolts and Nuts on EDG Exhaust Manifold Adapters
- AR 4220430; OPEVAL 18-001 Did Not Include DG Loading Considerations
- AR 4222277; NRC IDSupport 2MS042779X Strut is Offset From Clamp
- AR 4222709; Support 2MS04-2779X Strut Offset from Clamp (IR 4222277)
- AR 4224103; RM Discovered Within RX
- AR 4228241; Low Power Tip Set Not Completed Due to Tips OOS
- Drawing IT-7000-M-PS-04; Installation Tolerances Pipe Supports; Revision B
- EC 627315; Lost Parts EvaluationLost Parts Evaluation for L2R17; Revision 1
- EC 627557-000; L2C18 BOC Startup Nonadaptive Power Limit; Revision 0
- IR 4221979; Risk Assessment: ODM: FM Found in Jet Pump #16; 3/8/2019
- L-0036997; GNF S-0000-0142-0455, Analysis of Flow Blockage Consequences for LaSalle
Units 1 and 2 GNF2 New Fuel Introduction; Revision 0
- LSCS-UFSAR 8.3-3; Unit Class 1E A-C Power System; Revision 14
- LSCS-UFSAR 8.3; Unit Class 1E A-C Power System; Revision 14
- NEDO-10174; Consequences of a Postulated Flow Blockage Incident in a Boiling Water
Reactor; June 1970
- OPEVAL 18-001, EC 397773; 1A DG Soak Back Pump and 1A DG Circulating Oil Pump
Degraded, Nonconforming or Unanalyzed Condition; Revision 00
- TRM Appendix J, Amendment 72; Technical Requirements Manual Reload Licensing
Submittal / Core Operating Limits Report; Revision 16
- Operations Log; 2/7/2019 -2/8/2019
- Drawing M09-MS04-2779X; Main Steam (Sargent & Lundy); 4/29/1981
- VTIP J-0155; Diesel, Engine Maintenance Manual, Turbocharger; Undated
71111.19Post Maintenance Testing
- IR 4225511; PCRA: LOP-DG-09M Update 2DG006 Throttle Position; 3/1/2019
- LOP-HP-08; HPCS System Full Flow Test Operation; Revision 3
- Operations Log Search 2g33-f004 or rwc; 2/25/2019-3/8/2019
- Operations Log Search 3.5.2; 2/18/2019-3/8/2019
- PMID/RQ LOS-HP-M1 Unit 2 HPCS Operability Test; 3/2/2019
- WO 1809653-01; Integrated Division III ECCS Response Time; 2/21/2019
- WO 1810246-01; Integrated Division II ECCS Response Time; 3/3/2019
- WO 1885722-07; Replace U2 SRM A Dry Tube; 2/26/2019
- WO 4773112-05; EPN 2C51-N001D, Replace SRM D Detector;2/26/2019
- WO 4773112-07; Replace U2 SRM D Dry Tube per Program Engineering; 2/26/2019
- WO 4633636-23; Need WO to Replace 2VY03A Cooler Headers in L2R17; 3/1/2019
- WO 4633535-24; Need WO to Replace 2VY03A Cooler Headers in L2R17; 3/1/2019
- WO 4889667-01; LOS-NR-W1 U2 SRM Not-Full-In Rod Block Att 2A; 2/26/2019
- WO 4604621-01; 2B DG Start and Load Acceptance; 2/19/2019
- WO 4627480-08; Work Request for Gate ValvePart 21; 2/21/2019
- WO 4610925-02; Unit 2 2C51-N001 Test SRMs; 2/24/2019
- WO 4896236-02; 2C Tip Obstructed During Handcranking; 3/14/2019
- WO 4610925-03; Repair/Replace SRM Connectors; 2/26/2019
- WO 4627480-01; Work Request for Gate ValvePart 21; 2/28/2019
- WO 4627480-10; Work Request for Gate ValvePart 21; 2/28/2019
- WO 4640162-01; LOS-SC-R1 U2 SBLC Injection Test Att 2A; 2/28/2019
- WO 4627480-04; Work Request for Gate ValvePart 21; 3/1/2019
- WO 4627480-09; Work Request for Gate ValvePart 21; 3/1/2019
- WO 4627480-19; Work Request for Gate ValvePart 21; 3/1/2019
- WO 933907-21; 2E12-F336B: Inspect/Replace/Refurb Valve; 3/4/2019
- WO 937394-17; 2DG011: Inspect / Replace / Refurb Valve; 3/1/2019
71111.20Refueling Outage
-
- AR 4117757; RM1B33-F067B Vent Line Leak
- AR 4220430; Opeval 18-001 Did Not Include DG Loading Considerations
- AR 4220852; 2D RHRWS Pump Trip While Attempting to Start Pump
- AR 4220921; 2A RR Seal Trends During L2R17 Downshift
- AR 4221067; RM-L2R17 CR 26-35 Missing Position 40
- AR 4221091; Solenoid for 2B21-F032B Blowing Air
- AR 4221168; 2A IRM Spiked high Giving a Half Scram on a RPS
- AR 4221242; LLRT: Leakage on D Main Steam Line Exceeded Warning Limits
- AR 4221406; 2B21-F032A LLRT Above Alarm Limit
- AR 4221579; LLRT: Leakage on 2B21-F032A Exceeded Warning Limits
- AR 4221643; 2VQ031, 32, 40 LLRT Above Alarm Limit
- AR 4222277; NRC IDSupport M09-MS04-2779X Strut is Offset from Clamp
- AR 4223178; 2C SRM Inoperable Hard Upscale
- AR 4223178; 2C SRM Inoperable Hard upscale
- AR 4223349; U2 SRM C Signal Cable Cut Identified
- AR 4223660; L2R17 IVVI Shroud Head Bolt Indexing Pin Weld Crack
- AR 4223693; 1PL78J 02 Reading 1% Lower than 1PL76J and 1PL77J
- AR 4223771; CTHi Pot Testing B Phase of Generator Failed
- AR 4224052; New SRM Detector (Serial # 15D00KZU) Damaged As Found
- AR 4224082; WHR Fatigue Assessment/Waiver
- AR 4224106; NRC Question on 2DG27-20005X Lug Orientation
- AR 4224509; WHR Fatigue Assessment/Waiver
- AR 4225510; WHR Fatigue Assessment/Waiver
- AR 4225739; Piping Support 2LP02-2801C Failed ISI Inspection
- IR 4221979; Loss of Integrity Notification and Recovery Plan, U2 RX Vessel Foreign Material;
2/20/2019
- L2R17, OU-AA-103/104 Shutdown Safety Plan Review Board Actions; 12/18/2018
- L2R17, OU-AA-103/104 Shutdown Safety Plan; Revision 2
- LOS-RD-SR7; Channel Interference Monitoring; Revision 26
- Reactor Vessel Temperature Pressure Log; 3/8/2019
- WO 4832092; Scope Change Request (Items Deferred) 2B33-F3364B Handwheel Broke Off;
2/26/2019
- WO 1691124; Scope Change Request 2DG01P Replace with Stainless Steel Pump; 3/8/2019
- WO 1039048; Scope Change Request 2512-F050B Disassemble and Inspect Check Valve;
2/26/2019
- WO 4697130; Scope Change Request Replace the C LP Turbine Hood Spray Piping;
3/8/2019
- WO 4828720; Scope Change Request 2C11-D004 Commuter and Changeover Valves Do
Not Operate; 3/8/2019
- WO 4610323; Scope Change Request LES-NB-201B U-2 Div II ADS Relay Logic Test;
3/8/2019
- WO 4613792-01; Scope Change Request Main Generator Bump Test; 3/18/2019
- WO 4634160; Scope Change Request ANI Vessel Inspections and Relief Valve
Replacements; 3/19/2019
- WO 4816800; Scope Change Request Replace Bonnet Vent Valves on 2B33-F067A 1B;
3/8/2019
71111.22Surveillance Testing
- LMS-DG-01; Main Emergency Diesel Unit Surveillances (WO 1875298-01 Excerpt);
Revision 57
- LOP-DG-02; Attachment H, Hard Card for Restoration of DGs to Standby During Testing for a
LOOP; Revision 67
- LOS-DG-M3; Tech Spec Surveillance, Unit 1 HPCS DG Surv. 1B-Idle; 2/3/2019
- LOS-DG-M3; 1B (2B) Diesel Generator Operability Test; Revision 99
- LOS-RI-Q5; Tech Spec Surveillance; Unit 1 RCIC Cold-Quick Start (Wk 6) Attachment 1A;
2/10/2019
- LOP-DG-Q3; 1B(2B) Diesel Generator Auxiliaries Inservice Test; Revision 76
- Operations Log; Search 1B; 2/1/2019 -2/4/2019
- WO 4607966-01; RHR SDC PIV 2E12-F050A High Pressure Water Leak Test; 2/23/2019
- WO 461089-02; Scram Time Rods per LOS-RD-SR12 & Flush Rods at 00 or 48; 3/10/2019
- WO 4598642-01; LLRT, 2B21-F016, 2B21-F019; 2/20/2019
- WO 4627480-02; Work Request for Gate ValvePart 21; 2/23/2019
- WO 4627480-03; Work Request for Gate ValvePart 21; 3/1/2019
- WO 4611607-01; Reactor Vessel Leakage Test; 3/8/2019
71124.01 Radiological Hazard Assessment and Exposure Controls
- 71124.02 ALARA Planning and Controls
- ALARA Package for RWP LA-02-19-00405; L2R17 RB/DW TIP System Maintenance and
Hand Cranking; Revision 00
- ALARA Package for RWP LA-02-19-00506; L2R17 DW Scaffold Activities; Revision 00
- ALARA Package for RWP LA-02-19-00513; L2R17 DW Control Rod Drive (CRD)
Activities/Exchange; Revision 00
- ALARA Package for RWP LA-02-19-00518; L2R17 DW In Service Inspections (ISI) Preps;
Revision 00
- ALARA Package for RWP LA-02-19-00701; L2R17 Suppression Pool Diving; Revision 00
- ALARA Package for RWP LA-02-19-00805; L2R17 Turbine Building Sandblasting Activities;
Revision 00
- ALARA Package for RWP LA-02-19-00901; L2R17 Reactor Disassembly and Reassembly
Activities; Revision 00
- ALARA Package for RWP LA-02-19-00906; L2R17 Cavity Decontamination Activities;
Revision 00
- AR 04169612; L2M20: Radiological Conditions on B RR Seal; 09/02/2018
- AR 04169868; L2M20; DW Elevated Radiological Conditions Identified; 09/01/2018
- AR 04174710; RPAlpha and Air Sampling Program MRM Completed; 09/18/2018
- AR 04217520; NRC ID: Digital Plant Viewer Issue; 02/05/2019
- NISP-RP-05; Access Controls for High Radiation Areas; Revision 0
- Radiological Surveys of LaSalle Unit 2 RHR Alpha Corner Room and Upper Raceway; Various
Dates 07/25/2018 - 01/07/2019
- RP-AA-201; Dosimetry Issue, Usage and Control; Revision 29
- RP-AA-4002; Radiation Protection Refuel Outage Readiness; Revision 11
- RP-AA-401; Operational ALARA Planning and Controls; Revision 24
- RP-AA-401-1002; Radiological Risk Management; Revision 11
- RP-AA-403; Administration of the Radiation Work Permit; Revision 10
- RP-AA-460; Controls for High and Locked High Radiation Areas; Revision 34
- RP-AA-800; Long Term Source Inventory; 09/24/2018
- RP-AA-800; Source Inventory; Revision 8; 09/04/2018
- RWP-LA-02-19-00405; L2R17 RB/DW TIP System Maintenance and Hand Cranking;
Revision 00
- RWP-LA-02-19-00513; L2R17 DW Control Rod Drive (CRD) Activities/Exchange;
Revision 00
- RWP-LA-02-19-00805; L2R17 Turbine Building Sandblasting Activities; Revision 00
- RWP-LA-02-19-00901; L2R17 Reactor Disassembly and Reassembly Activities; Revision 00
- RWP-LA-02-19-00906; L2R17 Cavity Decontamination Activities; Revision 00
71153Follow-Up of Events and Notices of Enforcement Discretion
- 1E-1-4089AA; Drawing (Wiring Diagram), RHR Pump A Cubicle Cooler Fan 1VY01C
- AR 4024815; 0DG01P Tripped After Attempting to Shut Down Pump
- AR 4024815; Document ECAP for 0DG01P Tripped After Attempting to Shut Down Pump
- CR 4024815; 0DG01P Tripped After Attempting to Shut Down Pump, Corrective Action
Program Evaluation Report; 8/3/2017
- Investigation Review for AR 4024815; 0DG01P Tripped After Attempting to Shut Down
Pump; 8/1/2018
- LER 2017-007-00;Low Pressure Core Spray System Inoperable due to Loss of Cooling;
8/17/2017
- LOP-RH-25; RHR Full Flow Test Operation; Revision 7
- OP-AA-106-101-1005; Quarantine of Areas, Equipment and Records; Revision 2
- PI-AA-120; Issue Identification and Screening Process; Revision 8
- Core Standby Cooling SystemsTraining Diagram
- PI-AA-125-1003; Corrective Action Program Evaluation Manual; Revision 4
2