IR 05000259/2022012
| 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:
Report Numbers:
05000259/2022012, 05000260/2022012 and 05000296/2022012
Enterprise Identifier:
I-2022-012-0011
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
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 Aspect Report Section Barrier Integrity Green NCV 05000259/2022012-01 Open/Closed None (NPP)71153 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 Browns Ferry Nuclear Plant Unit 1, Pressure Boundary Leak on Residual Heat Removal System Low Pressure Coolant Injection Test Line 71153 Closed LER 05000259/2022-001-00 LER 2022-001-00 from Browns Ferry Nuclear Plant,
Unit 1 regarding Pressure Boundary Leak on Residual Heat Removal System Low Pressure Coolant Injection Test Line 71153 Closed
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 Aspect Report Section Barrier Integrity Green NCV 05000259/2022012-01 Open/Closed None (NPP)71153 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
Procedure
Type
Designation
Description or Title
Revision or
Date
Corrective Action
Documents
Condition Reports (CRs)
1775692, 1775693, 1774140, 1773954, 1745874,
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
Various
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
NCR-886
TVA returned motor
Rev. 0
22 NRC PI&R Inspection, the inspector provided an
observation based on the review of CR 1642929 (U2
"B" RPS half scrams). The technicians were not
interviewed during the investigation to see if any
insight could be gained on the cause of the bent
contact arm on the relay.
08/25/2022
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.
519' battery for cordless drill sitting on support steal
approx. 8' off ground
08/09/2022
CR 1795429
RHR Walkdown Observation: 2A RHR motor scaffold
on south side of motor is touching motor and conduit.
08/09/2022
CR 1795433
1A RHR Motor - North side - scaffold has approx. 1/2"
gap from motor.
08/09/2022
CR 1795810
NRC notification attached to CR1766933 (NRC Form
361) appears to be blank.
08/09/2022
CR 1796216
Procedural enhancements could be made that caution
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
08/11/2022
CR 1798172
Several interviewees (mainly recent new hires) were
not aware of the TVA Employee Concerns Program
(ECP) and that the ECP was a option for raising safety
concerns
08/22/2022
Corrective Action
Documents
Resulting from
Inspection
CR 1798174
During interviews with Security personnel, two items
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
08/23/2022
Engineering
EDN0248920111
250V DC Non-1E Main Battery, Chargers, Inverter and
Rev. 86
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Regulating Transformer Sizing
EDN0248920112
250 V DC Non-1E Main Battery System Voltage Drop
Calculation
Rev. 46
EDQ024820020042
250V DC Unit Batt Load Study, VD, SC, and Batt
Capacity for LOCA/LOOP, Station Blackout and NFPA
805 Analysis for Unit/shutdown Board Battery
Rev. 94
EDQ0999890047
Cable Ampacity Study-V4 and V5 cables in conduit,
V3 cables fed from Panel 9-9 and Battery Boards 1, 2,
and 3; Panel 8 and 9, and other saftey related V3
Cables
Rev. 82
EDQ2000870028
20V Voltage Drop Calculation
Rev. 87
Evaluations
MDQ0009992012000094
Rev. 7
OE Screening Committee Meeting: Covering OE from
2/15/20 to 12/28/20
2/30/2020
OE Screening Committee Meeting: Covering OE from
06/22/21 to 06/28/21
06/30/2021
OE Screening Committee Meeting: Covering OE from
2/14/21 to 12/20/21
01/26/2022
OE Screening Committee Meeting: Covering OE from
07/12/22 to 07/19/22
07/20/2022
Func 064A-B, 2-CKV-
076-0653
(a)(1) Plan
Rev. 3
Func 073-B &C, HPCI
(a)(1) Plan
Rev. 9
Func 099-B, RPS
(a)(1) Plan
Rev. 4
Miscellaneous
Func 573-B, Main Bank
Battery 5
(a)(1) Plan
Rev. 2
Operability
Evaluations
Operability
Determination
1649288, 1781455, 1678680, 1637517
Various
3-AOI-6-1
Feedwater Heater String/Extraction Steam Isolation
Rev. 0007
MCI-0-001-VL002
Main Steam Relief Valves Target Rock Model 7567
Disassembly, Inspection, Repair and Reassembly
Rev. 0056
NEDP-22
Operability Determinations and Functional Evaluations
Rev. 0022
NEDP-27
Past Operability Evaluations
Rev. 0007
Procedures
NPG-SPP-01.16
Condition Report Initiation
Rev. 0006
Inspection
Procedure
Type
Designation
Description or Title
Revision or
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
SA for CR 1670754
21 Maintenance & Technical Training
Comprehensive Assessment
08/02/2021
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
Nuclear Plant
06/11/2021
SSA2104
Radiation Protection / Radiological Waste Browns
Ferry Nuclear Plant
07/08/2021
SSA2107
Corrective Action Program
2/05/2021
Self-
Assessments
SSA2204
Maintenance
06/04/2022
Work Orders
Work Orders (WO)
2142871, 122066471, 122872922, 121649338,
21718721, 122205985, 120561853, 121384897,
2866305, 121724417, 122071657, 122071658,
2071659, 122071660, 121788877, 121851803,
119961886, 121798832, 122026684, 122075517,
2076896, 122092952, 121665266, 122565105
Various
98-011374-000
Crack in body seat drain weld for 3-FCV-068-003
identified during the performance of 3-SI-3.3.1.A,
ASME Section XI System Leakage Test
10/14/1998
Corrective Action
Documents
PER 961217
Two failures identified on the RHR System I Testable
Check Valve (3-FCV-074-0054) piping.
08/09/1996
Browns Ferry Nuclear Plant Nuclear Safety Culture
Evaluation
June 2018
Browns Ferry Nuclear Plant Nuclear Safety Culture
Evaluation
October
2019
2702
CONF
Miscellaneous
Seventh Independent Auditors Report of the TVA
2/26/2019
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Adverse Employment Action Process for Semester
Ending 12/31/2019
Eighth Independent Auditors Report of the TVA
Adverse Employment Action Process for Semester
Ending 6/30/2020
06/26/2020
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