IR 05000259/2022012

Biennial Problem Identification and Resolution Inspection Report 05000259 2022012, 05000260 2022012 and 05000296 2022012

ML22262A060
Person / Time
Site:	Browns Ferry
Issue date:	09/20/2022
From:	Wesley Deschaine NRC/RGN-II/DRP/RPB5
To:	Jim Barstow Tennessee Valley Authority
References
IR 2022012
Download: ML22262A060 (15)

Text

SUBJECT:

BROWNS FERRY NUCLEAR PLANT - BIENNIAL PROBLEM IDENTIFICATION AND RESOLUTION INSPECTION REPORT 05000259/2022012, 05000260/2022012 AND 05000296/2022012

Dear Mr. Barstow:

On August 26, 2022, the U.S. Nuclear Regulatory Commission (NRC) completed a problem identification and resolution inspection at your Browns Ferry Nuclear Plant and discussed the results of this inspection with Mr. Manu Sivaraman and other members of your staff. The results of this inspection are documented in the enclosed report.

The NRC inspection team reviewed the stations problem identification and resolution program and the stations implementation of the program to evaluate its effectiveness in identifying, prioritizing, evaluating, and correcting problems, and to confirm that the station was complying with NRC regulations and licensee standards for problem identification and resolution programs.

Based on the samples reviewed, the team determined that your staffs performance in each of these areas adequately supported nuclear safety.

The team also evaluated the stations processes for use of industry and NRC operating experience information and the effectiveness of the stations audits and self-assessments.

Based on the samples reviewed, the team determined that your staffs performance in each of these areas adequately supported nuclear safety.

Finally, the team reviewed the stations programs to establish and maintain a safety-conscious work environment, and interviewed station personnel to evaluate the effectiveness of these programs. Based on the teams observations and the results of these interviews the team found no evidence of challenges to your organizations safety-conscious work environment. Your employees appeared willing to raise nuclear safety concerns through at least one of the several means available.

The Regional Administrator has authorized an additional non-baseline inspection using Inspection Procedure 92702, Follow-up on Traditional Enforcement Actions Including Violations, Deviations, Confirmatory Action Letters, Confirmatory Orders, and Alternative Dispute Resolution Confirmatory Orders to verify the licensees implementation of the 2017 Confirmatory Order (CO) requirements because this follow-up inspection is an infrequent inspection and outside of the routine reactor oversight process (ROP) baseline inspections.September 20, 2022 One finding of very low safety significance (Green) is documented in this report. This finding involved a violation of NRC requirements. We are treating this violation as a non-cited violation (NCV) consistent with Section 2.3.2 of the Enforcement Policy.

If you contest the violation or the 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 II; the Director, Office of Enforcement; and the NRC Resident Inspector at Browns Ferry Nuclear Plant.

This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding.

Sincerely, Wesley D. Deschaine, Chief Reactor Projects Branch 5 Division of Reactor Projects Docket Nos. 50-259, 50-260, and 296 License Nos. DPR-33, DPR-52, and DPR-68

Enclosure:

As stated

Inspection Report

Docket Numbers: 05000259, 05000260 and 05000296

License Numbers: DPR-33, DPR-52 and DPR-68

Report Numbers: 05000259/2022012, 05000260/2022012 and 05000296/2022012

Licensee: Tennessee Valley Authority

Facility: Browns Ferry Nuclear Plant

Location: Athens, AL

Inspection Dates: August 08, 2022 to August 26, 2022

Inspectors: D. Hardage, Senior Resident Inspector N. Karlovich, Resident Inspector N. Peterka, Senior Project Engineer A. Ponko, Sr. Construction Inspector

Approved By: Wesley D. Deschaine, Chief Reactor Projects Branch 5 Division of Reactor Projects

Enclosure

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting a biennial problem identification and resolution inspection at Browns Ferry Nuclear Plant, in accordance with the Reactor Oversight Process. The Reactor Oversight Process is the NRCs program for overseeing the safe operation of commercial nuclear power reactors. Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information.

List of Findings and Violations

Pressure Boundary Leak on Residual Heat Removal (RHR) Low Pressure Coolant Injection (LPCI) Test Line Cornerstone Significance Cross-Cutting Report Aspect Section Barrier Integrity Green None (NPP) 71153 NCV 05000259/2022012-01 Open/Closed A Green self-revealing non-cited violation (NCV) of Technical Specification (TS) 3.4.4 was identified when the licensee failed to apply corrective actions from previous fatigue failures of American Society of Mechanical Engineers (ASME) Code Class 1 equivalent socket welded connections resulted in a through-wall piping leak on a test line upstream of the RHR and Shutdown Cooling test shut-off valve.

Additional Tracking Items

Type Issue Number Title Report Section Status LER 05000259/2022-001-01 LER 2022-001-01 for 71153 Closed Browns Ferry Nuclear Plant Unit 1, Pressure Boundary Leak on Residual Heat Removal System Low Pressure Coolant Injection Test Line LER 05000259/2022-001-00 LER 2022-001-00 from 71153 Closed Browns Ferry Nuclear Plant,

Unit 1 regarding Pressure Boundary Leak on Residual Heat Removal System Low Pressure Coolant Injection Test Line

INSPECTION SCOPES

Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.

OTHER ACTIVITIES - BASELINE

71152B - Problem Identification and Resolution Biennial Team Inspection (IP Section 03.04)

(1) The inspectors performed a biennial assessment of the effectiveness of the licensees Problem Identification and Resolution program, use of operating experience, self-assessments and audits, and safety conscious work environment (SCWE).

Corrective Action Program Effectiveness: The inspectors assessed the corrective action program's effectiveness in identifying, prioritizing, evaluating, and correcting problems. The inspectors also conducted an in-depth corrective action program review of the Residual Heat Removal (RHR)

System, RHR Service Water System, and the Control Room Emergency Ventilation System.

Operating Experience: The inspectors assessed the effectiveness of the licensees processes for use of operating experience,

Self-Assessments and Audits: The inspectors assessed the effectiveness of the licensees identification and correction of problems identified through audits and self-assessments.

Safety Conscious Work Environment: The inspectors assessed the effectiveness of the stations programs to establish and maintain a safety-conscious work environment.

71153 - Follow Up of Events and Notices of Enforcement Discretion Event Report (IP Section 03.02)

The inspectors evaluated the following licensee event reports (LERs):

(1) LER 50-259/2022-001-00 and -01 - Pressure Boundary Leak on Residual Heat Removal System Low Pressure Coolant Injection Test Line. (ADAMS Accession Nos.

ML E220531t052550 and E220316t044020). The inspection conclusions associated with this LER are documented in this report under Inspection Results Section.

OTHER ACTIVITIES

- TEMPORARY INSTRUCTIONS, INFREQUENT AND ABNORMAL

===92702 CONF - Enforcement Related Order Follow-Up-Only

The Regional Administrator has authorized an additional non-baseline inspection using Inspection Procedure 92702, Follow-up on Traditional Enforcement Actions Including Violations, Deviations, Confirmatory Action Letters, Confirmatory Orders, and Alternative Dispute Resolution Confirmatory Orders to verify the licensees implementation of the 2017 Confirmatory Order (CO) requirements because this follow-up inspection is an infrequent inspection and outside of the routine reactor oversight process (ROP) baseline inspections. The inspectors reviewed commitments associated with two items from Confirmatory Order EA-17-022, issued to TVA on July 27, 2017.

Enforcement Related Order Follow-Up-Only===

(1) Commitment V.1.d.1: This commitment required an independent third-party to perform quarterly audits for the first year after the date of issuance of the Confirmatory Order (CO), and semi-annually for the next two years, of the adverse employment action process. The inspectors reviewed the Seventh Independent Auditors Report of the TVA Adverse Employment Action Process for Semester Ending 12/31/2019, dated 12/26/2019 and the Eighth Independent Auditors Report of the TVA Adverse Employment Action Process for Semester Ending 6/30/2020, dated 6/26/2020. These are the final two semi-annual audits for Browns Ferry. The inspectors verified that the audits included a review of all adverse employment actions, periodical attendance at Executive Review Boards and a review of chilling effect mitigation plans (inclusive of recommendations as appropriate). The inspection of the audits was completed for Browns Ferry only.

The inspectors also performed a review of three recent Adverse Employment Action and Executive Review Board (ERB) packages at Browns Ferry. Specifically, the inspectors reviewed the background information and safety conscious work environment (SCWE) screening provided in the packages along with any associated SCWE Mitigation Plans.

Commitment V.1.e.1: This commitment required TVA to conduct an independent nuclear safety culture (NSC) assessment consistent with industry standard practices at Browns Ferry within one year of issuance of the CO, and another NSC assessment within approximately two years of the first assessment. TVA was further required to evaluate the results and develop, implement, and track to completion corrective actions to address weaknesses identified through the assessments and compare the results of the latter assessment with prior years survey results in an effort to identify trends. The inspectors reviewed the Browns Ferry Nuclear Plant NSC evaluations completed by Oak Ridge Associated Universities in June 2018 and October 2019. The inspectors verified if corrective actions were identified, developed and tracked to completion. The inspectors also verified if the results of the latter assessment were compared with prior years survey results in an effort to identify trends. The inspection was completed for Browns Ferry only.

INSPECTION RESULTS

Pressure Boundary Leak on Residual Heat Removal (RHR) Low Pressure Coolant Injection (LPCI) Test Line Cornerstone Significance Cross-Cutting Report Aspect Section Barrier Integrity Green None (NPP) 71153 NCV 05000259/2022012-01 Open/Closed A Green self-revealing non-cited violation (NCV) of Technical Specification (TS) 3.4.4 was identified when the licensee failed to apply corrective actions from previous fatigue failures of American Society of Mechanical Engineers (ASME) Code Class 1 equivalent socket welded connections resulted in a through-wall piping leak on a test line upstream of the RHR and Shutdown Cooling test shut-off valve.

Description:

On January 15, 2022, at 2320 Central Standard Time (CST), during a drywell entry for leak identification, Browns Ferry Nuclear Plant (BFN) Engineering personnel discovered a through-wall piping leak on a test line upstream of the RHR and Shutdown Cooling test shut-off valve. This test line is classified as ASME Code Class 1 piping and constitutes part of the Unit 1 Reactor Coolant System (RCS) pressure boundary. The leak was located just upstream of 1-SHV-074-0794A where the socket weld for the valve and the socket weld for the upstream sockolet overlapped. This leak was determined to be unisolable from the reactor pressure vessel (RPV). The licensee made an event notification in accordance with 10 CFR 50.72(b)(3)(ii)(A) - Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded. Operations personnel declared the Unit 1 LPCI system Loop I inoperable and maintained Unit 1 in Mode 4 or 5 until the leak was repaired.

This condition is assumed to have developed in September of 2021, when system monitoring detected a step change in unidentified drywell leakage. The licensee decided to repair the detected leak during a maintenance outage that was scheduled due to a previously identified fuel leak. The leaking RHR piping was identified during initial drywell entry.

Corrective Actions: A temporary modification was implemented to remove the test valves and vent piping to resolve leakage until addressed during the next Unit 1 outage. The test line was cut and capped and on January 20, 2022, at 1520 CST, Unit 1 LPCI Loop I was declared operable. The root cause of this event was small bore piping which was not specifically analyzed for fatigue failure vulnerability due to operational or resonance vibration. The corrective action for this event is to implement Engineering Change Packages for all small-bore piping with vulnerability to fatigue failure due to exceeding the endurance limit due to operational vibration.

Corrective Action References: CR 1747875

Performance Assessment:

Performance Deficiency: The licensees failure to apply corrective actions from fatigue failures of ASME Code Class 1 equivalent socket welded connections in the Units 2 and 3 drywells on RHR shutdown cooling lines and the 3A Recirculation Loop was a performance deficiency (PD). Specifically, the failures were addressed under Problem Evaluation Report (PER) 961217 and an extent of condition was performed under PER 98-011374-000. The corrective actions developed were not incorporated during the Unit 1 restart in 2007.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the RCS Equipment and Barrier Performance attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the inadequate application of corrective actions from previous fatigue failures of ASME Code Class 1 equivalent socket welded connections resulted in an un-isolable through wall leak in the test line upstream of the RHR and Shutdown Cooling test shut-off valve and was identified as RCS pressure boundary leakage.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The finding screened out in the review of the Barrier Integrity cornerstone as the PD was not related to pressurized thermal shock; therefore, the finding was addressed under the Initiating Events cornerstone. Since a reasonable assessment of degradation, could have resulted in exceeding the RCS leak rate for a small loss of coolant accident (LOCA), a detailed risk evaluation was performed by a regional Senior Risk Analyst (SRA) in accordance with IMC 0609, Appendix A, utilizing the Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) 8 Version 8.2.6, and the NRC Browns Ferry Unit 1 Standardized Plant Analysis Risk (SPAR) model version 8.61 dated 6/12/2019. The exposure period was from September 11, 2021, when indications of the leak were present until January 15 when the plant was taken to mode 4, a total of 126 days. The PD was conservatively modelled as an increase in the small LOCA frequency by two orders of magnitude given the leak upstream of 1-SHV-074-0794A had the potential to cause the socket to fail and initiate an unisolable small break loss of coolant accident if the entire socket failed. The dominant sequence was a Small Break Loss of coolant accident with failure of the turbine bypass valves, failure of high-pressure injection and operators failing to manually depressurize the reactor. The detailed risk evaluation estimated that the PD resulted in an increase in core damage frequency of < 1.0 E-6/year, a GREEN finding of very low safety significance.

Cross-Cutting Aspect: Not Present Performance. No cross-cutting aspect was assigned to this finding because the inspectors determined the finding did not reflect present licensee performance.

Enforcement:

Violation: TS LCO 3.4.4 requires, in part, operational LEAKAGE shall be limited to: a. No pressure boundary LEAKAGE; while in Modes 1, 2, and 3; otherwise, the unit shall be shut down and in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and Mode 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Contrary to the above, from approximately September 11, 2021, to January 15, 2022, while Unit 1 was in Mode 1 with an uninsolable pressure boundary leak present, the licensee operated in a condition prohibited by TSs because the unit had not been placed in Mode 3 and Mode 4.

Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.

EXIT MEETINGS AND DEBRIEFS

The inspectors verified no proprietary information was retained or documented in this report.

On August 26, 2022, the inspectors presented the biennial problem identification and resolution inspection results to Mr. Manu Sivaraman and other members of the licensee staff.

DOCUMENTS REVIEWED

Inspection Type Designation Description or Title Revision or

Procedure Date

71152B Corrective Action Condition Reports (CRs) 1775692, 1775693, 1774140, 1773954, 1745874, Various

Documents 1767368, 1784571, 1774143, 1785100, 1778716,

1782024, 1769257, 1794453, 1687343, 1698900,

1679626, 1680257, 1680682, 1769257, 1747875,

1642929, 1663776, 1747824, 1712139, 1712139,

1716499, 1700879, 1690583, 1612226, 1621239,

20905, 1699093, 1736015, 1640160, 1740513,

1769605, 1733846, 1705816, 1663775, 1640732,

1659001, 1779311, 17223229, 1776857, 1689113,

1699014, 1699015, 1736017, 1689250, 1740838,

27098, 1651234, 1639467, 1629973, 1660188,

1780920, 1692907, 1778527, 1777520, 1776697,

1783784, 1781532, 1781093, 1743378, 1736780,

1735693, 1725900, 1718130, 1715560, 1707957,

1706925, 1705189, 1688751, 1670628, 1660073,

1657879, 1715213, 1681154, 1651335, 1777288,

1743599, 1661852, 1788589, 1771842, 1751010,

1676114, 1640515, 1770005, 1723754, 1715260,

1687954, 1636300, 1768085, 1689190, 1696769,

1699014, 1649097, 1642037, 1660079, 1660775,

1434505, 1477448, 1504345, 1522472, 1540693,

1540684, 1540695, 1548744, 1553670, 1575596,

1578086, 1590864, 1628393, 1645959, 1649674,

1661315, 1664063, 1677968, 1681376, 1682390,

1695271, 1698453, 1696877, 1705020, 1705816,

1712139, 1712629, 1716190, 1716499, 1737479,

1737669, 1738803, 1739373, 1739548, 1745754,

1745983, 1755306, 1754502, 1754835, 1756427,

1756432, 1756448, 1759180, 1756456, 1766933,

1783069, 1784017, 1678515, 1788309, 1777794,

1766776, 1760216, 1755932, 1693940, 1679209,

1667993, 1634753, 1794696, 1784017, 1773876

Inspection Type Designation Description or Title Revision or

Procedure Date

NCR-886 TVA returned motor Rev. 0

Corrective Action 2022 NRC PI&R Inspection, the inspector provided an 08/25/2022

Documents observation based on the review of CR 1642929 (U2

Resulting from "B" RPS half scrams). The technicians were not

Inspection interviewed during the investigation to see if any

insight could be gained on the cause of the bent

contact arm on the relay.

CR 1795391 No indication above 0-XS-031-2201 as expected 08/10/2022

CR 1795425 3A Torus Suction Valve BFN-3-FCV-074-0001 on elev. 08/09/2022

519' battery for cordless drill sitting on support steal

approx. 8' off ground

CR 1795429 RHR Walkdown Observation: 2A RHR motor scaffold 08/09/2022

on south side of motor is touching motor and conduit.

CR 1795433 1A RHR Motor - North side - scaffold has approx. 1/2" 08/09/2022

gap from motor.

CR 1795810 NRC notification attached to CR1766933 (NRC Form 08/09/2022

361) appears to be blank.

CR 1796216 Procedural enhancements could be made that caution 08/11/2022

against too much examination of failed components if

there will be a chance that the component could be

send to the vendor for examination including Part 21

considerations

CR 1798172 Several interviewees (mainly recent new hires) were 08/22/2022

not aware of the TVA Employee Concerns Program

(ECP) and that the ECP was a option for raising safety

concerns

CR 1798174 During interviews with Security personnel, two items 08/23/2022

were brought up that were characterized as long

standing personnel safety issues that were not

resolved in a timely manner. 1) Flashing light on gate

N2 and, 2) Difficult operation of the Delta gates in

manual

Engineering EDN0248920111 250V DC Non-1E Main Battery, Chargers, Inverter and Rev. 86

Inspection Type Designation Description or Title Revision or

Procedure Date

Evaluations Regulating Transformer Sizing

EDN0248920112 250 V DC Non-1E Main Battery System Voltage Drop Rev. 46

Calculation

EDQ024820020042 250V DC Unit Batt Load Study, VD, SC, and Batt Rev. 94

Capacity for LOCA/LOOP, Station Blackout and NFPA

805 Analysis for Unit/shutdown Board Battery

EDQ0999890047 Cable Ampacity Study-V4 and V5 cables in conduit, Rev. 82

V3 cables fed from Panel 9-9 and Battery Boards 1, 2,

and 3; Panel 8 and 9, and other saftey related V3

Cables

EDQ2000870028 120V Voltage Drop Calculation Rev. 87

MDQ0009992012000094 Rev. 7

Miscellaneous OE Screening Committee Meeting: Covering OE from 12/30/2020

2/15/20 to 12/28/20

OE Screening Committee Meeting: Covering OE from 06/30/2021

06/22/21 to 06/28/21

OE Screening Committee Meeting: Covering OE from 01/26/2022

2/14/21 to 12/20/21

OE Screening Committee Meeting: Covering OE from 07/20/2022

07/12/22 to 07/19/22

Func 064A-B, 2-CKV- (a)(1) Plan Rev. 3

076-0653

Func 073-B &C, HPCI (a)(1) Plan Rev. 9

Func 099-B, RPS (a)(1) Plan Rev. 4

Func 573-B, Main Bank (a)(1) Plan Rev. 2

Battery 5

Operability Operability 1649288, 1781455, 1678680, 1637517 Various

Evaluations Determination

Procedures 3-AOI-6-1 Feedwater Heater String/Extraction Steam Isolation Rev. 0007

MCI-0-001-VL002 Main Steam Relief Valves Target Rock Model 7567 Rev. 0056

Disassembly, Inspection, Repair and Reassembly

NEDP-22 Operability Determinations and Functional EvaluationsRev. 0022

NEDP-27 Past Operability Evaluations Rev. 0007

NPG-SPP-01.16 Condition Report Initiation Rev. 0006

Inspection Type Designation Description or Title Revision or

Procedure Date

NPG-SPP-01.7.1 Employee Concerns Program Rev. 0006

NPG-SPP-03.5 Regulatory Reporting Requirements Rev. 0017

NPG-SPP-22.103 Performance Assessment and Monitoring Rev. 0015

NPG-SPP-22.300 Corrective Action Program Rev. 0024

NPG-SPP-22.500 Operating Experience Program Rev. 0015

NPG-SPP-22.600 Issue Resolution Rev. 0013

Self-SA for CR 1670754 2021 Maintenance & Technical Training 08/02/2021

Assessments Comprehensive Assessment

SA for CR 1688503 Electrical Maintenance Training Program 07/20/2021

SA for CR 1721138 Dose and Dose Rate Alarm Review Self Assessment 11/15/2021

SSA2002 Security and Safeguards Information (SGI) 01/30/2020

SSA2102 Operations Audit 05/13/2021

SSA2103 Chemistry and Environmental Monitoring Browns Ferry 06/11/2021

Nuclear Plant

SSA2104 Radiation Protection / Radiological Waste Browns 07/08/2021

Ferry Nuclear Plant

SSA2107 Corrective Action Program 02/05/2021

SSA2204 Maintenance 06/04/2022

Work Orders Work Orders (WO) 122142871, 122066471, 122872922, 121649338, Various

21718721, 122205985, 120561853, 121384897,

2866305, 121724417, 122071657, 122071658,

2071659, 122071660, 121788877, 121851803,

119961886, 121798832, 122026684, 122075517,

2076896, 122092952, 121665266, 122565105

71153 Corrective Action 98-011374-000 Crack in body seat drain weld for 3-FCV-068-003 10/14/1998

Documents identified during the performance of 3-SI-3.3.1.A,

ASME Section XI System Leakage Test

PER 961217 Two failures identified on the RHR System I Testable 08/09/1996

Check Valve (3-FCV-074-0054) piping.

2702 Miscellaneous Browns Ferry Nuclear Plant Nuclear Safety Culture June 2018

CONF Evaluation

Browns Ferry Nuclear Plant Nuclear Safety Culture October

Evaluation 2019

Seventh Independent Auditors Report of the TVA 12/26/2019

Inspection Type Designation Description or Title Revision or

Procedure Date

Adverse Employment Action Process for Semester

Ending 12/31/2019

Eighth Independent Auditors Report of the TVA 06/26/2020

Adverse Employment Action Process for Semester

Ending 6/30/2020

BFN-2019-013 Employee Review Board Package 10/08/2019

BFN-2020-05 Employee Review Board Package 01/08/2020

BFN-2020-13 Employee Review Board Package 04/17/2020

2