IR 05000259/2019002

NRC Integrated Inspection Report 05000259/2019002, 05000260/2019002, 05000296/2019002 and 07200052/2019001

ML19224B824
Person / Time
Site:	Browns Ferry
Issue date:	08/12/2019
From:	Omar Lopez-Santiago NRC/RGN-II, Division Reactor Projects II
To:	James Shea Tennessee Valley Authority
Reeder D
References
IR 2019001, IR 2019002
Download: ML19224B824 (29)

Text

{{#Wiki_filter:ust 12, 2019

SUBJECT:

BROWNS FERRY NUCLEAR PLANT - NRC INTEGRATED INSPECTION REPORT 05000259/2019002, 05000260/2019002, 05000296/2019002 AND 07200052/2019001

Dear Mr. Shea:

On June 30, 2019, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Browns Ferry Nuclear Plant. On August 2, 2019, the NRC inspectors discussed the results of this inspection with Mr. Lang Hughes and other members of your staff. The results of this inspection are documented in the enclosed report.

The NRC inspectors documented one finding of very low safety significance (Green) in this report. This finding did not involve a violation of NRC requirements. Additionally, two Severity Level IV violations without an associated finding are documented in this report. We are treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2.a of the Enforcement Policy.

The inspectors documented a licensee-identified violation which was determined to be of very low safety significance in this report. The NRC is treating this violation as a NCV consistent with Section 2.3.2.a of the Enforcement Policy.

If you contest the violations or significance or severity of the violations 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.

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 II; and the NRC Resident Inspector at Browns Ferry. 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/ Omar R. López-Santiago, Chief Reactor Projects Branch 5 Docket Nos. 05000259, 05000260, 05000296 and 07200052 License Nos. DPR-33 and DPR-52 and DPR-68 Enclosure: As stated cc w/ encl: Distribution via LISTSERV

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated 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. Licensee-identified non-cited violations are documented in report sections: 71114.04.

List of Findings and Violations Improper Operation of Automatic Voltage Regulator Results in Reactor SCRAM Cornerstone Significance Cross-Cutting Report Aspect Section Initiating Events Green [H.12] - Avoid 71153 FIN 05000296/2019002-01 Complacency Closed A self-revealed Finding of Browns Ferry procedure 3-OI-47 "Turbine-Generator System" was identified for the failure to properly operate the Unit 3 Automatic Voltage Regulator (AVR)while adjusting the reactive load on the main generator in order to lower grid voltage. Specifically, the licensee switched the AVR from Automatic to Manual and then took on too much reactive load causing a trip of the main turbine.

Units 1, 2 and 3, Standby Gas Treatment System Train C Inoperable Longer Than Allowed by Technical Specifications Cornerstone Significance Cross-Cutting Report Aspect Section Not Applicable Severity Level IV Not Applicable 71153 NCV 05000259, 260, 296/2019002-02 Closed A self-revealed Severity Level (SL) IV Non-Cited Violation (NCV) of Technical Specification (TS) 3.6.4.3, Standby Gas Treatment (SGT) System, was identified when the licensee determined that the SGT C subsystem was inoperable for longer than the 7 day allowed outage time following its failure to automatically start during a manual scram of Unit 3.

Unit 1, Main Steam Relief Valves Lift Settings Outside of Technical Specifications Required Setpoints Cornerstone Significance Cross-Cutting Report Aspect Section Not Applicable Severity Level IV Not Applicable 71153 NCV 05000259/2019002-03 Closed A self-revealed SL IV NCV of TS 3.4.3 was identified when the licensee discovered, through as found test results that two of the thirteen MSRVs that were removed during the Fall 2018 Unit 1 refueling outage had as found lift settings outside of the +/- 3 percent band required for their operability.

Additional Tracking Items Type Issue Number Title Report Section Status LER 05000259/2018-002- Automatic Reactor Scram 71153 Closed 01 Due to an Unanticipated Electro-Hydraulic Control Logic Condition.

LER 05000259/2018-007- Main Steam Relief Valves Lift 71153 Closed 00 Settings Outside of Technical Specifications Required Setpoints.

LER 05000260,05000296,0 Standby Gas Treatment 71153 Closed 5000259/2018-006-00 System Train C Inoperable Longer Than Allowed by Tech.

LER 05000296/2019-001- Automatic Reactor Scram 71153 Closed 01 Due to a Turbine Load Reject.

PLANT STATUS

Unit 1 operated at or near Rated Thermal Power (RTP) until June 1, 2019, when power was reduced to 69 percent of RTP for maintenance on the scram discharge volume. Power was returned to 100 percent RTP on June 2, and the unit remained at rated thermal power for the rest of the inspection period.

Unit 2 began the inspection period in a refueling outage. The unit was restarted on April 9 and followed their approved Extended Power Uprate (EPU) power ascension schedule until April 26, 2019, when power was reduced to 60 percent RTP for turbine stop valve (TSV)maintenance. Power returned to 87 percent RTP on April 26, and power ascension continued until June 4, when power was reduced to 50 percent RTP for Reactor Protection System (RPS)maintenance. Power was raised to 92 percent RTP on June 7, and power ascension continued until 100 percent RTP was reached on June 22.

Unit 3 began the inspection period at RTP. Power was reduced to 65 percent RTP on June 15 for a rod sequence exchange, and power was restored the same day. The unit remained at or near RTP for the rest 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.

71111.01 - Adverse Weather Protection Summer Readiness Sample (IP Section 03.01)

(1) The inspectors evaluated summer readiness of offsite and alternate alternating current (AC) power systems on June 12, 2019 through June 25, 2019. This is one of the three required inspections to close out URI 05000296/2019001-04 Unit 3 Notice of Unusual Event (NOUE) Caused by of a Loss of Offsite Power (LOOP) Resulting in a Reactor Scram.

71111.04 - Equipment Alignment Partial Walkdown Sample (IP Section 03.01)

The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:

(1) Unit 3 Emergency Diesel Generators (EDGs) on April 12, 2019

(2) Unit 1 and 2 B, C and D EDGs on May 6, 2019 while A EDG was out of service for maintenance

(3) Unit 3 Standby Liquid Control system on May 7, 2019

(4) Units 1, 2 and 3, Standby Gas Treatment (SGT) system on May 16, 2019 with 3D EDG was out of service and A and B SGT trains were protected

(5) Unit 3 EDGs on May 23, 2019 following maintenance on the 3D EDG

(6) Unit 2 Hardened Containment Vent System on May 24, 2019

(7) Unit 3, Control Rod Drive Hydraulic System on June 13, 2019

(8) Unit 2 Reactor Vessel Level Instrumentation System (RVLIS) on June 24, 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 3, Fire Area 16-O, Auxiliary Instrument Room 3 on April 17, 2019

(2) Units 1 and 2, Fire Area 16-A, Cable Spreading Room A on May 15, 2019

(3) Units 1, 2 and 3, Fire area 16-N, Computer room, Communications room, Communications battery and battery board rooms on June 20, 2019

(4) Units 1 and 2, Fire area 04, 4kV shutdown board room 1B on June 20, 2019

(5) Unit 3, Fire Area 03-03, Elevation 565'-593', Residual Heat Removal (RHR) heat exchanger rooms on June 21, 2019

71111.06 - Flood Protection Measures Inspection Activities - Internal Flooding (IP Section 02.02a.)

The inspectors evaluated internal flooding mitigation protections in the:

(1) Unit 3, Reactor Building 565 foot elevation on June 21, 2019

==71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01) (1 Sample)

(1) The inspectors observed and evaluated licensed operator performance in the Control Room during:

• Unit 2 Start-up activities on April 8 - 9, 2019
• Unit 2 testing and power ascension on April 24, May 13, June 18-19 and June 22, 2019
• Unit 2 EPU flow induced vibration data acquired at 91.9, 96.2, and 100 percent power Licensed Operator Requalification Training/Examinations (IP Section 03.02)==

(1) The inspectors observed and evaluated quarterly requalification training on May 20, 2019

71111.12 - Maintenance Effectiveness Routine Maintenance Effectiveness Inspection (IP Section 02.01)

The inspectors evaluated the effectiveness of routine maintenance activities associated with the following equipment and/or safety significant functions:

(1) Units 1, 2 and 3, Function 068-B, Anticipated Transient Without Scram (ATWS)-

Recirculation Pump Trip (RPT) System

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 and 2, Elevated risk due to EDG A nearing the beginning of the extended allowable outage time, May 7, 2019

(2) Unit 3, elevated risk due to Unit 3 HPCI being inoperable during a 3D EDG maintenance outage, May 16, 2019

(3) Units 1, 2 and 3, elevated risk due to maintenance outage of Main Battery Bank 1 during Conservative Operations Alert (COA) on June 27, 2019

71111.15 - Operability Determinations and Functionality Assessments Operability Determination or Functionality Assessment (IP Section 02.02)

The inspectors evaluated the following operability determinations and functionality assessments:

(1) Units 1, 2 and 3, Past operability evaluation associated with CR

===1440310, SGT system train B low flow condition

(2) Unit 3, Past operability evaluation associated with failure of RHR Loop 1 outboard injection valve 74-52, which failed to open as required for shutdown cooling on March 10, 2019

(3) Units 1, 2 and 3, Functionality Evaluation due to nozzle clogging of the deluge system for the Main bank transformers, Unit Station Service Transformers (USSTs), and Common Station Service Transformers (CSSTs) on April 5, 2019

(4) Unit 3, Operability and functionality evaluation associated with a suspected leak in 3C EDG cooling water heat exchanger on April 4, 2019

(5) Unit 2, Operability evaluation of the Reactor Protection System (RPS) with turbine stop valve #3 indicating closed on April 25, 2019

(6) Unit 3, Operability evaluation of control rod hydraulic control unit 22-03 after indicating low pressure with no alarm on May 1, 2019

(7) Unit 1, Operability evaluation associated with failure of Reactor Core Isolation Cooling (RCIC) steam admission valve to fully open or close during a surveillance on June 3, 2019

(8) Unit 1, Operability evaluation associated with failure of one input to RVLIS, which led to unplanned LCO entries on June 25, 2019

==71111.18 - Plant Modifications Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02) (1 Sample) The inspectors evaluated the following temporary or permanent modifications:

(1) Unit 2, change in RHR heat exchanger requirements due to Extended Power Uprate (EPU) conditions

71111.19 - Post-Maintenance Testing Post Maintenance Test Sample (IP Section 03.01)

[[IP sample::=

=IP 71111.18|count=9}}

The inspectors evaluated the following post maintenance tests:

(1) Unit 3 High Pressure Coolant Injection (HPCI) Gland Seal Condenser pump motor test following motor replacement, March 22, 2019

(2) Unit 2 HPCI turbine overspeed trip test following 10-year turbine maintenance, April 3, 2019

(3) Unit 2 Main Steam Relief Valves (MSRV) Manual Cycle Surveillance following pilot valve replacement on all MSRVs, April 9, 2019

(4) Unit 2 HPCI pump developed head and flow rate test at rated pressure following maintenance, April 9, 2019

(5) Units 1 and 2 EDG A operability test with large load reject following maintenance, May 8, 2019

(6) Unit 3, 3D EDG post maintenance testing following preventative maintenance on May 23, 2019

(7) Various Unit 2 integrated systems testing per 2-TI-700 at EPU conditions from April 17 to June 30 when the unit achieved the new licensed operating power of 3952 MWt

(8) Unit 1, Replacement of RPS voter #4 power supply on April 9, 2019

(9) Unit 2, RCIC System Rated Flow at Normal Operating Pressure on April 9, 2019

71111.20 - Refueling and Other Outage Activities Refueling/Other Outage Sample (IP Section 03.01)

(1) The inspectors evaluated the completion of Unit 2 refueling outage activities from April 1 to April 9, 2019. The inspection of the earlier portion of the outage was documented in the first quarter integrated inspection report Browns Ferry 05000259, 260, 296/2019-001.

71111.22 - Surveillance Testing The inspectors evaluated the following surveillance tests: FLEX Testing (IP Section 03.02)

(1) Units 1, 2 and 3, Reviewed inspection, maintenance and testing activities associated with the Flex portable turbine generators that was completed on March 25, 2019 Inservice Testing (IP Section 03.01) ===

(1) Units 1, 2 and 3 ASME Section XI System Pressure Test of North Header of the Emergency Equipment Cooling Water (EECW) System, WO 118062407 and

===120143309

(2) Units 1, 2 and 3 Residual Heat Removal Service Water (RHRSW) Pump D2 IST quarterly surveillance on May 14, 2019

(3) Units 1, 2, 0-SI-4.5.C.1(A2), RHRSW Pump A2 IST Group A Quarterly Pump Test, on April 11, 2019 Surveillance Tests (other) (IP Section 03.01) ===

(1) Units 1, 2 and 3 E Field AND X Field Detection System Calibration And Sensitivity Performance Tests on June 20, 2019

71114.01 - Exercise Evaluation Inspection Review (IP Section 02.01-02.11)

(1) The inspectors evaluated the biennial emergency plan exercise during the week of June 24, 2019. The exercise scenario simulated a flammable gas release that affected a 4kV (kilovolt) electrical room and led to the declaration of an Alert. A pre-existing fuel rod leak was exacerbated when three control rods near the leaker failed to insert upon a manual turbine and reactor trip that was initiated due to main turbine vibrations rising. Then the reactor water cleanup system failed to isolate, thus creating conditions that met the threshold for declaring a Site Area Emergency. As the reactor coolant chemistry increased to meet the simulated conditions for a General Emergency, this allowed the Offsite Response Organizations to demonstrate their ability to implement emergency actions.

71114.04 - Emergency Action Level and Emergency Plan Changes Inspection Review (IP Section 02.01-02.03)

(1) The inspectors evaluated submitted Emergency Action Level, Emergency Plan, and Emergency Plan Implementing Procedure changes during the week of June 24, 2019. This evaluation does not constitute NRC approval.

71114.06 - Drill Evaluation Drill/Training Evolution Observation (IP Section 03.02)

The inspectors evaluated:

(1) Integrated Training Drill on May 1, 2019

71114.08 - Exercise Evaluation Scenario Review Inspection Review (IP Section 02.01 - 02.04)

(1) The inspectors reviewed and evaluated in-office, the proposed scenario for the biennial emergency plan exercise at least 30 days prior to the day of the exercise.

OTHER ACTIVITIES - BASELINE

===71151 - Performance Indicator Verification The inspectors verified licensee performance indicators submittals listed below: EP01: Drill/Exercise Performance (IP Section 02.12) ===

(1) EP01: Drill & Exercise Performance for the period January 1, 2018, through March 31, 2019

EP02: ERO Drill Participation (IP Section 02.13) (1 Sample)

(1) EP02: Emergency Response Organization Drill Participation for the period January 1, 2018, through March 31, 2019 EP03: Alert & Notification System Reliability (IP Section 02.14) (1 Sample)

(1) EP03: Alert & Notification System Reliability for the period January 1, 2018, through March 31, 2019 MS05: Safety System Functional Failures (SSFFs) Sample (IP Section 02.04) (3 Samples)

(1) Unit 1 (April 1, 2018 - March 31, 2019)

(2) Unit 2 (April 1, 2018 - March 31, 2019)

(3) Unit 3 (April 1, 2018 - March 31, 2019)

MS07: High Pressure Injection Systems (IP Section 02.06) (3 Samples)

(1) Unit 1 (April 1, 2018 - March 31, 2019)

(2) Unit 2 (April 1, 2018 - March 31, 2019)

(3) Unit 3 (April 1, 2018 - March 31, 2019)

MS08: Heat Removal Systems (IP Section 02.07) (3 Samples)

(1) Unit 1 (April 1, 2018 - March 31, 2019)

(2) Unit 2 (April 1, 2018 - March 31, 2019)

(3) Unit 3 (April 1, 2018 - March 31, 2019)

71152 - Problem Identification and Resolution Annual Follow-up of Selected Issues (IP Section 02.03)

The inspectors reviewed the licensees implementation of its corrective action program related to the following issues:

(1) The inspectors conducted safety culture interviews with non-licensed operators, reactor operators, and senior reactor operators as a result of several safety culture related condition reports. The inspectors concluded the operators understood the various methods of reporting safety concerns and were willing to use them without fear of retaliation.

Semiannual Trend Review (IP Section 02.02) (1 Sample)

(1) The inspectors reviewed the licensees corrective action program for potential adverse trends in transient combustible control, seismic restraints, and maintenance rework that might be indicative of a more significant safety issue.

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

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

(1) LER 05000259/2018-007-00, Main Steam Relief Valves Lift Settings Outside of Technical Specifications Required Setpoints (ADAMS accession: ML1942A158):

The circumstances surrounding this LER are documented in the Results section.

The inspectors determined that it was not reasonable to foresee or correct the cause discussed in the LER therefore no performance deficiency was identified.

(2) LER 05000259/2018-006-00, Standby Gas Treatment System Train C Inoperable Longer Than Allowed by Technical Specifications (ADAMS accession:

ML18340A152): The circumstances surrounding this LER are documented in the Results section.

The inspectors determined that it was not reasonable to foresee or correct the cause discussed in the LER therefore no performance deficiency was identified.

(3) LER 05000296/2019-001-00, Automatic Reactor Scram Due to Turbine Load Reject (ADAMS accession: ML19128A102):

The circumstances surrounding this LER are documented in the Results section.

This completes one of the three required inspections to close out URI

===05000296/2019001-04 Unit 3 Notice of Unusual Event (NOUE) Caused by of a Loss of Offsite Power (LOOP) Resulting in a Reactor Scram.

(4) LER 05000259/2018-002-01, Automatic Reactor Scram Due to an Unanticipated Electro-Hydraulic Control Logic Condition (ADAMS accession: ML19087A244):

The inspectors reviewed the updated LER submittal. No new violations were provided in this LER. The previous LER submittal was reviewed in Inspection Report 05000259, 260, 296/2018-002.

OTHER ACTIVITIES

- TEMPORARY INSTRUCTIONS, INFREQUENT AND ABNORMAL 60855.1 - Operation of an Independent Spent Fuel Storage Installation at Operating Plants 1. Operation of an Independent Spent Fuel Storage Installation at Operating Plants (1 Sample) The inspectors evaluated the licensees activities related to long-term operation and monitoring of their independent spent fuel storage installation.

The inspectors evaluated the following change reviews:

• 0-SR-DCS3.1.2.1, Spent Fuel Storage Inspection
• MSI-0-079-DCS035, Dry Cask Campaign Guidelines
• NPG-SPP-05.8, Special Nuclear Material Control 60853 - Onsite Fabrication of Components and Construction of an Independent Spent Fuel Storage Installation 1. Onsite Fabrication of Components and Construction of an Independent Spent Fuel Storage Installation ===

The inspectors conducted a review of licensee and vendor activities in preparation for the concrete placement for the second section of the second Independent Spent Fuel Storage Installation (ISFSI) pad. The licensee will use the HI-STORM (Holtec International Storage Module) FW (Flood and Wind) System on the pad. The inspectors walked down the construction area of the ISFSI pad and examined the rebar installation, for compliance to licensee-approved drawings, specifications, procedures, and applicable codes, the Certificate of Compliance (CoC), and Technical Specifications (TSs). The inspectors also evaluated the concrete formwork installation for compliance to the licensee-approved drawings. The inspectors interviewed licensee and contract personnel to verify knowledge of the planned work. The inspectors also observed the actual concrete placement for the second section of the ISFSI slab and observed testing and sample collection by the independent laboratory to verify that the work was implemented according to approved specifications and procedures. Later, when the 28-day compression tests were completed by the independent laboratory, the inspectors reviewed the results to verify that the acceptance criteria were met. The inspectors noted that all tested samples satisfied the acceptance criteria.

INSPECTION RESULTS

Licensee-Identified Non-Cited Violation 71114.04 This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.

Violation: From July 2018 through September 2018, the licensee failed to maintain the effectiveness of its emergency plan by not updating Emergency Action Level (EAL) radiation monitor threshold values due to the Extended Power Uprate (EPU). As a result of an extent of condition review (from the Watts Bar Apparent Violation regarding its radiation monitors), TVA discovered that the primary containment radiation EAL values for the loss of Fuel Clad Barrier, RCS Barrier and potential loss of the Containment Barrier for the Unit 3 drywell monitors were lower than what they should have been in the current EPU state. Also, the licensee failed to identify parameters for several different radiation monitors that have input to EAL RU1 (Release of gaseous liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer) initiating condition that were not updated appropriately during the transition from NUMARC Revision 3 EALs to the NEI Revision 6 EALs, thus rendering the RU1 EAL ineffective.

Therefore, contrary to 10 CFR 50.54(q)(2) and 10 CFR 50.47(b)(4), BFN failed to maintain the effectiveness of its emergency plan by not updating EAL radiation monitor threshold values due to the Extended Power Uprate, thus rendering the affected EAL ineffective.

Significance: Green. The inspectors evaluated this issue as an ineffective EAL per IMC 0609, Appendix B, Figure 5.4-1. The inspectors screened this issue as a Green NCV (Licensee Identified) because prior to the inspection, BFN had conducted an extent of condition review and taken corrective actions. Also, given that the Initiating Condition was correct, an appropriate declaration wouldve likely been made, but in a degraded (delayed)manner.

Corrective Action References: CR 1418869, 1419082, 1429910, 1450520, 1451311, 1453905, 1453905 (Level 2 Report), 1462597.

Improper Operation of Automatic Voltage Regulator Results in Reactor SCRAM Cornerstone Significance Cross-Cutting Report Aspect Section Initiating Events Green [H.12] - Avoid 71153 FIN 05000296/2019002-01 Complacency Open/Closed A self-revealed Finding of Browns Ferry procedure 3-OI-47 "Turbine-Generator System" was identified for the failure to properly operate the Unit 3 Automatic Voltage Regulator (AVR)while adjusting the reactive load on the main generator in order to lower grid voltage. Specifically, the licensee switched the AVR from Automatic to Manual and then took on too much reactive load causing a trip of the main turbine.

Description:

On March 9, 2019, at approximately 2259 Central Daylight Time (CDT), Browns Ferry Unit 3 received an automatic Reactor SCRAM from 100 percent power as a result of a Turbine Load Reject. The automatic SCRAM occurred as a Licensed Reactor Operator (LRO)made adjustments to lower incoming reactive power on Unit 3 at the request of the Balancing Authority (BA). The request was made by the BA to address a high voltage condition on the 500 kV transmission system. During the adjustment, a LRO incorrectly operated a hand switch (HS), placing the Automatic Voltage Regulator (AVR) in manual control, and adjusted Mega Volt Amps Reactive (MVAR) beyond the Under Excitation Limiter protection setting. After two seconds, the generator excitation remained below the protection setting, which initiated a loss of excitation trip resulting in loss of the generator exciter field. Incoming MVAR and current increased on all three generator phases resulting in actuation of generator overcurrent protective relays and opening of the Unit 3 high-side breakers. The generator circuit breaker tripped causing a Turbine Load Reject and automatic SCRAM. The licensee's procedures for operation of the AVR required it to be operated in automatic per 3-OI-47 "Turbine Generator System" Section 6.1 "Normal Operation" Note 2.

Due to the electric plant lineup necessitated by outage work on Browns Ferry Unit 2, the offsite power to the 4kV shutdown boards was configured to manual transfer from 500 kV to 161 kV offsite power. The transfer of offsite power from 500 kV to 161 kV power took in excess of 15 minutes. Browns Ferry Unit 3 declared a Notification of Unusual Event (NOUE)due to loss of all offsite AC power. This finding is associated with Browns Ferry LER 50-296/2019-001-00 Automatic Reactor Scram Due to Turbine Load Reject.

Corrective Actions: A plastic cover was installed over the manual/auto select voltage regulator hand switch [HS] on Units 1, 2, and 3 to mitigate unintentional operation of the AVR from automatic to manual control. Units 1, 2, and 3 Operating Instruction procedure for the Turbine Generator System, was revised to prompt Operators to validate the initial status of the AVR prior to adjusting MVAR.

Corrective Action References: CR 1497448

Performance Assessment:

Performance Deficiency: The licensee switched the mode of the AVR from automatic to manual and took on reactive load in excess of the capacity of the Unit 3 main generator.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Protection Against External Factors attribute of the Initiating Events cornerstone. Specifically, the performance deficiency directly caused a reactor SCRAM and loss of offsite power resulting in the declaration of a NOUE.

Significance: The inspectors assessed the significance of the finding using Appendix A, Significance Determination of Reactor Inspection Findings for At - Power Situations for the initiating events cornerstone. The finding was determined to be Green because although the cause of the scram resulted in a LOOP, the licensee was able to restore offsite power and used the main condenser to complete the reactor coolant system cool-down to Mode 4.

Cross-Cutting Aspect: H.12 - Avoid Complacency: Individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Individuals implement appropriate error reduction tools. The inspectors determined that operations leaders and personnel have allowed negative human performance behaviors to exist. These behaviors include improper peer checks, stopping when unsure, and inadequate supervisory oversight.

Enforcement:

Inspectors did not identify a violation of regulatory requirements associated with this finding.

Units 1, 2, and 3, Standby Gas Treatment System Train C Inoperable Longer Than Allowed by Technical Specifications Cornerstone Severity Cross-Cutting Report Aspect Section Not Severity Level IV Not 71153 Applicable NCV 05000259, 260, 296/2019002-02 Applicable Open/Closed A self-revealed Severity Level (SL) IV Non-Cited Violation (NCV) of Technical Specification (TS) 3.6.4.3, Standby Gas Treatment (SGT) System, was identified when the licensee determined that the SGT C subsystem was inoperable for longer than the 7 day allowed outage time following its failure to automatically start during a manual scram of Unit 3.

Description:

The LER is associated with SGT C subsystem being inoperable for longer than the Unit 1, 2 and 3 TS Limiting Conditions for Operation 3.6.4.3. On October 6, 2018, SGT C subsystem failed to automatically start in response to a Primary Containment isolation signal that was received during a planned manual scram of Unit 3 for a forced outage. SGT C subsystem was manually started by operators following the failure to automatically start.

Operations declared SGT C subsystem inoperable and troubleshooting determined that the hand switch (0-HS-065-69A2) failed for SGT C subsystem fan motor. The licensee discovered that this hand switch had a broken phenolic barrier plate. The pieces of the broken barrier plate impeded switch operation and affected the continuity of the contacts.

Licensee operations staff requested a past operability evaluation (POE) for the SGT C subsystem.

This particular hand switch was previously identified (CR 1380403) as deficient on January 23, 2018 when SGT C subsystem was ran during a surveillance procedure, 2-SR-3.3.6.2.4(RX). The switch was documented as having a "catch" and would not secure the fan during several attempts. The licensee generated a work order (WO 119345859) to troubleshoot and repair/replace the hand switch. During troubleshooting, the contacts associated with the switch were verified to be in the correct position with the hand switch in the auto position. Based on this, operations determined SGT C subsystem would automatically start as required. On February 17, 2018 during the performance of logic testing for the primary containment isolation system, SGT C subsystem automatically started as required. SGT C subsystem operated successfully for plant operations up to its eventual failure on October 6, 2018.

Corrective Actions: The hand switch for SGT C subsystem was replaced with WO 119345859 and operability was restored on October 8, 2018. The licensee performed a POE to evaluate the affected safety function. The degraded switch affected the C subsystem function to automatically start, but it was capable of being manually initiated within 5 minutes of a valid signal. The licensee also verified that A and B SGT subsystems were operable during the time period evaluated. The licensee has taken actions to identify other switches affected by this condition, generated WOs to replace them and implemented guidance to inspect new switches for damage prior to installation.

Corrective Action References: CR 1454001, CR 1380403

Performance Assessment:

The NRC determined this violation was not reasonably foreseeable and preventable by the licensee and therefore is not a performance deficiency.

The inspectors determined that because the licensee performed initial trouble shooting of the switch contacts, verified that the switch contacts were in the correct position to automatically start the SGT C subsystem and successfully completed the surveillance requirement that specifically tests the primary containment isolation system to automatically start SGT in response to a simulated accident signal.

Enforcement:

The ROPs significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it is necessary to address this violation which impedes the NRCs ability to regulate using traditional enforcement to adequately deter non-compliance. Traditional Enforcement is being used because there was no performance deficiency using the Interim Guidance for Dispositioning Severity Level IV Violations with No Associated Performance Deficiencies (ML18158A220). This violation is characterized as a Severity Level IV NCV based on its similarity to a SLIV example 6.1.d.1 in the Enforcement Policy.

Violation: Browns Ferry Nuclear Plant, Unit 1, 2 and 3 TS Subsection 3.6.4.3, "Standby Gas Treatment (SGT) System", Condition A requires that, with one SGT subsystem inoperable, the subsystem must be restored to operable status within seven days. Condition B, requires, with the required action of Condition A not met in Modes 1, 2 or 3, that the applicable Unit be in Mode 3 within twelve hours and Mode 4 within thirty-six hours. Contrary to the above, SGT subsystem C was inoperable from September 27, 2018 when the switch was last operated successfully, to October 8, 2018 when the SGT subsystem C was declared operable following maintenance and the units did not complete the required actions for Conditions A and B in the required completion time.

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

Unit 1, Main Steam Relief Valves Lift Settings Outside of Technical Specifications Required Setpoints Cornerstone Severity Cross-Cutting Report Aspect Section Not Severity Level IV Not 71153 Applicable NCV 05000259, 260, 296/2019002-03 Applicable Open/Closed A self-revealed SL IV NCV of TS 3.4.3 was identified when the licensee discovered, through as found test results that two of the thirteen MSRVs that were removed during the Fall 2018 Unit 1 refueling outage had as found lift settings outside of the +/- 3 percent band required for their operability.

Description:

The LER was associated with two of the thirteen MSRVs as found setpoints. This was discovered on December 12, 2018, following as-found testing conducted on all thirteen MSRVs that were removed during the refueling outage. The licensee determined that one valve had corrosion bonding to its valve seat as a result of its platinum anti-corrosion coating flaking off. The second valve failed due to relaxation of the setpoint spring over time. The licensee determined that these two MSRVs were inoperable for an indeterminate period of time from November 2, 2016 (beginning of operating cycle), to October 14, 2018 (beginning of refueling outage). The inspectors reviewed the license event report and determined that the report adequately documented the summary of the event including the cause and potential safety consequences. The inspectors also reviewed other documents that indicate that this type of failure is a known industry issue associated with this type of valve.

Corrective Actions: The licensee replaced all thirteen of the Unit 1 MSRV pilot valves with refurbished valves during the Fall 2018 Unit 1 refueling outage. The licensee has corrective actions to ensure that pilot discs are prepared for platinum coating in accordance with the revised procedure and vendor recommendations. The currently installed refurbished valves had platinum coatings applied in accordance with the revised procedure, and as-left values were verified to be within +/- 1 percent of their setpoints.

Corrective Action References: CRs 962223, 1252419, 1475055

Performance Assessment:

The NRC determined this violation was not reasonably foreseeable and preventable by the licensee and therefore is not a performance deficiency.

Enforcement:

The ROPs significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it is necessary to address this violation which impedes the NRCs ability to regulate using traditional enforcement to adequately deter non-compliance. Traditional Enforcement is being used because there was no performance deficiency using the Interim Guidance for Dispositioning Severity Level IV Violations with No Associated Performance Deficiencies (ML18158A220). This violation is characterized as a Severity Level IV NCV based on its similarity to a SLIV example 6.1.d.1 in the Enforcement Policy.

Violation: Browns Ferry Nuclear Plant, Unit 1 TS Subsection 3.4.3, Safety/Relief Valves (S/RVs), Condition A, required that with one or more required S/RVs inoperable, that the unit be in Mode 3 within 12 hours and Mode 4 in 36 hours. Contrary to the above, two required S/RVs were inoperable from November 2, 2016, to October 14, 2018, and the unit did not enter Mode 3 and Mode 4 in 12 hours and 36 hours, respectively.

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 June 3, 2019, the inspector presented the ISFSI exit meeting to Chris Dunn, Plant Manager, and other members of the licensee staff.
• On June 28, 2019, the inspectors presented the Emergency Preparedness exit meeting to Chris Dunn, Plant Manager and other members of the licensee staff.
• On August 2, 2019, the inspectors presented the Quarterly exit meeting to Lang Hughes, Site Vice President and other members of the licensee staff.

DOCUMENTS REVIEWED

Inspection Type Designation Description or Title Revision or Procedure Date 60855.1 Corrective Action CR 1443278 and 60853 Documents CR 1499165 Procedures 0-SR-DCS3.1.2.1 Spent Fuel Storage Inspection Revision 23 MSI-0-079-DCS035 Dry Cask Storage Campaign Guidelines Revision 20 NPG-SPP-05.8 Special Nuclear Material Control Revision 11 Drawings 72564-00-117 Dry Fuel Expansion Project BFN ISFSI Pad #2, Layout/Coordinates 2564-00-118 Dry Fuel Expansion Project Expansion Project Details, Sheet 1 2564-00-119 Dry Fuel Expansion Project ISFSI Pad #2 Details, Sheet 2 Commercial Metals Browns Ferry ISFSI Project Revision 3 Rebar Fabrication Drawing R1.0 for Job # 40107362 Commercial Metals Browns Ferry ISFSI Project Revision 3 Rebar Fabrication Drawing R1.1 for Job# 40107362 Miscellaneous #CC-000149 S&ME, Inc. Concrete Compressive Strength Report 07/01/2019 2564-FCR-1 ISFSI Expansion Project: ISFSI Storage Pad #2 Revision 0 2564-FCR-2 ISFSI Expansion Project: ISFSI Storage Pad #2 Revision 0 2564-FCR-3 ISFSI Expansion Project: ISFSI Storage Pad #2 Revision 0 2564-FCR-4 ISFSI Expansion Project: ISFSI Storage Pad #2 Revision 0 2564-FCR-6 ISFSI Expansion Project: ISFSI Storage Pad #2 Revision 0 73564-FCR-5 ISFSI Expansion Project: ISFSI Storage Pad #2 Revision 0 CC-000149 S&ME, Inc. Concrete Compressive Strength Report 07/01/2019 CC-000150 S&ME, Inc. Concrete Compressive Strength Report 07/01/2019 CC-000151 S&ME, Inc. Concrete Compressive Strength Report 07/01/2019 CC-000152 S&ME, Inc. Concrete Compressive Strength Report General Engineering Plain and Reinforced Concrete Revision 8 Specification G-2 Inspection Type Designation Description or Title Revision or Procedure Date General Engineering Formwork for Concrete Revision 1 Specification G-8 General Engineering Earth and Rock Foundations and Fills during Revision Specification G-9 Construction, Modification and Maintenance for 0009 Nuclear Plants Work Orders 119823709 Excavate and Install Forms and Reinforcing Steel for ISFSI Pad #2 119823718 Install Concrete and Backfill for New ISFSI Pad 71111.01 Corrective Action CR 1521157 Documents CR 1526126 Drawings PIP-02-03 AC Electrical Distribution System, Browns Ferry 10/18/2018 Nuclear Plant Procedures 0-AOI-57-1A Loss of Offsite Power (161 and 500 KV)/Station Revision Blackout 110 0-AOI-57-1E Grid Instability Revision 11 0-GOI-200-3 Hot Weather Operations Revision 15 0-OI-57A Switchyard and 4160V AC Electrical System Revision 165 71111.04 Corrective Action Level 2 Evaluation Report for CR 1510961, June 4, Documents Unexpected RVLIS System Response During 2019 Maintenance CR 1506529 CR 1510961 CR 1513130 CR 1517153 CR 1518355 CR 1519039 CR 1519302 Drawings 0-47E820-1 Flow Diagram Control Rod Drive Hydraulic System Revision 32 0-47E865-11 Flow Diagram Heating and Ventilating Standby Gas Revision 32 Treatment System 1-47E865-1 Flow Diagram Heating and Ventilating Air Flow Revision 54 2-47E810-1 Flow Diagram Reactor Water Cleanup System Revision 43 Inspection Type Designation Description or Title Revision or Procedure Date 3-47E820-2 Flow Diagram Control Rod Drive Hydraulic System Revision 20 3-47E820-6 Flow Diagram Control Rod Drive Hydraulic System Revision 7 3-47E820-7 Flow Diagram Control Rod Drive Hydraulic System Revision 21 3-47E861-8 Flow Diagram - Cooling System & Lubrication Oil Revision 20 System Standby Diesel Generator 3D Procedures 0-GOI-300-1/ATT-9 Unit 3 Reactor Building Operator Round Logs Revision 257 0-OI-57D/ATT-3 Attachment 3 Electrical Lineup Checklist Revision 155 0-OI-65 Standby Gas Treatment System Revision 55 0-OI-82 Standby Diesel Generator System Revision 167 0-OI-82/ATT-1B Attachment 1B Standby Diesel Generator B Valve Revision Lineup Checklist 102 3-ARP-9-5B Alarm Response Procedure 3-XA-55-5B Revision 31 3-OI-82 Standby Diesel Generator System Revision 145 3-OI-85/ATT-1 Attachment 1 Valve Lineup Checklist Revision 71 3-SR-3.8.1.1(3D) Diesel Generator 3D Monthly Operability Test Revision 61 LCI-3-T-63-003 Standby Liquid Control Tank & Pump Suction Header Revision 4 Temperature NPG-SPP-10.2 Clearance Procedure to Safely Control Energy Revision 22 SII-2-F-85-763 Reactor Water Level Reference Leg Backfill System Revision 8 Work Orders WO 119604411 WO 120396810 71111.05Q Calculations MDN0009992012000100 Browns Ferry Nuclear Power Plant, Units 1, 2, and 3, Revision 8 Fire Risk Evaluations Fire Plans FPR-Volume 2 Fire Protection Report Volume 2 Revision 65 Miscellaneous NFPA 805, Fire Protection Requirements Manual Revision 9 NFPA 805 Fire Protection Report, Appendix F, Fire Revision 2 Area 03-03 NFPA 805 Fire Protection Report, Appendix F, Fire Revision 3 Area 16 Inspection Type Designation Description or Title Revision or Procedure Date Procedures NPG-SPP-18.4.7 Control of Transient Combustibles Revision 13 71111.06 Calculations NDN00099920070031 BFN Probabilistic Risk Assessment - Internal Flooding Revision 2 Analysis Drawings 3-47E858-1 Flow Diagram RHR Service Water System Revision 33 71111.11Q Procedures 2-GOI-100-1A Unit Startup and Power Operation Revision 2 2-SR-3.4.9.1(1) Reactor Heat-up and Cooldown Rate Monitoring Revision 29 71111.12 Corrective Action CR 1141448 Documents CR 1142007 CR 1308779 CR 1393646 Miscellaneous (a)(1) Plan for Function 068-B ATWS-RPT System Health Report for System 68 ( Reactor Recirculation System) Maintenance Rule Expert Panel Meeting Minutes BFN-50-7068 Reactor Water Recirculation System Revision 17 PM 500124368 Procedures 0-TI-346 Maintenance Rule Performance Indicator Monitoring, Revision 52 Trending, and Reporting - 10CFR50.65 Work Orders WO 117613302 WO 117618022 WO 118882711 WO 118882720 71111.13 Procedures BFN-ODM 4.18 Protected Equipment Revision 22 71111.15 Corrective Action Equipment Failure Investigation Checklist for CR Documents 1497471 CR 1110412 CR 1440310 CR 1459972 CR 1497471 CR 1504813 CR 1505425 CR 1507495 Inspection Type Designation Description or Title Revision or Procedure Date CR 1509558 CR 1513703 CR 1521607 CR 1522331 CR 1524013 CR 1527412 CR 1527729 Drawings 1-47E810-1 Flow Diagram Reactor Water Cleanup System Revision 20 2-47E610-1-2 Mechanical Control Diagram Main Steam System Revision 18 2-47E610-47-1 Mechanical Control Diagram Turbine Revision 11 2-47E610-47-10 Mechanical Control Diagram Turbine Revision 3 2-47E610-47-13 Mechanical Control Diagram Turbine Revision 5 2-730E915-9 Elementary Diagram Reactor Protection System Revision 27 3-45E779-22 Wiring Diagram 480V Shutdown Auxiliary Power Revision 20 Schematic Diagram 3-45E779-8 Wiring Diagram 480V Shutdown Aux Power Revision 28 Schematic Diagram 3-47E859-2 Flow Diagram Emergency Equipment Cooling Water Revision 28 3-47E861-7 Flow Diagram - Cooling System and Lubricating Oil Revision 20 System Standby Diesel Generator 3C Engineering Functionality Evaluation Documentation for CR Revision 0 Evaluations 1496533 Miscellaneous Browns Ferry Nuclear Standby Gas Treatment (SGT) System Plant Technical Specifications (TS) Section 3.6.4.3 Browns Ferry Nuclear Ventilation Filter Testing Program (VFTP) Plant Technical Specifications (TS) Section 5.5.7 FSAR Section 14.6 Analysis of Design Basis Accidents FSAR Section 5.3 Secondary Containment System Operability Past Operability Evaluation Documentation for CR Evaluations 1497471 Inspection Type Designation Description or Title Revision or Procedure Date Past Operability Evaluation Documentation for CR 21607 Past Operability Evaluation for CR 1440310 Procedures 0-TI-362(Bases) IST Program Bases Document Revision 14 1-OI-85 Control Rod Drive System Revision 46 SII-1-F-85-763 Reactor Water Level Reference Leg Backfill System Revision 6 Work Orders WO 117446675 WO 119906735 WO 120313166 WO 120389756 WO 120394634 71111.18 Calculations 18-159 Browns Ferry Nuclear Plant Unit 2 2A and 2C RHR Revision 0 Heat Exchanger Thermal Performance Test Report (3 October 2018) Procedures 0-TI-322 RHR Heat Exchanger Performance Testing Revision 5 Work Orders WO 116037712 71111.19 Calculations MDN299920030021 Condensate and Feedwater Hydraulic Model for Revision 10 Extended Power Uprate MDQ099920070022 Reactor Feedwater Pump Run-out Flow Hydraulic Revision 5 Analysis Corrective Action CR 1492943 Documents CR 1514774 CR 1514791 CR 1516625 CR 1516977 CR 1517145 CR 1517153 CR 1518340 CR 1518355 CR 1518711 Procedures 0-PMTI-82-1 (DG A) Diesel Generator A 2301A and DRU Setup & Tuning Revision 1 Instruction 0-PMTI-82-2 (DG A) Diesel Generator A Monthly Operability Test with Revision 1 Inspection Type Designation Description or Title Revision or Procedure Date Large Load Reject 2-SR-3.4.3.2 Main Steam Relief Valves Manual Cycle Test Revision 12 2-SR-3.5.1.7 HPCI Main & Booster Pump Set Developed Head & Revision 80 Flow Rate Test at Rated Reactor Pressure 2-TI-130 Main Steam Pressure Control Revision 8 2-TI-131 Feedwater Level Control System Revision 12 2-TI-700 EPU Master Startup Test Instruction Revision 2 2-TI-700/ATT-7 EPU Master Startup Test Instruction, Maximum FW Revision 0 Pump Run-out 3-OI-73 High Pressure Coolant Injection System Revision 61 3-PMTI-82-1(DG 3D) WO 119829057 PMT Diesel Generator 3D 2301A and Revision 2 DRU Setup and Tuning Instruction 3-PMTI-82-2(DG 3D) WO 119829057 PMT Diesel Generator 3D Monthly Revision 1 Operability with Large Load Reject MSI-2-073-GOV001 High Pressure Coolant Injection (HPCI) Turbine Revision 43 Overspeed Trip Test Work Orders 118428081 Unit1, Replace RPS voter #4 power supply 04/09/2019 119850912 Unit 2, RCIC System Rated Flow at Normal Operating 04/09/2019 Pressure 20346375 Unit 3 HPCI Gland Seal Condenser Pump Motor 03/18/2019 replacement WO 119017463 WO 119829057 WO 119830787 WO 120111817 71111.20 Corrective Action CR 1505244 Documents Procedures NPG-SPP-03.21 Fatigue Rule and Work Hour Limits Revision 21 71111.22 Corrective Action CR 1479434 Documents CR 1481165 CR 1515119 Drawings 0-47E839-5-ISI ASME Section XI Raw Water Chemical Treatment Revision 8 System Code Class Boundaries Inspection Type Designation Description or Title Revision or Procedure Date 1-47E858-1 Flow Diagram RHR Service Water System Revision 72 1-47E859-1-ISI Flow Diagram Emergency Equipment Cooling Water Revision 29 2-47E844-2-ISI ASME Section XI Raw Cooling Water Code Class Revision 14 Boundaries 2-47E859-1-ISI ASME Section XI Emergency Equipment Cooling Revision 14 Water System Code Class Boundaries 3-47E859-1-ISI ASME Section XI Emergency Equipment Cooling Revision 12 Water System Code Class Boundaries 3-47E859-2-ISI ASME Section XI Emergency Equipment Cooling Revision 7 Water System Code Class Boundaries Procedures 0-SI-4.5.C.1 (D2) RHRSW Pump D2 IST Group A Quarterly Pump Test Revision 8 0-SI-4.5.C.1(A2) RHRSW Pump A2 IST Group A Quarterly Pump Test Revision 10 2-SI-3.3.14.A ASME Section XI System Pressure Test of the North Revision 13 Header of the EECW System (ASME Section III Class 3) 3-SR-3.5.1.7 HPCI Main and Booster Pump Set Developed Head Revision 85 and Flow Rate Test at Rated Reactor Pressure PM Job Plan BFN-79376 825kW/480VAC CT Generator Maintenance Revision 0

                                            (Quarterly)

PM Job Plan BFN-79377 825kW/480VAC CT Generator Maintenance: 0-GEN- Revision 0 360-002B FLEX 480V Turbine Generator #2 PM Job Plan BFN-79385 1.1MW/4kVAC CT Generator Maintenance (Quarterly) Revision 0 PM Job Plan BFN-79387 1.1MW/4kVAC CT Generator Maintenance (Quarterly) Revision 0 SII-0-SEC-260-246 E Field and X Field Detection System Calibration and Revision 13 Sensitivity Performance Tests Work Orders WO 118062407 WO 119598650 WO 119603944 WO 119763675 WO 119763677 WO 119763681 WO 119763683 WO 120143309 Inspection Type Designation Description or Title Revision or Procedure Date 71114.06 Miscellaneous Browns Ferry Integrated Training Drill 05/01/2019 Procedures EPIP-1 Emergency Classification Procedure Revision 59 REP-Appendix A Browns Ferry Radiological Emergency Plan Revision 109 71151 Miscellaneous Unit 1, 2, and 3 Input records for the Heat Removal MSPI submission. 04/01/2018 to 03/31/2019 Units 1,2 and 3 Input records for the Safety System Functional Failure 04/01/2018 MSPI submission to 03/31/2019 71152 Corrective Action Site Trimester Performance Assessment 10/01/2018 Documents to 01/31/2019 CR 1482574 Emerging trend in valve rework CR 1504642 NRC Identified Trend - Continuing trend of unrestrained plant equipment CR 1506305 Increased trend of Rework CRs CR 1514561 Emerging trend of Significant Human Performance Related Events Within the Security Department CR 1517434 B computer room AHU trend indicates failure in 1-2 days CR 1519036 Develop ODMI per OPDP-11 for rising trend in DW identified leakage CR 1525220 Lowering Trend on Unit 2 EHC header pressure 71153 Level 2 Evaluation Report for CR 1454001 Drawings 0-45E772-1 Wiring Diagram 480V Stby Gas Trtmt Tn C Schematic Revision 23 Diagram 0-47E610-65-1 Mechanical Control Diagram Standby Gas Treatment Revision 50 System 2-45DS1655-211 Contact Development of Control and Instrument Revision 3 Switches Miscellaneous BFN-50-7065 General Design Criteria Document, Standby Gas Revision 20 Treatment Systems FSAR 5.3 Secondary Containment Systems Inspection Type Designation Description or Title Revision or Procedure Date Operability Past Operability Evaluation Documentation for CR Evaluations 1454001 Procedures 0-MEG-TRBSHT-001 Initial Troubleshooting Revision 0 3-SR-3.3.6.2.4(GRP 6) Group 6 PCIS Logic Revision 16 ECI-0-000-SWZ001 Replacement of Type SB Switches Revision 26 NPG-SPP-06.14 Guidelines for Planning and Execution of Revision 2 Troubleshooting Activities NPG-SPP-22.300 Corrective Action Program Revision 14 Work Orders WO 119345859 26| ]]