IR 05000259/2014008

Download: ML14274A559

Text

September 29, 2014

Mr. Joseph Vice President, Nuclear Licensing Tennessee Valley Authority 1101 Market Street, LP 3D-C Chattanooga, TN 37402-2801

SUBJECT: BROWNS FERRY NUCLEAR PLANT - NRC EVALUATION OF CHANGES, TESTS, AND EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS INSPECTION REPORT 05000259/2014008, 05000260/2014008 AND 05000296/2014008

Dear Mr. Shea:

On September 12, 2014, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Browns Ferry Nuclear Plant, Units 1, 2, and 3, and discussed the results of this inspection with Mr. K. Polson and other members of your staff. Additional inspection results were discussed with Mr. K. Groom of your staff on September 22, 2014. Inspectors documented the results of this inspection in the enclosed inspection report.

The NRC inspectors did not identify any findings or violations of more than minor significance.

In accordance with Title 10 of the Code of Federal Regulations 2.390, "Public Inspections, Exemptions, Requests for Withholding," of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response, if any, will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's Agencywide Documents Access and Management System (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,/RA/ Rebecca L. Nease, Chief Engineering Branch 1 Division of Reactor Safety

Docket Nos.: 05000259, 05000260, 05000296 License Nos.: DPR-33, DPR-52, and DPR-68

Enclosure:

Inspection Report 05000259/2014008, 05000260/2014008 and 05000296/2014008

w/Attachment:

Supplementary Information

cc: Distribution via Listserv September 29, 2014 Mr. Joseph

SUMMARY

Inspection Report (IR) 05000259/2014008, 05000260/2014008, 05000296/2014008; 08/25/2014-09/12/2014; Browns Ferry Nuclear Plant, Units 1, 2, and 3; NRC Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications.

This report covers a two-week, on-site inspection by two regional inspectors and one resident inspector. No findings or violations were identified. The Nuclear Regulatory Commission's (NRC's) program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 5, dated February 2014.

REPORT DETAILS

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R17 Evaluations of Changes, Tests, Experiments and Permanent Plant Modifications

a. Inspection Scope

Evaluations of Changes, Tests, and Experiments: The team reviewed seven safety evaluations performed pursuant to Title 10, Code of Federal Regulations (CFR) 50.59, "Changes, tests, and experiments," to determine if the evaluations were adequate and that prior NRC approval was obtained as appropriate. The team also reviewed 14 screenings where licensee personnel had determined that a 10 CFR 50.59 evaluation was not necessary. The team reviewed these documents to determine if:

• the changes, tests, or experiments performed were evaluated in accordance with 10 CFR 50.59 and that sufficient documentation existed to confirm that a license amendment was not required;

• the safety issues requiring the changes, tests or experiments were resolved;

• the licensee conclusions for evaluations of changes, tests, or experiments were correct and consistent with 10 CFR 50.59; and

• the design and licensing basis documentation used to support the change was updated to reflect the change. The team used, in part, Nuclear Energy Institute (NEI) 96-07, "Guidelines for 10 CFR 50.59 Implementation," Revision 1, to determine acceptability of the completed evaluations and screenings. The NEI document was endorsed by the NRC in Regulatory Guide 1.187, "Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments," dated November 2000.

Permanent Plant Modifications: The team reviewed seven permanent plant modifications that had been installed in the plant during the last three years. The modifications reviewed are listed below:

• Design Change Notice (DCN) 70488, Replace Flowserve (Anchor-Darling) Gate valve with Crane-Kalsi Sentinel gate valve, Rev. B

• DCN 64502, Replace the existing SPSVs "ASCO" with "AVCO", they have a 40 yr EQ life, Rev. A

• DCN 69939, Modify These TS to Qualify for Wetting and Spray from Moderate / Low Energy Line Break, Rev. A

• DCN 69959, ADS Logic Modification for Single Failure Proof Design, Rev. A

• DCN 69467, Replace CS & RHR Room Fan Motors, Fans, and Shafts, Rev. A

• Equivalent Change EQV 70945, Fabricate And Install a Limit Switch Follower Nut On 2-SHV-074-0087 Valve Stem, Rev. A

• DCN 70053, Replace EQ Damper Motors 0-FCO-65-4, -17, -26, -39, -52, -67, Rev. A The modifications were selected based upon risk significance, safety significance, and complexity. The team reviewed the modifications selected to determine if:

• the supporting design and licensing basis documentation was updated;

• the changes were in accordance with the specified design requirements;

• the procedures and training plans affected by the modification had been adequately updated;

• the test documentation as required by the applicable test programs had been updated; and

• post-modification testing adequately verified system operability and/or functionality. The team also used applicable industry standards to evaluate acceptability of the modifications and performed walkdowns of accessible portions of the modifications. Documents reviewed are listed in the Attachment.

b. Findings

No findings were identified.

4OA6 Meetings, Including Exit

On September 12, 2014, the team presented inspection results to Mr. K. Polson and other members of the licensee's staff. Additional inspection results were discussed with Mr. K. Groom of the licensee's staff on September 22, 2014. The team verified that no proprietary information was retained by the inspectors or documented in this report.

ATTACHMENT:

SUPPLEMENTARY INFORMATION

KEY POINTS OF CONTACT

Licensee personnel

R. Bates, Site Licensing Engineer

K. Groom, Design Engineering Manager

T. Mingus, Mechanical Design

J. Stone, Site Licensing Engineer

NRC personnel

C. Scott, Senior Resident Inspector (Acting), Browns Ferry Nuclear Plant

T. Stephen, Resident Inspector, Browns Ferry Nuclear Plant

LIST OF ITEMS OPENED, CLOSED, DISCUSSED, AND UPDATED

Opened

None

Closed

None

Discussed

None

Updated None

LIST OF DOCUMENTS REVIEWED

10

CFR 50.59 Evaluations 10

CFR 50.59 Evaluation Form,

DCN 69168A, Rev. 0 10

CFR 50.59 Evaluation Form,

DCN 69532A, Rev. 3

CFR 50.59 Evaluation Form,

DCN 69995A, Rev. 0

CFR 50.59 Evaluation Form,

DCN 70132, Rev. 0

CFR 50.59 Evaluation Form,

EDC 70660A, Rev. 2

CFR 50.59 Evaluation Form,

DCN 70834A, Revs. 0 and 1 10

CFR 50.59 Evaluation Form,

DCN 70835A, Rev. 0

10

CFR 50.59 Screenings Screening Review Form,

DCN 25853, Rev. 0

Screening Review Form,

DCN 51048A, Rev. 0 Screening Review Form,

DCN 51151A, Rev. 1 Screening Review Form,

DCN 69445A, Rev. 2

Screening Review Form,

DCN 69454A, Rev. 3

Screening Review Form,

DCN 69682, Rev. 1

Screening Review Form,

DCN 69907, Rev. 0 Screening Review Form,

DCN 69952A, Rev.1

Screening Review Form,

DCN 70054A, Rev. 1

Screening Review Form,

DCN 70162, Rev. 0

Screening Review Form,

DCN 70234A, Rev. 0

Screening Review Form,

DCN 70275A, Rev. 0

Screening Review Form,

DCN 70352, Rev. 0 Screening Review Form,

DCN 70490, Rev. 0

Licensing Bases Documents Technical Requirements Manual, Current Technical Specifications and Bases, Current Updated Final Safety Analysis, Current

Calculations

ED-Q0999-2003-0037, Fire Area/Zone Detailed Appendix R Safe Shutdown Separation Analysis Record - Fire Area 22, Rev. 13

EDQ 099920030048, Unit 1, 2, and 3 Appendix R Manual Action Requirements, Revs 26, 27, and 35

MDN 0085880002, ARI Venting Time of Scram Air Header, Revs. 7 and 8

MDQ 0073870190, HPCI Piping Pressure Drop, Rev. 17

MDQ 0999920053, Station Blackout- Multi-Unit HVAC and DG Availability Analysis, Rev. 10

MDQ 0999950057, Evaluation of Power Operated Gate Valves for NRC Generic Letter 95-07, Revs. 12 and 14

MDQ 0999980001, MOV Calculation Input Parameters, Rev. 13

MDQ 107320020089, MOV 1-FCV-73-0016, Operator Requirements and Capabilities, Rev. 4

MDQ 107420020051, MOV 1-FCV-074-0048 - Operator Requirements and Capabilities, Rev. 0

MDQ 199920020114, Evaluation of Stroke Times of

GL 89-10 MOVs Equipped with DC Motors, Rev. 7

MDQ 3074920421, MOV 3-FCV-74-07 Operator Requirements and Capabilities, Rev. 9

MDQ 3074920426, MOV 3-FCV-74-30 Operator Requirements and Capabilities, Rev. 9

NDQ 0064980007, Primary Containment Analysis, Rev. 7

NDQ 0999950014, Drywell Volume fractions for Use With Drywell Temperature Probes, Rev 14

Corrective Action Documents

PER 115333

PER 124904

PER 153228

PER 170722

PER 245386

PER 275699

PER 436575

PER 790144

PER 79705

PER 80014

PER 836156

PER 848134

PER 98262

Procedures

0-AOI-57-1A, Loss of Offsite Power (161 and 500 KV) / Station Blackout, Revs. 84 and 95

0-ARP-25-41B, BFNP Unit 0 Alarm Response Procedure, Rev. 8

0-ARP-25-41D, BFNP Unit 0 Alarm Response Procedure, Rev. 8 0-OI-23, Residual Heat Removal Service Water System, Revs. 94 and 97 0-OI-67, Emergency Equipment Cooling Water System, Revs. 97 and 102

0-SI-3.2.10.C, Verification of Remote Position Indicators for Emergency Equipment Cooling Water System Valves, Rev. 0 0-TI-362, Inservice Testing Program, Revs. 38 and 44

0-TI-362 (BASES), IST Program Bases Document, Revs. 3 and 6 1-AOI-100-2, Control Room Abandonment, Rev 21 1-ARP-9-3B, BFNP Unit 1 Alarm Response Procedure, Rev. 31

1-ARP-9-3F, Panel 9-3

XA-55-3F, Revs. 19 and 20

1-EOI Appendix-11A, Alternate RPV Pressure Control Systems MSRVs, Rev. 1

1-EOI Appendix 8G, Crosstie CAD to Drywell Control Air, Rev. 0 1-SIMI-80A, Primary Cooling Index, Rev. 0 1-SIMI-80B, Primary Cooling System Scaling and Setpoint Documents, Rev. 12

1-SIMI-80C, Primary Cooling System Calibration Data Sheets, Rev. 1

1-TI-82, Drywell Atmosphere Cooling System, Rev. 9

1/2-ARP-9-23A, BFNP Unit 1-2 Alarm Response Procedure, Rev. 5

1/2-ARP-9-23B, BFNP Unit 1-2 Alarm Response Procedure, Rev. 24 1/2-ARP-9-23D, BFNP Unit 1-2 Alarm Response Procedure, Rev. 24 2-AOI-100-2, Control Room Abandonment, Rev 57

2-SR-3.3.6.1.5 (1D), Main Steam Line Tunnel High Temperature Calibration, Rev. 9

3-AOI-100-2, Control Room Abandonment, Rev 22

3-ARP-25-4095, BFNP Unit 3 Alarm Response Procedure, Rev. 22 3-ARP-25-4096, BFNP Unit 3 Alarm Response Procedure, Rev. 22 3-ARP-25-4292, Browns Ferry Nuclear Plant Unit 3 Alarm Response Procedure, Rev. 23

3-ARP-25-4294, BFNP Unit 3 Alarm Response Procedure, Rev. 22

3-ARP-9-23A, BFNP Unit 3 Alarm Response Procedure, Rev. 21

3-ARP-9-23B, BFNP Unit 3 Alarm Response Procedure, Rev. 22

3-ARP-9-23C, BFNP Unit 3 Alarm Response Procedure, Rev. 22 3-ARP-9-23D, BFNP Unit 3 Alarm Response Procedure, Rev. 22

NPG-SPP-03.15, FSAR Management, Rev. 0

NPG-SPP-09.3, Plant Modifications and Engineering Change Control, Rev. 17

NPG-SPP-09.4, 10

CFR 50.59 Evaluations of Changes, Tests, and Experiments, Rev. 8

OPDP-1, Conduct of Operation, Rev. 33

SPP-8.3, Post Modification Testing, Rev. 6

Completed Procedures: 1-SR-3.5.1.7 (COMP), HPCI Comprehensive Pump Test, Rev. 21, dated 12/4/12

1-TI-82, Drywell Atmosphere Cooling System, Rev. 8, dated 1/16/11

General Design Criteria Documents

BFN-50-7023, Residual Heat Removal Service Water System, Revs 19 and 20

BFN-50-7067, Emergency Equipment Cooling Water System, Revs. 19 and 20

BFN-50-7073, High Pressure Coolant Injection System, Rev. 22

BFN-50-7074, Residual Heat Removal System, Rev. 11

BFN-50-7082, Standby Diesel Generator, Rev. 16

BFN-50-7307, Post-Accident Monitoring, Rev. 6

BFN-50-737, Backup Control, Rev. 5

BFN-50-C-7105, Pipe Rupture, Internal Missiles, Internal Flooding, Seismic Equipment Qualification and Vibration Qualification of Piping, Rev. 7

Drawings

0-105E2694, Process Diagram High Pressure Coolant Injection System for 3458MWT, Rev. 1 0-15E500-1, Unit 1 &2 Key Diagram of Standby Auxiliary Power System, Rev. 33

3-15E500-1, Unit 3 Key Diagram of Normal & Standby Auxiliary Power System, Rev. 66

0-15E500-2, Unit 1 &2 Key Diagram of Standby Auxiliary Power System, Rev. 27

0-45E610-67-2, Mechanical Control Diagram Emergency Equipment Cooling Water System, Rev. 36 0-45E724-1, Wiring Diagram 4160V Shutdown BD A Single Line, Revs. 29 and 30 0-45E724-2, Wiring Diagram 4160V Shutdown BD B Single Line, Revs. 33 and 34

0-45E732-1, Wiring Diagram 480V Diesel Auxiliary BD A Single Line, Revs. 40 and 43

0-45E732-3, Wiring Diagram 480V Diesel Auxiliary BD B Single Line, Revs. 36 and 38

0-47E861-5/R012, Flow Diagram - Cooling System and lubrication Oil System - Standby Diesel Generator A 0-47E861-6/R009, Flow Diagram - Cooling System and lubrication Oil System - Standby Diesel Generator B

0-47E861-7/R011, Flow Diagram - Cooling System and lubrication Oil System - Standby Diesel Generator C

0-47E861-8/R011, Flow Diagram - Cooling System and lubrication Oil System - Standby Diesel Generator D 0-8422409-2/R000, General Electric Drawing - Unit 1 and 2 Generator Diagram

1-47E225-101-1, Harsh Environmental Data Figures, Rev. 1

1-47E600-90, Mechanical Instruments and Controls, Rev. 0

1-47E610-67-1, Mechanical Control Diagram Emergency Equipment Cooling Water System, Rev. 39 1-47E812-1, Flow Diagram High Pressure Coolant Injection System, Revs. 35, 37 and 40

1-47E858-1, Flow Diagram RHR Service Water System, Revs. 64 and 66

1-47E859-1, Flow Diagram Emergency Equipment Cooling Water, Revs. 84 and 85

2-47E225-101-1, Harsh Environmental Data Figures, Rev. 2

2-47E600-90, Mechanical Instruments and Controls, Rev. 0 2-730E929 sh. 1, BFNP Unit 2 Elementary Diagram Automatic Blowdown System, Rev. 24

2-730E929 sh. 2, BFNP Unit 2 Elementary Diagram Automatic Blowdown System, Rev. 24

2-730E929 sh. 3, BFNP Unit 2 Elementary Diagram Automatic Blowdown System, Rev. 16

2-730E929 sh. 4, BFNP Unit 2 Elementary Diagram MSRV Auto Actuation Logic, Rev. 17

2-730E929 sh. 5, BFNP Unit 2 Elementary Diagram MSRV Auto Actuation Logic, Rev. 2 3-45E724-8, Wiring Diagram 4160V Shutdown BD 3EC Single Line, Revs. 35 and 37

3-45E724-9, Wiring Diagram 4160V Shutdown BD 3ED Single Line, Rev. 26 3-47E225-101-1, Harsh Environmental Data Figures, Rev. 3 3-47E600-90, Mechanical Instruments and Controls, Rev. 0

3-730-E918-1, BFNP Unit 3 E, Engineered Safeguard @ RCIC Instrumentation & control Separation Scheme, Rev. 2 3-730-E918-1, BFNP Unit 3 E, Engineered Safeguard @ RCIC Instrumentation & control Separation Scheme, Rev. 3 3-730-E929-1, BFNP Unit 3 Elementary Diagram Automatic Blowdown System, Rev. 17 3-730-E929-2, BFNP Unit 3 Elementary Diagram Automatic Blowdown System, Rev. 18

22409-3/R000, General Electric Drawing - Unit 1 and 2 Generator Diagram

BFNP Automatic Depressurization System Logic Bus A

BFNP Automatic Depressurization System Logic Bus B BFNP Automatic Depressurization System Basic Boolean BFNP FSAR Figure 7.4-3, Automatic Depressurization System Function Control Diagram (original) BFNP FSAR Figure 7.4-3, Automatic Depressurization System Mechanical Logic Diagram, (deleted) BFNP FSAR Figure 7.4-9, LPCI Break Direction-Logic Component Arrangement, (original)

BFNP FSAR Figure 7.4-9, Main Steam Automatic Depressurization System Mechanical Logic Diagram, (deleted)

Miscellaneous Documents

BFN-ENG-S-14-012, Conduct a Plant Modifications Inspection (using NRC document 711111.17T), dated 5/7/14

BFN-VTD-A613-0070, U-1 Scram Solenoid Pilot Valve Manifold Precautions, Engineering and Maintenance, Rev. 0

BFN-VTD-A701-0010, ACME Transformer General Catalog Industrial Transformers

BFN-VTD-C994-0100, Maintenance and Installation Manual

MI-241, Rev. 0, Rev. 1 BFNP System Description

SD-1A, Automatic Depressurization System, Rev 0

Browns Ferry NP Trending Report for U3R15, dated 9/17/12

DS-E2.0.2, Single Point Failure for Power Generation Reliability, Rev. 4

EDC 70660, Provide design output documentation for updating the SBO coping strategies and other supporting documents, Rev. A

EWR 12MEB073253, Crane Nuclear Design Report Number

DR-210, Rev. 0, dated 10/19/12 Letter from A. Thadani (NRC) to G. Sozzi (GE Nuclear Energy), Use of SHEX Computer Program and ANSI/ANS 5.1-1979 Decay Heat Source Term for Containment Long term Pressure and Temperature Analysis, dated 7/13/93 Letter from L. Raghavan (NRC) to J. Scalice (TVA), Issuance of Amendment RE: Power Uprate- Browns Ferry Plant, Units 2 and 3- (TAC Nos. M99711 and M99712), dated 9/8/98 Letter from L. Raghavan (NRC) to K. Singer (TVA), Browns Ferry Nuclear Plant, Unit 1- Issuance of Amendment Regarding Five Percent Uprate (TAC No. MD3048) (TS-431), dated 3/6/07 Letter from T. Abney (TVA) to Document Control Desk (NRC), Browns Ferry Nuclear Plant (BFN) - Units 2 and 3- Technical Specification (TS) Change

TS-384 - Request for License Amendment for Power Uprate Operation, dated 10/1/97 Letter from T. Abney (TVA) to Document Control Desk (NRC), Browns Ferry Nuclear Plant (BFN) - Units 2 and 3 Technical Specification (TS) No. 384 Supplement 1 - Request for License Amendment for Power Uprate Operation, dated 3/16/98

Letter from W. Russell (NRC) to P. Marriott (GE Nuclear Energy), Staff Position Concerning General Electric Boiling Water Reactor Power Uprate Program (TAC No. M79384), dated

9/30/91 NAMCO

EA-180 series vendor manual data sheets

Qualification Maintenance Data Sheets,

ITS-01, Rev. 18

Safety Assessment No. SABFEDCN930092, R0, DCN W25853A, dated 9/10/93

SAR Change Package No.19-020, dated 11/24/99

Screening Review Form,

DCN 64502A, Revs. 1 and 2

Screening Review Form,

DCN 69939, Rev. 0 Screening Review Form,

DCN 69959A, Rev. 0

SECY-93-067, Final Policy Statement on Technical Specifications Improvements, dated 3/17/93 Summary Report for 10

CFR 50.59 Evaluations, Fire Protection Report Revision, Technical Specifications Bases Changes, Technical Requirements Manual Changes, and NRC Commitment Revisions, dated 11/30/11 and 11/27/13 W79 980818-003, Final Power Uprate Task Report 4, Rev. 2, Primary Containment Evaluation, dated 8/18/98

Work Orders

114375726, 1-SR-2 Instrument Checks and Observations, dated 12/20/2013

Corrective Action Program Documents generated as a result of the inspection

SR 926600, 2014 10CFR50.59 Inspection- NRC Identified Calculation Discrepancy

SR 926998,

DCN 70132, Drywell Temperature NRC interview 71111-17-32

SR 927372, 2014 10CFR50.59 Inspection Item

ID 711111.17-43 identified

MD-Q0073-870190 reference errors

SR 928102, 2014 10CFR50.59 Inspection- MDQ0073870190

SR 929659 (PER 930971), The 50.59 evaluation does not clearly justify the "no" answer to question 2

SR 931867, Summer 2014 10CFR50.59 Modification Inspection 71111- NRC Identified

SR 931952, 2014 10CFR50.59 Inspection (71111) - NRC Identified

SR 931953, 2014 10CFR50.59 Inspection (71111) - NRC Identified

SR 932170, During Plant Mods Insp (71111) NRC Identified and Issue of concern regarding FSAR figure changes