NUREG-0847

Safety Evaluation Report

related to the operation of

Watts Bar Nuclear Plant,

Units 1 and 2

Docket Nos. 50-390 and 50-391

Tennessee Valley Authority

U.S. Nuclear Regulatory

Commission

Office of Nuclear Reactor Regulation

June 1982

1 00

,/
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NUREG-0847

Safety Evaluation Report

related to the operation of

Watts Bar Nuclear Plant,

Units 1 and 2

Docket Nos. 50-390 and 50-391

Tennessee Valley Authority

U.S. Nuclear Regulatory

Commission

Office of Nuclear Reactor Regulation

June 1982

ABSTRACT

The Safety Evaluation Report for the application filed by the Tennessee Valley

Authority, as applicant and owner, for a license to operate the Watts Bar

Nuclear Plant, Units 1 and 2 (Docket Nos. 50-390 and 50-391), has been prepared

by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory

Commission. The facility is located in Rhea County, Tennessee, near the Watts

Bar Dam on the Tennessee River. Subject to favorable resolution of the items

discussed in this report, the staff concludes that the facility can be operated

by the applicant without endangering the health and safety of the public.

Watts Bar SER iii

TABLE OF CONTENTS

Page

ABSTRACT ................................................................ iii

I INTRODUCTION AND DISCUSSION .................... ................... 1-1

1.1 Introduction ................................................. 1-1

1.2 General Plant Description ........ ........................... 1-6

1.3 Comparison with Similar Facility Designs ..................... 1-7

1.3.1 Comparison with the Sequoyah.Nuclear Plant ............ 1-7

1.3.2 Comparison with Other Facilities ...................... 1-10

1.4 Identification of Agents and Contractors ...................... 1-10

1.5 Summary of Principal Review Matters .......................... 1-11

1.6 Modifications to the Watts Bar Facility During the Course of

NRC Review ................................................... 1-12

1.7 Summary of Outstanding Issues ................................ 1-12

1.8 Confirmatory Issues .......................................... 1-12

1.9 License Conditions ............. ............................. 1-14

1.10 Unresolved Safety Issues ............................ .......... 1-15

2 SITE CHARACTERISTICS .............................................. 2-1

2.1 Geography and Demography ..................................... 2-1

2.1.1 Site Location and Description ......................... 2-1

2.1.2 Exclusion Area Authority and Control .................. 2-1

2.1.3 Population Distribution ............................... 2-6

2.1.4 Conclusions .......................................... 2-7

2.2 Nearby Industrial, Transportation, and Military

Facilities .... ............................................. 2-7

2.2.1 Transportation Routes ................................. 2-7

2.2.2 Nearby Facilities ..................................... 2-8

2.2.3 Conclusions ........................................... 2-9

2.3 Meteorology .................................................. 2-10

2.3.1 Regional Climatology .................................. 2-9

2.3.2 Local Meteorology ..................................... 2-12

2.3.3 Onsite Meteorological Measurements Program ............ 2-13

2.3.4 Short-Term (Accident) Diffusion Estimates ............. 2-14

2.3.5 Long-Term (Routine) Diffusion Estimates ............... 2-15

Watts Bar SER V
TABLE OF CONTENTS (Continued)

Page

2.4 Hydrologic Engineering ...................................... 2-16

2.4.1 Introduction ................................... .... 2-16

2.4.2 Hydrologic Description .............. .... ......... 2-16

2.4.3 Flood Potential ............... 2-17

2.4.4 Local Intense Precipitation in Plant Area ........ 2-19

2.4.5 Roof Drainage ............... ............... .... 2-19

2.4.6 Ultimate Heat Sink ................................... 2-19

2.4.7 Groundwater ......................................... 2-20

2.4.8 Design Basis. for Subsurface Hydrostatic Loading ......... 2-20

2.4.9 Transport of Liquid Releases .......................... 2-21

2.4.10 Flooding Protection Requirements and Technical

Specifications ....................................... 2-22

2.5 Geology and Seismology ....... ........................... 223

2.5.1 Geology..............................................2-23

2.5.2 Seismology .......................................... 2-24

.2.5.3 Surface Faulting ..................................... 2-32

2.5.4 Stability of Subsurface Materials and Foundations -.-. 2-32

2.5.5 Stability of Slopes ................................... 2-41

2.5.6 Embankments and Dams ................................. .2-42

2.6 References .......... ....................................... 2-42

3 DESIGN CRITERIA - STRUCTURE, COMPONENTS, EQUIPMENT, AND SYSTEMS ... 3-1

3.1 General ................................................... 3-1

3.1.1 Conformance with General Design Criteria ............... 3-1

3.1.2 Conformance with Industry Codes and Standards .......... 3-1

3.2 Classification of Structures, Systems, and Components ........ 3-1

3.2.1 Seismic Classification ........... .......---......... 3-1

3.2.2 System Quality Group Classification ................... 3-2

3.3 Wind and Tornado Loadings .................................. 3-4

3.3.1 Wind Loading .................. ....................... 3-4

3.3.2 Tornado Loading ...................................... 35

3.4 Flood Level (Flood) Design .................................. 3-6

3.4.1 Flood Protection ................................... 3-6

Watts Bar SER vi
TABLE OF CONTENTS (Cor ,-ued)

Page

3.5 Missile Protection ........................................... 3-7

3.5.1 Missile Selection and Description ..................... 3-7

3.5.2 Structures, Systems, and Components To Be Protected

From Externally Generated Missiles .................... 3-9

3.5.3 Barrier Design Procedures ............................. 3-10

3.6 Protection Against Dynamic Effects Associated with the

Postulated Rupture of Piping ............ I ................... 3-11

3.6.1 Plant Design for Protection Against Postulated Piping

Failures in Fluid Systems Outside Containment ......... 3-11

3.6.2 Determination of Break Locations and Dynamic

Effects Associated With the Postulated Rupture of

Piping ................................................. 3-12

3.7 Design Input ...................................................... 3-15

3.7.1 Seismic Input .............................................. 3-15

3.7.2 Seismic Analysis ........................................... 3-15

3.7.3 Seismic Subsystem Analysis ................................. .3-16

3.7.4 Seismic Instrumentation Program............................ 3-18

3.8 Design of Seismic Category I Structures ........................... 3-18

3.8.1 Steel Containment ......................................... 3-18

3.8.2 Concrete and Structural Steel Internal Structures .......... 3-20

3.8.3 Other Seismic Category I Structures ........................ 3-21

3.8.4 Foundations ................................................ 3-22

3.9 Mechanical Systems and Components ................................. 3-22

3.9.1 Special Topics for Mechanical Components ................... 3-22

3.9.2 Dynamic Testing and Analysis of Systems, Components,

and Equipment .............................................. 3-23

3.9.3 ASME Code Class 1, 2, and 3 Components, Component

Structures, and Core Support Structures ..................... 3-26

3.9.4 Control Rod Drive Systems ..................... ............ 3-30

3.9.5 Reactor Pressure Vessel Internals .......................... 3-31

3.9.6 Inservice Testing of Pumps and Valves ...................... 3-32

3.10 Seismic and Dynamic Qualification of Seismic Category I

Mechanical and Electrical Equipment ............................... 3-33

3.11 Environmental Qualification of Mechanical.and Electrical

Equipment ............................ .. ............................. 3-34

Watts Bar SER vii
TABLE OF CONTENTS (Continued)

Page

4 REACTOR ................. ......................................... 4-1

4.1 Introduction ........................... 4-1

4.2 Fuel System Design* .......................................... 4-1

4.2.1 Description ........................................... 4-1

4.2.2 Thermal Performance .................................. 4-2

4.2.3 Mechanical Performance ............. ; ................... 4-3

4.2.4 Surveillance ................... ...................... 4-8

4.2.5 Fuel Design Conclusions ........................... 4-9

4.3 Nuclear Design ............................................... .4-9

4.3.1 Design Bases .................................... ........ 4-9

4.3.2 Design Description .................................... 4-10

4.3.3 Analytical Methods .................... ............... 4-13

4.3.4 Summary of Evaluation Findings ........................ 4-13

4.4 Thermal-Hydraulic Design .................................... 4-14

4.4.1 Performance in Safety Criteria ........................ 4-14

4.4.2 Design Bases ......................................... 4-14

4.4.3 Thermal-Hydraulic Design Methodology .................. 4-15

4.4.4 Operating Abnormalities ............................... 4-17

4.4.5 Loose Parts Monitoring System ......................... 4-19

4.4.6 Thermal-Hydraulic Comparison .......................... 4-20

4.4.7 N-1 Loop Operation ........................ . ........ 4-20

4.4.8 Instrumentation for Inadequate Core Cooling

Detection (II.F.2) .................................... 4-20

4.4.9 Summary and Conclusion ............................... 4-23

4.5 Reactor Materials ............................................. 4-24

4.5.1 Control Rod Drive Structural Materials ................ 4-24

4.5.2 Reactor Internals and Core Support Materials .......... 4-24

4.6 Functional Design of Reactivity.Control Systems.............. 4-25

5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS ...................... 5-.1

5.1 Summary Description .......................................... 5-1

5.2 Integrity of Reactor Coolant Pressure Boundary ............... 5-1

5.2.1 Compliance With Codes and Code Cases .................. 5-1

5.2.2 Overpressure Protection ............................... 5-2

5.2.3 Reactor Coolant Pressure Boundary Materials ............ 5-3

5.2.4 Reactor Coolant Pressure Boundary Inservice

Inspection and Testing ................................ 5-5

5.2.5 Reactor Coolant Pressure Boundary Leakage Detection .. 5-6

Watts Bar SER viii
TABLE OF CONTENTS (Continued)

Page

5.3 Reactor Vessel ............................................. 5-8

5.3.1 Reactor Vessel Materials .............................. 5-8

5.3.2 Pressure-Temperature Limits ........................... 5-12

5.3.3 Reactor Vessel Integrity .............................. 5-13

5.4 Component and Subsystem Design ............................... 5-14

5.4.1 Reactor Coolant Pumps ................................. .5-14

5.4.2 Steam Generators ...................................... 5-14

5.4.3 Residual Heat Removal System. ......................... 5-16

5.4.4 Pressurizer Relief Tank ............................... 5-22

5.4.5 Reactor Coolant System Vents (II.B.1) .................. 5-22

6 ENGINEERED SAFETY FEATURES ........................................ 6-1

6.1 Engineered Safety Features Materials ......................... 6-1

6.1.1 Metallic Materials ..................................... 6-1

6.1.2 Organic Materials ..................................... 6-2

6.1.3 Postaccident Emergency Cooling Water Chemistry ......... 6-3

6.2 Containment Systems ......................................... 6-3

6.2.1 Containment Functional Design ......................... 6-4

6.2.2 Containment Heat Removal Systems ...................... 6-15

6.2.3 Secondary Containment Functional Design ............... 6-16

6.2.4 Containment Isolation System ........................... 6-17

6.2.5 Combustible Gas Control Systems ....................... 6-21

6.2.6 Containment Leakage Testing ........................... 6-22

6.2.7 Fracture Prevention of Containment Pressure

Boundary .............................................. 6-23

6.3 Emergency Core Cooling System ................................ 6-25

6.3.1 System Design ........................................ 6-25

6.3.2 Evaluation ............................................ 6-26

6.3.3 Testing .................................... ........... 6-30

6.3.4 Performance Evaluation ................................ 6-33

6.3.5 Conclusions .......................................... 6-34

6.4 Control Room Habitability ................................... 6-34

6.5 Engineered Safety Feature (ESF) Filter Systems ............... 6-35

6.5.1 ESF Atmosphere Cleanup Systems ........................ 6-35

6.5.2 Fission Product Cleanup System ........................ 6-36

6.5.3 Fission Product Control System ........................ 6-37

6.5.4 Ice Condenser as a Fission Product Removal System ..... 6-37

6.6 Inservice Inspection of Class 2 and 3 Components .............. 6-37

Watts Bar SER ix
TABLE OF CONTENTS (Continued)

Page

7 INSTRUMENTATION AND CONTROLS ..................................... 7-1

7.1 Introduction ..................... ........................... 7-1

7.1.1 General ............................................. 7-1

7.1.2 Comparison with Other Plants .......................... 7-1

7.1.3 Design Criteria ..................................... .7-1

7.2 Reactor Trip System ........................................ 7-3

7.2.1 System Description ................................... 7-3

7.2.2 Manual Trip Switches ...... ............................ 7-4

7.2.3 Testing of Reactor Trip Breaker Shunt Coils ......... 7-5

7.2.4 Anticipatory Trips ...... .................... ......... 7-5

7.2.5 Steam Generator Water Level Trip ....................... 7-5

7.2.6 Conclusions ........................................... 7-6

7.3 Engineered Safety Features Actuation System .................. 7-7

7.3.1 System Description .................................... 7-7

7.3.2 Containment Sump Level Measurement ..................... 7-8

7.3.3 Auxiliary Feedwater Initiation and Control ........ 7-9

7.3.4 Failure Modes and Effects Analysis .................... 7-9

7.3.5 IE Bulletin 80-06 ..................................... 7-10

7.3.6 Conclusions ........................................... 7-10

7.4 Systems Required for Safe Shutdown ........................... 7-11

7.4.1 System Description .................................... 7-11

7.4.2 Shutdown from Auxiliary Control Room ................... 7-12

7.4.3 Conclusions .... ...................................... 7-12

7.5 Safety-Related Display Instrumentation ....................... 7-12

7.5.1 System Description ................................... 7-12

7.5.2 Postaccident Monitoring System ........................ 7-13

7.5.3 IE Bulletin 79-27 ..................................... 7-14

7.5.4 Conclusions .......................... ................. 7-14

7.6 All Other Systems Required for Safety ........................ 7-14

7.6.1 System Description .................................... 7-14

7.6.2 Residual Heat Removal System Bypass Valves ............ 7-15

7.6.3 Upper Head Injection Manual Control ................... 7-16

7.6.4 Protection Against Spurious Actuation of Motor-

.Operated Valves ............. ......................... 7-16

Watts Bar SER X
TABLE OF CONTENTS (Continued)

Page

7.6.5 Overpressure Protection During Low Temperature

Operation ........................................... 7-17

7.6.6 Valve Power Lockout ................................. 7-17

7.6.7 Cold leg Accumulator Valve Interlocks and Postion

Indication ........................................... 7-17

7.6.8 Automatic Switchover from Injection to Recirculation

Mode .................................................. 7-18

7.6.9 Conclusions ............................................ 7-18

7.7 Control Systems Not Required for Safety ...................... 7-18

7.7.1 System Description ................................... 7-18

7.7.2 Safety System Status Monitoring System ................ 7-20

7.7.3 Volume Control Tank Level Control System ............... 7-22

7.7.4 Pressurizer and Steam Generator Overfill .............. 7-23

7.7.5 IE Information Notice 79-22 ........................... 7-23

7.7.6 Multiple Control System Failures ...................... 7-24

7.7.7 Conclusions ............................ ............... 7-24

7.8 NUREG-0737 Items............................... ............ 7-25

7.8.1 Relief and Safety Valve Position Indication (II.D.3) 7-25

7.8.2 Auxiliary Feedwater System initiation and'Flow

Indication (.II.E.1.2) ................................. 7-25

7.8.3 Proportional Integral Derivative Control

Modification (II.K.3.9) .............................. 7-25

7.8.4 Proposed Anticipatory Trip Modification (II.K.3.10) ... 7-26

7.8.5 Confirm Existence of Anticipatory Reactor Trip upon

Turbine Trip (II.K.3.12).............................. 7-26

8 ELECTRIC POWER SYSTEMS .......................................... 8-1

8.1 General .......................... 8-1

8.2 Offsite Electric Power System................. ........... 81..

8.2.1 Compliance with GDC 5 ........................... I ...... 8-1

8.2.2 Compliance with GDC 17 ....... 8-2

8.2.3 Compliance with GDC 18 ................................ .8-4

8.2.4 Evaluation Findings .................................. ... 8-5

8.3 Onsite Power Systems .......................................... 8-5

8.3.1 Onsite AC Power System Compliance with GDC 17 ......... 8-5

8.3.2 Onsite DC System Compliance with GDC 17 ................ 8-10

8.3.4 Evaluation Findings ................................... 8-24

Watts Bar SER xi
TABLE OF CONTENTS (Continued)

Page

9 AUXILIARY SYSTEMS ................................................. 9-1

9.1 Fuel Storage Facility .......................................... 9-1

9.1.1 New Fuel Storage ...................................... 9-1

9.1.2 Spent Fuel Storage .................................... 9-2

9.1.3 Spent Fuel Pool Cooling and Cleanup, System ............. 9-4

9.1.4 Fuel-Handling System ........... ....................... 9-6

9.2 Water Systems ................................................. 9-7

9.2.1 Essential Raw Cooling Water and Raw Cooling Water

Systems ............................................... 9-7

9.2.2 Component Cooling System (Reactor Auxiliaries

Cooling Water System) ................................. 9-10

9.2.3 Demineralized Water Makeup System .................... 9-11

9.2.4 Potable and Sanitary Water Systems .................... 9-12

9.2.5 Ultimate Heat Sink .................................... 9-12

9.2.6 Condensate Storage Facilities ......................... 9-13

9.3 Process Auxiliaries ............ ............................... 9-14

9.3.1 Compressed Air System ................................. 9-14

9.3.2 Process Sampling System ............................... 9-15

9.3.3 Equipment and Floor Drainage System ................... 9-18

9.3.4 Chemical and Volume Control System .................... 9-20

9.4 Heating, Ventilation, and Air Conditioning Systems .......... 9-21

9.4.1 Control Room Area Ventilation System .................. 9-21

9.4.2 Fuel Handling Area Ventilation System ................. 9-22

9.4.3 Auxiliary and Radwaste Area Ventilation System ........ 9-23

9.4.4 Turbine Building Area Ventilation System .............. 9-24

-9.4.5 Engineered Safety Features Ventilation System .......... 9-24

9.5 Other Auxiliary Systems ....................................... 9-25

9.5.1 Fire Protection ....................................... 9-25

9.5.2 Communication Systems .................................. .9-40

9.5.3 Lighting System ..: .................................... 9-46

9.5.4 Emergency Diesel Engine Fuel Oil Storage and

Transfer System ....................................... 9-48

9.5.5 Emergency Diesel Engine Cooling Water System .......... 9-53

9.5.6 Emergency Diesel Engine Starting Systems .............. 9-56

9.5.7 Emergency Diesel Engine Lubricating Oil System ........ 9-57

9.5.8 Emergency Diesel Engine Combustion Air Intake and

Exhaust System .......... ...... ........................ 9-59

Watts Bar SER xii
TABLE OF CONTENTS (Continued)

Page

10 STEAM AND POWER CONVERSION SYSTEM ................................. 10-1

10.1 Summary Description .......................................... 10-1

10.2 Turbine Generator ......................................... 10-1

10.2.1 Turbine Generator Design ............................ 10-1

10.2.2 Turbine Disc Integrity .............................. 10-3

10.3 Main Steam Supply System..................................... 10-4

10.3.1 Main Steam Supply System (up to and Including the

Main Steam Isolation Valves) ........................ 10-4

10.3.2 Main Steam Supply System ............................ 10-5

10.3.3 Steam and Feedwater Systems Materials ................ 10-5

10.3.4 Secondary Water Chemistry ........................... 10-6

10.4 Other Features ............................................ 10-7

10.4.1 Main Condenser ....................................... 10-7

10.4.2 Main Condenser Evacuation System .................... 10-8

10.4.3 Turbine Gland Sealing System ........................ 10-8

10.4.4 Turbine Bypass System ............................... 10-9

10.4.5 Condenser Circulating Water System .................. 10-10

10.4.6 Condensate Cleanup System ........................... 10-11

10.4.7 Condensate and Feedwater Systems .................... 10-12

10.4.8 Steam Generator Blowdown System..................... 10-13

10.4.9 Auxiliary Feedwater System .......................... 10-13

11 RADIOACTIVE WASTE SYSTEM ............................................ 11-1

11.1 Summary Description ..................................... 11-1

11.2 Liquid Waste Management .................................. 11-2

11.3 Gaseous Waste Management ................................... 11-3

11.4 Solid Waste Management System ............................... 11-5

11.5 Process and Effluent Radiological Monitoring and

Sampling Systems .......................................... 11-6

11.6 Evaluation Findings ......................................... 11-7

11.7 NUREG-0737 Items ............................................. 11-9

11.7.1 Wide Range Noble Gas, Iodine, and Particulate

Effluent Monitors (II.F.1(1) and II.F.1(2)) ......... 11-9

11.7.2 Primary Coolant Outside Containment (III.D.1.1) ..... 11-9

12 RADIATION PROTECTION .............................................. 12-1

12.1 General ....................... ............................. 12-1

12.2 Ensuring that Occupational Radiation Exposures Are

As Low As Reasonably Achievable ............................. 12-1

12.3 Radiation Sources ........................................... 12-3

Watts Bar SER xiii
TABLE OF CONTENTS (Continued)

Page

12.4 Radiation Protection Design Features ........... ............. 12-3

12.5 Dose Assessment ............................................. 12-6

12.6 Health Physics Program ....... ..... ..................... ...... 12-6

12.7 NUREG-0737 Items .......... ......... ........................ 12-7

12.7.1 Plant Shielding (II.B.2) ..................... ....... 12-7

12.7.2 High Range Incontainment Monitor (II.F.1(3))......... 12-7

12.7.3 Inplant Radioiodine Monitor (III.D.3.3) .............. 12-7

13 CONDUCT OF OPERATIONS ............................................... 13-1

13.1 Organizational Structure of Applicant ....................... 13-1

13.1.1 Management and Technical Organization ................ 13-1

13.1.2 Corporate Organization and Technical Support ......... 13-1

13.1.3 Plant Staff Organization ............................ 13-8

13.2 Training ........ ................... ...................... 13-14

13.2.1 Licensed Operator Training Program. 13-14

13.2.2 Training for Nonlicensed Personnel .................. 13-15

13.3 Emergency Preparedness Evaluation ........................... 13-18

13.3.1 Introduction ........................................ 13-18

13.3.2 Evaluation of the Emergency Plan .................... 13-18

13.3.3 Conclusions ......................................... 13-28

13.4 Review and Audit .. ........................................ 13-29

13.5 Plant Procedures .......................................... 13-32

13.5.1 Administrative Procedures ............................. 13-32

13.5.2 Operating and Maintenance Procedures ................ 13-37

13.5.3 NUREG-0737 Items ......................... .......... 13-39

13.6 Physical Security Plan ................................... 13-39

14 INITIAL TEST PROGRAM ............................ .................. 14-1

15 ACCIDENT ANALYSIS ................................................. 15-1

15.1 General Discussion ........................................... 15-1

15.2 Normal Operation and Anticipated Transients ................... 15-1

15.2.1 Loss-of-Coolant Transients ............................. 15-2

15.2.2 Increased Cooling Transients .......................... 15-3

15.2.3 Change in Coolant Inventory Transients ............... 15-3

15.2.4 Reactivity and Power Distribution Anomalies .......... 15-3

15.2.5 Conclusions ...... .................................... .15-7

Watts Bar SER xiv
TABLE OF CONTENTS (Continued)

Page

15.3 Limiting Accidents ............................ ...... ......... 15-7

15.3.1 Loss-of-Coolant Accident .............................. 15-8

15.3.2 Steamline Break ...................................... 15-10

15.3.3 Feedwater System Pipe Break.......... ................ 15-10

15.3.4 Reactor Coolant Pump Rotor Seizure .................... 15-11

15.3.5 Reactor Coolant Pump Shaft, Break...................... 15-11

15.3.6 Anticipated Transients.Without Scram .................. 15-11

15.3.7 Conclusions ........................................... 15-12

15.4 Radiological Consequences of Accidents ........................ 15-13

15.4.1 Loss-of-Coolant Accident .............................. 15-13

15.4.2 Main Steamline Break Outside of Containment ........... 15-17

15.4.3 Steam Generator Tube Rupture......... ................ 15-19

15.4.4 Control Rod Ejection Accident................ 15-21

15.4.5 Fuel-Handling Accident................................ 15-22

15.4.6 Failure of Small Lines Carrying Coolant

Outside Containment ................................... 15-24

15.4.7 Postulated Radioactive-Releases as a Result of

Liquid Tank Failures ................................. 15-25

15.5 NUREG-0737 Items ............... ............................... 15-25

15.5.1 Thermal Mechanical Report (II.K.2.13) ................ 15-25

15.5.2 Voiding in the Reactor Coolant System During

Transients (II.K.2.17) ................ ............... 15-26

15.5.3 Installation and Testing of Automatic PowerOperated Relief Valve Isolation System (II.K.3.1)

Report on Overall Safety Effect of PowerOperated Relief Valve Isolation System (II.K.3.2) ...... 15-26

15.5.4 Automatic Trip of Reactor Coolant Pumps (II.K.3.5) .... 15-26

15.5.5 Small-Break LOCA Methods (II.K.3.30) and

Plant-Specific Calculations (II.K.3.31)................ 15-26

15.6 Relative Risk of Low Power Operation ......................... 15-27

16 TECHNICAL SPECIFICATIONS ........................................... 16-1

17 QUALITY ASSURANCE ........................................... ....... 17-1

17.1 General........................................................ . 17-1

,17.2 Organization .......................................... 17-3

17.3 Quality Assurance Program ..................................... 17-5

17.4 Conclusion .................... ............................... 17-7

18 CONTROL ROOM DESIGN REVIEW ......................................... 18-1

18.1 General ....................................................... 18-1

18.2 Conclusions ................................................... 18-2

Watts Bar SER XV
TABLE OF CONTENTS (Continued)

Page

19 REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS ............. 19-1

20 COMMON DEFENSE AND SECURITY .......................................... 20-1

21 FINANCIAL QUALIFICATIONS ........................................... 21-1

22 FINANCIAL PROTECTION AND INDEMNITY REQUIREMENTS .................... 22-1

22.1 General ........... ........................................ 22-1

22.2 Preoperational Storage of Nuclear Fuel ........................ 22-1

22.3 Operating Licenses ............................................ 22-1

23 CONCLUSIONS....................... ............................. 23-1

APPENDICES

A CHRONOLOGY OF RADIOLOGICAL REVIEW OF WATTS BAR NUCLEAR PLANT,

UNITS 1 AND 2, OPERATING LICENSE REVIEW

B BIBLIOGRAPHY

C NUCLEAR REGULATORY COMMISSION UNRESOLVED SAFETY ISSUES

D EVALUATION OF THE APPLICANT'S CONTROL ROOM DESIGN

E PRINCIPAL CONTRIBUTORS

Watts Bar SER xvi
LIST OF TABLES

Page

1.1 Cross-Reference Table for NUREG-0737 Items ......................... 1-2

1.2 Comparison Between Principal Design Features of Watts Bar

and Other Ice Condenser PWR Facilities ............................. 1-8

2.1 Resident Population in Watts Bar Vicinity .......................... 2-6

4.1 Rod Bow Penalties ................................................. .4-18

4.2 Generic Margins .................. ............................... 4-19

4.3 Reactor Design Comparison .......................................... 4-21

6.1 Containment Design Parameters........ ............................... 6-5

9.1 Performance With Recommendations of NUREG/CR-0660 .................. 9-50

11.1 Design Parameters of Principal Components Considered in the

Evaluation of Liquid and Gaseous Radioactive Waste Treatment

Systems ................................ ........................ 11-4

11.2 Process and Effluent Radiation Monitoring System ................... 11-8

15.1 Radiological Consequences of Design-Basis Accidents ................ 15-14

15.2 Assumptions Used in the Calculation of the Radiological,

Consequences Following a Postulated Loss-of-Coolant Accident. 15-16

15.3 Assumptions Used to Evaluate the Radiological Consequences

Following a Postulated Main Steamline Break.Accident Outside

Containment....................................................15-18

15.4 Assumptions Used for the Calculation of the Radiological

Consequences Following a Postulated Steam Generator Tube

Rupture Accident ................................................... 15-20

15.5 Assumptions Used for Estimating the Radiological Consequences

Following a Postulated Control Rod Ejection Accident ............... 15-22

15.6 Assumptions Used for Estimating the Radiological Consequences

Following a Postulated Fuel-Handling Accident ...................... 15-23

17.1 Regulatory Guidance Applicable to the Quality Assurance

Programs.......................................................... 17-2

Watts Bar SER xvi i
LIST OF FIGURES

2.1 Map of the General Region of the Watts Bar Site .................... 2-2

2.2 Watts Bar Site (0-50 mi) ....................... ................. 2-3

2.3 Watts Bar Immediate Site Area .................................... 2-4

2.4 Watts Bar Plant Site Boundary and Exclusion Area ................... 2-5

2.5 Airways in the Area of the Watts Bar Nuclear Plant ................ 2-10

2.6 Watts Bar Design Response Spectrum (5% damping) and 50.and

84 Percentile Site-Specific Spectra (4% and 7% damping) ............ 2-31

13.1 Organization of the Tennessee Valley Authority..................... 13-2

13.2 Office of Power Organizational Chart .............................. 13-3

13.3 Division of Nuclear Power Organizational Chart ..................... 13-4

13.4 Plant Organizational Chart ....................................... 13-9

17-1 TVA Quality Assurance Organization ..................... ........... 17-4

Watts Bar SER xviii
SAFETY EVALUATION REPORT

RELATED TO THE OPERATION OF'THE

WATTS BAR NUCLEAR PLANT, UNITS 1 AND 2

1 INTRODUCTION AND DISCUSSION

1.1 Introduction

This report is a safety evaluation report (SER) on the application for an operating license (OL) for the Watts Bar Nuclear Plant, Units 1 and 2 (the facility),

based upon an application filed by the Tennessee Valley Authority (TVA or the

applicant). This report was prepared by the U.S. Nuclear Regulatory Commission

(the staff) and summarizes the results of the staff's radiological safety review

of the facility. The NRC Licensing Project Manager for the Watts Bar facility

is Thomas J. Kenyon. Mr. Kenyon may be contacted by calling (301) 492-7266 or

by writing to: Division of Licensing, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555. The NRC staff principal reviewers and its contractors for

this project are listed in Appendix E.

The application for a construction permit (CP) for the facility was filed with

the U.S. Atomic Energy Commission (now the U.S. Nuclear Regulatory Commission)

on May 14, 1971. Following staff review and a public hearing before the Atomic

Safety and Licensing Board (ASLB), Construction Permits Nos. CPPR-91 and CPPR-92

were issued on January 23, 1973, and the application was docketed for review on

October 4, 1976. The applicant has stated that construction of Watts Bar Units 1

and 2 will be complete and the facilities ready for fuel loading by August 1983

and August 1984, respectively.

Before issuing an OL for a nuclear power plant, the NRC staff is required to

review the effects of the plant on public health-and safety. The safety review

of the Watts Bar facility has been based on the Final Safety Analysis Report

(FSAR) that accompanied the application for an OL and Amendments 1 through 46

thereto. All of this information is available to the public for review at the

NRC Public Document Room at 1717 H Street, N.W., Washington, D.C., and at the

Chattanooga-Hamilton County Bicentennial Library, 1001 Broad Street, Chattanooga,

Tennessee. During the course of the review the staff has met a number of times

with the applicant, the suppliers, and their consultants to discuss the design,

construction, and proposed operation of the facility. As a consequence, additional information was requested, which the applicant provided in Amendments 1

through 46 to the FSAR.

This SER summarizes the results of the staff's radiological safety review of

the Watts Bar facility and delineates the scope of the technical details considered in evaluating the radiological safety aspects of its proposed operation.

The design of the station was reviewed against the Federal regulations, CP

criteria, and the NRC "Standard Review Plan for the Review of Safety Analysis

Reports for Nuclear Power Plants" (SRP), NUREG-0800, dated July 1981 (except

where noted otherwise). The SRP is written to cover a variety of site conditions and plant designs. Each section is written to provide the complete procedure and all acceptance criteria for all of the areas of review pertinent

Watts Bar SER 1-1
to the section. However, for any given application, the staff reviewers may

select and emphasize particular aspects of each SRP section as is appropriate

for the application. In some cases, the major portion of the review of a plant

feature may be done on a generic basis with the designer of that feature rather

than in the context of reviews of particular applications from utilities. In

other cases, a plant featuremay be sufficiently similar to that of a previous

plant so that a de novo review of the feature is not needed. For these and

other similar reasons, the staff may not have carried out in detail all of the

review steps listed in each SRP section in the review of every application.

Review and evaluation of compliance by the applicant to the licensing requirements established in NUREG-0660, entitled "NRC-Action Plan Developed as a Result

of the TMI-2 Accident," and NUREG-0737, "Clarification of TMI Action Plan Requirements," have been incorporated into the appropriate sections of this report.

Table 1.1 provides a cross reference relating the applicable TMI items to the

sections in this report where the items are discussed.

Table 1.1 Cross-reference table for NUREG-0737 items

Item Shortened title SER section(s)

I.A.1.1 Shift Technical Advisor 13.1, 13.2.2

I.A.1.2 Shift Supervisor 13.5.1

responsibilities.

I.A.l.3 Shift manning 13.5.1

I.A.2.1 Immediate upgrade of 13.2.2

RO and SRO training

and qualifications

I.A.2.3 Administration of 13.2.2

training programs

I.A.3.1 Revise scope and criteria 13.2'2

for licensing exams

I.B.1.2 Independent Safety 13.4.6

Engineering Group

I.C.1 Short-term accident and 13.5.2

procedure review

I.C.2 Shift and relief turnover 13.5.1

procedures

I.C.3 Shift supervisor 13.5.1

responsibility

I.C.4 Control room access 13.5.1

I.C.5 Feedback of operating 13.5.1

experience

I.C.6 Verify correct performance 13.5.1

of operating activities

I.C.7 NSSS vendor review of 13.5.2

procedures

Watts Bar SER 1-2
Table 1.1 (continued)

Item Shortened title SER section(s)

I.C. 8

I.D. 1

I.D. 2

Pilot monitor of selected

emergency procedures for

NTOLs

Control room design reviews

Plant safety-parameter display

console

I.G.1 Training during low-power

testing

II.B.1 Reactor coolant system vents

II.B.2 Plant shielding

II.B.3 Postaccident sampling

II.B.4 Training for mitigating core

damage

II.D.1 Relief and safety valve

test requirements

II.D.3 Relief and safety valve

position indication

II.E.1.1 Auxiliary feedwater system

evaluation

II.E.1.2 Auxiliary feedwater system

initiation and flow

indication

II.E.3.1 Emergency power for

pressurizer heaters

II.E.4.1 Dedicated hydrogen

penetrations

II.E.4.2 Containment isolation

dependability

II.F.1 Accident monitoring

instrumentation

(1) Noble gas monitoring

(2) Iodine particulate

sampling

(3) High-range incontainment

radiation monitor

(4) Containment pressure

(5) Containment water level

(6) Containment hydrogen

II.F.2 Instrumentation for detection

of inadequate core cooling

13.5.2

18.0

7.5.2, 13.3

14.0

5.4.5

12.7.1

9.3.2

13.2.2

3.9.3.3, 5.2.2

7.8.1

10.4.9

7.8.2

8.3.3.4

6.2.5

6.2.4

11.7.1

11.7.1

7.5.1, 12.7.2

6.2.1, 7.5.1

6.2.1

6.2.5

4.4.8

Watts Bar SER 1-3
Table 1.1 (continued)

Item Shortened title SER section(s)

I I -G.l1

I I.K. 2

II.K. 3

Power supplies for pressurizer relief valves and level

indicators

Orders on plants

13. Effect of HPI for smallbreak LOCA with no aux

feed

17. Voiding in RCS

19. Benchmark analysis seq.

AFW flow

Final recommendations, B&O

Task Force

1. Auto PORV isolation

system

2. Report on PORV failures

3. Reporting SV and RV

failures and challenges

5. Auto trip of RCPs

9. PID controller

modification

10. Applicant's proposed

anticipatory trip at

high power

12. Confirm anticipatory

trip upon turbine trip

17. ECCS outages

25. Power on pump seals

30. SB LOCA methods

31. Plant-specific methods

Emergency preparedness,

short term

Upgrade emergency support

facilities

Emergency preparedness

Primary coolant outside

containment

Inplant radioiodine

monitoring

Control room habitability

15.5.1

15.5.2

10.4.9

8.3.3.4

15.5.3

15.5.3

13.5.3

15.5.4

7.8.3

7.8.4

7.8.5

13.5.3

9.2.2

15.5.5

15.5.5

13.3

13.3

13.3

11.7.2

12.7.3

6.4

III. A.1. 1

III.A.1.2

III. A. 2

III. D.1. 1

III.D. 3.3

I IL .0.3.4

Watts Bar SER 1-4
Sections 2 through 18 of the SER address the staff's review and evaluation of

safety-related issues that have been considered during the review of the OL

application. Section 19 is reserved for the report of the Advisory Committee

on Reactor Safeguards. Section 20 considers whether the operation of the

facility will be inimical to the common defense and security. Section 21 used

to present the staff review and evaluation of the financial qualifications of

the applicant. However, as stated in this section, NRC regulations have been

amended to eliminate this review requirement. Section 22 describes the financial protection and indemnity requirements for preoperational storage of

nuclear fuel and operation of the facility. The conclusions of this report

are given in Section 23.

Appendix A is a chronology of NRC's principal actions related to the review of

the application. Appendix B is a bibliography of the references used during

the course of the review.* Appendix C is a discussion of how the "Unresolved

Safety Issues" relate to the Watts Bar application. Appendix D is an evaluation of the applicant's control room design. Appendix E is a list of principal

contributors.

As part of the NRC review of the Watts Bar facility for compliance with the

Commission's regulations, the staff asked the applicant to verify that the

Watts Bar facility meets the pertinent regulatory requirements in Title 10,

Code of Federal Regulations (CFR) 20, 50, and 100 (U.S. General Services

Administration). The applicant's response to this request, which was submitted on December 28, 1982, stated that the Watts Bar facility is in compliance

with all applicable regulations and requirements except where justification

for specific exemptions have been submitted by the applicant. 'Subject to the

applicant's adoption of the additional requirements imposed by the staff in

this Safety Evaluation Report, and the exemptions granted, the staff concurs

that the facility design is in compliance with these regulations and

requirements.

In accordance with the provisions of the National Environmental Policy Act

(NEPA) of 1969, Draft and Final Environmental Statements that set forth the

considerations related to the proposed construction and operation of the Watts

Bar Nuclear Plant were prepared by the staff. The Draft Environmental

Statement was issued on June 5, 1978 and the Final Environmental Statement on

December 29, 1978.

On April 26, 1982, a meeting was:held with TVA to discuss their position that

the Watts bar design and construction quality assurance (QA) program is

adequate to ensure that the facility will meet the requirements and commitments

of the license application once completed. TVA stated that their management

commitment to QA has been implemented through the programs and procedures

developed by their Division of Engineering Design and Division of Construction,

TVA's internal QA audit program, and the independent review activities of the

American Society of Mechanical Engineers (ASME) TVA's Nuclear Safety Review

*Availability of all material cited is described on the inside front cover of

this report. Where appropriate, specific references are listed alphabetically

at the end of chapters. NRC Regulatory Guides will be identified in full in the text the first time they are cited and will be referred to by number

thereafter. They are not included as reference lists or in Appendix B.

Watts Bar SER 1-5
Staff (NSRS) and outside management consultants. The applicant has proposed

to have the Institute of Nuclear Power Operations (INPO) perform an independent

audit of Watts Bar design and construction program before fuel loading. The

staff has not yet determined the acceptability of this proposal which is under

review by the staff and will be addressed in a supplement to the SER.

The review and evaluation of the Watts Bar facility for an OL is only one stage

in the continuing review by the staff of the design, construction, and operating

features of the facility. The proposed design of the facility was reviewed as

part of the CP review. Construction of the facility has been monitored in accordance with the inspection program of the staff.- During the OL review stage, the

NRC staff reviewed the final design to determine that the Commission's safety

requirements have been met. If an OL is granted, the Watts Bar facility must be

operated in accordance with the terms-of the OL and the Commission's regulations

and will be subject to the continuing inspection program of the staff.

1.2 General Plant Description

The Watts Bar Nuclear Plant is very similar in design to the Sequoyah and

McGuire nuclear plants. Each unit of the Watts Bar Nuclear Plant uses a

nuclear steam supply system (NSSS) incorporating a pressurized water reactor

(PWR) rated at 3411 MWt and a 4-loop reactor coolant system. In each of the

units, the reactor core is composed of fuel rods made of slightly enriched

uranium dioxide pellets enclosed in Zircaloy tubes with welded end plugs that

are grouped and supported into assemblies. The mechanical control rods consist

of clusters of stainless steel-clad boron carbide absorber rods that are inserted

into Zircaloy guide tubes located within the fuel assemblies. The core fuel is

loaded in three regions, each using fuel of a different enrichment of U-235.

The new fuel is introduced into the outer region, moved inward at successive

refuelings, and removed from the inner region to the spent fuel storage pool.

Water will serve as both the moderator and the coolant. It will be circulated

through the reactor vessel and core by four vertical, single-stage centrifugal

pumps, one located in the cold leg of each loop. The coolant water heated by

the reactor will be circulated through the four steam generators where heat

will be transferred to the secondary system to produce saturated steam, and

then it will be returned to the pumps to repeat the cycle.

An electrically heated pressurizer connected to the hot-leg piping of one of

the loops will establish and maintain the reactor coolant pressure and provide

a surge chamber and a water reserve to accommodate reactor coolant volume

changes during operation.

The steam produced in the steam generators will be used to drive a tandem

compound double-stage reheat turbine and will be condensed in a triple-shell

single-pass deaerating condenser. Cooling water from the cooling tower basin

will be pumped through the tubes of the condenser to remove the heat from, and

thus condense, the steam after it has passed through the main turbine. The

condenser condensate then will be pumped back to the steam generator to be

heated for another cycle. The condenser cooling water will be passed through

two natural draft cooling towers and returned to the cooling tower basin.

The reactor will be controlled by a coordinated combination of a soluble

neutron absorber (boric acid) and mechanical control rods whose drive shafts

Watts Bar SER 1-6
will allow the plant to accept step load changes of 10percent and ramp load

changes of 5 percent per minute over the range of 15 to 100 percent of full

power under normal operating conditions. With steam bypass, the plant will

also have the capability to accept a 50-percent step load rejection without

reactor trip.

Plant protection systems are provided that automatically initiate appropriate

action whenever a monitored condition approaches preestablished limits. These

protection systems will act to shut down the reactor, close isolation valves,

and initiate operation of the engineered safety features should any or all of

these actions be required.

Supervision and control of both the NSSS and the steam and power conversion

system for both units will be accomplished from the main control room.

The emergency core cooling system for the plant consists of accumulators, upper

head injection, and both high- and low-pressure injection subsystems with provisions for recirculation of the borated water after the end of the injection

phase. Various combinations of these features will ensure core cooling for the

complete range of postulated coolant pipe break sizes.

Each of the two NSSS is housed in a separate free-standing steel containment

structure within a reinforced concrete shield building. The containments employ

the ice condenser pressure-suppression concept. A common auxiliary building

adjacent to the containment structures houses the radioactive waste treatment

facilities, components of the engineered safety features, and various related

auxiliary systems for each unit. The units share a common fuel handling facility that contains a spent fuel pool and a new fuel storage facility.

The plant is supplied with electrical power by independent transmission lines

from offsite power sources. It has independent and redundant onsite emergency

power supplies capable of supplying power to shut down the plant safely or to

operate the engineered safety features in the event of an accident.

1.3 Comparison with Similar Facility Designs

Comparable principal parameters and features for the Watts Bar and other

recently licensed PWR ice condenser facilities are listed in Table 1.2.

1.3.1 Comparison with the Sequoyah Nuclear Plant

The TVA Watts Bar and Sequoyah nuclear plants are very similar in design. The

reactor cores are similar in design and each plant uses a free-standing stee-l

vessel with an ice condenser and a reinforced concrete shield building. The

plants' mechanical systems, containment systems, emergency core cooling systems,

instrument and control systems, electric power systems, auxiliary systems, radioactive waste systems, and steam and power conversion systems also are very similar

in design and materials; most differences that do exist are a result of the different vintages of the two plants. The CP for the Watts Bar facility was issued

on January 23, 1973, while the Sequoyah CP was issued on May 27, 1970. The OL

applications for the facilities were docketed on October 4, 1976 and January 31,

1974, respectively.

Watts Bar SER 1-7
In addition to the hardware similarities mentioned above, the organizational

structure of the facilities, the training programs for TVA personnel, and the

operating procedures for the facilities (for normal and abnormal conditions)

are very similar in nature. The commitments made by TVA in response to

NUREG-0737 also are very similar.

As a result of these similarities, the staff asked the applicant to prepare a

comparison of the major systems, components, structures, and features (both

hardware and software) between the two facilities. The intent of this comparison was to expedite the review of the Watts Bar facility by taking advantage

of the knowledge accrued by those reviewers who recently reviewed the Sequoyah

facility. The comparison was submitted by TVA by letter dated August 17, 1981.

Table 1.2 Comparison between principal design features of

Watts Bar and other ice condenser PWR facilities

Design features Watts Bar Sequoyah McGuire

Containment types Ice condenser Ice condenser Ice condenser

Containment design 15.0 12.0 7 15.0

pressure, psig

Containment volume, ft 3 1,191,400 1,191,500 1,196,000

Reactor thermal 3,411 3,411 3,411

power, MWt

Gross electrical 1,171 1,128 (Unit 1) 1,180

output, MWe 1,148 (Unit 2)

Total reactor coolant 140,300,00 138,100,00 140,300,000

flow-rate, lb/hr

Secondary steam

flow-rate, lb/hr 15,140,000 14,250,000 15,144,000

Fuel lattice 17 x 17 17 x 17 17 x 17

Number of fuel

assemblies 193 193 193

Number of fuel rods

per fuel assembly 264 264 264

Number of rod cluster

control assemblies 53 53 53

Reactor vessel

inside diameter,

in. 173 173 173

Overall height of

vessel and closure

head ft-in. 43-10 43-9-5/8 43-9-5/8

Watts Bar SER 1-8
Table 1.2 (Continued)

Design features

Reactor vessel

design pressure,

psig

Reactor vessel

design temperature, oF

Reactor vessel

minimum cladding

thickness, in.

Number of

recirculation loops

Average heat flux,

Btu/hr-ft 2

Maximum heat flux

Btu/hr-ft 2

Reactor coolant pump

flowrate, gpm

Steam generator

type

Steam generator

overall height (ft-in.)

Number of pressurizer

safety valves

Safety valve relieving

capacity, lb/hr

Number of poweroperated relief

valves (PORVs)

PORV-relieving capcity,

lb/hr

Turbine generator output, kW

Secondary steam pressure, psia

Secondary steam

temperature, OF

Watts Bar Sequoyah McGuire

2,485 2,485

650 650

1/8

4

189,800

440,300

101,000

W Model D-3

67-8

3

420,000

2

210,000

1,218,000

1,000

544

5/32

4

189,800

424,600

88,500

W Model 51

67-8

3

420,000

2

179,000

1,185,000

832

522

2,485

650

1/8

4

189,800

440,300

99,000

W Model D-2 (Unit 1)

D-3 (Unit 2)

67-8

3

420,000

3

210,000

1,220,000

975

542

Watts Bar SER 1-9
In light of this comparison, the Watts Bar review was performed on the following basis:

(1) Where systems were essentially the same as those of the Sequoyah nuclear

plant, the review performed on the Watts Bar system was confirmatory in

nature.

(2) Where systems were too dissimilar from those of the Sequoyah plant, or

where the depth of the applicant's comparison between the facilities'

systems was insufficient, a normal review was performed.

The areas where the Sequoyah-Watts Bar facility comparison was used as a major

tool in the review of Watts Bar are noted where applicable in this report.

The comparison between the two facilities was useful in the following areas:

* structural engineering

* auxiliary systems

* containment systems

* chemical engineering systems

* reactor systems

* effluent treatment systems

* operator licensing .procedures and test review

* emergency planning

* control room design

In many cases, the SER (and its supplements) for the Sequoyah nuclear plant

(NUREG-0011) is referenced; it may be obtained by writing to the address listed

in Section 1.3.2 below.

1.3.2 Comparison with Other Facilities

Many features of the design of the Watts Bar nuclear plant are similar to facilities (other than the Sequoyah nuclear plant) that the staff has evaluated and

approved previously. To the extent feasible and appropriate, the staff has

relied also on these other earlier reviews for those Watts Bar features that

were shown to be substantially the same as those previously considered. Where

this has been done, the appropriate sections of this report identify the other

facilities involved. The NRC SERs for these other facilities have been published

and are available for public inspection at-the NRC Public Document Room at

1717 H Street, N.W., Washington, DC 20555.

1.4 Identification of Agents and Contractors

The Westinghouse Electric Corporation (Westinghouse) is supplying the nuclear

steam supply system, including the first fuel loading and the turbine generators. TVA is responsible for the design of the remainder of the plant and for

all other architect-engineer functions, as well as for the construction and

operation of the plant.

Principal consultants utilized by TVA to perform selected design work andother

specialized services include Western Geophysical Engineering, Inc. for soil

foundation dynamic analyses; Engineering Data Systems, Inc. for seismic analysis

Watts Bar SER 1-10
of piping; and Chicago Bridge and Iron Company for design and construction of

the free-standing steel containments.

1.5 Summary of Principal Review Matters

NRC technical review and evaluation of the information submitted by the applicant considered, or will consider, the principal matters summarized below:

(1) *The population density and land use characteristics of the site environs

and the physical characteristics of the site (including seismology, meteorology, geology, and hydrology) to establish that these characteristics

have been determined adequately and have been given appropriate consideration i.n the plant design, and that the site characteristics are in accordance with the Commission's siting criteria of 10 CFR 100, taking into

consideration the design of the facility, including the engineered safety

features provided.

(2) The design, fabrication, construction and testing criteria, and expected

performance characteristics of the plant structures, systems, and components

important to safety to determine that they-are in accord with the Commission's General Design Criteria (GDC), Quality Assurance Criteria (QAC),

Regulatory Guides, and other appropriate rules, codes, and standards, and

that any departures from these criteria, codes, and standards have been

identified and justified.

(3) The expected response of the Watts Bar facility to various anticipated

operating transients and to a broad spectrum of postulated accidents.

Based on this evaluation, NRC determined that the potential consequences

of a few highly unlikely postulated accidents (design-basis accidents,

DBAs) would exceed those of all other accidents considered. The staff

performed conservative analyses of these DBAs to determine that the calculated potential offsite radiation doses that might result, in the very

unlikely event of their occurrence, would not exceed-the Commission's

guidelines for site acceptability given in 10 CFR 100.

(4) TVA's engineering and construction organization, plans for the conduct of

plant operations (including the organizational structure and the general

qualifications of operating and technical support personnel), the plans

for industrial security, and the planning for emergency actions to be

taken in the unlikely event of an accident that might affect the general

public, to determine that the applicant is technically qualified to operate the facility safely.

(5) The design of the systems provided for control of the radiological effluents from the Watts Bar facility to determine that these systems are capable of controlling the release of radioactive wastes from the facility

within the limits of the Commission's regulations in 10 CFR 20, and that

the equipment provided is capable of being operated by the applicant in

such a manner as to reduce radioactive release to levels that are as low

as reasonably achievable (ALARA) within the context of the Commission's

regulations in 10 CFR 50, and to meet the dose design objectives of

Appendix I to Part 50.

(6) TVA's quality assurance program for the operation of the facility to ensure

that the program complies with the Commission's regulations in 10 CFR 50,

Watts Bar SER 1-11
and that the applicant will have proper controls over the Watts Bar operations so that there is reasonable assurance that the facility can be operated safely and reliably.

1.6 Modifications to the Watts Bar Facility During the Course of NRC Review

During the review, the NRC staff met a number of times with the applicant's

representatives, contractors, and consultants to discuss various technical

matters related to the facility. Also, the staff made a number of site visits

to assess specific safety matters related to the station. The applicant made a

number of changes to the facility design as a result of NRC review. The staff

reviewed these design changes as well. Specific details concerning these changes

are included in amendments to the FSAR and in appropriate subsections of this

report.

1.7 Summary of Outstanding Issues

As a result of the NRC review of the safety aspects of the Watts Bar application, a number of *items remain outstanding at the time of issuance of this

report. Because the staff has not completed its review and reached final

positions in these areas, NRC considers these issues to be open. The staff

review of these items will be completed before a decision is made on issuing

an Operating License and will be reported in a supplement to this report. The

open items, with appropriate references to subsections of this report, are

listed below.

(1) Potential for liquefaction beneath ERCW pipelines and Class 1E electrical

conduit (2.5.4.4)

(2) Buckling loads on Class 2 and 3 supports (3.9.3.4)

(3) Preservice and inservice pump and valve test program (3.9.6)

(4) Seismic and environmental qualification of equipment (3.10, 3.11)

(5) Preservice and inservice inspection program (5.2.4 & 6.6)

(6) Pressure-temperature limits for Unit 2 (5.3.2, 5.3.3)

(7) Model D-3 steam generator preheater tube degradation (5.4.2.2)

(8) BTP-CSB 6-4 and containment isolation dependability (II.E.4.2) (6.2.4)

(9) H2 analysis review (6.2.5)

(10) Safety valve sizing analysis (WCAP-7769) (6.3)

(11) Compliance of proposed design change to the offsite power system to GDC 17

and 18 (8.2.2.2, 8.2.3)

(12) Fire Protection Program (9.5.1)

(13) Quality classification of diesel generator auxiliary system piping and-.

components (9.5.4-9.5.8)

(14) Diesel generator auxiliary systems design deficiencies (9.5.5, 9.5.7)-

(15) Physical Security Plan (13.6)

(16) Boron dilution event (15.2.4.4)

(17) Q-List (17.4)

1.8 Confirmatory Issues

At this point in the review there are some items that have essentially been

resolved to the staff's satisfaction, but for which certain confirmatory information has not yet been provided by the applicant. In these instances, the

applicant has committed to provide the confirmatory information in the near

Watts Bar SER 1-12
future. If staff review of this information does not confirm preliminary

conclusions on an item, that item will be treated as open and NRC staff will

report on its resolution in a supplement to this report.

The confirmatory items, with appropriate references to subsections of this

report, are:

(1) Design basis groundwater level for the ERCW pipeline (2.4.8)

(2) Material and geometric damping effect in SSI analysis (2.5.4.2)

(3) Analysis of sheetpile walls (2.5.4.3)

(4) Design differential settlement of piping and electrical components between

rock support structures (2.5.4.3)

(5) Upgrading ERCW system to seismic Category I (3.2.1, 3.2.2, 9.2.1)

(6) Seismic classification of structures, systems, and components important to

safety (3.2.1)

(7) Tornado missile protection of deisel generator exhaust (3.5.2, 9.5.4.1,

and 9.5.8)

(8) Steel containment building buckling research program (3.8.1)

(9) Pipe support baseplate flexibility and its effects on anchor belt loads

(IE Bulletin 79-02) (3.9.3.4)

(10) Thermal performance analysis (4.2.2)

(11) Cladding collapse (4.2.2)

(12) Fuel rod bowing evaluation (4.2.3)

(13) Loose parts monitoring system (4.4.5)

(14) Installation of residual heat removal system flow alarm (5.4.3)

(15) Natural circulation test (5.4.3)

(16) Dump valve testing (5.4.3)

(17) Protection against damage to containment from external pressure (6.2.1)

(18) Designation of containment isolation valves for main and auxiliary

feedwater lines and feedwater bypass lines (6.2.4)

(19) Compliance with GDC 51 (6.2.7)

(20) Insulation survey (sump debris) (6.3.3)

(21) Safety system set point methodology (7.1.3.1)

(22) Steam generator water level reference leg (7.2.5.9)

(23) Containment sump level measurement (7.3.2)

(24) IE Bulletin 80-06 (7.3.5)

(25) Overpressure protection during low temperature operation (7.6.5)

(26) Availability of offsite circuits (8.2.2.1)

(27) Nonsafety loads powered from the Class 1E ac distribution system (8.3.1.1)

(28) Low and/or degraded grid voltage condition (8.3.1.2)

(29) Diesel generator reliability qualification testing (8.3.1.6)

(30) Diesel generator battery system (8.3.2.4)

(31) Thermal overload protective bypass (8.3.3.1.2)

(32) Sharing of dc and ac distribution systems and power supplied between

Units I and 2 (8.3.3.2)

(33) Sharing of raceway systems between units (8.3.3.2)

(34) Testing Class 1E power systems (8.3.3.5.2)

(35) Evaluation of penetrations capability to withstand failure of overcurrent

protection device (8.3.3.6)

(36) Missile protection for diesel generator vent line (9.5.4.2)

(37) Component booster pump relocation (9.2.2)

(38) Electrical penetrations documentation (9.5.1.3)

(39) Compliance with NUREG/CR-0660 (9.5.4.1)

Watts Bar SER 1-13
(40) No-load, low-load, and testing operations for diesel generator (9.5.4.1)

(41) Initial test program (14.0)

1.9 License Conditions

There are several issues for which a license condition may be desirable to

ensure that staff requirements are met during plant operation. The license

condition may be in the form of a condition in the body of the operating

licenses, or a limiting condition for operation in the Technical Specifications appended to the licenses. These currently defined items, with appropriate references to subsections of this report, are:

(1) Relief and safety valve testing (II.D.1) (3.9.3.3, 5.2.2)

(2) Preservice/inservice testing of pumps and valves (3.9.6)

(3) Detectors for inadequate core cooling (II.F.2) (4.4.8)

(4) Inservice inspection program (5.2.4, 6.6)

(5) Installation of reactor coolant vents (II.B.1) (5.4.5)

(6) Accident monitoring instrumentation (II.F.1) (6.2.1, 6.2.5, 11.7.1, 12.7.2)

(7) Modifications to chemical feedlines (6.2.4)

(8) Containment isolation dependability (II.E.4.2) (6.2.4)

(9) Hydrogen control measure (II.B.4)(6.2.5)(App.C)

(10) Status monitoring system (7.7.2)

(11) Installation of acoustic monitoring system (II.D.3)(7.8.1)

(12) Diesel generator reliability qualification testing at normal operating

temperatures (8.3.1.6)

(13) DC monitoring and annunciation (8.3.2.2)

(14) Possible sharing of dc control power to ac switchgear (8.3.3.2.4)

(15) Testing of associated circuits (8.3.3.3.6)

(16) Testing of non-Class 1E cables (8.3.3.3.c)

(17) Low-temperature overpressure protection/power supplies for pressurizer

relief valves and level indicators (I.G.1) (8.3.3.4)

(18) Testing of reactor coolant pump breakers (8.3.3.6)

(19) Postaccident sampling system (II.B.3) (9.3.2)

(20) Fire protection program (9.5.1)

(21) Performance testing for communications systems (9.5.2)

(22) Diesel generator reliability (9.5.4.1)

(23) Secondary water chemistry monitoring and control program (10.3.4)

(24) Primary coolant outside containment (III.D.1.1) (11.7.2)

(25) Independent safety engineering group (I.B.1.2) (13.1)

(26) Use of experienced personnel during startup (13.1.2)

(27) Emergency preparedness (III.A.1.1, III.A.1.2, III.A.2) (13.3)

(28) Review of power ascension'test procedures and emergency operating procedures by NSSS Vendor (I.C.7)(13.5.2)

(29) Modifications to emergency operating instructions (I.C.8) (13.5.2)

(30) Report on outages of emergency core cooling system (II.K.3.17) (13.5.3)

(31) Initial test program (14)

(32) Effect of high pressure injection for small break LOCA with no auxiliary

feedwater (II.K.2.13) (15251)

(33) Voiding in the reactor coolant system (II.K.2.17) (13.5.2)

(34) PORV isolation system (II.K.3.1, II.K.3.2) (15.5.3)

(35) Automatic trip of the reactor coolant pumps during a small-break LOCA

(II.K.3.5) (15.5.4)

(36) Revised small-break LOCA analysis (II.K.3.30, II.K.3.3.31) (15.5.5)

(37) Control room design review (I.D.1) (18.0)

Watts Bar SER 1-14
1.10 Unresolved Safety Issues

Section 210 of the Energy Reorganization Act of 1974, as amended, reads as

follows:

Unresolved Safety Issues Plan

Section 210. The Commission shall develop a plan for providing for

specification and analysis of unresolved safety issues relating to

nuclear reactors and shall take such action as may be necessary to

implement corrective measures with respect to such issues. Such

plan shall be submitted to the Congress on or before January 1, 1978

and progress reports shall be included in the annual report to the

Congress thereafter.

In response to this reporting requirement, the NRC provided a report to the

Congress, NUREG-0410, in January 1978, which described the generic issues program of the Office of Nuclear Reactor Regulation (NRR) that had been implemented

early in 1977. The NRC program described in NUREG-0410 provides for the identification of generic issues, the assignment of priorities, the development of

detailed task action plans to resolve the issues, the projections of dollar and

personnel costs, continuing high-level management oversight of task progress,

and public dissemination of information related to the tasks as they progress.

Since the issuance of NUREG-0410, each annual report has described NRC progress

in resolving these issues.

The staff continuously evaluates the safety requirements used in its review

against new information as it becomes available. In some cases, the staff takes

immediate action or interim measures to ensure safety. In most cases, however,

the initial assessment indicates that immediate licensing actions or changes in

licensing criteria are not necessary. In any event, further study may be deemed

appropriate to make judgments as to whether existing requirements should be

modified. The issues being studied are sometimes called generic safety issues

because they are related to a particular class or type of. nuclear facility.

A discussion of these matters and the NRC program for the resolution of these

generic issues is provided in Appendix C to this report. The appendix includes

references to sections of this report for more specific discussions concerning

this facility.

Watts Bar SER 1-15

2 SITE CHARACTERISTICS

2.1 Geography and Demography

The July 1981 edition of the SRP (NUREG-0800) includes Chapter 2, Site

Characteristics. Watts Bar Units 1 and 2 were reviewed in accordance with

Sections 2.1.1, 2.1.2, and 2.1.3 of the SRP. The results of this review are

contained in Section 2.1 of this SER.

2.1.1 Site Location and Description

The Watts Bar nuclear plant is located on a 1770-acre site in Rhea County in

southeastern Tennessee. The site is located on the west bank of the Chickamauga Lake, a reservoir formed between the Chickamauga Dam and the Watts Bar

Dam on the Tennessee River. The plant is about 45 mi north-northeast of the

center of Chattanooga, Tennessee and 40 mi southwest of Oak Ridge, Tennessee.

Figures 2.1 and 2.2 show the general region of the Watts Bar site, while

Figure 2.3 shows the immediate site area.

The topography of the site varies from about 700 ft mean sea level (msl) adjacent to the Chickamauga Reservoir to 840 ft msl in the western portion of

the plant boundary. Hills on the opposite side of the reservoir rise from 700

ft msl to about 950 ft msl. The plant grade is 720 ft msl.

I0

The coordinates of Watts Bar Unit 1 are 350 36' 10.430" north latitude and 840

47' 24.267" west longitude. The universal transverse mercator coordinates are

3,941,954,20 m north and 700,189,94 m east.

2.1.2 Exclusion Area Authority and Control

The applicant has defined the exclusion area as a circular area within a 3940-ft

(1200-m) radius centered on a point 20 ft from the north wall of the turbine

building along the building centerline. The plant site boundary and the exclusion area are shown in Figure 2.4, which also shows the locatior within the

site of the principal plant structures. The exclusion area consists of the

site property plus small inlets of the Chickamauga Lake which penetrate

the site property and part of the Chickamauga Reservoir on the southeastern.

side of the plant. All of the land, including the mineral rights, is owned by

the United States and is in the custody of TVA.

There are no residences or designated recreational areas withinithe exclusion

area. Two cemeteries, the Leuty cemetery and an unnamed cemetery, are within

the exclusion area. Both are owned by the United States and are under the control of TVA. All persons buried in these cemeteries were reinterred in improved

cemeteries accessible by road. Access to both cemeteries is cohtrolled by the

applicant and is under the same restrictions as other portions of the exclusion

area.

Watts Bar SER 2-1
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Watts Bar SER 2-2
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Watts Bar SER 2-3
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atts Bar SER
No railroads or public highways traverse the exclusion area. A portion of the

Chickamauga Reservoir, which is navigable by commercial barge traffic, traverses the eastern portion of the exclusion area. In the event of emergency,

control of the movement of people on the waterway will be carried out by the

TVA security force at-the plant. (See Section 13.3 of this report for a discussion of emergency planning.) The Watts Bar coal-fired steam plant (owned

and operated by the TVA) is located within the exclusion area about 0.7 mi

northeast of the nuclear units. In an emergency, TVA also will exercise control over the movement of the employees of the steam plant.

On the basis of the ownership of the land within the exclusion area, including

the mineral rights, and because suitable arrangements have been made to control

traffic on that portion of Chickamauga Reservoir within the exclusion area,

the staff concludes that the applicant has the authority to determine all activities within the exclusion area, as required by 10 CFR 100. Moreover, based on

separation distance, the staff concludes that the fossil fuel plant activities

within the exclusion area will not interfere with normal operation of the nuclear

units.

2.1.3 Population Distribution

The resident population in the vicinity of the Watts Bar site is shown as a

function of distance in Table 2.1.

Table 2.1 Resident population in Watts Bar vicinity

0-1 0-2 0-3 0-4 0-5 0-10

Year mi mi mi mi mi mi

1970 35 190 540 1150 1750 10,735

1980 45 210 600 1295 2010 12,335

2020 55 225 650 1420 2280 14,225

The largest community Within 10 mi of the site is Spring City, whose nearest

boundary is about 7 mi northwest of the site and which'had a 1980 population

of 1944 persons. The applicant has chosen a low-population zone (LPZ) radius

of 3 mi. The 1980 population within the LPZ was 600 persons and is projected

to reach 650 persons in the year 2020.

The major transient population within 5 mi of the site is made up of visitors

at the Watts Bar resort and Watts Bar Dam reservation. The maximum 1980 population at these recreational areas is estimated to be 380 and 1400 persons,

respectively.

The applicant has indicated that the nearest densely populated center of

25,000 or more persons (as defined in 10 CFR 100) is the city of Oak Ridge,

Tennessee, which had a 1980 population of 27,552, and whose nearest boundary

is about 40 mi northwest of the Watts Bar site. Between the years 2000 and

2010, the population center is expected to shift to Athens, Tennessee (1980

Watts Bar SER 2-6
population 12,040). Athens is approximately 15 mi southeast of the site.

This distance is such that the population center distance will be at least

1-1/3 times the LPZ radius, as required by 10 CFR 100, and will not change

the 3-mi LPZ radius specified by the applicant.

The staff made an independent estimate of the 1970 population within a 50-mi

radius of the Watts Bar site based on the Bureau of the Census data. The

staff's value.of 679,280 persons is in good agreement with the applicant's value of 654,665 persons. The applicant projects that this population will

grow to about 905,370 by the year 2020. This represents a growth rate of 7.7

percent per decade for the period 1970 to 2020. The staff compared the appli- cant's population projections to those of the Bureau of Economic Analysis-for

economic areas 48, 49, and 50 (an area that encompasses Chattanooga and Oak

Ridge, Tennessee). This comparison showed that the applicant's projection of

7.7 percent per decade is conservative compared to the regional growth rate of

3.2 percent per decade using the Bureau of Economic Analysis areas.

2.1.4 Conclusion

On the basis of (1) the 10 CFR 100 definitions of the exclusion area, lowpopulation zone, and population center distance; (2) the staff analysis of the

onsite meteorological data from which the relative concentration factors (X/Q)

were calculated (see Section 2.3 of this report), and (3) calculated potential

radiological dose consequences of design-basis accidents (see Section 15 of

this report), the staff concludes that the exclusion area, LPZ, and population

center distance meet the criteria of 10 CFR 100 and are acceptable.

2.2 Nearby Industrial, Transportation, and Military Facilities

The July 1981 edition of SRP (NUREG-0800) includes Chapter 2, Site

Characteristics, as well as Chapter 3, Design of Structures, Components, and

Systems. The Watts Bar Units 1 and 2 were reviewed in accordance with

Sections 2.2.1, 2.2.2, 2.2.3, 3.5.1.5, and 3.5.1.6 of the SRP. The results of

this review are contained in Section 2.2 of this SER.

2.2.1 Transportation Routes

ýýThe nearest roads and their closest locations to the site are State Route 68,

which passes about 1 mi north of the plant, and the Old Dixie Highway, about 2

mi west of the plant. State Route 58 is about 3 mi east of the plant on the

opposite shore of the Chickamauga Reservoir. The nearest major highway is U.S.

Highway 27 (State Route 29), which passes about 6.5 mi west of the site. The

applicant indicated that the only known shipments of toxic materials within a

5-mi radius by barge and road routes are small quantities of chemicals shipped

to the Watts Bar steam plant on Route 68. The applicant states that the quan- tity of each shipment is small and would not produce a hazard to control room

personnel at the Watts Bar nuclear plant.

The separation distances between the Watts Bar nuclear plant and Old Dixie

Highway, State Route 58, and U.S. Highway 27 are sufficient to preclude adverse

effects on the plant from accidents on these routes.

Watts Bar SER '2-7
The nearest railroad is a spur line of the CNO&TP (Southern Railway System),

which runs parallel to State Route 68 about I mi north of the plant and which

terminates at the Watts Bar steam plant. This spur is used to ship heavy components to both the steam plant and the nuclear plant. The railroad will

continue to be. used as a coal supply route for the steam plant. The nearest

major railroad is the main line of the CNO&TP which runs parallel to U.S.

Highway 27 about 6.5 mi west of the site.

The Chickamauga Reservoir of the Tennessee River, which forms, the eastern boundary of the site, is used for barge transportation of commercial cargo. The

intake structure is protected against a potential barge impact by virtue of

its location approximately 800 ft from the edge of the Chickamauga Reservoir.

The intake structure is connected to the reservoir by an excavated channel that

is perpendicular to the main flow of the river. On the basis of the intake

structure location, the staff concludes that it is very unlikely that a drifting or runaway barge will impact the intake structure.

Fuel oil is shipped by barge past the Watts Bar site. In case of a fuel oil

barge accident, fire and dense smoke may result. The intake pumping station

is protected against fire by virtue of design and location. The pump suction

is taken from the bottom of the channel. Thus, even if fuel oil from a spill

should reach the intake station, entry of oil into the intake is unlikely

because the oil will float, and the pump suction is about 10 ft below the water

surface. All pumps and essential cables and instruments are protected from

fire because they are enclosed within concrete walls. The control room air

intake system will automatically isolate (see Section 6.4 of this report) when

it detects high smoke concentrations in the outside air supply. The staff concludes that barge accidents involving spillage of fuel oil and resulting fires

will not adversely affect the safe operation of the plant.

The control room air intake system will also automatically isolate (see Section 6.4 of this report) when it detects high chlorine concentrations in the

outside air supply. Consequently, accidents involving chlorine release will

not adversely affect the safe operation of the plant.

The Tennessee River barge traffic passing the Watts Bar site is expected to

remain at about its present level (total of 760,000 tons shipped in 1975) until

the possible future completion of the Tennessee-Tombigbee waterway, when the

total tonnage shipped is expected to increase to about 3 to 4 million tons per

year., Because this may significantly increase the frequency as well as the

type of hazardous materials shipped past the site, the staff will require that

the applicant provide a revised survey and analyses, as appropriate, of barge

traffic going past the Watts Bar site when the Tennessee-Tombigbee waterway is

completed. The staff concludes that the present barge traffic past the site

will not adversely affect the safety of the Watts Bar nuclear plant.

2.2.2 Nearby Facilities

The only significant nearby industrial facility is the Watts Bar steam plant,

which is at the edge of the exclusion area, about 0.7 mi northeast of the plant,

and employs about 100 persons. Because of the separation distance and nature

Watts Bar SER 2-8
of activities, the Watts Bar steam plant will not adversely affect the safety

of the Watts Bar nuclear plant.

There are no pipelines or military facilities within 5 mi. There are no airportswithin 10 mi of the site. The nearest airport is the Mark Anton airport,

about 12 mi southwest of the site. This field has lights and a hard-surface

runway 4500 ft long, but it has no commercial facilities. The nearest airport

with Commercial facilities is Lovell Field, about 45 miles south-southwest of

the site. Several low-level Federal airways pass near the site. V51 passes

about 1 mi northeast of the site, while V51W passes about 5 mi southwest and

V115W passes about 4 mi southeast of the site. These are shown in Figure 2.5.

At the request of the staff, the applicant examined'traffic along these routes

and performed an analysis of the probability of an aircraft impacting the plant.

As a result of this analysis, the applicant concluded that the probability of

an aircraft crash causing radiological consequences in excess of the guideline values given in 10 CFR 100 is within the acceptance criteria of SRP (NUREG-0800)

Section 2.2.3 (less than about 10-7 per year), and is, therefore, acceptable.

After reviewing this analysis, the staff concurs with this conclusion.

Several low-altitude military training routes, namely VR1052/1056 and SR102,

pass within 5 mi of the Watts Bar nuclear plant. Because the Department of

Defense has agreed to relocate such routes so that they are at least 5 mi from

an operating nuclear power plant, and because the staff has found such distances

acceptable, the staff will require satisfactory confirmation that these routes

have been relocated before fuel loading for the plant.

2.2.3 Conclusions

The staff review has been conducted based upon the criteria given in GDC 4 and

in SRP Section 2.2.3. The staff concludes, subject to a satisfactory resolution

of the above-mentioned items, that the plant is adequately protected and can

be operated with an acceptable degree of safety as a result of activities at

nearby transportation, industrial, and military facilities.

2.3 Meteorology

Evaluation of regional and local climatological information, including extremes

of climate and severe weather occurrences which may affect the design and siting

of a nuclear plant, is required to ensure that the plant is designed and operated

within the requirements of Commission regulations. Information concerning atmospheric transport and diffusion characteristics of a nuclear power plant site

is required for a determination that radioactive effluents from postulated accidental releases, as well as routine operational releases, are within Commission

guidelines. SER Sections 2.3.1 through 2.3.5 have been prepared in accordance

with the SRP (NUREG-0800) utilizing information presented in FSAR Section 2.3 as

well as responses to requests for additional information.

2.3.1 Regional Climatology

The climate of eastern Tennessee is generally continental, moderated by the

anticyclonic circulation of the Azores-Bermuda high pressure system, particularly in the late summer and early fall. The mean annual temperature in the

Watts Bar SER 2-9
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area is about 60'F with monthly mean temperatures ranging from about 40*F in

January to about 79°F in July. Annual precipitation is about 52 in.

The site is principally affected by storms originating in the Southwest and

Gulf Coast areas tracking northeastward along the western side of the Appalachian Mountains. About 76 thunderstorms can be expected to occur on about

56 days per year, being most frequent from May through August. Considering

the frequency of thunderstorms in the area, the applicant has estimated an

annual frequency of about 3 lightning strikes to ground per square kilometer.

Hail often accompanies severe thunderstorms. Hail with diameters 3/4 in. or

larger occurs about once each year.

Because the site is located east of the major tornado "alleys" of the midwest,

tornadoes are not frequent. The applicant has reported that in the period 1916

through November 1975 no tornadoes were observed in the county where the plant

is located. Examination of tornado occurrences in larger areas (for example,

a 1-degree latitude-longitude square containing the site or 10,000 mi2 area

containing the site) for periods of record 1955-1967, 1950-1976, and 1953-1974

indicates a computed recurrence interval for a tornado at the plant site of

about 1300 years. The design-basis tornado characteristics selected by the

applicant are: a rotational speed of 300 mph; a translational speed of 60 mph;

and a total pressure drop of 3 psi occurring at a rate of 1 psi/sec. These

tornado characteristics are somewhat different than those recommended in Regulatory Guide 1.76, which are: a rotational speed of 290 mph; a translational

speed of 70 mph; and, a total pressure drop of 3 psi occurring at a rate of

2 psi/sec. The maximum wind speed (360 mph) and the total pressure drop (3 psi)

are the same for both design-basis tornadoes; however, the rate of pressure

drop of the design-basis tornado selected by the applicant is one-half the

value recommended in Regulatory Guide 1.76. The applicant has stated that the

rate of pressure drop does not affect the design of nonvented and totally vented

structures. The applicant has reanalyzed the partially vented auxiliary building

with respect to the higher rate of pressure drop and determined that the limiting structural element was adequate. The acceptability of the applicant's

design of Category I structures with respect to design-basis tornado characteristics and load combinations is discussed in *Section 3.3.2 of this report.

The "fastest mile" wind speed reported at Chattanooga was 82 mph (March 1947).

The "fastest" wind speed at a height of 30 ft with a return period of 100 years

for theWatts Bar area is about 95 mph. The applicant has selected a wind speed

of 95 mph to be considered in the design of the shield building.

Another environmental loading to be considered in the design of safety-related

structures is the maximum weight of precipitation that can accumulate on roofs

due to a combination of snow, ice, and rain. The applicant has examined severe

and extreme loadings due only to accumulated snowfall rather than the more

likely (for eastern Tennessee) wintertime precipitation mixture of snow, ice,

and rain. Ice storms, which can plug drains and scuppers as well as disrupt

offsite power, are not uncommon in eastern Tennessee. For example, the applicant has presented information that ice accumulations of 0.6 in. have a return

period of 10 years for the Watts Bar area. For an operating basis load, the

applicant has assumed a value of 14 psf to be the weight of the 100-year return

period snowpack. To determine the design-basis operating load, the applicant

added the weight of a 28-in. snowfall reported at Westbourne, Tennessee, in

February 1960 (estimated to be 24 psf) to the weight of the 100-year return

Watts Bar SER 2-11
period snowpack. Although the extreme environmental loading of 38 psf proposed

by the applicant may be conservative for the maximum weight only caused by

snowfall, it is not clear that 38 psf is the maximum weight of precipitation

that can accumulate on roofs as a result of a combination of snow, ice, and

rain. However, the applicant has also stated that the minimum design live load

for the roof of any Category I structure is 50 psf. An extreme environmental

loading as a result of a combination of snow, ice, and rain is not likely to

exceed 50 psf. The applicant has further stated that for any severe environmental conditions (for example, the 48-hour PMWP) an increase to 0.9 yield is

justified, increasing the live load capacity to 300 psf. Roof loads as a

result of precipitation also are discussed in Section 2.4.2, and load combinations are discussed in Section 3.8.1 of this SER.

Large-scale episodes of atmospheric stagnation are not uncommon in the region,

particularly in late summer and early fall. In the period 1936-1970,' about 70

atmospheric stagnation cases totalling about 275 days were reported in the area.

As discussed above, the staff has reviewed available information relative to

the regional meteorological conditions of importance to the safe design and

siting of this plant. Based on this review, the staff concludes that the applicant has identified and considered appropriate regional meteorological conditions

in the design and siting of this plant, and, therefore, meets the requirements

of 10 CFR 100.10 and GDC 2. The design-basis tornado characteristics selected

by the applicant are somewhat different than the position set forth in Regulatory

Guide 1.76, "Design Basis Tornado for Nuclear Power Plants"; however, the

differences are not considered significant in the determination of an acceptable design-basis tornado for missile generation. Therefore, the requirement

of GDC 4 is met.

2.3.2 Local Meteorology

Climatological data from Chattanooga, Knoxville, Decatur, and Watts Bar Dam,

in addition to available onsite data, have been used to assess the local

meteorological characteristics of the site.

Extreme maximum temperatures of 108*F and 106°F and extreme minimum temperatures of -20 and -10OF have been observed at Decatur and Chattanooga, respectively. Mean monthly precipitation at Watts Bar Dam varies from about 5.6 in.

in March to about 2.9 in. in October. The maximum monthly precipitation

reported at Watts Bar Dam was 14.8 in., while the minimum monthly precipitation

reported is 0.0 in. The maximum 24-hour rainfall reported at Chattanooga was

about 7.6 in., while at Watts Bar Dam the maximum 24-hour rainfall reported

was about 5.3 in. Annual average snowfall in the area is about 9 in., although

about 18 in. of snow has been reported in 24 hours in Knoxville. The site area

is generally humid, particularly in the summer. Heavy fog can be expected on

about 35 days annually.

The plant is located along a stretch of the Tennessee River with a valley orientation generally southwest to northeast. The terrain rises rapidly away from

the plant site in most directions, with increases-on the order of several hundred feet within 2 mi of the plant. The local airflow pattern is bimodal,

reflecting channelling up and down the river valley. Wind data from the 30-ft

level of the onsite meteorological tower for the period July 1, 1973 through

June 30, 1975 indicate winds from the southwest and south-southwest occur about

Watts Bar SER 2-12
25 percent of the time while winds from the north-northeast and northeast occur

about 16 percent of the time. Winds from the east-southeast and southeast occur

least frequently, totalling only about 3 percent of the time. A similar pattern

is evident for winds at the 300-ft level, with winds from the southwest and

south-southwest occurring almost 30 percent of the time with winds from the

north-northeast and northeast occurring about 25 percent ofthe time. Winds

from the south-southwest at the 30-ft level have been observed to persist for

37 consecutive hours, and winds from the north-northeast have been observed to

persist for 26 consecutive hours.

The median wind speed at the 30-ft level is about 3 mph, and more than 95 percent of the winds occur with speeds less than 12 mph. Calm conditions (defined

as an hourly average wind speed below the starting threshold of the anemometer)

were reported only about 0.4 percent of the time for the 2-year period July

1973 to June 30, 1975. This frequency of observed calm conditions for this

2-year period of record contrasts markedly with the observed frequency of calm

conditions for data collected previously at the site. For the period July 1,

1971 to June 1972, calm conditions were reported for 11.6 percent of the time.

The applicant believes that the higher frequency of calm conditions and low

wind speed conditions in general for this period of record is a consequence of

tower location. A temporary onsite meteorological tower was located in a slight

topographic depression for the data collection period July 1971 to June 1972,

apparently resulting in sharply reduced wind speeds. A permanent onsite

meteorological tower was installed in a different location in May 1973, and

the temporary tower was decommissioned in September 1973. The applicant performed

a comparison of data collected while both towers were in operation from June

to September 1973 and concluded that the temporary tower was located in an area

affected by low-level "drainage" airflow (principally a nighttime phenomenon

where differential cooling of the ground surfaces causes cooler, more dense

air to flow towards lower terrain) resulting in abnormally high frequencies of

low wind speeds and very stable atmospheric conditions.

Inversions predominate at the Watts Bar site. Slightly stable (Pasquill type

"E"), moderately stable (Pasquill type "F") and extremely stable (Pasquill

type "G") conditions occur about 16 and 9 percent, respectively.

As discussed above, the staff has reviewed available information relative to

local meteorological conditions of importance to the safe design and siting of

this plant. Based on this review, the staff concludes that the applicant has

identified and considered appropriate local meteorological conditions in the

design and siting of this plant, and, therefore, meets the requirements of

10 CFR 100.10 and GDC 2.

2.3.3 Onsite Meteorological Measurements Program

The onsite meteorological measurements program at the Watts Bar site was initiated in June 1971 with the installation of a 130-ft temporary tower used to

provide data in support of the Construction Permit application. A permanent

296-ft meteorological tower was installed in May 1973 in a location about 2600 ft

west-southwest of the Unit I reactor building. Two 470-ft natural draft

cooling towers are located about 3087 ft northeast of the meteorological tower.

Wind speed and direction are measured at the 33-, 150-, and 296-ft levels of

the permanent tower; dry bulb temperature is measured at the 33-, 150-, and

296-ft levels; dewpoint temperature is measured at the 33-ft level; and solar

Watts Bar SER 2-13
radiation and precipitation are measured at about 3 ft above the ground.

Atmospheric stability is defined by the vertical temperature gradient determined between the 33-ft and 150-ft levels and between the 33- and 296-ft levels.

Data are recorded digitally. Apparently, before January 31, 1975, temperature

values were recorded only once per hour. After that time, temperature values

were recorded once per minute and hourly averages were determined based on 60

values. The representativeness of hourly averages based on just one value is

suspect. The impact of this averaging technique on the determination of

vertical temperature gradient is unknown; however, the stability distribution

for the period July 1973 to June 1975 appears reasonable. The applicant also

has examined the effects of the large natural draft cooling towers on wind

speed and wind direction measurements and concluded that no effects are evident

to date,.although the cooling towers are not in operation.

The applicant has provided 2 years (July 1, 1973 through June'30, 1975) of

meteorological data on magnetic tape. A joint frequency distribution for this

period was compiled using wind speed and direction from the 33-ft level, and

using vertical temperature gradient between the 33- and 150-ft levels as the

indicator of atmospheric stability. The joint data recovery for wind speed,

wind direction, and atmospheric stability was 93 percent.

The onsite meteorological measurement system now conforms to the guidance of

Regulatory Guide 1.23, "Onsite Meteorological Programs." The program in operation for the period July 1, 1973 to June 30, 1975 appears to have provided

adequate data to represent onsite meteorological conditions used for evaluations required by 10 CFR 100.10 and 10 CFR 50, Appendix I. The onsite data

appear to provide a reasonable basis for making estimates of atmospheric

dispersion conditions for assessing consequences of design basis accident and

routine releases from the plant.

To address the meteorological requirements for emergency preparedness planning

outlined in 10 CFR 50.47 and Appendix E to 10 CFR 50, the applicant will be

required to upgrade the operational meteorological measurements program to meet

the criteria in NUREG-0654, Appendix 2, "Criteria for Preparation and Evaluation

of Radiological Emergency Response Plans and Preparedness in Support of Nuclear

Power Plants." The upgrades must be in accordance with the schedule of

NUREG-0737, Item III.A.2. Although the effects of the cooling towers on

measurements made on the meteorologicalltower appear to be negligible, the

representativeness of the meteorological measurements with respect to airflow

from likely release points towards and around the cooling towers may be questionable. Distortions in airflow trajectories because of the presence of the

cooling towers should be considered in the development of the upgraded meteorological measurements program and modeling capability.

2.3.4 Short-Term (Accident) Diffusion Estimates

Short-term (less than 30 days) accidental releases from buildings and vents

were evaluated using the direction-dependent atmospheric dispersion model

described in Regulatory Guide 1.145 ("Atmospheric Dispersion Models for Potential

Accident Consequences Assessments of Nuclear Power Plants"), with consideration

of increased lateral dispersion during stable conditions accompanied by low

wind speeds. Two years (July 1, 1973 to June 30, 1975) of onsite data were

used for this evaluation. Wind speed and wind direction were measured at the

33-ft level and atmospheric stability was defined by the vertical temperature

Watts Bar SER 2-14
gradient determined between the 33- and 150-ft levels. A ground-level release

with a building wake factor, cA, of 850 m2 was assumed. The relative concentration (x/Q) for the 0- to 2-hour time period was determined to be 3.6 x 10-4

sec/m3 at a distance of 3900 ft from the reactor buildings in both the southeast and south-southeast sectors. The x/Q values for appropriate time periods

at the outer boundary of the low population zone (4828 m) are:

Time Period X/Q (sec/m3 )

B0-8 hours 5.0 x 10-s

8-24 hours 3.3 x 10-5

1-4 days 1.3 x 10-5

4-30 days 3.7 x 10-6

The applicant calculated short-term X/Q values according to the previously

accepted technique of a direction-independent atmospheric dispersion model

described in Regulatory Guide 1.4 ("Assumptions Used for Evaluating the Poten-

.tial Radiological Consequences of a Loss of Coolant Accident for Pressurized

Water Reactors") and SRP Section 2.3.4, Revision 0 (NUREG-75/087). Because the

model described in Regulatory Guide 1.145 provides a more realistic consideration of atmospheric dispersion during stable conditions accompanied by low wind

speeds (which predominate at the Watts Bar site) thereby reducing x/Q values,

the x/Q values calculated by the applicant for time periods less than 8 hours

are significantly more conservative than those calculated by the staff. For

example, the x/Q for the 0- to 2-hour time period at the exclusion area boundary

was calculated to be 1.1 x 10-3 sec/m3 , a factor of 3 higher than the value

calculated by the staff. The effect of increased lateral dispersion during

stable, low wind speed conditions is not as marked for longer time periods

and longer distances (for example, the LPZ distance compared to the exclusion

area boundary).

The staff's evaluation of dispersion, conditions for the CP application, made

using the direction-independent model with data from the temporary meteorological

tower, was higher than any values calculated for the OL application. For example,

the x/Q for the 0- to 2-hour time period for the CP review was determined to be

3.4 x 10-3 sec/m3 , almost 3 higher than the applicant's calculated value and

almost an order of magnitude higher than the staff's current calculation. The

reasons for the largedifferences are the improved data base (2 years of data

from a more representative tower location) and consideration of increased

lateral dispersion during low wind speed, stable conditions.

-The staff concludes that the atmospheric dispersion estimates provided by the

staff in this section are appropriate for assessments of the consequences of

radioactive releases for design-basis accidents in accordance with the requirements of 10 CFR 100.11. The staff concludes that the dispersion estimates

provided for the CP review although based on the best available data at the

time, were probably over conservative.

2.3.5 Long-Term (Routine) Diffusion Estimates

Annual average relative concentration (x/Q) and relative deposition (D/Q) values

were calculated using the constant mean wind direction model described in

Regulatory Guide. 1.111 ("Methods for Estimating Atmospheric Transport and

Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled

Reactors"). The same 2-year period of record as described in Section 2.3.4

Watts Bar SER 2-•15
was used for this evaluation. All releases were assumed to be at ground level

with mixing within the turbulent wake of plant structures. The results of the

constant mean wind direction model were adjusted to account for spatial and

temporal variations in airflow using the methodology described in NUREG-0324.

The staff concludes that representative atmospheric dispersion estimates have

been determined for demonstrating compliance with the numerical guides for doses

contained in 10 CFR 50, Appendix I. The assessment of the radiological impact

to man resulting from routine releases to the atmosphere is contained in the

staff',; environmental statement.

2.4 Hydrologic Engineering

2.4.1 Introduction

The staff has reviewed the hydrologic engineering aspects of the applicant's

design, design criteria, and design basis of safety-related facilities for the

Watts Bar plant. The acceptance criteria used as a basis for the staff evaluations are set forth in SRP Sections 2.4.1 through 2.4.14 (NUREG-0800). These

acceptance criteria include the applicable GDC, Reactor Site Criteria in 10 CFR 100,

and Standards for Protection Against Radiation in 10 CFR 20, Appendix B, Table II.

Guidelines for implementation of the requirements of the acceptance criteria

are provided in Regulatory Guides, ANSI standards, and BTPs in SRP Sections 2.4.1

through 2.4.14. Conformance to the acceptance criteria provides the bases for

concluding that the site and facilities meet the requirements of 10 CFR 20, 50

and 100 with respect to hydrologic engineering.

2.4.2 Hydrologic Description

The 1770-acre plant site is on the west bank of Chickamauga Lake, a reservoir

created by Chickamauga Dam, which is 57 mi downstream. The plant site is at

Tennessee river mile 528, some 1.9 mi downstream of Watts Bar Dam. Although

plant structures are generally located on what would normally be considered

high ground, design-basis flood conditions would inundate much of the site.

The elevations of plant features that could be affected by their submergence

during floods vary from 714.5 ft mean sea level (msl) for accesses to electrical conduits to 760.5 ft msl for emergency exits to the diesel generator building, which is above the design-basis flood level. Plant grade is elevation

728 ft msl. The plant has been designed to be shut down safely in the event of

a flood that could exceed plant grade.

The Tennessee River above the plant site drains 17,319 mi2 . At Watts Bar Dam

and Chickamauga Dam, the Tennessee River drains 17,310 and 20,790 mi 2 , respectively. Although precipitation is relatively evenly distributed throughout

the year, two general types of major flood-producing storms can cause severe

flooding; the cool season winter-type storm, and the warm season tropical-type

storm. Most floods in the vicinity of the plant site have, however, occurred

from January through mid-April and resulted from the cool season winter-type

storm.

Watts Bar Dam is composed of a 1234-ft-long earth embankment, a long (%1664-ft)

concrete section, and a 1339-ft-long saddle dam across a ground depression

southwest of the main structure. The normal reservoir elevation is 741 ft msl.

Watts Bar SER •2-16
Chickamauga Dam affects the water levels in Chickamauga Reservoir/Tennessee

River at the site. The normal reservoir level is elevation 682.5 ft msl. The

reservoir is about 1100 ft wide and 18 to 26 ft deep at the site at the normal

reservoir level.

There are 12 major upstream reservoirs in the TVA system, and 6 major upstream

reservoirs owned by the Aluminum Company of America. All 18 dams upstream were

considered in the TVA analysis offloods and water supply.

Flows in the Tennessee River at the site are largely controlled by Watts Bar

Dam. Since closure of the dam in 1942, the average daily stream flow has been

27,800 cubic feet per second (cfs). The maximum daily flow of record was

187,000 cfs and occurred in 1941. Minimum flows at the site also are controlled by the releases from Watts Bar Dam, and there have been periods of

several hours when no water has been released from either Watts Bar Dam,

Chickamauga Dam, or both. The lowest historical natural flow, or uncontrolled

flow, has been estimated to be 2,700 cfs. The historical water level at the

site was about elevation 697 ft msl.

Local drainage consists of nearby Yellow Creek, the plant area itself, and the

150-acre hillside area north of the plant. The flood potential from each source

also was considered by the applicant.

Both surface and groundwater are used locally by the public and industry. The

nearest downstream user of Tennessee River water is at Dayton, 24.2 mi downstream. There are 17 public and 8 industrial users of groundwater within 20 mi

of the site, and 51 wells within 2 mi of the site. The applicant has installed

three wells 2.5 mi northwest of the plant to supply at the rate of 16,000 gallons

per day (gpd), potable water, to the Watts Bar nuclear plant and to the Watts

Bar hydro plant and a nearby resort (200,000 gpd).

Cooling water is required for both nonsafety purposes and for vital equipment

necessary for plant shutdown. The intake flow requirement for the supply of

water for vital equipment (such as engineered safety features) is 75 cfs. The

cooling water is to be taken from the Tennessee River through an intake channel

and intake structure set some 800 ft into the river bank adjacent to the plant.

2.4.3 Flood Potential

The flood potential from several mechanisms has been analyzed by the applicant;

included were large- and small-scale precipitation events and seismic events

that could cause dam failures. Dam failures can occur from both severe precipitation and from seismic events.

The controlling set of conditions for design are those that could cause the

greatest water levels and associated impact forces. These conditions result

from a precipitation-induced probable maximum flood (PMF) in the Tennessee

River with concurrent failure of the Watts Bar Dam. The PMF would be produced

by a hypothetical storm centered upstream of Watts Bar Dam. The associated

peak flood flow rate at the plant was estimated to be 1,478,000 cfs. Maximum

water levels--including the effects of associated wind waves and a bore caused

by an assumed failure of Watts Bar Dam during the flood--were estimated to range

between elevations 739.4 and 740.1 ft msl, depending on the particular building

Watts Bar SER 2-17
in question. 'Additionally, flood conditions approachingthose of a severe

storm were determined to result from seismic events that are assumed to cause

dam failures.

There are 12 major dams (Watts Bar, Fort Loudoun, Watauga, South Holston,

Boone, Fort Patrick Henry, Cherokee, Douglass, Norris, Melton Hill, Fontana

and Tellico) above the Watts Bar plant whose failure could influence plant-site

flood levels. The applicant examined these dams individually and in groups to

determine if arbitrarily assumed seismic failures coincident with river flooding would create critical water levels at the plant. Specifically, the applicant examined two postulated combinations of natural events

(1) an earthquake of the severity of the operating basis earthquake (OBE) as

defined in Section 2.5 of this report, coincident with a flood discharge that is one-half of the PMF

(2) a safe shutdown earthquake (SSE) coincident with a 25-year flood

Neither of these conditions was estimated to result in water levels at the

plant site greater than that created by the PMF. The applicant also has estimated the time required for a flood wave above plant grade from a postulated

seismic dam failure(s) to reach plant grade. The minimum time is 28 hours for

the OBE failure of Fontana Dam during a flood equivalent to one-half the PMF.

This minimum time of 28 hours is greater than the stage I and II flood preparation time of 24 hours. The Fontana Dam failure also yields the highest flood

level (732.1 ft msl) at the site for seismic-induced dam failure. This level

is, however, well below the design flood level produced by the PMF.

Although Watts Bar Dam is only 1.9 miles upstream of the-site, its unexpected

failure--such as from a severe earthquake--would not result in water above

plant grade even if the failure were coincident with a 25-year flood. Combinations of more severe floods and failure of Watts Bar Dam would result in longer

warning times because of the time necessary for flood buildup.

The design flood for the site is a PMF, which would be likely to result in a

failure of the Watts Bar Dam. The maximum water levels noted above would

inundate much of the plant site. Special precautions would be necessary to

protect vital areas and equipment. Excluded would bethe turbine, control,

and auxiliary buildings, which would be allowed to flood. To ensure that the

necessary emergency shutdown procedures are in place before severe flood conditions could affect safety-related facilities, the applicant will use a flood

warning system and Technical Specification requirements that vary, depending

on the time of year. For the October 1 through April 15 winter season, shutdown procedures are to begin when the river is forecasted to reach elevation.

714.5 ft msl. The April 16 to September 30 summer level for initiation of

shutdown procedures is 726.5 ft msl.

Both the design-basis flood conditions and warning procedures have been developed by TVA using models of rainfall, runoff, dam failures, and flood wave

propagation. Where possible the models have been verified and calibrated

against historical data. Where no good data existed (such as for models of

dam failure), conservative assumptions were used in theoretical models. The

staff has reviewed the models, the assumptions used therein, and the verification, calibration, and application of the models to predict design-basis flood

Watts Bar SER 2-.18
conditions. In all cases the staff concluded that TVA estimates of design-basis

flood conditions were reasonable.

The staff concurs with the applicant's design-basis flood levels for Watts Bar

and concludes that they can be used as the basis for flood protection requirements to meet GDC 2 with respect to Tennessee River floods.

2.4.4 Local Intense Precipitation in Plant Area

The staff has reviewed the applicant's estimate of the probable maximum precipitation (PMP) for use in assessing the potential of locally induced flooding in

the plant area. The staff concludes that the PMP of 14.0 in. in 1 hour has been

estimated in-accordance with the suggested criteria of Regulatory Guide 1.59

("Design Basis Floods for Nuclear Power Plants") and is acceptable. The staff

has reviewed the applicant's provisions for conveying the runoff, resulting

from the local PMP, away from safety-related buildings and concludes that these

provisions for local site drainage meet the criteria recommended in Regulatory

Guide 1.59. The staff concludes that the Watts Bar design, in this regard,

meets the requirements of GDC 2 and, therefore, is acceptable.

2.4.5 Roof Drainage

The applicant has provided data and analyses to show that the roofs of safetyrelated buildings can support a combined snow and/or rainfall depth considerably in excess of the tops of parapet walls. Thus, PMP is not the controlling

design-basis event for the roofs of safety-related buildings. The staff has

reviewed the applicant's provisions for roof drainage and finds them to be

acceptable and to meet the requirements of GDC 2.

2.4.6 Ultimate/Heat Sink

The Tennessee River (Chickamauga Reservoir) is the ultimate heat sink (UHS)

for the Watts Bar plant. An 800-ft long riprapped channel connects the plant

intake structure with Chickamauga Reservoir/Tennessee River. A maximum intake

flow rate of 172 cfs is necessary for normal plant operation. Safe shutdown

under normal or emergency conditions requires about 75 cfs. Watts Bar Dam,

1.9 mi upstream, controls the flow and, with Chickamauga Dam downstream, helps

to control the reservoir/river level. The applicant also is the operator of

Watts Bar Dam and has committed to maintain an adequate flow and/or water

level in the river to shut down the plant safely, and to maintain shutdown.

In the remote event that both Watts Bar and Chickamauga Dams fail, the estimated "natural" (without dams) historic low flow of 2700 cfs would provide an

adequate flow for safety shutdown. A narrow section of the reservoir downstream

of the site is of sufficient depth to provide assurance that an adequate water

level also would exist to submerge the vital pumps properly.

The vital pumps are the essential raw cooling water (ERCW) and high pressure

fire protection (HPFP) pumps, which are located on the top deck (elevation

741.0 ft msl), within the walled enclosure of the intake pumping station. The

pumps are above the level of the design-basis flood'on the Tennessee River and,

thus, meet the requirements of GDC 2 with respect to flood hazards.

Watts Bar SER 2-;19
The intake channel is, however, susceptible to sediment deposition. Because

the plant is located just downstream of the Watts Bar Dam, the sediment deposition potential should not be excessive because the major portion of the river's

sediment load is deposited upstream of the dam. The applicant has provided

background sedimentation information and conservatively estimated an annual

sediment accumulation of less than 2.5 in. in the intake channel. The worstcase postulation of possible sediment blockage of the intake channel would be

the failure of Chickamauga Dam downstream and a minimum release of 2000 cfs

from the upstream Watts Bar Dam. The resultant minimum water level at the

intake channel would be 666.0 ft msl. The invert elevation of the intake

channel is about 660.0 ft msl. Thus, there would still be about a 6.0 ft

depth of water at the intake structure for safe shutdown of the plant.

The staff has reviewed and concurred with the applicant's sedimentation estimates. The staff also has made conservative estimates of sediment accumulation

during extreme flood events (less than I ft) and concludes that sedimentation

in the intake canal will not be significant and can be controlled under a formal

inspection and maintenance program as described in Regulatory Guide 1.127

("Inspection of Water-Control Structures Associated with Nuclear Power Plants").

The staff has reviewed the applicant's provisions for the safety-related water

supply and concludes that all provisions are acceptable and meet the suggested

criteria of Regulatory Guide 1.127. Therefore, the staff concludes that the

hydrologic aspects of the UHS meet the requirements of GDC 44.

2.4.7 Groundwater

The materials immediately beneath the'site are older terrace deposits and

alluvium that are poor waterbearing strata., At deeper levels, there are

discontinuous beds of Conasauga shale (more properly a combination of about 84

percent shale and 16 percent limestone), Chickamauga limestone, and Knox

dolomite. The Knox dolomite constitutes the primary regional aquifer where

water is found in solution'channels and openings formed along bedding planes

and joints. The general region is known as a Karst area where flowing water is

sometimes found in significant solution channels. At the site, groundwater in

contact with plant structures generally moves toward the river.

TVA has developed three wells 2.5 mi northwest of the plant drawing from the

Knox dolomite. The requirements placed on these wells include 16,000 gpd for

potable uses at the plant and 200,000 gpd for offsite uses.

2.4.8 Design Basis for Subsurface Hydrostatic Loading

The applicant has constructed an underdrain system around the structures in the

power block to reduce water pressures on the buildings and potential inleakage.

A single pipe conduit, fed and discharged by gravity, and porous backfill

constitute the primary features of the system. TVA has claimed the system

is not safety related under the provision of Branch Technical Position (BTP)

HMB/GSB1, SRP Section 2.4.13. TVA states that credit for the system's ability

to lower groundwater levels has not been taken in the design of safety-related

structures.

Watts Bar SER 2-20
The applicant stated in the FSAR that (1).buildings in the main plant area are

designed for a groundwater elevation of 710.0 ft msl, and (2) the elevation of

710.0 is maintained by a groundwater drainage system around the main plant area.

The applicant also stated that for the design of Category I structures not in

the main plant area the following criteria were used:

(1) For the miscellaneous ERCW structures, the elevation used for the normal

groundwater was 726.0 ft msl.

(2) For the intake pumping station and retaining walls, the elevation used

for normal groundwater was 685.0 ft msl. This elevation is maintained by

providing weep holes in the retaining walls.

(3) For the refueling water tanks, pipe tunnels and Class 1E electrical

systems structures, TVA used a groundwater elevation of 728.0 ft msl.

(4) For the diesel generator building, TVA used a groundwater elevation of

741.0 ft msl.

In response to staff questions on the underdrain system (not seismic Category I),

the applicant states: "Additional analysis has indicated, however, that the

stability and structural design of the WBN [Watts Bar Nuclear] main plant

structures are adequate for a groundwater level of elevation 726 (2 feet below

grade) combined with an earthquake loading condition (either [operating basis

earthquake] OBE or SSE [safe shutdown earthquake])."

The staff finds the design-basis groundwater levels for items (1) through (4)

above to be acceptable. The staff further concludes, on the basis of the above

applicant statement relevant to groundwater level in the main plant area, that

the design-basis groundwater elevation for buildings in the main plant area is

elevation 726.0 ft msl. The associated review of structural design capabilities

to withstand groundwater loads is contained in Section 3.8 of the SER.

The applicant has not identified a design-basis groundwater level for the ERCW

pipeline. Consequently, the staff has concluded that ground elevation must be

used as the design-basis groundwater level for the ERCW pipeline. The staff

will confirm this conclusion with TVA and will report its findings in a supplement to the SER.

The staff finds-that the above provisions for design-basis groundwater level

meet the requirements of GDC 2 and are acceptable. The staff will confirm this

conclusion with TVA and will report its findings in a supplement to the SER.

2.4.9 Transport of Liquid Releases

Normal releases of low level radioactivity in water from the plant will be made

through the plant discharge system. In addition, some accidental releases could

occur via the same route. TVA has evaluated the dilution of such releases in

the Tennessee River using conservative assumptions. At the closest major downstream intake, at Dayton, plant releases to the river would be likely to be

diluted by a factor of 3.6 x 108 or more.

TVA also evaluated the travel time and dilution that would occur if an acci-.

dental release were to be-made to groundwater. TVA concluded that contaminated

Watts Bar SER 2-21
water would take 812 days to reach the nearest point of known possible use, which

is a tributary of Yellow Creek. The point of groundwater discharge is 2600 ft

from the plant. TVA also estimated a dilution factor of about 9.8 x 106.

The staff has independently reviewed the potential for accidental release of

radioactive liquid to the groundwater at the Watts Bar plant and has concluded

that the applicant-identified pathway to Yellow Creek may not be the most

critical in terms of the potential for offsite contamination. The shortest

groundwater pathway to a potential user is to the southeast about 1600 ft to

the Tennessee River/Chickamauga Reservoir. However, there is a permanent,

passive, dewatering system at about elevation 710.0 ft msl that would capture

any radioactive liquid that might seep through the walls above the 710.0 ft msl

level. The dewatering system discharges, by gravity, to a holding pond and

then to the cooling tower blowdown discharge facilities. A radioactivity monitor on cooling tower blowdown is provided downstream of the holding pond and

will serve to detect any accidental releases that get into the dewatering system.

Details of radioactivity monitoring and Technical Specifications are provided

in Section 12 of this report.

Any radioactive liquid from an accidental tank failure flowing to lower building levels (below elevation 710.0 ft msl) would not escape from the buildings.

With groundwater at elevation 710.0 ft msl, there would be higher pressure on

the exterior building walls than on the interior.- Cracks in the building walls

or floor would result in groundwater entering the buildings rather than the

contaminated water seeping out of the buildings. However, if it were possible

for radioactive liquids to migrate from buildings, the direction would be toward

the Chickamauga Reservoir. The staff analyses show that the liquid transit

time would be greater than 267 days and the-dilution factor would be about

3.4 x 107. As discussed in Section 15.7.3 of this report, concentrations at

the nearest surface water in an unrestricted area from accidental failure of a

11,000-gal boric-acid-bottoms tank, assuming all of the contaminated water

reached the groundwater, would be less than the limits specified in 10 CFR 20,

Appendix B, Table II, Column 2.

2.4.10 Flooding Protection Requirements and Technical Specifications

Potential large magnitude floods resulting from either precipitation and

accompanying erosional-type (overtopping) dam failures or seismic dam failures

with and without coincident precipitation, can inundate the Watts Bar site and

flood some safety-related buildings. The applicant has prepared a very comprehensive flood emergency plan that allows for early flood forecasts, plant preparation time (at least 27 hours), and elaborate plant preparation procedures

to ready the nuclear plant for flood waters above plant grade and ensure continued safe shutdown of the plant during and after flood conditions.

A detailed discussion of the emergency plan is provided in Section 2.4.14 of

the FSAR. In summary, the applicant has documented his ability to provide,

through flood forecasting procedures, a minimum of 27 hours of preparation

time, divided into two stages (Stage 1-10 hours and Stage 11-17 hours) for any

Tennessee River flood that could inundate the Watts Bar plant site. Specific

details of the plan--such as target river levels, initiation of Stage I and

Stage II shutdown, communications--are incorporated in a Technical Specification

and limiting condition for operation.

Watts Bar SER 2-2.2
Based on its review of the information provided by the applicant and an onsite

audit, the staff finds both the flood emergency plan and the Draft Technical

Specification to be acceptable from a hydrologic engineering standpoint and

concludes that they meet the requirements of 10 CFR 50.36 and GDC 2.

2.5 Geology and Seismology

The, staff has reviewed all available relevant geologic and seismologic information obtained since the issuance of the CP-SER in 1972.

In the CP-SER, the staff concluded:

(1) Geologic and seismologic investigations and information provided by the

applicant and required by Appendix A to 10 CFR 100 provided an adequate

basis for determining that no capable faults exist at the plant site or

within 5 mi.

(2) Earthquakes that have occurred in the region cannot be related directly

to any faults in the area.

(3) Ground motion values of 0.18 g and 0.09 g are adequately conservative values

for the safe shutdown earthquake (SSE) and the operating basis earthquake

(OBE), respectively.

After careful review of the new information as provided and evaluated by the

applicant, the staff concludes that from a geologic standpoint there is no

basis for altering the CP-SER conclusions concerning the safety of the Watts

Bar nuclear plant.

However, since issuance of the CP-SER the staff has adopted an SRP and a

Regulatory Guide which affect the evaluation of ground motion. Further discussion of these is in Section 2.5.2.

After evaluating the FSAR and subsequent documents and information, the staff

has concluded that the applicant has (1) performed site and regional geologic

and geophysical investigations, (2) reviewed all available pertinent literature,

and (3) provided the staff with all information necessary to evaluate, assess,

and support the applicant's conclusions concerning the safety of the Watts Bar

site from the geologic and seismologic point of view. In addition, the staff

finds the applicant has satisfied the requirements of and is in compliance with

applicable portions of Appendix A to 10 CFR 100; SRP Sections 2.5.1, 2.5.2, and

2.5..3 (NUREG-0800); and Regulatory Guide 1.70, "Standard Format and Content of

Safety Analysis Reports for Nuclear Power Plants," Revision 2.

2.5.1 Geology

2.5.1.1 Summary of Regional and Site Features

The site is in the Southern Valley and Ridge Tectonic Province, which is characterized by folded and thrust-faulted Paleozoic age (600-250 million years before the present (mybp)) sedimentary rocks of varying erosional resistance,

which results in a series of northeast-trending ridges and intervening valleys. The site is underlain by shale and interbedded limestone of the Conasauga

Formation, which is Middle Cambrian in age (550 mybp).

Watts Bar SER 2-23
Overlying the bedrock is 30 to 46 ft of river terrace deposits of silt, clay,

micaceous sand, and quartz gravel, determined from isotopic dating to be a

minimum of 35,000 years old.

Structurally, the site is on the Kingston thrust sheet, one of a series of lowangle or bedding plane thrusts trending northeast-to-southwest that characterize

the Southern Valley and Ridge Tectonic Province. All of the folding and faulting, including the minor structural features in the Conasauga Formation at the

site (that is, minor folds and faults, bedding slips, and joint systems), are

interpreted as having resulted from the effects of the large-scale fault movements during the Appalachian Orogeny at the end.of the Paleozoic Era (250 mybp).

2.5.1.2 Tectonic Setting

Because earthquake activity cannot be reasonably associated with known geologic

structure in the Southern Appalachians, earthquakes are instead identified with

the tectonic province in which the site is located. In a report, Southern

Appalachian Tectonic Study (Seay, 1979), the applicant attempted to establish

new tectonic provinces based on geophysical and seismic information. The

subdivisions proposed divided the Southern Appalachians lengthwise into northern and southern parts along the New York-Alabama magnetic lineament and then

transverse to the Appalachian trend along northwest-southeast lines. An important result of these proposed province boundaries was to isolate the TVA nuclear

power plants from the Giles Gounty seismicity. However, staff consultants,

the U.S. Geologic Survey (USGS), in a status review (Nelson and McDowell, 1980)

rejected the TVA study conclusions because they did not provide an adequate

basis for the tectonic subdivisions for regulatory purposes. The NRC staff

adopted the USGS position. The applicant, therefore, indicates that the site

is located in the Southern Appalachian Tectonic Province. As defined, this

province is bounded on the east by the western margin of the Piedmont Province,

on the west by the western limits of the Cumberland Plateau, on the south by the

overlap of the Gulf Coastal Plain Province, and on the north by the re-entrant

in the Valley and Ridge Province near. Roanoke, Virginia.

In its review, the staff determined that the proposed site is within the Southern

Valley and Ridge Tectonic Province based on provinces which are more in accord

with those proposed by King (1969), Eardley (1962), Rodgers (1970) and Hadley and Devine (1974) for eastern North America. This province is bounded on the

east by the western extent of the Piedmont Province,* on the west by the Cumberland Plateau, on the south by the Gulf Coastal Plain, and on the north by the

northern part of-the Valley and Ridge Province.

2.5.2 Seismology

2.5.2.1 Introduction

Because the Watts Bar site lies within the Southern Valley and Ridge Tectonic

Province (as discussed in Section 2.5.1), in this and most eastern site evaluations, the controlling earthquake, or that event which defines the SSE, becomes

the maximum historical earthquake in the province within which the site is

located, following the procedures set forth in Appendix A to 10 CFR 100.

*In the staff's view, for purposes of nuclear power plant siting, the Blue

Ridge Province is considered as part of the Piedmont Province.

Watts Bar SER 2-24
During the CP review, the staff concluded that a modified Newmark-type (preRegulatory Guide 1.60," Design Response Spectra for Seismic Design of Nuclear

Power Plants") response spectrum anchored at 0.18 g was acceptable as the SSE

for the Watts Bar nuclear plant. This conclusion was based on the assumption

that the maximum historical earthquake (1897 Modified Mercalli Intensity VIII

earthquake at Giles County, Virginia) might recur anywhere within the Southern

Valley and Ridge Tectonic Province, and on adoption of recommended response

spectra and an acceleration anchor point.

In June 1981 the staff received a preprint of a research paper, "The Giles

County, Virginia Seismic Zone-Configuration and. Hazard Assessment," by Dr. G. A.

Bollinger. This paper was presented at a conference on earthquakes and earthquake engineering in the eastern United States in September 1981. In this

paper, Bollinger postulated the existence of a buried fault in Giles County.

He delineated this fault plane with the hypocenters of eight recent microearthquakes and four larger pre-1977 felt events. He used the largest extent of the

seismic zone defined by these hypocenters, including the maximum extent of the

hypocenter location errors in order to calculate the maximum possible fault

dimensions. Using these maximum dimensions, he then postulated a maximum earthquake of surface wave magnitude Ms=7 for this fault. Using Nuttli's (1981)

relation between Ms and mb for midplate earthquake results-in an mb of 6.3 for

this postulated event.

In view of the fact that the Giles County earthquake is the controlling earthquake for the seismic design of the Watts Bar nuclear plant, the staff asked

the applicant to provide a discussion on any effect Bollinger's hypothesis

might have on the site with respect to the following three items:

(1) the potential of the 1897 earthquake being associated with this specific

geologic structure

(2) the potential of an earthquake up to Ms=7.0 being located in Giles County,

and any far field ground motion effect (both peak values and response spectrum) at the site from an Ms=7.0 event in Giles County

(3) the potential of similar seismogenic structures being located near the

Watts Bar site and any effect at the site from earthquakes on these seismogenic structures

In FSAR Amendment 45, the applicant provided a response to the above items in

question. In addressing item (1) above, the applicant has provided arguments

which both lend support to the association of the 1897 earthquake with Bollinger's

inferred fault and arguments which tend to contradict the association. There

are no compelling arguments in this draft report that prove or disprove the

*association of the earthquake with the fault. Although the applicant pointed

out the potential benefits to its licensing position of associating the 1897

earthquake with a particular structure in Giles County, the applicant also

questioned the very existence of that structure by pointing out that four

(33 percent) of the 12 microearthquakes detected by Bollinger's seismic network and two (33 percent) of the six pre-1977 felt events do not fall within

the structure defined by Bollinger.

Watts Bar SER 2-25
In addressing item (2) above, the applicant argued that Bollinger stressed

throughout his report that he was postulating a "worst case" situation and

"not considering seismic risk for engineering purposes." The applicant also

pointed out that Bollinger used a significant number of qualifiers in his

arguments to stress the tentativeness of the assumptions and postulations

employed and that these are inferred from the data and not proven by them.

The applicant also argued that cascading several worst-case hypotheses yield

a result that is not directly applicable or appropriate for seismic design

criteria for specific sites in the region.

The applicant also estimated the ground motion effect at the Watts Bar-site of

a surface wave magnitude Ms=7.0 located in Giles County, Virginia at a distance

of about 250 mi, using Nuttli (1979) to calculate peak accelerations and velocities and Murphy and O'Brien (1978) to calculate peak accelerations. Both of

these sets of estimates were considerably less than the Watts Bar SSE.

The applicant also compared the peak acceleration and response spectra from 20

recordings of 10 earthquakes that ranged in magnitude from 6.3 to 7.75 and in

epicentral distance of from 64 to 230 mi. The applicant found that the ground

motions from these recordings are less than the Watts Bar SSE ground motions.

In addressing item (3) above, the applicant investigated the historical

seismicity of the region and states that he has found no significant evidence

from either microseismic research or from the alignments of historic epicenters

that suggests the existence of seismogenic structures in the Southern Appalachians

similar to that postulated by Bollinger in Giles County. The applicant does

recognize that there are localized areas of recurrent historical seismicity

but states there is not sufficient data to define a single source structure

using methods such as Bollinger did.

The staff has reviewed the Bollinger report and the applicant's responses to

staff questions and concludes that the site-specific spectrum adopted for the

Watts Bar plant by staff is adequately conservative to encompass any vibratory

ground motion which might result from the type of earthquake postulated by

Bollinger as a result of his research on the Giles County seismogenic structure.

This conclusion is based primarily on the following:

Bollinger characterizes his research as not being directly applicable or appropriate for seismic design criteria for specific sites in the region.

If all of Bollinger's assumptions are correct and there is a structure in Giles

County capable of a surface wave magnitude Ms=7 earthquake, then the event would

be related to that particular structure and could not be assumed to occur anywhere else in the tectonic province.

The staff made its own estimate of ground motion from an Ms=7 earthquake at a

distance of 250 mi from the site. This results in a vibratory ground motion

well below the Watts Bar SSE. Using Nuttli's (1979), Tera-Lawrence Livermore

Laboratory's (Jackson, 1980), and Hermann's (1981) attenuation relations for an

mb= 6 . 3 (Ms =7) earthquake at a distance of 250 mi, peak acceleration and velocity

values are obtained that, when used in conjunction with the amplification factors

in NUREG/CR-0098, result in a spectrum greatly below that of the SSE.

Watts Bar SER 2-26
There is no evidence at the present time to indicate that there are structures

similar to the type postulated by Bollinger nearer to the site.

Although Bollinger has postulated a seismogenic structure in Giles County, the

evidence presented is tentative and subject to interpretation. The staff posi- tion is that it would be premature to assume the existence of this structure

in the Watts Bar review.

Therefore, the staff still considers a possible recurrence of the 1897 maximum

Modified Mercalli Intensity VIII Giles County earthquake in the site vicinity

to be the controlling earthquake for determining the SSE.

Although the seismological and geological evaluation of this controlling earthquake has not been altered since the CP review, the staff has in the interim

adopted an SRP section and Regulatory Guides that have the effect of changing the response spectra and the anchor point acceleration for a Modified Mercalli

Intensity VIII earthquake from the values accepted for Watts Bar at the CP

review stage. Specifically, the staff would now characterize a Modified

Mercalli Intensity VIII earthquake with the more conservative spectrum speci- fied in Regulatory Guide 1.60, anchored at 0.25 g. The higher reference

acceleration is determined using the trend of the means relating peak acceleration to intensity shown by Trifunac and Brady (1975).

The SRP and Regulatory Guides represent approaches and practices the staff

considers acceptable to establish conformance with NRC regulations, but they

are notspecified as the only acceptable means of meeting the regulations.

Because the present staff practice, as represented by the SRP and Regulatory Guides, resulted in the identification of a seismic design basis acceptable to

the staff which exceeded the actual seismic design basis for the Sequoyah, Watts

Bar, and Bellefonte nuclear plants, the staff asked the applicant to provide information that would establish the adequacy of the seismic design of this

plant and other TVA plants in the region (Sequoyah and Bellefonte).

The applicant responded with several reports. These reports were extensively evaluated for the Sequoyah OL review. The following discussion represents an

updated evaluation, parts of which have been excerpted from the Sequoyah SER.

The first report, "Justification of the Seismic Design Criteria Used for the

Sequoyah, Watts Bar, and Bellefonte Nuclear Power Plants, Phase I," was submitted in April 1978. In it, the applicant argues that the present designs at

the three sites are adequate because

(1) The maximum intensity of the 1897 Giles County earthquake was really

VII-VIII rather than VIII.

(2) The intensity rating for the 1897 Giles County earthquake is soil biased

and that the same earthquake would result i'n a lower intensity at rock

sites.

(3) The intensity-acceleration relationship derived by Murphy and O'Brien (1978) is more appropriate than that found in Trifunac and Brady (1975) and should

be used in determining reference accelerations.

Watts Bar SER 2-27
(4) At foundation depth, earthquake-induced ground motion is less than that

at the surface.

The staff concluded that, although there was some validity in these arguments,

there were enough problems associated with each to preclude their use in

justifying the adequacy of the present seismic design. For example,

(1) The USGS and National Oceanic and Atmospheric Administration (NOAA) had

convened special panels to reevaluate the Giles County earthquake. The

majority of the panel members considered the maximum intensity to be VIII

even though some felt it to be a "weak" VIII (according to a letter from

W. A. Radlinski, USGS, to E. G. Case, NRC, dated February 12, 1976).

(2) Studies of earthquake response spectra suggest that, for similar earthquakes, spectra recorded on rock sites indicate stronger motion at frequencies of interest to nuclear power plants (greater than two Hertz) than

those recorded on soil sites (Trifunac, 1978). Other studies (Boore and

others, 1978, and Joyner and others, 1981) report no differences in peak

acceleration between soil and rock sites although they do report higher

values of peak velocity at soil sites than at rock sites.

(3) Although the Murphy and O'Brien (1978) intensity-acceleration relationship

may be more correct than the Trifunac and Brady (1975) relationship with

regard to peaks of acceleration, this does not necessarily mean that the

former is more appropriate for use in predicting reference accelerations.

Agbabian Associates (1977), for exampl.e, compared different design spectra and reference accelerations for given intensities with response spectra

determined from accelerograms recorded at the'same intensities. When used

in conjunction with the Regulatory Guide 1.60 spectrum, the more conservative Trifunac and Brady relationship was not found to predict an overly

conservative design (that is, exceed the mean plus one standard deviation

of the data in frequencies of interest).

(4) It is very difficult to predict the reduction in ground motion at shallow

depths. Staff and Advisory Committee on Reactor Safeguards positions on

the Davis-Besse Unit 1 Nuclear Power Plant (Docket No. 50-346) advise

against allowing for this reduction.

Thus, the staff then requested the applicant to

(1) Determine site-specific SSE response spectra from strong-motion records

of appropriate magnitude and distance.

(2) Determine site-specific SSE response spectra from strong-motion records

of appropriate intensity.

(3) Reevaluate original seismic structural and floor response analyses using

more realistic methods and material properties as well as site-specific

SSE response spectra.

(4) Reevaluate the OBE to determine if it meets the reasonability-of-occurrence

criterion of Appendix A to 10 CFR 100.

Watts Bar SER 2-28
(5) Compare the probability of the SSE being exceeded to other TVA plants that

conform to the SRP.

The applicant submitted two additional reports, "Justification of the Seismic

Design Criteria Used for the Sequoyah, Watts Bar and Bellefonte Nuclear Power

Plants, Phase II," and "Justification of the Seismic Design Criteria Used for

the Sequoyah, Watts Bar and Bellefonte Nuclear Power Plants, Phase II, Responses to NRC Questions 1 through 6."- The following review utilizes results from

these reports.

2.5.2.2 Site-Specific Earthquake

In order to compute site-specific response spectra, it is necessary to characterize the earthquake size, the epicentral distance (the distance between the

surface location of the earthquake and the site), and the site conditions (soil

or rock) being modelled. As mentioned above, the 1897 Giles County earthquake

was considered to have an epicentral intensity of VIII (Modified Mercalli).

There are relatively few recordings of strong ground motion at intensity VIII

and none (at least in the western United States) recorded at rock sites. This

fact, plus the dispute over the epicentral intensity (that is, VIII, weak VIII,

or VII-VIII) and the more dependable classification of strong-motion records

by magnitude, led the staff to seek a reliable magnitude estimate of the Giles

County event for this review. A study by Nuttli, Bollinger and Griffiths (1979)

has utilized empirically determined relationships between body wave magnitudes

of instrumentally recorded earthquakes and (1) their intensity fall-off with

distance, (2) their total felt area, and (3) the area within the intensity IV

isoseismals to determine magnitudes for historical earthquakes where only intensity data were available. Particular attention was given in this study to the

1897 Giles Gounty earthquake. The two most reliable estimates in the authors'

opinion (intensity fall-off with distance and the intensity IV isoseismal area)

each indicated a body wave magnitude of 5.8 for this event. The staff recommended use of this value and, to account for uncertainty, suggested the applicant assume a body wave magnitude of 5.8 ± 0.5 (5.3 to 6.3) for the Giles County

earthquake. In addition, the suggestion was made to include all strong-motion

records in this magnitude range recorded at rock sites at epicentral distances

of less than 12.5 to 15.5 mi. According to Gupta and Nuttli (1976), this is the

approximate distance range to which maximum intensities are felt in the central

United States. In addition, at these close distances, the strong differences

in seismic wave attenuation between earthquakes east and west of the Rocky

Mountains have not yet affected the ground motion (Nuttli, 1981). In the opinion

of the staff, this makes the direct use of strong-motion records from the west

for simulating ground motion in the east (where there are no appropriate records)

more credible and acceptable..

Aside from reflecting the uncertainties in characterization of the controlling

earthquake, establishing a range of magnitude and distance helps ensure an adequate amount of data. The applicant was able to collect 13 sets of strongmotion records (26 horizontal components) that fall within this range. Six

sets are from western US earthquakes (Helena, 1935; San Francisco, 1957; Parkfield, 1966; Lytle Creek, 1970; and Oroville, 1975), and seven sets are from a

sequence of well-recorded events that occurred in 1976 near Friuli, Italy. The

magnitudes of the earthquakes used range from 5.3 to 6.2 (average 5.7) and the

epicentral distances range from 4.4 to 16.8 mi (average 10 mi). The set of

records recorded at 16.8 mi represent the strongest ground motion in the data

Watts Bar SER 2-29
and were included to insure conservatism even though it fell slightly outside

the 15.5 mi guideline. The magnitude (ML) used for the western U.S. earthquakes

is the magnitude scale developed for southern California earthquakes by Richter

(1935). Chung and Bernreuter (1980) have investigated the link between the mb

of an eastern U.S. earthquake and the ML of a western U.S. earthquake releasing

the same amount of energy. Using their relationship results in differences of

less than 0.2 between western ML and eastern mb for earthquakes in the ML range

5.0 to 6.0. The staff knows of no relationship between Italian ML and eastern

U.S. mb but assumes that the differences are relatively small and comparable

to those between the eastern and western U.S. Response spectra from the 26

records were computed, and representative site-specific spectra were calculated.

No scaling for magnitude or distance was done because the parameter range chosen

is believed to represent the uncertainty in the definition of the controlling

earthquake causing maximum damage at the site. The data at different frequencies were found to best fit a log-normal distribution. The 50th and 84th percentile spectra were calculated.

The staff has concluded that the 84th percentile spectra represents an appropriately conservative representation of the site-specific earthquake. Although

it is not dictated by the regulations, the choice of this level is a judgement

which the staff considers a reasonable technical approach to establish a standard of acceptability which is conservative. The selection of the 84th percentile level (mean plus one standard deviation) is the level used in deriving

the Regulatory Guide 1.60 spectral shape; it is also the level of acceptability

being expressed by the staff in revisions of the SRP dealing with the use o-f

site-specific spectra. Aside from its use in Sequoyah, Watts Bar, and Bellefonte reviews, this has been the level approved in the other staff reviews

utilizing the site-specific spectra methodology (see, for example, Fermi

Unit 2 SER and San Onofre Units 2 and 3 SER).

Figure 2.6 is a plot of the Watts Bar design spectrum (5 percent damping) and

the 50th and 84th percentile site specific response spectra (4 percent and 7

percent damping). The damping values shown are those available for comparison. The Watts Bar design spectrum is exceeded by a small amount in the period

range 0.09 to 0.26 seconds by the 7 percent damping site-specific 84th percentile response spectrum, and in the period range 0.09 to 0.30 seconds by the

4 percent damping site-specific 84th percentile response spectrum. These comparisons apply to structures on rock foundations. For soil-supported structures, the applicant utilized techniques which assume conservative amplifications in the period of interest (less than 0.5 seconds).

2.5.2.3 Operating Basis Earthquake

As currently defined, the operating basis earthquake (OBE) used in the design

of the Watts Bar nuclear plant is 1/2 the SSE accepted by the staff in the

CP-SER. Based on his seismic hazard calculations, the applicant calculated a

return period for the OBE peak horizontal acceleration of 550 years. The staff

compared return periods calculated by the applicant for the OBE in the frequency

range of interest and arrived at values of from 200 to 700 years. Thus, the

OBE is acceptable in light of the Appendix A to 10 CFR 100 definition of the

Watts Bar SER 2-30
100 11I

7c~l

/0

WATTS BAR 5%

10

4%

7%/0

• o • •84th %/

.S

0

(h(

--

---

0.1

0.01 0.1 1 10

PERIOD (secs)

Figure 2.6 Watts Bar design response spectrum (5% damping) and 50 and

84 percentile site-specific spectra (4% and 7% damping)

Watts Bar SER 2-31
OBE as being "that earthquake which... could reasonably be expected to affect

the plant site during the operating life of the plant."

2.5.3 Surface Faulting

During construction, investigation by drilling and geologic mapping of excavations uncovered several faults and shear zones through the plantfoundation.

Detailed field mapping, tectonic analysis, and isotopic dating (as reported in

Section 2.5.3.2 of the FSAR) support the conclusion that the faults and shear

zones have minor displacement, are confined to the Conasauga Formation below

the site, and, because they do not offset terrace gravels where contacts are

exposed, are older than 35,000 years.

Therefore, the applicant has concluded, and the staff concurs, that no evidence

of surface displacement or capable faults within the meaning of Appendix A to

10 CFR 100 is present at the site.

2.5.4 Stability of Subsurface Materials and Foundations

2.5.4.1 General

The Watts Bar nuclear plant is located on the west bank of Chickamauga Lake'on

the Tennessee River approximately 2 mi downstream from Watts Bar Dam. The plant

is 31 mi upstream from the Sequoyah nuclear plant, which is also on Chickamauga

Lake. The Watts Bar plant site is situated in a bend of the Tennessee River

that has been covered by alluvial terrace deposits. The finished plant grade

around the reactor buildings and other safety-related buildings in the main

powerblock area is at elevation 728 ft msl.

The safety-related structures, systems, and components (SSC) which have been

reviewed for foundation stability are listed in Table 3.2-1 of the FSAR. The

seismic Category I structures at the site in the main powerblock area include

* the reactor building

auxiliary control building (control bay portion, auxiliary building

portion, waste packaging area, and condensate demineralizer waste

evaporator building portion)

* additional equipment buildings 1 and 2

* north steam valve rooms

* refueling water storage tanks foundations

* essential raw cooling water (ERCW) discharge overflow structure

ERCW standpipe structures I and II

pipe tunnels A and B from primary water storage tanks to the auxiliary

building

diesel generatorbuilding

Watts Bar SER 2-32
The seismic Category I structures related to the ultimate heat sink (UHS)

system at the Watts Bar site consist of

* intake channel

intake pumping station, along with the concrete and sheetpile retaining

walls

HPFP piping structures

ERCW piping structures

Class 1E electrical systems conduit banks, manholes and handholes

The source of water for the intake channel is the Chickamauga Lake on the

Tennessee River. At the Watts Bar site, the reservoir is about 1100 ft wide,

with depths ranging between 18 and 26 ft at the normal pool elevation of

682.5 ft. The intake channel is a man-made feature, extending approximately

800 ft from the edge of the reservoir to the intake pumping station. The

bottom of the channel is at elevation 660 ft and is 50 ft wide. The intake

pumping station is approximately 2200 ft from the main powerblock area.

2.5.4.2 Subsurface Conditions and Geotechnical Properties

The average thickness of the surficial soil deposits overlying the bedrock at

the site ranges from approximately 30 to 40 ft. The soil consists of a 10 to

20 ft thickness of alluvial deposits underlain by a 10 to 20 ft thickness of

terrace materials. In the powerblock area, the alluvium is composed of finegrained silts and clays, with micaceous sand and some quartz gravel; the underlying terrace deposits at this location are composed of pebbles, cobbles, and

small boulders of quartz embedded in a sandy clay matrix. Underlying these

soils is the bedrock composed of about 1000 ft of Conasauga Formation. This

formation consists of contorted gray-green shales interbedded with lenses of

limestones. The Conasauga shale is underlain by at least 2000 ft of the Lower

Cambrian Rome Formation, which consists of shale, siltstone, and sandstone.

The subsurface conditions along the seismic Category I channel, pipeline, and

conduit alignments are presented in Sections 2.5.4.4 of this SER.

The quality and properties of in situ soil and rock foundation materials beneath

seismic Category I structures, conduits, and pipelines were evaluated by the

applicant on the basis of field and laboratory investigations. The field

investigations included exploration drilling. In the powerblock area, the borings

were taken to elevation 635 ft, which is 55 ft below foundation grade and from

60 to 65 ft below the general elevation of the top of rock. Along seismic Category I channel, pipeline, and conduit alignments, the borings generally extended

about 10 ft below the top of shale. The soil exploration included soil sampling

using a split-spoon sampler and a fixed piston sampler. The Standard Penetration Test (blow count) information was obtained using the split-spoon sampler

in accordance with American Society for Testing and Materials (ASTM) method D-1586.

The rock was cored using NX wireline core barrels. Rock quality designation

(RQD) as well as percentage recovery were recorded. The staff finds the field

investigation program adequate and acceptable.

Watts Bar SER 2-33
A total of 72 Menard pressuremeter tests were also conducted on site. These

tests were conducted in six holes and were designed to determine the in situ

static modulus of elasticity of the shales at the site.

Geophysical studies performed at the site included seismic refraction surveys,

downhole surveys, a cross-hole shear wave survey, and continuous Birdwell 3D

sonic logging surveys.

Additional soil and rock characteristics required for the analyses of the static

and dynamic-stability of subsurface materials and foundations under seismic

Category I structures, channels, pipelines, and conduits were established using

laboratory tests which included static strength tests, moisture content and

Atterberg limits determinations, grain-size tests, and classification tests.

In the analysis of foundation shale bearing capacity, the applicant used an

undrained shear strength of 50 psi and angle of internal friction of 270 based

on laboratory shear tests. The static modulus of elasticity value.for the in

situ shale used by the applicant equals 10,000 psi. The staff finds these shear

strength parameters and the modulus of elasticity value for shale to be reasonable and acceptable.

The applicant determined the dynamic properties of shale from field geophysical

surveys. The compressional wave velocity values recorded ranged from 9,700 to

13,000 fps. The results of geophysical surveys are shown on FSAR Figure 2.5-109.

Based on results from the geophysical surveys of in situ soils, the applicant

has used a mean shear wave velocity of 1650 fps for the soil-structure interactions analyses of soil supported structures. The analyses use lumped mass

models on an elastic half space with strain independent soil properties. The

applicant has used a ±30 percent variation of the shear wave velocity to

account for uncertainty in soil properties. The-staff finds these soil proper- ties to be reasonable and acceptable. The applicant has indicated that for the

seismic analyses and soil a damping ratio of 10 percent was used for the soil

deposit underneath the structures. The staff finds this to be reasonable.

The staff will perform an audit of TVA's use of this damping ratio to confirm

that the applicant has properly accounted for the effect of the material damping and generic damping in the SSI analysis in the various modes of vibration

of the foundation of structures for the OBE and SSE conditions.

Results of the applicant's laboratory tests on in situ soils under Category I

structures are summarized in FSAR Tables 2.5-6 through 2.5-12. Results of

laboratory tests on the in situ basal gravel (terrace deposit) are presented

in FSAR Table Q362.28-1. The triaxial test results for the in situ terrace

deposits indicate an effective cohesion of 400 to 500 psf and an effective

angle of internal friction of 310 to 320. Based on laboratory tests the

applicant has conservatively adopted values of angle of internal friction for

basal gravel and shale to be 30' and 270, respectively. The staff finds the

above laboratory test results to be reasonable and acceptable.

The applicant has determined the groundwater level at the site, based on soil

borings in the powerblock area, to be at elevation 726 ft. This water level

is 2 ft below the 'final plant grade level. The applicant has installed a

permanent groundwater drainage system around the powerblock area to lower the

Watts Bar SER 2-34
groundwater elevation to 710 ft. Additional details of the groundwater conditions at the site are presented in Section 2.4 of this report.

2.5.4.3 Foundation Construction and Evaluation

Site preparation consisted of stripping surficial vegetation and unsuitable

material, excavating and backfilling (cut and fill) operations to attain a

nominal plant grade of 728 ft msl. After the top soil and organic materials

were stripped, in situ soil deposits were removed beneath the seismic Category I and other major structures supported on bedrock. The average depth of

excavation for the rock supported facility was approximately 40 ft below the

finished plant grade of 728 ft msl. During construction of the facilities,

the water in the excavated area from groundwater seepage or rainfall was controlled by using sumps in the excavation and pumping the water away from the

excavation.

Fill and Backfill Materials

Three types of backfill materials were used at the plant site beneath and

around seismic Category I structures: Class A backfill, granular fill, and

excavated soil from the intake channel.

Class A backfill was placed around all Category I structures at the site. The

backfill consisted of selected material placed in not more than 6-in. layers

and compacted to at least 95 percent of the maximum standard Proctor density

at or near optimum moisture content in accordance with ASTM D-698.

Granular fill material was placed to support some of the Category I structure

foundations (see FSAR Section 2.5.4.5.2). The fill was free from organic matter

and stones having a dimension larger than 1-1/4 in. It was compacted to at

least 70 percent relative density as determined by ASTM D-2049.

The fill material used to form the channel slopes in the intake channel and

the stabilizing dike to hold the channel open consisted of the soils excavated

from the intake channel. These soils were specified to be compacted to 95

percent maximum density at or near optimum moisture content in accordance with

ASTM D-698.

Stability of Foundations

Many of the Watts Bar seismic Category I structures in the powerblock area--

including the reactor buildings, auxiliary control building (control bay portion and auxiliary building portion), additional equipment buildings 1 and 2,

and north steam valve rooms--are supported directly on bedrock at a foundation

elevation of 683 ft. At the shore of Lake Chickamauga, the intake pumping

station housing the essential raw cooling water pumps, fire protection pumps,

and screen wash pumps is also founded on bedrock at elevation 648 ft.

,The waste packaging area of the auxiliary control building is supported at

elevation 725.25 ft on 27 ft of compacted granular fill underlain by bedrock.

The standpipe structures I and II are founded on about 20 ft and 14 ft, respectively, of in situ basal gravel overlying bedrock. The seismic Category I pipe

tunnels A and B from primary water storage tanks to the auxiliary building are

Watts Bar SER 2-35
also supported on the in situ basal gravel underlain by rock. The thickness

of the gravel layer under these tunnels is approximately 15 ft.

The diesel generator building, refueling water storage tank, and ERCW discharge

overflow structure are supported on granular fill overlying the in situ basal

gravel. The gravel is underlain by bedrock. The thickness of the compacted

fill under these structures ranges from 3 to 15 ft and the thickness of the

underlying in situ gravel ranges from 8 to 22 ft.

The condensate demineralizer waste evaporator building portion of the auxiliary

control building in the powerblock area and the ERCW pipe slabs at the intake

pumping station on the lake are supported on HP 12 x 74 and HP 12 x 53 steel

H-piles.

The earthfill at the intake pumping station is held back by two 32-ft long concrete retaining walls (wingwalls) that extend from the pumping station into

the intake channel. These concrete gravity retaining walls are keyed into rock

and are separated from the pumping station by expansion joints.

Two parallel sheetpile walls extend from each end of the intake pumping station

toward the main plant. These walls retain earthfill to elevation 710.0 ft.

The maximum projection of the walls above the adjacent sloping ground is 29 ft.

The ERCW pipeline and the Class IE electrical conduit are supported on the in

situ soils (alluvial and terrace deposits) underlain by rock.

Foundations on Rock

For all seismic Category I structures supported on bedrock, the in situ soils

were excavated to rock. All weathered material at thesurface of the bedrock

was then removed before establishing final grade for the base of the foundations. The freshly exposed shale was protected from deterioration by covering

the horizontal surfaces with at least 4 in. of concrete within 8 hours after

completion of excavation to final grade. Vertical surfaces were covered,

within 2 weeks after exposure, with a cast-in-place concrete wall.

During construction, a pocket of disintegrated shale was intersected in the

south wall of the reactor cavity for Unit 2 of the plant. A geologic investigation was made by the applicant to determine the conditions and extent of

this zone. The applicant concluded that the zone had its centroid at an

approximate elevation of 671 ft and the dimensions of the zone were approximately 8-ft wide, 12-ft long, and 5-ft thick. The outline of. the estimated

shape and extent of the zone is shown in FSAR Figures 2.5-140 through 2.5-143.

The disintegrated shale zone was satisfactorily grouted by the applicant with

a total of 255 gallons of grout using 22 bags of cement. During grouting, no

excessive grout takes were apparent. The applicant has provided reasonable

assurance that the disintegrated shale zone has been properly grouted. The

staff finds the procedure for establishing the general extent of disintegrated

shale zone to be reasonable and acceptable.

The rock-bearing capacity was evaluated and factors of safety were computed by

the applicant using conventional theories. The calculated allowable bearing

capacity provides a safety factor in excess of 3, which the staff agrees is

adequate for the support of the plant facilities on rock.

Watts Bar SER 2-36
The estimated settlements of structures on rock were computed by the applicant

by dividing the rock into layers and calculating the increase in vertical stress

at the center of each layer using the Boussinesq theory of stress distribution.

In this method of analysis, the loaded areas were assumed to be flexible. The

vertical deformation due to the increased stress for each layer was determined,

and the deformations for all layers were totalled to estimate the total settlement. The estimated total settlements for the structures founded on shale

range from approximately 0.6 to 0.8 in. The expected differential settlement

is computed to be less than 0.3 in. The measured settlements are close to or

less than the estimated settlements. The staff finds these differential and

total settlement values to be reasonable and acceptable, with the following

exception. The applicant has stated that the design 1-in. differential settlement between adjacent rock-supported structures was not incorporated into the

design of piping and electrical components that pass through adjacent rocksupported structures. This equipment is placed between the reactor building

of Unit 1 and the auxiliary building and between the reactor building of Unit 2

and the auxiliary building. By letter dated April 2, 1982, the applicant

has provided revised values of design differential settlements for these

piping and electrical components. The staff has reviewed this information and

finds the values to be reasonable. However, the staff has requested greater

detail of the applicant's analysis and will perform an audit of the applicant's

calculations to confirm these conclusions. The results of the confirmatory

audit will be reported in a supplement to the SER.

The applicant has designed all subsurface walls to resist static and dynamic

lateral earth pressures exerted by compacted Class A backfill. The computation procedure is consistent with the method described by Seed and Whitman

(1970) to calculate dynamic lateral loads in combination with static at-rest

pressures. This procedure is in accordance with the state of the art and is

acceptable to the staff.

Foundations on Granular Fill and In Situ Basal Gravel

The soil-supported Category I structures include

* the waste packaging area of the auxiliary control building

* standpipe structures I and II

* pipe tunnels A and B

* the diesel generator building

* the refueling water storage tanks

* the ERCW discharge overflow structure

The bearing capacities for the mat foundations of these structures were evaluated by the applicant using conventional methods. Because of the high strength

of the fill, basal gravel, and bedrock, and the loading and size of the mats

involved, the results of applicant's analyses show that bearing capacity factors

of safety for these foundations range from 9.5 to 20.0 which the staff agrees

are adequate. The computed factors of safety are listed in FSAR Table 362.28-4.

The applicant estimated the total and differential settlements for compacted

crushed stone and basal gravel underneath these bui.ldings to be negligible.

The observed settlements are less than 0.6 in. The staff concludes that the

estimated and observed settlements are reasonable and acceptable.

Watts Bar SER 2-37 .
Design lateral loads on the subsurface walls of these buildings were computed

by the applicant using a method consistent with the Seed and Whitman (1970) procedure. This state-of-the-art method is acceptable to the staff.

Pile Foundations

The condensate demineralizer waste evaporator (CDWE) building and the ERCW

piping and conduit support slabs are supported on steel H-piles. The applicant has provided the details of pile foundation design and installation in

FSAR Table Q362.32-2.

For the condensate demineralizer waste evaporator building, HP 12x74 piles were

driven through in situ soil with pile lengths ranging from 12 to 46 ft. The

soil parameters used in design are: angle of internal friction = 320,

cohesion = 0, and skin friction = 1800 psf. The allowable design capacity for

the piles is 260 kips for static compression and 325 kips for the SSE loading.

Piles supporting the CDWE were driven with a 15,000-ft-lb hammer until the rate

of penetration was less than 1/4 in. for five hammer blows. The piles were

also designed to resist the applied lateral loads as a result of OBE and SSE

conditions.

The piles under the ERCW piping and conduit support slabs were designed for

end bearing only; skin friction was not used in the design. HP 12x74 and

HP 12x53 piles ranging from 55 to 60ft in length were driven through Class A

compacted backfill to support the slabs. Piles supporting these slabs were

driven with a 15,000-ft-lb hammer until the rate of penetration was less than

1 in. for 48 hammer blows. Pile load tests were also conducted to confirm the

static load carrying ability of these piles. These piles also were designed

to resist the lateral loads as a result of-postulated earthquake loading conditions. There is no lateral load as a result of static conditions.

The applicant has provided reasonable assurance that proper procedures were

used in the design and installation of the pile foundations. The staff finds

these procedures acceptable.

Concrete Wingwalls and Sheetpile Retaining Walls

There are two concrete gravity-retaining walls holding back earthfill that

extend from the intake pumping station into the intake channel to a distance

of 32 ft. These wingwalls are keyed into rock. The factor of safety against

overturning of these wingwalls was computed by the applicant to be 3.96. The

staff agrees with the applicant that there is an adequate margin of safety in

the design of the retaining walls.

The two anchored sheetpile walls extend from each end of the intake pumping

station toward the main plant. These walls retain fill which supports ECCS

pipes and a service road. The maximum projection of these walls above the

backfill is 29 ft. The applicant has stated in the FSAR that he has investigated the stability of the sheetpile walls using saturated soil to an elevation

of 700 ft on one side of the wall, with no water on the opposite side. These

assumptions are reasonable and are acceptable to the staff. The applicant

previously stated in the FSAR that the sheet pile walls and anchorage system

adhere to allowable stresses and the walls are stable against sliding during

an SSE. However, by letter dated May 11, 1982, the applicant informed the

Watts Bar SER 2-38
staff that it has discovered errors in the final design calculations. As a

result of revised calculations the applicant has concluded that the sheet

pile walls will be overstressed in bending during an SSE.

The applicant has committed to instituting corrective field measures to reinforce these walls to eliminate such overstressing. The staff has requested

greater detail of the analysis and will perform an audit of the applicant's

calculations. The results of the confirmatory audit will be reported in a

supplement to the SER.

2.5.4.4 Liquefaction Potential

Many of the principal plant facilities are supported on mats or piles driven

into the underlying soil or weathered rock that is not susceptible to liquefaction. The remaining seismic Category I structures are founded on

compacted granular fill and the in situ basal gravel which the staff agrees are

also not susceptible to liquefaction under the postulated SSE loading. The

liquefaction potential of the in situ very-loose-to-loose submerged alluvial

silty sands and sandy silts that surround and support ERCW pipelines, electrical conduit, and the intake channel slopes is discussed below.

ERCW Pipeline

The ERCW piping runs from the intake structure on Lake Chickamauga to the powerblock area and furnishes water for cooling the reactor during emergency condition. There is additional fire protection piping along the same alignment.

The ERCW piping is approximately 5000 ft long. A major portion of this piping

has been founded in alluvial terrace deposits of loose-to-medium density. The

natural ground in the vicinity of the pipeline generally slopes downward on

one or both sides of the finished grade above the pipeline. The soil profile

along the seismic Category I ERCW pipeline was developed by the applicant from

the results of soil borings along the alignment of the pipeline. The borings

are approximately 100 ft apart. The applicant has presented the blow count

and soil classification data along with the results of stress-controlled

dynamic triaxial test data on soils susceptible to liquefaction along this

route. The staff has evaluated the results of the exploration and field and

laboratory test data for the soils beneath the ERCW pipelines. The available

data indicate the presence of very-loose-to-loose deposits of alluvial silty

sands and sandy silts as evidenced by blow counts ranging from 0 to 12 for

thicknesses ranging from 5 to 12 ft. Organic soil samples, indicating the

possible existence of unsuitable materials, were also recovered by the applicant

during the field investigations.

The staff reviewed the applicant's submittals and independently assessed the

liquefaction potential of the very-loose-to-loose alluvial, granular soils

underneath the ERCW pipelines and has concluded that these soils would be

susceptible to liquefaction for an SSE of 0.18 g horizontal ground motion on

rock. By letter dated March 17, 1982, the applicant submitted its evaluation

of the potential for liquefaction of the soils along the ERCW route. For

this analysis, TVA used an 0.18-g, peak acceleration at ground surface, which

redefined the site SSE. In addition the applicant used a modified evaluation

approach the basis of which is contained in a paper entitled "Evaluation of

Liquefaction Potential of Sand Deposits Based on Observations of Performance

in Previous Earthquakes" (1981), by H. B. Seed and I. M. Idriss. The paper

Watts Bar SER 2-39
recommends adding 7.5 to the normalized blow counts of silty sands below the

water table for the evaluation of the liquefaction potential of these soils,

and then applying the criterion previously developed for sands.

During a May 14, 1982 meeting, and by letter dated June 14, 1982, the staff

informed the applicant that the redefinition of the site SSE in the manner

proposed by the applicant was not acceptable. In addition, the staff informed

the applicant that use of the Seed and Idriss (1981) modified approach was not

acceptable to the staff at this time because it is not a proven approach for

evaluation of the potential for liquefaction of silty soils and the uncertainties in the data supporting this approach are too great to yield the level of

confidence appropriate for safety-related systems at a nuclear facility.

The staff is continuing its review of other information that has been submitted

by TVA and will pursue these matters with the applicant. The results of the

staff's evaluation will be presented in a supplement to this SER.

Class 1E Electrical Conduit Support

The Class 1E conduits run from the main powerblock area to the pumping station

on the shore of Lake Chickamauga. They furnish electrical power and control

for the pumps, valves, screens, control boards, and so forth at the intake

-pumping station.

The applicant has provided logs of borings along the approximately 2200 ft-long

alignment of the Class 1E electrical conduit. The borings are approximately

100 ft apart. The available borings indicate the presence of alluvial granular

terrace deposits of loose-to-medium density. The staff is concerned that a

portion of the alluvial granular soil deposit underneath the conduit (and in

the proximity of the ERCW pipelines) may be susceptible to liquefaction.

The measures taken to resolve the staff's concern of susceptability for liquefaction beneath the.ERCW pipelines would also be applicable to the affected area

beneath the Class 1E electrical conduit.

Intake Channel

The intake channel is approximately 800 ft long and extends from the end of

the reservoir to the intake pumping station. The bottom of the channel is at

elevation 660 ft; it is 50 ft wide at its bottom and has 4:1 (horizontal:

vertical) side slopes. The grade level at the crest of channel slopes is at

elevation 695 ft; the lake elevation is at 682.5 ft. The bedrock elevation

varies from 650 to 656 ft.

During excavation of the channel, a 15-ft-thick layer of silty sand extending

from elevation 665 ft to elevation 680 ft was found to be loose and potentially

susceptible to liquefaction. A dynamic two-dimensional analysis of the stability of intake channel slopes was performed. The applicant concluded that

this 15-ft-thick deposit of loose silty sands may liquefy under the postulated

SSE. Therefore, the applicant decided to remove this layer beyond the above

limits of excavation to a distance of 120 ft at elevation 665 ft and a distance of 45 ft at. elevation 695 ft (the crest of the slope). The excavated

material was compacted back in place to 95 percent of maximum density near its

optimum moisture content. The extent of the excavation and replacement of the

Watts Bar SER 2-40
loose material varied somewhat at different locations of the channel depending

upon the depth to bedrock. The major portion of the excavation and backfill

work was performed in a dry condition by building a dike at the mouth of the

intake channel near the reservoir. At the time of removal of this dike, some

of the excavation and backfill work was performed under water. In these areas,

the replacement fill used was rockfill instead of excavated material. Side

slopes of 4:1 also were used in these areas. The staff finds the methods

adopted by the applicant to ensure that soils under and around the intake

channel will remain stable during an SSE event to be acceptable. The stability of channel slopes is discussed in Section 2.5.5.

2.5.4.5 Conclusion

Based on the applicant's design criteria and,construction reports and based on

the results of applicant's investigations, laboratory and field tests, and

analyses presented in the FSAR, the staff has concluded that the site and plant

foundations will be adequate to safely supportthe Watts Bar nuclear plant and

to permit the safe operation of the ultimate heat sink in accordance with the

requirements of Appendix A to 10 CFR 100, pending satisfactory resolution of

the open item identified above.

2.5.5 Stability of Slopes

Except for the gently sloping ground that contains the ERCW pipelines discussed in Section 2.5.4.4, the slopes of the intake channel are the only

permanent seismic Category I slopes on the site. The finished grade at the

crest of the channel slopes is 695 ft. The bottom of the channel is at elevation 660 ft. The man-made slopes are graded to a slope of 4:1. Typical

as-built cross-sections of the channel slopes are shown on FSAR Figures 2.5-256

and 2.5-257. The cross sections of slopes near the reservoir end of the channel

where rockfill was used as replacement material in the excavated zone are shown

on FSAR Figures 2.5-258 and 2.5-259.

The stability of channel slopes was investigated for both static and dynamic

conditions. The soil strength parameters assumed in the analyses were based

on the laboratory test results on the remolded and in situ soils forming and

supporting the channel slopes. The effective-strength parameters used in the

slope stability analyses are as follows:

Lean clays cohesion =.1200 psf

angle of internal friction = 150

Remolded replacement soil cohesion = 1200 psf

angle of internal friction = 15'

In situ gravel cohesion = 0

angle of internal friction = 420

Rockfill cohesion = 0

angle of internal friction = 450

The stability of the intake channel slopes for static loading was investigated

using the slip-circle method of analysis. The design static conditions that

were considered for the slope stability are

Watts Bar SER 2-41
(1) normal operating condition with reservoir elevation 685 ft, groundwater

elevation 685 ft

(2) sudden drawdown of reservoir from elevation 685 ft to elevation 666 ft,

groundwater elevation 685 ft

(3) end of construction, channel dry and groundwater elevation 685 ft

The results of the applicant's analyses indicate that the channel slopes (4:1) have a minimum factor of safety of 2.5 against shear failure under the design static conditions. Based on these results, the staff concludes that for static

design loads the intake channel slopes are stable and present no safety hazard.

The applicant has made a seismic stability evaluation of the intake channel

slopes combining extreme events of sudden drawdown with the SSE. Pseudostatic

wedge analyses were performed to evaluate the dynamic stability of'slopes. The analyses of various cross sections of the channel slopes for the combination

of extreme events resulted in factors of safety ranging from 1.0 to 1.5.

The staff finds that the intake channel slopes have been analyzed by the

applicant in a reasonable manner, and, based on the results of analyses pre- sented by the applicant, the staff concludes that the intake channel slopes have an adequate margin of safety for all conditions noted in Regulatory Guide 1.27, Ultimate Heat Sink for Nuclear Plant, and meet the requirements of

Appendix A to 10 CFR 100. The intake channel slopes are, therefore, acceptable.

2.5.6 Embankments and Dams

There are no embankments 6r dams associated with the Watts Bar nuclear plant used for plant flood protection or for impounding cooling water required for

the operation of the plant, except the Chickamauga Dam, 57 mi downstream from

the Watts Bar. However, there are 12 major dams above the Watts Bar plant site

whose failure could influence plant-site flood levels. The applicant has

assessed the effects of the failure of the downstream Chickamauga Dam and the

12 upstream dams and has presented the results in FSAR Section 2.4. The staff

evaluation of these studies is in Section 2.4 of this report.

2.6 References

Agbabian Associates, "Correlation of Ground Response Spectra with Modified

Mercalli Site Intensity," Topical Report SAN/1011-113, 1977.

Bollinger, G. A., "The Giles County, Virginia, Seismic Zone Configuration

and Hazard Assessment," Conference on Earthquakes and Earthquake

Engineering in the Eastern United States, September 14-16, 1981,

Knoxville, TN.

Boore, D. M., W. B. Joyner, A. A. Oliver III, and R. A. Page, "Estimation of

Ground Motion Parameters," U.S. Geological Survey, Circular 795, 1978.

Chinnery, M. A., "A Comparison of the Seismicity of Three Regions of the

Eastern U.S.," Bulletin of the Seismological Society of America, Vol. 69,

pp. 757-772, 1979.

Watts Bar SER 2-42
Chung, D. H., and D. L. Bernreuter, "Regional Relationships Among Earthquake

Magnitude Scales," USNRC Report NUREG/CR-1457, 1980.

Eardley, A. J., Structural Geology of North America, Harper and Rowe, New York,

1962.

Gupta, I. N., and 0. W. Nuttli, "Spatial Attenuation of Intensities for Central

U.S. Earthquakes," Bulletin of the Seismological Society of America,

Vol. 66, pp 743-752, 1976.

Hadley, J. B., and J. F. Devine, "Seismotectonic Map of the Eastern United

States," U.S. Geological Survey Department of the Interior Publication

MF-620, 1974.

Joyner, W. B., D. M. Boore, and R. L. Porcella, "Peak Horizontal Acceleration

and Velocity from Strong-Motion Records Including Records from the 1979

Imperial Valley, California, Earthquake," U.S. Geological Survey, OpenFile Report 81-365, 1981.

King, P. B., "Tectonic Map of North America USGS," Department of the Interior,

Washington, DC, 1969.

Murphy, J. R., and L. J. O'Brien, "Analysis of a Worldwide Strong Motion Data

Sample To Develop an Improved Correlation Between Peak Acceleration,

Seismic Intensity and Other Physical Parameters," USNRC Report NUREG-0402

(prepared by Computer Sciences Corporation), 1978.

Nelson, A. E., and R. C. McDowell, "Status of Review of Southern-Appalachian

Study," forwarded by a letter dated April 9, 1980 from Robert Morris, USGS,

to Robert Jackson, NRC, 1980.

Nuttli, 0. W., "The Relationship of Sustained Maximum Ground Motion Acceleration and Velocity to Earthquake Intensity and Magnitude: Miscellaneous

Report S-73-1," Report 16, U. S. Army Corps of Engineers, 1979.

Nuttli, 0. W., "Similarities and Differences Between Western and Eastern United

States Earthquakes, and Their Consequences for Earthquake Engineering,"

Conference on Earthquakes and Earthquake Engineering in the Eastern United

States, September 14-16, 1981, Knoxville, TN.

Nuttli, 0. W., G. A. Bollinger, and D. W. Griffiths, "On the Relation Between

Modified Mercalli Intensity and Body-Wave Magnitude," Bulletin of the

Seismological Society of America, Vol. 69, 1979.

Richter, C. F., "An Instrumental Earthquake Magnitude Scale," Bulletin

of the Seismological Society of America, Vol. 25, 1935.

Rodgers, J., "The Tectonics of the Appalachians," New York, Intersciences

Publishers, 1970.

Seay, W. M., "Southern Appalachian Tectonic Study" Tennessee Valley Authority,

Knoxville, Tn, 1979.

Watts Bar SER 2-43
Seed, H. B., and I. M. Idriss, "Evaluation of Liquefaction Potential

of Sand Deposits Based on Observations of Performance in Previous

Earthquakes," ASCE 1981 Annual-Convention and Exposition, October 26-30,

St. Louis MO, 1981.

Seed, H. B., and R. V. Whitman, "Design of Earth Retaining Structures for

Dynamic Loads," ASCE Specialty Conference, Lateral Stresses and Earth

Retaining Structures, 1970.

Tennessee Valley Authority, "Justification of the Seismic Design

Criteria Used for the Sequoyah, Watts Bar, and Bellefonte Nuclear Power

Plants, Phase II."

"Justification of the Seismic Design Criteria Used for the Sequoyah, Watts

Bar, and Bellefonte Nuclear Power Plants, Phase II, Responses to NRC

Questions 1 through 6."

Trifunac, M.D., 1978.

Trifunac, M. D., and A. G. Brady, "On the Correlation of Seismic Intensity

Scales with the Peaks of Recorded Strong Motion," Bulletin of the

Seismological Society of America, Vol. 65, pp 139-162, 1975.

Watts Bar SER 2-44
3 DESIGN CRITERIA - STRUCTURE, COMPONENTS, EQUIPMENT, AND SYSTEMS

3.1 General

3.1.1 Conformance with General Design Criteria

In a letter dated December 28, 1981, the applicant stated that the Watts Bar

facility will meet all applicable NRC requirements and regulations, except

where justification for nonconformance has been submitted. The staff reviewed

the final design and the design criteria and concludes, subject to the applicant's adoption of the additional requirements imposed by the staff and the

exemptions granted as discussed in this report, that the facility has been

designed to meet the requirements of the GDC.

3.1.2 Conformance with Industry Codes and Standards

The staff review of structures, systems, and components relies extensively on

the application of industry codes and standards that have been used as accepted

industry practice. These codes and standards, as cited in this report and

attached bibliography (Appendix B), have been previously reviewed and found

accepted by the staff. They have been incorporated into the SRP.

3.2 Classification of Structures, Systems, and Components

3.2.1 Seismic Classification

GDC 2, in part, requires that nuclear power plant structures, systems, and

components important to safety be designed to withstand the effects of earthquakes without loss of capability to perform their safety function. These

plant features are those necessary to ensure (1) the integrity of the reactor

coolant pressure boundary, (2) the capability to shut down the reactor and

maintain it in a safe shutdown condition, or (3) the capability to prevent or

mitigate the consequences of accidents which could result in potential offsite

exposures comparable to 10 CFR 100 guidelines. The earthquake for which these

plant features are designed is defined as the SSE in 10 CFR 100 Appendix A.

The SSE is based on an evaluation of the maximum earthquake potential and it is

that earthquake, which produces the maximum vibratory ground motion, for which

structures, systems, and components important to safety are designed to remain

functional. Those plant features that are designed to remain functional if. an

SSE occurs are designated seismic Category I in Regulatory Guide 1.29. Regulatory Guide 1.29, "Seismic Design Classification," is the principal document

used in the staff review for identifying those plant features important to

safety that as a minimum, should be designed to seismic Category I requirements. The staff review of the seismic classification of (excluding electrical

features) of Watts Bar was performed in accordance with the guidance in SRP

Section 3.2.1, "Seismic Classification."

The Watts Bar structures, systems, and components important to safety that

are required to be designed to withstand the effects of an SSE and remain

Watts Bar SER 3-1
functional had been identified in an acceptable manner in Tables 3.2-1, 3.2-2a,

and 3.2-2b of the FSAR. These tables, in part, identify major components in

fluid systems, mechanical systems, and associated structures designated as seismic

Category I. In addition, piping and instrumentation diagrams in the FSAR identify

the interconnecting piping and valves and the boundary limits of each system

classified as seismic Category I. As a result of its review of Tables 3.2-1,

3.2-2a, and 3.2-2b and fluid system piping and instrumentation diagrams, the staff

concludes that the structures, systems, and components important to safety of

Watts Bar have been properly classified as seismic Category I items in conformance

with Regulatory Guide 1.29, except for the items discussed below.

Regulatory Position C.1.g of the guide identifies cooling water systems as

seismic Category I. Portions of the Watts Bar ERCWS which service areas containing essential equipment have been incorrectly classified and installed as

nonseismic Category I. The applicant is currently upgrading or replacing those

portions of the ERCWS that service air cooling units and chiller packages for

areas that contain essential equipment such as containment spray pump room

coolers, main control room air-conditioning water chillers, instrument room

water cooler, and centrifugal charging pump room coolers. Staff acceptance of

the ERCWS is contingent upon a satisfactory upgrading or replacement of those

portions of the system identified above to seismic Category I design requirements and verification that the applicant has adequately implemented these

revisions to the system. The staff will report on the resolution of this

confirmatory item in a Supplement to this Safety Evaluation Report.

In its review of FSAR Section 3.9, the staff confirmed that acceptable interfaces exist between seismic Category I and nonseismic Category I portions of

piping systems. All other structures, systems, and components that may be

required for operation of the facility are not required to be designed to

seismic Category I requirements, including those portions of seismic Category I

systems such as vent lines, fill lines, drain lines, and test lines on the

downstream side of isolation valves, and portions of these systems which are

not required to perform a safety function.

The applicant has combined FSAR Tables 3.2-2a and 3.2-2b in a recent amendment

to the FSAR. The staff will review the revised table to confirm that the revision does not affect the staff's conclusions regarding the acceptability of the

applicant's classifications. The staff will report its findings in a supplement to the SER. Pending confirmation of the above items, the staff concludes

that the structures, systems, and components important to safety at the Watts

Bar facility that are discussed above are designed to withstand the effects of

an SSE, remain functional, and are properly classified as seismic Category I

items in accordance with Regulatory Guide 1.29. This constitutes an acceptable

basis for satisfying, in part, the requirements of GDC 2, and they are, therefore,

acceptable.

3.2.2 System Quality Group Classification

GDC 1, requires that nuclear power plant systems and components important to

safety be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety function tobe performed. These

fluid system pressure-retaining components are part of the reactor coolant

pressure boundary (RCPB) and other fluid systems important to safety, where

Watts Bar SER 3-2
reliance is placed on these systems: (1) to prevent or mitigate the consequences

of accidents and malfunctions originating within the reactor coolant pressure

boundary, (2) to permit shut down of the reactor and maintain it in a safe shutdown condition, and (3) to retain radioactive material. Regulatory Guide 1.26,

"Quality Group Classification and Standards for Water, Steam, and Radioactive

Waste Containing Components of Nuclear Power Plants," is the principal document

used in the staff review for identifying on a functional basis the components

of those systems important to safety that are Quality Groups B, C, and D.

Section 50.55a of 10 CFR 50 identifies those American Society of Mechanical

Engineers (ASME) Code Section III, Class 1 components that are part of the RCPB.

Conformance of these RCPB components with Section 50.55a of 10 CFR 50 is discussed

in Section 5.2.1.1 of this Report. These RCPB components are designated in

Regulatory Guide 1.26 as Quality Group A. Certain other RCPB components which

meet the exclusion requirements of footnote 2 of the rule are classified Quality

Group B in accordance with Regulatory Guide 1.26. The staff review of the quality

group classification of pressure-retaining components of fluid systems important

to safety for Watts Bar was performed in accordance with the guidance in SRP

Section 3.2.2, "System Quality Group Classification."

The systems and components important to safety of Watts Bar have been identified

in an acceptable manner in FSAR Tables 3.2-1, 3.2-2a, and 3.2-2b. These tables,

in part, identify the major components in fluid systems such as pressure vessels,

heat exchangers, storage tanks, pumps, piping, and valves as well as in mechanical

systems such as cranes, refueling platforms, and other miscellaneous handling

equipment. In addition, FSAR piping andinstrumentation diagrams identify the

quality group classification boundaries of the interconnecting piping and valves.

As a result of its review of FSAR Tables 3.2-1, 3.2-2a, and 3.2-2b and the fluid

system piping and instrumentation diagrams, the staff concludes that pressureretaining components have been properly classified as TVA Class A, B, C, D, G,

and H components and are in general in conformance with Regulatory Guide 1.26,

except for the item discussed below.

Regulatory Position C.2.a of the guide identifies cooling water systems as constructed to Quality Group C standards. Portions of the ERCWS for Watts Bar have

been incorrectly classified and installed to Quality Group D standards. As noted

.in Section 3.2.1, the applicant is of upgrading or replacing those portions of

the ERCWS that service areas containing essential equipment. Staff acceptance

of the ERCWS is contingent upon a satisfactory resolution of this matter and

verification that the applicant has adequately implemented these revisions to

the system. The staff will report on the resolution of this issue in a supplement to this SER.

The codes and standards used in the construction of TVA Classes A, B, C, D, G,

or H components are identified in FSAR Tables 3.2.4, 3.2-5, and 3.2-6. The

staff finds the summary list of codes and standards used in the construction

of components acceptable. The applicant also has used the American Nuclear

Society (ANS) Safety Classes 1, 2a, 2b, 3, and Non-Nuclear Safety (NNS) as

defined in the August 1970 draft of ANS-18.2, "Nuclear Safety Criteria for the

Design of Stationary Pressurized Water Reactor Plants," in the classification

of system components. Safety Classes 1, 2a, 2b, 3, and NNS generally correspond to the NRC Quality Groups A, B, C, and D in Regulatory Guide 1.26 and have

been used by the applicant to supplement the TVA classification system. A

summary of the relationship of the NRC Quality Groups, TVA Classes, and ANS

Safety Classes are as follows:

Watts Bar SER 3-3
NRC Quality Group PWR Safety Classes TVA Classes

A 1 A

B 2a B

C 2b, 3 C, D

D NNS G, H

The staff has reviewed the use of ANS Safety Classes in TVA Classes in Table 3.2-1,

and finds the classification of components acceptable. Quality Group A (Safety Class 1) components of the RCPB will be constructed* in accordance with the ASME

Boiler and Pressure Vessel Code, Section III, Division 1, Class 1. Components in fluid systems important to safety that are classified Quality Group B (Safety Class 2a) will be contructed in accordance with ASME. Code, Section III, Division 1,

Class 2. Components in fluid systems important to safety that are classified

Quality Group C (Safety Classes 2b and 3) will be constructed in accordance with

ASME Code, Section III, Division 1, Class 3. Components in fluid systems that

are classified Quality Group D (Safety Class NNS) will be constructed to the

following codes as appropriate: ASME Code, Section VIII, Division 1 or 2;

ANSI B31.1.0 Power Piping; and storage tank codes such as API 620 and API 650.

The staff concludes that design of the components in fluid systems important to

safety in conformance with the ASME Code, the Commission's regulations, and the

guidance provided in Regulatory Guide 1.26 and ANS-18.2, provides assurance that

component quality is commensurate with the importance of the safety function of

these systems. This constitutes an acceptable basis for satisfying the requirements of GDC 1 and is,-therefore, acceptable.

3.3 Wind and Tornado Loadings

3.3.1 Wind Loading

All seismic Category I structures exposed to wind forces were designed to withstand the effects of design wind. The design wind specified has a velocity of

95 mph based on a recurrence interval of 100 years.

The procedures that were used to transform the wind velocity into pressure loadings on structures and the associated vertical distribution of wind pressures and gust factors are in accordance with American Society of Civil Engineers - Paper No. 3269. The staff finds these procedures, which were used to determine

the loadings on seismic Category I structures induced by the design wind specified for the plant, acceptable because they provide a conservative basis for engineer- ing design to ensure that the structures will withstand such environmental forces.

*Constructed, as used herein, is an-all-inclusive term comprising materials certification, design, fabrication, examination, testing, inspection, and certification required in the manufacture and installation of components.

Watts Bar SER 3-4
The use of these procedures provides reasonable assurance that, in the event of

design basis winds, the structural integrity of the plant seismic Category I

structures will not be impaired; consequently, seismic Category I systems and

components located within these structures are adequately protected and will

perform their intended safety function if needed. The applicant's conformance

with these procedures is an acceptable basis for satisfying the applicable

requirements of GDC 2.

3.3.2 Tornado Loading

All seismic Category I structures exposed to tornado forces and needed for the

safe shutdown of the plant were originally designed to resist a design-basis

tornado consisting of the following parameters: a 300-mph tangential wind

velocity; a 60-mph translational wind velocity; a total pressure drop of 3 psi

at the rate of 1 psi per second; and a postulated spectrum of tornado missiles

that was appropriate.

As stated in Section 2.3.1 of this SER, the rate of pressure drop was one-half

the value recommended by Regulatory Guide 1.76. Accordingly, the applicant was

asked to reanalyze the affected Category I structures for the faster pressure

drop of 2 psi per second as recommended by Regulatory Guide 1.76.

The applicant responded by stating that the rate of pressure drop does not affect

the design of nonvented and totally vented structures. The applicant did reanalyze the partially vented auxiliary building with respect to the higher rate of

pressure drop and found that only the refueling room was affected with a differential pressure change from 0.66 psi to 0.72 psi. The limiting structural element

in this room was the roof which was reanalyzed and found to be adequate.

The procedures that were used to transform the tornado wind velocity into pressure

loadings are similar to those used for the design wind loadings as discussed in

Section 3.3.1 of this report. The tornado missile effects were determined using

procedures discussed in Section 3.5.3 of this report. The total effect of the

design tornado on seismic Category I structures was determined by appropriate

combinations of the individual effects of the tornado wind pressure, pressure

drop, and tornado associated missiles. Structures were arranged on the plant

site and protected so that collapse of structures not designed for the tornado

would not affect other safety-related structures.

The staff finds that the procedures used to determine the loadings on structures induced by the design-basis tornado specified for the plant are acceptable

because they provide a conservative basis for engineering design to ensure that

the structures withstand such environmental forces.

These procedures provide reasonable assurance that, in the event of a designbasis tornado, the structural integrity of the plant structures that have to

be designed for tornados will not be impaired; consequently, safety-related

systems and components located within these structures will be adequately protected to perform necessary safety functions as required. Conformance with

these procedures is an acceptable basis for satisfying the applicable requirements of GDC 2.

Watts Bar SER 3-5
3.4 Water Level (Flood) Design

3.4.1 Flood Protection

To ensure conformance with the' requirements of GDC 2 with respect to protection

against flooding, the staff reviewed the overall plant design and protection

against the effects of flooding for those structures, systems, and components

whose failure could prevent shutdown or result in an uncontrolled release of

s'ignificant radioactivity: The applicant provided protection against the

inundation and the static and dynamic effects of postulated flooding of safetyrelated structures, systems and components by the "incorporated barrier" method

as defined in Regulatory Guide 1.102, "Flood Protection for Nuclear Power

Plants."

At Watts Bar, the probable maximum flood (PMF) level is 738.1 ft msl and the

design-basis flood (DBF) including wind-wave run-up is 740.1 ft msl (see

Section 2.4 of this SER for further discussion). All safety-related facilities, systems, and components are housed in seismic Category I structures, and

protection is provided from flooding up to plant grade at elevation 728.0 ft

msl. These Category I structures include the Unit 1 and Unit 2 reactor buildings, the auxiliary building, control building, diesel generator building, and

ERCW intake structure.

Only the reactor buildings will be maintained dry during flood-mode operation.

Walls and penetrations are designed to withstand all static and dynamic forces

imposed by the DBF. The exterior walls are not waterproofed; minor seepage

through the concrete walls is collected in the reactor building sump and pumped

out to the tritiated drain collector tank in the auxiliary building.

The diesel generator building will remain dry during flooding because the

lowest access into the building is above the DBF. All other structures,

including the service, turbine, control, and auxiliary buildings are expected

to flood as the water exceeds their grade level entrances. Safety-related

equipment that will experience flooding have been identified. The applicant

has stated that all equipment required to maintain plant safety during the

flood and for 100 days after the beginning of the flood is either designed to

operate submerged or located above the maximum flood level. The Watts Bar

flood warning plan provides for a minimum of 27 hours for preparation for flood

operation and sufficient time to accomplish a controlled reactor shutdown to

the hot-shutdown condition. The-reactor will remain in the hot-shutdown condition for the duration of flood operation. The staff has reviewed the applicant's list of safety-related equipment, which will be affected by flooding,

and concurs that the unavailability of those systems and components will not

prevent or adversely affect the accomplishment of a controlled reactor shutdown

to the hot-shutdown condition for the duration of the flood operation (up to

100 days).

Within structures, the protection of safety-related systems and components

against flooding caused by the failure of fluid systems is described under

Section 9.3.3 of this report.

Based on its review of the design criteria and bases, and safety classification

of safety-related equipment necessary for maintenance of safe plant shutdown

Watts Bar SER 3-6
during and following flood conditions, the staff concludes that the design of

the facility for flood protection conforms to the requirements of GDC 2 with

respect to protection against natural phenomena, conforms to the guidelines

of Regulatory Guide 1.102 concerning flood protection, and is, therefore,

acceptable.

3.5 Missile Protection

3.5.1 Missile Selection and Description

3.5.1.1 Internally Generated Missiles (Outside Containment)

Protection against postulated internally generated missiles outside containment

associated with plant operation--such as missiles generated by rotating or

pressurized equipment as identified in the requirements of GDC 4--is provided

by any one or a combination of compartmentalization, physical barriers, separation, and equipment design. The physical arrangements of components and systems

is the primary means of providing protection to safety-related equipment and

from damage resulting from internally generated missiles. Safety-related systems

are physically separated from nonsafety-related systems, and redundant components

of safety-related systems are physically separated such that a potential missile

cound not damage or incapacitate' both trains of the system. Stored fuel (new

and spent) is protected from damage from internal missiles that could result in

radioactive release as identified in the guidelines of Regulatory Guide 1.13,

"Spent Fuel Storage Facility Design Basis," by the fuel pool walls and by locating new and spent fuel in an area with no high-energy piping systems or rotating

machinery in the vicinity.

The applicant has provided an evaluation of potential missile sources on the

basis that a single failure could result in their becoming potential missiles

and has identified the turbine for the steam driven auxiliary feedwater pump,

and temperature/pressure wells and sensors to be the only credible missiles

outside containment. The evaluation verified that the physical barriers and

separation provide adequate protection for safety-related systems and components against the effects of these missiles. The staff concurs with the applicant's assumptions and evaluation for potential missiles outside containment.

Protection of safety-related components and stored fuel from the effects of

turbine missiles is discussed in Section 3.5.1.3 of this report.

The staff has reviewed the adequacy of the applicant's design to maintain the

capability for a safe plant shutdown in the event of internally generated

missiles outside containment. The staff concludes that, through the use of

compartmentalization, barriers, separation, and equipment design, the design

is in conformance with the requirements of GDC 4 with respect to missile protection; this meets the guidelines of Regulatory Guide 1.13 as it relates to

protection of spent fuel from missiles and is, therefore, acceptable.

3.5.1.2 Internally Generated Missiles (Inside Containment)

Protection against postulated internally generated missiles inside containment

associated with plant operation such as missiles generated by rotating or pressurized equipment as identified in the requirements of GDC 4 is provided by any

one or a combination of barriers, separation, and equipment design. The primary

Watts Bar SER 3-7
means of protection of safety-related equipment from damage resulting from

internally generated missiles is provided by shield walls within the reactor

building.

The applicant has provided an evaluation of potential missile sources inside

containment. The only credible missile sources are from high energy systems.

The credible missiles are identified as

(1) control rod drive mechanism housing plug, drive shaft assembly, and drive

shaft with mechanism

(2) pressurizer safety valve bonnet

(3) pressurizer relief block valve bonnet

(4) pressurizer relief valve bonnet

(5) pressurizer spray valve

(6) reactor coolant system and reactor coolant pump temperature and pressure

sensors (and wells)

(7) pressurizer heater elements

Kinetic energy, impact area, and penetration were determined for each of the

credible.missiles. The applicant's analysis verified that structures, shields,

or barriers provide adequate protection for safety-related equipment from impact

by those identified primary missiles and any secondary missiles generated. The

staff concurs with the applicant's assumptions and evaluation for potential

missiles inside containment.

The applicant has analyzed the potential for the reactor coolant pump flywheel

to become a missile source as a result of flywheel failure in accordance with

the guidelines of Regulatory Guide 1.14, "Reactor Coolant Pump Flywheel Integrity." The applicant's analysis evaluated the materials integrity of the flywheel under assumed overspeed conditions of the pump as a result of pipe break

at the pump discharge. This analysis verified that failure of the flywheel

does not occur and, thus, it is not a postulated missile source. The staff

concurs with the applicant's analysis.

The staff has reviewed the applicant's design to maintain the capability for

a safe plant shutdown in the event of internally generated missiles inside

containment. Through the use of barriers, separation, and equipment design,

the design is in conformance with the requirements of GDC 4 with respect to

missile protection and the guidelines of Regulatory Guide 1.14 regarding the

reactor coolant pump flywheel integrity, and is, therefore, acceptable.

3.5.1.3 Turbine Missiles

According to GDC 4, nuclear power plant structures, systems, and components

important to safety shall be appropriately protected against dynamic effects,

including the effects of turbine missiles.

The turbine generator trains at Watts Bar Units 1 and 2 have peninsular orientations relative to their reactor buildings. This configuration places the spent

fuel pools, control room, diesel generator building, ECCS,'and valve rooms 1A,

1B, 2A, and 2B, as well as the auxiliary building outside the low trajectory

turbine missile strike zones. The staff review indicates that ERCW conduit run

is the only safety-related system located within the low trajectory turbine

Watts Bar SER 3-8
missile strike zones. However, the position of ERCW conduit run relative to

the turbine is such that it is adequately protected by the presence of the

concrete turbine pedestal, turbine deck floor, and turbine building walls.

On the basis of its review, the staff concludes that the risk as a result of

potential turbine missiles for Watts Bar is acceptably low and that the facility layout is acceptable.

3.5.1.4 Missiles Generated by Natural Phenomena

GDC 2 requires that structures, systems, and components essential to safety be

designed to withstand the effects of natural phenomena; GDC 4 requires that

these same plant features be protected against missiles. The missiles of concern are those generated by natural phenomena such as tornadoes. The applicant

has selected three missile-spectra on which the analysis is based. Each spectrum includes the weight, velocity, kinetic energy, impact area, and height for

missiles in a tornado zone-1 site, as identified in-Regulatory Guide 1.76

Spectra A and C are applicable for seismic Category 1 structures. Spectrum B

is applicable only for the diesel generator building doors. Spectrum A is the

missile spectrum used for the design of those Category I structures at Watts

Bar identical to those structures at the Sequoyah facility.. Spectrum C is the

missile spectrum used for the design of Category I structures which are not

identical to those at Sequoyah (the ERCW intake structure). The missile

spectra used for Watts Bar were reviewed and approved at the CP stage before

issuance of the SRP. The staff has reevaluated the missile spectra and concl.udes that they are representative of missiles at the site and are, therefore,

acceptable. Discussion of the protection afforded safety-related equipment

from the identified tornado missiles is provided in Section 3.5.2 of this

report. Discussion of the adequacy of barriers and structures designed to

withstand the effects of the identified tornado missiles is provided in

Section 3.5.3 of this report.

Based on its review of the three missile-spectra, the .staff concludes that

they were appropriately selected and conform to the requirements of GDC 2 and

4, with respect to protection against natural phenomena and missiles, and the

guidelines of Regulatory Guide 1.76, with respect to identification of tornado

missiles generated by natural phenomena, and are, therefore, acceptable.

3.5.2 Structures, Systems, and Components To-Be Protected from Externally

Generated Missiles

GDC 4 states that all structures, systems, and components essential to the

safety of the plants shall be protected from the effects of missiles. The

applicant has identified all safety-related structures, systems, and components

requiring protection from externally generated missiles. These safety-related

structures are designed to withstand the effects of postulated tornado generated missiles, including vertical missiles, without damage to safety-related

equipment. The tornado spectrum is discussed in Section 3.5.1.4 of this

report.

The applicant has identified the plant outdoor features, including air intakes

and exhaust, which may be required to perform a safety function coincident with

or following the occurrence of a tornado. All safety-related structures are

Watts Bar SER 3-9
designed to withstand postulated tornado-generated missiles including vertical

missiles, without damage to safety-related equipment. All safety-related

systems and components and stored fuel are located within tornado-missile

protected structures or are provided with tornado-missile barriers with the

exception of the refueling water storage tanks (RWST). Failure of the RWST

will not compromise the safe controlled shutdown of the plant because the

required makeup for shutdown is usually obtained from the boric acid storage

tanks via the chemical volume and control system, located in the auxiliary

building (See Section 9.3.4 of this report for further discussion).

The diesel generator exhaust stacks which protrude approximately 2 ft above the

roof grade are susceptible to damage or incapacitating flow blockage due to

tornado missile impact. (The diesel generator system and components are

further discussed in Sections 8.3 and 9.5.4 of this report.) By letter dated

June 4, 1982, the applicant has committed to provide modifications to protect

the diesel generator from degraded operation as a result of a tornado-generated

missile impacting the diesel generator exhausts. The staff finds this commitment acceptable pending review of the details of the modification. The results

of this confirmatory review will be reported in a supplement to the SER.

Essential piping from the outdoor and ERCW intake structure is protected from

missiles thoughout its length by missile protection slabs or missile protection

slabs or seismic Category I pipe tunnels. The UHS, which is the Tennessee

River waterway complex, does not require protection from externally generated

missiles. (The UHS and the ERCW systems are discussed in Sections 9.2.5 and

9.2.1 of this report,.) Thus, the requirements of GDC 4 with respect to missile

protection and the specific guidance of Regulatory Guide 1.13, 1.27 ("Ultimate

Heat Sink for Nuclear Power Plants"), and 1.117 ("Tornado Design Classification")

concerning tornado-missile protection for safety-related systems, structures,

and components including stored fuel and UHS are met.

Based on its review, the staff concludes that those safety-related structures

systems, and components identified by the applicant as requiring protection

from externally generated missiles conform to the requirements of GDC 4 with

respect to missile and environmental effects and the guidelines of Regulatory

Guide 1.27 concerning the protection of safety-related plant features from

-tornado missiles, and are, therefore, acceptable pending review of the details

of the modifications to prevent degraded diesel generator operations.

3.5.3 Barrier Design Procedures

The analysis of structures, shields, and barriers to determine the effects of

missile impact was accomplished in two steps. In the first step, the potential

damage that could be done by the missile in the immediate vicinity of impact

was investigated. This was accomplished by estimating the depth of penetration

of the missile into the impacted structure. Furthermore, secondary missiles

were prevented by fixing the target thickness well above that determined, using

established methods of impactive analysis. The.equivalent.loads of missile

impact, whether the missile is environmentally generated or accidentally

generated within the plant, were combined with other applicable loads as is

discussed in Section 3.8 of this report.

The design procedures used to determine the effects and loading on seismic

Category I structures by design-basis missiles selected for the plant provide a

Watts Bar SER 3-10
conservative basis for engineering design to ensure adequate protection from

the effects of missile impacts.

The use of this information provides reasonable assurance that, in the event of

design-basis missiles striking seismic Category I structures, the structural

integrity of structures would not be impaired or degraded to an extent that

would result in a loss of required protecton. Seismic Category I systems and

components located within these structures are, therefore, expected to be

adequately protected against the effects of missiles. Conformance with these

missile protection design procedures is an acceptable basis for satisfying the

requirements of GDC 4.

3.6 Protection Against Dynamic Effects Associated with the Postulated Rupture

of Piping

3.6.1 Plant Design for Protection Against Postulated Piping Failures in Fluid

Systems Outside Containment

GDC 4 requires that systems and components important to safety be appropriately

protected against dynamic effects, including the effects of missiles, pipe

whipping, and discharging fluids, that may result from equipment failures. To

meet this requirement with regard to protection against pipe breaks outside

containment, the applicant should design the plant in accordance with the

criteria of BTP ASB 3-1. These criteria concern failure in high and moderate

energy fluid systems.

The applicant has identified all high- and moderate-energy piping systems in

accordance with these guidelines and has also identified those systems requiring protection from postulated piping failures.

The plant design accommodated the effects of postulated pipe breaks and cracks,

including pipe whip, jet impingement, and environmental effects. Protection of

safety-related systems and components are provided by physical separation,

enclosure within suitably designed structures, pipe whip restraints, and equipment shields. To be consistent with BTP ASB 3-1, the applicant used separation

as the primary means of protection, and where separation was not feasible, one

of the other acceptable methods of protection was used.-

The plant design includes the ability to sustain a high-energy pipe break accident coincident with a single active failure and retain the capability for safe

cold shutdown. For postulated pipe failures, the resulting effect will not preclude the habitability of the control room, and will not cause the loss of function of power supplies or controls and instrumentation needed to complete a safety

action as indicated in BTP ASB 3-1.

The main steam and feedwater lines outside the containment between the main

steam valve room and the turbine building are nonseismic Category I, and are

routed around the reactor building, auxiliary building, and control building

walls, open to the outside environment. Therefore, a subcompartment pressure

analysis as a result of a postulated pipe break was not required. The location of essential equipment in the vicinity was established to verify its

appropriate environmental qualification. The environmental qualification of

essential equipment, including the main steam and feedwater isolation Valves,

Watts Bar SER 3-11
auxiliary, feedwater system flow control valves, atmospheric dump valves, and

auxiliary feedwater turbine steam supply valves, is discussed in Section 3.11

of this report. The applicant provided a pressure analysis for the other

subcompartments outside containment that contain high-energy p.iping. The

results of this analysis indicated that safe shutdown of the plant is not

affected by the postulated break in these high-energy lines.

The applicant has analyzed the effects of moderate-energy line breaks outside

containment on safety-related systems. The moderate-energy systems also are

designed to meet the criteria set forth in BTP ASB 3-1. The staff has evaluated the applicant's analysis and concluded that it demonstrates that a postulated crack in a moderate-energy line will not cause a loss of function of any

safety-related system.

Based on its review, the staff conclude that the plant design and protection of

those areas and systems required for safe plant shutdown following postulated

events, including the combination of pipe failures and single active failure,

meets the criteria of BTP ASB 3-1 as related to the protection of safetyrelated systems and components from postulated piping system failures. Thus,

the design meets the requirements of GDC 4 regarding protection against pipe

failures in fluid systems outside containment and is, therefore, acceptable.

3.6.2 Determination of Break Location Dynamic Effects Associated with the

Postulated Rupture of Piping

GDC 4, "Environmental and Missile Design Bases," of 10 CFR 50, Appendix A,

requires that structures, systems, and components important to safety shall be

designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and

postulated accidents, including LOCA. These structures, systems, and components

shall be appropriately protected against dynamic effects, including the effects

of missiles, pipe whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power plant.

The staff review, conducted in accordance with SRP Section 3.6.2 (NUREG-0800),

pertains to the methodology used for protecting safety-related structures,

systems, and components against the effects of postulated pipe breaks both

inside and outside containment. The staff also reviewed the criteria used to

determine the location, size, and orientation of postulated failures and the

methodology used to calculate the resultant pipe whip and jet impingement loads,

which might affect nearby safety-related structures, systems, and components.

Reactor Coolant System (RCS) Main Loop Piping

The applicant has referenced Topical Reports WCAP-8082-P-A and WCAP-8172-A,

"Pipe Breaks for the LOCA Analysis of the Westinghouse Primary Coolant Loop,"

as the basis for determining the postulated design-basis pipe break locations

in the RCS. The break locations orientation and the transients used in the

Watts Bar analysis are identical to the transients described in WCAP-8082. A

comparison of the Watts Bar plant-.specific analysis with the WCAP-8082 reference analysis indicates that thermal moments at three locations for the Watts

Watts Bar SER 3-12
Bar analysis exceed the referenced analysis thermal moments, however, the

resulting stresses and usage factors for these locations are within acceptable

limits.

Details of the actual limited displacement break flow areas and the actual

break separation have been provided as well as structural details of the pipe

whip restraints and break locations. Both areas are acceptable.

It is the staff's position that the above referenced reports provide an acceptable basis for postulating breaks in the reactor coolant system piping, and

that the implementation of the criteria therein-provides a level of protection

equivalent to that resulting from the application of the criteria of Regulatory

Guide 1.46 ("Protection Against Pipe Whip Inside Containment").

Staff evaluation has shown that the applicant's RCS dynamic analysis criteria,

dynamic analysis models for jet thrust forces, are consistent with SRP Section 3.6.2.

Systems Other than RCS Main Loop

The criteria for defining break and crack locations and configurations, the

analytical methods used to define the forcing functions to verify the integrity

and operability of mechanical components, component supports, and piping systemsare adequate and in compliance with SRP Section 3.6.2. High energy pipe rupture

analysis has been provided for systems both inside and outside the containment.

Dynamic analysis has been provided for the mainsteam and feedwater system. No

break exclusion zones in the containment penetration region have been postulated

in these systems. Therefore, SRP Section 3.6.2 acceptance criteria relative to

augmented inservice inspection is not applicable to these systems. A listing

of the high-energy systems that are considered for pipe rupture analysis has

been provided. A summary has been provided of the results of the analyses of

these systems to demonstrate that essential systems, components, and supports

will not be impaired as a result of high-energy pipe break. SRP Section 3.6.2

specifies that exceptions taken by the applicant to the pipe break location and

configuration criteria in the section must be identified and the bases clearly

justified. As discussed in FSAR Question 010.26, the applicant has used definitions of high-energy fluid systems and moderate-energy fluid systems that differ

from those of BTP ASB 3-1 Appendix A.

Regulatory Guide 1.46 high-energy piping definition considers two regions of

temperature and pressure combinations that TVA has not classified as high energy.

One of these pressure temperature combinations has a maximum operating temperature

> 200'F and pressure <275 lb/in. 2 . This condition does not exist in any of the

piping analyses and, therefore, would not be of any consequence. The other

pressure temperature combination has a maximum operating pressure > 275 lb/in. 2

and temperature <200'F. This region of operation occurs in piping subject to

pump discharge pressure, or subject to leakage past pressure isolation valves

in lines connecting the reactor coolant system to low pressure systems. Piping

fitting these categories was investigated and it was determined that in all

cases, insufficient energy was available to produce equipment damage.

The assumption of a jet profile expansion half angle of 200 as stated in the

FSAR is less conservative than the assumption of a half angle not exceeding 100

Watts Bar SER 3-13
as stated in SRP Section 3.6.2-111.3. The applicant has, however, provided the

following additional justification for the design adequacy of safety-related

systems exposed to the consequences of jet impingement loads determined on the

basis of a 200 expansion half angle. Most of the general superstructure is large

enough in area so that the entire jet is intercepted and the total force is,

therefore, the same regardless of the jet expansion half angle used. The smaller

structures which intercept only a portion of the jet are loaded more severely by

the 100 jet. However, all intercepted active components are routinely assumed

to fail when immersed in the jet. Since the 200 jet is assumed by the applicant

to have an equivalent reach to the 100 jet and would impinge on a greater number

of targets, the staff concludes that the applicant has provided equivalent protection to that which would result from a jet profile based on a 10° half angle.

In the analysis of jet forces from pipe breaks of equal or lesser pipe size and

wall thickness, an assumption has been made in FSAR Section 3.9.3.4.2.1e that

these jet forces will not be considered during faulted conditions. The criteria

in SRP Section 3.6.2 and Regulatory Guide 1.46 dealing with this subject specifically covers only pipe impacts and was not intended to be extrapolated to deal

with jet impingement. By letter dated February 8, 1982, the applicant, provided

analysis to show that sufficient provisions have been included in the design of

the Watts Bar facility to mitigate unacceptable damage from jet forces from pipe

breaks of equal or lesser pipe size and wall thicknesses.

Based on its review of FSAR 3.6.2, the staff concludes that the pipe rupture

postulation and the associated effects are adequately considered in the plant

design, and, therefore, are acceptable and meet the guidelines of SRP Section 3.6.2 and the requirements of GDC 4. This conclusion is based on

(1) The proposed pipe rupture locations have been adequately assumed and the

design of piping restraints and measures to deal with the subsequent

dynamic effects of pipe whip and jet impingement provide adequate protection to the integrity and functionality of safety-related structures,

systems, and components.

(2) The provision for protection against dynamic effects associated with pipe

ruptures of the reactor coolant pressure boundary inside containment and

the resulting discharging fluid provide adequate assurance that design

basis loss-of-coolant accidents will not be aggravated by sequential

failures of safety-related piping, and emergency core cooling system

performance will not be degraded by these dynamic effects.

(3) The proposed piping and restraint arrangement and applicable design considerations for high and moderate energy fluid systems inside and outside

of containment, including the reactor coolant pressure boundary, will

provide adequate assurance that the structures, systems, and components

important to safety that are in close proximity to the postulated pipe

rupture will be protected. The design will be of a nature to mitigate

the consequences of pipe ruptures so that the reactor can be safely shut

down and maintained in a safe shutdown condition in the event of a

postulated rupture of a high- or moderate-energy piping systems inside or

outside of containment.

Watts Bar SER 3-14
3.7 Seismic Design

3.7.1 Seismic Input

The seismic Category ILstructures, systems, and components have been designed

for a safe shutdown earthquake having a maximum horizontal acceleration of

0.18 times the acceleration of gravity and a maximum vertical acceleration of

0.12 times the acceleration of gravity. These values are identical with those

of Sequoyah. As stated in Section 2.5 of this report, the site-specific

response spectra that were developed for Sequoyah also apply to the Watts Bar

site. These exceeded the original design spectra of Watts Bar.

In his submittal dated December 9, 1981, the applicant subsequently discussed

the adequacy of the Watts Bar structural design with respect to the new sitespecific spectra. Two major facts are mentioned. First, the Sequoyah structural design has been qualified for adequacy with respect to new spectra. The

staff agreed with the applicant and so stated in the Sequoyah Safety Evaluation

Report. The applicant also stated that the Watts Bar structures are much

stronger than those of Sequoyah; it has more reinforcing bars in the concrete

and a heavier steel containment structure. Because the new spectra are the

same for both plants, the above facts provide sufficient bases for the staff

to approve the Watts Bar design upon consideration of the new spectra.

To account for a possible soil difference at the two different sites, the applicant performed soil amplification analyses beginning from rock-based artificial

earthquake time histories to the base of the structures. The resultant spectra

were compared with those of the 84th percentile and it was found that there

exists a margin of factor of approximately two in maximum acceleration for all

frequencies. The staff finds that such margin justifies the adequacy of the

soil-supported structure for 84th percentile site-specific spectra.

Additional reasons for approving the seismic structural design are discussed

in Sections 3.7.2, 3.8.1, and 3.8.2.

Damping values used in the dynamic analyses are as presented in FSAR Table 3.7.2.

They are conservative values in relation to Regulatory Guide 1.61 ("Damping Values

for Seismic Design of Nuclear Power Plants").

3.7.2 Seismic Analysis

The scope of review of the seismic system and subsystem analysis for the plant

included the seismic analysis methods for all seismic Category I structures,

systems, and components. It included review of procedures for modeling, seismic

soil-structure interaction, development of floor response spectra, inclusion of

torsional effects, and evaluation of seismic Category I structure overturning.

The review included design criteria and procedures for evaluating interaction

of nonseismic Category I structures and piping with seismic Category I structures and piping and effects of parameter variations on floor response spectra.

The review also included criteria and seismic analysis procedures for reactor

internals and seismic Category I piping buried outside the containment.

The system and subsystem analyses were performed by the applicant on an elastic

basis. Modal response spectrum multidegree of freedom and time history methods

Watts Bar SER 3-15
form the bases for the analyses of all major seismic Category I structures,

systems, and components. When the modal response spectrum method was used,

governing response parameters were combined by the square root of the sum of

squares (SRSS) rule. However, the absolute sum of the modal responses was used

for modes with closely spaced frequencies. Three components of seismic motion

were considered: two horizontal and one vertical. Floor spectra inputs to be

used for design and test verifications of structures, systems, and components

are generated from the time history method, taking into account variation of

parameters by peak widening. A vertical seismic system dynamic analysis was

employed for all structural amplification in the vertical direction. Torsional

effects and stability against overturning were considered.

The lumped soil-spring approach was used to evaluate soil-structure interaction

effects on structural responses.

The staff concludes that the seismic system and subsystem analysis procedures

and criteria used by the applicant as discussed above provide an acceptable

basis for the seismic design.

3.7.3 Seismic Subsystem Analysis

The review performed under SRP Section 3.9.2 of NUREG-0800 included the applicant's analysis of all seismic Category I piping systems, which was discussed

in FSAR Section 3.7.3. In addition to normal operating loads, this analysis

also considers abnormal loadings such as an earthquake.

The scope of review of the seismic subsystem analysis for Watts Bar included

the seismic analysis methods for all seismic Category I piping systems, equipment, and components. It included review of procedures for modeling, combination of the three components of earthquake motion, combinaton of modal responses,

inclusion of torsional effects of eccentric masses, and determination of composite damping. The review included design criteria and procedures for evaluation

of the interaction of nonseismic Category I piping systems with seismic Category I piping systems, for evaluation of relative support motions of piping

systems that interconnect two or more components or span two or more floors.

The review also included criteria and seismic analysis procedures for reactor

internals and seismic Category I buried piping outside containment.

The subsystem analyses were performed by the applicant on an elastic basis.

Modal response spectrum multidegree of freedom, equivalent static load method

of analysis, and tests formed the basis for the analyses of all major seismic

Category I piping systems, equipment, and components. To ensure that an adequate number of lumped masses or degrees of freedom were used in the analyses

and to support the validity of the NSSS modeling techniques, the applicant

provided test results for reactor coolant loop primary piping and components

for the staff's review. It was noted that the measured frequencies and mode

shapes compared favorably to the calculated results. In addition, the benchmarking of computer codes used in the seismic analyses provided additional

assurance that the balance-of plant (BOP) piping systems, equipment, and components are properly evaluated. When the modal response spectrum method was used,

governing response parameters were combined by the SRSS rule. The absolute

sum of the modal responses was used for modes with closely spaced frequencies

for other than NSSS analyses.

Watts Bar SER 3-16
Regulatory Guide 1.92 and SRP Section 3.9.2 of NUREG-0800 recommend that the

maximum structural responses of each of the three components of earthquake

motion should be combined by taking the SRSS of the maximum codirectional

responses caused by each of the three components of earthquake motion at a

particular point of the structure or of the mathematical model. The applicant

considers seismic excitation in each of two orthogonal horizontal directions

and separately combines the associated responses using absolute sum with the

responses caused by excitation in the vertical direction. The larger of these

two combinations at a particular point of the structure is then used in the

seismic evaluation. The applicant's procedure for combining spatial components

has been previously used for other facilities and have been approved by the

staff as an acceptable alternative to the Regulatory Guide 1.92 ("Combining

Modal Responses and Spatial Components in Seismic Response Analysis") specified

methods.

Vertical ground accelerations were assumed to be two-thirds of the horizontal

ground accelerations and constant vertical load factors were employed only

where analysis showed sufficient vertical rigidity to preclude significant

vertical amplification in the seismic subsystem being analyzed.

As stated in the FSAR Section 3.7.3.10.2, constant vertical load factors are

not used for design of safety-related systems and components within the NSSS

scope of supply. With respect to BOP equipment, the static coefficient method

as prescribed in IEEE 344-1975, paragraph 5.3 was used. The peak acceleration

value of the applicable response spectrum was increased by a factor of 1.5 and

applied as an equivalent static load factor to the entire mass of the equipment

being evaluated. The static load method was used only for evaluation of

structural integrity and not for demonstration of operability or functional

capability of safety-related equipment.

The reactor internals were qualified on a generic basis for a Westinghouse four

loop plant. The response spectra and loads used in the generic analysis envelope the Watts Bar plant-specific response spectra and loads by a considerable

margin. The adequacy of the control rod drive mechanism when subjected to

seismic loadings is verified by a combination of test and analyses as djscussed

in Section 3.9.4 of this report. The seismic analysis of the reactor internals

is in accordance with Regulatory Guide 1.92 and system structural damping values

are in accordance with Regulatory Guide 1.61.

Based on its review of FSAR Section 3.7.3, the staff concludes that the applicant has met the guidelines of SRP Section 3.9.2 of NUREG-0800 and the relevant

requirements of GDC 2 with respect to demonstrating design adequacy of all

Category I systems, components, equipment, and their supports to withstand

earthquakes by meeting the regulatory positions of Regulatory Guides 1.61 and

1.92 and by providing an acceptable seismic systems analysis procedure and

criteria. The scope of review of the seismic system analysis included the

seismic analysis methods of all Category I systems, components, equipment, and

their supports. It included review of procedures for modeling, inclusion of

torsional effects, seismic analysis of Category I piping systems, seismic

analysis of multiple-supported equipment and components with distinct inputs,

justification for the use of constant vertical static factors and determination

of composite damping. The review has included design criteria and procedures

for evaluation of the interaction of non-Category I piping with Category I

Watts Bar SER 3-17
piping. The review also has included criteria and seismic analysis procedures

for reactor internals and Category I buried piping outside containment.

The system analyses are performed by the applicant on an elastic basis. Modal

response spectrum multidegree of freedom and time history methods form the bases

for the analyses of all major Category I systems, components, equipment, and their

supports. When the modal response spectrum method is used, governing response

parameters are combined by the SRSS rule. The double sum of the modal responses

are used for modes with closely spaced frequencies. The absolute sum of the

largest horizontal response and the vertical response is used for both the time

history and response spectrum methods. Floor spectra inputs to be used for design

and test verifications of systems, components, equipment, and their supports are

generated from the time history method, taking into account variation of parameters by peak widening. A vertical seismic system dynamic analysis will be

employed for all systems, and components, equipment, and their supports where

analyses show significant structural amplification in the vertical direction.

3.7.4 Seismic Instrumentation Program

The type, number, location, and utilization of strong motion accelerographs to

record seismic events and to provide data on the frequency, amplitude, and

phase relationship of the seismic response of the containment structure comply

with Regulatory Guide 1.12, Revision 1, "Instrumentation for Earthquakes."

Supporting instrumentation is being installed on seismic Category I structures,

systems, and components in order to provide data for the verification of the

seismic responses determined analytically for such seismic Category I items.

The installation of the specified seismic instrumentation in the reactor containment structure and at other seismic Category I structures, systems, and

components, that complies with Regulatory Guide 1.12 constitutes an acceptable

program to record data on seismic ground motion was well as data on the frequency and amplitude relationship of the response of major structures and

systems. A prompt readout of pertinent data at the control room can be expected

to yield sufficient information to guide the operator on a timely basis for the

purpose of evaluating the seismic response in the event of an earthquake. Data

obtained from such installed seismic instrumentation will be sufficient to

determine that the seismic analysis assumptions and the analytical model used

for the design of the plant are adequate and that allowable stresses are not

exceeded under conditions where continuity of operation is intended. Provision

of such seismic instrumentation complies with Regulatory Guide 1.12.

3.8 Design of Seismic Category I Structures

3.8.1 Steel Containment

Each reactor and its cooling system is enclosed in a separate containment structure which consists of a free-standing cylindrical steel shell, hemispherical

dome and reinforced concrete base located within a separate reinforced concrete

reactor (or shield) building. The containment was designed, fabricated, constructed and tested as a Class MC vessel in accordance with Subsection NE of

the ASME Code, Section III, Division 1. Loads include an appropriate combination of dead and live loads, thermal loads, seismic, and LOCA-induced loads,

including pressure and jet forces.

Watts Bar SER 3-18
The analysis of the containment was based on the elastic thin shell theory.

The'allowable stress and strain limits are those delineated in the applicable

sections of Subsection NE of the ASME Code, Section III, Division 1, for the

various loading conditions.

The staff has reviewed the criteria used in the analysis, design, and construction of the steel containment structure. Based on its review, the staff

concludes that the criteria account for anticipated loadings and postulated

conditions that may be imposed upon the structure during its service lifetime

are in conformance with established critieria, codes, standards, and guides

acceptable to the staff, in accordance with criteria now included in SRP

Section 3.8.2.

The use of these criteria, the loads and loading combinations, the design and

analysis procedures, the structural acceptance criteria, the materials, quality

control programs, and special construction techniques, and the testing and inservice surveillance requirements, provide reasonable assurance that, in the

event of earthquakes and various postulated accidents occurring inside and outside the containment, the structure will withstand the specified conditions

without impairment of structural integrity or safety function. (A seismic

Category I concrete shield building protects the steel containment from the

effects of wind and tornados and various postulated accidents occurring outside

the shield building. Conformance with these criteria constitutes an acceptable

basis for satisfying the applicable requirements of GDC 2, 4, 16, and 50.

The ultimate static positive internal pressure capability of the containment

shell was assessed by the applicant. The source of this pressure is burning of

accumulated hydrogen Within the containment. A detonation pressure time history

was used for local structure evaluation and has a triangular shape. The pressurerises from zero to a value of 180 psig in 0.1 millisecond then decaps to zero

in an additional 0.4 milliseconds. This curve was used to assess the local

dynamic effects of a local detonation.

The steel shell, its anchorage into the concrete base mat, penetrations, personnel lock and equipment hatch were assessed. The results of the static analysis

shows the minimum pressure retaining capacity is 53.5 psig and is based on the

ASME minimum code material strengths. The next limiting item is the electrical

penetrations which is rated by the vendor at 60 psig.

The local effects of a detonation of the hydrogen gas in the vicinity of the

containment shell, using the above described pressure time history, were

assessed. This assessment shows the effects of the pressure time history to be

local in nature and that they do not control the pressure retaining capacity.

The values found for Watts Bar are similar to those found in other ice condenser containments that have been analyzed. The containment analysis is

acceptable to the staff.

In the evaluation of the steel containment buckling, the applicant assumed an

axisymmetric shell for simplicity even though equipment hatches and other

penetrations introduce discontinuities to the shell at these openings. The

applicant stated that the shell was reinforced around the openings and, hence,

the use of the axisymmetric model was justified. The staff expressed concern

Watts Bar SER 3-19
that adequacy of such reinforcement was not well demonstrated. For this

reason, the staff has initiated an experimental research program at Los Alamos

Scientific Laboratory; the results are expected Within a few months. The

applicant agreed to verify his code with respect to experimental data when the

results from Los Alamos become available.

Los Alamos published its preliminary results in NUREG/CR-2165 dated June 1981.

Tests in the report indicated that reinforcement of the opening by the area

replacement method of ASME, which is used in the design by the applicant,

restored original buckling strength. At the same time, a theoretical calculation in the same report indicated that complete restoration of the buckling

strength is unlikely. The staff is continuing research for a further understanding. However, the applicant introduced a load factor for LOCA load to

compensate such uncertainty as opening reinforcement effect. The load factor

is 1.25 and the LOCA load is the primary load for causing compressive stress in

the containment shell. Moreover, the shell buckling design was based on

criteria that provide a safety factor of two for all loading conditions. The

current staff position and Regulatory Guide 1.57 ("Design Limits and Loading

Combinations for Metal Primary Reactor Containment System Components") recommend

a minimum safety factor of two. Therefore, the staff concludes that there is no

likelihood of buckling and the Watts Bar containment design is acceptable.

The staff requests the applicant to inform it of the progress of confirmatory

verification of TVA buckling methodology with the previously mentioned research

program. Data needed for the verification, such as shell geometry and loading

conditions, are available at this time.

3.8.2 Concrete and Structural Steel Internal Structures

The containment interior structures consist of a shield wall around the reactor,

secondary shield walls, and other interior walls, compartments, and floors, and

the ice condenser structural systems. The ice condenser unit consists essentially of a well insulated cold storage room in which ice is contained in

perforated metal baskets. The spaces between are supported by a steel frame.

The ice condenser system is contained in the annulus formed by the containment

wall and the crane wall, circumferentially over a 3000 of arc. A refueling

canal and equipment hatch are located in the remaining 600 of arc. The

interior structures were designed in accordance with the American Concrete

Institute (ACI) Standard 318-63 for concrete and the American Institute of

Steel Construction (AISC) specifications for structural steel that are acceptable to the staff.

The containment internal structures were designed and proportioned to remain

within limits established by us under the various load combinations. These

limits are based generally on the ACI Standard 318-63 and on the AISC specification for concrete and steel structures, respectively, modified as appropriate

for load combinations that are considered extreme.

The use of these criteria; the loads and loading combinations; the design and

analysis procedures; the structural acceptance criteria; the materials, quality

control programs and special construction techniques; and the testing and inservice surveillance requirements provide reasonable assurance that, in the

event of earthquakes and various postulated accidents occurring within the

Watts Bar SER 3-20
containment, the interior structures will withstand the specified design conditions without impairment of structural integrity or the performance of required safety functions. Conformance-with these criteria constitutes an acceptable basis for satisfying the applicable requirements of GDC 2 and 4.

3.8.3 Other Seismic Category I Structures

The shield building for each unit is a reinforced concrete structure consisting

of a foundation mat, a cylindrical wall, and a shallow dome. The foundation

mat, common to the shield building, containment vessel, and other interior

structures, forms the only structural tie between these structures. The shield

building was designed in accordance with the applicable criteria of ACI Standard 318-63, including modifications as required by the staff.

Seismic Category I structures other than the containment and its interior structures are of structural steel and concrete, or seismic Category I masonry walls.

The seismic Category I masonry walls are discussed later in this section. The

structural components of the steel and concrete structures consist of slabs,

walls, beams, and columns, except for the shield building which is of shell construction. The design method for concrete was in accordance with that specified

in the ACE Standard 318. Structural steel components were designed in accordance

with the AISC specifications. These documents are acceptable to the staff.

The criteria used in the the analysis, design, and construction of the steel

and concrete seismic Category I structures to account for anticipated loadings

and postulated conditions that may be imposed upon each structure during its

service lifetime are in conformance with established criteria, codes, standards,

and specifications acceptable to the staff, as now included in SRP Section 3.8.4.

The use of these criteria, the loads and loading combinations, the design and

analysis procedures, the structural acceptance criteria, the materials, quality

control, and special construction techniques, and the testing and inservice

surveillance requirements, provide reasonable assurance that, in the event of

winds, tornados, earthquakes and various postulated accidents occurring within

the structures, the structures will withstand the specified design conditions

without impairment of structural integrity or the performance of required safety

functions. Conformance with these criteria, codes, specifications, and standards

constitutes an accepatable basis for satisfying the applicable requirements of

GDC 2 and 4.

The general requirements with respect to materials, testing, analysis, design,

construction, and inspection related to the design and construction of Category I

masonry walls conform to the ACI 531-79 code. Conformance with the ACI 531-79

code is acceptable to the staff.

The loads and load combinations used in the analysis and design of Category I

walls are in conformance with staff criteria and are, therefore, acceptable.

The criteria used in the analysis and design of Category I masonry walls to

account for anticipated loadings that may be imposed on the structures during

their service lifetime are in conformance with the staff's criteria for masonry

walls, and with codes, standards and specifications acceptable to the staff.

The staff concludes that in the event of earthquakes and various postulated

Watts Bar SER 3-21
accidents, the Category I masonry walls will withstand the specified design condition without impairment of structural integrity. Conformance with these

criteria constitutes an acceptable basis for satisfying, in part, the requirements of GDC 2 and 4.

3.8.4 Foundations

The foundation of the containment is a concrete mat. It was analyzed to determine the effects of the various combinations of loads expected during the life

of the plant. Analysis was accomplished by means of selected structural codes

taking into account bending moment, shear, and soil pressure for a plate on a

elastic foundation. Foundations of the other major structures, such as the

fuel building, auxiliary building, and main control areas consist, likewise,

of reinforced concrete mats. Foundations were designed in accordance with the

ACI Standard 318.

The use of these criteria as defined by applicable codes, standards, and

specifications, the loads and loading combinations, the design and analysis

procedures, the structural acceptance criteria, the acceptance criteria, the

materials, quality control, and special construction techniques, and the testing and inservice surveillance requirements provide reasonable assurance that,

in the event of winds, tornados, earthquakes, and various postulated events,

seismic Category I foundations will withstand the specified design conditions

without impairment of structural integrity and stability or the performance of

required safety functions. Conformance with these criteria, codes, specifica- tions, and standards constitutes an acceptable basis for satisfying the applicable requirements of GDC 2 and 4.

3.9 Mechanical Systems and Components

The review performed under SRP Sections 3.9.1 through 3.9.6 ofNUREG-0800 (July 1981) pertains to the structural integrity and inservice testing program of

various safety-related mechanical components in the plant. The staff review is

not limited to ASME Code components and supports, but extends to other components such as control rod drive mechanisms, reactor internals, ventilation

ducting, cable trays, and safety-related piping designed to industry standards

other than the ASME Boiler and Pressure Vessel Code. The staff reviewed such

issues as load combinations, load combination methodology, allowable stresses,

methods of analysis, summary results, preoperational testing, and inservice

testing of pumps and valves. The staff review must determine that there is

adequate assurance of a mechanical component performing its safety-related

function under the required combinations of normal operating conditions, system operating transients, postulated pipe breaks, and seismic events.

3.9.1 Special Topics for Mechanical Components

The review performed under SRP Section 3.9.1 pertains to the design transients,

computer programs, experimental stress analysis and elastic-plastic analysis methods that were used in the analysis of seismic Category 1 ASME Code and nonCode items. The applicant has provided a complete list of transients to be

used in the design and fatigue analysis of all Code Class 1 and CS components

and of component supports and reactor internals within the RCPB.- The number of

events postulated for each transient has been included and is acceptable.

Watts Bar SER 3-22
Computer programs were used in the analysis of specific components. A list of

the computer programs used in the dynamic and static analyses to determine the

structural and functional integrity of these components is included in the FSAR

along with a brief description of each program. Design control measures, as

stated under 10 CFR 50, Appendix B, require that verification of the computer

programs be included. The applicant has provided verification for all of the

listed computer programs.

The Argonne National Laboratory was contracted by the staff to perform an

independent analysis of a sample piping system in the Watts Bar plant. The

pressurizer safety and relief valve piping was chosen as the sample system.

This analysis has been completed. It not only verifies that the sample piping

system meets the applicable ASME Code requirements, but also provides a check on

the applicant's ability to correctly model and analyze its piping systems.

Based on its review of FSAR Section 3.9.1, the staff concludes that the design

transients and resulting loads and load combinations with appropriate specified

design and service limits for mechanical components is acceptable and meets the

guidelines of SRP Section 3.9.1 (NUREG-0800) and the relevant requirements of

GDC 1, 2, 14, 15, 10 CFR 50, Appendix B, and 10 CFR 100, Appendix B. This conclusion is based on the following:

(1) The applicant has met the relevant requirements of GDC 14 and 15 by

demonstrating that the design transients and resulting loads and load

combinations with appropriate specified design and service limits, which

the applicant has used for designing Code Class 1 and CS components and

supports, and reactor internals provide a complete basis for design of

the reactor coolant pressure boundary for all conditions and events

expected over the service lifetime of the plant.

(2) The applicant has met the relevant requirements of GDC 2 and 10 CFR 100,

Appendix A, by including seismic events in design transients that serve

as design bases to withstand the effects of natural phenomena.

(3) The applicant has met the relevant requirements of 10 CFR 50, Appendix B,

and GDC 1 by having submitted information that demonstrates the applicability and validity of the design methods and computer programs used for

the design and analysis of seismic Category I Code Class 1, 2, 3 and CS

structures, and non-Code structures within the present state-of-the-art

limits and by having design control measures that are acceptable to ensure

the quality of the computer programs.

3.9.2 Dynamic Testing and Analysis of Systems, Components, and Equipment

The staff has reviewed the criteria, testing procedures, and dynamic analyses

employed by the applicant to ensure the structural integrity and functionality

of piping systems, mechanical equipment, and their supports under vibratory

loadings. This principal document used in this review is SRP Section 3.9.2

(NUREG-0800). This review encompassed several areas, each of which is

described below.

Watts Bar SER 3-23
3.9.2.1 Piping Preoperational and Startup Testing Program

As discussed in FSAR Section 3.9.2.1, piping vibration, thermal expansion,

and dynamic effects testing will be conducted during a preoperational testing

program. The purpose of these tests is to ensure that the piping vibrations

are within acceptable limits and that the piping system can expand thermally in a manner consistent with the design intent. During the Watts Bar preoperational and startup testing program, the applicant will test various piping

systems for abnormal steady-state or transient vibration and for restraint of

thermal growth. Systems to be monitored include all safety-related piping

systems designated as ASME Code Class 1, 2, and 3. Also included in the

program are those portions of high- and moderate-energy piping not defined as

ASME Code Class 1, 2, and 3 whose failure could reduce the functioning of any

safety-related feature.

Based on its review of FSAR Section 3.9.2.1, the staff concludes that the applicant has met the guidelines of SRP Section 3.9.2 (NUREG-0800) and the relevant

requirements of GDC 14 and 15 with respect-to the design and testing of the

reactor coolant pressure boundary. The vibration, thermal expansion, and dynamic

effects test program that will be conducted during startup and initial operation

on specified high and moderate energy piping and their associated restraints and

supports ensures that design conditions are not exceeded during normal operation,

including anticipated operational occurrences. The tests provide adequate assurance that the piping and piping supports have been designed to withstand vibrational dynamic effects resulting from valve closures, pump trips, and other

operating modes associated with the design-basis flow conditions. In addition,.

the tests provide assurance that adequate clearances and free movement of snubbers

exist for unrestrained thermal movement of piping and supports during normal

system heatup and cooldown operations. The planned tests will develop loads

similar to those experienced during reactor operations.

Because of the potential safety significance of failed snubbers in safety-related

systems and components, the staff requires that snubber operability be ensured

through a program of preservice examination and preoperational testing. The applicant's program of preservice examination-and preoperational testing of snubbers is in accordance with the requested format of Question 112.39 and is

acceptable.

3.9.2.2 Seismic Analysis

This subject is discussed in Section 3.7.3 of this SER.

3.9.2.3 Preoperational Flow Induced Vibration Testing of Reactor Internals,

The applicant has identified Indian Point Unit 3 as the prototype design

for a four-loop plant with regard to the testing of reactor internals for

flow-induced vibrations. The significant differences between the Watts Bar

and Indian Point designs are the 17 x 17 fuel assembly, the use of neutron pads rather than a thermal shield, and the upper head injection (UHI) upper internals assembly. The applicant has referenced the testing of the Trojan nuclear

plant in connection with the effect of neutron pads on core barrel excitation

and the preoperational test program for the Sequoyah plant in connection with

Watts Bar SER 3-24
the vibrations of upper internals. The staff has reviewed the design similarities of these reference plants to Watts Bar and have determined that the test

data from these plants is applicable to Watts Bar. The applicant has proposed

additional confirmatory vibration testing and subsequent visual inspection as

part of the Watts Bar preoperational test to provide added confirmation of the

capability of the structural elements of the reactor internals to sustain

flow-induced vibrations. The proposed program is consistent with Regulatory

Guide 1.20, "Vibration Measurements on Reactor Internals."

Based on its review of FSAR Sections 3.9.2.3 and 3.9.2.4, the staff concludes

that the applicant has met the relevant requirements of SRP Section 3.9.2

(NUREG-0800) and GDC 1 and 4 with respect to the reactor internals. (1) being

designed and tested to quality standards commensurate with the importance of

the safety functions being performed, (2) being appropriately protected against

dynamic effects by meeting the regulatory positions of Regulatory Guide 1.20

for the conduct of preoperational vibration tests, and (3) having a preoperational vibration program planned for the reactor internals that provides an

acceptable basis for verifying the design adequacy of these internals under

test-loading conditions comparable to those that will be experienced during

operation. The combination of tests, predictive analysis, and post-test inspection provide adequate assurance that the reactor internals will, during their

service lifetime, withstand the flow-induced vibrations of reactor without loss

of structural integrity. The integrity of the reactor internals in service is

essential to ensure the proper positioning of reactor fuel assemblies and unimpaired operation of the control rod assemblies to permit safe reactor operation

and shutdown.

3.9.2.4 Dynamic Analyses of the Reactor Coolant System

The applicant has performed dynamic system analyses to confirm the structural

adequacy of the reactor internals and unbroken loops of the reactor coolant

pressure boundary, including supports for the combined loads as a result of a

simultaneous LOCA and safe shutdown earthquake (SSE). Based on its review of

FSAR Section 3.9.2.5, the staff concludes that the applicant has met the guidelines of SRP Section 3.9.2 (NUREG-0800) and GDC 2 and 4 with respect to the

design of systems and components important to safety to withstand the effects

of earthquakes and the appropriate combinations of the effects of normal and

postulated accident conditions with the effects of the SSE by providing a

dynamic system analysis. This provides an acceptable basis for confirming the

structural design adequacy of the reactor internals and unbroken piping loops

to withstand the combined dynamic loads of a postulated LOCA and the SSE. The

analysis provides adequate assurance that the combined stresses and strains in

the components of the reactor coolant system and reactor internals will not

exceed the allowable design stress and strain limits for the materials of construction, and that the resulting deflections or displacements at any structural

element of the reactor internals will not distort the reactor internals geometry

to the extent that core cooling may be impaired. The methods used for component

analysis have been found to be compatible with those used for the system analysis.

The proposed combinations of component and system analyses are, therefore,

acceptable. The assurance of structural integrity of the reactor internals

under LOCA conditions for the most adverse postulated loading event provides

added confidence that the design will withstand a spectrum of lesser pipe breaks

and seismic loading events.

Watts Bar SER 3-25
3.9.3 ASME Code Class 1, 2, and 3 Components, Component Supports, and Core

Support Structures

The staff review under SRP Section 3.9.3 (NUREG-0800) is concerned with the

structural integrity and operability of pressure retaining components, their

supports, and core support structures which are designed in accordance with the

rules of the ASME Code, Section III, or earlier industry standards. The review

is divided into four parts, each of which is briefly described below.

3.9.3.1 Loading Combinations, Design Transients, and Stress Limits

The applicant has described acceptable load combinations, design transients,

and stress limits for all ASME Code Class 1, 2, and 3 components, component

supports, and core support structures. For primary equipment and piping, the

LOCA and SSE were combined by SRSS methods. For equipment supports, LOCA and

SSE were combined by absolute sum.

The applicant has provided yield to tensile ratios for materials used in structural supports and has verified that AISC-allowable stresses have been met at

temperature. The stress limits used for component support bolting have been

specified and these are less than yield stress at temperature for all loading

conditions. The staff finds this criteria to be acceptable.

Based on the information in FSAR Section 3.9.3.1, the staff concludes that the

applicant has met the guidelines of Section 3.9.3 (NUREG-0800), the requirements of 10 CFR 50.55a and GDC 1, 2, and 4 with respect to the design and service load

combinations and associated stress and deformation limits specified for ASME,

Code Class 1, 2, and 3 components by ensuring that systems and components

important to safety are designed to quality standards commensurate with their

importance to safety and that these systems can accommodate the effects of

normal operation as well as postulated events such as LOCAs and the dynamic

effects resulting from earthquakes. The specified design and service combinations of loadings as applied to ASME Code Class 1, 2, and 3 pressure retaining

components in systems designed to meet seismic Category I~standards are such

as to provide assurance that in the event of an earthquake affecting the site

or other service loadings due to postulated events or system operating transients,

the resulting combined stresses imposed on system components will not exceed

allowable stress and strain limits for the materials of construction. Limiting

the stresses under such loading combinations provides a conservative basis for

the design of system components to withstand the most adverse combination of

loading events without loss of structural integrity.

3.9.3.2 Pump and Valve-Operability Assurance Program

This review under SRP Section 3.9.3 evaluates the criteria which the applicant

has used to assure operability of active pumps and valves. Active pumps and

valves are those that must perform a mechanical motion to shut down the plant

or mitigate the consequences of an accident. For instance, under accident

conditions certain valves must open or close and certain pumps are required to

start. On the other hand, passive pumps and valves are only required to maintain their position during an accident.

Watts-Bar SER .3-26
After the review of this section, the staff finds that the component operability assurance program for ASME Code Class 1, 2, and 3 active valves and

pumps provides adequate assurance of-the capability of such active components

(1) to withstand the imposed design and service load without loss of structural

integrity, and (2) to perform necessary "active" functions and as valve closure

or opening or pump operation, during postulated events and conditions expected

when plant shutdown is required. The specified component operability assurance

test program complies with SRP Section 3.9.3 and satisfies the applicable portions of GDC 1, 2, and 4.

3.9.3.3 Design and Installation of Pressure Relief Devices

The review objective for this section is to ensure the adequacy of the design

and installation of pressure relief devices so that integrity of the relieving

devices is maintained during the functioning of one or more of the relief

devices.

Based on its review of FSAR Section 3.9.3.3, the staff concludes that the

applicant has met the guidelines of Section 3.9.3 (NUREG-0800) and the requirements of 10 CFR 50.55a and GDC 1, 2, and 4 with respect to the criteria used

for design and installation of. ASME Code Class 1, 2, and 3 overpressure relief

devices by ensuring that safety and relief valves and their installations are

designed to standards that are commensurate with their safety functions, and

that they can accommodate the effects of discharge due to normal operation as

well as postulated events such as LOCAs and the dynamic effects resulting from

the safe shutdown earthquake. The relevant requirements of GDC 14 and 15 also

are met with respect to ensuring that the reactor coolant pressure boundary

design limits for normal operation, including anticipated operational occurrences, are not exceeded. The criteria used by the applicant in the design and

installation of ASME Class 1, 2, and 3 safety and relief valves provide adequate assurance that, under discharging conditions, the resulting stresses will

not exceed allowable stress and strain limits for the materials of construction.

Limiting the stresses under the loading combinations associated with the actuation of these pressure relief devices provides a conservative basis for the

design and installation of the devices to withstand these loads without loss

of structuralintegrity or impairment of the overpressure protection function.

Relief and Safety Valve Test Requirements

Item II.D.1 of NUREG-0737 requires that applicants conduct testing to qualify

the reactor coolant system relief and safety valves, block valves, and associated

piping and supports, under expected operating conditions for design-basis transients and accidents.

The applicant has stated that he is a participant in the Electric Power Research

Institute (EPRI)-PWR safety and relief valve test program. Generic testing of

PWR relief and safety valves, associated piping, and supports are included in

this test program.

A description of the EPRI/PWR test program was provided to NRC by EPRI in

December 1979 and an updated revision of the program description was provided

in July 1980. The staff has reviewed these descriptions and is in agreement

that the NUREG-0737 technical requirements for relief and safety valves and

Watts Bar SER 3-27
associated piping and supports can be met on satisfactory completion of the

test program.

By letter dated July 1, 1981, from R. C. Youngdahl to Harold R. Denton, the PWR

Owners Group reported on the status of the EPRI/PWR safety and relief valve

test program and requested and extension of the completion dates specified in

NUREG-0737. The Owners Group stated its intention to develop an expanded test

matrix to obtain more information with respect to the effects of inlet piping

configurations and adjustments of ring settings on safety Valve operation.

On July 17, 1981, the staff met with EPRI and the PWR Owners Group representatives to review the status of the test program and discuss the expanded test

matrix. Although the exact number of additional tests will have to be determined as the program progresses, the test program managers estimated that it

could take from 4 to 8 months longer than the original test completion date of

July 1, 1981, to complete the expanded test program.

The PORV test results reported to date indicate that, although the initial

screening criteria were not met in some cases, all PORVs tested will function

in the primary mode to relieve pressure. Additional PORV tests have been

performed to evaluate the effect of variable water seal temperature on valve

closure times.

The safety valve test results indicate a need for additional information regard- ing the effects of inlet piping configuration and adjusting ring settings on

safety valve operation.

Based on a review of the EPRI test program to date, the staff has concluded that

the program represents a fully responsive effort to meet Commission requirements and that the additional testing proposed will provide needed information to

ensure that the technical requirements of Item II.D.1 of NUREG-0737 will be met.

The preliminary testing data received to date have not uncovered problems with

safety and relief valves that indicate the potential for occurrences that fall

outside the safety evaluation of operating plants, and the staff believes that

good cause has been shown to extend the NUREG-0737 completion date for PORV and

safety valve testing so that the expanded EPRI program may be carried to orderly completion. On this basis, the test completion dates were extended. Although

the extended test completion date was April 1, 1982, notification has been

received that all valve testing had been completed before the end of 1981.

The PWR Owners Group has not at thistime committed to a formal generic block

valve test program. During the EPRI steam testing of PORVs at the Marshall

facility, some testing was performed on several readily available block valves

even though there was no commitment to a test program.

At the July 17, 1981 meeting with the staff, the PWR Owners Group presented its

conclusion that it need not commit to any PORV block valve testing beyond that

already completed by EPRI for the following reasons:

(1) The isolation of a stuck-open PORV is not required to safely shut down a

plant, as shown by PWR NSSS vendor analyses.

(2) Post TMI plant procedures address means for dealing with stuck-open PORVs.

Watts Bar SER 3-28
(3) EPRI testing of PORVs has demonstrated that PORVs perform well over a wide

range of accident conditions.

The PWR Owners Group is, therefore, satisfied with the degree of operability

demonstrated at the Marshall facility for PORV block valves.

A report intended to justify these conclusions and the Owners Group assertion

that further block valve testing is not warranted has been received by the

staff in June 1982. After review of the block valve report, the staff will

determine further actions to ensure compliance with the NUREG-0737 block valve

qualification requirement. Nevertheless, it is the applicant's responsibility

to fulfill the NUREG-0737 requirements and provide plant-specific submittals by

July 1, 1982 or fuel load, whichever is later, that demonstrate that the block

valves are qualified.

By letter dated March 11, 1982, TVA submitted a revised response to Item II.D.1

of NUREG-0737. The response references the September 29, 1981 letter from

Darrell G. Eisenhut, which defines the revised schedule for the completion of

Item II.D.1 and the requirements for submittal of all generic and plant specific

information for staff review. The applicant stated that he intends to provide

this information to NRC before Unit 1 fuel load. The applicant also has stated.

that the submittal of the necessary plant-specific evaluations required by

NUREG-0737 depends on the preparation and submittal of the final valve test

reports to the utility by EPRI.

The staff requires that these submittals be made by fuel load in accordance

with the schedule of NUREG-0737 and the September 29, 1981 clarification letter

on this matter. After review of these submittals, the staff will report its

findings in a supplement to this SER.

3.9.3.4 Component Supports

To be acceptable, the component supports must provide adequate margins of.

safety under all plant operating conditions. Staff review under this section

includes the combination of loadings that were considered for each component

support within a system, including the designation of the appropriate service

stress limits for these loadings. The specified design basis loading combinations and allowable limits used for the design of safety-related ASME Code

Class 1, 2, and 3 component supports in systems classified as seismic Category I

provide assurance that all significant loadings are considered in the event of

an earthquake or an upset-, emergency-, or faulted-plant transient. With the

exception of the reactor coolant pump columns, the 2/3 critical buckling limit

was met for all Class 1 component supports based on absolute summation of LOCA

and SSE loads. For the reactor coolant pump columns, the limit was met when

LOCA and SSE loads were combined by SRSS method and actual material properties

were used. The staff finds this an acceptable basis for meeting the requirements of NUREG-0484, Revision 1. By letter dated January 29, 1982, the applicant provided analysis to show that short columns (L/r <100) do not tend to

buckle and that the design concepts at the Watts Bar facility preclude the use

of columns with L/r >100 (L is the effective length of the column and r is the

radius of gyration). Although the staff concurs with the applicant's conclusions

on columns with L/r >100, for columns with L/r <100, the staff disagrees with

the applicant's presentation and has given TVA a position stating the margin

Watts Bar SER 3-29
to be maintained to critical buckling. The applicant has not yet responded to

this position. The staff will report its findings in a supplement to the

SER.

The applicant has responded to IE Bulletin 79-02 for the Watts Bar facility in

a letter dated July 7, 1980. The staff will review the applicant's response

with respect to pipe support baseplate flexibility and its effect on anchor

bolt loads. Upon completion of its review of this submittal, the staff will

report its findings in a supplement'to the SER.

Based on its review of FSAR Section 3.9.3.4 and contingent upon the satisfactory resolution of the above items, the staff concludes that the applicant has met the

guidelines of Section 3.9.3 (NUREG-0800) and the requirements of 10 CFR 50.55a and

GDC 1, 2, and 4 with respect to the design and service load-combinations and

associated stress and deformation limits specified for ASME Code Class 1, 2, and

3 component supports by ensuring that component supports important to safety are

designed to quality standards commensurate with their importance to safety, and

that these supports can accommodate the effects of normal operation as well as

postulated events such as LOCAs and the dynamic effects resulting from the SSE.

The combination of loadings (including system operating transients) considered

for each component support within a system, including the designation of the

appropriate service stress limit for each loading combination, has met the posi- tions and criteria of Regulatory Guides 1.124 and 1.130 and are in accordance

with NUREG-0484, Revision 1. The specified design and service loading combinations used for the design of ASME Code Class 1, 2, and 3 component supports in

systems classified as seismic Category I provide assurance that in the event of

an earthquake or other service loadings as a result of postulated events or system

operating transients, the resulting combined stresses imposed on system components

will not exceed allowable stress and strain limits for the materials of construction. Limiting the stresses under such loading combinations provides a conservative basis for the design of support components to withstand the most adverse

combination of loading events without loss of structrual integrity.

Class CS component evaluation findings are covered in SER Section 3.9.5 in

connection with reactor internals.

3.9.4 Control Rod Drive Systems

The staff review under Section 3.9.4 (NUREG-0800) covers the design of the

hydraulic control rod drive system up to its interface with the control rods.

The information in FSAR Section 3.9.4 relative to the analyses and tests

performed to ensure the structural integrity and functionality of this system

during normal operation and under accident conditions was reviewed. In addition, the staff reviewed the life-cycle testing performed to demonstrate the

reliability of the control rod drive system over its 40-year life. Based on

its review of the above information, the staff concludes that the design of

the control rod drive system is acceptable and meets the guidelines of SRP

Section 3.9.4 (NUREG-0800) and the requirements of GDC 1, 2, 14, 26, 27, and

29, and 10 CFR 50.55a. This conclusion is based on the following:

(1) The applicant has met the requirements of GDC 1 and 10 CFR 50.55a with

respect to designing components important to safety to quality standards

commensurate with the importance of the safety functions to be performed.

Watts Bar SER 3-30
The design procedures and criteria used for the control rod drive system

are in conformance with the requirements of appropriate ANSI and ASME

Codes.

(2) The applicant has met the requirements of GDC 2, 14, and 26 with respect

to designing the control rod drive system to withstand effects of earthquakes and anticipated normal operation occurrences with adequate margins

to ensure its reactivity control function and with extremely low probability of leakage or gross rupture of reactor coolant pressure boundary.

The specified design transients, design and service loadings, combination

of loads, and limiting the stresses and deformations under such loading

combinations, are in conformance with the requirements of appropriate ANSI

and ASME Codes and acceptable regulatory positions specified in SRP

Section 3.9.3.

(3) The applicant has met requirements of GDC 27 and 29 with respect to

designing the control rod drive system to ensure its capability of

controlling reactivity and cooling the reactor core with appropriate

margin, in conjunction with either the emergency core cooling system or

the reactor protection system. The operability assurance program is

acceptable with respect to meeting system design requirements in observed

performance as to wear, functioning times, latching, and overcoming a

stuck rod.

3.9.5 Reactor Pressure Vessel Internals

The staff evaluated the load combinations, allowable stress, and deformation

limits that were used in the design of the reactor internals. The staff review

also included the structural integrity of reactor internals under the combination of loads which could result from postulated events such as the SSE and LOCA.

The review was performed in accordance with SRP Section 3.9.5 (NUREG-0800).

Although the reactor internals were not constructed to subsection NG of the

ASME Code, all of the Code requirements were satisfied with the exception of

the application of a Code N-Symbol stamp and the preparation of Code specificplant stress report. The staff reviewed a summary of stresses, deformation,

and usage factors for the Watts Bar reactor internals and agrees that allowable

limits were satisfied. In FSAR Amendment 45 the applicant clearly defined the

extent of compliance with the ASME Code and indicated that allowable limits

have been met.

Based on its review of FSAR Section 3.9.5. the staff concludes that the design

of the reactor internals is acceptable and meets the guidelines of SRP Section 3.9.5 (NUREG-0800) and the requirements of GDC 1, 2, 4, and 10 and 10 CFR

50.55a. This conclusion is based on the following:

(1) The applicant has met the requirements of GDC 1 and 10 CFR 50.55a with

respect to designing the reactor internals to quality standards commensurate with the importance of the safety functions to be performed. The

design procedures and criteria used for the reactor internals are in

conformance with the requirements of Subsection NG of the ASME Code,

Section III.

Watts Bar SER 3-31
(2) The applicant has met the requirements of GDC 2, 4, and 10 with respect

to designing components important to safety to withstand the effects of

earthquake and the effects of normal operation, maintenance, testing, and

postulated loss-of-coolant accidents with sufficient margin to ensure that

capability to perform its safety functions is maintained and the specified

acceptable fuel design limits are not exceeded.

The specified design transients, design and service loadings, and combination

of loadings as applied to the design of the reactor internals structures and

components provide reasonable assurance that in the event of an earthquake or

of a system transient during normal plant operation, the resulting deflections

and associated stresses imposed on these structures and components would not

exceed allowable stresses and deformation limits.for the materials of construction. Limiting the stresses and deformations under such loading combinations

provides an acceptable basis for the design of these structures and components

to withstand the most adverse loading events which have been postulated to occur

during service lifetime without loss of structural integrity or impairment of

function.

3.9.6 Inservice Testing of Pumps and Valves

The inservice testing program for pumps and valves is intended to demonstrate

that they will maintain operational readiness at any time during the plant

life. These tests and parameter measurements to detect long-term degradation

are performed in accordance with the rules of Section XI of the ASME Code to

verify that pumps and valves operate successfully when called upon. Staff

review under SRP Section 3.9.6 covers the applicant's program for preservice

and inservice testing of pumps and valves and emphasizes those areas of the

test program for which the applicant requests relief from the requirements of

Section XI of the ASME Code.

By letter dated October 31, 1981, TVA submitted a preliminary inservice program

for testing of pumps and valves, including required systems drawings, and

requested relief from certain pump and valve testing requirements. The staff

has not completed its detailed review of the applicant's submittal. However,

based on a preliminary review, the staff finds that it is impractical within

the limitations of design, geometry, and accessibility for the applicant to

meet certain ASME Code requirements. Imposition of those requirements would,

in the view of.the staff, result in hardships or unusual difficulties without a

compensating increase in the level of quality or safety. Therefore, based on

the resolution of the following concerns, and pursuant to 10 CFR 50.55a(g)(2)

and (g)(4)(i), the relief that the applicant has requested from the pump and

valve testing requirements of the ASME Code is granted for that portion of the

initial 120-month period during which the staff completes its reviews. The

applicant will be required to comply with Section XI of the ASME Code endorsed

by 10 CFR 50.55a 12 months before issuance of an operating license and/or the

Technical Specifications, the staff finds the Watts Bar inservice testing program

for pumps and valves acceptable for the period described above.

One area of concern during staff review was the periodic leak testing of pressure

isolation valves.

Watts Bar SER 3-32
There are several safety systems connected to the reactor coolant pressure

boundary that have design pressure below the rated reactor coolant system (RCS)

pressure. There are also some systems which are rated at full reactor pressure

on the discharge side of pumps but have pump suction below RCS pressure. In

order to protect these systems from the RCS pressure, two or more isolation

valves are placed in series to form the interface between the high-pressure RCS

and the low-pressure systems. The leaktight integrity of these valves must be

ensured by periodic leak testing to prevent exceeding the design pressure of

the low-pressure systems and thus causing an intersystem LOCA. Periodic leak

testing of pressure isolation valves shall be performed after all disturbances

to the valve are complete. The pressure isolation valves to be tested are

listed in the Technical Specifications.

The staff requires that the Watts Bar Technical Specifications contain limiting

conditions for operation that will require plant shutdown or system isolation

when the leakage limits are not met. The Technical Specifications will include

surveillance requirements which state the acceptable frequency of leak rate

testing and the acceptable value for leakage rate. The Technical Specifications

as discussed in the section will be based on the latest revision of NUREG-0452,

"Standard Technical Specifications for Westinghouse Pressurized Water Reactors."

However, the Technical Specification for the Watts Bar facility have not yet

undergone staff review. Based on the Standard Technical Specifications and the

applicant's commitment to perform periodic leak rate testing of pressure isolation valves between the reactor coolant system and low pressure systems, the

staff concludes that there is reasonable assurance that the design pressure of

the low-pressure systems will not be exceeded, and therefore, an intersystem

LOCA will not occur. This meets, in part, the requirements of GDC 55 of

Appendix A to 10 CFR 50.

3.10 Seismic and Dynamic Qualification of Seismic Category I Mechanical and

Electrical Equipment

The safety evaluation of the seismic and dynamic qualification of safetyrelated equipment consists of (1) a review of the methodology standards and

procedures as described in the relevant sections of the FSAR and (2) an onsite

audit of selected equipment to examine installation and to verify the completeness and adequacy of the qualification program including documentation. The

objective of the onsite audit is to develop the basis for the staff judgement

on the adequacy of the applicant's entire equipment qualification program based

on the results of a detailed review of a limited number of selected equipment.

A satisfactory finding by the staff can only be based on the acceptability of

the qualification methodology for the entire list of safety-related equipment

as judged by the site audit.

The staff has performed a review of the methodology and procedures for the

seismic and dynamic qualification program described in FSAR Sections 3.9.2 and

3.10 for the Watts Bar facility. The applicant's seismic and dynamic qualification criteria are generally based on IEEE 344-1971. Because the currently

acceptable standard is the IEEE 344-1975, various approaches are described in

the FSAR for equipment supplied by the reactor system vendor and those procured

under the applicant's purchase specifications to indicate how the requirements

of the IEEE 344-1975 standard can be met. The most significant approach is the

Watts Bar SER 3-33
conservatism of the input used for the original qualification. The conservatism argument is equipment specific, that is, depending on how conservative the

input loading is with respect to the design loading, and whether, or not the

equipment response i.s dominated by response from a single natural mode.

The seismic hazard for the Watts Bar site was redefined by the staff during the

operating license review. The seismic hazard is now specified in terms of a

site-specific spectrum which corresponds to the 84th percentile spectrum shape

derived from a number of recorded time histories of ground motion. The sitespecific spectrum is applicable for rock supported structures. For structures

that are supported by soil at the Watts Bar site, the applicant has performed

an amplification study for the soil overburden and used an amplified seismic

input motion. It is necessary to demonstrate that the seismic input for equipment qualification developed from (1) the use of site specific response spectra

for rock supported structures, and (2) the use of amplified input from sitespecific spectra for soil supported structures are'adequately covered by the

actual input spectra used to qualify equipment important to safety and that the

intent of the criteria described in the Regulatory Guides 1.92 and 1.100 are

satisfied.

The staff performs an onsite audit to judge the completeness and adequacy of

the implementation of the applicant's equipment qualification program for

seismic and dynamic loads. In a December 9, 1981 letter, the applicant submitted the results of ongoing studies to demonstrate such adequacy. The onsite

audit was performed during the last week of April 1982. The information

obtained during the audit and the December 9, 1981 submittal are under evaluation; the results will be reported in a future supplement to this report.

3.11 Environmental Qualification of Mechanical and Electrical Equipment

In December 1979,.the staff issued guidance for the environmental qualification

of safety-related electrical equipment in NUREG-0588,-"Interim Staff Position

on Environmental Qualification of Safety-Related Electrical Equipment." By

letters dated February 5, 1980 and February 21, 1980, the staff requested TVA

to review the environmental qualification documentation for each item of

safety-related electrical equipment which could be exposed to a harsh environment so as to identify the degree to which the associated environmental qualification program complies with the staff's position as described in this NUREG.

Further, where there are deficiencies, the staff requested the applicant to

commit to corrective action (requalification, replacement, relocation, etc.)

consistent with the requirements to establish qualification. If fuel loading

occurs before complete qualification can be obtained, justification for operation until the corrective actions are completed must be provided.

The Commission Memorandum and Order, CLI-80-21, dated May 23, 1980, directs

that safety-related electrical equipment be in compliance with NUREG-0588. The

applicable requirements for the Watts Bar facility are in Category II of

NUREG-0588. Full qualification of all equipment must be achieved by fuel load

or the Commission's deadline, whichever is later. TVA has indicated that their

environmental qualification program for safety-related electrical equipment

will be submitted in March 1983. The information requested in Appendix E of

NUREG-0588 must be furnished for review by the staff. To determine the extent

Watts Bar SER 3-34
to which safety-related mechanical equipment complies with the relevant requirements of GDC 4, the staff will review information submitted by the applicant.

Included in this review will be the equipment list, the environments for which

qualification must be demonstrated, and the qualification methods.

After the information requested has been received, the staff will conduct an

audit of the applicant's central files of qualification documentation for

safety-related electrical and mechanical equipment. The staff will report its

findings in a supplement to the SER.

Watts Bar SER 3-35

4 REACTOR

4.1 Introduction

The Watts Bar nuclear steam supply system (NSSS) i.s supplied by Westinghouse

Electric Corporation and is designed to operate at a core thermal power of

3411 MWt. Sufficient margin exists to ensure that fuel damage will not occur

during steady-state operation or anticipated operational occurrences.

The NSSS is a four-loop Westinghouse design having a minimum primary coolant

flow rate of 144.8 x 106 lb/hr. The reactor coolant and moderator is light

water at a pressure of 2250 psia. The reactor core is made up of 193 fuel

assemblies having identical mechanical designs but different fuel pellet enrichments. Each assembly is comprised of a 17 x 17 array containing 264 fuel rods.

The center-position is used for incore instrumentation while the remaining

24 thimbles are used for rod cluster control assemblies, neutron source assemblies, and burnable poison rods.

The design of the Watts Bar reactors is similar to that of the McGuire reactors.

4.2 Fuel System Design

4.2.1 Description

The fuel assemblies proposed for Watts Bar will consist of 264 fuel rods, 24

guide thimbles, and 1 instrumentation thimble arranged in a 17 x 17 array. The

instrumentation thimble will be at the center of the assemblies and will facilitate the insertion of neutron detectors. The guide thimbles will provide channels for inserting various reactivity controls. The fuel rods will contain

uranium dioxide ceramic pellets hermetically clad in Zircaloy-4 tubes. The

fuel assembly structure is held together by Zircaloy thimble tubes and the stainless steel fuel assembly nozzles at the top and bottom. Alignment and transverse spacings will be maintained by eight spacer grids separated uniformly

along the axis of the assembly.

All fuel rods will be internally prepressurized with helium during final welding

to reduce cladding compressive'stresses during service. The level of prepressurization is designed to preclude cladding flattening. The specific level of

prepressurization will depend on the planned fuel burnup and will be determined

before the Technical Specifications are established.

The Watts Bar fuel assembly design (17 x 17 array) is identical to that used

in a number of operating reactors manufactured by Westinghouse. Fuel design

differences between Watts Bar and another recently licensed Westinghouse reactor

(McGuire) are shown in Table 4.1-1 of the Watts Bar FSAR. There are no significant fuel-related changes except for the addition of boron carbide as an absorber

material.

Watts Bar SER 4-1
All aspects of the Westinghouse fuel design are based upon mechanical tests,

inreactor operating experience, andengineering analyses. Additionally, the

inreactor performance of the design is subject to the continuing surveillance

programs of Westinghouse and individual utilities. These programs provide

confirmatory and current design performance information.

4.2.2 Thermal Performance

In its evaluation of the thermal performance of the reactor fuel, the staff

assumes that densification of the uranium oxide fuel pellets may occur during irradiationin light-water reactors. The initial density of the fuel pellets

and the size, shape, and distribution of pores within the fuel pellets influence the densification phenomenon.

Briefly stated, inreactor densification (shrinkage) of oxide fuel pellets

(1) may reduce gap conductance, and hence increase fuel temperatures, because

of a decrease in pellet diameter; (2) may increase the linear heat generation

rate because of the decrease in pellet length; and (3) may result in gaps in

the fuel column as a result of pellet-length decreases (these gaps produce.

local power spikes and the potential for cladding creep collapse).

The engineering methods to be used by Westinghouse to analyze the densification

effects on fuel thermal performance have been previously submitted to the staff

in Westinghouse Topical Report WCAP-8219 and approved for use in licensing. The methods include testing, mechanical analyses, thermal and hydraulic analyses,

and accident analyses. The results of the staff's review are in the AEC report, "Technical Report on Densification of Westinghouse PWR Fuel" (May 1974). Additional information on densification methods can be found in "The Analysis of

Fuel Densification," NUREG-0O85.

A Westinghouse fuel thermal-performance code known as PAD-3.1 described in

attachments to correspondence from Westinghousb to the AEC* was used for the

Watts Bar safety analysis. A more recent Westinghouse fuel thermal performance

code, known as PAD-3.3 (Westinghouse Topical Report WCAP-8785), has also been

approved by the NRC. The more recent code, which contains revised models for

fission gas release, helium solubility, fuel swelling, and fuel densification,

was not used for the Watts Bar safety analysis.

The PAD-3.3 code addresses a concern over enhanced fission gas release at high

burnup. Because the previous version of the code, PAD-3.1, does not contain

the models necessary to analyze this effect, the staff's safety evaluation of

PAD-3.3 stated that future fuel performance analyses must be done with the

revised version of the code, PAD-3.3, according to a February 9, 1979 letter

from J. Stolz, NRC, to T. Anderson, Westinghouse.

The use of the earlier PAD-3.1 code, rather than the more recent PAD-3.3 code,

is generally acceptable because the earlier code produces more conservative

*Letters from R. Salvatori, Westinghouse, to D. Knuth, AEC, NS-SL-518 (December 22, 1972), NS-SL-521 (December 29, 1972), NS-SL-527 (January 1, 1973),

NS-SL-543 (January 12, 1973) (proprietary), and NS-SL-544 (January 12, 1973).

Watts Bar SER 4-2
thermal conditions than the revised code. This margin, however, does not exist

for high burnup fission gas release and fuel rod internal pressure calculations.

In a letter dated September 22, 1981, the applicant stated that the more recent

code, PAD-3.3, will be used to analyze the fuel thermal performance (including

fission gas release and rod internal pressure) at Watts Bar. The staff, therefore, concludes that the Watts Bar fuel performance analysis will be performed in an acceptable manner. The staff will confirm the acceptability of the result

of these calculations when they are submitted and report its findings in a

supplement to this report.

For the Watts Bar safety analysis, revised internal fuel rod pressure criteria,

as described in an approved Westinghouse Topical Report WCAP-8963-A, were used.

Briefly stated, these criteria allow the fuel rod internal pressure to exceed

the external system pressure. The approved criteria are: (1) the internal

pressure is limited so that the fuel-to-cladding gap does not increase during

steady-state operation, and (2) extensive departure from nucleate boiling (DNB)

propagation does not occur during postulated transients and accidents. Provided

that the as-docketed analyses are bounded by the conditions predicted by PAD-3.3

(in WCAP-8785), the staff concludes that these rod pressure criteria will be

satisfied for fuel burnups up to the peak target burnup.

The staff has reviewed Westinghouse Topical Report WCAP-8377, which describes

the details of a revised cladding-flattening model that, for a given fuel

region, predicts initial flattening time for pressurized rods containing relatively stable fuel. This revised analysis was based on the results of TV examinations of irradiated fuel rods, and the results indicated that the original

flattening model (Westinghouse Topical Report WCAP-7982) significantly underpredicted the time of collapse. The COLLAP computer code is used to perform

these calculations. The revised model was accepted for use in'safety analyses related to licensing subject to provisions specified in the safety evaluation

report (discussed in a staff memorandum from V. Stello to R. C. DeYoung,

January 14, 1975), which did not permit alterations to the specified curves

used as input to the model. In a letter dated September 22, 1981, the applicant stated that the cladding collapse model will be used in a manner consistent with conditions given in the staff's safety evaluation of WCAP-8377.

Therefore, the staff concludes that the cladding-collapse calculations will be

performed in an acceptable manner. The staff will confirm the acceptability of

the results of these calculations when they are submitted and report its findings in a supplement to this report.

4.2.3 Mechanical Performance

Although Westinghouse manufactures fueldesigns other than that used in Watts

Bar, substantially all of the Westinghouse inreactor operating experience with

Zircaloy-clad fuel rods is applicable because the 17 x 17 fuel design used is

only a slight mechanical extrapolation from the 14 x 14 and 15 x 15 fuel designs

used previously. Westinghouse accumulated operating experience with more than

30 commercial PWRs provides confidence in the acceptable performance of the fuel

assembly design. This inreactor experience represents more than 70 million MWhr

of thermal power generation.

Watts Bar SER 4-3
Fuel Rod Bowing

The consideration of fuel rod bowing in the 17 x 17 design was initially

analyzed by Westinghouse in Topical Report WCAP-8346. Subsequently, Westinghouse withdrew that analysis in light of new information and documented new

findings in WCAP-8692. The latter report was revised and was subsequently

resubmitted at NRC request, and is currently under review. The staff, however,

issued an interim safety evaluation report on Westinghouse fuel rod bowing in

April 1976, and in February 1977, issued a generic interim evaluation report.

In the February 1977 report, the staff gave approval to a burnup-dependent

approach to rod bowing, presented a formulation to extrapolate bow magnitudes

for new designs, and described a factor that increases rod bowing from asmeasured values (cold dimensions) to those in reactor (hot dimensions).. The

interim method is conservative and acceptable for use for Watts Bar.

In a letter dated September 22, 1981, the applicant stated that rod bow penalties consistent with the interim method will be used. The applicant has also

referenced partial bow DNB test data from Westinghouse.* Rod bow penalties

based on these data have been approved by the staff** and the approach is acceptable for use at Watts Bar. Although the staff concludes that fuel rod bowing

calculations will be performed in an acceptable manner, final resolution of this

issue will require that the applicant (1) identify in the basis of the Technical

Specifications any plant-specific or generic margins (credits) used to offset

the reduction in DNBR due to fuel rod bowing and (2) incorporate the residual

rod bowing penalty into the Technical Specifications. The staff will ensure

that the Technical Specifications will reflect the above items.

LOCA Loads

The fuel mechanical response to combined seismic and vertical LOCA loads from

postulated double-ended hot- and cold-leg breaks was analyzed in Topical

Report WCAP-8288. The staff found the methodology acceptable (NRC memorandum

from Stello to DeYoung, May 16, 1974) for 17 x 17 assemblies with either seven

or eight grid spacers. Westinghouse subsequently postulated a new asymmetric

(horizontal) hydraulic load caused by a postulated pipe break within the biological shield. Westinghouse has performed a preliminary analysis that indicated that the fuel assemblies will be able to accommodate this load. In a

letter dated March 1, 1976, Westinghouse stated that although the experiments

and calculational techniques supplied in WCAP-8288 may be applicable in assessing the adequacy of the fuel assembly to withstand these loads, it would be

expected that they would be reviewed on a plant-by-plant basis. The applicant

has performed an analysis for the most limiting primary system coolant pipe

break and states that the maximum grid'impact force for the combined LOCA and

safe shutdown earthquake (SSE) is approximately 35 percent of the minimum grid

strength. This results in an acceptable safety factor of 2.86. Therefore, the

response for Watts Bar fuel assemblies for seismic and LOCA loads has been

analyzed with acceptable methodology, and the results show that the assemblies

will be capable of accommodating such loads without crushing.

*Letters from Westinghouse to NRC, NS-TMA-1760, April 19,1978; NS-CE-1580,

October 24, 1977; and NS-CE-1598, November 8, 1977.

**Letter from J. F. Stolz, NRC, to T. Anderson, Westinghouse, April 5, 1975.

Watts Bar SER 4-4
Cladding Strain

Limitations on power rate changes will affect pellet/cladding interaction (PCI),

which is being reviewed as a generic item. The Westinghouse 17 x 17 fuel rod

design used in Watts Bar incorporates features that, compared with older Westing- house fuel designs, reduce cladding strain as a result of PCI. These features

include (1) pellet chamfering, (2) higher rod prepressurization, (3) lower linear

heat rating, and (4) smaller cladding diameter-to-thickness ratio. Based on the

available experimental and commercial reactor data, these design features should

result in a reduction or delay of PCI failures to later in the fuel design life.

Although the failure thresholds are probably lower at high burnup than at low

burnup, the fuel duty is also less severe. While PCI interaction is being studied

generically to determine if licensing criteria should be revised, the Watts Bar

design satisfies current criteria. Should licensing criteria related to PCI

change, the effects of such a change would be reviewed for all plants, including

Watts Bar.

The staff has reviewed the safety aspects of waterlogging fuel rod failures. A

1978 NRC survey (NUREG-0303) of available information included results of tests

in the capsule driver core at SPERT and the Japanese test reactor NSRR, and

observations of waterlogging failures in test and commercial reactors. The

survey concluded that (1) operating restrictions to reduce PCI also reduce the

potential for waterlogging failures during transients, (2) tests to simulate

accident conditions produced the worst waterlogging failures, and (3) there is

no apparent threat from waterlogging failures to the overall coolability of the

core or to safe reactor shutdown. The staff agrees that the evaluation of waterlogging failures as presented in the Watts Bar FSAR is correct and, therefore,

is acceptable.

Fuel assembly fretting and wear test results from 17 x 17 fuel assemblies were

reported in the Westinghouse hydraulic flow test report (WCAP-8279). These

tests with a seven-grid assembly indicated that fuel rod wear under both normal

and transient operating conditions was within the Westinghouse-predicted values

and that, even for fuel rods with deliberately damaged grid cells, the wear was

within acceptable limits. Westinghouse has since submitted the tests of an

eight-grid 17 x 17 fuel assembly flow test (Westinghouse letter NS-CE-640 to

V. Stello, NRC, May 15, 1975). These tests, which simulated actual inreactor

conditions, showed that the 17 x 17 fuel assembly design was acceptably resistant to fuel rod fretting at grid contact locations.

Guide Thimble Tube Wear

In response to the staff's attempt to "assess• the susceptibility and impact of

guide thimble tube wear in Westinghouse plants, two meetings were held with

Westinghouse, and information was submitted by the applicant* on the issue.

This information consisted of guide thimble tube wear measurements taken on

irradiated fuel assemblies from Point Beach Units 1 and 2 (two-loop plants

using 14 x 14 fuel assemblies). Also described were a mechanistic wear model

*Letter from L. M. Mills, TVA, to L. S. Rubenstein, NRC, November 27, 1975,

and Westinghouse letter NS-2238, to H. R. Denton, NRC,.April 29, 1980.

Watts Bar SER 4-5
(developed from the Point Beach data) and the impact of the model's wear

predictions on the safety analyses of plant designs.

Westinghouse believes that its fuel designs will experience less wear than that

reported in some other NSSS designs because the Westinghouse designs use thinner,

more flexible control rods that have relatively more lateral support in the

guide tube assembly of the upper core structure. Such construction provides the

housing and guide path for the rod cluster control assemblies (RCCAs) above the

core and thus restricts control rod vibration as a result of lateral exit flow.

Also, Westinghouse believes that its wear model conservatively predicts guide

thimble tube wear and that even with the worst anticipated wear conditions (both

in the degree of wear and the location of wear) its guide thimble tubes will be

able to fulfill their design functions.

The staff concluded that the Westinghouse analysis probably accounts for all of

the major variables that control this wear process. Because of the complexities

and uncertainties in (1) determining contact forces, (2) surface-to-surface wear

rates, (3) forcing functions, and (4) extrapolations of these variables to the

new 17 x 17 fuel assembly design, the staff required several near-term OL applicants to submit to a surveillance program. For acceptability, the minimum

objective of the program was to demonstrate that there is no occurrence of hole

formation in rodded guide thimble tubes.

To satisfy this request for confirmation of the Westinghouse analytical predictions, a cooperative owners group was established which sponsored a program

to obtain postirradiation examnination (PIE) data from Salem Unit 1. Members

participating in the program were Duke Power Company, TVA, Public Service Electric and Gas Company, Pacific Gas and Electric Company, Virginia Electric and

Power Company, and Westinghouse. The surveillance data were obtained'on fuel

discharged from Unit 1 of Salem, a RESAR-3S plant similar to those owned by

each participating utility. The top 20 in. of all 24 rodded thimbles in each

of 6 fuel assemblies were selected for examination. Four of these assemblies

had experienced one cycle of operation and two had experienced two full cycles

of operation.

The assemblies were selected from various core locations in order to assess

any specific wear dependency on core location (for example, by virtue of their

different locations within the core, the assemblies had been subjected to

various coolant flow velocities in the control. rod cluster guide column above

the upper core plate.)

The purpose for the examinations was to .determine the presence of any wear

holes of a size greater than 1/16 in. in diameter; however, the experimental

techniques actually empolyed allowed for a much greater resolution of the

acquired wear. Specifically, the boroscope and television monitor that were

used permitted qualitative resolution of wear depths corresponding as little

as surface scratches.

There were no wear holes-or unusual wear found in any of the 144 thimble tubes

that were examined. Occasional longitudinal scratches were observed, but the

scratches had no appreciable depth. The most pronounced scratches were typically 3- to 8-in. long 1/30 sec/in, wide.

Watts Bar SER 4-6
Accordingly, the staff concludes that the Westinghouse 17 x 17 fuel assembly

design is resistant to guide thimble tube wear, and therefore, acceptable for

use in the Watts Bar reactor.

Core Hardware

An additional fretting problem has arisen in some fuel rods that are adjacent

to baffle plate joints on the periphery of the core. The baffle plates are not

always tightly joined, and pressure differences across the baffle sometimes

result in cross flow impingement on nearby fuel rods. In several instances,

this baffle jetting has resulted in failure of one or two isolated fuel rods.

Baffle jetting is a problem for a limited number of plants that utilize downflow baffle cooling with a large pressure differential across the baffle plates.

More recent Westinghouse designs, which include Watts Bar, utilize a baffle

upflow design. An advantage of this design is that the pressure differential

across the baffle plate is small, which eliminates the potential for excess

bypass flow. Thus, the potential for damage as a result of baffle jetting is

eliminated.

Recent experiences in Westinghouse plants have indicated an actual or potential

operating problem with some of the core hardware items. Specifically, these

items are spacer grids and control rodlet fingers. Spacer grid damage to a

significant number of grids occurred at Salem Unit 1 during a refueling operation. Westinghouse and NRC have issued notices (NRC Information Memo 19) with

recommended revisions to operating procedures that should eliminate this problem.

At Salem, eight control rodlet fingers in a single reactor core failed from

stress corrosion cracking, allowing single control rodlets to be inserted into

fuel assemblies. The probable cause for stress corrosion cracking has been

identified as a tapping lubricant, and this lubricant has been eliminated from

the manufacturing process. The most significant effect of dropped rodlets is

in the core physics area. Local regions of power depression will occur and

power tilts may result. Westinghouse has issued a letter (NS-TMA-2214, March 14,

1980) on this subject with recommendations for increased attention to the hot

zero power flux maps. No significant safety problem from dropped rodlets is

anticipated at the Watts Bar facility should these recommendations be followed.

In summary, for the actual or potential problem areas for the two core-hardware

items discussed above, no design changes have been required for the items

concerned. The staff concludes that these problems have been satisfactorily

addressed for Watts Bar, provided the facility follows the Westinghouse

recommendations.

Materials

The staff has been generically evaluating three materials models that are used

in ECCS evaluations. Those models predict cladding rupture temperature, cladding burst strain, and fuel assembly flow blockage. The staff has (1) discussed

its evaluation with vendors and other industry representatives, (2) published

NUREG-0630, "Cladding Swelling and Rupture Models for LOCA Analysis," which

concluded that the Westinghouse cladding models were nonconservative, and

(3) required licensees to confirm that their operating reactors would continue

Watts Bar SER 4-7
to be in conformance with 10 CFR 50.46 if the NUREG-0630 models were substituted for the present materials models in their ECCS evaluations and certain

other compensatory model changes were allowed (letter from D. G. Eisenhut to

all operating LWRs, November 9, 1979, and memorandum from H. R. Denton to the

Commissioners, November 26, 1975).

Although revised Westinghouse cladding models have been reviewed and accepted

the staff's review of the new ECCS evaluation model has not been completed.

Therefore, the staff requested that supplemental calculations be performed for

the ECCS analyses in the Watts Bar FSAR with the materials models of NUREG-0630

unless a new ECCS model is available. In a letter dated September 22, 1981,

the applicant submitted the results of the supplemental ECCS calculations which

show that Watts Bar will continue to meet the acceptance criteria of 10 CFR 50.46

if the FQ peaking factor is reduced to 2.31. Consequently, the Technical Specifications must reflect-the new Watts Bar FQ value of 2.31. The staff, therefore,

concludes that the issue of cladding swelling and flow blockage has been

resolved for Watts Bar.

4.2.4 Surveillance

At the Watts Bar facility, performance of the fuel is indirectly monitored by

measurement of the activity of the primary coolant for compliance with Technical Specification limits. Westinghouse has proposed a fuel surveillance program for several plants that will use the 17 x 17 fuel assemblies. A summary

of this program is given in the fuel rod bowing report, WCAP-8692. This program includes lead assemblies in Surry Units 1 and 2 and the initial core loadings for Trojan, Beaver Valley Unit 1, Farley Unit 1, and Salem Unit 1.

The Surry units each have two lead burnup 17 x 17 fuel assemblies. One of the

lead assemblies in each unit has removable rods. These assemblies were carefully

measured prior to insertion and will be examined between cycles for dimensional

changes, fretting corrosion near the spacer grids, fuel rod bowing, axial gamma

distribution, cladding defects, and surface deposits. Inspections after two

cycles in Unit 1 and Unit 2 have revealed no anomalies. Unit 2 assemblies have

completed three full cycles with an estimated burnup of 28,000 MWd/t.

The other four reactors included in the-surveillance program will each have an

initial core loading of 17 x 17 fuel assemblies (Trojan, Beaver Valley Unit 1,

Farley Unit 1, and Salem Unit 1). Each core will include a removable rod

assembly except for Beaver Valley. Only two of the four, however, will be

examined as part of the 17 x 17 fuel assembly surveillance program, and these

will be selected on the basis of the first two to actually reload fuel. The

surveillance program includes visual examination (100 percent TV scanning) of

the initially loaded (first-core) fuel assemblies to be removed during the first

three refueling outages. If any anomalies are detected, further examination

will be performed using the removable fuel rod assemblies.

The first visual examination has been completed at Trojan Unit 1, and the

results show the fuel assemblies with burnups up to 17,000 MWd/t to be in

excellent condition. The staff will continue to monitor this fuel surveillance

program, and, if a need is indicated by the results, the staff will require

modifications to the Watts Bar facility.

Watts Bar SER 4-8
4.2.5 Fuel Design Conclusion

On the basis of the Watts Bar fuel safety analysis, confirmatory data from both

in reactor and out-of-reactor tests, and satisfactory experience with this fuel

type in other operating reactors, the staff concludes that the fuel for Watts

Bar will perform its function adequately and that all applicable requirements

have been met. The staff will confirm that the applicant's actions in the areas

of (1) thermal performance analysis, (2) cladding collapse, and (3) fuel rod

bowing are completed and will report its findings in a supplement to this report.

The staff's review has been based on the applicable requirements related to

the reactor fuel as described in SRP (NUREG-75/087) Section 4.2, "Fuel System

Design;" 10 CFR 50.46; 10 CFR 50, Appendix A (GDC 10); 10 CFR 50, Appendix K;

and Regulatory Guides 1.4, 1.25 ("Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Release of Radioactivity in Liquid and Gaseous

Effluents from Light Water-Cooled Nuclear Power Plants"), 1.77 ("Assumptions

Used for Evaluating a Control Rod Ejection Accident for Pressurized Water

Reactors"), and 1.126 ("An Acceptable Model and Related Statistical Methods

for the Analysis of Fuel Densification").

4.3 Nuclear Design

The Watts Bar nuclear plant has a reactor core consisting of 193 fuel assemblies-of the Westinghouse 17 x 17 design. The core has a design heat output of

3411 MWt and is essentially identical to the W. B. McGuire reactor.

4.3.1 Design Bases

The staff has reviewed the design bases and functional requirements used in

the nuclear design of the fuel and reactivity control systems of the Watts Bar

reactor. The basic requirement for the core and control system is that the

consequences of each event be appropriate to the category for that event. To

meet this requirement, several specific design bases are presented in the FSAR.

These include

(1) specification of acceptable fuel design limits

(2) specification of a negative prompt feedback coefficient

(3) requirement that power oscillations be inherently damped or that the

control system be capable of detecting and suppressing them

(4) requiring a control and monitoring system which automatically initiates'

a rapid negative reactivity insertion to prevent exceeding fuel design

limits in normal operation and anticipated transients

(5) requirement that the design of the control system be such that no single

malfunction or operator error lead to a violation of fuel design limits

(6) requirement that shutdown be assured even when the single rod cluster

control assembly (control rod) of highest worth is assumed to be stuck

out of the core

Watts Bar SER 4-9
(7) requiring a chemical shim system capable of controlling power changes in

normal operation and of bringing the reactor to cold shutdown

(8) requirement that the control system, when combined with the engineered

safety features, be capable of controlling reactivity changes during

accident conditions

(9) requirement that reactivity insertion rates and amounts be controlled so

that only limited local yielding occurs in the pressure boundary and the

core remains in a coolable geometry following a reactivity insertion

accident

Based on its review, the staff concludes that the nuclear design bases presented

in the FSAR are in conformance with GDC 10, 11, 12, 13, 20, 25, 26, 27, and 28,

and are, therefore, acceptable.

4.3.2 Design Description

The FSAR contains the description of the first cycle fuel loading, which consists

of three different enrichments and has a first cycle length of approximately

I year. The enrichment distribution, burnable poison distribution, soluble

poison concentration, and higher isotope (actinide) content as a function of

core exposure are presented. Values presented for the delayed neutron fraction

and prompt neutron lifetime at beginning and end of cycle are consistent with

those normally used and are acceptable.

4.3.2.1 Power Distribution

The design bases affecting power distribution are

(1) the peaking factor in the core will not be greater than 2.31 during normal

operation at full power in order to meet the initial conditions assumed in

the LOCA analysis

(2) under normal conditions (including maximum overpower) the peak fuel power will not produce fuel centerline melting

(3) the core will not operate during normal operation or anticipated operational occurrences, with a power distribution that will cause the departure from nucleate boiling ratio to fall below 1.3 (W-3 correlation with

modified spacer effect).

The applicant has described how the core will be operated and power distributions monitored to ensure that these limits are met. The core will be operated

in the constant axial offset control mode, which has been shown to result in

peaking factors less than 2.32 for both constant power and load following

operation.

Two types of instrumentation systems'are provided to monitor core power distribution measurements. Excore detectors are used to monitor core power, axial

offset, and aximuthal tilt, and moveable incore detectors permit detailed power

distributions to be measured. These systems are used in operating reactors

supplied by Westinghouse, and the staff finds their use acceptable for Watts

Bar.

Watts Bar SER 4-10
4.3.2.2 Reactivity Coefficients

The reactivity coefficients are expression of the effect on core reactivity of

changes in such core conditions as power, fuel and moderator temperature, moderator density, and boron concentration. These coefficients vary with fuel

burnup and power level. The applicant has presented values of the coefficients

in the FSAR and has evaluated the uncertainties of these values. The staff has

reviewed the calculated values of reactivity coefficients and has concluded

that they adequately represent the'full range of expected values. The staff

has reviewed the reactivity coefficients used in the transient and accident

analyses and conclude that they conservatively bound the expected values,

including uncertainties. Further, moderator and power Doppler coefficients

along with boron worth are measured as part of the startup physics testing to

ensure that actual values are within those used in these analyses.

4.3.2.3 Control

To allow for changes in reactivity as a result of reactor heatup, load following, and fuel burnup with consequent fission product buildup, a significant

amount of excess reactivity is built into the core. This excess reactivity

is controlled by a combination of full length contol rods and soluble boron.

Soluble boron is used to control reactivity changes as a result of

(1) moderator density and temperature changes from ambient to operating

temperatures

(2) equilibrium xenon and samarium buildup

(3) fuel depletion and fission product buildup (that portion not controlled by

lumped burnable poison)

(4) transient xenon resulting from load following

Control rods are used to control reactivity change as a result of

(1) moderator reactivity changes from hot zero to full power

(2) fuel temperature changes (Doppler reactivity changes)

Burnable poison rods placed in some fuel assembliesare used for radial flux

shaping and to control part of the reactivity change because of fuel depletion

and fission product buildup.

The applicant has provided data to show that adequate control exists to satisfy the above requirements. Moreover, the data show enough additional control rod

worth to provide a hot shutdown effective multiplication factor less than the

design basis value of 0.984 during initial and equilibrium fuel cycles with the

most reactive control rod stuck out of the core. In addition, the chemical and

volume control system will be capable of shutting down the reactor by adding

soluble boron and maintaining it shut down in the cold, xenon-free condition at

any time in core life. These two systems satisfy the requirements of GDC 26.

Watts Bar SER 4-11
Comparisons have been made between calculated and measured control rod bank

worth in operating reactors and in critical experiments. These comparisons

lead to the conclusion that bank worth may be calculated to within approximately 10 percent. In addition bank worth measurements are performed as part

of the startup test program to ensure that conservative values have been used

in safety analyses.

Based on these comparisons, the staff concludes that the applicant has made

suitably conservative assessments of reactivity control requirements and that

adequate control'rod worths have been provided to ensure shutdown capability.

Provision is made in the design for the use of part-length control rods. However, the applicant has stated that the use of part-length rods is not presently

contemplated. All analyses have been performed without part-length rods and

therefore the use of part-length rods will be prohibited.

4.3.2.4 Control Rod Patterns and Reactivity Worths

The full-length control rods are divided into two categories--shutdown rods and

regulating rods. The shutdown rods are always completely out of the core when

the reactor is at operating conditions. Core power changes are made with regulating rods which are nearly out of the core when it is operating at full power.

Regulating rod insertion will be controlled by power-dependent insertion limits

required in the Technical Specifications to ensure that

(1) there is sufficient negative reactivity available to permit rapid shutdown

of the reactor with adequate margin

(2) the worth of a control rod that might be ejected is not greater than that

which has been shown to have acceptable consequences in the safety analyses

The staff has reviewed the calculated rod worths and the uncertainties in these

worths, and concludes that rapid shutdown capability exists at all times in core

life, assuming the most reactive control rod assembly is stuck out of the core.

4.3.2.5 Stability

The stability of the Watts Bar core to xenon induced spatial oscillations is

discussed in the FSAR. The overall negative reactivity (power) coefficient

provides assurance that the reactor will be stable against total power oscillation. The applicant also concluded that sustained radial or aximuthal xenon

oscillations are not possible. This conclusion is based on measurements on an

operating reactor of the same dimensions that showed stability against these

oscillations. The staff concurs with this conclusion.

This core is predicted to be unstable with respect to axial xenon oscillations

after about 12,000 MW days per ton of exposure. The applicant has acceptably

shown that axial xenon oscillations may be controlled by the regulating rods to

prevent reaching any fuel damage limits.

4.3.2.6 Criticality of Fuel Assemblies

Criticality of fuel assemblies outside the reactor is precluded by adequate

design of fuel transfer and storage facilities. The applicant presents

Watts Bar SER 4-12
information on calculational techniques and assumptions used to ensure that

criticality is avoided. The staff has reviewed this information against the

criteria which will be employed and finds them acceptable.

4.3.2.7 Vessel Irradiation

Values are presented for the neutron flux in various energy ranges at mid-height

of the pressure vessel inner boundary. Core flux shapes calculated by standard

design methods are input to a transport theory calculation (Sn) which results

in a neutron flux of 2.8.x 1010 neutrons per cm2 per second having energy greater

than 106 electron-volts at the inner vessel boundary. This results in a fluence

of 2.8 x 1019 neutrons per cm2 for a 40-year vessel life with an 80-percent use

factor. The methods used for these calculations are state of the art, and the

staff concludes that acceptable analytical procedures have been used to calculate the vessel fluence. The staff also will review the requirements for surveillance programs and the pressure-temperature limits for operation.

4.3.3 Analytical Methods

The applicant has described the computer programs and calculational techniques

used to obtain the nuclear characteristics of the reactor design. The calculations consist of three distinct types, which are performed in sequence: determination of effective fuel temperatures, generation of macroscopic few-group

parameters, and space-dependent few-group diffusion calculations. The programs

used (LASER, TWINKLE, LEOPARD, TURTLE, andPANDA) have been applied as part of

the applications for most earlier Westinghouse-designed nuclear plant facilities, and the predicted results have been compared with measured characteristics

obtained during many startup tests for first cycle and reload cores. These

results have validated the ability of these methods to predict experimental

results. The staff, therefore, concludes that these methods are acceptable for

use in calculating the nuclear characteristics of the Watts Bar core.

4.3.4 Summary of Evaluation Findings

The applicant has described the computer programs and calculational techniques

used to predict the nuclear characteristics of the reactor design and has provided examples to demonstrate the ability of the analyses to predict reactivity

and physics characteristics of the Watts Bar plant.

To allow for changes of reactivity as a result of reactor heatup, changes in

operating conditions, fuel burnup, and fission product buildup, a significant

amount of excess reactivity is designed into the core. The applicant has

provided substantial information relating to core reactivity balances for the

first cycle and has shown that means have been incorporated into the design to

control excess reactivity at all times. The applicant has shown that sufficient control rod worth is available to make the reactor subcritical with an

effective multiplication factor no greater than 0.984 in the hot condition at

any time during the cycle with the most reactive control rod stuck in 'the fully

withdrawn position. On the basis of its review, the staff concludes that the

applicant's assessment of reactivity control requirements over the first core

cycle is suitably conservative and that adequate negative worth has been provided

Watts Bar SER 4-13
by the control system to ensure shutdown capability. Reactivity control requirements will be reviewed for additional cycles as this information becomes available, The staff concludes that nuclear design bases, features, and limits have

been established in conformance with the requirements of GDC 10, 11, 12, 13, 20,

25,-26, 27, and 28.

4.4 Thermal Hydraulic Design

4.4.1 Performance and Safety Criteria

The performance and safety criteria for the Watts Bar facility, as stated in

FSAR Section 4.4.1 are

(1) Fuel damage (defined as penetration of the fission product barrier, i.e.,

the fuel rod clad) is not expected during normal operation and operational

transients (Condition I) or any transient conditions arising from faults

of moderate frequency (Condition II). It is not possible, however, to

preclude a very small number of rod failures. These will be within the

capability of the plant cleanup system and are consistent with the plant

design bases.

(2) The reactor core can be brought to a safe state following a Condition III

event with only a small fraction of fuel rods damaged (see above definition) although sufficient fuel damage might occur to preclude immediate

resumption of operation.

(3) The reactor core can be brought to a safe state and the core can be kept

subcritical with acceptable heat transfer geometry following transients

arising from Condition IV events.

4.4.2 Design Bases

The performance and safety criteria listed above are implemented through the

following design bases.

4.4.2.1 Departure from Nucleate Boiling

The margin to departure fromnucleate boiling at any point in the core is

expressed in terms of the departure.from nucleate boiling ratio (DNBR). The

DNBR is defined as the ratio of the heat flux required to produce departure

from nucleate boiling at the calculated local coolant conditions to the actual

local heat flux. The thermal-hydraulic design basis, as stated in FSAR Section 4.4.1.1, for the prevention of DNB is

Departure from nucleate boiling (DNB) will not occur on at least

95 percent of the limiting fuel rods during normal operation and

operational transients and any transient conditions arising from

faults of moderate frequency (Condition I and II) events at a 95

percent confidence level.

Watts Bar SER 4-14
4.4.2.2 Fuel Temperature

The fuel temperature design basis given in FSAR Section 4.4.1.2 is

During modes of operation associated with Condition I and Condition II events, the maximum fuel temperature shall be less than

the meltng temperature of U02 . The U02 melting temperature for

at least 95 percent of the peak kW/ft fuel rods will not be

exceeded at the 95 percent confidence level.

This design basis is evaluated in Section 4.2, of this report.

4.4.2.3 Core Flow

Section 4.4.1.3 of the FSAR states the core flow design basis as, "A minimum

of 92.5 percent of the thermal flow rate will pass through the fuel rod region

and be effective for fuel rod cooling."

4.4.2.4 Hydrodynamic Stability

The hydrodynamic stability design basis in FSAR Section 4.4.1.4 is, "Modes of

operation associated with Condition I and II events shall not lead to hydrodynamic instability."

4.4.3 Thermal-Hydraulic Design Methodology

4.4.3.1 Departure from Nucleate Boiling

The thermal-hydraulic design analyses were performed using the W-3 critical

heat flux (CHF) correlation in conjunction with a THINC-IV analysis. THINC-IV

is an open channel computer code which determines the coolant density, mass

velocity, enthalpy, vapor void, static pressure, and DNBR distribution along

parallel flow channels within a reactor core.

The W-3 correlation was developed from data obtained from experiments conducted

with fluid flowing inside single heated tubes. As test procedures progressed

to the use of rod bundles instead of tubes, the correlation was modified to

include the effects of "R" mixing vane grids and axially nonuniform power

distributions.

The applicant has proposed a minimum departure from nucleate boiling of 1.30

to ensure that there is a 95-percent probability at a 95-percent confidence

level that critical heat flux will not occur on the limiting fuel rod. The

use of the W-3 CHF correlation with a minimum DNBR of 1.30 was previously

approved by the staff.

A description of the THINC-IV computer code is given in Westinghouse Topical

Report WCAP-7956, "THINC-IV An Improved Program For Thermal-Hydraulic Analysis

of Rod Bundle Cores." The design application of the THINC-IV program is given

in detail in WCAP-8054, "Application of the THINC-IV Program to PWR Design."

Both WCAP-7956 and WCAP-8054 have been reviewed and approved by the staff..

Watts Bar SER 4-15
The design calculational procedure, using THINC-IV, is to perform a core-wide

analysis followed by a hot assembly and hot subchannel analysis. A reload

review of a pressurized water reactor, not of Westinghouse design, showed that

the input parameters Used in the thermal-hydraulic analysis of the initial core

did not bound future cycles (that is, beyond the first cycle). The staff asked

the applicant whether his methods appropriately bound future cycles. The applicant responded that the safety analysis is intended to be valid for all plant

cycles and the input parameter values used in the reference safety analysis were

selected to bound the values expected in all subsequent cycles. He further

stated that when all .of the parameters for a given accident are bounded, the

reference safety analysis is valid; however, if a parameter is not bounded,

further evaluation is necessary. This further evaluation is to confirm that

the margin of safety, as defined in the basis for any Technical Specifications,

is not reduced.

Based on the information given above, the staff concludes that the applicant

has adequately considered future cycle operation in the design safety analyses.

During the course of its review, the staff requested that the applicant provide

the radial pressure gradient in the upper plenum and at the core outlet and

that he discuss and support, by calculations, the differences in hot channel

pressure drop, flow, enthalpy rise, and minimum DNBR relative to the assumption

of a uniform pressure at the core boundaries.

The staff has reviewed a November 2, 1977 letter from C. Eicheldinger (Westinghouse) to J. Stolz (NRC) which described THINC-IV analyses using a cosine upper

plenum radial pressure gradient with a maximum value of 5 psi at the core center

and 0 psi at the periphery. The results of these analyses showed that the

effects of a core pressure distribution on the minimum DNBR is negligible. The

staff conducted a similar sensitivity study using COBRA-IV. The staff results

also showed that the effects are small. Based on these analyses, the staff

concludes that.the use of a uniform exit pressure gradient in the Watts Bar

thermal-hydraulic design is acceptable.

Based on its findings that the critical heat flux correlation and the thermalhydraulic computer code used by the applicant have been previously approved by

the staff, the applicant has appropriately bounded future cycles inhis safety

analyses, and the use of a uniform core exit pressure gradient has been adequately justified. The staff concludes that the DNB design methodology used

in the design of Watts Bar is acceptable.

4.4.3.2 Core Flow

The core flow design basis requires that the minimum flow which will pass through

the fuel rod region and be effective for fuel rod cooling is 92.5 percent of the

primary coolant flow rate. The remainder of the flow, called bypass flow, will

be ineffective for cooling because it will take the following bypass paths:

(1) flow through the spray nozzles into-the upper head

(2) flow into the rod cluster control rod guide thimbles

(3) leakage from the vessel inlet nozzle directly to the vessel outlet nozzle

(4) flow between the baffle and barrel

(5) flow in the gaps between the fuel assemblies

Watts Bar SER 4-16
The amount of bypass flow is determined by a series of hydraulic resistance

calculations on the core and vessel internals and verified by model flow tests.

Because the amount of bypass flow is consistent with approved plants of similar

design, the staff concludes that the core bypass flow used in the design

analysis, 7.5 percent, is acceptable.

4.4.3.3 Hydrodynamic Instability

For steady-state, two-phase heated flow in parallel channels, the potential for

hydrodynamic instability exists. The applicant stated that the core was stable

because Westinghouse reactors will not experience any Ledinegg instability over

Condition I and II operational ranges. In addition, open channel configurations

(which are a feature of Westinghouse PWRs) are more stable than closed channel

configurations. This was shown by flow stability tests that were conducted at

pressures up to 2200 psia. The results showed that for flow and power levels

typical of reactor conditions, no flow oscillations could be induced.

Also, a method developed by Ishii for evaluating density wave stability in

parallel closed channel systems was used to assess the stability of typical

Westinghouse reactor designs. The results indicate that a large margin to

density wave instability exists. Finally, data from numerous rod bundle tests,

which were performed over wide ranges of operational conditions, show no

evidence of premature DNB or of inconsistent data that might be indicative of

flow instabilities in the rod bundles.

The staff is conducting a generic study of the hydrodynamic stability of light

water reactors. Limitations to the thermal-hydraulic design resulting from the

staff study will be compensated for by appropriate operating restrictions; however, none are anticipated.

In the interim, the staff concludes that past operating experience, flow stability experience, and the thermal-hydraulic design of Westinghouse PWRs serve as

a basis for issuance of an operating license.

4.4.4 Operating Abnormalities

4.4.4.1 Fuel Rod Bowing

A significant parameter that affects the thermal-hydraulic design of the core

is rod-to-rod bowing within fuel assemblies. The Westinghouse methods for

predicting the effects of rod bow on DNBR are under review by the staff. Therefore, the magnitude of rod bow as a function of burnup was evaluated based on

interim methods which have been previously accepted. The resultant reduction

in the departure from nucleate boiling ratio due to rod bow is given in Table 4.1.

Before the Technical Specifications are issued the staff will ensure that the

thermal margin reductions given above have been accommodated using an acceptable

method.

For plants designed by Westinghouse,.the staff has approved the following generic

margins in Table 4.2 (see "Interime Safety Evaluation Report on the Effects of

Fuel Rod Bowing on Thermal Margin Calculations for Light Water Reactors,"

December 1976), which may be used to offset the reduction in DNBR as a result

of rod bowing.

Watts Bar SER 4-17
Table 4.1 Rod bow penalties

Burnup DNBR penalty

(MWD/MTU) (%)

0 0

3,500 0

5,000 0

10,000 2.15

15,000 4.64

20,000 6.74

25,000 8.59

30,000 10.27

35,000 13.07

40,000 19.09

Plant-specific margins which could be available are:

(1) The Technical Specifications minimum flow rate is greater than the design flow rate.

(2) The Technical Specifications maximum Tavg is less than the design Tavg.

(3) The trip set points are more limiting than the thermal-hydraulic analysis indicates.

The basis of the Technical Specifications will include any plant-specific or generic margin that may be used to offset the reduction in DNBR as a result of

rod bowing.

4.4.4.2 Crud Deposition

Crud deposition in the core and an associated change in core pressure drop and flow have been observed in some PWRs not of Westinghouse design. In response to a staff question, the applicant stated that (1) operating experience on Westinghouse reactors indicate that a flow resistance allowance for crud deposition is not required; (2) the effects of crud enters into the calcula- tions by use of a surface roughness factor three times greater than those obtained from operating Westinghouse PWRs; and (3) reduced flow would be observed by four different means: (1) flow meters, (2) a reduction in reactor

power if in an automatic control mode or an increase of AT across the core if in a manual mode, (3) core exit thermocouple readings, and (4) incore flux

maps.

Based on the information given above, the staff concludes that the applicant adequately addressed its concerns relative to uniform or preferential crud deposition in the core. Appropriate surveillance requirements to recognize rapid crud buildup will be included in the Technical Specifications.

Watts Bar SER 4-18
Table 4.2 Generic margins

Margin % reduction

Use of a design minimum DNBR of 1.30 instead of the

95/95 DNBR limit of 1.28 1.6

Reduction in fuel rod pitch for the hot channel analysis 1.7

Use of a thermal diffusion coefficient (TDC of 0.038

instead of a TDC of 0.051 1.2

Addition of an extra grid in the design of the Westinghouse 17 x 17 fuel assembly relative to the 15 x 15

fuel design 2.9

Use of a 0.86 multiplier on the modified spacer factor

(Fs) of the W-3 correleation instead of a 0.88 multiplier 1.7

Maximum generic margin which may be claimed 9.1

4.4.5 Loose Parts Monitoring System

The applicant has provided a description of the loose parts monitoring system

(LPMS) which will be used at Watts Bar. The design will include two sensors at

each natural collection region on the exterior surface of the reactor coolant

boundary. The system will be designed to remain functional for a seismic event

(up to and including the operational basis earthquake (OBE)) that does not

require plant shutdown. The system will be capable of detecting a metallic

loose part impacting within 3 ft of a sensor and having a kinetic energy of

0.5 ft-lb. That portion of the system which is located inside containment will

be tested to ensure compatability with the operating environment. Alert levels

will be set to detect a loose part having an impact of 0.5 ft-lb during plant

shutdown. Because operating conditions will cause variation in the amount of

background noise, the sensitivity will be the same percentage of the background

noise.

The staff has evaluated the Watts Bar LPMS by comparing it with the equipment

and procedures used on other comparable plants, taking into account pertinent

differences, and the requirements of Regulatory Guide 1.133 ("Loose-Part Detection Program for the Primary System of Light-Water-Cooled Reactors"). The staff

requested the applicant to provide a commitment to supply a report on the Watts

Bar LPMS which contains (1) an evaluation of the LPMS for conformance to Regulatory Guide 1.133, (2) a description of the operator training program, and (3) a

description of the system hardware operation and implementation of the loose

parts detection program after startup testing.

On February 25, 1982, the applicant submitted his response containing information regarding items (1) and (2) above.

Based on its review of the applicant's LPMS submittals and pending staff review

and approval of the final LPMS startup report, the staff concludes that an

Watts Bar SER 4-19
acceptable loose parts monitoring program will be implemented at the Watts Bar

facility. The staff will ensure that the limiting conditions for operation and surveillance requirements for the LPMS are included in the Technical

Specifications.

4.4.6 Thermal-Hydraulic Comparison

Table 4.3 lists the thermal-hydraulic design parameters for the Watts Bar facility and compares them to values for the Trojan and Sequoyah facilities.

The Watts Bar units were designed to operate at the same thermal power as the

Sequoyah and Trojan plants. The W-3 CHF correlation and THINC-IV computer

program were used in the design of all of the plants. The major differences

between the Watts Bar and Trojan plants are a higher nominal inlet temperature,

a lower flow rate, and a lower maximum heat flux. The higher nominal inlet

temperature for the Watts Bar units results in a decrease in the thermal margin.

However, the higher flow rate and the lower maximum heat flux compensate and

result in the minimum DNBR, at nominal conditions, increasing from 1.71 for

Trojan to 1.72 for Watts Bar. Therefore, the net change for Watts Bar is a

slightly greater thermal margin than for Trojan.

The differences in the Watts Bar and Sequoyah designs are a higher flow rate and

inlet temperature for Watts Bar. These increases result in a lower minimum DNBR

for Watts Bar, 1.72 for Watts Bar as compared to 1.81 for Sequoyah. Hence, the

thermal margin available to Watts Bar is less than that available to Sequoyah.

Both the Sequoyah and Trojan facilities have been previously reviewed and

approved by the staff. The comparability of the Watts Bar plant to the

Sequoyah and Trojan facilities supports the conclusion that the Watts Bar

thermal-hydraulic design is acceptable.

4.4.7 N-1 Loop Operation

N-1 loop operation is when one of the reactor's coolant loops is out of service.

Thus, only three coolant loops are available to supply coolant to the reactor

core...

In response to a staff question, the applicant stated that he did not wish to

exercise the option to operate in the N-1 mode. The staff will require that

the Technical Specifications include appropriate provisions to ensure that this

type of operation is prohibited.

4.4.8 Instrumentation for Inadequate Core-Cooling Detection (II.F.2)

In response to Item II.F.2 of NUREG-0737, the applicant has transmitted information on instrumentation for detection of inadequate core cooling (ICC) in

September 14, 1981, October 29, 1981, and March 11, 1982 letters.

Analysis and procedures for the detection of ICC using existing'instrumentation

have been developed in conjunction with the Westinghouse Owners' Group. The

applicant has committed to incorporate this guidance into plant procedures before

fuel load. The existing ICC system should be evaluated to the requirements of

Watts Bar SER 4-20
Table 4.3 Reactor design comparison

Watts Bar Trojan Sequoyah

Parameter Units 1&2 SER, SER

Performance characteristics

Reactor core heat output, (MWt) 3411 3411 3411

System pressure, psia 2250 2250 2250

Minimum DNBR 1.30 1.30 1.30

Typical cell 2.05 2.04 2.22

Thimble cell, 1.72 1.71 1.81

Critical heat flux correlation W-3 W-3 W-3

Coolant flow

Total flow rate (106 lb/hr) 144.8 132.7 133.8

Effective flow rate for heat

transfer (106 lb/hr) 133.0 126.7 127.8

Average velocity along fuel

rods, (ft/s) 16.6 15.7 15.6

Effective core flow area (ft 2 ) 51.1 51.1 51.1

Coolant temperature (OF)

Nominal reactor inlet 559.1 552.7 545.7

Average rise in core 62.7 66.9 67.8

Pressure drop across tore (psi) 25.8±3.9 - 24.3±2.4

Heat transfer, 100 % power

Active heat transfer surface

area, (ft 2 ) 59,700 59,700 59,700

Average heat flux,

(BTU/hr-ft 2 ) 189,800 189,800 189,800

Maximum heat flux,

(BTU/hr-ft 2 ) 440,300 574,500 474,500

Average linear heat rate

(kW/ft) 5.44 5.44 5.44

Maximum thermal output (kW/ft) 12.6 13.6 12.2

Appendix B. Evaluations are being performed to determine to what extent modifications should be made to upgrade this system. This evaluation, along with-an

implementation schedule, will be available by fuel loading on Watts Bar Unit 1.

In addition to the primary method for detecting ICC described above, the applicant will provide instrumentation to measure water level in the reactor vessel

down to the hot-leg piping and between the hot leg and bottom of the reactor

vessel. This instrumentation will be designed and qualified in accordance with

safety grade, Class 1E, requirements including redundancy and emergency power.

Watts Bar SER 4-21
Subcooling Monitor

This instrumentation is part of the existing ICC instrumentation. TVA has not

provided the functional design description or the applicant's evaluation for

conformance of this instrumentation with Item II.F.2 of NUREG-0737. During a

June 2, 1982 telecon, TVA agreed to submit this information. The staff will

review this information when it is received and will report its findings in a

supplement to the SER.

Incore Thermocouple System

The Watts Bar Nuclear Plant incore thermocouples are located at the core exit

for each quadrant and, in conjunction with core inlet RTD data, are sufficient

to provide an indication of the radial distribution of the coolant enthalpy

rise across representative sections of the core. Sixteen (four per quadrant)

of the core-exit thermocouples will be designated as postaccident monitoring

(PAM) sensors.

The primary operator display is a computer-driven printer. A spatially oriented core map is available on demand; it indicates the temperature at each core

exit thermocouple location. The printout range is 200OF to 2300'F.

Alarm capability is provided in conjunction with the subcooling monitor that

uses the average of all the thermocouple (T/C) readings in the calculations.

A backup analog readout is provided with the capability of selective reading

of any T/C in the system. The range of the system is 0 to 700*F. However, TVA

has agreed to.extend the temperature range of the backup analog readout to

2300OF per the requirements of Item II.F.2 of NUREG-0737.

Another means of obtaining this data is reading the raw signals (T/C and reference junction output) with portable test equipment. These data are available

in the control building and would be accessible under all conditions should

the primary and backup display devices fail.

Reactor Vessel Level Instrumentation

The reactor vessel level instrumentation (RVLI) system was designed to provide

direct readings of vessel level that can be used by the operator. This RVLI

system does not replace existing systems and is not coupled to safety systems,

but acts only to provide additional information to the operator.

Redundant displays are provided for the two sets. Level information based on

all three differential pressure measurements is presented. Correction for

reference leg densities is automatic. Any error conditions such as out-ofrange sensors or hydraulic isolators are automatically displayed on the affected

measurements.

The RVLI is to be used in conjunction with a coolant subcooling readout to

determine the state and transient behavior of the reactor coolant system. The

reactor vessel wide-range level indication will read on scale with all four

Watts Bar SER 4-22
reactor coolant pumps running during normal operation from 0-to-lOO-percent full

power. With all pumps shut down, the indicator will provide a direct indication

of water level in the reactor vessel.

Staff Evaluation

The staff has reviewed the applicant's September 14 and October 29, 1981 and

March 11, 1982 submittals regarding Item II.F.2 of NUREG-0737 and finds that

(1) The applicant's response is incomplete with respect to the itemized documentation requirements of NUREG-0737 Item II.F.2. All nine items must be

addressed.

(2) The described incore thermocouple system, including the primary and backup

operator display, is not acceptable with respect to NUREG-0737 Item II.F.2

Attachment 1 and Appendix B requirements; for example, the applicant should

evaluate the ICC system for conformance to Appendix B.

(3) the description of the RVLI is not acceptable. A complete description of

the instrumentation, the vendor, testing programs, analysis, and installation schedule is required.

Therefore, the staff will require the applicant to provide this information on

a schedule that will permit completion of its review before fuel load. The

staff will report its findings in a supplement to this report.

4.4.9 Summary and Conclusion

The thermal-hydraulic design of the core for the Watts Bar nuclear plant has

been reviewed. The acceptance criteria used as the basis for evaluation are

set forth in SRP (NUREG-0800) Section 4.4, Section II, "Thermal and Hydraulic

Design Criteria." The scope of the review included the design criteria, the

core design, and the steady-state analysis of the core thermal-hydraulic

performance. The review concentrated on the difference between the proposed

core design and those designs that have been previously reviewed and found

acceptable by the staff. The staff concluded that all such differences were

acceptable. The applicant's thermal-hydraulic design analyses were performed

using analytical methods and correlations which have been previously reviewed

and approved by the staff.

The staff concludes that the core has been designed with appropriate margin to

ensure that acceptable fuel design limits are not exceeded during steady-state

operation and anticipated operational occurences. The thermal-hydraulic design

of the core, therefore, meets the requirements of GDC 10 and is acceptable for

preliminary design approval. This conclusion is based on the applicant's

analyses of the core thermal-hydraulic performance which were reviewed by the

staff and found to be acceptable. The applicant has committed to a preoperational and initial startup test program in accordance with Regulatory Guide 1.68

("Initial Test Programs for Water-Cooled Nuclear Power Plants") to measure and

confirm the thermal-hydraulic design aspects. The staff has reviewed the applicant's preoperational and initial startup test program and has concluded that

it is acceptable. However, before final design approval and issuance of an

operating license, the staff will require the applicant to supply the documentation required by Item II.F.2 of NUREG-0737. In addition, the staff must complete

Watts Bar SER 4-23
its review of the report and the results of the startup tests on the loose parts

monitoring system. These issues will be addressed in a supplement to this SER.

4.5 Reactor Materials

4.5.1 Control Rod Drive Structural Materials

The staff concludes that the materials used for the construction of the control

rod drive structure are acceptable and meet, in part, the requirements of GDC 1,

14 and 26 and 10 CFR 50.55a.

The applicant has met these requirements by assuring that the design, fabrication, and testing of the materials used in the control rod drive structural are

of high quality and adequate for structural integrity.

The mechanical properties of structural materials selected by the applicant for

the Watts Bar control rod system components that are exposed to the reactor

coolant satisfy the stringent procurement specification requirements applied by

the designer and conform with the staff's position in SRP Section 4.5.1 ' that

the yield strength of cold-worked austenitic stainless steel should not exceed

90,000 psi.

The controls imposed upon the austenitic stainless steel of the mechanisms

conform to the recommendations or to acceptable alternatives of Regulatory

Guides 1.31, "Control of Ferrite Content in Stainless Steel Weld Metal," and

1.44, "Control of the Use of Sensitized Stainless Steel." Fabrication and

heat treatment practices performed in accordance with these recommendations

provide added assurance that stress corrosion cracking will not occur during

the design life of the component. The austenitic stainless steel materials

are furnished in solution heat treated condition. Sensitization is avoided

by not permitting heat treatment in the temperature range of 800 to 15009F.

Fabrication and heat treatment practices performed as stated above provided

added assurance that stress corrosion cracking will not occur during the

design life of the components. The compatibility of all materials used in

the control rod systems in contact with the reactor coolant satisfies the

criteria of Articles NB-2160 and NB-3120 of Section III of the ASME Boiler

and Pressure Vessel Code. Martensitic stainless steel has been given tempering treatments in accordance with the positions in SRP Section 4.5.1.

Cleaning and cleanliness control have been performed to provide adequate contamination control of components during fabrication, shipment, and storage.

Conformance with the Code and Regulatory Guides cited above, conformance with

staff positions on the allowable maximum yield strength of cold-worked austenitic stainless steel, and generally the tempering or aging temperature of

martensitic and precipitation-hardened stainless steel constitute an acceptable

basis for meeting the requirements of GDC 1, 14, and 26 and 10 CFR 50.55a.

4.5.2 Reactor Internals and Core Support Materials

The staff concludes that the materials used for the construction of the'reactor

internals and core support are acceptable and meet, in part, the requirements

of GDC 1 and 10 CFR 50.55a.

Watts Bar SER 4-24
The applicant has met the requirements of GDC 1 and 10 CFR 50.55a by ensuring

that the design, fabrication, and testing of the materials used i^n the reactor

internals and core support structure are of high quality and adequate for

structural integrity. The controls imposed upon components constructed of

austenitic stainless steel satisfy the intent of the recommendations of Regulatory Guides 1.31 and 1.44.

The materials used for the construction of components of the reactor internals

and core support structure have been identified by specification and found to

be in conformance with the requirements of NG-2000 of Section III and Parts A,

B, and C of Section II (or equivalent specification) of the ASME Code. As

proven by extensive tests and satisfactory performance, the specified materials

are compatible with the expected environment, and corrosion is expected to be

negligible.

The controls imposed on the reactor coolant chemistry provide reasonable assurance that the reactor internals and core support structure will be adequately

protected during operation from-conditions which could lead to stress corrosion

of the materials and loss of component integrity.

The material selection, fabrication practices, examination and testing procedures, and control practice performed in accordance with these recommendations

provide reasonable assurance that the materials used for the reactor internals

and core support structure will be in a metallurgical condition to preclude

service deterioration. Conformance with requirements of the ASME Code and the

recommendations of the Regulatory Guides constitute an acceptable basis for

meeting the requirements of GDC 1 and 10 CFR 50.55a.

4.6 Functional Design of Reactivity Control Systems

The functional design of the reactivity control systems for the facility have

been reviewed to confirm that they meet the various reactivity control conditions for all modes of operation. These are

(1) the capability to operate in the unrodded, critical, full power mode

throughout plant life

(2) the capability to vary power level from full power to hot shutdown and

ensure control of power distributions within acceptable limits at any

power level

(3) the capability to shutdown the reactor in a manner sufficient to mitigate

the effects of postulated events discussed in Section 15.0 of this report

The control rod drive system (CRDS), the safety injection system (SIS), and the

chemical and volume control system (CVCS) are the reactivity control systems at.

Watts Bar.

The CRDS is composed of control rod drive mechanisms (CRDMs) to which the rod

cluster control assemblies (RCCAs) are attached. The RCCAs are divided into

two categories; control and shutdown.

Watts Bar SER 4-25
The CRDM is a magnetically operated jack. The magnetic jack is an arrangement

of three electromagnets which are energized in a controlled sequence to insert

or withdraw RCCAs in discrete steps.

The control category RCCAs may be automatically inserted or withdrawn to compensate for changes in reactivity associated with power level changes and power

distribution, variations in moderator temperature, or changes in boron concentration. The shutdown category RCCAs, which are fully withdrawn during power

operations, are used solely to insert large amounts of negative reactivity to

shut Jown the reactor. (See Section 4.3 of this report for further discussions

of these features.)

The RCCAs are the primary shutdown mechanism for normal operation, accidents

and transients. They insert automatically upon a reactor trip signal. Concentrated boric acid solution is injected by the SIS in the event of a LOCA,

steamline break, loss of normal feedwater flow, steam generator tube rupture,

or RCCA ejection, thereby complying with GDC 20, which requires that automatic

.protective systems be provided (1) to ensure that specified acceptable fuel

design limits are not exceeded and (2) to sense accident conditions and to

initiate operation of systems and components important to safety.

Failure of electrical power to a RCCA will result in the insertion of that

assembly, as will shearing of the connection between the rod cluster control

assembly and control rod drive mechanism. Single failure of a rod cluster control assembly is considered in transient and accident analyses that include the

most reactive rod cluster control assembly stuck outside the core. Analysis of

accidental withdrawal of a rod cluster control assembly is found to have acceptable results. This conforms to the requirements of GDC 23, "Protection System

Failure Modes," and 25, "Protection System Requirements for Reactivity Control

Malfunctions."

During power operations, the operability of each RCCA is tested at least once

every 31 days. At every refueling shutdown, rod drop tests are performed for

each RCCA to demonstrate the ability of the assemblies to meet required drop

times. The RCCA design is such that a single failure will not result in loss

of the protection system nor will a loss of redundancy occur as a result of

removal of a channel or component from service. This is discussed further in

Section 7.2 of this SER. The foregoing periodic testing, reliability, and

redundancy conform to the requirements of GDC 21.

The SIS is automatically actuated to inject borated water into the reactor

coolant system (RCS) upon receipt of a safety injection actuation signal (SIAS).

The SIS pumps take suction from the refueling water storage tank (RWST). The

SIS is discussed further in Section 6.3 of this SER.

The CVCS is designed to accommodate slow- or long-term reactivity changes such

as that caused by fuel burnup or by variation in the xenon concentration resulting from changes in reactor power level. The CVCS is used to control reactivity

by adjusting the dissolved boron concentration in the reactor coolant system.

The boron concentration is controlled to obtain optimum RCCA positioning; to

compensate for reactivity changes associated with variations in coolant temperature, core burnup, xenon concentration; and to provide shutdown margin for

maintenance and refueling operations or emergencies. A portion of the CVCS

(the charging pumps, the boric acid pump discharge, and the boric acid makeup

Watts Bar SER 4-26
tanks) injects a high concentration boron solution into the reactor coolant

system to help ensure plant shutdown in the event of a SIAS. The boric acid

concentration in the reactor coolant system is controlled by the charging and

letdown portions of the CVCS.

The CVCS can maintain the reactivity of the reactor within required bounds by

means of the automatic makeup system to replace minor leakage without significantly changing the boron concentration in the reactor coolant system. Dilution of the reactor coolant system boron concentration required for the reactivity losses occurring as a result of fuel depletion is accomplished by manual

action. The CVCS is discussed further in Section 9.3.4 of this SER.

The concentration of boron in the reactor coolant system is changed manually

under the following operating conditions:

(1) startup - boron concentration decreased to compensate for moderator

temperature and power increase

(2) load follow - boron concentration increased or decreased to compensate

for xenon transients following load changes

(3) fuel burnup - boron concentration decreased to compensate for burnup

(4) cold shutdown - boron concentration increased to compensate for increased

moderator density as a result of cooldown

Soluble poison concentration is used to control slow operating reactivity

changes. If necessary, RCCA movement can also be used to accommodate such

changes, but assembly insertion is used mainly to control anticipated operational occurrences even with a single malfunction, such as a stuck rod. In

either case, fuel design limits are not exceeded. The soluble poison control

is capable of maintaining the core subcritical under conditions of cold shutdown, which conforms to the requirements of GDC 26.

The reactivity control systems, including the addition of concentrated boric

acid solution by the SIS, are capable of controlling all anticipated operational changes, transients, and accidents, except possibly the small-break LOCA. (For

further information on the performance of the charging and borating portion of

the CVCS with respect to small-break LOCAs, see Sections 6.3 and 15.3.5 of this

SER.) All accidents are calculated with the assumption that the most reactive

RCCA is stuck out and cannot be inserted, which complies with the requirements

of GDC 27.

Compliance with the requirements of GDC 28, is discussed in Sections 4.3 and

15.0 of this report.

Based on its review, the staff concludes that the reactivity control system functional design meets the requirements of GDC 21, 23, 25, 26, and 27 with

respect to reliability and testability, fail safe design, malfunction protection design, redundancy and capability, and combined systems capability and, is,

therefore, acceptable.

Watts Bar SER 4-27

5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS

5.1 Summary Description

Each reactor coolant system (RCS) consists of four similar heat transport loops

connected to the reactor pressure vessel. Each loop contains a reactor coolant

pump, steam generator, and associated piping. In addition, the system includes

a pressurizer, a pressurizer relief tank, interconnecting piping, and instrumentation necessary for operational control. All of these components are located

within the containment building.

During operation, the RCS transfers the heat generated in the core to the steam

generators where steam is produced to drive the turbine generator. Borated

demineralized water is circulated in the RCS at a flow rate and temperature

consistent with achieving the reactor core thermal-hydraulic performance. The

coolant also acts as a neutron moderator and reflector and as a solvent for the

neutron absorbing boric acid used for chemical shim control.

The RCS pressure boundary provides a second barrier against the release of

radioactivity generated within the reactor, and it is designed to ensure a high

degree of integrity throughout the life of the plant.

The RCS pressure changes during normal operation are controlled by the use of

the pressurizer where water and steam are maintianed in equilibrium by electrical

heaters and water spray. Spring-loaded safety valves and poweroperated relief

valves are mounted on the pressurizer and discharged to the pressurizer relief

tank where steam is condensed and cooled by mixing with water.

5.2 Integrity of Reactor Coolant Pressure Boundary

5.2.1 Compliance with Codes and Code Cases

5.2.1.1 Compliance with 10 CFR 50.55a

The components of the Watts Bar reactor coolant pressure boundary (RCPB) as

defined by the rules of 10 CFR 50.55a have been properly classified in FSAR

Table 5.2-1 as ASME Section III, Class 1 components. These components are

designated TVA Class A and are in accordance with the guidance in Regulatory

Guide 1.26. The ASME Section III Code Editions and Addenda used in the construction of these TVA Class A components are, at a minimum, those required

to ensure compliance with 10 CFR 50.55a. In addition to the TVA Class A

components of the RCPB, certain lines that perform a safety function and that

meet the exclusion requirements of footnote 2 of the rule are classified TVA

Class B. These are in accordance with the guidance provided in Regulatory

Position C.1 of Regulatory Guide 1.26 and are constructed as ASME Section III,

Class 2, components. Valve leakage monitoring system lines that do not perform

a safety function and that meet the exclusion requirements of footnote 2 of the

rule are classified TVA Class G on the downstream side of the isolation valve.

Watts Bar SER 5-1
The staff concludes that construction of the RCPB components in conformance

with the appropriate ASME Code editions and addenda and the Commission's

regulations provides assurance that component quality is commensurate with the

importance of the RCPB safety function, constitutes an acceptable basis for

satisfying the requirements of GDC 1, and, is therefore, acceptable.

5.2.1.2 Applicable Code Cases

The applicant has identified specific ASME Code cases in FSAR Section 5.2.1.4

whose requirements have been applied in the construction of pressure-retaining

ASME Section III, Class 1, components within the RCPB (Quality Group A). The

staff has reviewed the Code cases in Section 5.2.1.4. The basis for acceptance

in the staff review has been the Code cases found to be-acceptable in Regulatory

Guides 1.84 ("Code Case Acceptability-ASME Section III, Design and Fabrication")

and 1.85 ("Code Case Acceptability-ASME Section III, Materials") and the Code

cases previously found to be acceptable by the staff for plants similar to Watts

Bar, before publication of the Regulatory Guides. The staff concludes that

compliance with the requirements of these Code cases will result in a component

quality level that is commensurate with the importance of the safety function

of the RCPB, constitutes an acceptable basis for satisfying the requirements of

GDC 1, and is, therefore, acceptable.

5.2.2 Overpressure Protection

Overpressure protection of the reactor coolant systems is provided by three

pressurizer safety valves and two power-operated-relief valves (PROVs), all

discharging through a common header to the pressurizer relief tank. The basis

for acceptance of the overpressure protection system is that it conforms to the

ASME Code, Section III, and GDC 15.

Each of the PORVs has the capacity to relieve 210,000 pounds mass per hour of

saturated steam at 2350 psig. These relief valves are designed to limit

pressurizer pressure to a value below the high pressure reactor trip set point

for a 10 percent step load decrease and to prevent unnecessary safety valve

action. Each PORV can be individually isolated by remotely operated stop

valves in the event of excessive leakage.

The applicant has referenced Westinghouse Topical Report WCAP-7769 to describe

the sizing of the Watts Bar pressurizer safety valves and has provided additional plant-specific analyses to demonstrate the adequacy of the sizing basis.

Each pressurizer safety valve is spring loaded and has a relieving capacity. of

420,000 pounds mass per hour of saturated steam at 2485 psig. The combined

capability of two of these three valves is adequate to prevent the pressurizer

pressure from exceeding the ASME Code, Section III, limit of 110 percent design

pressure following the worst reactor coolant system pressure transient (identified to be a 100 percent load rejection resulting from a turbine trip with

concurrent loss of main feedwater). This event was analyzed with no credit

taken for operation of RCS relief valves, steamline relief valves, steam dump

system, pressurizer level control system, and pressurizer spray.

In the analysis, the applicant has taken credit for a high pressurizer pressure

trip (the first safety grade primary system trip), but has stated that, without

Watts Bar SER 5-2
that credit, an overtemperature AT trip (second safety grade trip) would terminate the transient without significantly more severe results. The evaluation

is supported by a generic sensitivity study of required safety valve flow rate

versus trip parameter presented in WCAP-7769. The staff has requested additional

information on the details of this calculation and will report, its findings in

a supplement to this SER.

The above analyses were performed using the LOFTRAN Code, a digital simulation

that includes point neutron kinetics, RCS (including the reactor vessel), hot

leg, primary side of the steam generator and cold leg, secondary side of the

steam generator, pressurizer, and pressurizer surge line. This code is currently

under review by the staff. The review has progressed to the point that there

is reasonable assurance that the conclusions based on these analyses will not

be appreciably altered by completion of the analytical review. If the final

approval of LOFTRAN indicates that any revisions to the analyses are required,

the effect of these changes on Watts Bar will be evaluated, and the staff will

require implementation, if indicated.

The safety valves are designed in accordance with ASME Code, Section III and

periodic testing and inspection are performed in accordance with Section XI.

In FSAR Chapter 14, the applicant has described the preoperational test program, which includes testing of the pressure relieving device discussed in

this SER section, and has indicated that these tests would be conducted in full

compliance with the intent of Regulatory Guide 1.68. GDC 32 is discussed in

Section 5.2.4. Item II.D.1 of NUREG-0737 requires performance testing of relief

and safety valves (discussed in Section 3.9.3.3) and Item II.D.3 (Indication

of Relief and Safety Valve Position) is discussed in Section 7.8.1. Subject

to the resolution of the above issues by the applicant, the staff concludes

that the overpressure protection provided for Watts Bar at hot operating

conditions will comply with the guidelines of SRP Section 5.2.2 and the requirements of GDC 15.

With regard to operation when the RCS is at low temperature (during startup and

shutdown), the staff is concerned about potential reactor vessel damage as a

result of an overpressure transient. The applicant has provided a description

of the low temperature overpressure protection design, describing a system that,

once manually-enabled, will autmomatically monitor RCS pressure and temperature

and will provide backup control of the pressurizer PORVs. Proposed administrative

procedures to reduce the potential for low temperature overpressure events also

were addressed. Additionally, the applicant has discussed design low temperature overpressure transients and further details on the administrative procedures

associated both with operation of the system and with reducing the potential for

overpressure events. The applicant has committed to provide plant-specific

analyses, which will dictate governing Technical Specifications.

The staff finds the conceptual design of the applicant's low temperature overpressure protection system to be acceptable; the acceptability of the system's

electrical design is discussed in Sections 7.6.5 and 8.3.3 of this report.

5.2.3 Reactor Coolant Pressure Boundary Materials

GDC 14, 15, and 31 of 10 CFR 50 require that the reactor coolant pressure

boundary (RCPB) have an extremely low probability of abnormal leakage and be

Watts Bar SER 5-3
designed with sufficient-margin to ensure that the design conditions are not

exceeded during normal operation and anticipated failure of the RCPB is

minimized.

The materials used for construction of components of the RCPB have been identified by specification and found to be in conformance with the requirements of

Section III of the ASME Code.

Fracture toughness of components in the RCPB are discussed in Section.5.3.2.

The materials of construction of the RCPB exposed to the reactor coolant have

been identified, and all of the materials are compatible with the expected

environment, as proven by extensive testing and satisfactory performance.

General corrosion of all materials, except carbon and low alloy steel, is

negligible. There are small unclad low alloy steel surfaces exposed to reactor

coolant in the Watts Bar Unit I reactor vessel inlet nozzles. For this

material in this environment, the corrosion rate data show the required use of

reinforcements for nozzles has not been violated when considering the maximum

cross sectional area of the grind outs, including the maximum depth of penetration as a result of corrosion. Inservice inspection is not required because

experience, to date, shows the corrosion rates in this reactor coolant environ-*

ment will not be of concern.

As the thermal insulation used on the RCPB is the reflective stainless steel

type, the presence of chlorides and other compounds that can cause intergranular stress corrosion cracking of stainless steels is prevented, and

Regulatory Guide 1.36, "Non-metallic Thermal Insulation for Austenitic Steel"

is not applicable.

The controls imposed on reactor coo'lant chemistry are in conformance with the

recommendations of Regulatory Guide 1.44 and provide reasonable assurance that

the RCPB components will be adequately protected during operation from conditions that could critically lead to stress corrosion of the materials and loss

of structural integrity of a component. The instrumentation and sampling

provisions for monitoring reactor coolant water chemistry provide adequate

measurement capability for detecting significant changes on a timely basis.

Compliance with the recommendations of these Regulatory Guides constitutes an

acceptable basis for satisfying the applicable requirements of GDC 14 and 31.

The controls imposed on welding preheat temperatures are not in total conformance with the recommendations of Regulatory Guide 1.50, "Control of Preheat

Temperature for Welding Low Alloy Steels." However, the acceptance of WCAP7

8577 by the staff allows an alternative to regulatory position 2, which was

followed. The applicant also did not meet regulatory position 1.b, which

requires that weld procedure qualifications be performed at the minimum preheat

temperature. The staff agrees that qualification within the range of preheat

temperature allowed by ASME Code is acceptable because it is not possible to

control the temperature of a welding qualification plate to a given temperature

with no tolerances. Accordingly, it is the staff position that the controls

imposed provide reasonable assurance that cracking of components made from low

alloy steels will not occur during fabrication and minimize the possibility of

subsequent cracking as a result of hydrogen being retained in the weldment.

Watts Bar SER 5-4
The recommendations of Regulatory Guide 1.34, "Control of Electroslag Weld

Properties," are not applicable because electroslag welds were not used in the

RCPB.

The controls imposed upon austenitic stainless steel are either in accordance

with Regulatory Guides 1.31, and 1.44, or, if they are not in accordance with

these Regulatory Guides, the positions and actions taken have previously been

accepted by the staff.

The material selection, fabrication practices, examination procedures, and

protection procedures performed provide reasonable assurance that the austenitic stainless steel in the reactor coolant pressure boundary will be in a

metallurgical condition which precludes susceptibility to stress corrosion

cracking during service. Conformance with these Regulatory Guides constitutes

an acceptable basis for meeting in part the requirements of GDC 1 and 14.

5.2.4 Reactor Coolant Pressure Boundary Inservice Inspection and Testing

GDC 32, of Appendix A, of 10 CFR 50, requires, in part, that components that

are part of the RCPB be designed to permit periodic inspection and testing of

important areas and features to assess their structural and leaktight integrity.

To ensure that no deleterious defects develop during service, selected welds

and weld-affected-zones (HAZ) will be inspected periodically. The applicant

has stated that the RCPB is designed so that provisions for access required by Paragraph IWA-1500 of Section XI of the ASME Code can be followed. Methods

have been developed to facilitate the remote examination of those areas of the

reactor vessel not readily accessible to examination personnel.

10 CFR 50.55a(g) defines the detailed requirements for the preservice and

inservice inspection programs for light-water-cooled nuclear power facility

Class 1 components. Based on a construction permit date of January 23, 1973,

this section of the Code of Federal Regulation requires that a preservice

inspection program be developed and implemented using at least the edition and

addenda of Section XI of the ASME Code in effect 6 months before to the date of

issuance of the construction permit. Also, the initial inservice inspection

program mu~t comply with the requirements of the latest edition and addenda of

Section XI of the ASME Code in effect 12 months before the date of issuance of

the operating license, subject to the limitations and modifications listed in

10 CFR 50.55a(b).

The applicant has stated that the preservice inspection program shall be in

accordance with Summer 1975 Addenda, 1974 Edition of Section XI of the ASME

Code to the extent practical as presently required by 10 CFR 50.55a. The

preservice inspection program, including revisions submitted by letters dated

May 19 and June 4, 1982, is currently under staff review and additional information has been requested from the applicant.

The staff has recently identified a problem concerning the effectiveness of

ultrasonic examination techniques to examine the primary piping system.

Certain ultrasonic techniques may not be adequate to consistently detect and

reliably characterize service-induced flaws during the inservice inspection of

thick-wall cast stainless steel components to acceptance standards of Paragraph IWB-3500 of Section XI. The staff considers this issue of effectiveness

Watts Bar SER 5-5
of the ultrasonic examination of thick-walled stainless steel components generic

in nature and will resolve the issue on a generic basis.

Specific areas where ASME Code examination requirements cannot be met will be

identified as performance of the preservice examination progresses. The staff's

complete evaluation of the preservice inspection (PSI) program will be presented

in a supplement to this report after the applicant submits the required examination information and identifies all plant-specific areas where ASME Code

Section XI requirements cannot be met and provides a supporting technical

justification.

The initial inservice inspection program has not been submitted by the applicant.

The staff will evaluate the program after the applicable ASME Code edition and

addenda can be determined based on Section 50.55a(b) of 10 CFR 50, but before

the first refueling outage when inservice inspection begins. The conduct of

periodic inspections and hydrostatic testing of pressure-retaining components

of the RCPB in accordance with the requirements of Section XI of the ASME Code

and 10 CFR 50 will provide reasonable assurance that evidence of structural

degradation or loss of leaktight integrity occurring during service will be

detected in time to permit corrective action before the safety functions of a

component are compromised. Compliance with the preservice and inservice

inspections required by this Code and 10,CFR 50 constitutes an acceptable basis

for satisfying the inspection requirements of GDC 32.

5.2.5 Reactor Coolant Pressure Boundary Leakage Detection

A limited amount of leakage is to be expected from components which comprise

the RCPB. Means are provided for detecting and identifying this leakage as

practical in accordance with the requirements of GDC 30. Leakage is classified

into two types, identified and unidentified. Expected leakage is considered

identified leakage and is monitored, limited, and separated from other leakage

(unidentified) by directing it to closed collection systems as identified in

the guidelines of position C.1 of Regulatory Guide 1.45, "Reactor Coolant

Pressure Boundary Leakage Detection System."

Identified Leakage

The sources, disposition, and indication of identified leakage are as follows:

(1) Valve stem leakage is directed either to the reactor coolant drain tank or

to the pressurizer relief tank.

(2) Reactor coolant pump shaft seal leakage is directed tc the reactor coolant

drain tank. The seals are equipped with temperature sensors; excessive

leakage is alarmed and monitored in the control room.

(3) Reactor vessel closure head leakage is monitored for temperature increase

which is alarmed in the control room.

(4) Pressurizer safety and relief valve leakage is directed to the pressurizer

relief tank. Each safety valve line and relief line is provided with a

temperature sensor that initiates a high temperature alarm in the control

room when excessive leakage is detected.

Watts Bar SER 5-6
Unidentified Leakage

Unidentified leakage, which includes steam generator tube or tube sheet and

intersystem leakage, is monitored by several devices as identified in Positions C.2, C.3, and C.4 of Regulatory Guide 1.45. Indication of unidentified

leakage from the RCPB is provided by the containment air particulate monitors,

containment radiogas monitors, and the containment floor and equipment drain

sump level monitors. The containment air particulate and radiogas monitors

operate continuously to detect the emission of radiation from particulates and

gas. Leakage that is collected by the equipment and floor drains is routed to

the containment sump. This sump is equipped with a level monitor that transmits

a signal to level change rate instrumentation that, in turn, is converted to

flow rate indication. The sensitivity of these monitors is such that leaks of

1 gpm are detectable in less than 1 hr and meet the guidelines of Position C.5

of Regulatory Guide 1.45. Additional sources of indication of unidentified leakage include containment pressure, temperature, and humidity indicators.

The methods of detecting intersystem leakage are dependent on the interfacing

system. The applicant has identified the possible intersystem leakage paths,

which included the safety injection system (SIS), upper head injection system

(UHI), accumulators, steam generators, residual heat removal system (RHR), and

the component cooling system (CCS). Leakage is detected and monitored by the

use of pressure, temperature, and level sensors. Leakage also is indicated by

the measured amount of radiation activity and changes in the inventory balance

of any given system.

Leakage across the SIS cold-leg and hot-leg check valves is detected by pressure transmitters and is indicated and alarmed in the control room. Leakage

across the UHI check valves is detected by an increase in water level in the

UHI accumulator surge tank which is indicated and alarmed in the control room.

Leakage across the cold-leg accumulator check valves is detected by an increase

in accumulator inventory and is indicated and alarmed in the control room.

Steam generator tube leakage is monitored by the condenser air vacuum pump

exhaust radiation monitors, steam generator radiation monitors, or routine

steam generator water samples. Leakage past the RHR injection check valves is

detected by pressure transmitters and is indicated and alarmed in the control

room. Reactor coolant system leakage into the CCS is detected with radiation

monitors located downstream of the CCS heat exchangers, and is indicated,

recorded, and alarmed in the control room. If leakage is alarmed and confirmed

in a flow path with no indicators, the staff will require that the Technical

Specifications include that a water inventory material balance be initiated

within 1 hr to determine the extent of the leakage.

Indicators and alarms that monitor leakage have been provided in the control

room. The air particulate and radiogas monitors are capable of remaining

functional during and after a safe shutdown earthquake. The component cooling

system liquid effluent monitors and steam generator blowdown liquid sample

monitors will be functional following a safe shutdown earthquake. The design

of the leakage detection systems readily permits calibration, inspection, and

testing for operability during plant operations as identified in the guidelines

of Regulatory Guide 1.45, Positions C.6, C.7, and C.8.

Watts Bar SER 5-7
Based on its review, the staff concludes that the RCPB leakage detection systems

are diverse and provide reasonable assurance that identified and unidentified

primary system leakage will be detected in a timely manner. The systems meet

the requirements of GDC 30 with respect to RCPB leakage detection and identification, and the guidelines of Regulatory Guide 1.45 with respect to RCPB

leakage detection system design, and are, therefore, acceptable.

5.3 Reactor Vessel

5.3.1 Reactor Vessel Materials

GDC 31, requires, in part, that the RCPB be designed with sufficient margin to

ensure that when the boundary is stressed under operating, maintenance, and

testing conditions, it behaves in a nonbrittle manner and the probability of

rapidly propagating fracture is minimized. GDC 32 requires,Ain part, that the

RCPB be designed to permit an appropriate material surveillance program for the

reactor pressure boundary.

The staff has reviewed the materials selection, toughness requirements, and the

extent of materials testing in accordance with the above GDC. The ferritic

'materials were specified to meet the toughness requirements of the 1971 Edition

of the ASME Boiler and Pressure Vessel Code,.Section III.

Appendix G, "Fracture Toughness Requirements," and Appendix H, "Reactor Vessel

Material Surveillance Requirements," of 10 CFR 50 specify the fracture toughness requriements for the ferritic materials of the RCPB. The Watts Bar

ferritic materials were qualified by impact testing in accordance with the ASME

Code, Section III, with Addenda through Winter 1971 and evaluated by the

applicant in accordance with Appendix G, Section III of the 1971 Edition, 1972

Summer Addenda. The edition and addenda of the ASME used by the applicant to

qualify and evaluate the RCPB materials meet the requirements of 10 CFR 50.55a.

5.3.1.1 Compliance with Appendix G, 10 CFR 50

The staff has evaluated the applicant's FSAR to determine the degree of

compliance with the fractureotoughness requirements of Appendix G, 10 CFR 50.

The evaluation indicates that the applicant has met all the requirements of

Appendix G, except for Paragraphs III.B.4, IV.A.1, and IV.B. The evaluation of

each of these areas and the applicant's use of RCPB materials that have minimum

yield strength over 50 ksi follows.

Paragraph L.A of Appendix G requires the applicant to demonstrate to the

Commission on an individual case basis the adequacy of the fracture toughness

of any ferritic material that has a specified minimum yield strength over

50 ksi and is used in pressure-retaining components of the RCPB.

Westinghouse Topical Report WCAP-9292, "'Dynamic Fracture Toughness of ASME SA508 Class 2a and ASME SA-533 Grade A Class 2 Base and Heat-Affected Zone

Material and Applicable Weld Metals," contain fracture toughness data from a

sufficient number of heats of SA-533. Grade A Class 2 and SA-508 Class 2a steels

to demonstrate the generic fracture thoughness of these high strength materials.

The applicant indicates that the conclusion in report WCAP-9292 concerning

Watts Bar SER 5-8
SA-533 Grade A Class 2 and SA-508 Class 2a in the RCPB are applicable to Watts

Bar. The applicant indicates that these high strength materials were used in

pressurizer and steam generator RCPB components in Watts Bar.

Based on report WCAP-9292, the staff considers the applicant's use of SA-533

Grade A Class 2 and SA-508 Class 2a steels in pressurizer and steam generator

RCPB components acceptable.

Paragraph III.B.4 of Appendix G requires that the testing personnel shall be

qualified by training and experience and should be able to perform the tests in

accordance with written procedures. For Watts Bar component testing, no

written procedures were in existence. However, the applicant has supplied

sufficient information to demonstrate that the intent of Paragraph III.B.4 has

been met. The individuals performing the testing were qualified by on-the-job

training and past experience. Because these tests are relatively routine in

nature and are continually being performed in the laboratory that conducted

these tests, it is unlikely that the tests were conducted improperly. Consequently, the staff concludes that an exemption for not performing the tests

in accordance with written procedures is justified.

Paragraph IV.A.1 of Appendix G requires that RCPB materials be tested to the

requirements of NB 2330 of the ASME Code. Paragraph NB 2330 of the ASME Code

requires that a reference temperature, RTNDT, be determined for each ferritic

material of the RCPB and that this reference temperature be used as a basis for

providing adequate margins of safety for reactor operation. The value of RTNDT

is defined in the ASME Code as the higher of either (1) the nil ductility

temperature, as determined by the dropweight test, or (2) a temperature of 60OF

less than the temperature at which 50 ft-lb energy and 35 mils lateral expansion is achieved, as determined by the Charpy V-notch (CVN) impact test. The

CVN impacts test for base metal is to be conducted using specimens oriented

in the transverse direction.

The applicant has complied with the requirements of Paragraph IV.A.1 of

Appendix G for all RCPB materials except for reactor vessel base metals that

are located outside the beltline region and the beltline intermediate-to-lowershell root weld in Watts Bar.

For Watts Bar, the value of RTNDT for all base metal outside the reactor vessel

beltline was determined in accordance with Paragraph IV.A.1 except that the CVN

specimens were oriented in the longitudinal rather than transverse direction.

To compensate for specimen orientation, the temperatures at which the 50 ft-lb

energy levels would have been achieved for transverse specimens were estimated

at 20F' higher than the temperature indicated from the longitudinal data.

Based on CVN impact data in Welding Research Council Bulletin 217, and Electric

Power Research Institute Reports, NP-121 Volume II, April 1976, and NP-933

December 1978, and CVN impact data submitted by other applicants, the staff

concludes that adding 20°F to compensate for the effect of orientation is not

conservative. To compensate for specimen orientation, the temperature at which

50 ft-lb energy and 35 mils lateral expansion would occur for transversely

Watts Bar SER 5-9
oriented specimens must be estimated at 30F' higher than the temperature

indicated from the longitudinal CVN impact data.

The staff has evaluated the Watts Bar RCPB base metal by adding 30F* to the

temperature at which the longitudinally oriented CVN impact specimens had 50

ft-lb energy and 35 mils lateral expansion. This analysis indicates the intermediate shell forgings in Watts Bar.are limiting throughout the operating life

of the reactor vessels. The applicant has provided transversely oriented CVN

impact data from these intermediate shell forgings that are located in the

reactor vessel beltlines.

The value of the RTNDT for beltline intermediate to lower shell weld in the

weld root area of Watts Bar were not determined in accordance with Paragraph

IV.A.1 because the dropweight tests were not performed and the CVN impacts

tests were performed at a single temperature. These CVN impact tests results

were not performed at a test temperature high enough to detemine an RTNDT for

the beltline root weld material. The applicant assumed an RTNDT for this weld

material of O*F. The applicant submitted dropweight data and CVN impact data

from other weld materials which had been welded by same manufacturer using the

same process, flux type, and weld wire type as the weld metal in the root weld.

Based on these data, the staff considers the assumed RTNDT conservative, and

the weld will not be limiting for operation of the reactor vessel.

The applicant has not complied with the explicit requirements of Paragraph

IV.A.1 of Appendix G for determining the RTNDT forferritic reactor vessel base

metals located outside the beltline region and the beltline intermediate-tolower-shell root welds because these materials were procured before publication

of Appendix G, 10 CFR 50. However, based on the above discussion, an exemption

to testing these materials to the requirements of Paragraph IV.A.1 of Appendix G,

10 CFR 50 is justified.

Paragraph IV.B of Appendix G:requires that the reactor vessel beltline materials

have a minimum upper shelf energy, as determined from Charpy V-notch impact

tests on unirradiated specimens in accordance with Paragraph NB-2322.2(a) of

the ASME Code, of 75 ft-lb, unless it can be demonstrated to the Commission

by appropriate data and analyses that lower values of upper shelf energy

provide adequate margin for deterioration from irradiation.

All Watts Bar reactor vessel beltline material meets Paragraph IV.B, Appendix G,

requirements except for the intermediate shell forging in the beltline of Unit 1.

The intermediate shell forging in the beltline of Unit I has an unirradiated

CVN upper shelf impact energy ranging from 60 to 64 ft-lb, which is less than

the 75 ft-lb required by Paragraph IV.B of Appendix G, 10 CFR 50. To determine

if Paragraph IV.B of Appendix G, 10 CFR 50, can be satisfied for this forging,

the staff has evaluated the effect of neutron fluence on the forging's upper

shelf impact energy. Based on the conservative method of evaluation in

Regulatory Guide 1.99, Revision 1, "Effect of Residual Elements on Predicted

Radiation Damage to Reactor Vessel Materials," this forging's upper shelf

impact energy will remain above the safety margins required by Appendix G, 10

CFR 50 for 4.3 effective full power years. Although the intermediate shell

Watts Bar SER 5-10
forging's predicted upper shelf impact energy is marginal, the actual fracture

toughness of this forging will be monitored by a materials surveillance program

throughout its service life. The CVN impact tests from the material in the

surveillance capsule will be used to establish limitations on the pressure and

temperature of Unit 1 that will ensure that the intermediate shell forging's

material properties remain above the safety margins required by Appendix G,

10 CFR 50. In addition, the marginal upper shelf fracture toughness of this

forging will be reevaluated using the improved engineering method and safety

criteria developed under generic Task A-11, when this task is completed. Based

on the above discussion, the staff has determined that an exemption from the.

minimum upper shelf energy requirements of Paragraph IV.B of Appendix G is

justified.

5.3.1.2 Compliance with Appendix H, 10 CFR 50

The toughness properties of the reactor vessel beltline materials wil.l be

monitored throughout the service life of Watts Bar by a materials surveillance

program that must meet the requirements of ASTM Standard E 185-73, "Standard

Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels," and

Appendix H, 10 CFR 50. The staff has evaluated the applicant's information for

degree of compliance to these requirements and found that the applicant has met

all the requirements of Appendix H, 10 CFR 50.

5.3.1.3 Conclusions on Compliance with Appendices G and H, 10 CFR 50

Based on its evaluation, the staff finds that the applicant has met all the

requirements of Appendices G and H, 10 CFR 50, except for Paragraph III.B.4,

I.V.A.1, and IV.B of Appendix G. The staff technical evaluation has not

identified any practical methods by which the existing Watts Bar reactor

vessels can comply with the specific requirements of Paragraphs III.B.4,

IV.A.1, and IV.B of Appendix G, 10 CFR 50. However, the alternate methods

proposed to demonstrate compliance with these paragraphs of Appendix G have

been reviewed and evaluated, and have been found to demonstrate that the

safety margins required by Appendix G have been achieved.

Compliance with Appendices G and H and the fracture toughness requirements of

Section III of the ASME Code ensures that the ferritic components in the

primary coolant pressure boundary will behave in a nonbrittle manner, that the

probability of rapidly propagating fracture is minimized, and that an appropriate material surveillance program exists to monitor radiation damage for the

reactor pressure boundary. Compliance with the requirements of the NRC regulations and the specified codes and standards satisfies the requirements of

the Commission's GDC 31 and 32.

Based on the foregoing, pursuant to 10 CFR 50.12, exemptions from the specific

requirements of Appendix G of 10 CFR 50, as discussed in Section 5.3.1.1, are

authorized by law and can be granted without endangering life or property or

the common defense and security and are otherwise in the public interest. The

staff concludes that the public is served by not imposing certain provisions of

Appendix G of 10 CFR 50 that have-been determined to be either impractical or

would result in hardship or unusual difficulties without a compensating increase in the level of quality and safety.

Watts Bar SER 5-11
Furthermore, the staff has determined that the granting of these exemptions

does not authorize a change in effluent types of total amounts nor an increase

in power level and will not result in any significant environmental impact.

The staff has concluded that these exemptions would be insignificant from the

standpoint of environmental impact and, pursuant to 10 CFR 51.5(d)(4), that an

environmental impact appraisal need not be granted in connection with this

action.

5.3.2 Pressure-Temperature Limits

Appendices G and H of 10 CFR 50 describe the conditions that require pressuretemperature limits for the RCPB and provide the general bases for these limits.

These appendices specifically require that pressure-temperature limits must

provide safety margins for the RCPB at least as great as the safety margins

recommended in the ASME Code, Section III, Appendix G, "Protection Against

Non-Ductile Failure." Appendix G to 10 CFR 50 requires additional safety

margins whenever the reactor core is'critical, except for low-level physics

tests.

The following pressure-temperature limits imposed on the RCPB during operating

and testing conditions are reviewed to ensure that they provide adequate safety

margins against nonductile behavior or rapidly propagating failure of ferritic

components as required by GDC 31:

(1) preservice hydrostatic tests

(2) inservice leak and hydrostatic tests

(3) heatup and cooldown operations

(4) core operation

The staff has reviewed the pressure temperature limits for Unit 1; the

applicant has not yet provided pressure temperature limits for Unit 2. The

staff will review the Unit 2 pressure-temperature limits as part of the Unit 2

Technical Specifications and will report its findings in a supplement to the

SER.

As discussed in Section 5.3.1.1, the staff has determined that the fracture

toughness of the limiting beltline material in Unit 1 will remain above the

safety margins required by Appendix G, 10 CFR 50 for 4.3 effective full power

years. Therefore,'the staff will grant approval for. the use of the applicant's

pressure-temperature curves for 4.3 effective full power years. After this

period of time, the pressure-temperature limits will be calculated from the

actual shift in the RTNDT of the materials in the reactor vessel's material

surveillance program.

The pressure-temperature limits to be imposed on the reactor coolant system for

all operating and testing conditions to ensure adequate safety margins against

nonductile or rapidly propagating failure must be in conformance with

established criteria, codes, and standards to be acceptable to the staff. The

use of operating limits based on these criteria, as defined by applicable

regulations, codes, and standards, will provide reasonable assurance that

nonductile or rapidly propagating failure will not occur, and will constitute

and acceptable basis for satisfying the applicable requirements of GDC 31.

Watts Bar SER 5-12,
5.3.3 Reactor Vessel Integrity

The staff has reviewed the FSAR sections related to the reactor vessel integrity

of Watts Bar. Although most areas are reviewed separately in accordance with

other review plans, reactor vessel integrity is of such importance that a

special summary review of all factors relating to reactor vessel integrity is

warranted.

The staff has reviewed the information in each area to ensure that it is

complete and that no inconsistencies exist that would reduce the certainty of

reactor vessel integrity. The areas reviewed (all in Section 5.3.1) are:

(1). design

(2) materials of construction

(3) fabrication methods

(4) operating conditions

The staff has reviewed the above factors as they contribute to Watts Bar reactor

vessel structural integrity of and concludes that the applicant has complied

with Appendices G and H, 10 CFR 50, except for paragraphs III.B.4, IV.A.1, and

IV.B of Appendix G. for which the applicant has provided sufficient information

to 'justify exemptions (see Section 5.3.1.1).

Paragraph III.B.4 of Appendix G requires the applicant to qualify the personnel

performing the CVN impact testing according to written procedures. Although

personnel were not qualified by written procedures, the individuals conducting

the CVN impact tests were qualified by on-the-job training and past experience,

which the staff concludes is adequate and sufficient to justify an exemption to

paragraph III.B.4 of Appendix G.

Paragraph IV.A.1 of Appendix G requires that a reference temperature, RTNDT, be

determined per paragraph NB-2330 of the ASME Code for each ferritic material in

the RCPB. Although the applicant did not determine the RTNDT per paragraph

NB-2330 of the ASME Code for each ferritic material, the critical RTNDT for

operating, maintenance, and testing conditions has been determined based on

additional information available in the literature and additional data supplied

by the applicant. Therefore, the staff has concluded that an exemption to

paragraph IV.A.1 of Appendix G is justified.

Paragraph IV.B of Appendix G requires that the reactor vessel beltline materials

have a minimum upper shelf CVN energy of 75 ft-lb Unless it can be demonstrated

that lower values of upper-shelf CVN energy less than 75 ft-lb. The staff has

determined that this material will have adequate margins for irradiation

deterioration or a minimum of 4.3 effective full power years. After this

period, the adequacy of this material will be determined from the fracture

toughness test results of the reactor vessel material surveillance program.

Therefore, the staff concludes than an exemption to paragraph IV.B is justified.

The staff has reviewed all factors contributing to the structural integrity of

the reactor vessel and concludes there are no special considerations that make

it necessary to consider potential reactor vessel failure for Watts Bar Unit 1.

Watts Bar. SER 5-13
The staff will complete the evaluation of Watts Bar Unit 2 after receipt of its

pressure-temperature limits as part of the review of the Unit 2 Technical

Specifications and will report its findings in a supplement to the SER.

5.4 Component and Subsystem Design

5.4.1 Reactor Coolant Pumps

5.4.1.1 Pump Flywheel Integrity

GDC 4 requires, in part, that nuclear power plant structures, systems,'and

components important to safety be protected against the effects of missiles

that might result from equipment failures. Because reactor coolant pump

flywheels have large masses and rotate at speeds of approximately 1200 rpm

during normal operation, a loss of flywheel integrity could result in high

energy missiles and excessive vibration of the reactor coolant pump assembly.

The safety consequences could be significant because of possible damage to the

reactor coolant system, the containment, or the engineered safety features.

Adequate margins of safety and protection against the potential for damage from

flywheel missiles can be achieved by the use of suitable material, adequate

design, and inspection. The flywheels have been fabricated from SA-533

Grade B, Class 1 steel. This material has been produced by a process that will

minimize flaws and improve fracture toughness, and has been cut, machined,

finished, and inspected in accordance with Section III of the ASME and

Regulatory.Guide 1.14, "Reactor Coolant Pump Flywheel Intergrity," Revision 1.

The reactor coolant pump has been designed for a speed 125 percent that of the

normal synchronous speed of the motor (approximately 1500 rpm). The margin

against failure for the flywheel is significantly higher because the minimum

speed for ductile failure is estimated to be.much higher than 125 percent of

operating speed for flywheels of the design used at Watts Bar. The nominal

operating temperature is specified to be 110°F. The applicant has stated that

the NDTT of the flywheel material is no higher than 10F and CVN impact energy

at 70'F if greater than 50 ft-lbs. These fracture toughness data indicate that

the RTNDT of the flywheel material is less than 10OF and that the normal operat

ing temperature of the flywheel will be 100°F above RTNDT. Based on the fracture

toughness data and normal operating temperatures of the flywheel, the applicant

has demonstrated compliance with Regulatory Guide 1.14, paragraph C.l.b. The

staff concludes that the reactor coolant pump flywheels in Watts Bar have a

margin of safety against flywheel missiles equivalent to that recommended in

Regulatory Guide 1.14. Compliance with Regulatory Guide 1.14 provides a basis

acceptable to the'staff for satisfying the requirements of GDC 4.

5.4.2 Steam Generators

5.4.2.1 Steam Generator Materials

The staff concludes that the materials specified for the steam generator are

acceptable and meet the requirements of GDC 1, 14, 15, and 31, and Appendix B

Watts Bar SER 5-14
of 10 CFR 50. This conclusion is based on the considerations discussed

below.

The applicant has met the requirements of GDC 1 with respect to codes and

standards by assuring that the materials for use in Class 1 and 2 components

were fabricated and inspected in conformance with codes, standards, and specifications acceptable to the staff. Welding qualification, fabrication, and

inspection during manufacture and assembly of the steam generator was done in

conformance with the requirements of Sections III and IX of the ASME Code.

The requirements of GDC 14 and 15 have been met by ensuring that the RCPB and

associated auxiliary systems have been designed, fabricated, erected, and

tested so as to have an extremely low probability of abnormal leakage, of

rapidly propagating failure, and of gross rupture during normal operation and

anticipated operational occurrences.

The primary side of the steam generator is designed and fabricated to comply

with Class 1 criteria of the ASME Code as required by the staff. The secondary

side of the pressure boundary of the steam generator was designed, manufactured,

and tested to Class 1 criteria although the required classification is ASME

Code Class 2.

The crevice between the tube sheet and the inserted tube will be minimal

because the tube will be expanded to the full depth of insertion of the tube

in the tube sheet. The tube expansion and subsequent positive contact pressure between the tube and the tube sheet will preclude a buildup of impurities

from forming in the crevice region and reduce the probability of crevice

boiling.

The requirements of GDC 31 have been met with respect to the fracture toughness

of the ferritic materials since the pressure boundary materials of ASME Class 1

components of the steam generator will comply with the fracture toughness

requirements and tests of subarticle NB-2300 of Section III of the ASME Code.

The-materials of the ASME Class 2 components of the steam generator will comply

with the fracture toughness requirements of subarticle NB-2300 of Section III

of the ASME Code.

The requirements of Appendix B of 10 CFR 50 have been met because the applicant

has committed to the onsite cleaning and cleanliness control during fabrication

which will conform to the recommendations of Regulatory Guide 1.37, "Quality

Assurance Requirements for Cleaning of Fluid Systems and Associated Components

of Water-Cooled Nuclear Power Plants."

Reasonable assurance of the satisfactory performance of steam generator tubing

and other steam generator materials (except for the generic concern regarding

the potential for tube degradation in Westinghouse Model D steam generators

discussed in Section 5.4.2.2) is provided by (1) the design provisions and

manufacturing requirements of the ASME Code, (2) secondary water monitoring and

control, and (3) the limiting of condenser inleakage. The controls described

above combined with conformance with applicable codes, standards, staff

positions, and Regulatory Guides constitute an acceptable basis for meeting, in

part, the requirements of GDC,1, 14, 15, and 31 and Appendix B of 10 CFR 50.

Watts Bar SER 5-15
5.4.2.2 Westinghouse Model D Steam Generator Tube Degradation Potential

Recently, a generic problem concerning the potential for tube degradation

caused by flow-induced vibration in the preheater section of Westinghouse

Model D-3 steam generators has been identified in nondomestic plants.

Significant tube wall reductions in the preheater section of the Ringhals 3

(Sweden) and Almaraz 1 (Spain) have been indicated by eddy current testing.

The staff is monitoring developments closely at the McGuire facility--the only

operating domestic plant with Model D steam generators--and at foreign plants

to evaluate its impact on the other domestic facilities with similar steam

generator designs. Westinghouse has initiated a program to resolve this

problem. This includes the use of scale model testing, analytical studies,

wear testing, periodic eddy current testing of operating units, and the use of

diagnostic internal and external instrumentation in selected operating units.

Details of the Westinghouse efforts to date were presented to the NRC on

February 19, 1982. Because the Watts Bar facility uses Westinghouse Model D-3

steam generators, the staff requested (letter dated January 22, 1982) the

applicant to address his proposed actions to resolve this problem by indicating

(1) what instrumentationwas to be used for the detection of flow-induced

vibrations and (2) what procedures were to be followed for testing and startup

of the Watts Bar facility.

By letter dated March 1, 1982, the applicant responded that it was his understanding from Westinghouse that one Model D-3 unit and one Model D-4 unit have

been equipped with internal and external instrumentation. After the data from

these units are collected and evaluated, the applicant stated that he will be

able to better define any future action that might be required. Pending the

results of this investigation, the applicant stated he has not changed the

startup test program at Watts Bar as described in the FSAR.

During a May 19, 1982 meeting, the applicant informed the staff that a Model D-2/

D-3 steam generator preheater modification design review panel composed of

representatives from TVA, Duke Power Company, and South Carolina Electric and

Gas Company was being established. All panel members own nuclear power facilities with Model D-2/D-3 steam generators. The purpose of the independent design

review panel is (1) to evaluate the final proposed modifications. to the preheater to ensure that applicable regulatory criteria are met, (2) to verify

the functional adequacy of the proposed design, and (3) to ensure the safety

acceptability of the design.

Because of the safety concerns relative to steam generator tube damage, the

staff considers the potential for such damage to have significant safety

implications. However, because the onset of tube degradation does not begin

until exceeding power levels greater than 50 percent, the staff concludes that

this issue does not affect fuel load and low power licensing. The staff will

continue to monitor developments of the Westinghouse program, and evaluate

the applicant's proposed resolution upon submittal. The staff will address

this matter in a supplement to the SER.

5.4.3 Residual Heat Removal System

The residual heat removal (RHR) system is used in conjunction with the main

steam and main or auxiliary feedwater systems during normal plant shutdown to

Watts Bar SER 5-16
reduce reactor coolant system temperatures from hot conditions to cold shutdown

and to provide core cooling during initial refueling operations. The principal

basis for the staff review of the RHR system is its conformance to the requirements in GDC 34. The residual heat removal system is operated in several other

modes. These are

(1) startup - connected to chemical and volume control systems, acting as an

alternate letdown path to control reactor pressure

(2) refueling - used for refilling the refueling canal

(3) emergency core cooling system - the RHR system is aligned during power

operation and hot shutdown for low pressure coolant injection into the

reactor coolant system as an integral part of the emergency core cooling

system

Initially, heat is rejected by the steam generators to the condenser or atmosphere. When the reactor coolant temperature and pressure have been reduced to

approximately 350'F and 400 psi, the RHR system is put into operation to reduce

the reactor coolant temperature to the refueling temperature. The system takes

its suction from one hot leg through two isolation valves in series. The discharge from each of the two parallel residual heat removal pumps passes through

a separate heat exchanger, which is cooled by the component cooling water system.

The flow is then discharged to the four reactor cold legs.

The applicant has indicated that under normal shutdown conditions the RHR

system is capable of removing residual heat from the reactor in accordance with

the requirements of GDC 34; with only one residual heat removal train in

operation, the time required to cool the reactor coolant system to the cold

shutdown condition (200*F) is 34 hours; with both pumps in operation, the

cooldown time is 16 hours. The staff has reviewed the component cooling water

system to ensure that sufficient cooling capability is available to the heat

exchangers. The acceptability of this cooling capability and its satisfaction

of GDC 44, 45, and 46 are discussed in Section 9.0.

The staff has reviewed the description of the RHR system and the piping and

instrumentation drawings to determine whether the system can be operated with

or without offsite power and assuming a single failure.

The two RHR pumps are connected to separate buses which can be powered by

separate diesel generators in the event of loss of offsite power. The parallel

trains containing the RHR pumps and heat exchangers provide redundancy of the

major components.

The two isolation valves in series in the suction line each have a bypass line

containing a normally closed, motor-operated valve. This alternate path can be

used if one of the normal isolation valves cannot be opened. The staff

requires that the valves in the bypass lines be normally locked closed at all

times with the power removed-from the operators. Strict administrative keylock

control must be demonstrated. If the alternate bypass lines are utilized the

staff requires that the malfunctioning main isolation valve be corrected and

the valve in the bypass line locked closed (with power removed) before the

plant is repressurized. The staff also noted that an inadvertent closure of

Watts Bar SER 5-17
one of the main isolation valves during RHR operation would result in loss of

suction and potential failure of the RHR pumps. The applicant has committed to

provide, before startup, an RHR flow alarm that will alert the operator to

initiate alternate tooling modes (such as open bypass) in the event of loss RHR

pump suction. The staff finds this acceptable pending verification by the

staff that the alarm is installed.

To address the spurious mispositioning of.RHR valves (such as RHR spray valves),

the applicant has described revisions to the operator training program and

emergency procedures to respond to a mispositioning occurrence. The staff

finds these measures acceptable. Any additional considerations that result

from the generic review of this area will be applied to Watts Bar.

The staff has reviewed the applicant's instruments and controls necessary to

operate the RHR system and finds that they are acceptable for operating the

RHR system. Discussion of the adequacy of instrumentation and control system

design is addressed in Section 7.0.

RHR overpressure protection is provided by relief valves on the suction line

and each of the discharge lines. The suction line relief valve has a capacity

of 900 gpm at 450 psig, which is sufficient to relieve flow equivalent to two

charging pumps. The relief valves in the discharge lines have a capacity of

20 gpm at a pressure of 600 psig to protect the system from leakage past the

check valves. These relief valves are adequate to protect the RHR system from

overpressurization.

The RHR system is designed to provide an adequate isolation between the reactor

coolant system and RHR system when the reactor coolant system is above the

design pressure of the residual heat removal (600 psia) as follows:

(1) There are two separate and redundant motor-operated isolation valves

between the RHR pump suction line and the reactor coolant system (see

previous discussion of bypass valves). These valves are interlocked with

one of the independent reactor coolant system pressure signals. Valve

opening is prevented until the reactor coolant system pressure falls to a

value of 425 psig. Already opened valves are closed when the reactor

coolant system pressure rises to 750 psig. This arrangement satisfies the

staff position on the suction valves.

(2) There are two check valves and an open motor-operated valve on each

residual heat removal discharge line to protect the system from the

reactor coolant system pressure during operation. The Watts Bar design

features permit leak testing of each check valve separately during plant

operation to fulfill the staff requirements for high/low pressure isola--

tion with two check valves. The leak testing program for these valves is

discussed in Section 3.9.6 of this report.

The design provisions and staff requirements noted above (see Section 3.9.6)

satisfy the staff requirements for system isolation as specified in BTP RSB

5-1. Provisions for detecting leakage into the RHR system are discussed in

Section 5.2.7 of this report.

The planned preoperational and startup test program provides for demonstrating

the operation of the RHR system. Conformance with Regulatory Guide 1.68 is

Watts Bar SER 5-18
discussed in Chapter 14. Any additional testing requirements which result from

the staff review of the applicant's compliance with RSB 5-1 are discussed below..

The RHR system is housed within a structure that is designed to-withstand

tornadoes, floods, and seismic phenomena in accordance with GDC 2, as discussed

in Section 3.0. The system seismic requirements (Regulatory Guide 1.29) and

quality standards (10 CFR 50.55a and Regulatory Guide 1.26) are discussed in

Chapter 3.

The RHR system capability to withstand pipe whip inside containment, as

required by GDC 4 and Regulatory 1.46, is discussed in Section 3.6. Protection

against piping failures outside of containment in accordance with GDC 4 is

discussed in Section 3.6. Item III.D.1.1 of NUREG-0737 concerning leakage

reduction of primary coolant from systems outside containment is discussed in

Section 11.7.2.

As noted above, the RHR system serves both during normal shutdown cooling, and

emergency low pressure cooling as part of the emergency core cooling system

(ECCS). However, both functions are mutually exclusive, because the*RHR system

is aligned for ECC except for normal cooldown below reactor coolant conditions

of 350*F and 425 psig. When the RHR system is alignedfor normal shutdown

cooling, the suction paths from the refueling water storage tank (RWST) are

closed and the suction paths from the hot legs are opened. When the RHR system

is aligned for ECC operation the suction paths from the RWST are kept open, and

the two suction paths (from one hot.leg) are isolated each by two

motor-operated valves in series. A separate residual heat removal system is

provided for each unit, thus satisfying GDC 5. Compliance with the intent of

Regulatory Guide 1.1 concerning NPSH is discussed in Section 6.3.

All RHR lines, including instrument lines, have containment isolation features;

their satisfaction of the requirements of GDC 56 and 57 and Regulatory

Guide 1.11 is discussed in Section 6.2.

The staff asked the applicant to address the capability of the RHR system to

meet BTP RSB 5-1, "Design Requirements for the Residual Heat Removal System."

The applicant has shown that the Watts Bar facility, under normal conditions,

can be brought to cold shutdown by the operator performing actions in the

control room (complying with the applicable portions of GDC 19).

Four processes are involved in taking the plant from hot standby to cold shutdown conditions. These are: (1) removal of residual heat and stored energy;

(2) circulation of the reactor coolant; (3) boration of the reaction coolant

to the cold shutdown boron concentration and coolant makeup; and (4) depressurization. With loss of offsite power, the reactor coolant pumps, main

condenser, and the main feedwater pumps are unavailable. Heat removal and

coolant circulation under natural circulation conditions is then controlled by

use of the steam generator atmospheric dump valves and the auxiliary feedwater

system.

The four air-operated atmospheric dump valves at Watts Bar (one per steam

generator) are seismic Category I. Electrical power is obtained from the 125-V

vital battery system. The most limiting single failure would be loss of air

Watts Bar SER 5-19
supply or power, which prevents actuation of dump valves from the main control

room. The valves could be operated by manual action (outside of containment)

to correct for this single failure. Because this is a control function, the

applicant has committed to perform tests to confirm the feasibility of.this

type of manual action. The staff finds this acceptable pending verification

that the dump values function properly. Mechanical failure could prevent

opening of a single dump valve. Manual action to correct for this failure

would involve closing of an upstream isolation valve and replacement or repair

of the dump valve. Alternatively, natural circulation tests may justify plant

cooldown with one failed dump valve.

The water supply to the auxiliary feedwater system is provided initially from

the nonseismic condensate storage tank. This supply is backed up by the

seismic Category I emergency raw cooling water system (ERCWS). The supply is

transferred automatically to the ERCWs via fully qualified automatic admission

valves to maintain adequate net positive suction head at the auxiliary

feedwater pumps.

During a normal plant cooldown from hot standby conditions, the Chemical and

Volume Control System (CVCS) letdown line from the reactor coolant system (RCS)

would be used during the initial boration to the required boron shutdown concentration and while the RCS inventory is controlled during the cooldown.

Loss of the nonseismic air supply results in loss of letdown as a result of the

air-operated valves failing closed in the letdown line. The CVCS makeup

control system also could be unavailable because of the loss of nonsafety-grade

control circuits of the nonseismic air supply. Under these conditions,

boration without letdown could still be accomplished using safety grade

equipment. Borated water (12 w/o boric acid) could be supplied to the suction

of the centrifugal charging pumps from one of the three boric acid tanks using

one of the two boric acid transfer pumps. The tanks, pumps, and associated

piping are seismic Category I. Thecapacity of one boric acid tank is

sufficient to provide boration to the required shutdown concentration. Makeup

above that provided by the boric acid tanks is obtained from the refueling

water storage tank. Borated water from the centrifugal charging pumps can be

supplied to the RCS via the normal charging, and reactor coolant pump seal

injection flow paths or via the boron injection tank (12 w/o boric acid) path.

The effect of valve failures due either to loss of air supply or postulated

single failure is mitigated either by manual actions to correct the failure or

use of an alternate injection path.

Calculations, based on the staff review of the applicant's injection of borated

water with 12 w/o boric acid, indicate that the available volume in the pressurizer steam space is greater than thatneeded to achieve a cold shutdown

boron concentration in the RCS without taking full credit for-letdown or

contraction of the primary coolant in cooldown. In addition, the available

volume for borated water injection without letdown, which results from the

contraction of the primary coolant, is much larger than that required to cool

and, hence, depressurize the pressurizer to 425 psig by injection of borated

water through the pressurizer spray. This pressure must be reached to permit

shutdown cooling with the RHR system.

Under natural circulation conditions, the normal supply for the pressurizer

spray from the cold legs of two, coolant loops is lost. In this case, the

Watts Bar SER 5-20
pressurizer spray can be supplied by flow from the centrifugal charging pumps

through a line branching off from the charging pumps of the CVCS. This supply

could be lost by a single failure involving either closing of a single valve in

the supply line or opening of one of several valves in lines connected to the

supply line. If manual actions to correct for such failures were not

successful, a backup method of depressurization would involve opening either of

the two seismic Category I power-operated relief valves of the pressurizer that

discharge to the pressurizer relief tank. The pressurizer relief tank is not

designed for continuous operation and does not have safety grade equipment to

provide for intermittent operation. Hence, these actions might result in

rupture of the tank rupture disc and a release to containment. The isolation

valves (and their bypass valves) in the suction line from the RCS to the RHR

pumps, which must be opened to achieve long-term cooling with the RHR system,

are designed to withstand the environmental conditions following a steamline

break inside containment and, thus, would be qualified to withstand the less

severe conditions resulting from this type of operation.

BTP RSB 5-1 requires that a natural circulation test with supporting analysis

be conducted to demonstrate the ability to cool down and depressurize the plant

and to demonstrate that boron mixing is sufficient under such circumstances.

Comparison with performance of previously tested plants of similar design may

be substituted for these tests, if justified. The applicant intends to

reference tests to be conducted at Diablo Canyon nuclear plant to meet this

requirement. The applicant has referenced a preliminary assessment of a

comparison between Sequoyah and Diablo Canyon plants that might affect boron

mixing under natural circulation and has included a preliminary comparison of

Watts Bar versus Sequoyah. These assessments indicate that the results of the

Diablo Canyon tests and supporting analyses are expected to satisfy BTP RSB 5-1

requirements for Watts Bar. However, the staff cannot reach a conclusion on

this matter until the Diablo Canyon results have been reviewed and their

applicability to Watts Bar evaluated and confirmed. If the Diablo Canyon

results have been reviewed and their applicability to Watts Bar evaluated. If

the Diablo Canyon tests are not completed or do not provide satisfactory

results, the applicant will be required to submit such test results applicable

to the Watts Bar design before startup after the first refueling. This testing

is not necessary for first-cycle operation of Watts Bar because the major

purpose of the natural circulation test is to obtain information on the time

needed to take the plant from hot standby to the cut-in point of the RHR system

under conditions such as extended loss-of-offsite power when the RCS pumps are

not available. It would be preferable to run this test after the reload when

the decay heat is relatively large. This would result in more meaningful test

data and testing under conditions more representative of those occurring over

the 40-year plant life.

The Watts Bar design has bypass lines around the RHR suction isolation valves.

Valves in these lines have the same qualification as the "normal" isolation

valves that they bypass and have the same closure interlocks. Because this

bypass feature eliminates the suction isolation valve single failure (closed),

which could prevent availability of the RHR system, this is acceptable to the

staff.

The staff concludes that with satisfaction of the natural circulation test

requirement (discussed above), Watts Bar meets current requirements for Class 2

Watts Bar SER 5-21
plants specified in RSB 5-1 and GDC 34. The design for Watts Bar to achieve

cold shutdown is, therefore, acceptable.

The staff is conducting generic studies of the required reliability and

capability of shutdown heat removal systems under Item II.E.3.2 of NUREG-0660.

Another staff study described in Item II.E.3.3 of NUREG-0660 involves the

desirability and possible requirements for diverse heat removal systems. The

Watts Bar facility must comply with the results of the staff's generic review

of these items when completed.

As discussed above, the staff finds that the RHR system has suitable redundancy

in components and features. Suitable interconnections and isolation capabilities are provided to ensure that with either onsite or offsite power the RHR

system can perform its normal safety function, assuming a single failure, as

required GDC 34, and with operator actions only from the control room under

normal conditions as required by GDC 19.

5.4.4 Pressurizer Relief Tank

The nonsafety-related (quality group D, nonseismic) pressurizer relief tank

(PRT), located inside containment, is designed to receive and condense normal

discharges of steam from the primary system safety and relief valves up to

110 percent of the pressurizer steam volume above the full power water level

set point without a release to the containment atmosphere.

The tank normally contains water and a predominantly nitrogen atmosphere.

Steam is discharged through a sparger pipe under the water level where condensing and mixing occurs. The tank is equipped with an internal spray and drain

used to coo] the tank following a discharge. Tank overpressurization protection is provided by two ruptured discs, whose release capacity is equal to or

greater than the combined relief capacities of the pressurizer safety valves.

Level, pressure, and temperature indicators and alarms are provided in the

control room to inform the operator of the PRT condition. Failure of the

PRT does not affect the integrity of the reactor coolant pressure boundary nor

does it affect the capability to shut down the plant safely. Thus the requirements of GDC 2 and the guidelines of Regulatory Guide 1.26 and position C.2 of

Regulatory Guide 1.29 are met. The PRT is, therefore, acceptable.

5.4.5 Reactor Coolant System Vents (II.B.1)

NUREG-0737 requires installation of reactor coolant system and reactor vessel

head high point vents that are remotely operated from the central room.

In a submittal dated September 14, 1981, the applicant committed to providing

the same RCS vent system for Watts Bar as is provided for Sequoyah (NRC Staff

approval documented in Sequoyah SER Supplement 5, NUREG-0111, June, 1981). The

vent system will be installed before fuel loading, and the applicant has

committed using the to venting guidelines being developed by the Westinghouse

Owners Group. The staff finds this commitment acceptable and will require modification of Watts Bar venting guidelines if they are indicated by the staff

generic review.

Watts Bar SER 5-22
6 ENGINEERED SAFETY FEATURES.

6.1 Engineered Safety Features Material

Engineered safety features (ESFs) material was reviewed in accordance with the

SRP (NUREG-0800), on the basis of information contained in FSAR Section 6.1.1,

6.1.2, and 6.1.3 as well as responses to requests for additional information.

6.1.1 Metallic Materials

GDC 1, 14, and 31, and 10 CFR 50.55a have been met with respect to ensuring an

extremely low probability of leakage, rapidly propagating failure, and gross

rupture. This is shown since the materials selected for the ESFs satisfy

Appendix I of Section III of the ASME Code, Parts A, B, and C of Section II of

the Code, and the staff position that the yield strength of cold-worked stainless steels'shall be less than 90,000 psi.

The sections of the ASME Code applicable to this plant allowed waiving fracture

toughness testing of ferritic materials in Class 2 and 3 systems. However,

based on results of impact testing by (1) other applicants of the same specification steels and (2) fracture toughness data presented in NUREG-0577, the staff

concludes that the fracture-toughness properties of the ESF ferritic materials

would satisfy the fracture toughness requirements of the current version of the

Code.

The controls on the use and fabrication of the austenitic stainless steel of

the systems met most of the requirements of Regulatory Guides 1.31 and 1.44.

Where the recommendations of these guides were not met, the actions taken by

the applicant have been approved by the staff. The fabrication and heat treatment practices performed provide added assurance that the probability of stress

corrosion cracking will be reduced during the postulated accident time interval.

Conformance with the ASME Code and Regulatory Guides, and the staff positions

mentioned above constitutes an acceptable basis for meeting the requirements of

GDC 1, 4, 14, 35, and 41; Appendix B to 10 CFR 50; and 10 CFR 50.55a under which

the systems are to be designed, fabricated, and erected so that the systems can

perform their functions as required. GDC 1, 14, and 31 and Appendix B to

10 CFR 50 have been met with respect to ensuring that the reactor coolant boundary and associated auxiliary systems have an extremely low probability of leakage, rapidly propagating failures, and gross rupture. The controls placed on

concentrations of leachable impurities in nonmetallic thermal insulation used

on components of the ESFs are in accordance with the requirements of Regulatory

Guide 1.36.

The protective coating systems have been qualified by tests acceptable to the

staff. This qualification provides reasonable assurance that the coating systems will not degrade the operation of the ESF by delaminating, flaking, or

peeling.

Watts Bar SER 6-1
The coatings applied are inhaccordance with Regulatory Guide 1.54, "Quality

Assurance Requirements for Protective Coatings Applied to Water-Cooled Nuclear

Power Plants."

Conformance with this Regulatory Guide provides a basis for meeting the requirements of Appendix B to 10 CFR 50.

The requirements of GDC 4, 35, and 41 and Appendix B, 10 CFR 50, have been met

with respect to compatibility of ESF components with environmental conditions

associated with normal operation, maintenance, testing, and postulated accidents.

This includes LOCAs because the controls on the pH and chemistry of the reactor

containment sprays and the emergency core cooling water following a LOCA or

design-basis accident (DBA) are adequate to reduce the probability of stress

corrosion cracking of the austenitic stainless steel components and welds of

the ESF systems in containment throughout the duration of the postulated accident to completion of cleanup.

Also, the control of the pH of the sprays and cooling water, in conjunction

with controls on selection of containment materials, is in accordance with Regulatory Guide 1.7, "Control of Combustible Gas Concentrations in Containment

Following a Loss-of-Coolant Accident," and provides assurance that the sprays

and cooling water will not give rise to excessive hydrogen gas evolution resulting from corrosion of containment metal or cause serious deterioration of the

materials in containment.

The controls committed to be placed on component and system cleaning are in

accordance with Regulatory Guide 1.37 and provide a basis for the finding that

the components and systems have been protected against damage or deterioration

by contaminants as stated in the cleaning requirements of Appendix B, 10 CFR 50.

6.1.2 Organic Materials

This evaluation is conducted to verify that protective coatings applied inside

containment meet the testing requirements of ANSI N101.2, "Protective Coatings

(Paints) for Light Water Nuclear Reactor Containment Facilities," 1972, and the

guidelines of Regulatory Guide 1.54.

Compliance with these requirements provides assurance that the protective coatings will not fail under DBA conditions and will not generate significant quantities of solid debris or combustible gas that could adversely affect the engineered safety features.

The applicant has committed to use protective coating systems inside containment that meet the testing requirements of ANSI N101.2. The applicant stated

that the protective coatings will meet the quality assurance (QA) guidelines

of Regulatory Guide 1.54 with proposed alternatives to ANSI N101.4, "Quality

Assurance for Protective Coatings Applied to Nuclear Facilities," 1972: The

proposed alternatives to ANSI N101.4 were developed by TVA for Watts Bar and

other nuclear projects.

The proposed alternatives to.ANSI N101.4 include (1) assigning to TVA the

responsibility of specifying testing for the quality control of the coating

materials, the manufacturing process, the applicators and surface preparation,

and inspection and (2) deleting nonmandatory QA requirements of ANSI N1O1.4.

Watts Bar SER 6-2
The staff has reviewed the proposed alternatives and found them acceptable, as

indicated in NRC/TVA meeting summary dated April 14, 1978.

The staff concludes that the protective coating systems and their applications

are acceptable and meet the requirements of Appendix B to 10 CFR 50. This conclusion is based on the coating systems and their applications meeting (1) the

positions of Regulatory Guide 1.54, with an acceptable alternate to ANSI N101.4

(1972) and (2) the testing requirements of ANSI N101.2. The coating systems

chosen by the applicant have been qualified under conditions which take into

account the postulated DBA conditions.

The control of combustible gases that can potentially be generated from the

organic materials and from qualified and unqualified paints is reviewed in

Section 6.2.5 of this report. The consequences of solid debris that can potentially be formed from unqualified paints are reviewed in Section 6.2.2.

6.1.3 Postaccident Emergency Cooling Water Chemistry

This review is related to providing and maintaining the proper pH of the containment sump water and recirculated containment spray water following a DBA

to reduce the likelihood of stress corrosion cracking of austenitic stainless

steel.

The applicant will use borated water (concentration >2000 ppm boron) from the

refueling water storage tank during the injection phase of containment spray.

The borated water from the containment spray drains to the sump. The ice in

the ice condenser contains sodium tetraborate equivalent to a concentration of

2000 ppm as boron. The melted ice also drains to the containment sump and,

after mixing, will raise the sump water pH to >7.0. Mixing is achieved as the

solution is continuously recirculated from the sump to.the containment spray

nozzles during the recirculation phase of containment spray.

The staff evaluated the pH of the water (mixture of refueling water storage

tank and ice condenser borated water) in the containment sump. The staff verified by independent calculations that sufficient sodium tetraborate is available to raise the containment sump water pH above the minimum 7.0 level to

reduce the probability of stress corrosion cracking of austenitic stainless

steel components. The staff will include surveillance requirements in the

plant Technical Specifications to verify that sufficient sodium tetraborate is

contained in the ice condenser.

On the basis of the above evaluation, the staff concludes that the postaccident

emergency cooling water chemistry meets the minimum pH acceptance criterion-of

SRP Section 6.1.1, the positions of BTP MTEB 6-1, and the requirements of GDC 14.

It is, therefore, acceptable.

6.2 Containment Systems

The containment systems for Watts Bar Units 1 and 2 include dual containment

structures, containment heat removal systems, containment isolation systems,

containment combustible gas control systems, secondary containment (annulus)

emergency gas treatment systems, and the auxiliary building gas treatment

system. The plant will utilize an ice condenser-type pressure suppression

containment similar to that of the Donald C. Cook and the Sequoyah plants.

Watts Bar SER 6-3
The staff has reviewed the applicant's design, design criteria, and design

bases for containment systems. The acceptance criteria used as the basis for

the staff evaluation are set forth in certain sections of the SRP, NUREG-0800.

Specifically, the Watts Bar application was reviewed using SRP Sections 6.2.1,

"Containment Functional Design"; 6.2.2, "Containment Heat Removal Systems";

6.2.3, "Secondary Containment Functional Design"; 6.2.4. "Containment Isolation

System"; 6.2.5, "Combustible Gas Control in Containment"; and 6.2.6, "Containment Leakage Testing." These acceptance criteria include the applicable GDC,

Regulatory Guides, Branch Technical Positions, and industry codes and standards

as specified in the above cited sections of the SRP. The results of-the staff's

review are discussed below.

The design of the Watts Bar plant containment is very similar to the containment design for the Sequoyah Nuclear Plant, which has been previously reviewed

by the staff. Both plants utilize the dual containment concept, with a freestanding steel primary containment within a reinforced concrete shield building.

Volumes and plant arrangement within the primary containment are nearly'identical. Basic differences between the Watts Bar and Sequoyah facilities are slight

and are limited to the design of the reactor coolant system, containment spray

system, containment design pressure, and containment internal structures.

Table 6.1 indicates the differences in principal containment parameters between

the two plants. These differences have been evaluated and the results of the

evaluation are addressed in the appropriate sections of this report.

The primary containment, which has a net free volume of about 1,191,000 ft 3 , is

divided into three major subvolumes, including a 383,000-ft 3 compartment enclosing the reactor system, a 110,400-ft 3 ice condenser compartment enclosing

the ice condenser, and a 698,000-ft 3 upper compartment.

The basic performance and design evaluation of the ice condenser system has

been the subject of both analyses and experimental programs. These efforts are

described in the "Staff Evaluation of Tests Conducted-to Demonstrate the Functional Adequacy of the Ice Condenser Design," dated April 25, 1974, and provide

the basis for the staff's evaluation of the containment functional design.

6.2.1 Containment Functional Design

6.2.1.1 Containment Structure

The containment for each unit of the Watts Bar facility consists of a primary

containment vessel and a shield building, and the common auxiliary building.

The primary containment vessel is a free-standing, welded steel structure consisting of.a vertical cylinder, a hemispherical dome, and a concrete base mat

with steel membrane. The shield building is a medium-leakage concrete structure

enclosing the containmment vessel and is designed to provide for the collection,

mixing,'.holdup, and controlled release of containment vessel fission product

leakage following an accident. The interior of the primary containment vessel

is divided into three compartments: (1) a lower compartment, which houses the

reactor and reactor coolant system; (2) the ice condenser compartment, which

houses the energy-absorbing ice bed in which steam is condensed; and (3) the

upper compartment, which accommodates the air displaced from the other two

volumes during-a postulated LOCA and steamline break accident.

Watts Bar SER 6-4
Table 6.1 Containment design parameters

Watts Bar Sequoyah*

Reactor containment volumes (net free volume)

Upper compartment (ft 3 )

Ice condenser (ft 3 )

Lower compartment (ft3 )

Total containment volume (ft 3 )

Reactor containment air compression ratio

ECCS reactor power used for containment

analysis (MWt)

Design energy release to containment

Initial blowdown mass release (Ib)

Initial blowdown energy release (Btu)

Ice condenser parameters

Weight of ice in condenser (lb)

Containment return air fan flow (ft 3/min)

Containment spray flow (LOCA analysis)

One spray train inoperable

Upper compartment (gpm)

Lower compartment (gpm)

One RHR pump inoperable

Upper compartment (gpm)

Lower compartment (gpm)

Total spray (gpm)

Containment design pressure (psig)

*These values are taken from the Sequoyah SER.

698,000

110,400

383,000

1,191,400

1.43

698,000 110,500

383,000

1,191,500

1.43

3,579 3,582

493,000 318 x 106

2.45 x 106

40,000

4,000

0

2,000

0

6,000

15.0

543,330 334.6 x 106

2.45 x 106

40,000

4,7500

2,000

0

6,750

12.0

Watts Bar SER 6-5•
The intermediate, or ice condenser compartment, is an enclosed annular compartment encompassing most of the perimeter of the containment structure. Borated

flake ice is stored within the ice condenser compartment in 48-ft-long cylindrical perforated metal baskets. The ice contained in the baskets is provided

to condense the steam released in the event of a LOCA or a steamline break

accident. The ice condenser system at Watts Bar is essentially identical to

that at Sequoyah.

The staff will require the applicant to weigh the ice in a large statistical

sample (approximately 50 percent) of the 1944 ice baskets in each unit, following their initial ice loading. The staff will require that this information be

used in statistical analyses to determine (1) the initial distribution of ice

in the ice condenser, (2) the minimum amount of ice loaded into the ice condenser at a 95-percent level of confidence, and (3) appropriate subdivision of the

ice condenser into groups of bays to be utilized in the periodic ice weight

surveillance program. The matters will be resolved in conjunction with the

staff's development of the Technical Specifications.

In an effort to provide the earliestpossible indication of the actual sublimation rate for the D. C. Cook Unit 1 ice condenser, its licensee, the American

Electric Power Company, implemented a program to measure periodically the weight

of selected ice baskets in the ice condenser, and has weighed a sample of ice

baskets on numerous occasions. The results of the ice basket weighing program

have indicated that the average sublimation rate of 2 to 3 percent per year is

significantly greater than the expected rate of about 0.5 percent per year, and

slightly greater than the maximum design sublimation rate of 2 percent per year.

The results also have shown that ice sublimation does not occur uniformly over

the cross-sectional area of the ice condenser. Baskets adjacent to the crane

and containment wall cooling ducts lose ice at a greater rate than baskets

located in the interior of the ice condenser. Interpretation of the data from

the ice basket weighing program has been complicated by the fact that variations

in original ice loading techniques resulted in three distinct groups of ice

baskets within the ice condenser having significantly different mean basket

weights. The frequent weighing programs conducted at the D. C. Cook facility

have provided early identification of the ice condenser loss rates and patterns,

the opportunity to develop corrective modifications and procedures, and has

ensured the safety of continued operation of the plant.

The ice weighing program at Sequoyah Nuclear Plant Unit 1 has detected an even

greater rate of ice loss, on the order of 4 to 5 percent per year. This high

loss rate necessitated, during the first fuel cycle, a reanalysis of the DBAs

to-permit a reduction in the requirement in the plant Technical Specifications

for total ice weight.

Based on the above discussion of current ice condenser operating experience,

the staff will require the applicant to institute a periodic ice basket

weighing program for each unit at Watts Bar, similar to the program being conducted at the Sequoyah plant. The staff has recommended that the applicant

continue to evaluate the equipment and techniques available for ice loading to

achieve an initial ice inventory that is uniformly distributed. The staff will

pursue the development of a suitable periodic ice weighing program with the

applicant during the development of Technical Specifications for the operation

of the plant and will include appropriate operating limits to ensure an acceptable margin of safety.

Watts Bar SER 6-6
During normal plant operation, the ice bed is maintained at about 15*F by a

redundant refrigeration system. Refrigeration ducts and insulation on the ice

condenser walls serve to minimize heat introduction to the ice. Thirty chiller

units are provided in the containment but only twenty-one of the units are

required to operate at any time to maintain the design temperature of 15'F

within the ice bed. In the unlikely event that a complete loss of the refrigeration system occurs, the insulation within the ice condenser is sufficient to

prevent substantial melting of the ice for a minimum period of 7 days, which

allows adequate time for safe plant shutdown.

Inlet and outlet doors are provided at the bottom and top of the ice condenser

compartment. In the event of a LOCA, the lower inlet doors will open as a

result of the pressure rise in the lower compartment caused by the release of

the reactor coolant to the lower compartment. The differential pressure will

then cause air, entrained water, and steam to flow from the lower compartment

into the ice condenser. The resulting pressure rise, due principally to the

air mass in the ice condenser, will cause the doors at the top of the ice condenser to open and allow the air to flow from the ice condenser into the upper

compartment. Steam will be condensed as it contacts the ice contained in the

baskets in the ice condenser compartment; therefore, steam will not appear in

the upper compartment.

Complete steam condensation is ensured because of the ice mass and geometrical

arrangements of the ice columns. Developmental testing by Westinghouse has

confirmed this phenomemon. The staff's evaluation of the test programs was

completed in conjunction with the review of the D. C. Cook plant, and was

reported in the SER for that plant and in the report, "Staff Evaluation of

Tests Conducted To Demonstrate the Functional Adequacy of the Ice Condenser

Design," dated April 25, 1974.

An operating deck separates the upper and lower compartment and ensures that

steam and air flow resulting from a LOCA is directed through the-ice condenser

to the upper compartment rather than through uncontrolled bypass paths. Condensation of the steam in the ice limits the containment pressure to approximately 8.0 psig between the time reactor blowdown is complete and the time that

meltout of the ice bed occurs. Ice meltout is predicted to occur about 69

minutes after onset of a design-basis LOCA. Following ice meltout, the rise in

the containment pressure as a result of the release of decay energy from the

core is limited by the containment spray system.

The lower compartment is divided into a number of subcompartments formed by

internal equipment, structures, and components. The pressure responses within

these subcompartments were analyzed by the applicant using the transient mass

distribution (TMD) computer code developed by Westinghouse. The code is

described in a nonproprietary Westinghouse Topical Report, "Ice Condenser Containment Pressure Transient Analysis Methods," WCAP-8078. The code provides

a means for computing pressures, temperatures, heat transfer rates, and mass

flow rates as functions of time and location throughout the containment. The

staff reviewed the mathematical description of this code during the review of

the D. C. Cook plant and found it to be acceptable for calculating the shortterm pressure response in subcompartments. The pressure response within the

subcompartments is different from the overall pressure response of the containment only during the early blowdown phase of the accident, that is, up to about

10 seconds following the occurrence of the break.

Watts Bar SER 6-7
Following the early blowdown phase of the accident, the pressure and temperature

responses of the upper and lower compartments are analyzed with the Westinghouse

LOTIC-1 computer program. This program has been described in Westinghouse

Topical Report WCAP-8354, "Long Term Ice Condenser Containmment Code - LOTIC

Code." The staff has completed a generic review of the LOTIC-1 computer program

through the NRC's topical report evaluation program and has concluded that the

.LOTIC-1 code is acceptable for the calculation of the long-term ice condenser

containment response to postulated LOCAs (see NRC letter to Westinghouse dated

May 3, 1978).

Maximum Pressure and Temperature Analysis

The applicant has performed containment analyses on a spectrum of reactor

coolant system (RCS) and secondary system pipe ruptures to verify the containment functional design pressure and temperature and to establish the pressure

and temperature conditions for environmental qualification of safety-related

equipment located inside containment. The containment functional analyses

include the assumption of the most limiting single active failure and the

availability or unavailability of offsite power, depending on which results in

the highest containment temperatures and pressures.

The applicant used the LOTIC-1 computer program to calculate the long-term containment pressure and temperature response to a spectrum of LOCAs. LOTIC-1 is

a computer program similar to the COCO code, which has been used to analyze the

containment pressure transients for other types of containments. The main differences between these computer codes lie in the methods by which the heat

removal systems are modeled. LOTIC-1 includes features for modeling the heat

removal capabilities of the ice and has-provisions to calculate the pressure

response of the containment. The containment upper and lower compartments and

the ice condenser are modeled as control volumes in the code to represent the

physical geometry of the containment. Conservation of mass and energy are

applied and equations are solved by appropriate numerical procedures.

The applicant has provided the mass and energy release data for a spectrum of

LOCA break sizes and locations in the RCS. The break locations include the

cold- leg piping at the suction and discharge sides of the primary coolant pump

and the hot leg (see Table 6.2.1-15 in the Watts Bar FSAR). The effects of.

single failures on the mass and energy release rates were included by bounding

the possible effects with two cases, that is, the maximum safeguards case where

no single failures are assumed and the minimum safeguards case where the single

failure assumed is the loss of one emergency diesel. The staff has reviewed

the applicant's spectrum of breaks and the single failures considered and finds

them acceptable.

The mass and energy released to the containment are considered in terms of the

blowdown, refill, reflood, and post-reflood phases. The method used by the

applicant to compute the mass and energy release rates from postulated reactor

coolant pipe breaks for the containment functional analyses is documented in

Topical Report WCAP-8312A, "Westinghouse Mass and Energy Release Data for Containment Design," which was approved by the staff in a letter dated March 12,

1975. Therefore, the applicant's mass and energy release rate data for postulated reactor coolant pipe breaks are acceptable for use in the containment

functional analyses.

Watts Bar SER 6-8.
The staff has also reviewed the plant containment parameters and has found them

to be conservative for the evaluation of the long-term containment response to

postulated LOCAs. Using the above mass and energy release rate data, containment input parameters, and the LOTIC-1 ice condenser containment analysis program, the applicant has calculated a peak containment pressure of 12.3 psig and

a peak containment temperature of 230°F for the worst-case LOCA, a double-ended

rupture of the RCS pump suction cold leg. This is also the limiting break for

the Sequoyah plant, and the analysis and results for LOCAs at Watts Bar and

Sequoyah are very similar. Because the peak calculated containment pressure of

12.3 psig is less than the 15.0-psig containment design pressure, the staff

finds the applicant's long-term containment response calculations for LOCAs

acceptable.

The applicant also has analyzed the effect of steam bypassing the ice condenser

on the containment pressure response. Drain lines in the floor of the refueling canal are provided to allow water sprayed into the upper compartment to

return to the containment sump. These drains represent a bypass path. The

applicant has included in the containment analysis the effect of this bypass

area (2.2 ft 2 ). The applicant has also provided analyses which indicate that

more than 40 ft 2 of bypass area can be accommodated in the design without the

design pressure of the containment being exceeded.

The staff has reviewed the applicant's analysis of the maximum differential

pressures which could exist in the reverse direction (that is, upper compartment

to lower compartment) during a LOCA. The applicant's methods of analysis and

assumptions are conservative. The applicant has calculated a maximum reverse

differential pressure of 1.05 psig. The design reverse differential pressures

*of the operating deck and ice condenser lower inlet doors are 9.6 and 8.6 psig,

respectively.

The applicant has calculated the containment response to a spectrum of main

steamline breaks (MSLBs) using the LOTIC-3 computer program. This program has

been described in Supplement 2 to the Westinghouse Topical Report WCAP-8354.

The staff has completed a generic review of the LOTIC-3 code and has concluded

that the LOTIC-3 code is acceptable for the calculation of long-term ice condenser containment response to postulated secondary system pipe break accidents

(see NRC letter to Westinghouse dated May 3, 1978). The applicant also has

presented information to show that the calculated temperature transient inside

the Watts Bar containment following a postulated small MSLB accident is conservatively predicted by the analyses presented in Supplement 2 to WCAP-8354.

These analyses were performed for a "generic" ice condenser plant using the

LOTIC-3 computer code to demonstrate the adequacy of the code for ice condenser

long-term transient analyses for secondary system ruptures. At the staff's

request, additional small MSLBs were analyzed, extending the spectrum down to a

0.1-ft2 break size.

The MSLB analyses for both the Watts Bar and the Sequoyah plants were done

using Watts Bar input parameters. The Watts Bar parameters are more conservative than the Sequoyah parameters, and the generic plant parameters are more

conservative than either the Watts Bar or Sequoyah parameters. Thus, identical

analyses were done for both Watts Bar and Sequoyah. Therefore, the staff finds

the Watts Bar MSLB analysis acceptable on the basis that an identical-analysis

has already been found acceptable for the Sequoyah plant.

Watts Bar SER 6-9 ̧
The maximum calculated containment atmosphere temperature occurs in the lower

compartment because all steam is effectively removed in the ice condenser.. The

maximum calculated lower compartment temperature and pressure for the generic

plant are 327°F and 4.5 psig, and results from a postulated small-split break

of about 0.6-ft 2 area at 30-percent power level with the assumed failure of the

auxiliary feedwater runout protection system. The results of analyses of the

spectrum of breaks smaller than this break have shown that the peak calculated

temperatures are slightly less for these breaks but that superheated conditions

in the containment lower compartment are slightly prolonged.

The staff concludes that the results of the LOTIC-3 analyses for the generic

ice condenser plant result in containment temperatures more conservative than

those which would be calculated specifically for the Watts Bar plant and are,

therefore, acceptable for use on Watts Bar. The staff will use the results of

the spectrum of steamline breaks to assess the equipment qualification tests

performed on those instruments and equipment located in the containment lower

compartment that are required to perform safety functions and monitor the course

of the accident.

Based on the above review of the applicant's containment pressure and temperature functional analyses, the staff concludes that the applicant has satisfactorily demonstrated the adequacy of the containment functional design following

onset of a LOCA or MSLB.

Protection Against Damage from External Pressure

The containment vessel is designed for an external pressure of 2.0 psig.

Inadvertent operation of the spray system and/or return air fan systems would

cause a reduction in the containment pressure. The applicant has stated that

such transients would not cause containment external pressure to exceed the

design value of 2.0 psig. The staff has requested, and the applicant has

agreed to furnish for staff review the analyses performed to support the applicant's statement of design adequacy. This information has been submitted by

letter dated June 4, 1982 and currently is under review by the staff. A simplified analysis by the staff indicates that the containment external design pressure value would not be exceeded by the transients mentioned above., Therefore,

the staff considers this to be a confirmatory item and will confirm in a supplement to this SER the contention that containment external pressure transients

will not exceed the design value of 2.0 psig.

Containment Instrumentation

The Watts Bar design includes instrumentation to monitor containment pressure

and containment water level, as required by NUREG-0737, Item II.F.1, Attachments 4, "Containment Pressure Monitor," and 5, "Containment Water Level

Monitor."

Four indications of the containment pressure are provided in the main control

room. The existing pressure indicators have a range of -1 to 15 psig. The

applicant has stated that redundant, continuous containment pressure indication with a range up to four times the design pressure of the containment will

be provided by the fuel load date. In accordance with the implementation-schedule of NUREG-0737, the staff will condition the license to require acceptable

Watts Bar SER 6-10
containment pressure monitors to be installed and operational at Watts Bar

before the full-power operating license is issued.

Watts Bar has a containment water level monitoring system that is the same as

the one installed at Sequoyah; the monitors at both Watts Bar and Sequoyah satisfy the provisions of NUREG-0737, with the exception of their being environmentally qualified in accordance with Regulatory Guide 1.89, "Qualification

of Class IE Equipment for Nuclear Power Plants" (NUREG-0588). In accordance

with the implementation schedule of NUREG-0737, the staff will condition the

license to require acceptable containment water level monitors to be installed

and operational at Watts Bar before the full-power operating license is issued.

6.2.1.2 Subcompartment Analysis

Following the onset of a postulated reactor coolant pipe rupture, differential

and local pressures build up in the subcompartments of the lower containment

compartment as high-energy fluid is released and transported throughout the

various regions. The pressure magnitudes depend upon the volumes of the subcompartments, interconnecting vent flow paths, mass flow behavior, and the

thermodynamic behavior within the pressure nodes. During this phase of the

transient, flow to the upper containment compartment is rising but pressure is

still near its initial pressure. It is during this time that the peak operating deck differential pressure and peak subcompartment differential pressures would

be experienced. As the blowdown continues, the pressure in the upper compartment rises, and about 10 seconds after the start of blowdown the upper compartment reaches a peak pressure approximately equal to the lower compartment pressure, that is, about 8.0 psig. The primary factor in producing this upper

compartment peak is the displacement of air from the lower compartment through the ice columns into the upper compartment.

The LOCA short-term mass and energy release rate data used in the subcompartment

analyses were developed from computer programs described in WCAP-8312A (which

was approved by the staff in a letter dated March 12, 1975) and therefore, are

acceptable.

The applicant used the TMD computer program to calculate the short-term pressures, temperatures, heat transfer rates, and mass flow rates as a function of

time and location throughout the. containment, including the containment compartments, following either a LOCA or MSLB. The model includes a nodalization

scheme of 50 volumes representing the containment to analyze the pressure

response of each of the subcompartments within the lower compartment, including

dead-ended compartments, the ice condenser compartment, the upper compartment,

and the steam generator enclosures.

TMD was developed specifically to analyze the short-term pressure response of

the ice condenser system. The mathematical modeling in TMD is similar to that

of the SATAN-V blowdown code in that the analytical solution is developed by

considering the conservation equations of mass, momentum and energy, and the

equation of state,,and uses the control volume technique for simulating spatial

variation. The governing equations for TMD are somewhat different from those

in SATAN-V in that a two-phase (liquid water droplets and steam-air vapor),

two-component (air-water) system is considered. The staff has reviewed these

mathematical differences between SATAN-V and TMD and concurs that TMD has

maintained the conservatisms incorporated in SATAN-V. The TMD calculates the

Watts Bar SER 6-11
critical flow of a two-compartment, two-phase fluid (air, steam, and water)

assuming a thermal equilibrium condition. However, a correction factor, which

was determined by Westinghouse to account for experimental data on applicable

flow regimes, is then applied to the calculated critical flow. The correction

factor as used in the code increases the critical flow up to 20 percent through

the compartments as the quality of the fluid decreases. The applicant refers

to this increased critical flow as "augmented" flow. The net effect results in

a lower compartment differential pressure when compared to a nonaugmented flow

regime. The use of the augmented flow factor results in less conservatisms

than use of the thermal equilibrium correlation.

Following the staff's review of the experimental data and analysis performed

during the review of the D. C. Cook facility, the staff informed the applicant

that the use of this correction factor for the type of analysis being performed

could not be justified. In response to a staff request, theapplicant repeated

the short-term containment pressure response using the latest version of-the

TMD code with a nonaugmented or unity flow correlation.

The applicant's heat transfer model of the ice condenser used in the TMD code

was- based on the results of full-scale testing done by Westinghouse during 1968

and 1969. The test program used an ice basket design which had the equivalent

of an 81-percent basket open surface area (that is, exposed ice for condensing).

Structural problems with the initial basket design required a change in basket

design to accommodate the postulated loads. The redesign resulted in a decrease

in exposed surface area from the previous design. In 1973, Westinghouse reactivated the ice condenser full-scale test facility to provide final verification of the ice condenser functional performance with the redesigned ice basket

and final designs of all the ice condenser internal structures. Results of the

staff-review of the 1973-1974 full-scale. ice condenser tests have been presented

in an April 1974 report, "Staff Evaluation of Tests Conducted To Demonstrate

the Functional Adequacy of the Ice Condenser Design." The test program resulted

in a further redesign of the basket to achieve acceptable heat transfer between

steam and the ice in the baskets. A new heat transfer correlation was derived

for the TMD code to conservatively predict the ice condenser performance with

the new basket design. The latest version of the TMD code that was used for

the reanalysis of the short-term containment transient uses the new heat transfer correlation and a compressibility factor which is used with the subsonic

incompressible flow equations to include the effects of compressible fluid flow.

The applicant used the TMD code without the augmented critical flow correlation

to analyze the reactor cavity response to a LOCA. The reactor cavity was modeled

by 50 nodes within the cavity structures and 11 nodes external to the structures.

The annulus between the reactor vessel and the shield wall was broken into axial

and circumferential nodes. The applicant has calculated the reactor cavity

response to an assumed 127-in. 2 break of the reactor cold-leg pipe at the pressure vessel nozzle to pipe weld.

The applicant has also used the TMD code to analyze the response of the steam

generator enclosures to a double-ended steamline rupture and the pressurizer

enclosure to a double-ended rupture of the pressurizer spray line. The applicant has performed nodalization sensitivity studies and has developed a fivenode model for the steam generator enclosure and a four-node model for the

pressurizer enclosure. These models are similar to the models developed for

Watts Bar SER 6-12
the analyses of steam generator and pressurizer enclosures at Unit 2 of the

D. C. Cook facility, the McGuire facility, and the Sequoyah nuclear plant.

The following is a comparison of the subcompartment analyses performed at the

Watts Bar and Sequoyah plants.

Subcompartment pressure analyses were performed for-the lower compartment and

ice condenser, the reactor cavity, the steam generator enclosures, and the

pressurizer enclosure for both plants. The lower compartment and ice condenser,

reactor cavity, and pressurizer models use the same number of nodes at both

Sequoyah and Watts Bar. The differences in the values used for volumes and

flow areas are to account for geometric differences between the plants. These

geometric differences are most noticeable in the reactor cavity where design

changes were made to reduce the asymmetric pressures that were found at Sequoyah.

A steam generator enclosure at Sequoyah was modeled using nine nodes. The

Watts Bar model uses five nodes. The steamline break location in the enclosure

at Watts Bar is the nozzle at the top, the center of the steam generator. At

Sequoyah, in addition to the nozzle breaks, a break in the second elbow at one

side of the steam generator enclosure was required to be postulated. This

builtin asymmetry required a more detailed model.

All analyses were performed using the Westinghouse TMD code. Double-ended pipe

breaks were assumed to occur in all areas except the reactor cavity. The

reactor cavity break sizes were 100 and 127 in. 2 at Sequoyah and Watts Bar,

respectively, due to physical restraints limiting the maximum break size.

Results showed that for the lower compartment, the Watts Bar peak absolute and

differential pressures were slightly higher than those calculated for Sequoyah.

For breaks in the reactor cavity, the Watts Bar pressures are slightly lower.

The pressurizer results are the same since both used the Watts Bar model blowdown values. The Sequoyah steam generator enclosures show higher differential

pressures across the generators and lower differential pressures on the structures than were calculated for Watts Bar.

The calculated differential pressures acting on structures and components for

all subcompartments at Watts Bar do not exceed the corresponding design values.

Also, the applicant's results compare favorably with the results of the staff's

confirmatory analyses performed with the COMPARE-MOD 1 computer code. The Watts

Bar analysis of asymmetric loading due to differential pressures for the reactor

cavity is acceptable, based on comparison with the previously accepted Sequoyah

analysis, which is slightly less conservative than the Watts Bar analysis.

Based on the above, for the subcompartment analyses, the staff finds the applicant's method of analysis, modeling assumptions, and results to be acceptablefor the structural evaluation of the containment interior compartments and component supports.

6.2.1.3 Minimum Containment Pressure Analysis for Performance Capability

Studies on the ECCS

Appendix K to 10 CFR 50 of the Commission's regulations requires that the effect

of operation of all the pressure-reducing systems and processes installed in

the containment be included in the ECCS evaluation. For the purpose of this

evaluation, it is conservative to minimize the containment pressure. The

Watts Bar SER 6-13
reflood rate in the core will then be reduced because of the resistance to

steam flow in the reactor coolant loops.

Following a LOCA, the pressure in the containment building will be increased by

the addition of steam and water from the primary reactor system to the containment atmosphere. After initial blowdown, heat transfer from the core, primary

metal structure, and steam generators to the emergency core cooling water will

produce additional steam. This steam, together with any emergency core cooling

water spilled from the primary system, will flow through the postulated break

into the containment. This energy will be released to the containment duringboth the blowdown and later operational phases (that is, the reflood and postreflood phases).

Energy removal occurs within the containment by several means. Steam condensation on the containment walls and on internal structures serves as a passive

energy heat sink that becomes effective early in the blowdown transient. Subsequently, the operation of the containment heat removal systems such as containment sprays will remove steam from the containment atmosphere. In an ice

condenser-type containment, energy is removed as the mixture of steam, air,

and water, passes through the ice condenser (that is, when the mixture is forced

from the containment lower compartment to the upper compartment).

The ECCS containment pressure calculations for Watts Bar were done using the

Westinghouse ECCS evaluation model. The containment response calculations were

performed using the Westinghouse LOTIC-2 containment code. The staff has

reviewed the LOTIC-2 code and has concluded that the LOTIC-2 code is acceptable

for the calculation of minimum containment pressure response for ice condenser

plants. Although the staff has accepted the methods used to calculate containment pressure response, justification of the plant-dependent input parameters

used in the analysis of containment pressure response are required to be submitted for review on a plant-by-plant basis. This information was submitted

in the Watts Bar FSAR. The applicant has evaluated the containment net-free

volume, the passive heat sinks, operation of the Containment heat removal systems, and containmept initial conditions with regard to the conservatism for

the ECCS analysis. The containment heat removal systems were assumed to operate

at their maximum capacities, and minimum operational values for the spray water

and service water temperatures were assumed. The staff finds these assumptions

to be acceptable and in accordance with BTP CSB 6-1, "Minimum Containment Pressure Model for PWR ECCS Performance Evaluation." The staff also has found the

mass and energy release data used for this analysis to be acceptable.

The staff has concluded that the plant-dependent information used for the ECCS

containment pressure analysis for Watts Bar is reasonably conservative; therefore, the calculated containment pressures'are in accordance with the applicable

provisions of Appendix K to 10 CFR 50.

6.2.1.4 Summary and Conclusions

Based on the preceeding evaluations, and pending confirmation of the acceptability of the containment external pressure transient analysis and imposition

of the two license conditions concerning containment pressure and water level

monitors, the staff concludes that the containment functional design is acceptable and meets the requirements of GDC 16, 50, 52, and 53.

Watts Bar SER 6-14
6.2.2 Containment Heat Removal Systems

The energy released to the containment during a design-basis LOCA is absorbed

by the ice condenser. However, after the ice bed has melted, mass and energy

will continue to be released to the containment. Therefore, containment spray

systems are provided to maintain the containment pressure in the long term

below the containment design pressure and eventually reduce the containment

pressure to about atmospheric pressure.

Containment spray for the Watts Bar plant is provided by two redundant spray

systems. Each system is designed to provide the cooling capacity required to

maintain the peak pressure at less than design pressure for the full spectrum

of break sizes. Each of the redundant spray systems delivers 4000 gpm to the

containment from one containment spray pump and heat exchanger and 2000 gpm

from one residual heat removal pump and heat exchanger. The containment spray

pump is actuated by a containment high-high pressure signal set at approximately

3 psig; containment spray starts at about 30 seconds after the beginning of a

high-energy line break accident. Containment spray from the residual heat

removal (RHR) pump is manually initiated 1 hour after the beginning of the

accident.

The containment is equipped with redundant return air fan systems. Each of the

two 100-percent-capacity return air fan systems uses a 40,000-cfm fan to force

air from the upper compartment back to the lower compartment. This occurs.

after reactor coolant system blowdown and subsequent core reflooding are

completed.

The return air fans are used to return air from the upper compartment to the

lower compartment after the compression peak is reached, and thus ensure a

homogeneous mixture of'steam and air throughout the containment during the

long-term pressure peak. The return air fans are actuated by the containment

high-high pressure signal, but fan startup is delayed for 10 minutes to provide

an increased backpressure during core reflood.

The applicant has provided a malfunction analysis and other information which

demonstrate independence of the redundant spray trains and return air fan

systems. Each spray train has its own recirculation piping suction inlet from

a common sump. The sump is constructed with a grating in the inlet to prevent

the ingestion of debris. The spray nozzles are the limiting component in the

containment spray systems and are not subject to clogging by particles less

than 0.25 in. diameter. Spray droplet size at design conditions, spray pattern,

and header locations are acceptable. The applicant has provided analyses, which

the staff has reviewed, to ensure that adequate net positive suction head is

available at both the containent spray pump and RHR pump inlets during flow

from both the refueling water storage tank and the containment sump, without

taking credit for increased containment pressure following a LOCA. This is in

accordance with the guidance of Regulatory Guide 1.1, "Net Positive Suction

Head for Emergency Core Cooling and Containment Heat Removal System Pumps."

The applicant used the LOTIC-1 code to demonstrate the long-term capability of

minimum containment heat removal systems (one complete train of spray and

return air systems) to maintain the containment pressure below design pressure

for the design-basis LOCA. The staff has reviewed the applicant's containment.

Watts Bar SER 6-15
pressure and temperature response as calculated by the LOTIC-1 code. and concludes that the design of the containment heat removal systems is acceptable.,

Provisions are made in the containment spray system and the return air system

to permit inservice inspection of the system components and functional testing

of active components in both systems.

Based on the foregoing, the staff concludes that the design of the containment

heat removal systems is acceptable and meets the requirements of GDC 38, 39 and

40.

6.2.3 Secondary Containment Functional Design

The Sequoyah and Watts Bar plants have essentially the same secondary containment system designs. The secondary containment system includes the shield

building, auxiliary building secondary containment enclosure (ABSCE), auxiliary

building gas treatment system (ABGTS),.and emergency gas treatment system

(EGTS).

The secondary containment functional design goal is to preclude airborne fission products that may result from a LOCA from reaching the atmosphere in an

uncontrolled manner. The system, therefore, has the following design

objectives:

(1) maintain an effective barrier for gases and vapors that may leak from the

primary containment

(2) delay the release of such gases or vapors

(3) dilute and filter any gases or vapors before their releasi to the environs

The shield building is a reinforced concrete structure which completely surrounds the steel primary containment structure, providing a barrier around the

primary containment. Penetration seals on isolation valves and leakoff facilities are provided to limit the amount of inleakage into the annulus.

The ABSCE structure provides a barrier to preclude the uncontrolled release of

airborne radioactivity which may emanate from systems which it houses. The

penetrations, seals, and isolation valves are similar to those in the shield

building. Redundant, low leakage isolation dampers are provided for ventilation ducts penetrating the ABSCE boundary. Access to the ABSCE is provided

through airlocks.

The EGTS is divided into two subsystems: (1) an annulus vacuum control

subsystem and (2) the air cleanup subsystem..

The annulus vacuum control subsystem is a fan and duct network designed to keep

the annulus at a negative pressure (-5.0 in. wg) with respect to atmosphere

during normal operation; this will be included as a limiting condition for

operation in the facility Technical Specifications. There is no need for this

subsystem following an accident. The annulus vacuum control subsystem contains

two 100-percent-capacity (1000-cfm) fans.

The air cleanup subsystem operates following a LOCA to maintain.the annulus at

a negative pressure with respect to auxiliary building and primary containment

atmospheres and to remove particulates and vapors that may contain radioactive

Watts Bar SER 6-16
nuclides. Annulus pressure control is accomplished by adjusting the fraction

of the airstream that is returned to the annulus with that exhausted to the

atmosphere. An exhaust rate of 100 cfm is required to maintain the annulus

pressure at -0.5 in. wg. The air cleanup subsystem consists of two air cleanup

units (ACUs), each having 100-percent capacity at 4000 cfm. The ACUs are

started either by a Phase A isolation signal or manually.

The ABGTS operates to filter and exhaust air from all areas of the ABSCE while

maintaining a building pressure of -0.25 in. wg during building isolation. The

ABGTS uses the same ventilation ductwork that is used for normal operation.

Each ABGTS ACU is controlled by one airflow control module (consisting of a

modulating damper, differential pressure sensor, and transmitter) which controls the auxiliary building negative pressure by varying the amount of air

drawn from the auxiliary building. The ABGTS is composed of two 100-percent-°

capacity ACUs, with each rated at 9000 cfm.

There are two basic control modes for the ABGTS: (1) one ACU operating and the

other on standby (the standby unit comes on automatically upon a low-flow

signal from the operating fan) and (2) starting both ACUs manually and manually

placing one unit in a standby status approximately 30 minutes after startup.

The ABGTS is started automatically upon receipt of one of the following:

(1) Phase A containment isolation signal from either reactor unit

(2) high-radiation signal from the fuel-handling area radiation monitors

(3) high-radiation signal from the auxiliary building exhaust vent monitors

(4) high temperature in the auxiliary building intakes for the general supply

fan

During the interim between startup of Unit 1 and completion of construction of

Unit 2 at Watts Bar, an interim ABSCE will be defined, very much as was done at

Sequoyah. The staff will require, through the Technical Specifications, the

same measures to be taken at Watts Bar as were taken at Sequoyah to ensure the

operability of this interim ABSCE. Primarily, this consists of preoperational

and periodic testing to demonstrate the capability of the ABGTS to draw down

and maintain the pressure for the interim ABSCE configuration at -0.25 in. wg

during building isolation, as effectively as it would for the final ABSCE

configuration. This particularly applies to maintaining the spent fuel storage

area and the ESF pump rooms at 0.25-in. wg pressure.

The applicant has acceptably identified potential bypass leakage paths past the

annulus and into the ABSCE, and will leak test these paths in accordance with

the requirements of Appendix J to 10 CFR 50 to determine total bypass leakage.

The staff concludes that the secondary containment functional design is acceptable and meets the requirements of GDC 4, 16, and 43. This conclusion is based

on meeting the regulatory position in BTP CSB 6-3, "Determination of Bypass

Leakage Paths in Dual Containment Plants," and the fact that the Watts Bar

secondary containment functional design is the same as that for Sequoyah, which

has previously been reviewed by.the staff and found acceptable.

6.2.4 Containment Isolation System

The design objective of the containment isolation system is to allow the normal

or emergency passage of fluids through the containment boundary while preserving

Watts Bar SER 6-17
the ability of the boundary to prevent or limit the escape of fission products

in the course of postulated accidents. This section, therefore, is concerned

with the isolation of fluid systems which penetrate the containment boundary,

including the design and testing requirements for isolation barriers and actuators. The isolation barriers include valves, closed piping systems, and blind

flanges.

The containment isolation systems for the Sequoyah and Watts Bar plants are

designed to meet the criteria of Appendix A to 10 CFR 50 and SRP Section 6.2.4.

The Sequoyah and Watts Bar plants have very similar containment isolation systems; the only differences result from plant system differences. The major

differences between the plants are discussed below.

'(1) Penetrations X-8A,-8B,-8C, and-8D - In addition to the main and auxiliary

feedwater lines, Watts Bar has four feedwater bypass lines (one per steam

generator). These feedwater bypass lines provide another path for supplying feedwater to the.steam generator. These feedwater bypass lines penetrate primary containment (penetrations X-8A through X-8D) and attach to

the secondary side of the steam generator. The auxiliary feedwater lines

at Watts Bar connect to the feedwater bypass lines. The auxiliary feedwater lines to steam generators 1 and 4 join to the corresponding feedwater

bypass lines outside containment, and the auxiliary feedwater lines to

steam generators 2 and 3 penetrate containment (X-40A and -40 B) and connect to the respective feedwater bypass line inside containment.

With respect to Sequoyah, feedwater bypass lines are not used and penetrations X-8A through X-8D do not exist. At Sequoyah, the auxiliary feedwater lines connect directly to the main feedwater lines outside containment and downstream of the feedwater isolation valves.

The isolation provisions for the Watts Bar feedwater system have deficiencies similar to those found in the Sequoyah feedwater system .(see Section

6.2.4, Sequoyah Safety Evaluation Report (NUREG-O011), Supplement 5, June 5,

1981). These deficiencies are discussed below.

GDC 57 requires, for a closed system inside containment (such as the feedwater system), that each line have at least one isolation valve outside

containment that is either automatic, locked closed, or capable of remote

manual operation. A simple check valve may not be used as the isolation

valve.

Each of the four main feedwater lines and each of the four feedwater bypass

lines have no designated isolation valve. However, outside the containment

penetration and within the safety-grade (Quality Group B) portion of the

system, is an automatic or remote manual valve that meets the requirements

of GDC 57 for an isolation valve. Therefore, if the automatic-or remote

manual valve is designated as the containment isolation valve, the requirements of GDC 57 are met for each main feedwater and feedwater bypass line.

Each of the four auxiliary feedwater lines also has no designated containment isolation valve. However, upstream of each containment penetration

are remote manual valves which may be designated containment isolation

valves. These valves are in the safety-grade (Quality Group C) portion of

the system. Therefore, if the remote manual valves (LCVs 3-175, 3-171,

Watts Bar SER 6-18
3-171A, 3-174, 3-164, 3-164A, 3-173, 3-156, 3-156A, 3-172, 3-148, and

3-148A) are designated as containment isolation valves, the requirements

of GDC 57 are met for the auxiliary feedwater system.

For each of the four main feedwater lines and four feedwater bypass lines,

and for two of the auxiliary feedwater lines, there is a 1-in.-diameter

chemical feed line joining the main line downstream of the isolation

valve. Each of these 1-in. lines has a simple check valve as an isolation

valve. The applicant must modify each of these lines to satisfy the

requirements of GDC 57. To meet these requirements, the applicant must

install a safety-grade isolation valve of an appropriate type (such as,

automatic, remote manual, or locked-closed manual, but not simple check)

in the safety-grade portion of each line. The staff will require the

applicant to complete these modifications before initial criticality,

unless the applicant can justify that additional time is needed. In that

case, the staff will condition the license to require completion of the

modifications by a date to be mutually determined by the staff and

applicant.

As mentioned above, this-situation at Watts Bar is very similar to the

situation at Sequoyah, and Sequoyah is being required to make similar

modifications to its feedwater system.

When the modifications described above are in place at Watts Bar, the

isolation provisions for these lines will be acceptable.

(2) Penetration X-17 - The remote manual isolation valve in this RHR line is

outside containment at Watts Bar and inside containment at Sequoyah.

Containment isolation for both systems is composed of a closed system

outside containment and check valves inside containment.

Additionally, at both plants, a water seal is placed on the penetration by using the remote manual valves and the RHR pumps. These provisions are

acceptable.

(3) Penetrations X-19A and -19B - The containment spray suction lines connect

to the RHR sump suction lines outboard of the RHR sump suction isolation

valves at Sequoyah. At Watts Bar, however, the containment spray suction

lines connect inboard of the RHR sump suction isolation valves. For this

reason, each containment spray suction line at Watts Bar has a motoroperated containment isolation valve which is remote manually operated

from the main control room.. This is acceptable.

(4) Penetrations X-27A, -27B, -27C, and -27D - At Sequoyah, the steam generator

sample lines connect to the steam generator blowdown lines outboard of

primary containment and inboard of the outer isolation valve. Each of

these sample lines is isolated by an air-operated gate valve which receives

the Phase A isolation signal.

At Watts Bar, the steam generator sample lines penetrate primary containment (penetrations X-27A through X-27D) and connect directly to the steam

generator. Each sample line has two containment isolation valves which

receive the containment isolation signal - one valve immediately inside

Watts Bar SER 6-19
containment and one valve immediately outside containment. These provisions are acceptable.

These penetrations (X-27A through X-27D) are seal-welded spares at

Sequoyah.

(5) Penetrations X-40A and -40B - The auxiliary feedwater lines at Sequoyah

connect to main feedwater lines, but at Watts Bar the auxiliary feedwater

lines connect to feedwater bypass lines as described in (1) above. Containment isolation in these lines, however, is the same for both plants.

These provisions are acceptable, pending completion of the modifications

described in (1) above.

(6) Penetrations X-102 and -104 - At Sequoyah, these penetrations are used for

the auxiliary feedwater test lines. Isolation of these lines is accomplished by a closed system inside containment (secondary side of the steam

generator) and a locked.closed manual isolation valve outside containment.

At this time, there are no corresponding lines at Watts Bar and penetrations X-102 and X-104,are seal-welded spares. These provisions are

acceptable.

(7) Penetration X-107 - The RHR supply line at Watts Bar has a bypass line

around the inner containment isolation valve. The bypass line, which does

not penetrate containment, contains an inline containment isolation valve

which is motor operated and is remote manually operated from the main

control room. This is acceptable.

At Sequoyah, there is no corresponding bypass line.

(8) Penetrations X-111, -112, and -113 - At Sequoyah, these penetrations are

used for the vacuum relief lines which employ two isolation valves outside

containment.

At Watts Bar, these penetrations are seal-welded spares, which is

acceptable.

The Watts Bar plant-also meets all the requirements of Item II.E.4.2, "Containment Isolation Dependability," of NUREG-0737, except for subsection (6) concerning containment purging during normal operation. The applicant had not:•

fully addressed the provisions of BTP CSB 6-4, "Containment Purging During

Normal Plant Operations." By letter dated May 14, 1982, TVA submitted this

information, which is currently under review. The staff will report its findings regarding this item in a supplement to the SER.

With the condition that certain modifications (described above) to the chemical

feed lines be made, and with the exception of the issue of containment purging

during normal plant operations, the staff concludes that the containment functional design for containment isolation is acceptable and meets the requirements

of GDC 1, 2, 4, 16, 54, 55, 56, and 57 and Appendix K to 10 CFR 50. The basis

for this conclusion is that the Watts Bar-and Sequoyah containment isolation

systems are the same, except for the differences detailed above, and, where different, the Watts Bar provisions are in compliance with the regulations cited in

this paragraph. The design also complies with'the requirements of Item II.E.4.2

Watts Bar SER 6-20
of NUREG-0737, except for the open item concerning containment purging, which

will be addressed in a supplement to the SER.

6.2.5 Combustible Gas Control Systems

Following onset of a LOCA, hydrogen may accumulate within the containment as a

result of (1) metal-water reaction between the fuel cladding and the reactor

coolant, (2) radiolytic decomposition of the postaccident emergency cooling

water, and (3) corrosion of metals (zinc and aluminum) by emergency core coolant

and containment spray solutions. The applicant will provide redundant Westinghouse electrical thermal hydrogen recombiners designed to limit the hydrogen

concentration within the containment to below a limit of 4 v%. These recombiners

are the same as those installed at Sequoyah. However, the applicant's FSAR does

not include a discussion concerning the analysis of the production and accumulation of hydrogen within containment from the sources listed above. After review

of the applicant's March 5, 1982 submittal, the staff requested additional information. After receiving and reviewing this information, the staff will report

its findings in a supplement to the SER.

The Westinghouse electric thermal hydrogen recombiner system incorporates several

design features that are intended to ensure the capability of the system to be

operated in the event of an accident. Among these are

(1) seismic Category I design

(2) IEEE-279 requirements for the wiring and electrical equipment

(3) Quality Group B design

(4) protection from missile and jet impingement'from broken pipes

(5) redundancy to the extent that no single component failure can disable both

recombiners

(6) separate power supplies for each heater

Each of the two 100-percent-capacity electric recombiners is capable of processing 100 scfm of containment atmosphere for postaccident hydrogen control. The

staff has reviewed tests that have been conducted for a full-scale prototype

and a production recombiner. The tests consisted of proof-of-principle tests,

testing on a prototype recombiner, environmental qualification testing, and

functional tests for a production recombiner. (These tests are described in

WCAP-7820 and its Supplements 1-4.) The results of these tests demonstrated

that the recombiner should be capable of properly recombining hydrogen in a

post-LOCA containment environment. The recombiner system is designed to seismic

Category I criteria and to the IEEE-279 requirements for an ESF. These recombiners are the same as those used in the Sequoyah plant. Because these recombiners are situated inside containment, the requirements of Item II.E.4.1 of

NUREG-0737, "Dedicated Hydrogen Penetrations," do not apply; that is, no penetrations are used.

The containment return air fans, which begin operation 10 minutes after the

start of an accident, provide adequate mixing of the upper and lower containment compartment volumes. Two redundant hydrogen collection systems are provided to prevent the accumulation of hydrogen in the containment dome or in

dead-ended subcompartments in the lower compartment. These areas are continuously vented by diverting a portion of the return air fan flow through the

collection system; therefore, the potential for local hydrogen pocketing is

limited. These systems are the same as those used at the Sequoyah plant.

Watts Bar SER 6-21
In accordance with Regulatory Guide 1.7, the applicant also has provided a

containment purge system for postaccident containment atmosphere cleanup.

Capability is provided for periodic inservice inspection, operability testing,--

and leak rate testing of these systems and components.

Redundant hydrogen monitors are provided. They meet the provisions of Item

II.F.l of NUREG-0737, "Additional Accident Monitoring Instrumentation" (Attachment 6, "Containment Hydrogen.Monitor,") with the exception that they are not

environmentally qualified in accordance with Regulatory Guide 1.89 (NUREG-0588).

In this and other respects, the hydrogen monitors are the same as the monitors

installed at the Sequoyah plant, which have been-found acceptable by license

condition until the first refueling outage at Sequoyah. However, the staff

will require that qualified monitors be installed and operational at Watts Bar

in accordance with the implementation provisions of NUREG-0737 before the fullpower operating license is issued.

The measures being adopted by the applicant to control hydrogen that may be

generated during a degraded core accident are discussed in Appendix C to this

SER, under Unresolved Safety Issue A-48, "Hydrogen Control Measures and Effects

of Hydrogen Burns on Safety Equipment."

In summary, the staff is currently reviewing the additional information provided

by the applicant with respect to the production and accumulation of hydrogen

within the containment, and the analysis of the functional capability of the

systems provided to reduce combustible gas concentrations within the containment. Also, requirements on the proper qualification of the hydrogen monitors

must be satisfied. The remaining aspects of the design of the combustible gas

control system are acceptable and meet the requirements of 10 CFR 50.44 and

50.46, and GDC 5, 41, 42, and 43. The staff will report its conclusions

regarding the remaining open item in supplement to the SER.

6.2.6 Containment Leakage Testing

The Watts Bar plant containment design includes the provisions and features

required to satisfy the testing requirements of Appendix J, 10 CFR 50. The

design of the containment penetrations and isolation valves permits periodic

leakage rate testing at the pressure specified in Appendix J, 10 CFR 50.

Included are those penetrations that have resilient seals and expansion bellows,

such as airlocks, emergency hatches, refueling tube blind flanges, and electrical penetrations.

The applicant has designed the Watts Bar plant'containments so that there is no

potential path by which containment leakage could bypass both the emergency gas

treatment system and the auxiliary building gas treatment system and reach the

environs untreated. The applicant has identified systems for which throughline or penetration leakage could bypass the annulus and be released within the

areas of the auxiliary building which are treated by the auxiliary building gas

treatment system. The applicant has committed to perform local leak rate tests

in accordance with the requirements of Appendix J to 10 CFR 50 and to limit the

total potential leakage, which could bypass the emergency gas treatment system

and be treated by the auxiliary building gas treatment system, to 10 percent of

Watts Bar SER 6-22
the containment design leakage rate (0.25 percent per day by weight of the

containment atmosphere) at 15.0 psig.

The proposed reactor containment leakage testing program complies with the

requirements of Appendix J to 10 CFR 50. Such compliance provides adequate

assurance that containment leak-tight integrity can be verified periodically

throughout service lifetime on a timely basis to maintain such leakage within

the limits of the Technical Specifications.

Maintaining containment leakage rates within such limits provides reasonable

assurance that, in the event of any radioactivity releases within the containment, the loss of the containment atmosphere through leak paths will not be in

excess of acceptable limits specified for the site.

Based on the foregoing, the staff concludes that the containment leak testing

program is acceptable and meets the requirements of GDC 52, 53, and 54;

Appendix J to 10 CFR 50; and 10 CFR 100.

6.2.7 Fracture Prevention of Containment Pressure Boundary

The staff safety evaluation review assessed the ferritic materials in the Watts

Bar Units 1 and 2 containment system that constitute the containment pressure

boundary to determine if the material fracture toughness is in compliance with

the requirements of GDC 51, "Fracture Prevention of Containment Pressure

Boundary."

GDC 51 requires that under operating, maintenance, testing, and postulated

accident conditions, (1).the ferritic materials of the containment pressure

boundary behave in a nonbrittle manner and (2) the probability of rapidly

propagating fracture is minimized.

The Watts Bar plant containment system includes a freestanding containment

vessel enclosed by a reinforced concrete shield building. The ferritic

materials of the containment pressure boundary that were considered in the

staff assessment are those which have been applied in the fabrication of the

containment vessel, equipment hatch, personnel lock, penetrations, and fluid

system components including the valves required to isolate the system. These

components are the parts of the containment system which are not backed by

concrete and must sustain loads during the performance of the containment

function under the conditions cited by GDC 51.

The Watts Bar containment pressure boundary consists of ASME Code Class 1,

Class 2, and Class MC components. In late 1979, the staff reviewed the fracture toughness requirements of the ferritic materials of Class MC, Class 1, and

Class 2 components which typically constitute the containment pressure boundary.

Based on this review, the staff determined that the fracture toughness requirements contained in ASME Code editions and addenda typical of those used in the

design of the Watts Bar containment may not ensure compliance with GDC 51 for

all areas of the containment pressureboundary. The staff initiated a program

to review fracture toughness requirements for containment pressure boundary

materials for the purpose of defining those fracture toughness criteria that

most appropriately address the requirements of GDC 51. Before completing this

study, the staff elected to apply in its licensing reviews of ferritic containment pressure boundary materials the criteria for Class 2 components identified

Watts Bar SER 6-23
in the Summer 1977 Addenda of Section III of the ASME Code. Because the fracture toughness criteria that have been applied in construction typically differ

in Code classification and Code edition and addenda, the staff has chosen the

criteria in the Summer 1977 Addenda of Section III of the Code to provide a

uniform review, consistent with the safety function of the containment pressure

boundary materials. Therefore, the staff has reviewed the Class 1, Class 2,

and Class MC components of the Watts Bar Units 1 and 2 containment pressure

boundary according to the fracture toughness requirements of the Summer 1977

Addenda of Section III for Class 2 components.

Considered in the review are components of the containment system which are

load bearing and provide a pressure boundary in the performance of the containment function under operating, maintenance, testing, and postulated accident

conditions as addressed in GDC 51. These components are the containment vessel,

equipment hatch, personnel airlock, penetrations, and elements of the main steam

and main feedwater systems.

In some cases, materials were not tested or were tested in accordance with the

ASME Code edition and addenda in effect at the time of the test but which the

staff does not consider to be appropriate within the context of Summer 1977

Addenda, Section III, Class 2 Rules. The staff's assessment of fracture toughness of these materials is based on the metallurgical characterization of these

materials and fracture toughness data presented in NUREG-0577, "Potential for

Low Fracture Toughness and Lamellar Tearing on PWR Steam Generator and Reactor

Coolant Pump Supports," October 1979, and ASME Code Section III, Summer 1977

Addenda Subsection NC.

The metallurgical characterization of these materials, with respect to their

fracture toughness, was developed from a review of how these materials were

fabricated and what thermal history they experienced during fabrication. The

metallurgical characterization of these materials, when correlated with the data

presented in NUREG-0577 and the Summer 1977 Addenda of the ASME Code Section III,

provided the technical basis for the staff's evaluation of compliance with Code

requirements.

Based on its review of the available fracture toughness data and material fabrication histories, and the use of correlations between metallurgical characteristics and material fracture toughness, the staff concludes, based on.the

receipt of confirmatory information, that the ferritic components in the Watts

Bar Units 1 and 2 containment pressure boundary meet the fracture toughness

requirements that are specified for Class 2 components by the 1977 Addenda of

Section III of the ASME Code. Compliance with these Code requirements provides

reasonable assurance that the Watts Bar reactor containment pressure boundary

materials will behave in a nonbrittle manner, that the probability of rapidly

propagating fracture will be minimized, and that the requirements of GDC 51 are

satisfied.

The information required from the applicant will confirm that the lowest temperatures which will be experienced by the limiting materials of the reactor

containment pressure boundary under the conditions cited by GDC 51 will be in

compliance with the temperatures identified in the staff's analysis.

Watts Bar SER 6-24
The applicant states that the requirements will be met even with minimum engi- neered safeguards available, such as the loss of one emergency power bus, with

offsite power unavailable.

6.3 Emergency Core Cooling System

The emergency core cooling system (ECCS) is designed to provide core cooling as

well as additional shutdown capability for accidents that result in significant depressurization of the reactor coolant system. These accidents include

mechanical failure of the reactor coolant system piping up to and including the

double-end break of the largest pipe, rupture of a control rod drive, spurious relief valve operation in the primary and secondary fluid systems, and breaks

in the steam piping.

The principal bases for acceptance of this system are conformance with 10 CFR

50.46 and GDC 35.

6.3.1 System Design

The ECCS consists of an upper head injection system (UHI) and low pressure accumulator tanks that are passive systems that are actuated when the reactor

coolant pressure falls below preset values. The active components of the ECCS

are high head, medium, and low pressure pumps that are actuated by the safety injection signal. Following a postulated LOCA, the passive and active injec- tion systems will operate, and after the water inventory in the refueling

water storage tank has been depleted, the long-term recirculation mode will be

activated.

The UHI system consists of a borated-water-filled tank connected to a nitrogen tank that is pressurized. When the reactor system pressure falls below the UHI

pressure, water will be injected into the top of the core.

The low pressure accumulator system consists of four tanks that contain water

covered by nitrogen pressurized to approximately 400 psia. When the reactor

coolant system pressure falls to a level below the pressure-in the tanks, water

is forced into the four cold legs.

The high pressure injection mode of operation, upon actuation by a safety injection signal, will consist of the operation of two high head centrifugal

pumps that provide high pressure injection of concentrated boric acid solution

into the reactor coolant system. Also part of the high pressure injection mode

are two safety injection pumps which will take their suction from the refueling

water storage tank with a boron concentration of 2000 ppm.

The refueling waterstorage tank (RWST) is the water supply for the injection phase when ECCS actuates. It has a total volume of 378,000 gal with a minimum

inventory of 370,000 gal. The RWST is equipped with four 15-kW heaters to

maintain borated water above the minimum temperature (60'F) for boron precipi- tation. A 28-in. vent at the top of the RWST has a rain hood and is designed

to ensure against freezing or plugging. As'a further assurance of adequate venting, an 8-in. overflow line, is provided. The nominal minimum inventory

(370,000 gal) ensures 227,000 gal for injected water, about 74,000 gal to

accomplish switchover, and an additional 35,000 gal left in the RWST (to

account for NPSH and unusable volume), with about 34,000 gal for postulated

Watts Bar SER 6-25
instrument uncertainties. Elsewhere in this section the staff has reviewed net

positive section head (NPSH) calculations and addressed flooding elevation for

sump testing to maintain consistency with RWST volume allocations.

Based upon the above considerations and acceptable consequences to NPSH and

sump performance design, the staff finds the RWST capacity acceptable.

Low pressure safety injection (LPSI) consists of two residual heat removal

pumps that take their suction from the RWST. When the RWST level drops to

120,000 gals, the LPSI suction is automatically switched to the containment

emergency sump.

6.3.2 Evaluation

Single Failure

The staff has reviewed the system description and piping and instrumentation

drawings to ensure that abundant core cooling will be provided during the

initial injection phase with or without offsite power and assuming a single

failure. The UHI subsystem is aligned for injection, through two parallel

lines with normally open isolation valves, when the primary pressure drops

below the UHI set pressure. An inadvertent valve closure in either discharge

line will not preclude UHI. Each UHI discharge line has two isolation valves

in series that are closed automatically when a low level in the UHI accumulator

is reached. Failure of a single valve to close will not prevent isolation of

the UHI accumulator. The cold-leg accumulators have normally open isolation

valves in their discharge lines. These valves will have power removed from the

motor operators to preclude inadvertent closure during the ECCS injection phase.

There are two pumps in each of the three different active injection systems.

The pumps in each system are connected to separate power buses and would be

powered from separate diesel generators in the event of loss of offsite power,

as required by GDC 17. Thus, at least one pump in each injection train would

be actuated. The high and intermediate head injection.systems contain parallel

valves in the suction and discharge lines, thus ensuring system function even

if one valve fails to open. 'The low head injection system is normally aligned

so as to not require any valve actuation during the injection phase.

The applicant will remove power to the normally open valves in the hot-leg

injection lines to preclude inadvertent actuation during the initial injection

phase. The isolation valves in the crossover lines between the RHR pump discharge, and the suction of the high pressure pumps will be closed and power

removed to preclude inadvertent overpressurization of these lines. Similarly,

the applicant will remove power to the normally open.valves identified below:

(1) residual heat removal discharge valves

(2) refueling water storage tank to safety injection pump suction valve

(3) refueling water storage tank to residual heat removal suction valve

(4) safety injection pump miniflow line

(5) safety injection pump discharge to cold legs

These power removals will preclude inadvertent actuation of these valves which

potentially could result in unacceptable ECCS performance. The applicant's

proposed method for locking out valves uses a keylock in the control room, with

-Watts Bar SER 6-26
covers to reduce the likelihood of unwanted power restoration. The acceptability of this method is discussed in Section 7.6.

To address NUREG-0138, Issue 4, concerning emergency procedures which may

permit early manual reset of the safety injection signal during the injection

phase following a loss of offsite power, the applicant has committed to use

emergency operating procedures which preclude manual reset of the safety

injection signal for at least 10 minutes following a safety injection signal.

The staff finds this acceptable.

The staff has reviewed the submergence of ECCS equipment following an assumed

LOCA. The applicant has indicated that no powered electrical equipment in the

ECCS would become submerged and, therefore, ECCS performance would not be

affected. The staff finds this acceptable.

About 15 hours after a LOCA, the ECCS will be realigned to simultaneously

inject into both the cold and hot legs to preclude the excessive buildup of

boron concentration in the core. The applicant has determined this realignment

time (15 hours) using analytical methods that conservatively provide coolant

volume, boron solubility, and flow requirements (50 gpm) for heat removal.

The staff has reviewed the design to ensure that there is sufficient redundancy

in the injection paths to meet design requirements even with a single failure.

During the long-term recirculation cooling phase of ECCS, leak detection is

required to identify passive ECCS failures outside containment, such as pump

seal failures. The applicant has provided a system of water level monitors

(which meet IEEE-279, except single failure as discussed below), located in

each compartment that contains ECCS components. With this system the limiting

leak (assumed to be 50 gpm) would be detected and isolated within 30 minutes.

The applicant has calculated that the total leakage in 30 minutes would not

compromise long-term cooling. Leak rates of less than 50 gpm would result in

scenarios in which the detection (alarm) time would be longer, but the time

available for operator response would also be longer. The staff finds the

system acceptable because it provides a means of identifying and isolating a

passive failure in the ECCS outside containment.

The staff has asked the applicant to address two single-failure scenarios. The

first scenario ("charging pump deadheading") postulates a small break in which

the RCS pressure remains at or near the RCS safety-valve setpoint. The applicant has addressed the staff's concern about potential damage to the charging

pumps as a result of inadequate flow against this pressure by providing a modification to the Watts Bar system which automatically makes a miniflow path

available to the charging pumps whenever an "S" signal is present and the normal

miniflow path i~s isolated. The staff finds this proposal acceptable. Additional considerations that arise from generic review of this area will be

applied to Watts Bar. The second scenario postulates a failure in volume

control tank level instrumentation, diverting letdown away from the volume

control tank and permitting continued charging pump suction from the volume

control tank, with eventual cavitation of the charging pump(s). The applicant

has addressed this scenario indicating that diversion of letdown flow to a

holdup tank (on high level in the volume control tank (VCT rather than to the

VCT) and automatic opening of a charging pump suction path from the RWST (on

low VCT level) are both initiated independently by either of two diverse VCT

Watts Bar SER 6-27
level transmitters. In addition, the applicant has indicated that for this

scenario only one charging pump would normally be operating with two others

in standby as backup; one charging pump has adequate capacity for long-term

shutdown makeup requirements. Control room indications and alarms would alert

the operator to the above occurrences and assist in diagnosing the event.

Based on the foregoing discussion, the staff finds the applicant's response

acceptable.

Based on its review of the design features discussed above, the staff concludes

that the ECCS complies with the single failure criterion of GDC 35.

Qualification of Emergency Core Cooling System

The ECCS design to meet seismic Category I requirements in compliance with

Regulatory Guide 1.29 and its housing in structures designed to withstand a

safe shutdown earthquake and other natural phenomena as required by GDC 2 are

discussed in Section 3.0. The equipment design to meet Quality Group B in

compliance with Regulatory Guide 1.26 is discussed in Section 3.2.

The ECCS protection against missiles inside and outside containment by the

design of suitable reinforced concrete barriers which include reinforcedconcrete walls and slabs and conformation to GDC 4 is discussed in Section 3.5.

The protection of the ECCS from pipe whip inside and outside containment is

discussed in Section 3.6.

The active components of the ECCS design to function under the most severe duty

loads including safe shutdown earthquake is discussed in Section 3.9. The ECCS

design to permit periodic inspection in accordance with ASME Code, Section XI,

which constitutes compliance with GDC 36, is discussed in Section 3.9.

The ECCS incorporates two subsystems that serve other functions. The RHR system provides for decay heat removal during reactor shutdown; at other times

the RHR system is aligned for ECCS operation. The centrifugal charging pumps

are utilized for maintaining the required volume of primary fluid in the reactor

coolant system. On an ECCS actuation signal, the system is aligned to ECCS

operation and the chemical and volume control system function is isolated. In

neither case (RHR or centrifugal charging) does the normal system use impair

its capability to function as an integral portion of the ECCS.

Each reactor unit has a separate ECCS; however, portions are housed in a common

auxiliary building. The individual components within the building are separated

by barriers, and the installation has been reviewed for possible flooding, as

discussed in Section 3.4 and in preceding paragraphs. The design constitutes

demonstration that the ECCS is not shared by the two units, in compliance with

GDC 5.

Instrumentation and Control

The ECCS is initiated automatically on (1) low pressurizer pressure, (2) high

containment pressure, (3) high differential pressure between any two out of

three steam generators, or (4) high steam flow coincident with low average

temperature or low steam pressure. As noted above, the cold-leg accumulator

and upper head injection subsystems actuate automatically when the reactor

Watts Bar SER 6-28
coolant pressure decreases to a value below that at which the ECCS subsystems

are maintained. This meets the requirements of GDC 20.

Equipment status indication provided in accordance with the requirements of

Regulatory Guide 1.47, "Bypassed and Inoperable Status Indication for Nuclear

Power Plant Safety Systems,!' is discussed in Section 7.3. Automatic acuation

is provided by redundant signals, whose diversity is noted above. The ECCS may

also be manually actuated, monitored, and controlled from the control room, as

required by GDC 19. The instrumentation needed to monitor and control ECCS

equipment following a LOCA has been reviewed. The applicanthas committed to

provide an alarm (low flow) to alert the operator to a postulated degradation

of ECCS performance (such as sump clogging). In addressing procedures that

would assist the operator in responding to an ECCS flow degradation, the applicant has stated that emergency operating instructions would include checks for

inadequate core cooling and transfers to inadequate core cooling procedures if

indicated by the checks. Complemented by appropriate procedures, the instrumentation at Watts Bar can provide sufficient information so that the operator

can maintain adequate core cooling following an assumed LOCA. Environmental

qualification of electrical components of the emergency core cooling system

will be discussed in Section 3.11 upon receipt of TVA's response to NUREG-0588.

Functional Design

By letter dated February 12, 1982, the applicant submitted an analysis of NPSH

available to ECCS pumps. A worst-case condition was identified as an RHR pump

drawing suction from the reactor containment sump during the recirculation mode

post-LOCA while supplying water directly to the reactor vessel and indirectly

through high head pumps and while containment spray pumps are drawing water

from the same sump. The flow rate assumed for the RHR pump was 5500 gpm, conservatively higher than the expected pump runout flow rate. Because Watts Bar

has an ice condenser containment, the applicant has been allowed to assume a,

subcooled temperature for the sump water. The applicant will perform preoperational RHR runout flow tests for the worst-case flow condition. For these

tests the flow must be measured to be less than 5500 gpm to verify the conservatism in the NPSH analysis assumption.

In his analysis, the applicant has assumed the sump water to be at 190'F, based

on the possibility of a break in the area of the sump (not the design-basis

event) which would yield higher sump water temperatures. TVA calculates

14.4 ft of excess NPSH available.

The applicant has assumed a combined flow rate of 8000 gpm for the two containment spray pumps, whereas previous analyses submitted by the applicant had

assumed a corresponding flow rate of 9500 gpm. The applicant has not justified

this difference. In addition, the applicant has specified a containment water

level based upon an assumed 250,000 gal of injected water. As discussed in

the staff's review of RWST capacity, when instrument error considerations are

factored into the analysis, 227,320 gal are ensured, which results in a lower

flooding elevation (about 10 in. lower). Adjusting the available NPSH by

10 in. to account for the lower elevation, and about 2 ft to conservatively

bound the higher flow rate, the staff has estimated the excess NPSH available

to be 11.5 ft.

Watts Bar SER 6-29
Based on the use of acceptable analysis techniques, suitably conservative

assumptions, confirmatory inplant testing that will verify the conservatism of

the assumed RHR flow rate, and the staff's calculated results which indicate

that at least 11.5 ft of excess NPSH are available for a worst-flow case, the

staff concludes that the ECCS pumps will be provided with adequate NPSH for

all modes of operation. Therefore, the NPSH design of Watts Bar ECCS pumps is

acceptable.

Boron injection tank pressure is indicated in the control room with a high

pressure alarm. All accumulators have pressure indication in the control room

with high and low pressure alarms.

The valve arrangements on the ECCS discharge lines have been reviewed with

respect to adequate isolation between the reactor cooling system and the lower

pressure ECCS. In some lines, this isolation is provided by two check valves

in series with a closed isolation valve. Other discharge lines have only two

check valves in series. The acceptability of this arrangement is discussed in

Section 3.9.6.

Containment isolation features for all ECCS lines, including instrument lines,

the requirements of GDC 56 and Regulatory Guide 1.11, "Instrument Lines Penetrative Primary Reactor Containment," are discussed in Section 6.2.

The safety injection lines are protected from intersystem leakage by relief

valves in both suction header and discharge lines. Intersystem leakage detection is described in Section 5.2.7 for the RHR and safety injection pump

systems.

Switchover from the injection to recirculation phase is initiated at the low

RWST level set point (120,000 gals) with the automatic realignment of RHR pump

suction to the sump. The operator then completes the alignment of other ECCS

equipment leaving about 47,000 gals of water inventory in the RWST. As discussed in Section 6.3.1, the staff has reviewed the adequacy of the remaining

RWST capacity to complete the switchover without impairment of the ECCS and

found that capacity acceptable.

The applicant has proposed to divert emergency core cooling from the reactor

coolant systems to the auxiliary spray headers in the containment if only one

containment spray train is operational. This would be accomplished by isolating the direct injection of the RHR pumps into the cold legs and diverting

the RHR discharge to the safety injection train and the auxiliary spray line.

The applicant has stated that this procedure would not be required for at least

1 hour after an assumed LOCA accident. After 1 hour, the core flow from onesafety injection pump would be greater than 1.5 times the required core cooling

flow. The valve to the auxiliary spray header will be interlocked so that it

cannot be opened unless the sump suction valve for its respective RHR pump is

open. The staff finds this proposed diversion acceptable.

6.3.3 Testing

The applicant demonstrates the operability of the ECCS by subjecting all

components to preoperational and periodic testing, as recommended by Regulatory

Guides 1.68 and 1.79, "Preoperational Testing of Emergency Core Cooling System

for Pressurized Water Reactors," and GDC 37.

Watts Bar SER 6-30
Preoperational Test

One of these tests is to verify system actuation, namely the operability of all

ECCS valves initiated by the safety injection signal, the operability of all

safeguard pump circuitry down through the pump breaker control circuits, and

the proper operation of all valve interlocks.

Another test is to check the cold-leg accumulator system and injection line to

verify that the lines are free of obstructions and that the accumulator check

valves and isolation valves operate correctly. The applicant performs low

pressure blowdown of each accumulator to confirm the line is clear and check

the operation of the check valves.

Preoperational tests performed with the upper head injection to demonstrate

hydraulic resistances, absences of nitrogen entrainment vortexing, level set

points,*and isolation function are discussed in Section 14.

The ECCS pumps consist of the high head injection pumps, the RHR pumps, and the

safety injection pumps. The applicant uses the results of these tests to

evaluate the hydraulic and mechanical performance of these pumps delivering

through the flow paths for. emergency core cooling. The pumps will be operated

under both miniflow (through test lines) and full-flow (through the actual

piping) conditions.

By measuring the flow in each pipe, the applicant makes the adjustment necessary to ensure that no one branch has an unacceptably low or high resistance.

The staff requires an analysis of the results to verify that there is sufficient total line resistance to prevent excessive runout of the.pumps and adequate net positive suction head under the most limiting system alignment.

To show compliance with Regulatory Guide 1.79, the applicant must show that the

maximum flow rate predicted from the test results confirms the maximum flow

rate used in the net positive suction head calculations under the most limiting

conditions. The applicant must show that the minimum acceptable flow used in

the LOCA are met by the measured total pump flow and a relative flow between

the branch lines.

Based on its review of the test objectives disc'ussed above, the staff concludes

that the preoperational test program conforms to the recommendations of Regulatory Guides 1.68 and 1.79 and is acceptable pending successful completion of

the program. Additional discussion of the preoperational test program is

presented in Section 14. of this SER.

Sump Debris

The safety issue of containment sump performance under post-LOCA conditions can

be viewed as consisting of two parts: (1) containment sump hydraulic performance (such as providing adequate NPSH to the recirculation pumps with up to

50 percent of the screen area blocked) and (2) the effects of debris. The

first part is discussed in this section under NPSH and sump tests. The second

part, sump debris, must be resolved by addressing four objectives:

(1) containment, insulation, and sump design

(2) housekeeping

Watts Bar SER 6-31
(3) alarms to alert the operator to ECCS degradation

(4) actions to be taken in the event of ECCS degradation

The first objective is a containment and sump design such that post-LOCA debris

will not compromise ECCS performance by clogging the sump. All forms of potential debris are considered, with insulation debris being of principal concern.

The information provided in the applicant's response is not sufficient to make

a direct plant-specific assessment of the Watts Bar design versus sump debris.

However, there is sufficient similarity between the Watts Bar design and the

Sequoyah design (previously reviewed and approved) for the staff to conclude

that the Watts Bar design (predominantly metal reflexive insulation) against

sump debris is acceptable. The applicant (per the staff's review requests)

has verbally agreed to provide a detailed survey of insulation materials before

startup following the first refueling outage, to confirm this conclusion.

The second objective is to ensure by appropriate housekeeping procedures that

the containment "as-licensed" cleanliness is maintained. The applicant has

cited Technical Specifications which address this concern. The staff finds

this acceptable.

The third objective is ensuring that adequate alarms are available to alert the

operator to ECCS degradation if it should occur despite the precautions discussed above. The applicant has addressed this concern by committing to provide an alarm on low LPI flow.

The fourth objective is to have available equipment and procedures to respond

to an ECCS degradation occurrence. The applicant has addressed this concern by

stating that emergency operating instructions would include guidance to alert

the operator to the symptoms of inadequate core cooling. Emergency operating

procedures are included in the review of Section 13 of this report.

Based on the considerations noted above, and subject to the acceptability of

the detailed survey of insulation materials, the staff finds the Watts Bar

design against sump debris to be acceptable.

Sump Testing

The applicant has provided a report describing 1:4 scale model tests which were

carried out to demonstrate ECCS sump performance during the recirculation mode.

The tests simulated several flow conditions and configurations encompassing

those anticipated for the Watts Bar ECCS. These tests verified that modifications to the sump design similar to those made at Sequoyah would be effective

in ensuring acceptable sump performance at Watts Bar for submergences greater'

than 7-ft, 10-in. The recommended modifications have been made for Watts Bar,

and, even with the reduced flooding level ensured by RWST analyses, the submergence at Watts Bar should be about 12 ft. Based on the applicant's implementation of report recommendations, similarity of model parameters to Watts

Bar, and test results verifying the performance of the modified sump design,

the staff finds that these tests meet the intent of applicable portions of

Regulatory Guides 1.68, 1.79, and 1.82, "Sumps for Emergency Core Coolings

and Containment Spray Systems," and are, therefore, acceptable.

Watts Bar SER 6-32
Periodic Component Tests

Routine periodic testing of the ECCS components and all necessary support

systems at power will be performed. Valves which actuate after a LOCA are

operated through a complete cycle, and pumps are operated individually in this

test on their miniflow lines, except the charging pumps, which are tested by

their normal charging function. The applicant has stated that these tests will

be performed in accordance with ASME Code, Section XI, to satisfy GDC 37.

6.3.4 Performance Evaluation

The ECCS must provide abundant core cooling to minimize fuel and clad damage in

accordance with the requirements of 10 CFR 50.46. Topical Report WCAP-8479,

"Westinghouse Emergency Core Cooling System Evaluation Model Application to

Plants Equipped With Upper Head Injection," describes the Westinghouse calcuiational model for a PWR with ice condenser containment and upper head injection

(UHI) systems. The staff has reviewed and approved the Westinghouse evaluation

model for analyzing LOCAs in UHI plants. The applicant has submitted LOCA

analyses to satisfy the requirements of 10 CFR 50.46(b) and Appendix K. The

staff has found these analyses to be acceptable as discussed in Section 15.3.1.

The applicant has described his shutdown procedures to close and lock out power

to accumulator isolation valves while shutting down as follows:

(1) At 1000 psig, the operator will maintain pressure and cool down the RCS to

less than 425°F.

(2) At 1000 psig and less than 4250 F, the operator will close and lock out the

accumulator isolation valves.

The applicant has presented an evaluation of a LOCA during shutdown assuming no

credit for the accumulators and a worst single failure in the remaining ECCS

equipment. This evaluation, based on a LOCA 2-1/2 hours after shutdown, has

shown a peak cladding temperature of 1671°F at time of downcomer filling, which

is below the typical design-basis LOCA analysis temperature (1775 0 F) for a

corresponding event time. In the scenario, after downcomer filling, the LOCA

during the shutdown case would experience more rapid core reflooding and lower

power in the fuel than the design-basis event and thus a smaller subsequent

temperature rise. Therefore, the calculated peak-cladding temperature for the

shutdown case would be lower and the event (LOCA at shutdown) less limiting

than the design-basis LOCA. Because the calculated results of a LOCA at

shutdown are less than those anticipated for the design-basis LOCA, the staff

finds the proposed accumulator lockout procedure acceptable.

The applicant has also analyzed the consequences of a moderate energy line

break in the RHR system immediately after RHR initiation (at about 400 psig).

The analyses indicate that an alarm on low level would alert the operator to

the event 62 minutes before core uncovery, giving him sufficient time to

manually initiate ECCS.

Analyses of steamline breaks and post-LOCA long-term cooling design (most

limiting cases) demonstrate the capability of the ECCS to perform its reactivity control function as required by GDC 27.

Watts Bar SER 6-33
The staff is conducting a generic review of degraded cores (NUREG-0660,

Item II.B.8). Should this generic review prescribe any further action

applicable to the Watts Bar facility, thestaff will require the applicant to

comply.

6.3.5 Conclusions

The staff finds that the applicant's ECCS design meets the requirements of

10 CFR 50.46(b) and GDC 5, 25, 35, and the recommendations in Regulatory

Guide 1.1 and BTP RSB 6-1, and, therefore, is acceptable.

6.4 Control Room Habitability

Based on its evaluation, the staff finds that the calculated toxic gas and

radiological consequences are within the acceptance criteria contained in SRP

Section 6.4 and the design of the control room emergency ventilation system is

acceptable for preventing significant toxic gas and radiological exposure to

operating personnel in the control room.

The control room design meets GDC 4, with respect to "structures, systems and

components shall be designed to accommodate the effects of and to be compatible

with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents ..... " This conclusion is based on the

following:

Isolation of the control room occurs automatically upon the actuation

of a safety injection signal or upon indication of any of the following in the outside air supply high radioactivity, temperature, chlorine or smoke concentrations any of the following in the outside air

supply: The operators havethe capability of manually purging smoke

or fumes from the control room.

Specifically with respect to GDC-19, the applicant has protected the control

room operators against'radiation by the use of shielding and by the installation of a filtration system to remove airborne contaminants. After an accident, isolation occurs automatically in response to the accident signal

(safety injection) or the high gaseous radioactivity signal for inlet air.

This places the control room ventilation system in a pressurization mode such

that 200 cfm of pressurized air is supplied through adsorbers while 4000 cfm is

recirculated through redundant particulate and carbon filtration components.

In summary, the staff review was performed in accordance with SRP Sections

2.2.1, 2.2.2, 2.2.3, and 6.4, and Regulatory Guides 1.78 ("Assumptions for

Evaluating the Habitability of a Nuclear Power Plant Control Room During a

Postulated Hazardous Chemical Release") and 1.95 ("Protection of Nuclear Power

Plant Control Room Operators Against an Accident Chlorine Release"). The staff

finds that the control room habitability systems are adequate to provide safe,

habitable conditions within the control room under both normal and accident

conditions without personnel receiving radiation exposures in excess of 5 rems

whole body, or its equivalent to any part of the body, for the duration of the

accident.

Watts Bar SER 6-34
As a result, the staff concludes that the control room satisfies the requirements of: (1) Item III.D.3.4 of NUREG-0737 and (2) GDC-19 and is, therefore,

acceptable for a full power license.

6.5 Engineered Safety Feature (ESF) Filter Systems

6.5.1 ESF Atmosphere Cleanup System

The engineered safety feature atmosphere cleanup systems for Watts Bar Nuclear

Plant consist of process equipment and instrumentation to control the releases

of radioactive materials in gaseous effluents (radioactive iodine and particulate matter) following a design-basis accident. In the Watts Bar application,

there are four filtration systems designed for this purpose; the emergency gas

treatment system, the auxiliary building gas treatment system, the reactor

building purge system, and the main control room emergency air cleanup system.

6.5.1.1. Emergency Gas Treatment System (Air Cleanup Subsystem or EGTS)

The function of the air cleanup subsystem of the emergency gas treatment system

(EGTS) is to collect and process the leakage of radioactive materials from the

containment into the annular'space between the containment and the concrete

enclosure structure following a LOCA. The system is designed to maintain a

slight negative pressure in the annulus following a LOCA. The air cleanup subsystem is a redundant system. Each train has a design capacity of 4000 cfm and

includes the following components: demister, heating coil, high efficiency

particulate air (HEPA) filter, two banks of carbon adsorbers in series, HEPA

filter, and fan. A quencher-type water supply is provided for flooding the

charcoal in case of fire. The equipment and components are designed to Quality

Group C and seismic Category I, and are located in a seismic Category I

structure.

The staff has determined that the air cleanup subsystem of the EGTS is designed

in accordance with the guidelines of Regulatory Guide 1.52 ("Design, Testing,

and Maintenance Criteria for Post-accident Engineered-Safety-Feature Atmosphere

Cleanup System Air Filtration and Absorption Units of Light-Water-Cooled Nuclear

Power Plants"). In its evaluation, the staff has assigned the system decontamination efficiencies of 99 percent for elemental and organic iodine and 99 percent for particulates. Based on this evaluation, the staff finds that the

EGTS is designed to control the releases of radioactive materials in gaseous

effluents in accordance with applicable regulations following a postulated DBA.

6.5.1.2 Auxiliary Building Gas Treatment System (ABGTS)

The function of the auxiliary building gas treatment system (ABGTS) is to collect and process leakage from the fuel-handling and waste packaging areas of

the auxiliary building during accidents. The system is designed to maintain a

slight negative pressure in the auxiliary building following an accident. The

AGBTS is a redundant system. Each train has a design capacity of 9,000 cfm of

air and includes the following components: heater, prefilter, HEPA filter,

carbon adsorber, and fan. The equipment and components are designed to Quality

Group C and seismic Category I and are located in a seismic Category I

structure.

Watts Bar SER 6-35
The staff has reviewed the ABGTS in accordance with the guidelines of Regulatory Guide 1.52. In its evaluation, the staff has assigned the system decontamination efficiencies of 99 percent for elemental and organic iodine and

99 percent for particulates. Based on this evaluation, the staff finds that

the ABGTS is designed to control the releases of radioactive materials in gaseous effluents in accordance with applicable regulations following a postulated

DBA.

6.5.1.3 Reactor Building Purge Ventilation System (RBPVS)

The function of the reactor building purge ventilation system (RBPVS) is to

ensure that activity released inside containment from a refueling accident or

fuel-handling accident is treated prior to discharge to the environment. The

ESF portions of the RBPVS are redundant. Each train has a design capacity of

14,000 cfm of air and includes the following components: prefilter, HEPA filter, carbon adsorber, and fan. The equipment and components are designed to

Quality Group C and seismic Category I and are located in a seismic Category I

structure.

The staff has reviewed the RBPVS in accordance with the guidelines of Regulatory Guide 1.52. In its evaluation, the staff has assigned the system decontamination efficiencies of 95 percent for elemental and organic iodine and

99 percent for particulates. Based on this evaluation, the staff finds that

the RBPVS is designed to control the releases of radioactive materials in

gaseous effluents in accordance with applicable regulations following a

postulated DBA.

6.5.1.4 Main Control Room Emergency Air Cleanup System

The function of the main control room emergency air cleanup system (MCREACS) is

to supply nonradioactive air to the control room after a DBA and to pressurize

the control room. This system will permit operating personnel to remain in the

control room following a DBA. The MCREACS is a redundant system, with each

system having an intake design capacity of 200 cfm of air and recirculating

design capacity of 4000 cfm of air. Each system contains the following components: HEPA filter, carbon adsorber, and fan. Cooling coils are also provided for relative humidity control. The equipment and components are designed

to Quality Group C and seismic Category I and are located in a seismic Category I structure.

The staff has reviewed the MCREACS in accordance with the guidelines of Regulatory Guide 1.52. In its evaluation, the staff has assigned the system decontamination efficiencies of 95 percent for elemental and organic iodine and

99 percent for particulates. Based on this evaluation, the staff finds that

the RBPVS is designed to control a suitable control room environment following

a postulated DBA.

6.5.2 Containment Spray as a Fission Product Cleanup System

The Watts Bar design does not take into consideration the containment sprays to

perform as a fission product air cleanup system in its accident analyses.

Therefore, the staff did not review the containment spray design for this

function.

Watts Bar SER 6-36
6.5.3 Fission Product Control System andStructures

The discussion of the Watts Bar secondary containment as a fission product

control system is included in Sections 6.2.3, 6.5.1, and 15.4.1 of this SER.

6.5.4 Ice Condenser as a Fission Product Removal System

The ice condenser is designed to remove iodine from the postaccident atmosphere

passing through the ice beds. Sodium tetraborate is added to the ice to enhance

the iodine adsorption characteristics of the ice. Technical Specifications

require a minimum ice pH of 8.5 whenever the reactor is critical.

The ice condenser iodine removal effectiveness is a function of the flow rate

through the alkaline ice beds and the mole-fractions of air and steam in the

flow. Based on the expected conditions following a postulated LOCA, the ice

condenser iodine removal effectiveness is expected to be high from the initiation of the accident until melt out of the ice beds has occurred. However, it

is difficult to establish assured minimum values for the flow rate and molefraction of air prior to startup of the recirculation fans. Therefore, in its

model of ice condenser effectiveness, the staff has assumed that the alkaline

ice will remove iodine from the 40,000 cfm flow through the ice beds established

by the recirculation fans, commencing from the fan startup at 10 minutes and

ceasing at the earliest ice bed melt out (60 minutes after the design-basis

LOCA). A minimum value of 30 percent efficiency for the removal of elemental

iodine was assumed during this period.

6.6 Inservice Inspectionof Class 2 and 3 Components

GDC 36, 39, 42, and 45 require, in part, that the Class 2 and 3 components be

designed to permit appropriate periodic inspection of important component parts

to ensure system integrity and capability. Section 50.55a(g) of 10 CFR 50

defines the detailed requirements for the preservice inspection programs for

light water cooled nuclear power facility components. Based on the construction permit date of January 23, 1973, this section of the regulations requires

that a preservice inspection program for Class 2 and 3 components be developed

and implemented using at least the edition and addenda of Section XI of the

ASME Code in effect 6 months prior to the date of issuance of the construction

permit. Also, the initial inservice inspection program must comply with the

requirements of the latest edition and addenda of Section XI of the ASME Code

in effect 12 months prior to the date of'issuance of the operating license,

subject to the limitations and modifications listed in 10 CFR 50.55a(b).

The Watts Bar preservice inspection program and revisions submitted in letters

dated May 19 and June 4, 1982, are currently under staff review. The program is

based on the 1974 Edition of Section XI of the ASME Code with addenda through

the Summer 1975 Addenda. Additional information has been requested from the

applicant in order to allow completion of the review. Specific areas where ASME

Code examination requirements cannot be met will be identified as performance of

the preservice examination progresses. The staff's complete evaluation of the

PSI program will be presented in a supplement to this SER after the applicant

submits the required examination information and identifies all plant specific

areas where ASME Code Section XI requirements cannot be met and provides a

supporting technical justification.

Watts Bar SER 6-37
The initial inservice inspection program has not been submitted by the applicant. The staff will evaluate the program after the applicable ASME Code

Edition and Addenda can be determined based upon 10 CFR Section 50.55a(b), but

before the first refueling outage, when inservice inspection commences.

Compliance with the

Code and 10 CFR 50,

requirements of GDC

preservice and inservice inspections required by the ASME

constitutes an acceptable basis for satisfying applicable

36, 39, 42, and 45.

Watts Bar SER 6-38
7 INSTRUMENTATION AND CONTROLS

7.1 Introduction

7.1.1 General

The staff evaluated the Watts Bar instrumentation and control systems using

the criteria set forth in SRP Table 7-1 (NUREG-75/087, Revision 1). These

criteria include the GDC of 10 CFR 50, Appendix A, the Institute of Electrical

and Electronic Engineers (IEEE) Standards (including IEEE 279-1971, "Criteria

for Protection Systems for Nuclear Power Generating Stations"), Regulatory

Guides, and BTPs.

In addition to reviewing FSAR material, the staff also reviewed drawings

referenced inthe FSAR, held several design review meetings with the applicant,

and conducted a site visit to review the physical arrangements and installation

of the instrumentation and controls. The review concentrated on (1) those

areas where the design differs from the plants referenced as similar by the

applicant (most notable, the Sequoyah facility), and (2) areas which have

remained of concern during reviews of other similar plants.

Seismic and environmental qualification of the instrumentation and control

systems is addressed in SER Sections 3.10 and 3.11.

7.1.2 Comparison with Other Plants

The applicant states in the FSAR that the design of the Watts Bar NSSS is

similar to that of the Sequoyah, Donald C. Cook, and Trojan nuclear plants.

The only significant differences identified between these plants are the

physical configuration of the safety-related display instrumentation, the

load-rejection capabilities, and the rod-position indication (analog or digital

systems). For the comparison of the design of the Watts Bar secondary cycle,

the applicant refers to the Sequoyah, D. C. Cook, and Zion Plants. The

applicant further indicates that Sequoyah should be used as the primary

reference plant. The applicant provided a comparison of the Watts Bar and

Sequoyah designs in a letter dated August 17, 1981.

7.1.3 Design Criteria

The staff review of the design bases and the descriptions provided in the FSAR

for the safety-related systems shows that the design complies with the criteria

the staff reviewed and approved for the CP issuance. The staff also finds that

the criteria used for the design of the Watts Bar instrumentation and control

systems are in general agreement with SRP guidelines.

In the FSAR, the applicant discusses the extent to which the recommendations

of the guidelines are used in the Watts Bar design.

Watts Bar SER 7-1
7.1.3.1 Safety System Set Point Methodology

The methodology for establishing NSSS and BOP safety system set points for

Watts Bar is the same as for Sequoyah. The staff reviewed this methodology

and finds it acceptable for Watts Bar. Because the primary function of this

information is to confirm the adequacy of set points specified in the plant

Technical Specifications, the staff will audit the set point methodology during

its review of the Technical Specifications. In addition, the applicant committed to provide a list of all safety-related functions and a summary of

the data used in the set point analysis to facilitate the review of the proposed

Technical Specifications.

7.1.3.2 Response Time Testing

To ensure that'the response times of each protective function of the reactor

trip system (RTS) and the engineered safety features actuation system (ESFAS)

are within the limits assumed in the accident analysis, Technical Specifications require testing the time response at specified intervals. The periodic

test frequency specified in the Technical Specifications, as stated by the

applicant, is developed primarily on the basis of past operating experience.

If the response time of any protection channel were found to be marginal or to

require frequent adjustments of instrumentation because of changes on plant

conditions, the test frequency would be increased.

In response to a staff request for additional information to evaluate the

design adequacy of the testing system, the applicant provided a detailed discuss ion on the design provisions, test equipment, and test procedures. As

stated by the applicant, testing the response time of each protective function

from sensor through actuated equipment in one step is not practical because of

the system complexity. Consequently, the response time test is performed in

several discrete overlapping segments. The response time of each segment is

measured separately and then combined with the response times of other segments

to obtain overall response time of each protection function.

The segmented test methodology uses built-in test and calibration features,

thereby minimizing the number of modifications required. The segments into

which the protection systems are divided for testing are the analog instrumentation, solid-state 'logic protection system, and the final actuation devices

(see Sections 7.2 and 7.3 below). The specialized test equipment includes

hydraulic signal generators to test the response of the pressure, level, a flow

sensors, and the loop current step response (LCSR) test method for measuring

the response of the resistance temperature detectors (RTDs). Also, current

and voltage sources are used to generate certain analog and digital test

signals. The nuclear instrumentation detectors are excluded from the test.

The testing scheme described above is in agreement with the requirements of

IEEE Standand 338-1975 and Regulatory Guide 1.118; thus the design of the

testing systems is acceptable.

Watts Bar SER 7-2
7.2 Reactor Trip System

7.2.1 System Description

The reactor trip system (RTS) automatically shuts down the reactor whenever

established limits of safe operation are reached. To accomplish its function,

the RTS includes instrumentation channels to monitor various plant variables

(process and nuclear) pertinent to the reactor safety. Whenever a monitored

variable reaches a set limit, the associated instrumentation channel trips a

bistable. This turns off power to the relays that provide inputs (voltage/no

voltage) corresponding to the condition (normal/abnormal) of the measured

parameter to the solid-state logic protection system consisting of two redundant

trains, A and B. This system combines the input signals in an established logic

combination and generates trip signals (no voltage) to the undervoltage coils

of the reactor trip circuit breakers whenever the necessary combination of input

signals occur. Opening either of the two breakers interrupts power to the rod

control power supply, and the rods fall, by gravity, into the core.

In addition to the automatic trip of the reactor described above, means also

are provided for manual trip by the operator. The manual trip consists of

two switches. Actuation of either switch removes the voltage from the undervoltage coils and energizes the shunt trip coils of both reactor trip breakers.

In addition, the reactor will be tripped by actuating either of the two manual

switches for safety injection (see Section 7.3).

The trips included in the RTS are listed below. The first number in parentheses

after each trip parameter is the number of coincident trips required; the second

number is the number of redundant channels provided.

(1)

(2)

(3)

(4)

(5)

(6)

(7)

(8)

(9)

(10)

(11)

(12)

(13)

(14)

(15)

(16)

(17)

(18)

(19)

power-range high neutron flux (low and high settings)

intermediate-range high neutron flux

source-range high neutron flux

power-range high positive neutron flux rate

power-range high negative neutron flux rate

overtemperature AT Trip

overpower AT trip

pressurizer low pressure

pressurizer high pressure

pressurizer high-water level

low reactor-coolant flow (2/3 in any

reactor coolant pump bus undervoltage

reactor coolant pump bus underfrequency

low feedwater flow (1/2 steam/feedwater flow

mismatch in coincidence with 1/2 low steam

generator water level) (1/2 in any

low-low steam generator water level (2/3 in any

safety injection (see Section 7.3)

turbine-generator trip

(a) low auto stop oil pressure, or.

(b) turbine stop valve close

general warning alarm

manual

(2/4)

(1/2)

(1/2)

(2/4)

(2/4).

(2/4)

(2/4)

(2/4)

(2/4)

(2/3)

loop)

(2/4)

(2/4)

loop)

loop)

(1/2)

(2/3)

(4/4)

(2/2)

(1/2)

Watts Bar SER 7-3
The trip variables shown in the list above are monitored directly and their

functions are self explanatory. Exceptions are the overtemperature AT,. the

overpower AT, and the general warning alarm. The overtemperature AT protects

against a low departure-from-nucleate-boiling-ratio (DNBR). The set point for

this trip is continuously calculated by analog circuitry that receives inputs

from instrumentation channels measuring several different variables. The

overpower AT protects against excessive local linear power density. Several

variables are monitored, and the setpoint is continuously calculated by analog

circuitry. The general warning alarm system monitors various conditions, suchas power supply output, test switch position, and so forth, in the solid-state

logic protection system. If any of the monitored conditions in a train is

abnormal, the alarm relay for that train is de-energized. This actuates the

train trouble annunciator in the control room. If an abnormal condition occurs

simultaneously in both trains, the reactor is automatically tripped.

Some of the trips shown in the list above are not effective below or above

certain power levels. The source range high neutron flux trip, the intermediaterange high neutron flux trip, and the power-range high neutron flux low-setting

trip can be manually blocked above set power levels. These blocks are automati--

cally removed when the power level decreases below the set value. Trip numbers

(8), (10), (11), (12), (13) and (17) in the list above are automatically blocked

below set power levels. The blocks are automatically removed when the power

increases above the set value.

The RTS includes provisions for testing the system operation. The testing is

carried out in steps, a part of the system at a time, in a sequence that provides

the necessary overlap to ensure complete system operability. All of the system

can be tested at power, except for the manual reactor trip and the safety

injection manual trip switches because actuation of these switches would trip

the reactor. Also, the nuclear channel trips that are not effective above

certain power levels are tested at reduced power levels and at shutdown. The

analog process channel testing is performed by introducing dummy input signals

into the instrumentation channels and observing the tripping of the appropriate

output bistables. The logic matrices of the solid-state logic protection

system are tested by introducing test logic input signals in all combinations

of the trip and no-trip logic. For testing the reactor trip breakers, bypass

breakers are provided. Bypassing the trip functions during testing is required

for the source and intermediate range nuclear channels that are arranged in

one-out-of-two logic to trip.

The staff reviewed the generic aspects of the testing scheme for the solid-state

logic protection system in connection with review of the Callaway plant.

Plant-specific aspects were reviewed with the Watts Bar applicant, and the

equipment was inspected during the site visit. The staff finds the testing

provisions of the solid-state logic protection system and its actuated devices

to be acceptable.

7.2.2 Manual Trip Switches

As described by the applicant, two manual reactor trip switches are provided.

Each of the switches has outputs to both train A and train B reactor trip

breakers. The same arrangement applies to the manual safety injection switches

that also are used to trip the reactor.

Watts Bar SER 7"4
Having each of the manual trip switches actuate breakers of both trains increases

the system reliability. The staff was concerned, however, about the effect of

this arrangement on the separation of the trains. In response to this concern,

the applicant stated, and the staff verified during the site visit, that separate

contact blocks are used on the switches for each of the trains and that barriers

are provided where needed to meet the separation criteria applicable to redundant

Class 1E systems. Based on this information, the design is acceptable.

7.2.3 Testing of Reactor Trip Breaker Shunt Coils

A related concern involves the testing of the manual reactor trip switch

operation. Each of the switches controls power to both the undervoltage coil

and the shunt trip coil of each breaker. De-energizing the undervoltage coil

or energizing the shunt coil trips the breaker. As a result, tripping a breaker

when testing a manual switch does not confirm that both breaker trip coils

operate as intended. The staff expressed its concern to the applicant and

advised him that the staff will require in the Technical Specifications that

the shunt coils be tested at least once each refueling outage.

7.2.4 Anticipatory Trips

It is the staff's position that all inputs to the RTS, regardless of their

classification, must be seismically and environmentally qualified and conform

to the requirements of IEEE Standard 279-1971. An exception to this requirement is, however, allowed for the'anticipatory trips that are not required for safety

actions in the accident analyses. The exception is that the sensors and the

associated cabling are not required to be in a seismic Category I structure.

The only trip classified as anticipatory by the applicant is the reactor trip

on turbine trip. The signals that generate this trip are the low autostop oil

pressure and closure of the steam stop valves. For the low autostop oil

pressure, three channels are provided and are arranged in a two-out-of-three

trip logic. A four-out-of-four valve-closed trip logic is employed for the

four steam stop valves. The above trip signals are arranged in four channels.

Three of the channels include a low autostop oil pressure signal and a closed

signal from one of the four steam-stop valves. As stated by the applicant,

the design of the reactor trip on turbine trip meets all of the requirements

of IEEE Standard 279-1971, including separation, redundancy, and testability. The applicant also states that no failure of the seismically nonqualified

circuits or devices would cause any adverse consequences on the protection

system operation.. In particular, the input isolation devices for the protection

system are qualified to a voltage level in excess of any voltage to which the

sensors are exposed, and the sensors are of a type that is qualified in these

areas. Based on its review, the staff finds the design of the anticipatory

reactor trip on turbine trip acceptable.

7.2.5 Steam Generator Water Level Trip

Three water level measurement channels are provided for each of the steam

generators and are arranged in a two-out-of-three logic to trip the reactor on

low-low water level. Isolated outputs of these channels also are used as input signals to the steam generator water level control system. In addition, each

Watts Bar SER 7-5
steam generator's water level is a variable included in the postaccident

monitoring system (PAMS).

Reference Leg

The water level measurement channels use differential pressure transmitters.

The measurement accuracy of such a system is affected by several factors. Of

primary importance is the increase in the indicated water level caused by a

decrease of the water density in the reference leg resulting from an increase

in the ambient temperature as a result of a high-energy-line break. For such

an accident, the steam generator water level provides the primary trip function,

and the trip set points must be selected to ensure that the action required by

the safety analyses will be initiated throughout the range of temperatures that

can be expected. This issue was addressed for operating reactors in IE

Bulletin 79-21. There are several approaches that can be taken to alleviate

the temperature-dependence problem, such as insulation of the reference leg or

electrical compensation of the measured signal to reflect-temperature changes.

In a March 10, 1982 letter the applicant committed to adopt one of the above

steps before fuel loading. The staff finds this commitment acceptable and will

report its review of the final design in a supplement to the SER.

Low-Level Trip

Another item of concern pertaining to.the low-low steam generator water level

reactor trip is the number of instrumentation channels provided. For Watts

Bar, three channels are provided; one of these also is usedto provide a signal

to the water level control system. A high failure of the channel used for control would not only prevent the supply of water to the steam generator, but

also would make the channel ineffective in the trip logic on low water level.

This would reduce the low-level protection logic from two out of three to two

out of two, which would not meet the single-failure criterion. The staff

finds that this design does not introduce a concern for adverse control and.

protection system interaction because the level channel that is used for control

is independent of the two remaining level channels that are used in combination

with the feed flow/steam flow mismatch reactor trip. Therefore, the staff finds

the Watts Bar design does not violate the single-failure criterion.

7.2-.6 Conclusions

The review of the RTS included the initiating circuits, logic, bypasses, -interlocks, redundancy, diversity, and actuation devices used to implement reactor

shutdown. The review included descriptive information, functional logic di-agrams,

functional instrumentation and electrical diagrams, and partial physical arrangement drawings, as well as the applicant's design basis and its relation to the

design for the RTS. The review also included the means for identification of

cables and equipment and periodic testing capability.

In addition to the acceptance criteria discussed above, the staff's basis for

acceptance includes the staff's previous review and acceptance of similar systems.

It has been verified that the Watts Bar RTS is functionally the same as that

for the Sequoyah plant which has been previously reviewed and accepted.

Watts Bar SER 7-6
Items audited during the staff review include conformance to system redundancy

and diversity; single failure; both electrical and physical independence; identification of control boards, equipment, cables, and cable trays; and system testing

and inoperable status surveillance criteria.

The staff finds that the Watts Bar RTS conforms to applicable regulations,

guides, technical positions, and industry standards and is acceptable subject

to confirmation of the items identified above.

7.3 Engineered Safety Features Actuation System

7.3.1 System Description

The engineered safety features actuation system (ESFAS) monitors selected plant

parameters, and whenever predetermined safety limits are reached, the system

sends actuation signals to the appropriate engineered safety features (ESF)

and the auxiliary support systems equipment. Typical accidents that. require

actuation of the ESF systems are a loss of primary coolant and steamline breaks.

The plant variables that are monitored by the analog circuitry of the ESFAS

include pressurizer pressure, steamline pressures and flows, steamline differential pressure, containment pressure, and reactor coolant average temperature.

Whenever a monitored variable reaches a set limit, the associated instrumentation

channel trips a bistable. This turns off power to the relays that provide inputs

(voltage/no voltage) corresponding to the condition (normal/abnormal) of the

measured parameter to the solid-state logic protection system, which consists

of two redundant trains, A.and B. Whenever a required logic combination of

inputs is received by the solid-state logic protection system, each train

operates an appropriate master relay. Contacts of these relays are used to

operate slave relays that, in turn, provide contacts to actuate various ESF

system equipment.

The ESFAS signals and the plant conditions that generate these signals are shown

below. The first number in parentheses after each parameter indicates the number

of coincident trips required', and the second number is the number of redundant

channels provided.

(1) Safety Injection

(a) manual (1/2)

(b) high containment pressure (2/3)

(c) high differential pressure between steamlines (2/3 in any steam line)

(d) pressurizer low pressure (2/3)

(e) high steam flow in 2/4 steamlines coincident

with low-low T or low steamline pressure (1/2 in any two lines) avg

(2) Containment Spray and Containment Isolation, Phase B

(a) manual (two sets, two switches per set) (1/2 sets)

(b) containment pressure high-high (2/4)

Watts Bar SER 7-7
(3) Containment Isolation, Phase A

(a) manual (1/2)

(b) automatic safety injection (see second through fifth items

for function (1) above)

(4) Steamline Isolation

(a) manual (1/1 for any loop)

(b) high steam flow in 2/4 steamlines coincident

with low-low Tavg or low steam pressure (1/2 in any two lines)

(c) containment pressure high-high (2/4)

(5) Feedwater Line Isolation

(a) safety injection (see function (1) above)

(b) steam generator level high-high (2/3)

In addition to actuation of various equipment as needed to fulfill the functions

listed above, some of the ESFAS signals also are employed in the actuation or

realignment of the following systems:

(1) diesel generators

(2) ventilation systems (containment and control room)

(3) essential raw cooling water system

(4) component cooling water system

(5) auxiliary feedwater system

(6) emergency gas treatment system

(7) reactor trip system (see Section 7.2)

The positions of the ESF remotely operated valves are indicated by lights

located next to the control switch. Also, a light on the monitor panel is

actuated if a valve is in the automatically actuated position. The inoperable

or bypassed status of the ESF systems is addressed in Section 7.7.2.

The testing of the analog instrumentation channels and the solid-state logic

protection system is carried out in the same way as described for the RTS in

Section 7.2. Most of the ESF final actuators also can be fully tested during

online operation. The remaining actuators whose operation is not compatible

with continued online plant operation are checked by continuity testing. As

noted in Section 7.2, the staff reviewed the testing provisions in detail and

finds them acceptable.

7.3.2 Containment Sump Level Measurement

One of the permissives to the initiation logic for the automatic switchover

from the injection to the recirculation mode of the emergency core cooling

system (ECCS) is provided by water in the containment sump reaching a set level.

The sump water level is monitored by four level measurement channels utilizing

differential pressure transmitters. The staff was concerned that debris in the

sump could block the inlets to the differential pressure transmitters and result

in a loss of the permissive signal to the switchover logic. The applicant-has

committed to make a design change to provide protection to prevent debris from

Watts Bar SER 7-8
entering-the level sensors. The staff finds this commitment to be acceptable.

Once the final design has been submitted, the staff will report its findings

in a supplement to the SER.

7.3.3 Auxiliary Feedwater Initiation and Control

In the event of a loss of the main feedwater supply, the auxiliary feedwater

(AFW) system supplies sufficient feedwater to the steam generators to remove

the energy stored in the primary system. The two plant units have separate

AFW systems except for certain shared support facilities. Each system has two

470-gpm electric-motor-driven pumps, and one 940-gpm turbine-driven pump. Each

of the electric pumps serves two steam generators; the turbine-driven pump serves

all four. The preferred sources of water for the AFW system are the two condensate storage tanks. The backup water supply is provided by the essential raw

cooling water (ERCW) system.

The pumps start automatically on a loss of offsite power, loss of both main

feedwater pumps, or a safety-injection signal. The electric-motor-driven pumps

also start automatically on a two-out-of three low-low water level signal from

any steam generator; the turbine-drive pump starts automatically on a two-outof-three low-low level signal from any. two steam generators. All pumps also

can be started either remote-manually or locally. A modulating level control

valve (which is normally closed) between each pump and each steam generator

fed by the pump receives an opening signal on a low!-low water level in the

steam generator. After an accident, the operator can take manual control of

the system-level operation by blocking the initiating signal with a handswitch.

However, if another initiation signal occurs, the control will again revert

to automatic.

The staff reviewed applicable information in Sections 7.3 and 10.4.9"of the

FSAR. In response to a request for additional information, the applicant gave

a detailed presentation of the automatic initiation, operation, reset, and

control of the AFW system. The applicant stated that the auxiliary feedwater

control circuitry, is safety grade, Class 1E, is powered from a power source

connected to the emergency power system, and is testable on line. A manualinitiation capability that is independent of the automatic initiation is also

provided. The design of the auxiliary feedwater system for the Watts Bar plant

is similar to that for the Sequoyah plant for which a failure mode and effects

analysis has been performed and documented in the FSAR. The staff finds that

the design of the auxiliary feedwater system is acceptable. Acceptability of

the TMI Action Plan requirements are addressed in Section 7.8.3.

7.3.4 Failure Modes and Effects Analysis

After the staff advised the applicant that the FSAR did not provide a failure

modes and effects analyses of the safety features actuation system (SFAS) as

required by Regulatory Guide 1.70, the applicant referenced Westinghouse Topical

Report WCAP-8584. The applicant stated that the Topical Report covers the SFAS

and that the BOP design satisfies the interface criteria of WCAP-8584. The

staff finds this acceptable.

Watts Bar SER 7-9
7.3.5 IE Bulletin 80-06

IE Bulletin 80-06 calls for review of engineered safety features, with the objective of ensuring that no device will change position solely because of the

reset action. The applicant has stated that the requested reviews have been

performed, and has committed to perform the confirmatory tests requested by the

Bulletin. The applicant stated that the feedwater isolation valves, the main

feedwater check valve bypass valves, the upper tap main feedwater isolation

valves, the steam generator blowdown isolation valves, and the RHR heat exchanger outlet flow control valves do not remain in the emergency mode after

ESF reset. Design modifications are being made for these components to ensure

that the valves will remain in the emergency mode after ESF reset. The staff

finds the design modifications to be acceptable subject to its review of the

electrical schematics, which are not currently available. The staff will

report its findings in a supplement to this report.

7.3.6 Conclusions

The ESFAS includes the instrumentation and controls used to detect a plant

condition requiring operation of an ESF system, to initiate action of the ESF,

and to control its operation. The scope of the review of the ESFAS included

single-line, function logic and schematic diagrams, and descriptive information

for the ESFAS and those auxiliary supporting systems that are essential to the

operation of either the ESFAS o'r the ESF systems. The review included the

applicant's design criteria and design bases for the ESFAS and the instrumentation and controls of auxiliary supporting systems. The review also included

the applicant's analyses of the manner in which the design of the ESFAS and

the auxiliary supporting systems conform to the design criteria.

The bases for acceptance in the staff review has been the conformance of the

applicant's designs, design criteria, and design bases for the ESFAS and

necessary auxiliary supporting systems to the GDC and applicable Regulatory

Guides, BTPs and industry standards as listed in Table 7.1 of the FSAR.

During reactor operation, the basis for ESFAS acceptability will be the

successful completion of the overlapping tests performed on the initiating

systems. Checks of process indications verify operability of the sensors.

Analog checks and tests verify the operability of the analog circuitry from

the input of these circuits to and including the logic input relays, except

for the input relays associated with the containment spray functions, which

are tested during the solid-state logic testing. Solid-state logic testing

also checks the digital signal path from and, including logic input relay

contacts, through the logic matrices and master relays, and performs continuity

tests on the coils of the output slave relays. Final actuator testing operates

the output slave relays and verifies operability of those devices that require

safeguards actuation and that can be tested without causing plant upset. A

continuity check is performed on the actuators of the untestable devices.

Operation of the final devices is confirmed by control board indication and

visual observation that the appropriate pump breakers close and automatic

valves have completed their travel.

Watts Bar SER 7-10
The basis for acceptability for the ESF interlocks will be control board indication of proper receipt of the signal upon introducing the required input at

the appropriate set point.

The staff has audited the following and found them acceptable: conformance to

system redundancy and diversity; single failure; both electrical and physical

separation; identification of control boards, equipment, cables and cable trays;

and system testing and inoperable status surveillance criteria.

On the basis of its review, the staff finds that the ESFAS conforms to the

applicable regulations, guides, BTPs and industry standards and is acceptable,

subject to confirmation of the items identified above.

7.4 Systems Required for Safe Shutdown

7.4.1 System Description

The applicant states that securing and maintaining the plant in safe condition

can be achieved by appropriate alignment of selected systems that normally serve

a variety of operational functions. The capabilities that the selected systems

must provide to maintain a safe shutdown are

(1) boration

(2) adequate supply of auxiliary feedwater

(3) residual heat removal

The systems and components that are required to be functioning to achieve and

maintain hot shutdown include

(1) auxiliary feedwater pumps

(2) charging and boric acid transfer pumps

(3) essential raw cooling water pumps

(4) component cooling water pumps

(5) instrument air compressors

(6) reactor containment fan cooler units

(7) diesel generators

(8) charging flow control valves

(9) letdown orifice isolation valves

(10) auxiliary feedwater control valves

(11) pressurizer heater control

The following variables need to be monitored and indicated:

(1) water level for each steam generator

(2) pressure for each steam generator

(3) pressurizer water level

(4) pressurizer pressure

The controls for all of the equipment and the indicators listed above are in

the main control room. In addition, an auxiliary control room is provided that

allows the plant to be maintained in a hot shutdown condition or taken to cold

shutdown should the main control room become uninhabitable. Transfer switches

allow transferring the controls from the main control room to the auxiliary

Watts Bar SER 7-11
control room. Placing a transfer switch in the auxiliary position gives an

audible alarm in the main control room. Systems requiring infrequent alignment

have their auxiliary controls at the motor control centers.

7.4.2 Shutdown from Auxiliary Control Room

The auxiliary control room provides a means to shut down the reactor in case

the control room is uninhabitable. The staff questioned the applicant about

the need for a test to prove the adequacy of design of the auxiliary control

room controls and instrumentation to shut down the reactor safely and to maintain the plant in this condition. In response, the applicant stated that a

test to shut down the reactor using instrumentation to the auxiliary control

room was performed on Sequoyah Unit 1 from an initial power level of 30 percent.

The test was successful, and the cold shutdown was reached in 36 hours as

required by BTP RSB 5.1 using safety-grade equipment. Because the design of

the systems required for safe shutdown of the Watts Bar plant is very similar

to that of Sequoyah, the applicant does not consider it necessary to perform a

similar test at Watts Bar. The staff finds this acceptable.

During the site visit, the staff noted that steam generator level is not

displayed on the auxiliary control panel. Although the auxiliary controllers

are of the indicating type, they are located on the infrequently used transfer

switch panels'behind the auxiliary panel. The applicant has agreed to display

level information for each steam generator on the L-1O auxiliary control panel

by the time of fuel loading. The staff finds this acceptable.

7.4.3 Conclusions

The systems required for safe shutdown include the sensors, initiating circuitry,

logic elements, interlocks, redundant features, actuated devices, and auxiliaries

that provide the instrumentation and control functions that prevent the reactor

from returning to criticality and provide means for adequate residual heat

removal from the core, containment, and other vital components and systems.

The scope of the review of the systems required for safe shutdown included the

single-line and schematic diagrams and the descriptive information for these

systems and for the auxiliary systems essential for their operation.

The basis for acceptability has been the conformance of the applicant's designs,

design criteria, and design bases for the systems required for safe shutdown

and their essential supporting auxiliaries to the GDC, applicable Regulatory

Guides, BTPs, and industry standards, as listed in Table 7-1 of the FSAR, and

the review and acceptance of similar systems.

On the basis of its review, the staff finds that the systems required for safe

shutdown are acceptable.

7.5 Safety-Related Display Information

7.5.1 System Description

The function of the safety-related display instrumentation is to provide

information to the reactor operator to enable the operator to perform required

Watts Bar SER 7-12
safety functions during normal, abnormal, and postaccident conditions. The

instrumentation for monitoring the operating conditons during normal operation

(including anticipated operational occurrences) covers the following systems:

(1) nuclear instrumentation

(2) reactor coolant system

(3) reactor control system

(4) containment

(5) feedwater and steam systems

The monitored variables include neutron flux, flows, temperatures, fluid levels,,

pressures, and control rod positions.

The postaccident monitoring system (PAMS) instrumentation enables the operator

to take correct action during the course of a Condition I, II, III, or IV

accident and during the postaccident recovery. The variables that are

monitored by the PAMS instrumentation are

(1) reactor coolant temperature (T and T

(2) reactor coolant pressure ht cold

(3) pressurizer water level.

(4) pressurizer water and vapor temperature

(5) refueling water storage tank level

(6) boric acid tank level

(7) essential raw cooling water flow

(8) steamline pressure

(9) steam generator water level

(10) containment pressure

(11) containment sump temperature

(12) containment airborne radiation

In general, a minimum of two redundant instrumentation channels with separate

indicators is provided for each variable listed above, with one of the channels

recorded or logged on the plant computer. Single channels only are provided

for the pressurizer water and vapor temperatures. Also one channel only per

header is provided for the essential raw cooling water flow, with two'headers

per unit. For the containment airborne radiation monitoring, one-monitor is

provided in the containment upper compartment and one in the lower compartment.

7.5.2 Postaccident Monitoring System

The PAMS provides information readouts to enable the operator to perform

required manual functions and to determine the effect of manual actions take

after a reactor trip as a result of a Condition II, III, or IV event, as

defined in Chapter 15. The applicant did not use Regulatory Guide 1.97,

"Instrumentation for Light-Water-Cooled Nuclear Power Plants To Assess Plant

and Environs Conditions During and Following an Accident," for the design

because the design preceeded the publication and implementation dates of the

Guide. The applicant developed his own design criteria using applicable

requirements of IEEE Standard 279-1971. These include the requirements for

redundancy (either duplicate or functionally related channels), isolation,

separation, Class lE power,.qualification, and the capability for verifying

the operability of the monitoring channels.

Watts Bar SER 7-13
The staff finds that the safety-related display instrumentation is acceptable

for initial plant operation. For the long term, the staff has proposed, as

Commission policy, that conformance to these requirements be addressed in the

broader context of the requirements for emergency response capability. This

would include the evaluation of designs and implementation schedules for the

safety parameter display system, control room design review, upgraded emergency

operating procedures, technical support center, operational support center, and

the emergency response facility in addition to Regulatory Guide 1.97. Based on

the review of the instrumentation provided for postaccident monitoring, the

staff concludes that there is substantial conformance to Regulatory Guide 1.97.

Therefore, this item will be addressed in the broader context of subjects noted

above.

7.5.3 IE Bulletin 79-27

The staff requested that the applicant review the adequacy of emergency

operating procedures to be used by control room operators to attain safe shutdown upon loss of any Class 1E or non-Class 1E bus supplying power to safetyor nonsafety-related instruments and to control systems. This issue was

addressed for operating reactors through IE Bulletin 79-27. In a letter dated

March 1, 1982, the applicant responded to the concerns of IE Bulletin 79-27.

The effects on the plant and the ability to reach and maintain cold shutdown

upon loss of power to the individual Class IE and non-Class IE instrument buses

were evaluated. The applicant has determined that loss of power to any one of

the buses would not prevent reaching and maintaining cold shutdown from an

initial full power condition. Also, in the course of this evaluation, the.

alarms and indications are available to the operator in the main control room.

The appropriate emergency procedures are being prepared for use at the Watts

Bar plant. The applicant concluded that no deficiencies existed based on the

capability to achieve shutdown conditions using plant procedures. The staff

agrees with this conclusion and finds it acceptable.

7.5.4 Conclusions

On the basis of its review, the staff finds that the safety-related display

instrumentation conforms to the applicable regulations, Regulatory Guides,

Branch Technical Positions, and industry standards, and is acceptable.

7.6 All Other Systems Required for Safety

7.6.1 System Description

The other systems required for safety are

120-V AC and 125-V DC Vital Plant Control Power System

Four electrically and physically independent channels of 120-V ac vital power

are provided. Each channel consists of an inverter and a distribution panel.

Each channel has access to a normal, a standby, and a maintenance supply, and

is supplied by a separate battery. The four 125-V vital dc power batteries

are located in individual rooms to provide physical separation.

Watts Bar SER 7-14
Residual Heat Removal System Isolation Interlock

There are two motor-operated isolation valves in series in the inlet line from

the reactor coolant system to the residual heat removal system. In addition,

bypass valves are provided for each of those valves to provide an alternate

path in case of a valve failure. The isolation valves are normally closed.

Interlocks, provided by pressure monitoring channels, prevent these valves from

opening if the pressure is greater than 425 psig, and close the valves

automatically when the pressure increases above 750 psig.

Refueling Interlocks

Electrical interlocks are provided for minimizing the possibility of damage to

the fuel during fuel-handling operations. The interlocks are either redundant,

have a backup of another interlock, or are monitored.and actuate an alarm on

failure.

Upper Head Injection Interlocks

Redundant upper head injection (UHI) lines are provided with four hydraulically

operated accumulator isolation valves, two in each of the two lines. Their

function is to prevent the nitrogen gas from entering the reactor coolant

following the injection of the borated water from the accumulator. The valves

are normally open during operation and close on a low-level signal from the

appropriate accumulator water level channel. Following valve closure, a motordriven gagging device may be inserted in the valve operator to prevent reopening

of the valve on loss of the hydraulic pressure to the operator. The valve

position is indicated by lights operated by valve stem limit switches. An alarm

in the main control room, operated by a valve position limit switch, is actuated

if the valves are not open when the reactor coolant pressure is raised to about

1900 psig.

Accumulator Motor-Operated Valves

The cold-leg accumulator valves are motor-operated normally open valves

controlled from the main control board. Interlocks open these valves automatically on safety injection signal "S" or when the pressure is above the safety

injection unblock pressure (with the main control switch in "Auto" position).

The interlocks also prevent the valves from closing as long as the "S" signal

is present, and an alarm in the main control room will be activated if the

valve is not open when required. The valve position, open and closed, is

indicated by lights at the manual control station.

7.6.2 Residual Heat Removal System Bypass Valves

A bypass valve is provided for each of the residual heat removal system inlet

line isolation valves to provide an alternate path in case an isolation valve

fails to open when required. During normal plant operation, the bypass valves

are closed and the power to the bypass valve operators is locked out to prevent

inadvertent opening, but the valves can be opened by use of handwheels. In

the present design, valve position indication is provided at the breaker cabinet

but not in the main control room. After the staff expressed concern that the

valves could be opened without a warning to the operator, the applicant agreed

Watts Bar'SER 7-15
to provide printout and audible alarming of the valve position in the main

control room. In FSAR Amendment 47, the applicant stated that the design was

modified so that RHR letdown bypass valves are alarmed on the plant computer

with an audible alarm and an automatic printout in the main control room when

either of the valves is not in its fully closed position. The bypass valves

are also interlocked with RCS pressure transmitters PT-405 and PT-403,

respectively. The staff finds this design acceptable.

7.6.3 Upper Head Injection Manual Control

The UHI system includes two accumulators, one filled with borated water, the

other with pressurized nitrogen gas. During normal plant operation, the

contents of the two accumulators are separated by a membrane in the line connecting them and their pressure, the check valves in the lines connecting the

water accumulator to the reactor coolant system (RCS) will open, and the borated

water will be forced into the RCS by the nitrogen gas.

In addition to the check valves, each of the two lines from the water accumulator

to the RCS has two hydraulically operated valves, connected in series. One of

the functions of these valves is to prevent the nitrogen gas from entering the

RCS following injection of the borated water. This is accomplished automatically

by closing all valves on low accumulator water level signal.

The hydraulic valve operators are served by a common nonsafety-grade hydraulic

system. This system includes a reservoir for the hydraulic fluid and a pump

to charge the hydraulic accumulators used to close the valves on low level.

The pump also provides the motive power to position the valves individually by

manual control to their open or closed positions for periodic testing and

during plant startup and shutdown operations. There are, however, no provisions for manual control of the isolation valves using the safety-grade

hydraulic accumulators. After the staff expressed concern to the applicant,

he agreed to provide a manual actuation switch in the main control room for

each valve. In FSAR Amendment 47, the applicant stated that the design was

modified to include a manual switch located on the UHI hydraulic panel

in the main control room in parallel with the low accumulator water level

signal for manual initiation of valve closure. Each low level channel with

its associated manual closure switch is physically separated and electrically

isolated from the other channels. The staff finds the design acceptable.

7.6.4 Protection Against Spurious Actuation of Motor-Operated Valves

The applicant has identified in the FSAR the motor-operated valves whose

spurious operation could result in the loss of a system safety function.

Means have been provided to preclude such.spurious actuation. The design

consists of modified control circuits for these valves to ensure that no single

failure will be able to energize the opening or closing coils of the valve

operator. To protect against inadvertent operation of a valve by the operator,

hinged clear protective covers are provided on the main control board panel

over each respective control room switch. To guard against a single fault in

the electrical control circuit resulting in an inadvertent operation of a valve,

contacts from the control switches are. provided in both sides of the power

supply to the opening and closing coils. This design is acceptable to the staff.

Watts Bar SER 7-16
7.6.5 Overpressure Protection During Low Temperature Operation

The overpressure protection during low temperature operation is provided by automatic actuation of the pressurizer PORVs. In its review of'the automatic

control logic for these valves, the staff found that a failure resulting in a

high ouput signal from either of the two auctioneers would prevent both of the valves from opening when needed.- Because no indication of the auctioneered

safety output signals is provided to the operator, such a failure could remain

undetected. Furthermore, even if such a failure would be detected, the system would remain inoperative because no capability to manually arm the system to

replace a failed permissive signal from the auctioneer is provided. By letter

dated February 12, 1982, the applicant has committed to install switches on the

main control board for the operator to manually arm this system. The manual

arming will be included in the operating procedure when the reactor coolant

temperature is equal to or below the set point and before beginning the filling operation. The system arming will be reset when the system is brought back

above the system temperature set point for arming. The staff will define the

arming set point in the Technical Specifications, and the instruments for

overpressure protection will be under periodic surveillance test. The staff

finds this design acceptable subject to its review of the updated drawings and

FSAR descriptions to be submitted by the applicant. The staff will report on

its final conclusions in a supplement to the SER.

7.6.6 Valve Power Lockout

BTP PSB-18 addresses power lockout during normal reactor operation for valves

whose inadvertent operation could affect plant safety. In the Watts Bar design, this requirement is satisfied by adding a special lockout breaker in the power feed to the valve below the main circuit breaker. It also is required that

redundant valve position indication must be provided to the reactor operator regardless of the power lockout. For all such valves, redundancy of position indication is provided by independently powered limit switches mounted on the

valve stems, which actuate annunciators on the control board when the valves

are not correctly positioned for ESF actuation. The staff finds this design acceptable.

7.6.7 Cold Leg Accumulator Valve Interlocks and Position Indication

A motor-operated isolation valve is provided between each safety injection tank

and the reactor coolant (primary) system. -The valve receives an-opening signal when either the primary coolant system pressure exceeds the safety injection unblock pressure as specified in the Technical Specification, or when injection signal "S" is present with the main control switch in "Auto" position. Administratively, manual block of the automatic opening of the valve at pressures above the safety injection unblock pressure can be achieved by the operator by depressing and holding the control switch in the "Close" position. The manual

block is required to perform periodic test of the valve leakage when the system is at pressure. After the valve is closed, the operator-removes power. After

the test, the valve is reopened by administrative procedures. The valve also

is closed during plant shutdown.

As noted above, the power to the valve is removed after closing for testing and during'shutdown, and after the valve is opened during normal plant opera- tion. This is accomplished by opening the lockout breaker in the power feed

Watts Bar SER 7-17
to the valve operator (see Section 7.6.6 above). Gear and stem limit switches

are separately powered. The valve position indication in the control room is

redundantly available regardless of the power lockout to the valve.

The staff concludes that the designs of the cold-leg accumulator isolation

valve interlocks and the valve position indication are in accordance with the

requirements of BTPs ICSB-4 and PSB-18 and are acceptable.

7.6.8 Automatic Switchover from Injection to Recirculation Mode

The injection mode of emergency core cooling is initiated by the safety

injection signal "S" described in Section 7.3.1 of this report. In the injection mode all pumps take suction from the refueling water storage tank (RWST);

in the recirculation mode the pumps take suction from the containment sump.

The realignment is accomplished automatically. The actuation signal is generated

by a coincidence of the following: two-out-of-four RWST level measurement

channels are below low-low level set points, two-out-of-four sump level measurement channels are above high level set point, the safety injection signal is

present, and the valve control switch is in the "Auto" position.

In response to a staff request, the applicant provided additional information

confirming conformance to the requirements of IEEE Standard 279. The redundancy requirements are met by four level measurement channels for both the RWST

and the sump. Each of the four channels-in each of the two groups is supplied

from a separate power source. Two-out-of-four coincidence logic is used, and

separate train A and train B switchover actuation signals are generated.. Train

separation and independence are maintained from-level sensors to the actuated

equipment. The level measurement analog circuitry and the solid-state logic

are testable. The safety injection signal that provides an interlock to the

switchover actuation logic is retained by latching it; it is not removed by

action of the main safety injection reset that is used by the operator per

emergency procedures to remove the safety injection signal to other equipment

prior to realignment to the recirculation mode. The staff finds the design of

the automatic switchover from the injection to the recirculation mode acceptable.

7.6.9 Conclusions

On the basis of its review, the staff finds that the systems covered in

Section 7.6 conform to the applicable regulations, Regulatory Guides, BTPs and

industry standards, and are acceptable subject to confirmation of the items

identified above.

7.7 Control Systems Not Required for Safety

7.7.1 System Description

The systems specified by the applicant as not required for safety are

Reactor Control System

This system enables the plant to follow automatically the load changes and to

maintain reactor coolant average temperature (T avg) within established limits.

Watts Bar SER 7-18
Rod Control System

This system adjusts the reactor power by manual or automatic control of the

control rod position. In the automatic mode, the full-length rod control

system receives rod speed and direction signals from the T control system avg

(see above). In the manual control mode, the rods are moved at a prescribed

fixed speed. The system provides the operator with rod position indication,

and it alarms if the rod position deviates from preset limits or if the

required core reactivity shutdown margin is not available as a result of

excessive ,rod insertion.

Plant Control System Interlocks

These interlocks prevenit withdrawal of rods when the limits of DNBR or kw/ft

are approached, and these inhibit automatic turbine load change as required by

the NSSS.

Pressurizer Pressure Control System

This system maintains the pressurizer pressure at an established level by

operating heaters and sprays in the pressurizer and by controlling the PORVs.

Pressurizer Water Level Control System

The function of this system is to control the pressurizer water level as a

function of average coolant temperature by actuating the charging and letdown

systems.

Steam Generator Water Level Control

Each steam generator is equipped with a three-element flow controller that

adjusts the feedwater control valve as required to maintain a programmed water

level which is a function of nuclear power. In:addition, the water pump speed

is varied to maintain a programmed pressure differential between the steam

header and the feed pump discharge header.

Steam Dump Control System

This system permits the plant to accept a step decrease in load of 50 percent

without causing reactor trip by dumping the excess steam to the condenser

and/or to the atmosphere. It also permits bringing the plant to equilibrium

no load conditions without-actuation of safety/relief valves following a reactor

trip, and it-maintains the plant at no-load conditions to permit a controlled

cooldown.

Incore Instrumentation

The incore instrumentation includes thermocouples at fixed core positions and

provides information on neutron flux distribution by using movable neutron

detectors that can be positioned anywhere along the length of selected fuel

assemblies.

Watts Bar SER 7-19
Boron Concentration Measurement System

This system provides information on the boron concentration in the RCS.

Safety System Status Monitoring System

Safety system status monitoring is addressed in Section 7.7.2.

7.7.2 Safety System Status Monitoring System

FSAR Section 7.7.1.3.6 described a status monitoring system (SMS) that was to

be provided to meet the intent of Regulatory Guide 1.47," Bypassed and

Inoperable Status Indication for Nuclear Power Plant Safety Systems." As

described in that section, this system consisted of a small dedicated computer

with a cathode-ray tube (CRT) display in the main control room (MCR).

After the issuance of NUREG-0696, and the draft of NUREG-0700, which emphasizes

good human factors engineering in MCRs, the applicant decided to re-evaluate

the approach 'for handling the Regulatory Guide 1.47 requirements.

There are functional similarities between Regulatory Guide 1.47 and NUREG-0696

concerning the information to be displayed in the MCR. Both documents call

for high level plant status information to be presented automatically to the

MCR operator.

Regulatory Guide 1.47 calls for status information on a system basis to ensure

that safety systems are ready to operate in the event they are called upon.

NUREG-0696 calls for plant status information at a high functional level to

assist the MCR personnel in evaluating the overall safety status of the plant.

Both types of information need to be provided using good human factor principles

so plant personnel may rapidly understand the data.

In addition to the similarity in function there is also commonality in portions

of the data base needed for these two requirements.

Therefore, primarily based on the following reasons, the applicant determined

that the Regulatory Guide 1.47 function and the NUREG-0696 function should be

incorporated into the same computer system:

(1) similarity in functions

(2) intergrated operator interface for the MCR

(3) commonality in the data base

A common computer system will provide an integrated approach to data collecting,

data processing, and data display.

Because of the timing of the NUREG-0696 requirements and the extended durations

required for developing and checking out computer software, the Regulatory

Guide 1.47 function will not be operational in this system by fuel loading.

However, the applicant will have implemented an administrative procedure to

handle the intent of Regulatory Guide 1.47 in the interim. This procedure has

been implemented at previous operating plants and proven to be successful.

Watts Bar SER 7,20
Plant administrative procedures require that

(1) The alignment of all systems and components important to safety be verified

prior to unit startup. (Equipment important to safety is defined as the

RCS (pressure boundary components) and associated pressurizer and pressure

relief systems; the residual heat removal system; ESF systems; ESFs electric

power systems; and cooling water systems necessary to operate the above

systems.)

(2) Changes in the alignment of any system important to safety be recorded on

the applicable status sheet of the configuration control system.

(3) Shift personnel being relieved communicate information on any abnormal

plant condition including temporary conditions.

(4) System operability be demonstrated before a system is returned to service.

(5) Approval by the shift supervisor or his/her representative be received

prior to the performance of any activitiy on any systems important to

safety or any activity that may affect systems important to safety. The

shift supervisor or his/her representative is notified when an activity

authorized to be performed on a system important to safety is completed

or a change occurs in the scope of the activity.

Plant operating instructions require completion of a startup checklist before

unit startup. This checklist is used to verify correct alignment of all

systems important to safety. In addition, alignment of systems important to

safety are reviewed each shift. Any time a critical component is changed from

its normal position or condition, an entry is made on the applicable status sheet

of the configuration control system. Panel checklists are reviewed each shift

to verify that proper panel alignment exists for all systems important to

safety.

It is TVA's opinion that this verification function can be performed adequately

by an assistant unit operator (AUO) and that the use of licensed unit operators

is not necessary. TVA contends that the AUO has sufficient training and familiarity with plant systems to ensure correct system alignment and that this

policy will allow the licensed operator to remain in the control room.

Systems in which second verification of system alignment has been required are

(1) auxiliary feedwater

(2) emergency core cooling

(3) emergency gas treatment

(4) essential raw cooling water

(5) reactor coolant

(6) component cooling water

(7) containment spray

(8) residual heat removal

(9) emergency diesel generators

(10) upper head injection

(11) spent fuel pit coolant

Watts Bar SER 7-21
In addition to the administrative procedures just described, the safety channel

bypasses are annunciated in the MCR and safety system-valve alignment monitor

lights are provided in the MCR. These as well as other indications and alarms

in the MCR provide information to assist the operators in determining if any

safety system is bypassed or inoperative.

The applicant'has also revised the description of the system in FSAR

Amendment 44 to define the criteria applicable to the future system. The staff

finds that these interim measures, which are comparable to the provisions in

presently operating plants, provide adequate treatment of the concerns of

Regulatory Guide 1.47 until the first refueling outage. The staff review of

safety system status monitoring will be integrated with the emergency response

facility review. The operating license will be conditioned to require

conformance with Regulatory Guide 1.47 before startup from first refueling.

7.7.3 Volume Control Tank Level Control System

The volume control tank (VCT) is a part of the chemical and volume control

system (CVCS). The CVCS provides several services to the RCS, such as maintenance of programmed water level in the pressurizer, supplying water to the

reactor coolant pump seals, coolant purification, and others. In addition,

the centrifugal charging pumps of the CVCS also serve as the high-head safety

injection pumps in the emergency core cooling system (ECCS).

A potential concern has been raised by the NSSS supplier, Westinghouse Electric

Corporation, for an adverse control and protection system interaction. A single

random failure in the VCT level control system could cause the letdown flow to

be diverted to the liquid holdup tank. Without operator intervention, the VCT

could empty, causing the operating charging pump (three pumps are provided,

one is normally operating) to be damaged as a result of a loss of suction fluid.

Assuming that one charging pump is out of service as allowed by Technical

Specifications, there would be only one remaining undamaged pump available,

and the ECCS would not meet the single-failure criterion.

In response to the concern raised by Westinghouse, the applicant has reviewed

the design of the VCT level control system for the Watts Bar plant. Two level

measurement channels are involved in the control of VCT water inventory. One

of these channels provides a control signal for the diversion valve to the holdup tank; the other provides a signal to the automatic makeup control system.

The downscale failure of either of these channels would not affect any safety

function. The effect of an upscale (high) failure of the level measurement

channels would differ somewhat for each channel. In either case the letdown

would be diverted to the holdup tank. Automatic makeup, however, would be

retained in case of an upscale failure of the channel used in the control of

the diversion to the holdup tank. The effect of this failure would be a slight

increase, in the VCT level over a period of time. An upscale failure of the

level channel used for the control of the makeup would result in a loss of the

makeup, and the water level in the VCT would continue to decrease. In either

case, an alarm would alert the operator who would take actions required by

procedures. As stated by the applicant, the controls involved with VCT

inventory maintenance are among the most operated and observed controls on the

main control board. All operators are cognizant of VCT inventory system failure

Watts Bar SER 7-22
modes and required responses. Based on the applicant's statements, the staff

considers the concern raised by Westinghouse on the VCT control system failure

satisfactorily resolved.

7.7.4 Pressurizer and Steam Generator Overfill

For the pressurizer, three water level monitoring channels are provided and

arranged in a two-out-of-three logic to trip the reactor on high level. One

of these three channels is used for the pressurizer level control. Similarly,

three level monitoring channels are provided for each of the steam generators

and are arranged in a two-out-of-three high-level logic for the feedwater

isolation. Again, one of these channels is used for level control. Because a

failure in the common part of the channel selected for control would affect

both control and protection functions, the staff was concerned about overfilling

resulting from the control system demanding maximum feedwater flow and the

protection system not providing the protection required.

In response to the staff concern, the applicant provided the results of analyses

of the consequences of possible overfill. For the pressurizer, four cases were

presented (with or without pressurizer spray available, and combinations of

charging pump operation). In all cases, the time for filling the pressurizer

after a high-level alarm was more than the 10-min operator time limit.

The steam generator could be filled in approximately 4 min after the high-level

alarm set point is reached. Analysis, however,.showed that no failure of the

main steamline components would result because of overfilling. Also, no safety

concern with the turbine was identified. Furthermore, in addition to the three

level measurement channels provided by the NSSS supplier to close all feedwater

valves as described above, the applicant has provided two level monitoring

channels arranged in a two-out-of-two logic to close the main feedwater isolation valve.

Based on the information above, the staff finds the concern of overfilling

resolved.

7.7.5 IE Information Notice 79-22

The staff requested the applicant to determine whether the harsh environments

associated with high energy line breaks might cause control system malfunctions

and result in consequences more severe than those of Chapter 15 analyses, or

beyond the capability of operators or safety systems.

The applicant has performed a systematic evaluation of the environmental

effects resulting from high energy line breaks inside and outside containment

upon nonsafety-related systems for the Sequoyah plant. The conclusion of that

evaluation was that a safe shutdown can be achieved even if a postulated event

is compounded by environmentally induced inappropriate control system actuation.

The applicant stated that the control and logic functions of the Watts Bar plant

are identical to the Sequoyah plant; therefore, the Sequoyah evaluation is

applicable to Watts Bar. In FSAR Amendment 47, the applicant further identified

the instruments and alarms available for the operator to take appropriate actions.

Watts Bar SER 7-23
Based on its review of the applicant's analysis of consequential failure of

control system components as a result of high energy line breaks and the

operator's action to preclude unacceptable failures, the staff considers this

matter resolved.

7.7.6 Multiple Control System Failures

The staff requested that the applicant (1) identify any power sources, sensors,

or sensor impulse lines that provide power or signals to two or more control

systems and (2) demonstrate that failures of these power sources, sensors, or

sensor impulse lines will not result in consequences outside the bounds of the

Chapter 15 analyses or beyond the capability of-operators or safety systems.

The applicant submitted a response describing the results of a study to determine

the effect on the plant of

(1) loss of power to all control systems powered by a single power supply

(2) failure of each instrument sensor which provides a signal to two or more

control systems

(3) break of any sensor impulse line which is used for sensors providing

signals to two or more control systems

The applicant provided a summary of the events resulting from each postulated

failure and identified the specific Class II Chapter 15 analysis that delineates

the bounding consequences of the failure.

The staff has reviewed the bases for the applicant's study and concludes that,

with reasonable assurance, the consequences of single failure within the control

systems are bounded by analyses in Chapter 15 of the FSAR.

Unresolved Safety Issue A-47, "Safety Implications of Control Systems," will

,address control systemdesigns and.the need for any control system design

modifications. The applicant will be required to address staff guidance which

may result from the resolution of the unresolved safety issue.

7.7.7 Conclusions

The control systems not required for safety control include such system

parameters as reactivity, primary system pressure, feedwater flow, and turbine

speed. The scope of the review included descriptive information, functional

logic, instrumentation and electrical diagrams, and partial physical arrangement

drawings. The review included the applicant's design bases and analyses.

The basis for acceptance in the review has been conformance of the applicant's

designs, design criteria, and design bases for the control systems not required

for safety to the GDC, applicable Regulatory Guides, BTPs, and industry standards

as listed in FSAR Table 7-1, as well as the staff review and acceptance of

similar systems. It has been verified that control systems not required for

safety are functionally the same as for the Sequoyah plant previously reviewed

and accepted by the staff.

Watts Bar SER 7-24
On the basis of its review, the staff finds that the control systems not

required for safety acceptable subject to imposition of the license condition

identified in Section 7.7.2.

7.8 NUREG-0737 Items

7.8.1 Relief and Safety Valve Position Indication (ll.D.3)

The TMI Action.Plan requires that the reactor coolant system relief and safety valves be provided with a positive indication of valve position in the control

room, derived from a reliable valve-position detection device or a reliable

indication of flow in the discharge pipe.

As stated by the applicant, stem-operated limit switches provide direct

indication of the pressurizer PORV position. The safety valve position is

monitored by sensing the temperature downstream of the valve and by monitoring the pressurizer relief tank temperature, pressure, and fluid levels. In addition, the applicant has provided an acoustic monitoring system for both safety- and power-operated relief valves on Unit 1 and he has committed to provide them on Unit 2 before fuel loading. As stated by the applicant, the valve

position indication, alarms, and environmental and seismic qualification of

components conform with the Action Plan requirements. On this basis, the staff

finds the design acceptable, pending confirmation of the installation of the

acoustic monitoring system on Unit 2 before fuel load.

7.8.2 Auxiliary Feedwater System Initiation and Flow Indication (II.E.1.2)

Consistent with GDC 20, the TMI Action Plan requires a timely initiation of

the auxiliary feedwater system (AFWS). Also, consistent with the requirements in GDC 13, a safety-grade auxiliary feedwater flow indication powered from

emergency buses is required.

The staff review of the applicant's response shows that the auxiliary feedwater

automatic initiation system meets the applicable requirements of II.E.1.2, and

is, therefore, acceptable. The staff also finds that redundant transmitter

loops are provided for the auxiliary feedwater flow indication for each steam

generator. The transmitters are safety-grade and are powered from separate

power sources connected to the emergency power system. The staff finds that

the automatic initiation of the auxiliary feedwater flow and the flow indication

satisfy the guidelines of NUREG-0737, and are, therefore, acceptable.

7.8.3 Proportional Integral Derivative Controller Modification (II.K.3.9)o

This NUREG-0737 item calls for implementation of a Westinghouse recommendation

to modify the PORV proportional integral derivative controller to prevent derivative pressure action from opening the PORV. Two options are provided.

The applicant has satisfied this requirement by implementing the option of

setting the derivative time constant equal to zero.

Watts Bar SER 7-25
7.8.4 Proposed Anticipatory Trip Modification (II.K.3.10)

The applicant has not proposed a change in the interlock for reactor trip on

turbine trip; therefore this item is not applicable to Watts Bar.

7.8.5 Confirm Existence of Anticipatory Reactor Trip upon Turbine Trip

(II. K.3.12)

The licensee has confirmed that the Watts Bar facility has an anticipatory

reactor trip on turbine trip, which satisfies this NUREG-0737 item.

Watts Bar SER 7-26
8 ELECTRIC POWER SYSTEMS

8.1 General

The bases for the staff's evaluation of the applicant's designs, design criteria,

and design bases for the Watts Bar electric power systems are set forth in the

SRP (NUREG-0800), Section 8.1, Table 8-1, "Acceptance Criteria and Guidelines

for Electric Power Systems." These acceptance criteria and guidelines include

the applicable GDC, BTPs, Regulatory Guides, and NUREG reports. Conformance

to the applicable GDC and guidelines cited above provides sufficient basis for

acceptance of the electric power systems.

The following subsections provide the staff's evaluation of the offsite and

onsite electric power system design and how it meets-the requirements of the

above-cited acceptance criteria. The staff also will visit the site to view

the installation and arrangement of electrical equipment and cables, to review

confirmatory electric drawings, and to verify test results for the purpose of

verifying the adequacy of the design and proper implementation of the design

criteria. The confirmatory site visit will be completed before the license is

issued, and if any problems are found, they will be addressed in a supplement

to this report.

The conclusions in the following subsections are subject to acceptable implementation of any design changes that may be required as a result of the staff's

site visit.

8.2 Offsite Electric Power System

The safety function of the offsite power system (assuming the onsite power

system is not functioning) is to provide sufficient capacity and capability to

ensure that the structures, systems, and components important to safety perform

as intended. The objective of the staff review is to determine that the

offsite power system satisfies the requirements of GDC 5, 17, and 18 and will

perform its design function during all plant operating and accident conditions.

8.2.1 Compliance with GDC 5

The applicant has met the requirements of GDC 5, "Sharing of Structures,

Systems, and Components," with respect to sharing of circuits of the preferred

power system. The two offsite preferred power circuits at Watts Bar are

shared between Units 1 and 2. In accordance with GDC 5, it must be demonstrated that this sharing of circuits will not significantly impair the Class

1E power systems' ability to perform their safety functions. In accordance

with Section 8.1.1 of IEEE Standard 308-1974, it is the staff position that as

a minimum, the capacity of each preferred circuit must be sufficient to operate

the ESF for a design-basis accident on one unit as well as those systems

required for concurrent safe shutdown on the remaining unit.

Watts Bar SER 8-1
The applicant has documented in Section 8.2.2 of the FSAR that each station

service transformer has sufficient capacity to supply the essential safety

auxiliaries of one unit under LOCA accident conditions, in addition to the

power required for shutdown of the nonaccident unit. This capacity meets the

above stated position and is acceptable.

Subsequently, the applicant by letter dated October 9, 1981, proposed a system

design change. The design change adds two additional service transformers to

improve the capability of the offsite preferred circuits. The applicant has

stated that each of these service transformers can supply the Class 1E power

system for both units, with one unit in a design-basis-accident and the other

unit in a concurrent full-load rejection, without exceeding the transformers'

self-cooling power rating. The proposed change exceeds the capacity requirements of the above-stated position and is acceptable.

8.2.2 Compliance with GDC 17

The applicant has met (except as noted below) the requirements of GDC 17,

"Electric Power Systems," with respect to the offsite power systems'

(1) capacity and capability to permit functioning of structures, systems, and

components important to safety; (2) provisions to minimize the probability of

losing electric power from any of the remaining supplies as a result of, or

coincident with, the loss of power generated by the nuclear power unit or loss

of power from the onsite electric power supplies; (3) physical independence of

circuits, and (4) availability of circuits.

The Watts Bar plant is interconnected to the electric grid system through six

161-kV transmission lines that terminate on an existing 161-kV switchyard

(Watts Bar Hydro Plant Switchyard) 1.5 mi from the plant. The 161-kV lines

enter the switchyard by way of a number of physically separate and independent

rights of way. In addition, five hydro generators and four steam-driven

generators terminate at the switchyard.

The 161-kV switchyard consists of circuit breakers, disconnect switches,

transformers, buses, and associated equipment arranged so that each incoming

or outgoing transmission line can be connected to one or to both main buses

through circuit breakers. Switchyard protective relays include transmission

line protective relays and switchyard bus differential relays. These relays

are backed up by switchyard bus breakup relays and by switchyard circuit

breaker failure relays.

Offsite power from the switchyard to the onsite Class 1E distribution system

is from two independent immediate-access circuits. Each of the two circuits

is routed from the switchyard through a 161-kV transmission line and 161-to6.9-kV transformer (common station service transformer) to the onsite Class 1E

distribution system.

The onsite Class 1E distribution system consists, in part, of two redundant

and independent 6.9-kV buses, each capable of being fed from either of the

above-described offsite circuits. Offsite power is normally supplied to the

onsite distribution system from the plants main generator through a 22.5-to6.9-kV transformer (unit station service transformer) and 6.9-kV switchgear

(unit start board) to the onsite Class 1E distribution system. For any unit

Watts Bar SER 8-2
generator trip, offsite power is automatically transferred from the normal

supply to the two preferred offsite circuits.

The results of the applicant's grid stability analysis indicated that loss of

the largest capacity being supplied to the grid, loss of the largest load from

the grid, loss of the most critical transmission line, or l~oss of both units

themselves will not cause grid instability.

The following items address concerns raised during the 'staff review and their

resolution or status.

8.2.2.1 Availability of Offsite Power Circuits

GDC 17, in part, requires that each offsite circuit be designed to be available

in sufficient time following a loss of all onsite ac power supplies and other

offsite power circuits, to ensure that specified acceptable fuel design limits

and design conditions of the reactor coolant pressure boundary are not exceeded.

In accordance with this GDC 17 requirement, it is the staff position that the

period of time that the station can remain in a safe condition assuming no ac

power must be greater than the time required to re-establish ac power from the

offsite grid to the onsite Class 1E distribution buses. Therefore, the switchyard circuit breaker control scheme should be such that any incoming transmission line, switchyard bus, or any path to the onsite safety-related distribution buses can be isolated so that ac power can be re-established to the onsite

Class 1E buses through one of the two offsite circuits. The capability to

re-establish an offsite circuit is generally achieved with separate and redundant breaker tripping and closing devices that are actuated by redundant dc

battery supplies and control power distribution systems. Air stored under pressure in accumulators or spring energy is generally used to open and/or close

breakers independent of ac power.

Control power supplies and distribution systems for the switchyard (Watts Bar

Hydro Plant Switchyard) circuit breakers and protective relaying has not been

sufficiently described in the FSAR to determine compliance with the abovedescribed position.

Subsequently, the applicant by letter dated October 9, 1981, provided a proposed revision to the FSAR describing the Watts Bar Hydro Plant Switchyard.

The switchyard is separated into two parts, with each part providing power to

one of the two preferred offsite circuits. Each part contains a 250-V battery

.with associated battery charging and distribution systems. The two parts are

separated by normally open motor-operated disconnect switches. Switchyard

breakers are opened by spring energy and closed by air stored in accumulators.

Based on the additional information, the staff concludes that the offsite,

power system circuits at the Watts Bar Hydro Plant Switchyard meet GDC 17 and

are acceptable pending documentation of the additional information in the FSAR.

8.2.2.2 Minimizing the Probability of Losing All AC Power

In accordance with GDC 17, provisions must be included in the Watts Bar design

to minimize the probability of losing electric power from any of the remaining

supplies as a result of, or coincident with, the loss of power generated by

Watts Bar SER 8-3
the nuclear power unit, the loss of power from the transmission network, or

the loss of power from the onsite electric power supplies.

The Watts Bar offsite power circuit design includes a number of automatic

transfers of power between the two offsite preferred circuits. Failure of one

circuit causes automatic transfer to the remaining offsite circuit. It is

implied in the FSAR that these transfers will temporarily overload the remaining offsite circuit until the plant operator can reduce load. The staff

expressed concern that the temporary overload will cause degraded voltages on

the onsite Class 1E buses and thus cause loss of both offsite circuits due to

a single event. With the automatic transfer design, the staff determined that

the capacity and the minimizing of the probability of losing all ac power

required by GDC 17 were not met.,

The applicant, by letter dated October 9, 1981, proposed a system design

change. The proposed change eliminates the automatic transfers between the

two preferred offsite circuits, eliminates the 30-second delay time before

transfer to the preferred circuits, and thus, eliminates the above-described

concerns. However, the proposed system design change adds two new service

transformers with related distribution system components to the offsite power

system. A system description and an analysis sufficient to demonstrate compliance with GDC 17 and 18 had not been provided by the applicant in his

October 9, 1981 submittal.

By letter dated January 7., 1982, the applicant reinstated and justified the

automatic transfer design that was assumed to be eliminated in the design

described by letter dated October 9, 1981. The applicant has been informed

that the requested description and analysis have not been provided in the

FSAR and was not contained in the information submitted by letters dated

April 15 and May 19, 1982. This item will be pursued with the applicant and

the results will be reported in a supplement to this report.

8.2.3 Compliance with GDC 18

The applicant has met (except as noted below) the requirements of GDC 18,

"Inspection and Testing of Electric Power Systems," with respect to the capa- bility to test systems and associated components during normal plant operation

and the capability to test the transfer of power from the nuclear power unit,

the offsite preferred power system, and the onsite power system.

By letter dated October 9, 1981, the applicant proposed a design change to the

offsite power system. The design change includes an automatic fast transfer

between the normal (unit station service transformer) source and the preferred

source. Information describing the capability to test this transfer during

power operation had not been provided in the October 9, 1981 submittal.

By letter dated January 7, 1982, the applicant indicated an additional design provision for automatically transferring power from common station service

transformer (CSST) C to CSST D or from CSST D to CSST C. Information describing the capability to test this transfer during power operation also has not

been provided. This item has been discussed with the applicant and the resolution will be reported in Section 8.2.2.2 of a supplement to this report.

Watts Bar SER 8-4
8.2.4 Evaluation Findings

The review of the offsite power system for the Watts Bar plant covered singleline diagrams, station layout drawings, schematic diagrams, and descriptive

information. The basis for acceptance of the offsite power system was conformance of the design criteria and bases to the GDC. The staff concludes that

the plant design meets the requirements of GDC 5, 17, and 18, and conforms to

applicable guidelines of Regulatory Guides and BTPs and is acceptable, except

as noted in the preceding sections.

8.3 Onsite Power Systems

The safety function of the onsite-power system (assuming the offsite power

system is not functioning) is to provide sufficient capacity and capability to

ensure that the structures, systems, and components important to safety,

perform as intended. The objective of the review is to determine that the

onsite power system satisfies the requirements of GDC 2, 4, 5, 17, 18, and 50

and will perform its intended function during all plant operating and accident

conditions.

The onsite power system consists of an ac power system and a dc power system.

Compliance with GDC 2, 4, 5, 18, and 50 relates to both ac and dc systems and

is evaluated in Section .8.3.3 of this report. Compliance with GDC 17 is

evaluated in 'Section 8.3.1 for the ac system and Section 8.3.2 for the dc

system.

8.3.1 Onsite-AC Power System Compliance with GDC 17

The applicant has met (except as noted below) the requirements of GDC 17 with

respect to the onsite ac system's (1) capacity and capability to permit functioning of structures, systems, and components important to safety; (2) independence, redundancy, and testability to perform their safety function assuming a

single failure; and (3) provisions to minimize the probability of losing electric

power from any of the remaining supplies as a result of, or coincident with, the

loss of power generated by the nuclear power unit or the loss of power from the

transmission network.

The ac power system for each unit is comprised of two redundant and independent

distribution system divisions, each powered by one of two redundant standby

diesel generators or by either of two offsite circuits. Each distribution

system division includes a 6.9-kV shutdown board, 480-V shutdown boards, 480-V

motor control centers, and distribution system cables to accommodate the

voltage requirements of the safety loads. The safety loads for each unit are

distributed among these two electrical divisions in such a manner that the

operation of any one of the two is all that is required to meet minimum safety

requirements. There are no automatic transfers of electrical loads between

these two redundant electrical divisions.

Each of the diesel generators and selected associated auxiliaries is located

in a separate compartment within a seismic Category 1 structure.

Each of the diesel generators has a continuous rating, 2000-hour rating, and

2-hour rating of 4400 kW, 4750 kW, and 5225 kW, respectively. Each diesel

Watts Bar SER 8-5
generator will be automatically started by an undervoltage signal from its

respective 6900-V emergency bus or a~safety injection signal. Upon loss of

offsite power, the 6900-V.emergency buses will be automatically isolated from

all supply sources, and all associated 6900-V motor loads will be tripped.

Further, all diesel generator sets will be connected automatically to their

respective emergency bus, and, under accident conditions or loss of offsite

power, the safety loads will be automatically connected in a predetermined

sequence to their respective diesel generator.

The following items address concerns raised during the staff review and their

resolution or status.

8.3.1.1 Nonsafety Loads Powered from the Class 1E AC Distribution System

Present regulatory guidelines for OL applications allow the connection of

nonsafety loads in addition to the required safety loads to the Class 1E

distribution system if it can be shown that the connection of the nonsafety

loads will not degrade the Class 1E power system below an acceptable level.

The Watts Bar design provides for the connection of both safety and nonsafety

loads to the Class 1E ac distribution system. The capacity of the diesel

generator would be exceeded if the total electrical load associated with its

attendant distribution system is connected simultaneously. To preclude this

from occurring, the design includes load shedding on loss of bus voltage and

subsequent sequencing of loads. The load shedding and sequencing design

ensures that the diesel generator and its associated distribution system

capacities will not be exceeded when there is a loss of offsite power.

Preoperational test results will demonstrate the above diesel generator

capacities and load shedding features.

In addition, when there is a station trip with offsite power being transferred

to the preferred offsite source or there is an accident signal, it appears,

based on information in the FSAR, that nonsafety-related loads are not automatically disconnected and that accident loads are automatically loaded with

the non-Class 1E loads on the Class 1E distribution system, causing possible

overload of the Class 1E distribution system and the preferred offsite circuits.

Subsequently, the applicant, by letter dated October 9, 1981, proposed a new

design that consists of dedicated transformer for each of the preferred offsite

circuits. Each transformer and circuit is capable of supplying all connected

loads, Class 1E and non-Class 1E. This meets GDC 17 and is acceptable pending

incorporation of the new design in the FSAR.

8.3.1.2 Low and/or Degraded Grid Voltage Condition

The Millstone Unit 2 low grid voltage occurrence brought into focus the potential common mode failure of redundant safety-related electrical equipment that

could result from a degraded grid voltage condition. As a result, the staff

developed various positions to ensure that the requirements of GDC 17 will be

satisfied with regard to making provisions to minimize the probability of

losing electric power from any of the remaining supplies as a result of, or

coincident with, the loss of power generated by the nuclear power unit, the

loss of power from the transmission network, or the loss of power from the

onsite electrical power supplies. These ,provisions for maintaining the independence between the offsite and onsite emergency power systems are emphasized

Watts Bar SER 8-6
in IEEE Standard 308-1974, which states that preferred offsite and the standby

onsite emergency power supplies shall not have a common mode failure between

them.

The positions that the staff have developed are being used in the evaluation

of electrical power designs for operating plants, and CP and OL applications.

The applicant was made aware of these positions, which have been incorporated

into SRP Appendix 8A as BTP PSB-1.

The applicant documented that the Watts Bar design will be modified as shown

on FSAR Figure 040.62-1 to meet BTP PSB-1. By letter dated October 9, 1981,

the applicant provided additional descriptive information to support the conclusion that the Watts Bar design, once modified, will be in conformance with

positions B-1 and B-2 of BTP PSB-1. The staff concludes that the proposed

design meets BTP positions and is acceptable. In addition, design implementation will be verified as part of the site visit/ drawing review.

In regard to positions 3 and 4, the applicant has documented that the auxiliary

power system meets these positions and that the analyses will be verified in

the preoperational testing program. This meets the staff positions and is

acceptable, pending verification of the analyses. The staff will verify the

test results.

8.3.1.3 Possible Interconnection Between Redundant Divisions and Possible

Sharing of AC Power Systems Between Units

FSAR Figure 8.1-2 shows that the redundant 6.9-kV shutdown boards for Units 1

and 2 are interconnected with the shutdown boards of Unit 2 through a shutdown

utility bus. The applicant by letter dated October 9, 1981, stated that the

subject interconnections would be eliminated. This item is, therefore, resolved.

8.3.1.4 Diesel Generator Protective Trips

In accordance with position 7 of Regulatory Guide 1.9, Revision 2 ("Selection,

Design, and Qualification of Diesel-Generator Units Used or Standby (Onsite)

Electric Power Systems at Nuclear. Power Plants"), the Watts Bar diesel generator

protective trip design provides for bypassing, as described in Section 8.3.1.1

of the FSAR. At a September 17 and 18, 1981, meeting, the staff verified with

the applicant the bypassing scheme for diesel generator protective trip as shown

on drawing numbers 45W760-82-2 and 45W760-82-5. The staff concludes, therefore,

that the design meets the guidelines of Regulatory Guide 1.9 and is acceptable.

8.3.1.5 Diesel Generator Reliability

The evaluation of this item is addressed in Section 9.5.4.1.

8.3.1.6 Diesel Generator Reliability Qualification Testing

The applicant has documented that the diesel generator sets being used at Watts

Bar were tested successfully by the supplier in accordance with BTP EICSB 2.

Because EICSB 2 has been incorporated into Regulatory Guide 1.9, the applicant

was asked to describe the degree of compliance to positions of Regulatory

Watts Bar SER 8-7
Guide 1.9 and IEEE Standard 387-1977. In addition, the staff requested that

the diesel generator qualification test report be provided for staff audit.

In regard to Regulatory Guide 1.9, Revision 2, the applicant stated that the

diesel generators were tested successfully in compliance with all positions of

the guide except positions C5, C8, C9, and C14.

Positions C5 and C9 deal with environmental and seismic qualification of the

diesel generators. These items will be evaluated in Sections 3.10 and 3.11 of

this .'eport.

Position C8 deals with surveillance systems used to facilitate trouble diagnosis. The protective trips at Watts Bar have indication to show which protective device operated but they do not have a first-to-operate indication

recommended by the Regulatory Guide. An analysis of the diesel problem must

be made to determine which device operated first. A first to operate indication may be desirable to facilitate trouble diagnosis-but is not considered of

sufficient importance to safety to require backfit. The staff, thus, concludes that the Watts Bar design is acceptable.

Position C14 deals with the sequence in which the load capability qualification

test is to be performed. Regulatory Guide 1.9 recommends that the 2-hour

short-time rated-load test be performed before the 22-hour continuous ratedload test versus after the 22-hour test as required by IEEE Standard 387-1977

and as performed by the applicant. The basis for the Regulatory Guide recommendation is that the load sequence more closely follows the loading the

diesel would experience under accident conditions. The applicant, in justification of his load test sequence, documented by letter dated October 9, 1981,

that his test sequence is a more severe test than the testing sequence that is

recommended by the Regulatory Guide. The staff agrees with the applicant's

justification and find the Watts Bar testing sequence to be acceptable.

In regard to IEEE Standard 387-1977, the applicant has stated, by letter dated

March 3, 1982 that the diesel generators were tested successfully in compliance

with all sections of the standard except for.Sections 5.4, 6.3, and 6.3.2(4).

(1) Sections 5.4 and 6.3 deal with environmental and seismic qualifications

of diesel generators. These exceptions will be evaluated in Section 3.10

and 3.11 in a supplement to this report.

(2) Section 6.3.2(4) deals with 30 of the 300 acceptance qualifications tests

being performed at normal operating temperature. The 30 tests were not

performed. The staff requires that the capability of the diesel to startat normal operating temperature be demonstrated before fuel load. The

staff will condition the license to ensure that these tests will be

performed before fuel load.

In regard to requested diesel generator qualification test report, the applicant

indicated that a report was provided by letter dated December 30, 1975, on the

Sequoyah nuclear plant docket and that the Watts Bar diesel generators are the

same as those at Sequoyah. Because the staff has been unable to locate the

subject report for evaluation as it relates to the Watts Bar design, the staff

Watts Bar SER 8-8
has requested that a duplicate of the test report be made available for its

audit verification.

8.3.1.7 Possible Interconnection Between Redundant Divisions Through the

Normal and Alternate Power to the Battery Charger

The normal power feed to each of the battery chargers is from its associated

division's ac distribution system. The alternate (manually implemented) power

feed is from the opposite division's ac distribution system, as shown on FSAR

Figure 8.1-3. Thus a possible interconnection exists between redundant divisions. To prevent an operator error that would parallel redundant sources, a.

manual transfer switch is provided that is interlocked so that when the normal

power is being supplied the alternate source switch is open. This manual

transfer switch may be acceptable, to prevent an operator error from paralleling redundant divisions, but it will not preclude a single failure of the

transfer switch from paralleling redundant divisions. It is, therefore, the

staff position that (1) the circuit breaker at the 480-V shutdown board.for

the alternate feed to the battery charger be kept open and be alarmed in the

control room when it is closed and (2) the manual transfer switch must be

alarmed in the control room if it is in the alternate supply position.

Other possible interconnection between divisions exist for each of a number of

components that are manually transferred between divisions. These. components

have been listed by the applicant in FSAR Table 8.3-10. Each of these components also must meet the above-described positions.

The applicant by letter dated October 9, 1981 documented compliance with these

positions, with the exception of alarming the breakers and transfer switch

associated with the spent fuel pit pump. Because the power source for the

pump is from two different units versus redundant divisions, the staff finds

this exception for the spent fuel pump acceptable.

The subject of interconnections is further discussed in Section 8.3.3.2 of

this report.

8.3.1.8 Application of the Single Failure Criterion to Manually Controlled

Electrically Operated Valves

With regard to safety-related manually controlled, electrically operated

valves, the staff asked the applicant to provide

(1) an evaluation of all safety-related fluid systems to identify all such

valves whose failure (that is, failure to operate on demand or undesired

spurious operation) could result in the loss of capability to perform a

system safety function

(2) a description of the means provided to meet the single-failure criterion

in safety-related fluid systems where it is identified that a single

failure, as defined above, would result in the loss of capability to

perform the system safety function

Watts Bar SER .8-9
In response, the applicant identified*17 such valves and documented in Section 7.6.6 of the FSAR that the design for these valves consists of modified

control circuits. The modified circuit utilizes redundant contacts which are

wired before and after each opening and closing coil. Based on its review of

the information provided by the applicant, the staff concluded that the above

provisions are in accordance with BTP ICSB 18 of SRP Appendix 8-A, with the

exception of redundant valve position indication.

Subsequently, the applicant stated that the method of locking out power with

the required redundant instrumentation is shown on electrical drawing 45W760-

63-2. Based on this drawing, the staff concludes that the design meets the

staff's position and is acceptable.

8.3.1.9 Physical Independence

Physical independence criteria for the redundant onsite ac power system is the

same as that for the onsite dc system and is addressed in Section 8.3.3 of

this report.

8.3.2 Onsite DC System Compliance with GDC 17

The applicant has met (except as noted below) the following requirements of

GDC 17 with respect to the onsite dc system's (1) capacity and capability to

permit functioning of structures, systems, and components important to safety;

(2) independence, redundancy, and testability to perform their safety function

assuming a single failure; and (3) provisions to minimize the probability of

losing electric power from any of the remaining supplies as a result of, or

coincident with, the loss of power generated by the nuclear power unit or the

loss of power from the transmission network.

The dc power system for each unit is comprised of two redundant and independent

distribution systems each powered by a 125-V battery or a battery charger.

Each distribution system division includes a 125-V dc distribution board and

system cables to accommodate the dc control power requirements of the safety

loads. Each division supplies control power to its associated ac distribution

system and loads. Also each division supplies power to two 120-V ac vital

instrumentation and control inverters. One of the two inverters supplies

vital loads of the opposite unit so that there is sharing of dc and ac supplies

between units for vital instrumentation-and control power. The dc power system

is also comprised of a separate diesel generator battery system associated with

each of the diesel generators.

The paragraphs below address concerns raised during the staff review and their

resolution or status.

8.3.2.1 Battery Charger Capacity

Two battery chargers, one per division, are the primary source of power for dc

loads. Each battery charger is sized as documented in the FSAR to continuously

supply the normal loads and maintain the batteries in the design maximum-charged

state or to recharge the batteries from the design discharge state within an

acceptable time interval while supplying the normal loads. This sizing for the

battery charger meets subparagraph 5.3.4 of IEEE Standard 308-1971. GDC 17,

Watts Bar SER 8-10
however, does not restrict the battery charger-supplied load to that of the

steady-state condition during normal operation. Accordingly, it is the staff

position that the capacity of the battery charger supply should be based on

the largest combined demands of the various steady-state loads (Class 1E and

non-Class 1E loads) and the charging capacity to restore the battery fromthe

design minimum-charge state to the fully charged state, irrespective of the

status of the plant during which these demands occur (position c.l.b of Regulatory Guide 1.32, "Criteria for Safety-Related Electric Power Systems for

Nuclear Power Plants").

FSAR Section 8.3.2.1.1 states that the battery chargers at Watts Bar have a

rated capacity of 200 amps, and the loading, as indicated by Tables 8.3-19

through 8.3-26 of the FSAR, appeared to be in excess of 500 amps.

However, the applicant at a September 17 and 18, 1981, meeting and by letter

dated October 9, 1981, clarified the design and showed that the worst case

loading of a dc bus is 131.94 amps. This loading is within the 200-amp

capacity of the charger, meets the above-described staff positions, and is

acceptable.

8.3.2.2 DC System Monitoring and Annunciation

The specific requirements for dc power system monitoring are derived from the

generic requirements in Sections 5.3.2(4), 5.3.3(5), and 5.3.4(5) of IEEE

Standard 308-1974, and in Regulatory Guide 1.47. In summary, these general

requirements state that the dc system components (batteries, distribution

systems and chargers) shall be monitored to the extent to show that they are

ready to perform their intended function.

Accordingly, the following indications and alarms of the Class 1E dc power

system status shall be provided in the control room:

(1) battery current (ammeter-charge/discharge)

(2) battery charger output current (ammeter)

(3) dc bus voltage (voltmeter)

(4) battery discharge alarm

(5) dc bus undervoltage and overvoltage alarm

(6) dc bus ground alarm (for ungrounded system)

(7) battery breaker(s) or fuse(s) open alarm

(8) battery charger output breaker(s) or fuse(s) open alarm

(9) battery charger trouble alarm (one alarm for a number of abnormal

conditions that are usually indicated locally)

The staff has concluded that the above-cited monitoring, augmented by the

periodic test and surveillance requirements that are included in the Technical

Specifications, provide reasonable assurance that the Class 1E dc power system

is ready to perform its intended safety function.

Compliance of the Watts Bar 125-V station and diesel generator battery system

to the above-stated position was discussed with the applicant at a September 17

and 18, 1981 meeting. By letter dated January 7, 1982, the applicant provided

additional information describing the extent to which the Watts Bar design meets

the above-stated position. In regard to the 125-V dc system, the applicant

Watts Bar SER 8-11
indicated that all of the above defined indications and alarms are provided in

the control room except battery charger output current. Battery charger output

current is monitored locally rather than in the main control room as required

by the above staff positions. In justification of the Watts Bar design, the

applicant indicated that a battery charge/discharge meter is provided in the

control room that alarms on battery discharge. This alarm provides the control

room operator with immediate indication of a problem in the battery charging

system as a result of battery charger output current malfunction. The applicant concludes and the staff agrees that this alarm, with the other indications

and alarms, provides sufficient indication that the 125-V dc power system is

ready to perform its intended safety function. The staff, thus, concludes that

the indications and alarms provided are acceptable.

In regard to the diesel generator 125-V dc battery, the applicant indicated

that only battery charger alarms are provided in the control room. The following indications are provided locally: battery current and voltage, battery

charger current, and voltage, and battery system ground detection. The staff

concludes that the indication provided is inconsistent with the indication provided for the station battery system, does not provide reasonable assurance that

the diesel generator battery is ready to perform its intended safety function,

and is, therefore, unacceptable. The staff requires that the same indications

and alarms that are provided for the 125-V station battery system be provided

for the 125-V diesel generator battery system or, as a minimum, the following

abnormal conditions shall be alarmed in the control room: (1) battery circuit

open, (2) battery charger circuit open, (3) dc bus ground fault, (4) dc bus

undervoltage, (5) dc overvoltage, (6) battery charger failure, and (7) battery

discharge. In addition, the staff requires that the following parameters also

shall be monitored: (1) dc bus voltage, (2) battery circuit input current,

(3) battery circuit output current, and (4) battery charger output current.

The license will be conditioned to ensure that these indications and alarms are

installed and monitored.

8.3.2.3 Availability of the Battery Supplies to Vital Instrument Buses

Section 5.3.3.(3) of IEEE Standard 308-1974 requires that the battery be

immediately available during normal operation and following, the loss of power

from the ac system. In addition, Section 5.4 of IEEE Standard 308-1974 requires

that vital instrumentation power be supplied in such a manner as to preserve

the vital i nstrumentation's reliability. The Watts. Bar design provides a

normal power feed to the vital instrumentation from the 480-V ac shutdown

boards through an auctioneering circuit. The auctioneering circuit switches

power supplies from the ac normal supply to the dc system supplies (battery or

battery charger) on loss of ac power. The staff was concerned that this

needed switching on loss of ac power will not ensure the vital instrumentation's

reliability. By letter dated October 9, 1981, the applicant described the

subject auctioneering circuit. The circuit allows a very smooth transfer from

the dc regulated power supply (the ac source)*to the battery power source

during a loss of ac power. Thus, the power to the vital instrumentation loads

is not interrupted during transfer, will maintain the required vital instrumentation's reliability, meets Section 5.3.3(3) of IEEE Standard 308-1974, and

is acceptable.

Watts Bar SER 8-12
8.3.2.4 Diesel Generator Battery System

Each diesel generator has been supplied with a battery system to supply control

power required by the diesel generators. The sizing of the batteries, battery

charges, and distribution system has been described in FSAR Amendment 46. A

system description and applicable design criteria have also been provided in

the FSAR.

The staff concludes that the information adequately describes the diesel

generator 125-V dc system and is acceptable. However, the requested analysis

(similar to that provided for the vital 125-V dc control power system), demonstrating compliance with the GDC and applicable Regulatory Guides and IEEE

standards has been provided by a January 7, 1982 letter from the applicant.

Pending the incorporation of the analysis in the FSAR, the staff concludes

that the analysis adequately shows compliance with the applicable GDC, Regulatory Guides, and IEEE standards, and is acceptable.

8.3.2.5 Nonsafety Loads Powered from the DC Distribution System and Vital

Inverters

Present regulatory guidelines for OL applications allows the connection of

nonsafety loads--in addition to the required safety loads--to the Class 1E

distribution system if it can be shown that the connection of the nonsafety

loads will not degrade the Class 1E power system below an acceptable level.

The Watts Bar design provides for the connection of both safety and nonsafety

loads to the Class 1E distribution system and vital instrumentation and control

inverters.

The vital instrumentation and control inverters have a rated capacity of

167 amps. The maximum load connected to any one inverter is 125 amp, as shown

in FSAR Tables 8.3-12 through 8.3-18. The inverters have the capacity to

supply all connected loads both Class 1E and non-Class 1E, and thus meet

review guidelines for a plant of this vintage and are acceptable.

The dc distribution system batteries and battery chargers provide the required

power to the Class 1E and non-Class 1E connected loads. The battery chargers

have a rate capacity of 200 amp, with the maximum possible loading on each

charger in excess of 500 amp. (See Section 8.3.2.1 of this report for the

evaluation of the capacity of the battery chargers.)

The applicant has stated in Section 8.3.2.1.1 of the FSAR that the batteries

have the capacity to supply all connected loads (Class 1E and non-Class 1E)"

for a minimum of 2 hours and that the batteries will be tested-periodically in

accordance with the Technical Specifications to ensure this capacity. This

meets review guidelines and is acceptable.

8.3.2.6 Physical Independence

Physical independence criteria for the redundant onsite dc power system is the

same as that for the onsite ac system and is addressed in Section 8.3.3.3 of

this report.

Watts Bar SER 8-13
8.3.3 Common Electrical Features and Requirements

This section presents common electrical features and requirements of the

onsite ac and dc power system that deal with distinct aspects of the onsite ac

and dc power systems.

8.3.3.1 Compliance with GDC 2 and 4

The applicant has met (except as noted below) the requirements of GDC 2,

"Design Basis for Protection Against Natural Phenomena," and GDC 4, "Environmental and Missile Design Bases," with respect to structures, systems, and

components of the onsite ac and dc power system being capable of withstanding

the effects of natural phenomena (such as earthquakes, tornadoes, hurricanes,

and floods), missiles, and environmental conditions associated with normal

operation and postulated accidents. The onsite power system and components

(1) are located in seismic Category I structures, which provide protection

from-the effects of tornadoes, tornado missiles, turbine missiles, and external floods; (2) have been given a quality assurance designation "Class 1E";

(3) have been designated to be seismically and environmentally qualified;,and

(4) are to be designed to accommodate or to be protected from the effects of

missiles and environmental conditions associated with normal operation and

postulated accidents.

The following items address concerns raised during the staff review and their

resolution or status.

8.3.3.1.1 Submerged Electrical Equipment as a Result of a Loss-of-Coolant

Accident

It is the staff's concern that following a LOCA, fluid (from the reactor

coolant system (RCS) and from operation of the emergency core cooling systems

(ECCS)) may collect in the primary containment and reach a level that may

cause certain electrical equipment located inside the containment to become

submerged and thereby rendered inoperable. Both safety- and nonsafety-related

electrical equipment is of concern, because its failure may cause electrical

faults that could compromise the operability of redundant emergency power

sources or the integrity of containment electrical penetrations. In addition,

the safety-related electrical equipment that may be submerged is also of

concern if it is required to mitigate the consequences of the accident for

both the short-term and long-term ECCS functions and for containment isolation.

The applicant in Section 8.3.1.2.3 of the FSAR and in subsequent responses

stated that "There is no Class 1E equipment required to operate during or

after a LOCA or main steamline break (MSLB) that will be submerged." This

statement may be misleading. The applicant has implied in Section 8.3.1.2.3

of the FSAR that electrical sensors, electric motors for motor-operated valves,

and electric solenoids for air-operated valves will be submerged and are

safety related. Containment isolation valves are considered Class 1E and are

submerged as indicated in Section 8.3.1.2.3 of the FSAR. The staff position

is that all Class 1E equipment (rather than only those that are required to

operate) be located above the maximum LOCA water level or that it be designed

and qualified for submerged operation.

Watts Bar SER 8-14
In regard to non-Class 1E equipment, the applicant identified a number of

nonsafety-related components in Section 8.3.1.2.3 of the FSAR that may be

submerged following a LOCA. Submergence of energized non-Class 1E components

may cause electrical shorts and opening of power supply circuit breakers.

Since the staff is uncertain as to what effect-electrical shorts may have on

the functioning of Class 1E systems and electrical equipment, it is the staff

position that all non-Class 1E equipment that is associated by being powered

by a Class 1E supply or by having its cable routed in a Class 1E raceway must

be disconnected on V LOCA signal.

By letter dated January 7, 1982, the applicant indicated that Class 1E equipment required to operate after a LOCA will be relocated above the flood level.

Class 1E and associated non-Class 1E equipment not required to operate after a

LOCA either will be (1) automatically de-energized in the event of a LOCA,

(2) disconnected from its power source during normal operation (power lockout),

or (3) analyzed to show (given proper operation of overcurrent protection)

that the resultant short circuit as a result of equipment submergence will not

degrade the Class 1E power systems.

With the exception of circuits to be analyzed, the above described relocations

and design provisions meet the staff position and are acceptable. Implementation of automatic de-energization design provisions will be verified as part

of the site visit/drawing review. Surveillance requirements for components

with power lockout provisions will be pursued as part of the Technical Specifications review.

By letter dated March 3, 1982, the applicant provided a listing of the components being analyzed with a description and results of the analysis. The

analysis concludes that protective devices are coordinated so that all shorted

components will be isolated by a protective device without tripping the circuits main power protective device. Thus, submergence of Class 1E and associated non-Class 1E equipment or circuits will not prevent Class 1E power

systems from performing their intended safety function. Based on the results

of this analysis, the staff concludes that'there is reasonable assurance that

Class 1E power systems will not be lost or degraded as a result of submergence

of Class HE and associated non-Class 1E equipment not required after a LOCA.

In regard to power systems functional aspect, the resolution regarding submergence of equipment is acceptable pending incorporation of surveillance

requirements for the protective devices in the Technical Specifications.-

8.3.3.1.2 Thermal-Overload Protective Bypass

Motor-operated valves with thermal overload protection devices for the valve

motors are used in safety systems and their auxiliary supporting systems.

Operating experience has shown that indiscriminate application of thermal

overload protection devices to the motors associated with these valves could

result in needless hindrance to successful completion of safety functions.

Regulatory Guide 1.106, "Thermal Overload Protection for Electric Motors on

Motor-Operated Valves" (November 1975), recommends bypassing during accident

conditions or properly selecting the setpoints for the thermal overloads,

supplemented with periodic testing of these devices, as acceptable methods to

be implemented in the design of motor-operated valves.

Watts Bar SER 8-15
The applicant has indicated that the Watts Bar design is in compliance with

Regulatory Guide 1.106 and that the design has been changed to include bypass

of thermal overload protective devices. This design change meets Regulatory Guide 1.106 and is acceptable. The staff will verify this aspect of the

design during the electrical drawing review to be conducted with the applicant.

8.3.3.2 Compliance with GDC 5

The applicant has met (except as noted below) the requirements of GDC 5, "Sharing of Structures, Systems, and Components," with respect to structures,

systems, and components of the ac and dc onsite power systems.

8.3.3.2.1 Sharing of DC Distribution Systems and Power Supplies

Between Units 1 and 2

Each unit is provided with four vital instrument power buses. The vital buses

receive power from the dc distribution system and power supplies. The dc

system for Unit I supplies power to vital buses I and II for Unit 2, and the

dc system for Unit 2 supplies power to vital buses III and IV for Unit 1, as

described on Figure 8.1-3 and in Section 8.3.2.5 of the FSAR. The applicant

was requested to provide a description of how the shared system design satisfies the recommendations of Regulatory Guides 1.6 ("Independence Between

Redundant Standby (Onsite) Power Sources and Between Their Distribution

Systems") and 1.81 ("Shared Emergency and Shutdown Electric Systems for MultiUnit Nuclear Power Plants"). Regulatory Guide 1.81 guidelines do not permit

sharing of dc systems. GDC 5 requires that dc systems not be shared unless it

can be shown that such sharing will not significantly impair the ablity of the

dc system to perform its safety functions.

By letter dated January 7, 1982, the applicant provided the results of an

analysis of the vital 125-V dc system with respect to position C.2 of Regulatory Guide 1.81. Based on these results, imposition of appropriate Technical

Specifications for shared dc systems, and other information documented in the

FSAR, the staff concludes that there is reasonable assurance that the sharing

will not significantly impair the ability of the dc system from performing its

safety function. The system meets the requirements of GDC 5 and is acceptable.

Surveillance requirements for shared dc systems will be pursued as part of the

Technical Specifications review.

8.3.3.2.2 Sharing of AC Distribution Systems and Standby Power Supplies

Between Units 1 and 2

In Section 8.3.1.1 of the FSAR, the applicant states that the standby power

system serving each unit is divided into two redundant power trains with-loads

connected so that failure of an entire train will not prevent the redundant

train from performing the required safety function. The applicant was asked

to provide a description of how the design satisfies position C2 of Regulatory Guide 1.81. By letter dated October 9, 1981, the applicant provided the

description of compliance. This description appears to contradict Section 8.3.1.1 of the FSAR in that Section 8.3.1.1 implies two diesels (one per

unit) can be lost rather than only one diesel out of four, as stated in the

October 9, 1981 letter. Similarly, Section 8.3.1.1 contradicts Sections 8.1.4

and 3.1.2.1 of the FSAR. Sharing of onsite power systems (ac and dc) had not

Watts Bar SER 8-16
been adequately described or analysed in Section 8.3 of the FSAR as required.

For example the essential raw cooling water (ERCW) that is needed to mitigate

an accident in Unit 1 is dependent on pumps powered from Unit 2 to meet the

single-failure criterion. This sharing of ac power systems had not been

described or analysed in Section 8.3 of the FSAR. By letter dated January 7,

1982, the applicant provided the subject description and analysis. Based on

this letter and imposition of appropriate Technical Specifications for shared

ac power systems, the staff concludes that the design meets the guidelines of

Regulatory Guide 1.81 and is acceptable pending revision of the FSAR that

reflects requirements of the shared safety systems. Surveillance requirements

for sharing of ac systems will be pursued as part of the Technical Specifications review.

8.3.3.2.3 Sharing of Raceway Systems Between Units

In Section 8.3.2.4.3 of the FSAR, the applicant states that Unit 1 train A

cables may be routed in the same raceways as Unit 2 train A cables and Unit 1

train B cables may be routed in the same raceway as Unit 2 train B cables.

Thus, raceway systems may be shared between units.

At a September 17 and 18,.1981 meeting, the applicant was requested to identify

each shared raceway-and provide the results of an analysis in accordance with

GDC 5. The analysis must demonstrate that failure of all circuits located in

each shared raceway will not significantly impair the capability of safety

systems to mitigate an accident in one unit and an orderly and safe shutdown

and cooldown of the remaining unit.

The applicant, by letter dated October 9, 1981, provided the results of an

analysis for one example where the cables associated with the Units 1 and 2

ERCW systems share a common raceway. Based on the results of this analysis,

the staff concludes that the design meets the guidelines of Regulatory

Guide 1.81 and the requirements of GDC 5, and is acceptable for this example.

For the remaining raceways that are shared between units, the staff requested

a similar analysis be provided for staff audit verification that identifies

the shared raceways with analyses results that demonstrate Compliance with

GDC 5. By letter dated January 7, 1982, the applicant stated that cables for

Units 1 and 2 of the same division may share a common raceway, but are physically separated from cables associated with the redundant division so that,

given a failure of common division A cables, division B cables will not fail.

Division B cables are sufficient to shut down both reactors given an accident

in one and safe shutdown in the other. Based on the information presented to

date, the design meets GDC 5 and is, therefore, acceptable. Confirmation of

the shared raceway design's conformance with GDC 5 will be conducted as part

of the drawing review/site visit.

8.3.3.2.4 Possible Sharing of DC Control Power to AC Switchgear

The Watts Bar design provides primary and backup control power feeds from each

dc power source to the 6.9-kV and 480-V switchgear. These power feeds allow

dc loads associated with Unit 1 division 1 to be powered from Unit 2 division

1 dc power sources,

Watts Bar SER 8-17
Possible interconnection between redundant divisions through normal and alternate power to various loads is addressed in Section 8.3.1.7 of this report.

Based on discussions with the applicant, it was the staff's understanding that

all interconnections between divisions were identified in FSAR Table 8.3-10

and will meet the staff positions defined in Section 8.3.1.7 of this report.

The subject interconnection between dc divisions is, however, not identified in

FSAR Table 8.3-10. It is the staff's requirement that all possible interconnections between redundant divisions through normal and alternate power sources

to various loads be identified in FSAR Table 8.3-10 regardless of the source

of power and will meet the'staff positions identified in Section 8.3.1.7.

With the imposition of this requirement, the staff considers this item resolved.

By letter dated January 7, 1982, the applicant indicated for the subject dc

system interconnection that the correct switch alignment will be verified by

administrative procedures every 7 days rather than being verified immediately

by alarm as required by the staff position identified in Section 8.3.1.7 of

this report.

Because incorrect switch position can only interconnect battery systems associated with the same ac power division, the staff finds the 7-day administrative procedure proposed by the applicant to be acceptable.

8.3.3.3 Physical Independence (Compliance with GDC 17)

(1) Physical Identification of Electrical Cables

In accordance with the specific identification criteria of IEEE Standard

384-1974 and Regulatory Guide 1.75, Revision 1 ("Physical Independence of

Electric Systems), it is the staff position that Class 1E cables must be

marked at a sufficient number of points (5-ft intervals) to facilitate

visual verification that the installation is in conformance with separation

criteria. Cables must be marked either before or during cable installation.

The Watts Bar design for cable identification, as documented in the FSAR,

utilizes color-coded tags at cable terminations and colored markings on

raceways at conspicuous intervals. Because cables have been marked only at

terminations rather than at a sufficient number of points, the staff cannot

conclude that there is sufficient assurance that cables have been installed

in conformance with separation criteria. Therefore, the staff required

that an independent inspection agency perform a cable-routing audit of both

Class 1E and non-Class 1E cables to ensure that cables are routed in accordance with acceptable separation criteria. The cable audit includes

10 percent of the installed cables.

The applicant, by letter dated October 9, 1981, indicated that the required cable-routing audit has been performed and was provided to the

NRC by letter dated February 24, 1981. The audit report identified

numerous violations of the Watts Bar separation criteria. Resolution of

these violations is currently underway by TVA.

(2) Associated Circuits

Section 8.3.1.4.3 of the FSAR indicates that when a non-Class 1E cable is

routed in a Class 1E tray, the non-Class 1E cable (or any cable in the

Watts Bar SER 8-18
same circuit as the non-Class 1E cable) will not be routed in a cable

tray containing the Class 1E circuits redundant counterpart. This routing does not meet the guidelines of Regulatory Guide 1.75 and IEEE Standard 384-1974. IEEE Standard 384 requires that once the non-Class 1E

cable is routed with a Class 1E circuit, the non-Class 1E cable must be

treated as an associated cable. An associated cable can only be routed

with its associated Class 1E cables. The associated cable is, thus,

separated from other non-Class 1E cables as well as the redundant Class

1E cables. The Watts Bar cable routing criteria allows non-Class 1E

cables associated with redundant Class 1E cables to share a common nonClass 1E raceway. Thus, the independence-of redundant cable systems may

be compromised through the non-Class 1E cables sharing a common raceway.

It is the staff's position that non-Class 1E cables should meet the guidelines of positions 4, 6, and 7 of Regulatory Guide 1.75 and Sections 4.5

and 4.6.2 of IEEE Standard 384-1974. By letter dated March 3, 1982, the

applicant submitted an analysis of associated circuits in accordance with

section 4.6.2(1) of IEEE Standard 384-1974. Associated circuits, not

separated from other non-Class 1E cables or protected from design basis

events, were identified and analyzed. The analysis demonstrated that

electrical faults, caused by failure of the associated cables, will not

compromise the independence of redundant Class 1E cable systems. The

analysis verified that the cable's associated protective device will

clear the imposed fault condition (in an acceptable time period) without

exceeding the 12t rating for the cable.

Based on the results-of this analysis, the staff concludes that, given

proper operation of protection devices, there is reasonable assurance that

Class 1E power systems and their distribution circuits will not be lost

or degraded as a result of failure of non-Class 1E cables routed with

Class 1E cables.

To.ensure proper operation of protective devices, the staff requires that

they be of a high quality commensurate with their importance to safety

and be periodically tested. With the imposition of this requirement as a

condition to the Watts Bar license, the staff considers this item to be

acceptable with respect to the functional aspects of electric power

systems. Surveillance requirements for the protective devices will be

reviewed with the Technical Specifications.

(3) Separation Criteria Between Class 1E and Non-Class 1E Circuits

Section 8.3.1.4.3 of the FSAR indicates that non-Class 1E cable trays are

separated from Class 1E trays by 8 in. of air space vertically and 6 in.

of air space horizontally. In addition, nonsafety circuits may be routed

in the same conduit as Class 1E circuits at terminal equipment. Separation between an open top non-Class 1E cable tray and a Class 1E conduit

has not been described in the FSAR. This separation of non-Class 1E from

Class 1E circuits does not meet the guidelines of section 4.6.1 of IEEE

Standard 384-1974. By letter dated March 3, 1982, the applicant submitted

additional information concerning separation between Class 1E and nonClass 1E circuits at Watts Bar. Separation between 1E and non-Class 1E

trays is 12 in. vertical (tray bottom to tray bottom) and 6 in. horizontal (tray side to tray side). Separation may be reduced to 9 in. vertical

Watts Bar SER 8-19
when solid bottom trays are used. Separation between conduits and open

top cable trays was not described.

In justification of the above described separation, the applicant has

documented and substantiated by reference to tests or analysis (1) that a

cable fault will be cleared by a protection device well before ignition

temperature of cable insulation is reached and (2) that an electrically

initiated fire may occur in cables routed in non-Class 1E cable trays,

but that it is only minutely probable for the fire to propagate enough to

challenge cables routed in adjacent Class 1E trays. Based on circuit

protection, results of fire tests, and fire protection modifications, the

applicant judges that the spacing of Class 1E and non-Class 1E cable

systems provides adequate separation.

Given proper operation of protective devi ces, the staff agrees with the

applicant's judgment and concludes that there is reasonable assurance

that Class 1E power systems and their distribution circuits will not be

lost or degraded as a result of failure of non-Class 1E cables that are

routed closer to Class 1E circuits than allowed by Regulatory

Guide 1.75.

To ensure proper operation of protective devices, the staff requires that

non-Class 1E cables, routed closer to Class 1E cables than allowed by

Regulatory Guide 1.75, be identified with their associated protective

devices. The protective devices shall be of a high quality commensurate

with their importance to safety and be periodically tested. With the

impos'ition of this requirement as a condition to the Watts Bar license,

the staff considers this item to be acceptable. Surveillance requirements

for protective devices will be reviewed with the Technical Specifications.

(4) Fire Protection

The fire protection system for Watts Bar is evaluated in Section 9.5.1 of

this report.

8.3.3.4 Compliance with NUREG-0737 Items

Two TMI items relating to GDC 17 are identified in NUREG-0737. These items

are II.E.3.1, Emergency Power Supply for Pressurizer Heaters, and II.G.1,'

Emergency Power for Pressurizer Equipment. The background, the NUREG position,

clarification of the positions, and the applicant's descriptions of compliance

for items II.E.3.1 and II.G.1 is provided in letters from the applicant dated

September 14, 1981, and March 11, 1982.

The staff's evaluation of each of these items follows.

Emergency Power for Pressurizer Equipment (II.G.1)

NUREG-0737 requires the applicant to ensure that the motive and control components of the PORVs and the associated block valves, and the pressurizer level

indication instrument channels are capable of being supplied from either the

offsite power source or the emergency power 'source (during a loss-of-offsitepower event). In addition, the motive and control power connections to the

Watts Bar SER 8-20
emergency buses for this equipment should be qualified in accordance with

safety-grade requirements.

The applicant has stated, in part, that the relief valves (PORVs) and their

associated block valves are powered from opposite power trains. Powering the

valves from opposite power trains meets the objective of this TMI Action Plan

Item but does not meet the recommendations of BTP RSB 5-2 for overpressurization protection while operating at low temperature. To meet the objective of

this TMI Action Plan Item (as indicated by clarification 2), the power supplies

for the PORVs and block valves can be from different emergency power sources,

(for example, PORV on dc power and the associated block valve on ac power),

both power sources emanating from the same division. This single division

orientation of power sources also is in conformance with BTP RSB 5-2 recommendations for reactor overpressurization protection at low temperature.

Therefore, the license will be conditioned to ensure that the PORV and block

valves be powered from the same power train but from different buses.

Emergency Power Supply for Pressurizer Heaters (II.E.3.1)

The applicant, in response to clarification 7 of item II.E.3.1, has stated:

"The pressurizer heaters are automatically shed from the emergency power

sources upon the occurrence of a safety injection actuation signal (SIS)."

This statement meets' clarification 7 and is acceptable; however, the applicant

also has stated: "In the event of a loss of offsite power and safety injection

signal, two backup heater banks rated at 485 kW each can be manually activated

by hand switches in the main control room." Based on this second statement,

the design appears not to meet clarification 7. By letter dated January 7,

1982, the applicant provided additional information and clarification. Based

on information in the applicant's September 14, 1981 and January 7, 1982

letters, the staff concludes that the Watts Bar design meets the guidelines of

Item II.E.3.1 of NUREG-0737 and is acceptable.

8.3.3.5 Compliance with GDC 18

The applicant has met the requirements of GDC 18 with respect to the onsite ac

and dc power system. The onsite power system is designed to be testable during

operation of the nuclear power generating station as well as during those intervals when the station is shut down.

The following items address the concerns raised during the staff's review and

their resolution:

8.3.3.5.1 Compliance with Regulatory Guides 1.108 and 1.118

The applicant, by letter dated October 16, 1981, has documented that the FSAR

will be revised to indicate full compliance'to Regulatory Guide 1.118

("Periodic Testing of Electric Power and Protection Systems"). The staff finds

this acceptable.

By letter dated October 9, 1981, the applicant has identified and provided

justification for two exceptions to Regulatory Guide 1.108 ("Periodic Testing

of Diesel Generator Units Used as'Onsite Electric Power Systems at Nuclear

Power Plants"). The first exception deals with first-out annunciation. This

Watts Bar SER 8-21
exception is evaluated in Section 8.3.1.6 of this report. The second exception

addresses position C.2.a(2): "Demonstrate proper operation for design-accidentloading sequence to design-load requirements and verify that voltage and

frequency are maintained within required limits." The applicant has stated

that loads will be sequenced as designed, except the pumps will be operated

with their miniflow connections open and not at full flow. By letter dated

January 7, 1982, the applicant provided the following additional justification:

(1) A full flow test will be done as part of the preoperational testing program;

(2) full load tests are accomplished periodically by paralleling the diesel

generator with the offsite power system; and (3) the most severe disturbance of

voltage and frequency occur when starting-a motor; thus, whether a motor is

operating at full flow or miniflow, this disturbance is the same.

The primary difference between a test using miniflow rather than full flow is

that the load at the beginning of each load step may be less than full load.

Regulatory position C.2.a(2) recommends that proper operation of the diesel

generator be demonstrated at least once every 18 months for design-accidentloading-sequence to design-load requirements. The intent of this position is

to test the sequencing of actual loads onto the diesel generator within the

correct time sequence interval. It is not the intent, even though it may be

preferable, to require that actual loads be run at full flow conditions.

Actual full design load need only be applied to the extent practical. The

staff concludes that use of the miniflow valves during the subject test (every

18 months) will provide reasonable assurance of proper operation of the diesel

generator, meets the intent of Regulatory Guide 1.108. It is, therefore,

acceptable.

8.3.3.5.2 Testing of One of Two Class 1E Power Systems Versus One of

Four Systems

Section 8.3.1.1 of the FSAR states that tests are to be performed on only one

of the two power trains per unit at any one time and that testing of one power

train does not prevent the remaining power train from performing its intended

safety function. These statements contradict other statements in the FSAR and

are misleading. This item was discussed with the applicant at a September 17

and 18, 1981, meeting. In response, the applicant, by letter dated October 9,

1981, documented that only one of the four power trains of the plant (Unit I

and Unit 2) would be tested at one time. Therefore, pending documentation of

this. commitment in the FSAR, the staff concludes that this item is acceptable.

8.3.3.6 Compliance with GDC 50

The applicant has met the requirements of GDC 50 with respect to electrical

penetrations containing circuits of the safety and nonsafety onsite power system.

GDC 50 requires, in part, that the reactor containment structures, including

penetrations, be designed so that the containment structure and its internal

compartments can accommodate, without exceeding the design leakage rate and with

sufficient margin, the calculated pressure and temperature conditions resulting

from any LOCA. Therefore, with regard to electrical penetrations, the main

objective of the staff review is to determine that the electrical penetration

assemblies are designed to withstand, without the loss of mechanical integrity,

the maximum available fault current versus time conditions that could occur given

single random failures of circuit overload protective devices as recommended by

Watts Bar SER 8-22
position Cl of Regulatory-Guide 1.63, Revision 1, "Electric Penetration Assemblies

in Containment Structures for Water Cooled Nuclear Power Plants."

The applicant, in FSAR Section 8.1.5.3 and in a subsequent response, has

indicated compliance to Regulatory Guide 1.63 with the exception of position

C1. In regard to position C1, the applicant has documented (1) that current

limiting fuses have been added in series with circuit breakers for the 480-V

low voltage circuits, (2) that redundant overcurrent protective relays energize

separate trip coils on a single protective circuit breaker for 6900 V medium

voltage circuits, and (3) the penetrations for Watts Bar are designed to

withstand without loss of mechanical integrity the maximum fault current

available at the penetration for the time required for the protective device

to clear the fault.

The information provided in the FSAR does not provide sufficient content and

justification for the staff to reach a conclusion of acceptability. Control

and instrument cables have not been addressed. Protection for the 6900-V

cables consists of one versus the required two circuit breakers. The capability of the penetration during LOCA environmental conditions to withstand the

maximum fault current available at the penetration has not been addressed.

Representative sample test results showing the fault current versus time

capability of protective devices in relation to the capability of electric

penetrations to withstand the fault current for each type of cable has not

been addressed. Testability and physicalseparation and independence of

protective devices have not been addressed. Other non-Class 1E penetration

circuits cables also have not-been addressed.

It is the staff position, in accordance with the recommendations of Regulatory

Guide 1.63, that penetration circuit overload protection systems meet the

criteria of IEEE Standard 279-1971, (single failure, testability and qualification for Class 1E circuits, and single failure and testability for non-Class

1E circuits). The staff position follows the guidelines that have been established for implementation of Regulatory Guide 1.63 for other recent OL applications. The staff requires redundant fault-Current protective devices

(circuit breakers and/or fuses) on each circuit, either of which would limit

and isolate a fault current surge to a value less than the surge for which the

penetration is qualified.

Subsequently, the applicant, by letter dated October 9, 1981, provided a

proposed revision to the FSAR describing the extent and justification of his

compliance to the guidelines of Regulatory Guide 1.63. Redundant over-current devices (breaker and fuse) have been provided for all penetration circuits

except very low energy instrumentation circuits and the 6900-V reactor coolant

pump circuits. The instrumentation circuits have currents in the milliampere

range, are current limited so that fault currents will not be a problem with

containment integrity, and are, therefore, acceptable. The reactor coolant

pump circuits (when power is supplied during normal operation from the unit

transformers) have only one rather than the required two breakers.. The one

breaker, however, has two trip coils, two relays, two current transformers,

and two separate sources of dc control power. The applicant has performed a

reliability study that shows that one breaker with two trip circuits is 98.3

percent as reliable as the two breakers in series. In addition, when power is

supplied from the common station transformer during shutdown or after unit

Watts Bar SER 8-23
trip, the Watts Bar design has the required two breakers in series. Except for

the mechanical portion of the breaker, the electrical design meets the single

failure criterion and independence requirements of Regulatory Guide 1.63.

Because experience at operating plants indicates that a large number of mechanical breaker failures have occurred, the staff finds that the justification presented for one breaker by the applicant is unacceptable. Therefore, the license

will be conditioned to ensure that redundant fault current protective devices

be provided in series in accordance with position 1 of Regulatory Guide 1.63

before fuel load.

In regard to testability of fuses, the applicant has indicated that fuses.will

be periodically checked for proper size, deterioration, and the connection's

tightness and cleanliness. In addition, the staff will require, and include

as part of its Technical Specification, periodic measurement of the fuses and

terminal connection resistance.

In addition to compliance to Regulatory Guide 1.63, the staff asked the applicant at a September 17 and 18, 1981, meeting to provide coordinated time versus

current curves of the two fault devices in relation to the electrical penetration. The staff also asked for the test report with results that substantiate

the curves provided for the penetration. By letter dated January 7, 1982, the

applicant provided the requested curves and test report. Based on information

provided, the staff concludes that the design meets the guidelines of Regulatory Guide 1.63 and is acceptable with one exception. The penetration curves

are not plotted beyond 10-sec duration. It is not clear from the curves and

test report that the thermal capability of the penetration has been adequately

demonstrated for relatively long-time-duration fault currents (10 to 1000 sec).

Thus, the staff requires a re-evaluation of the penetrations capability to

verify that the penetrations can withstand, without seal failure, the total

range of available time-current characteristics assuming a single failure of

any overcurrent protective device. The results of the re-evaluation will be

verified as part of the staff confirmatory site visit. If the results are

found unacceptable, the staff will condition the license to ensure that the

protective devices trip set point is correct so that position 1 of Regulatory

Guide 1.63 will be met.

8.3.4 Evaluation Findings

The review of the onsite ac and dc power system for the Watts Bar plant covered

single-line diagrams, station layout drawings, schematic diagrams, and descriptive information. The basis for acceptance of the onsite power systems in the

staff's review was conformance of the design criteria and bases to the Commission's regulations as set forth in the General Design Criteria. The staff concludes that the plant design meets the requirements of GDC 2, 4, 5, 17, 18,

and 50; conforms to applicable Regulatory Guides and NUREG reports; and is

acceptable, except as noted in preceding sections.

Watts Bar SER 8-24
9 AUXILIARY SYSTEMS

The staff has reviewed the design of the auxiliary systems necessary for safe

reactor operation, shutdown, or fuel storage, or whose failure might affect

plant safety-related objectives, and the way these objectives are achieved.

The auxiliary systems necessary for safe reactor operation or shutdown include

the essential raw cooling water (ERCW) system, the component cooling system

(CCS), the ultimate heat sink (UHS), essential portions of the compressed air

system, essential portions of the chemical and volume control system (CVCS),

the heating, ventilating and air conditioning (HVAC) systems for the control

room and essential portions of the reactor and auxiliary buildings, and the

auxiliary feedwater system (AFWS). The auxiliary systems necessary to ensure

the safety of the fuel storage facility include new fuel storage, the spent

fuel pool cooling and cleanup system, fuel-handling systems, and the fuel

handling building HVAC system.

The staff has also reviewed nonsafety-related auxiliary systems to verify that

their failure will not affect safety-related equipment, prevent safe shutdown

of the plant, or result in unacceptable release of radioactivity to the environment. These systems include: the demineralized water makeup system, potable

and sanitary water system, condensate storage facilities, raw cooling water

system, equipment and floor drainage system, nonessential portions of the compressed air system, nonessential portions of the CVCS, and HVAC systems for

nonessential portions of the reactor auxiliary building and the turbine

building.

9.1 Fuel Storage Facility

9.1.1 New Fuel Storage

The new fuel storage facility which is housed within the seismic Category I

auxiliary building consists of the seismic Category I storage racks and the

concrete storage vault that contains the storage racks. The new fuel storage

vault, shared by both reactor units, provides dry storage for the new fuel

assemblies. The auxiliary building protects the new fuel storage facility from

the effects of tornadoes, tornado-generated missiles, and flooding (refer to

Sections 3.4.1 and 3.5.2 of this SER). The new fuel storage facility is

located above the probable maximum flood (PMF) level and therefore will not be

affected if the auxiliary building experiences flooding. Thus, the requirements

of GDC 2, "Design Bases for Protection Against Natural Phenomena," and the

guidance of Regulatory Guide 1.29 are satisfied.

The new fuel storage facility is not located in the vicinity of any moderate

or high energy lines or rotating machinery. Physical protection for the new

fuel from internally generated missiles and the effects of pipe breaks is

provided by separation (see Sections 3.5.1.1 and 3.6.1 of this SER), thus

satisfying the requirements of GDC 4, "Environmental and Missile Design Bases."

Watts Bar SER 9-1
The facility is designed to store unirradiated fuel assemblies. Accidental

damage to the fuel would release relatively minor amounts of radioactivity that

would'be accommodated by the fuel handling building ventilation system. Thus,

the requirements of GDC.61, "Fuel Storage and Handling and Radioactivity

Control," are satisfied.

The new fuel storage racks are designed to store the fuel assemblies at a

center to center spacing of 21 in. This spacing is sufficient to ensure that

the effective multiplication factor Keff does not exceed 0.98 with fuel of the

highest enrichment in place, assuming possible sources of moderation such as

aqueous foam or mist. The value of Keff is less than or equal toý0.95, assuming

fully flooded conditions. The design of the racks precludes the insertion of

fuel assemblies in other than the prescribed locations. The racks are designed

to withstand the impact of a fuel assembly dropped from the maximum lift height

of the fuel-handling bridge crane without causing an unsafe geometric spacing

of the new fuel assemblies. Seismic Category I sectional steel covers are

provided over the new fuel vault to protect the racks and fuel from dropped

objects. Thus, the requirements of GDC 62 are satisfied.

Radiation monitoring equipment for the new fuel storage area is provided and

is evaluated in Section 12 of this SER. It satisfies the requirements of

GDC 63, "Monitoring Fuel and Waste Storage."

Based on its review, the staff concludes that the new-fuel storage facility, is

in conformance with the requirements of GDC 2, 4, 61, 62, and 63 as they relate

to new fuel protection against natural phenomena, missiles, pipe break effects,

radiation protection, prevention of criticality, and radiation monitoring, and

to the guidelines of Regulatory Guide 1.29 relating to seismic design. It is,

therefore, acceptable.

9.1.2 Spent Fuel Storage

Nuclear reactor plants include storage facilities for the wet storage of spent

fuel assemblies. The safety function of the spent fuel pool and storage racks

is to maintain the spent fuel assemblies in a subcritical array during all

credible storage conditions. In accordance with SRP Section 9.1.2 of NUREG-08O,

the staff has reviewed the compatibility and chemical stability of the materials

(except the fuel assemblies) wetted by the pool water.

The pool liner, rack structure, and fuel storage tubes are of welded stainless

steel construction with a neutron absorber sandwiched between the stainless

steel sheets. The neutron absorber is marketed under the trade name of Boraflex. Boraflex is composed of boron carbide powder in a rubber-like silicone

polymeric matrix. The pool is filled with borated demineralized water.

The stainless steel is compatible with the storage pool environment. In this

environment of oxygen-saturated borated water, the corrosive deterioration of

the type 304 stainless steel should not exceed a depth of 6.00 x 10-s in. in

100 years, which is negligible relative to the initial thickness.. Dissimilar

metal contact corrosion (galvanic attack) between the stainless steel of the

pool liner, rack lattice structure, fuel storage tubes, and the Inconel and

the Zircaloy in the spent fuel assemblies will not be significant because all

Watts Bar SER 9-2
of these materials are protected by highly passivating oxide films and are

therefore at similar potentials. The Boraflex poison material is composed of

nonmetallic materials and therefore will not develop a galvanic potential in

contact with the metal components. Boraflex has undergone extensive testing

to study the effects of gamma irradiation in various environments and to verify

its structural integrity and suitability as a neutron absorbing material.

The annular space;which contains the Boraflex is vented to the pool. Venting

of the annulus will allow gas generated by the chemical degradation of the

silicone polymer binder during heating and irradiation to escape; it will also

prevent bulging or.swelling of the inner stainless steel tube.

From the evaluation discussed above, the staff concludes that the corrosion

that will occur in the spent fuel storage pool environment should be of little

significance during the 40-yr life of the plant. Components in the spent fuel

pool are constructed of alloys that have a low differential galvanic potential

between them and have a high resistance to general corrosion, localized corrosion, and galvanic corrosion.

The staff, therefore concludes that the environmental compatibility and stability

of the materials used in the spent fuel storage pool are adequate based on test

data and actual service experience in operating reactors. On the basis of its

review (in accordance with SRP Section 9.1.2 of NUREG-0800), the staff finds

that the selection of appropriate materials of construction by the applicant

meets the requirements of GDC 61, having a capability to permit appropriate

periodic inspection and testing of components, and GDC 62, preventing criticality by maintaining its structural integrity of components, and is, therefore,

acceptable.

The spent fuel storage facility, which is housed in the seismic Category I

auxiliary building, consists of the spent fuel storage racks, the lined spent

fuel pool, and the fuel transfer canal/spent fuel pool gates, all of which are

seismic Category I. The spent fuel pool, shared-by both reactor units, will

provide high density underwater storage for 1312 fuel assemblies. The auxiliary

building, in which the spent fuel pool is located, is not only protected against

seismic events, but it is also protected against flooding up to elevation

728.0 ft. Flooding of the lower elevations of the auxiliary building will not

adversely affect safe storage of spent fuel. Thus, the requirements of GDC 2

and the guidelines of Regulatory Guides 1.13, 1.29, 1.102 ("Flood Protection

for Nuclear Power Plants"), and 1.117 ("Tornado Design Classification") are

satisfied.

The fuel pool is not located in the vicinity of any high energy lines or

rotating machinery. Physical separation is provided to protect the spent fuel

from internally generated missiles and the effects of pipe breaks (see Sections

3.5.1.1 and 3.6.2 of this SER). Thus, the requirements of GDC 4, "Environmental

and Missile Design Bases," are satisfied.

The high density fuel racks are designed to store the fuel assemblies in an

array that limits the effective multiplication factor to 0.95 or less, assuming

the array is fully flooded with nonborated water. The racks are designed to

preclude the inadvertent placement of a fuel assembly in other than a design

Watts Bar SER 9-3
storage-location. The racks can withstand the impact of a dropped fuel assembly

without unacceptable damage to the fuel and can withstand the maximum uplift

forces exerted by the fuel-handling bridge crane.

The applicant has stated that, under normal plant operating conditions, no

loads will be carried over the spent fuel pool. Thus, the requirements of

GDC 61 and 62 and the guidelines of Regulatory Guide 1.13 concerning fuel

storage facility design, are satisfied.

The design of the storage pool includes a leak chase network behind the pool

liner welds to detect and collect leakage through the welds, a pool water level

monitoring system, and radiation monitoring systems with indications and alarms

in the control room. These features satisfy the requirements of GDC 63.

Based on its review, the staff concludes that the spent fuel storage facility

is in conformance with the requirements of GDC 2, 4, 61, 62 and 63 as related

to protection against natural phenomena, missiles, pipe break effects, radiation

protection, prevention of criticality, and monitoring provisions, and the

guidelines of Regulatory Guides 1.13 and 1.29 concerning design and protection

against seismic events, and is, therefore, acceptable.

9.1.3 Spent Fuel Pool Cooling and Cleanup System

The spent fuel pool cooling and cleanup system (SFPCCS) was reviewed in accordance with the SRP (NUREG-0800, July 1981) ultilizing information presented in

FSAR Section 9.1.3 as well as responses to requests for additional information

and generally available reference materials.

The spent fuel pool cooling and cleanup system is designed to remove the decay

heat generated by stored spent fuel assemblies from the fuel pool water. Additional functions of the system include clarifying and purifying the water in

the spent fuel pool, transfer canal, and refueling water storage tanks. The

system design incorporates two essential trains of equipment, each train consisting of one fuel pool pump and one heat exchanger, a single separate nonessential purification train with two refueling water purification pumps, two

refueling water purification filters, one spent fuel pool filter and one demineralizer, and a single nonessential skimmer train consisting of one pump,

filter, and strainer. An essential spare fuel pool pump is provided and is

capable of operation in either cooling train.

Chemical analysis downstream of the filter and ion exchanger will be made by

batch sampling at 1-week intervals for chemical impurities, twice per month

for pH and boron concentration, and once a montht for gross beta and gamma

radioactivities. The applicant provided the radioactivity and chemical impurity

limits to be maintained in the pool water. The ion exchanger will be replaced

when the chemical impurities exceed the specified limits. A high pressure drop

in the filter is the basis for changing the cartridge in the filter. Area

radiation monitors are also provided.

The staff determined that the spent fuel pool cleanup system (1) provides the

capability and capacity of removing radioactive materials, corrosion products

and impurities from the pool water, and thus meets the requirements of GDC 61

as it relates to appropriate filtering systems for fuel storage; (2) is capable

Watts Bar SER 9-4
of reducing occupational exposure to radiation by removing radioactive products

from the pool water, and thus meets the requirements of Section 20.1(c) of

10 CFR 20, as it relates to maintaining radiation exposures ALARA; (3) confines

radioactive materials in the pool water into the demineralizer and filters,

and thus meets Regulatory Position C.2.f(2) of Regulatory Guide 8.8, "Information Relevant To Ensuring that Occupational Radiation Exposures at Nuclear

Power Stations Will Be As Low As Is Reasonably Achievable," as it relates to

reducing the spread of contaminants from the source; and (4) removes suspended

impurities from the pool water by filters, and thus meets Regulatory Position

C.2.f(3) of Regulatory Guide 8.8, as it relates to removing crud through

physical action. On the basis of this evaluation, the staff concludes that the

spent fuel pool cleanup system is acceptable.

With two trains of the spent fuel pool cooling and cleanup system in operation,

the cooling system is designed to maintain the pool water temperature at or

below 120'F following a back-to-back one-third core refueling of both units.

The FSAR also stated that a pool water temperature of 150*F can be maintained

when only one cooling train is in operation following a full core discharge

subsequent to a normal refueling. This temperature is based on a 60-day delay

period before the fuel assemblies of the full core are placed i/n the spent fuel

pool.

The system design includes the capability for periodic visual inspection and

preventive maintenance. Periodic testing of components will be performed in

accordance with plant procedures. All connections to the spent fuel pool are

designed to preclude possible siphon draining of the pool water. The safetyrelated cooling and cleanup system (CCS) provides cooling water to the fuel

pool heat exchanger and transfers its heat to the ERCW system.

During flood operations the ERCW system will replace the CCS as the heat sink

for the spent fuel cooling water because the CCS will not be operable. (See

Sections 9.2.1 and 9.2.2 for additional information.) Thus, the requirements

of GDC 44, GDC 45, and GDC 46 are satisfied.

Also, if a flood occurs during a refueling, cooling of the refueling canal and

the spent fuel pool will be accomplished by circulating water through the RHR

system via the spent fuel cooling pumps. These pumps are situated above the

design-basis flood and will not be affected during flooding of the auxiliary

building. Each fuel pool pump can be powered from an independent emergency

power bus. Normal makeup to the fuel pool is provided by the demineralized

water system. Seismic makeup is provided from the refueling water storage tank

(RWST). The system incorporates control room alarmed pool water level, water

temperature, and building radiation level monitoring systems. Thus, the

requirements of GDC 61 and 63 and the guidelines of Regulatory Guide 1.13

concerning spent fuel pool design are met.

The essential portions of the system are housed in the seismic Category I flood

and tornado-missile-protected auxiliary building. Because flooding of the

auxiliary building is expected during a design-basis flood, equipment required

for flood/post-flood operation has been identified and is adequately protected.

The spent fuel pool cooling pumps are such equipment and are located above the

PMF level. The SFPCCS is designed to Quality Groups B and C and seismic

Category I requirements, with the exception of the purification and cleanup

Watts Bar SER 9-5
portion, which is designed to Quality Group D and nonseismic Category I requirements. Therefore, the design satisfies'the requirements of GDC 2 and the guidelines of Regulatory Guides 1.13, 1.26, and 1.29.

The system is housed in seismic Category I structures and is separated from

other high and moderate energy piping systems, thus satisfying the requirements

of GDC 4.

The applicant has provided a comparison of the system to that which was designed

for Sequoyah. The staff has reviewed the information provided in the comparison and concurs with the applicant's findings. The staff reaffirms that the

conclusions stated in the Sequoyah SER (NUREG-O011) are applicable to Watts

Bar.

The staff further reviewed the system for compliance with the applicable GDC,

Regulatory Guides, and BTPs and concludes that the system design meets the

requirements of GDC 2, 4, 44, 45, 46, 61, and 63 with respect to protection

against natural phenomena and missiles, cooling water capability, inservice

inspection, functional testing, fuel cooling and radiation protection, and

monitoring provisions, and the guidelines of Regulatory Guides 1.13, 1.26,

and 1.29 relating to the system's design, quality group and seismic classification. It is, therefore, acceptable.

9.1.4 Fuel-Handling System

The fuel-handling system (FHS) provides the means of handling and transporting

both new and spent fuel. The FHS consists of equipment necessary for the safe

handling of new and spent fuel assemblies, the spent fuel cask, and for the

safe disassembly, handling, and reassembly of the reactor vessel heads and

internals during refueling operations. The handling of fuel during refueling

is controlled by a series of interlocks to ensure that fuel-handling procedures

are maintained. The design precludes the possibility of raising spent fuel

assemblies above the safe shielding depth.

The entire system is housed within the seismic Category I and tornado-missile--

protected auxiliary building and reactor building. All equipment associated

with the FHS, with the exception of the fuel-handling machines located in the

refueling and transfer canal, is located above the probable maximum flood level.

The reactor building polar crane and the auxiliary building crane are designed

to seismic Category I requirements so that they will not fail in a manner which

results in unacceptable consequences such as fuel damage or damage to safetyrelated equipment. The design thus satisfies the requirements of GDC 2 and the

guidelines of Regulatory Guide 1.29.

The spent fuel cask pit, located beside the spent fuel pool, is separated from

the pool by a reinforced concrete wall and a fuel pool gate. The walls of the

spent fuel pool extend to the fuel handling floor to prevent the dropped cask

from tipping into the pool. The cask is transported along a prescribed travel

path by the auxiliary building crane to minimize the possibility of damage to

spent fuel or safety-related equipment located below the floor in the event of

a cask drop. The crane is interlocked in a.fashion as to prevent the carrying

of loads over the spent fuel pool during plant operation. Thus, the requirements of GDC 61 and the guidelines of Regulatory Guide 1.13 and BTP ASB 9-1,

Watts Bar SER 9-6
"Overhead Handling Systems for Nuclear Power Plants," are satisfied for handling of the spent fuel cask.

Load drop analyses for the reactor building polar crane and auxiliary building

crane have not been completed.-These analyses are under way as part of the

applicant's response to NUREG-0612, "Control of Heavy Loads at Nuclear Power

Plants," which was transmitted to the applicant for action by generic NRC

letters dated December 22, 1980, and February 3, 1981. NUREG-0612 resolved

Generic Task A-36 and provides guidelines for necessary changes to ensure safe

handling of heavy loads once the plant becomes operational. The applicant will

submit the results of the review against NUREG-0612 guidelines by August 1982.

By letter dated October 28, 1981, the applicant has committed to perform and

implement the interim actions outlined in Enclosure 2 of the December 22, 1980,

generic letter. These interim measures deal with safe load paths, procedure,

operator training, and crane inspections, testing, and maintenance. With the

implementation of these interim actions, the staff concludes that the heavy

load handling system is acceptable for licensing without completion of the

staff's evaluations of the applicant's response to NUREG-0612.

The applicant has provided a comparison of the system to that which was designed

for Sequoyah. The staff has reviewed the information provided in the comparison and concurs with the applicant's findings and reaffirms that the conclusions

stated in the Sequoyah SER are applicable to Watts Bar.

The staff has further reviewed the system for compliance with the applicable

GDC, Regulatory.Guides, and BTPs and concludes that the fuel handling system

conforms to the requirements of GDC 2 and 61 with respect to its protection

against natural phenomena and safe fuel handling and the guidelines of Regulatory Guides 1.13 and 1.29 with respect to crane interlocks and seismic classification, and is, therefore, acceptable. The staff further concludes that

implementation of the interim actions of NUREG-0612 before the OL is received

provides reasonable assurance of safe handling of heavy loads until NUREG-0612

can be fully implemented. The system is, therefore,,acceptable.

9.2 'Water Systems

9.2.1 Essential Raw Cooling Water and Raw Cooling Water Systems

The essential raw cooling water (ERCW) system and the raw cooling water (RCW)

system supply cooling water to the plant from the ultimate heat sink (discussed

in Section 9.2.5 of this SER).

The safety-related seismic Category I ERCW system provides essential auxiliary

support functions to the engineered safety features (ESF) of the plant and

supplies cooling water to various heat loads in both the safety and nonsafety

portions of each unit. Provisions are made to ensure a continuous flow of

cooling water to those systems and components necessary for plant safety either

during normal operation or under accident conditions. The systems and components serviced by the ERCW are the component cooling heat exchangers, containment spray heat exchangers, emergency diesel generators, emergency makeup for

the component cooling system, control building air conditioning system, auxiliary building ventilation coolers (for ESF equipment), containment ventilation

system, air compressors, reactor coolant pump (RCP) motor coolers, control rod

Watts Bar SER 9-7
drive ventilation coolers, and instrument room coolers. Additionally, the ERCW

supplies cooling water to the residual heat removal (RHR) heat exchangers,

spent fuel cooling heat exchangers, RCP thermal barrier heat exchangers, and

the ice machine refrigeration condenser only during flood mode operation when

flooding exceeds elevation 728.0 ft msl. The ERCW also serves as the seismic

backup water supply to the auxiliary feedwater system if the normal supply from

the condensate storage tank is unavailable.

Supply water for the ERCW is from the UHS (Tennessee River) via the ERCW intake

pumping station (IPS). This seismic Category I flood-and-tornado-missileprotected structure, located at the end of an intake channel approximately

800 ft from the reservoir, houses the same quantities of ERCW equipment (eight

pumps, four travelling screens, strainers, and screen wash pumps) that the

Sequoyah "new', ERCW intake structure does, in addition to housing the nonsafetyrelated RCW pumps and safety-related high pressure fire protection pumps. The

ERCW system configuration at Watts Bar is the same-as that of Sequoyah with

respect to the redundant shared trains, of which either train can safely shut

down both units with the opposite train inoperable. Four independent supply

headers direct the ERCW flow to the plant from the ERCW pumping station; two

separate discharge headers return the heated ERCW to the yard. The headers

are arranged and fitted with isolation valves such that a critical crack in

either header can be isolated and will not jeopardize the safety functions of

the system. Although the operation of only two pumps on the same header is

sufficient to supply all cooling water requirements for the two-unit plant for

unit cooldown, refueling or postaccident operation, eight ERCW pumps are

supplied for the facility--four of which automatically start on an ESF actuation signal (two pumps per header), thus providing full redundancy. All pump

motors, screen motors, screen wash pump motors, backwashing strainer motors,

and motor-operated valves can be powered from emergency sources. The design of

the ERCW system ensures that system function is maintained assuming a single

active component failure coincident with a loss of offsite power. Thus, the

requirements of GDC 5 and 44 are met.

The ERCW system is designed to Quality Group B and C and seismic Category I

requirements. However, during construction, portions of piping leading to

HVAC coolers or chillers which service areas containing essential equipment

were not installed to Quality Group B or C requirements. The staff determination of the system's acceptability in this regard is discussed in Sections 3.2.1

and 3.2.2 of this SER. Components of the system are located in seismic Category I structures that provide protection against tornadoes, tornado-generated

missiles, and flooding (see to Sections 3.4.1 and 3.5.2 of this SER). ERCW

piping between the IPS and the auxiliary building and between the auxiliary

building and the ERCW standpipe is seismic Category I. The ERCW piping from

the standpipe to the cooling tower is TVA Class H (nonseismic). The ERCW

piping is buried to protect the piping from tornado missiles. Pump motors,

valve operators, and controls are located above the level of the PMF in the

seismic Category IIPS.

In a December 24, 1981, letter, the applicant has proposed the use of cement

mortar lining in the carbon steel yard piping portion of this system to inhibit

corrosion. Based on testing information and data provided by the applicant,

the staff concludes that the cement mortar lining will not fail during an SSE

in a manner that will prevent the ERCW system from delivering the required

Watts Bar SER 9-8
cooling water flow to the necessary systems and components needed for safe

shutdown.

The ERCW pumps and pump motors are housed in a Category I structure that

shields against horizontal and vertical tornado missiles. Although the roof of

the structure shields the pump motors from vertical missile, the motors are

exposed to the effects of the environment. The ERCW pump motors are designed

and weatherproofed to operate in such environmental conditions. The staff

finds, this acceptable. Pumps and pump motors inside the pumphouse are physically separated from each other to preclude coincident damage to redundant

equipment from pipe rupture, equipment failure, and missile generation.

The ERCW travelling screens are located in the same Category I structure that

houses the ERCW pumps. These screens are protected from the effects of tornadogenerated missiles, and are designed to function in an exposed atmospheric

environment. The applicant has stated that, through the use of administrative

procedures, all travelling screens will be turned on during periods when

conditions for surface water freezing water present. Thus, the staff concludes,

that adverse weather conditions will not affect the cooling function of the

ERCW system. Although the ERCW system operates during normal plant operations,

components and controls are periodically inspected and tested. All portions of

the system are accessible to permit inservice inspections and testing as

required. Thus, the requirements of GDC '2, 4, 45, and 46 are satisfied.

Based on its review, the staff concludes that the ERCW system meets the requirements of GDC 2, 4, 5, 44, 45, and 46 with respect to protection against natural

phenomena, missiles and environmental effects, the sharing of essential systems,

heat removal capability, inservice inspection and functional testing, and the

guidelines of Regulatory Guide 1.102 with respect to flood protection.

The raw cooling water (RCW) system which services the balance of plant cooling

and makeup requirements (cooling water to the turbine-generator, auxiliary

equipment within the turbine building and nonessential air conditioning equipment within the auxiliary building, makeup to the water treatment plant, a

source of makeup to the condenser circulating water system, the primary nonqualified source of cooling water for the ice condenser system, and miscellaneous water requirements such as maintenance and cleaning) is a shared,

nonsafety-related system. The RCW system is designed so that no component can

adversely affect the function of any safety-related system. The RCW pumps

located in the ERCW intake pumping station are completely separated from any

safety-related equipment. RCW system piping within seismic Category I structures are seismically qualified to the extent required to ensure that a safe

shutdown earthquake in combination with normal operating conditions will not

cause flooding, water impingement, or damage as a result of-falling onto

safety-related equipment, thus conforming to the guidelines of Regulatory

Guide 1.29, Position C.2, "Seismic Design Classification."

The applicant has provided a comparison which states that the design of the

system is essentially the same as that of Sequoyah. The only major difference

in the RCW systems between the two plants is that Watts Bar maintains the use

of seven RCW pumps as compared to five pumps used at.Sequoyah. This difference

is the result of the lesser capacity of the Watts Bar pumps. The staff has,

Watts Bar SER 9-9
reviewed the information provided in the comparison and concurs with the

applicant's findings.ý

Based on its review, the staff concludes that a failure within the nonseismic

RCW system will not compromise the ability of safety-related systems to perform

their intended functions; is in conformance with the guidelines of Regulatory

Guide 1.29, Position C.2; and is, therefore, acceptable.

9.2.2 Component Cooling System (Reactor Auxiliaries Cooling Water System)

The component cooling system (CCS), a safety-related system designed to seismic

Category I and Quality Group B and C requirements, provides cooling water to

various plant components, and rejects heat to the ERCW system. It serves as

an intermediate cooling loop between radioactive or potentially radioactive

heat sources and the ERCW. The systems served by the CCS are: RHR, CVC,

safety injection system (SIS), waste disposal, spent fuel pool cooling and

cleaning, sampling, and containment spray.

The CCS consists of five component cooling system pumps, four thermal barrier

booster pumps, three heat exchangers, two surge tanks, and one component cooling system pump seal water collection unit, and associated valves, piping and

instrumentation. The component cooling system is a shared system of two

trains, each train having the capability to provide the maximum cooling requirement for both units under all design-basis plant conditions. A "spare" train

is available and can be remotely aligned to supplement either the Unit 1 or the

Unit 2 train when it is required. These equipment trains are sufficiently

independent so that a single active failure of any one component in one train

will not preclude safe plant operations in either unit. Thus, the requirements

of GDC 5 and 44 have been satisfied. Components will be periodically tested

and inspected during plant operation to ensure operability, thus satisfying

the requirements of GDC 45 and GDC 46.

The system is located in seismic Category I structures and is protected from

the effects of internally and externally generated missiles. With the exception of the component cooling booster pumps, the CCS is not protected against

the effects of flooding. Flood waters exceeding plant-grade entrances will

flood the lower elevations of the auxiliary building, which will result in the

flooding of CCS components. During a site visit, it was noted that the component cooling booster pumps were located at an elevation below the PMF level.

Because these pumps may be required to operate during flooding, the applicant

has committed to relocate them above the PMF level before receipt of the

operating license. During flood operation, the ERCW will provide the

necessary cooling water supply to all essential and safeguards components

normally serviced by the CCS. Thus, the requirements of GDC 2 and 4 have

been satisfied.

During normal plant operations, three component cooling system pumps are

required to produce the minimum flow for both units. Three pumps are also

required for shutdown. In response to staff concern regarding loss of component-cooling system flow to the RCPs as a result of a single failure in the

common supply line, the applicant has'demonstrated that there is sufficient

time between the annunciation of the loss of component cooling and pump damage

for required operator action to mitigate the consequences of such an event.

Watts Bar SER 9-10
Low flow indications and alarms for each RCP are provided in the control room.

High bearing temperature alarms are also provided to indicate the loss of

component cooling system flow. The applicant's evaluation has shown that the

time span between the indication/alarm of the loss of CCS flow and required RCP

trip is greater than 10 minutes. Actual tests conducted by Westinghouse on

similar reactor coolant pumps confirm that reactor coolant pumps are capable of

operating for 10 minutes with loss of cooling water to the oil coolers without

pump damage.

The applicant has provided a comparison of the system to that designed for

Sequoyah. The staff has reviewed the information provided in the comparison

and concurs with the findings. The staff thus reaffirms that the conclusions

stated in the Sequoyah SER (NUREG-O011) are applicable to Watts Bar.

The staff has further reviewed the system for compliance with the applicable

GDC, Regulatory Guides, and BTPs and concludes that the CCS meets the requirements of GDC 2, 4, 5, 44, 45, and 46 with respect to its protection against

natural phenomena, missiles and environmental effects, sharing of systems, heat

removal capability, inservice inspection and functional testing, and the

guidelines of Regulatory Guides 1.26, 1.29, 1.102, and 1.117 with respect to

the system's quality group and seismic qualification, and flood and tornado

missile protection. Thus, it is acceptable.

Effect of Loss of AC Power on Pump Seals (II.K.3.25)

NUREG-0737 required determination of the consequences of a loss of cooling

water to the reactor coolant pump seal coolers.

In a submittal dated September 14, 1981, the applicant stated that water

systems cooling the RCP seals are automatically powered by the emergency diesel

generators in the event of loss of offsite power. The staff finds this

acceptable.

9.2.3 Demineralized Water Makeup System

The nonsafety-related (Quality Group D, nonseismic Category I) demineralized

water makeup system provides treated and demineralized water to various systems

and components within the plant. Makeup water for the condensate storage tank

is supplied directly from the water treatment plant. The system, which is

shared between both units, serves no safety-related function. Thus, the system

design is consistent with the requirements of GDC 2 and the guidelines of

Regulatory Guide 1.26.

Pipe supports for those portions of the system routed through the reactor

building, auxiliary building, or control building are designed for seismic

loading to prevent damage to adjacent safety-related equipment necessary for

safe shutdown, thus conforming to the guidelines of Regulatory Guide 1.29,

The applicant's comparison of the demineralized water makeup system for Watts

Bar and Sequoyah indicated that the system design for the two facilities differ

only as to equipment characteristics, because of supply vendors and system

capacity differ. The water distributions and system operations and effluent

quality are essentially the same for the two plants.

Watts Bar SER 9-11
The applicant has provided a comparison of the system to that which was designed

for Sequoyah. The staff has reviewed the information provided in the comparison

and concurs with the findings.

Based on its review, the staff concludes that the design of the demineralized

water and makeup system meets the requirements of GDC 2 as related to its protection against natural phenomena and the guidelines of Regulatory Guides 1.26

and 1.29, Position C.2, as related to quality group and seismic classification.

9.2.4 Potable and Sanitary Water Systems

The potable water system provides potable water to all facilities on the site.

The system 'is designed to Quality Group D, nonseismic Category I requirements.

Portions of the system housed within seismic Category I structures are designed

not to fail in a way that would adversely affect the function of safety-related

equipment. The system design at Watts Bar is essentially identical to that of

Sequoyah. Two differences are noted however. Watts Bar has a lower daily

potable water requirement (7500 gpd) compared to Sequoyah (12,000 gpd). Also,

the source of potable water for the two plant facilities is different. The

potable water supply for Watts Bar is provided by two wells and a pumping and

storage tank network. This system is designed, constructed, and operated by

the applicant. The Sequoyah potable water supply is provided from a utility

district water system.

The sanitary water systems of the two plant facilities are essentially identical. There are no cross connections between the potable and sanitary water

systems and potentially radioactive systems and, therefore, inadvertent contamination is prevented. Protection from flooding for safety-related equipment

resulting from failure of the system is discussed in Section 9.3.3 of this SER.

Failure of the system does not affect plant safety as described above;

Based on its review, the staff concludes that the system meets the requirements

of GDC 2 with respect to the need for protection against natural phenomena and

meets the guidance of Regulatory Guides 1.26 and 1.29 regarding its seismic

and quality group classification, and is, therefore, acceptable.

9.2.5 Ultimate Heat Sink

The ultimate heat sink (UHS) provides the heat dissipation capability for the

reactor and its essential auxiliaries through the ERCW and component cooling

water systems during normal operation, shutdown, refueling, and accident conditions. The UHS for the Watts Bar facility is the Tennessee River, including

associated retaining structures and the channel connecting the source with, but

not including, the intake structure. Water required for all plant operations

and cooling is channeled to the essential raw cooling water intake pumping

station (ERCW-IPS). Distribution is via the ERCW system and the raw cooling

water (RCW) system and the condenser circulating water (CCW) system. Makeup

for the CCW is provided by the ERCW and RCW system discharge (see also Sections 9.2.1 and 10.4.5 of this SER).

The applicant performed a heat transfer analysis assuming conservative worst

site meteorology to verify the performance capability of the UHS. The analysis

shows that the UHS is capable of providing sufficient cooling water for plant

Watts Bar SER 9-12
systems for 30 days under all normal and accident conditions. The intake

channel, which extends from the IPS to the original river bed, is periodically

dredged to provide clean access to the river under low flow conditions. The

normally exposed and submerged portions of the channel are dredged and are

seismically qualified to preclude the possibility of channel blockage as a

result of earth or mud slides. The UHS is also designed to accommodate the

most severe combination of events considered credible to occur, such as simultaneous occurrence of safe shutdown earthquake, a LOCA in one unit and safe

shutdown of the other, loss of offsite power, and loss of upstream and/or

downstream dams, either individually or concurrently. The design described

above ensures that adequate heat removal capability to maintain plant safety

is provided by the UHS for all modes of operation including accidents coincident with a single active failure. Thus, the requirements of GDC 44, and the

guidelines of Regulatory Guide 1.27, Positions C.1, C.3, and C.4 regarding the

ability of the UHS to maintain proper system temperature under all modes of

operation are met.

The immenseness of the UHS in conjunction with its geographic location precludes the impairment of cooling capability as a result of tornado-missile

impact and environmental conditions, and-additional protection is not required.

Therefore, the UHS is consistent with the requirements of GDC 2 and 4.

The applicant has provided a comparison of the.Watts Bar system to that which

was designed for Sequoyah. The staff has reviewed the comparison and concurs

with the applicant's findings. Thus the staff reaffirms that the conclusions

stated in the Sequoyah SER (NUREG-O011) are applicable to Watts Bar.

The staff has further reviewed the system for compliance with the applicable

GDC, Regulatory Guides, and BTP, and concludes that the UHS design conforms to

the requirements of GDC 2, 4, and 44 with respect to the need for protection

from natural phenomena and missiles, and heat removal capability, and the

guidelines of Regulatory Guide 1.27 as related to the functional and design

requirements of the UHS, and is, therefore, acceptable.

9.2.6 Condensate Storage Facilities

The nonsafety-related (Quality Group D, nonseismic) condensate storage facility

stores and supplies treated water for various plant functions; it includes all

components and piping associated with the system from the storage tank to the

points of connection or interfaces with other systems. The staff review has

determined that the system is capable of fulfilling the normal operating

requirements of the facility for storage of condensate with the necessary

component redundancy. The system was evaluated and found to have no functions

necessary for achieving safe reactor shutdown or for accident prevention or

mitigation.

The two outdoor condensate storage tanks (CSTs), which reserve 200,000 gal for

each auxiliary feedwater (AFW) system, are not seismic Category I, flood, or

missile protected because the tanks are not required for maintaining plant

safety. Though the CSTs, are the preferred source for the AFW, the seismic

Category I backup water source is the ERCW. A CST rupture or overflow would

allow the water to drain to the turbine building sump or the Tennessee River

by way of the holding pond. Protection for safety-related equipment against

Watts Bar SER 9-13
internal flooding caused by failure of the system is discussed in Section 9.3.3

of this SER. Thus, the system is consistent with the requirements of GDC 2

and the guidelines of Regulatory Guide 1.26, and Position C.2 of Regulatory

Guide 1.29.

The applicant has provided a comparison of the system to that which was designed

for Sequoyah. The staff has reviewed the information provided in the comparison

and concurs with the applicant's findings.

Based on its review, the staff concludes that the system is consistent with

the requirements of GDC 2 with respect to the need for protection against

natural phenomena, conforms to the guidelines of Regulatory Guides 1.26 and

1.29 as related to the quality group and seismic classification, and is,

therefore, acceptable.

9.3 Process Auxiliaries

9.3.1 Compressed Air System

The compressed air system, designed to provide a reliable supply of oil-free

air for pneumatic instruments, valves, controls, and services, is a shared

system and is divided into two subsystems: (1) the'station control and service

air system and (2) the auxiliary control air system. The station control and

service air system is nonsafety related (Quality Group D, nonseismic Category I)

and is not designed to function during accident conditions. This system

normally supplies air to both trains of the auxiliary control air system, but

is automatically isolated when the system's output pressure drops below a

preset value. The auxiliary control air system, which is required to achieve a

safe reactor shutdown or to mitigate the consequences of an accident, is

comprised of two fully redundant trains and is sized and equipped so that the

required air supply is provided for both units under all design-basis accident

conditions. Redundancy and train separation have been provided to ensure that

the system can maintain its necessary safety functions assuming a design-basis

event followed by a single active failure. The auxiliary control air system is

designed to Quality Group B and C and seismic Category I requirements, and is

powered from separate and independent Class 1E power sources. All piping and

components of those portions of the compressed air system required for containment isolation are designed to Quality Group B and seismic Category I requirements. The components of the auxiliary control air system are located in

seismic Category I, tornado-missile-protected structures, and are located above

the PMF level. Therefore, the system satisfies the requirements of GDC 2 and 5

and the guidelines of Regulatory Guides 1.26 and 1.29.

The auxiliary control air system compressors start automatically when the

station control and service air system supply pressure falls below 80 psig.

Isolation of the auxiliary control air system from the main air supply occurs

whenever the system pressure falls below 78 psig. This ensures that all vital

equipment will receive air from the appropriately assigned subsystem under all

conditions including safe shutdown earthquake and maximum possible flood.

Because of differences in system capacity requirements between the Watts Bar

and Sequoyah facilities, Watts Bar has added a fourth compressor, one extra

air dryer, and associated ancillary components. Instrument air compressors

Watts Bar SER 9-14
are powered from emergency busses and are manually loaded onto the emergency

diesel generators following a loss of offsite power.

The applicant has provided a comparison of the system to that which was designed

for SequoyahW The staff has reviewed the information provided in the comparison and concurs with the applicant"'s findings. Thus, the staff reaffirms that

the conclusions stated in the Sequoyah SER (NUREG-O011) are applicable to Watts

Bar.

The staff has further reviewed the system for compliance with the applicable

GDC, Regulatory Guides, and BTPs and concludes-that the design of the system

meets the requirements of GDC 2 and 5 with respect to protection against

natural phenomena, and the sharing of systems and components; the guidelines

of Regulatory Guides 1.26 and 1.29 regarding its quality group and seismic

classification; and is, therefore, acceptable.

9.3.2 Process Sampling System

The process sampling system was reviewed in accordance with the SRP (NUREG-0800),

on the basis of FSAR Section 9.3.2 as well as responses to requests for additional information, and generally available reference materials.

A separate process sampling system is furnished for each unit. Each sampling

system is designed to provide liquid and gaseous samples drawn from the primary

and secondary coolant systems, the associated auxiliary system process streams,

and the spent fuel pool cleanup system. Provisions are made to ensure that

representative samples are obtained from well-mixed streams or volumes of

effluent by the selection of proper sampling equipment and location of sampling points as well as proper sampling procedures. In the event of an accident,

all sample lines which pass through the containment are automatically isolated

by fail-closed valves on either side of the containment.

The process sampling system includes piping, valves, heat exchangers, and other

components associated with the system from the point of sample withdrawal from

a fluid system up to the analyzing station, sampling station, or local sampling

point. The staff review included the provisions proposed to sample all principal fluid process streams associated with plant operation and. the applicant's

proposed design of these systems including the location of sampling points, as

shown on piping and instrumentation diagrams.

The staff determined that the proposed process sampling system meets (1) the

requirements of GDC 13 to monitor variables that can affect the fission process

for normal operation, anticipated operational occurrences, and accident conditions by sampling the reactor coolant, the ECCS core flooding tank, the refueling water storage tank, the boric acid mix tank,.and the boron injection tank

for boron concentrations; (2) the requirements of GDC 13 and 14 to monitor

variables that can affect the reactor coolant pressure boundary and to ensure

a low probability of abnormal leakage, rapidly propagating failure, and gross

rupture, respectively, by sampling the reactor coolant and the secondary

coolant for chemical impurities that can affect the reactor coolant pressure

boundary; (3) the requirements of GDC 26 to control the rate of reactivity

changes by sampling the reactor coolant, the refueling water storage tank, and

the boric acid mix tank for boron concentration; and (4) the requirements of

Watts Bar SER 9-15
GDC 64 to monitor for radioactivity that may be released from normal operations,

including anticipated operational occurrences, and from postulated accidents,

by sampling the reactor coolant, the pressurizer tank, the steam generator

blowdown, the secondary coolant condensate treatment waste, the sump inside

containment, the containment atmosphere, and the gaseous radwaste storage tank

for radioactivity.

The staff further determined that the proposed process sampling system meets

(1) the requirements of 10 CFR 20.1(c) to keep radiation exposures ALARA and

of GDC 60 to control the release of radioactive materials to the environment

by purging and draining sample streams back to the system or origin or to an

appropriate radwaste treatment system and by providing redundant isolation

valves that fail in the closed position; and (2) the requirements of GDC 63 to

detect conditions that may result in excessive radiation levels in fuel storage

and radioactive waste systems by sampling the spent fuel pool water and the

gaseous radwaste storage tank for radioactivity.

The staff also determined that the proposed process sampling system meets the

quality standards requirements of GDC 1 and the seismic requirements of GDC 2

by designing the sampling lines and components of the process sampling system

to conform to the classification of the system to which each sampling line and

component is connected, in accordance with the regulatory positions C.1, C.2,

and C.3 of Regulatory Guide 1.26, the regulatory positions C.1, C.2, C.3, and

C.4 of Regulatory Guide 1.29, and the guidelines of Regulatory Guide 1.97.

Based on its evaluation, the staff concludes that the process sampling system

meets the relevant requirements of 10 CFR 20.1(c) and GDC 1, 2, 13, 14, 26,

60, 63, and 64, and is, therefore, acceptable.

Postaccident Sampling Capability (II.B.3)

NUREG-0737 requires that the applicant provide the capability to promptly

obtain and analyze samples of the reactor coolant system and of the containment

atmosphere under serious accident conditions without radiation doses to individuals exceeding 5 rems whole body or 75 rems to any extremity. Quantitative

analysis is required for certain radionuclides that are indicators of the

degree of core damage, for hydrogen in the containment atmosphere, and for

certain chemicals (such as boron).

In letters dated September 14 and 15, 1981, the applicant committed to a postaccident sampling system that meets the requirements of item II.B.3 in NUREG-0737

and meets the sampling and analysis requirements of Regulatory Guide 1.97,

Revision 2. The applicant will provide the capability to obtain and analyze

samples, within 3 hours of the time a decision is made to sample, from the

reactor coolant system, containment sump, and containment atmosphere under

accident conditions, and to obtain grab samples for offsite analyses. The

applicant also will provide onsite radiological and chemical analysis capabilities to quantify hydrogen levels in containment afmosphere; radioactive

isotopes (noble gases, iodine and cesium isotopes and nonvolatile isotopes);

and dissolved gases, chloride, and boron concentrations in liquid samples.

Sample lines are routed to a shielded sample station in an accessible area.

An isolated auxiliary system is not required to be operational for the postaccident sampling system to be used. Provisions will be made (1) to permit

Watts Bar SER 9-16
sampling containment atmosphere under both positive and negative pressures and

(2) to purge the sample lines for reducing plate out, for minimizing sample

loss or distortion, for preventing blockage of sample lines, for appropriate

disposal of samples, and for passive flow restrictions. Sufficient shielding

will be provided to meet the'requirements of GDC 19, assuming the source term

defined in Regulatory Guide 1.4' All equipment in the postaccident sampling

facility, which may be required to be operational during an accident, will have

access to two independent offsite power supplies. The sampling system and

containment isolation valves will be environmentally qualified. By design,

high-pressure carrier gas in chromatographic analyses cannot enter the reactor

coolant system. Dissolved oxygen at levels less than 0.1 ppm can be analyzed

in a laboratory.

Based on the above evaluation, the staff determined that the provisions in the

proposed postaccident sampling facility will partially meet the requirements

of item II.B.3 in NUREG-0737. However, the applicant has not provided sufficient information on the following: the procedure relating radioactive isotopes to estimated reactor core damage; the time required for chloride analysis;

demonstration of the capability of analyzing grab samples; the required frequency of backup sampling following an accident; the procedures for onsite

radiological and chemical analyses; the provisions to restrict background

radiation in the analysis facility; the accuracy, range, and sensitivity of the

radiological and chemical analyses; verification of the sample station ventilation adequacy; provisions to return residues of sample collection to a closed

system; the testing frequency and type of testing for the postaccident sampling

system; and operator training requirements for postaccident sampling. Hence,

the staff has proposed the l~icense conditions stated below.

Implementation of the requirement is not necessary before low power operation

because only small quantities of radionuclide inventory will exist in the

reactor coolant system and therefore will not affect the health and safety of

the public. Before the plant exceeds 5 percent power operation, the applicant

must demonstrate the capability to promptly obtain reactor coolant samples in

the event of an accident in which there is core damage, consistent with the

following conditions:

(1) Provide a procedure for relating radionuclide gaseous and ionic species

to estimated core damage.

(2) Verify that chloride analysis can be completed within 4 days following an

accident which requires postaccident sampling.

(3) Demonstrate the capability of analyzing the grab samples and verify that

equipment provided for backup sampling shall be capable of providing at

least one sample per day for 7 days following onset of the accident and

at least one sample per week until the accident condition no longer

exists.

(4) Describe the procedures for onsite radiological and chemical analyses and

provide the accuracy, range, and sensitivity of these analyses in an accident chemistry and radiation environment (that is, the presence of large

amounts of fission products and a high radiation field in the samples).

Watts Bar SER 9-17
(5) Verify that provisions are available to restrict background radiation

levels such that the sample analyses will provide results with a range of

accuracy within a factor of 2.

(6) Verify that the sensitivity of onsite liquid sample analysis capability

is such as to permit measurement of nuclide concentration in the range

from approximately 1 pCi/g to 10 Ci/g.

(7) Verify that the ventilation exhaust from the sample station will be

filtered with charcoal adsorbers and high-efficiency particulate air

filters.

(8) Verify that the residues of sample collection will be returned to containment or to a closed system.

(9) Provide information on testing frequency and type of testing to ensure

long term operability of the postaccident sampling system and on operator

training requirements for postaccident sampling-.

In addition to the above licensing conditions, the staff is conducting a

generic review of accuracy and sensitivity for analytical procedures and

online instrumentation to be used for postaccident analysis. The staff will

require that the applicant submit data supporting the applicability of each

selected analytical chemistry procedure or online instrument, along with documentation demonstrating compliance with the licensing conditions 4 months

before the facility exceeds 5 percent power operation, but review and approval

of these procedures will not be a condition for full power operation. If the

staff generic review determines a specific procedure is unacceptable, the staff

will require that the applicant make modifications as determined by its generic

review. The operating license should be conditioned for the items stated

above.

ALARA for Postaccident Sampling

By submittal dated August 21, 1981, TVA committed to procedural changes and

plant modifications which will assure that radiation exposures during postaccident sampling will be ALARA. The commitment and interim procedures

described by the applicant should enable postaccident sampling that is within

the guidelines of GDC 19. These actions meet the staff ALARA position for

this item and are acceptable for fuel loading and full power operations.

9.3.3 Equipment and Floor Drainage System

The nonsafety-related .(Quality Group D, nonseismic Category I) equipment and

floor drainage system (EFDS) include all piping from equipment or floor drains

to the sump, sump pumps, collector tanks, and piping necessary to carry effluents through separate systems.

The liquid drain system is segregated into three basic systems, each designed

to accommodate and collect a particular type of effluent. Effluents are categorized as tritiated, nontritiated and nonchromated, and component cooling

system effluents. In the auxiliary building, discharges are collected in the

floor and equipment drain sump or tritiated sump, and are then subsequently

Watts Bar SER 9-18
pumped to their respective drain collector tanks. Turbine building discharge

is collected in the sump and sampled before it is discharged to the environment.

In the reactor building, most equipment drains are for tritiate-deaerated

liquids, which are pumped to the reactor coolant drain tank. All other floor

and equipment drains are piped to the containment floor and equipment drain

sump. Component cooling system (CCS) effluents are returned to the CCS surge

tank. The containment penetration for the containment sump pump discharge line

is designed to seismic Category I and Quality Group B requirements, and is

located in a seismic Category I, flood-, tornado-, missile-, and environmentally

protected structures, thereby satisfying the requirements of GDC 2 and 4..

All exposed drain piping is in the control building and other areas within

seismic Category I structures where ESF equipment is located and supported so

that its failure will'not compromise plant safety, prevent safe shutdown, or

result in an uncontrolled release of radiation. Thus, the guidelines of

Position C.2 of Regulatory Guide 1.29 are met.

The staff review considered those safety systems needed to provide safe plant

shutdown and the physical location of those systems with regard to potential

inplant flooding. Because of their location at the lowest elevation in the

auxiliary building, the ESF equipment rooms, which contain the residual heat

removal and containment'spray pumps required for safe plant shutdown under

accident conditions, were considered of particular importance with respect to

provisions for prevention of water accumulation. These safety-related pumps

are mounted on pedestals of a minimum height of 12 in. Each of these pump

rooms and area is provided with separate independent sumps and emergency drains

for removal of water accumulation. Equipment and floor drains and sumps are

capable of handling all anticipated drainage. Level indicators and alarms are

provided in the control room for monitoring all sump operating modes. Water

level monitoring at the lowest auxiliary building elevation is provided by

redundant seismic Category I, Class 1E level switches mounted on the floor;

alarm in the control room when the water reaches a depth of 3 in. This will

allow for sufficient time for operator action before an unacceptably high level

is reached in these areas. Adequately sized drains are provided in other

critical areas of the plant to provide for carrying off sufficient water to

prevent room flooding. Thus, the system design meets the requirements of GDC 2

and the guidance of Regulatory Guide 1.26, and Position C.2 of Regulatory

Guide 1.29, because its failure will not compromise plant safety or prevent a

safe shutdown. In addition, the guidelines of BTP ASB 3-1, "Protection Against Postulated Piping Failures in Fluid Systems Outside Containment," concerning.

internal plant flooding as a result of postulated piping failures are met.

Based on its review, the staff concludes that adequate protection against

flooding of safety-related equipment and areas and protection against the

inadvertent release of potentially radioactive liquids to the environment

through plant drainage paths is provided by the measures discussed above to

justify the nonsafety classification.of the equipment and floor drainage

system. The staff concludes that the system is in conformance with the requirements of GDC 2 and 4 with respect to the need for protection against natural

phenomena and the protection afforded against environmental effects (flooding)

and with the guidelines of Regulatory Guides 1.26 and 1.29 and BTP ASB 3-1

concerning seismic and quality group classification and flooding as a result of

postulated piping failures. The system is, therefore, acceptable.

Watts Bar SER 9-19
9.3.4 Chemical and Volume Control System

The chemical and volume control system (CVCS) was reviewed in accordance with

the SRP (NUREG-0800), on the basis of FSAR Section 9.3.4 as well as responses

to requests for additional information.

The CVCS is designed to control and maintain reactor coolant inventory and to

control the boron concentration in the reactor coolant through the process of

charging (makeup) and letdown (drawing off). The system is also designed to

provide seal-water injection flow to the reactor coolant pumps, control the

primary water chemistry unit activity level by-ion exchange and chemical

addition, and process the reactor coolant to recover boron and makeup water.

An essential portion of the system consists of two centrifugal and one positive

displacement charging pumps. These pumps are used during normal operation and

also for high pressure safety injection when the ECCS is required to function.

(The ECCS is evaluated in Section 6.3.)

The volume control tank serves as a volume surge for the reactor coolant

system, to provide for control of hydrogen concentration in the reactor coolant

and to provide a reservoir of makeup for the charging pumps.' The boric acid

makeup system provides for boron additions to compensate for reactivity changes

and to provide shutdown margin for maintenance and refueling operations or

emergencies. The charging and letdown portions of the system are designed to

seismic Category I requirements and contain redundant active components and an

alternate flow path in order to meet the single failure criterion.

The CVCS (including the boron recovery system) includes components and piping

associated with the system from the letdown line of the primary systemto .the

charging-lines that provide makeup to the.primary system and the reactor

coolant pump seal water system.

The basis for acceptance in the staff review has been conformance of the applicant's design of the CVCS with the following regulations and regulatory guides:

(1) the requirements of GDC 1 and the guidelines of Regulatory Guide 1.26 by

assigning quality group classifications to system components in accordance with

the importance of the safety function to be performed; (2) the requirements of

GDC 2 and the guidelines of Regulatory Guide 1.29 by designing safety-related

portions of the system to seismic Category I requirements; (3) the requirements

of GDC 14 by maintaining reactor coolant purity and material compatibility to

reduce corrosion and thus reduce the probability of abnormal leakage, rapid

propagating failure, or gross rupture of the reactor coolant pressure boundary;

(4) the requirements of GDC 29 as related to the reliability of the CVCS to

provide negative reactivity to the reactor by supplying borated water to the

reactor coolant system in the event of anticipated operational occurrences;

(5) the requirements of GDC 33 and 35 by designing the CVCS with the capability

to supply reactor coolant makeup in the event of small breaks or leaks in the

reactor coolant pressure boundary and to function as part of ECCS assuming a

single-failure coincident with loss of offsite power; and (6) the requirements

of GDC 60 and 61 with respect to confining radioactivity by venting and collecting drainage from the CVCS components through closed systems.

Based on the review of the CVCS, and the requirements for system performance

of necessary functions during normal, abnormal, and accident conditions, the

Watts Bar SER 9-20
staff concludes that the design of the CVCS and supporting system meets the

requirements of GDC 1, 2, 14, 29, 33, 35, 60, and 61, and is, therefore,

acceptable.

9.4 Heating, Ventilation, and Air Conditioning System

9.4.1 Control Room Area Ventilation System

The control building heating, ventilating, air-conditioning (HVAC), and air

cleanup systems are designed to maintain suitable environmental conditions

throughout the building for equipment operation and safe occupancy under all

plant operating conditions. (See Section 6.4 of this SER for further information regarding control room habitability.) The building pressurizing air

supply system is designed to maintain the control building at a positive

pressure relative to the outdoors and to the adjoining building at all times.

The control room HVAC system components are provided with two 100-percent

capacity units. Each meets the single failure criterion, and automatic switchover occurs if one of the ,units fail. These components include

(1) main control room air-conditioning system, water chillers, air handling

units, and piping

(2) control building pressurizing air supply fans

(3) main control room emergency air cleanup supply fans and filter assemblies

(4) main control room emergency pressurizing air supply fans

The control building outside air intakes are provided with radiation monitors,

high temperature detectors, chlorine detectors, and smoke detectors. Indicators

are provided with the chlorine detectors and radiation monitors. Main control

room annunciation is provided for each type of monitor or detector. Isolation

of the main control room air intakes occurs automatically upon the actuation of

a safety injection signal from either unit or upon indication of high radiation,

high temperature, chlorine, or smoke concentrations in the outside air supply

stream to the building. Ventilation is then provided by recirculation of air

through emergency filters. Main control room isolation may also be accomplished

manually at any time by the control room operators. Thus, the design conforms

to the requirements of GDC 19 and the guidelines of Regulatory Guide 1.95.

Two minor differences in the system design exist between the Watts Bar and

Sequoyah facilities: (1) at Watts Bar the capacity of the air handling units

(AHUs) is greater than those at Sequoyah, and (2) chilled water is used as the

cooling medium for the AHUs-at Watts Bar, whereas a refrigerant is used at

Sequoyah.

The system is housed in seismic Category I structures, and is flood- and

tornado-missile protected. Essential portions of the system are physically

separated from high-energy piping systems and are protected against postulated

internally generated missiles. Essential portions of the system are designed

to Quality Group C seismic Category I requirements, and are located above the

Watts Bar SER 9-21
pmf level. Thus, the design satisfies the requirements of GDC 2 and the

guidelines of Regulatory Guides 1.26 and 1.29.

The system is physically separated from high-energy systems and is thus protected from the effects of postulated pipe failure in high-energy systems (see

to Section 3.6.1 of this SER). The control room HVAC equipment is protected

from internally and externally generated missiles and is provided with suitable

drainage for protection against flooding as a result of moderate-energy piping

system failure and failures in nonseismic water systems (see Sections 3.5.1.1

and 9.3.3 of this SER). The outside air intakes are located in the auxiliary

building and are tornado-missile protected. The design thus satisfies the

requirements of GDC 4 and the guidelines of BTP ASB 3-1.

The applicant has provided a comparison of the system to that which was designed

for Sequoyah. The staff has reviewed the information provided in the comparison

and concurs with the applicant's findings. The staff reaffirms that the

conclusions stated in the Sequoyah SER (NUREG-O011) are applicable to Watts

Bar. The staff has further reviewed the system for compliance with applicable

GDC, Regulatory Guides, and BTPs and concludes that the control building HVAC

system conforms to the requirements of GDC 2, 4, and 19 with respect to providing protection against natural phenomena, environmental effects and missiles,

and to providing an adequate environment for equipment operation and protection

to permit access for occupancy of the control room under accident conditions,

and to the guidelines of Regulatory Guides 1.26, 1.29, 1.95, and BTP ASB 3-1

relating to the quality group and seismic classification, protection against

chlorine release, and high- and moderate-energy pipe breaks. It is, therefore,

acceptable.

9.4.2 Fuel-Handling Area Ventilation System

The fuel-handling area ventilation system (FHAVS), a subsystem of the auxiliary

building ventilation system, serves the spent fuel pool area, penetration

rooms, and the fuel, waste, and cask handling areas. The system is designed to

maintain acceptable environmental conditions for personnel access, to protect

and maintain safe equipment operations, and to limit the release of radioactivity to the environment.

The system is classified as nonessential (nonsafety related), except for those

portions that may be used by the auxiliary building gas treatment system

(ABGTS) to mitigate the consequences of a fuel-handling accident, which are

classified as essential. The nonessential portion of the system is separated

from the essential portions so that its failure will not prevent essential

safety functions. During an accident, the FHAVS is shut down and all environmental control is handled by the ABGTS (described in Section 6.2.3 of this

SER). All duct work, dampers, and grills essential to operation of the ABGTS

are designed to seismic Category I, Quality Group C requirements. All other

system components are designed to Quality Group D, nonseismic Category I

requirements. The FHAVS is located within seismic Category I structures. All

safety-related components of the system are protected from the effects of

floods, tornado missiles, and internally generated missiles. Thus, the requirements of GDC 2 and 4 and the guidelines of.Regulatory Guides 1.26, 1.29, and

1.117 are met.

Watts Bar SER 9-22
As a subsystem of the auxiliary building ventilation system, all essential

functions of this system are powered by two trains of emergency electrical

power upon the loss of offsite power. Upon receipt of a high radiation or

containment isolation signal, detectors will transmit their signals via redundant trains. This isolates the FHAVS and automatically starts the ABGTS.

Isolation of the essential portion of the system is accomplished by the automatic disengagement of the auxiliary building supply and exhaust fans and the

fuel. handling exhaust fans,.and the closure of isolation dampers, thus conforming to the guidelines of Regulatory Guide 1.13.

The applicant has provided a comparison of the Watts Bar system to that designed

for Sequoyah. The staff has reviewed the information provided in the comparison

and concurs with the applicant's findings.

The staff has further reviewed the system for conformance with applicable GDC,

Regulatory Guides, and BTPs and concludes that the system conforms to the

requirements of GDC 2 and 4 with respect to protection against natural phenomena

and missiles and the guidelines of Regulatory Guides 1.13, 1.26, 1.29 and 1.117

with respect to protection against radioactive releases, quality group and

seismic classification, and protection against tornado missiles. It is,

therefore, acceptable.

9.4.3 Auxiliary and Radwaste Area Ventilation System

The auxiliary building and radwaste area ventilation systems (ARAVS) serve all

areas of the auxiliary building including the fuel handling area (Section 9.4.2

of this SER) and the radwaste area. The ARAVS consists of the following subsystems:

(1) building air supply and exhaust system

(2) building cooling system

(3) safety features equipment coolers, shutdown board room air-conditioning

system

(4) auxiliary board room air-conditioning system

(5) shutdown transformer room ventilation system

(6) miscellaneous ventilation and air-conditioning systems

Of the above, the safety-related subsystems are the safety features equipment

coolers, the shutdown board room air-conditioning system, and the auxiliary

board room air-conditi~oning system. These subsystems are housed in seismic

Category I structures and are protected against tornado missiles and flooding.

All essential equipment is located above the PMF level. These safety-related

systems are designed to Quality Group C, seismic Category I requirements, and

are supplied by emergency Class 1E power supply in the event of a loss of

offsite power.ý They are also designed to ensure that a singl.e failure will

not result in the loss of a safety-related function. Thus, the system meets

the requirements of GDC 2 and 4 and the guidelines of Regulatory Guides 1.26,

1.29, and 1.117.

Watts Bar SER 9-23
The applicant has provided a comparison of the Watts Bar system to that which

was designed for Sequoyah. The staff has reviewed the information provided

in the comparison and concurs with the applicant's findings. The staff reaffirms

that the conclusions stated in the Sequoyah SER (NUREG-O011) are applicable to

Watts Bar.

The staff has further reviewed the system for compliance with the applicable GDC, Regulatory Guides, and BTPs and concludes'that the ARAVS conform to the

requirements of GDC 2 and 4 with respect to protection against natural

phenomena and environmental effects and missiles and to the guidelines of

Regulatory Guides 1.26, 1.29, and 1.117 regarding quality group and seismic

classification and the effects of tornado missiles. It is, therefore,

acceptable.

9.4.4 Turbine Building Area Ventilation System

The turbine building HVAC system, classified as nonessential Quality Group D,

nonsafety related, is designed to maintain an acceptable environment for personnel and the nonessential equipment served during normal plant operations.

The system is separated from safety-related plant systems and areas. Failure

of the system will not compromise the operation of any essential plant systems

or result in an unacceptable release of radioactivity and is, therefore, consistent with the guidelines of Regulatory Guide 1.26 and Position C.2 of

Regulatory Guide 1.29.

The applicant has provided a comparison of the.Watts Bar system to that which

was designed for Sequoyah. The staff has reviewed the information provided in

the comparison and concurs with the applicant's findings.

The staff has further reviewed the system for compliance with the applicable

GDC, Regulatory Guides, and BTPs and concludes that the system design conforms

to the guidelines of Regulatory Guides 1.26 and 1.29, Position C.2, as related

to quality group and seismic classification, and is, therefore, acceptable.

9.4.5 Engineered Safety Features Ventilation System

The engineered safety features (ESF) ventilation system provides a suitable

and controlled environment for essential equipment and components during normal

plant operating conditions and accident conditions. These components include

(1) the essential raw cooling water pumps and ancillary equipment

(2) the diesel generator and associated components

(3) ESF equipment such as the containment spray pumps, the residual heat

removal pumps, and the centrifigal charging pumps

At the essential raw cooling water (ERCW) pumping station, the ERCW pumps are

cooled by natural convection; electrical and mechanical equipment are cooled

mechanically. Failure of one of the two ventilation fans in a mechanical

equipment room will not prevent the operation of any safety-related equipment.

Failure of the electrical equipment room ventilation subsystem will not affect

any safety-related components or functions.

Watts Bar SER 9-24
The diesel generator ventilating system is required to maintain the temperature

of the diesel generator rooms between 40*F and 120'F during all operating

conditions. Components of the diesel generator ventilating system are housed

in seismic Category I structures and are protected from flooding, tornado

missiles, and internally generated missiles. The diesel generator room-main

exhaust fans for each diesel unit are provided in redundant pairs to ensure

that a single failure in the ventilating system will not prevent a safetyrelated component from accomplishing its function. Upon the loss of offsite

power, each diesel generator will provide emergency electrical power to its

associated ventilation components. The diesel generator room exhaust fans,

battery hood exhaust fans, electrical board room exhaust fans, and all associated ductwork, fittings, and dampers are designed to Quality Group C and

seismic Category I requirements; thus, the requirements of GDC 2 and 4 and the

guidelines of Regulatory Guides 1.26 and 1.29 are satisfied.

The auxiliary building safety feature ventilation system provides ventilation

to the ESF components, the reciprocating charging pump, the component cooling

booster pumps, and the spent fuel cooling pumps. All components of this

system--including air cooling units, fans, ductwork, dampers, and grilles are

designed to Quality Group C, seismic Category I requirements--are enclosed

in seismic Category I structures. All essential portions of the system are

located above the PMF level. During the accident mode of operation, emergency

electrical power will be provided to the ESF pumps and their corresponding

coolers or fans. If high radiation levels are detected, this system will be

isolated and airflow patterns and cleanup will be established and maintained

by the ABGTS (as discussed in Section 6.5 of this SER). Thus, the require- ments of GDC 2 and 4 and the guidelines of Regulatory Guides 1.26, and 1.29,

are met.

The applicant has provided a comparison of the Watts Bar system to that which

was designed for Sequoyah. The staff has reviewed the information provided in

the comparison and concurs with the applicant's findings.

The staff has further reviewed the system for conformance with the applicable

GDC, Regulatory Guides, and BTPs and concludes that the system conforms to the

requirements of GDC 2 and 4 with respect to protection against natural phenomena and environmental effects and missiles and the guidelines of Regulatory

Guides 1.26 and 1.29 regarding quality group and seismic classification. It

is, therefore, acceptable.

9.5 Other Auxiliary Systems

9.5.1 Fire Protection

9.5.1.1 Introduction

The staff reviewed the Watts Bar fire protection progam re-evaluation and fire

hazards analysis submitted by the applicant by letter dated April 18, 1977,

September 8, 1980, and August 28, 1981. Watts Bar is an identical two-unit

facility. Because Units 1 and 2 are of the same design except as noted, the

comments below apply to both units.

Watts Bar SER 9-25
The Watts Bar re-evaluation was in response to a staff request that the applicant review his fire protection program against the guidelines of Appendix A to

BTP ASB 9.5-1, "Guidelines for Fire Protection for Nuclear Power Plants." As

part of the review, the staff visited the plant site to examine the relationship

of safety-related components, systems, and structures in specific plant areas

to both combustible materials and to fire detection and suppression systems.

The overall objective of the review was to ensure that in the event of a fire

at Watts Bar, personnel and plant equipment would be adequate to safely shut

down the reactor, to maintain the plant in a safe shutdown condition, and to

minimize the release of radioactivity to the environment.

The staff review included an evaluation of the automatic and manually operated

water and gas fire suppression systems, the fire detection systems, fire

barriers, fire doors and dampers, fire protection administrative controls, and

the fire brigade size and training.

The applicant has been informed that all fire protection modifications have to

be implemented before fuel load.

Staff consultants, Gage-Babcock and Associates, Inc., participated in the

review of the fire protection program and in the preparation of this SER, and

concur with the staff findings.

9.5.1.2 Fire Protection Systems Description

Water Supply Systems

The water system is common to both units and consists of four ASME Section III

seismic Category I high pressure vertical turbine motor-driven pumps, each

rated at 1590 gpm at 330-ft head. The pumps are located in the seismic Category I intake pumping station. A 3-hour-fire-rated barrier is provided to

separate two fire pumps from the other two.

The fire pump motors are powered by the 480-V shutdown boards. In the event

of loss of offsite power, each fire pump power supply is automatically connected to a separate emergency diesel generator. Supervised alarm circuits,

indicating fire pump motor running condition and loss of line power on the

line side of the switchgear, are provided in the main control room for each

pump.

The staff was concerned that an exposure fire in the control building (elevation 708 ft) might damage the relays for the fire pumps and prevent automatic

operation of the pumps. At the request of the staff, the applicant has verbally

agreed to relocate two (one primary and one secondary) of the relays and

separate them by at least 20 ft. This will be documented in a future amendment

to the FSAR.

Water supply for the fire pumps is taken from Chickamauga Reservoir and is

considered unlimited for fire protection purposes. An underground fire main

loop is provided to serve both units. Sectional isolation valves are provided

so that maintenance may be performed on the loop or portions of the loop for

one unit without affecting fire-fighting capability of either unit. The

sectional isolation valves in the underground loop are mechanically locked in

Watts Bar SER 9-26
position and Technical Specification surveillance is placed upon supervision of

valve position to ensure proper system alignment. The yard fire main loop is

cross-tied between units. The fire protection headers are pressurized through

an interconnection with the raw water system, with the pressure being maintained

by two 10,000-gal raw water tanks on the auxiliary building roof. The raw water

system is automatically isolated when the fire pumps start.

The diesel generator building has a single feed from the underground fire main

into the building feeding a preaction sprinkler system and manual hose station.

A single active failure or crack in a moderate-energy, line can impair both the

primary and backup fire suppression systems. By letter dated August 28, 1981,

the applicant agreed to provide an additional fire water feed for the diesel

generator building so that the primary or secondary suppression is assured.

Based on its review, the staff concludes that the fire water supply system

meets the guidelines of Section C.2 of Appendix A to BTP 9.5-1 and, therefore,

is acceptable.

Sprinkler and Standpipe Systems

Automatic sprinkler systems and hose station standpipe systems are separately

connected to the yard main or to headers within buildings fed from each end of

the building; therefore, a single failure cannot impair both sprinkler systems

and hose stations. Fixed-water spray systems and sprinkler systems are designed

according to the requirements of NFPA Standard No. 13, "Standard for Installation of Sprinkler Systems," and NFPA Standard No. 15, "Standard for Water Spray

Fixed System." Hose stations are provided for the lower levels of the intake

pumping station; however, these hose stations cannot reach the upper elevations.

By letter dated August 28, 1981, the applicant agreed to modify the present

standpipe system by extending it and adding an additional interior manual hose

station. Manual hose stations are located throughout the plant to ensure that

an effective hose stream can be directed to any safety-related area in the

plant. The system is designed according to the requirements of NFPA Standard

No. 14, "Standpipe and Hose System for Sizing, Spacing, and.Pipe Support

Requirements." Pipe and pipe hangers of the fire protection system located in

seismic Category I structures are designed for seismic requirements to ensure

the integrity of other essential equipment in the same area.

Valves in the fire protection system are not electrically supervised; however,

all valves whose misalignment would prevent proper operation of the system will,

be mechanically locked in their normal position. Technical Specifications

surveillance is placed on supervision of valve position to ensure proper system

alignment.

Areas that are equipped with water suppression systems are

Control Building

Elevation 755 ft

(1) mechanical equipment room

(2) janitor's closet

(3) corridor

(4) kitchen

Watts Bar SER 9-27
(5) toilet

(6) locker room

(7) instrument calibration

(8) chart storage

(9) shift engineer's office

(10) record storage vault

(11) PSO engineering shop

(12) control room air cleanup charcoal filters

Elevation 692 ft

(1) mechanical equipment rooms

(2) 250-V battery room 1 and 2

(3) 24- and 48-V battery room

(4) communications room

(5), corridor

(6) secondary alarm station

Elevation 729 ft

(1) spreading room

Diesel Generator Building

Elevation 742 ft

(1) pipe gallery and corridor

Intake Pumping Station

Elevation 710 ft

(1) electrical equipment room

Reactor Building

(1) reactor coolant pumps

(2) annulus area (division interactions)

Turbine Building

(1) numerous areas of building

Auxiliary Building

Elevation 772 ft

(1) 480-V board rooms

(2) 125-V vital battery rooms

(3) 480-V transformer rooms

(4) mechanical equipment rooms

(5) HEPA filter plenum rooms

Watts Bar SER 9-28
Elevation 782 ft

(1) control rod drive equipment rooms

(2) pressure heater transfer rooms

Elevation 757 ft

(1) auxiliary control room

(2) 6.9-kV and 480-V shutdown board rooms

(3) 125-V vital battery rooms

(4) personnel and equipment access

(5) reverse osmosis equipment room

(6) reactor building equipment hatches

(7) reactor building access rooms

(8). emergency gas treatment filters

(9) auxiliary control instrument rooms

Elevation 737 ft

(1) common area

(2) hot instrument shop

(3) heating and vent

(4) ventilation and purge air

(5) GF fuel detector room

(6) auxiliary building gas treatment system filters

Elevation 733 ft

(1) valve gallery

(2) decontamination room

Elevation 729 ft

(1) waste package areas

(2) fuel transfer valve room

Elevation 713 ft

(1) auxiliary building common area

(2) air lock

(3) titration room

(4) radiochemical laboratory

(5) counting room

(6) pipe gallery

(7) volume control tank rooms

(8) sample rooms

(9) pipe gallery

(10) containment purge air exhaust filters

Elevation 692 ft

(1) auxiliary feedwater pump rooms

(2) pipe gallery

(3) charging pump room

Watts Bar SER - 9-29
(4) safety injection pump rooms

(5) cast decontamination collection tank room

(6) spent resin tank room

(7) valve gallery

(8) waste evaporator package room

(9) 'auxiliary waste evaporator packaging room

(10) corridor

(11) chemical drain tank room

During its site visit, the staff observed that throughout the auxiliary building elevations 713 ft and 737 ft, numerous cable trays, conduit, and equipment

obstruct the overhead sprinkler discharge. The staff was concerned that the

floor elevation might not be protected against a fire by the overhead sprinkler

system. By letter dated August 28, 1981, the applicant agreed to modify design

so that the sprinkler heads be either relocated or additional heads installed

to enable complete sprinkler coverage on the floor below.

The staff has reviewed the design criteria and bases for the water suppression

systems and concludes that these systems, with the additional sprinkler systems

and standpipe system modifications to be installed, meet the guidelines of

Appendix A to BTP 9.5-1 and are, therefore, acceptable.

Gas Suppression System

A low pressure total flooding:carbon dioxide (C0 2 ) system is provided for the

following areas:

(1) standby diesel generator rooms lA-A, 2A-A, 1B-B, 2B-B

(2) turbine lube oil dispensing room

(3) computer room

(4) paint shop and storage room

(5) auxiliary instrument rooms

(6) spreading room

(7) 480-V board room

(8) lube oil storage room

(9) fuel oil transfer room

(10) lube oil purification room

The C02 system for the diesel generator building is automatically actuated by

thermal-rate-compensated detectors. The system can also be actuated manually.

A 20-sec time delay gives personnel time to clear the area before CO2 is discharged. Actuation of the system provides alarms and annunciates in the main

control room. Fire dampers, provided in each air supply and exhaust to the

room, will automatically close to isolate the room in the event of a fire. A

manually operated total flooding CO2 system, with a 2-min delay, is also provided for the cable spreading room, which will be used only as a backup system

to the automatic sprinkler system. Actuation of the system will alarm locally

and in the main control room.

The CO2 systems are designed and installed according to NFPA Standard No. 12,

"Carbon Dioxide Extinguishing Systems."

The staff concludes that these systems satisfy the provisions of Appendix A to

BTP 9.5-1 and are, therefore, acceptable.

Watts Bar SER 9-30
Fire Detection System

The fire detection system consists of initiating devices, local control panels,

remote transmitter-receiver providing remote multiplex (MUX) function, computerized multiplex central control equipment, and power supply. The types of

detectors used are photoelectric and ionization (products of combustion),

thermal (heat sensors) and infrared. Fire detection systems will give an

audible and visual alarm, and also annunciation in the control room. .Local

audible and/or visual alarms are also provided.

The system is electrically supervised for ground and open wiring faults in the

detection, power supply, alarm, and MUX data transmission circuits. Supervision

is Class A in the detection and data transmission circuits and Class B in local

audible alarm circuits. A wiring fault in the above circuits results in an

audible and visual trouble indication at both the local and control locations.

The fire detection system is powered from a single 120-V ac distribution panel.

The panel is provided with a manual transfer switch to allow normal or alternate power feed from the Class 1E 480-V ac control and auxiliary building

ventilation boards. The ventilation boards are automatically connected to the

emergency diesel generators on loss of offsite power.

By.letter dated September 9, 1980, the applicant stated that the fire alarm

system is electrically supervised for ground and open wiring faults in the

detection, power supply, alarm, and MUX data transmission circuits. Supervision is Class A in the detection and data transmission circuits and Class B in

local audible alarm circuits. A wiring fault in these circuits results in an

audible and visual trouble indication at both the local and control locations.

In a February 18, 1982 meeting, the applicant committed to provide supervision

of valve actuation circuits (for preaction sprinkler system, and deluge systems)

so that a single break or ground fault condition will initiate a visual and

audible trouble indication in the control room. Pending confirmation of this

commitment, the staff finds this supervision complies-with the requirements of

NFPA Standards 13 and 72D, and is, therefore, acceptable.

The fire detection systems have been installed or will be installed according

to NFPA Standard 72D, "Standard for the Installation, Maintenance, and Use of

Proprietary Protection-Signalling Systems."

The staff has reviewed the fire detection systems to ensure that fire detectors

are adequate to provide detection and alarm of fires that could occur. It has

also reviewed-the fire detection system's design criteria to ensure that it

conforms to the applicable sections of NFPA 72D. Once the applicant has

confirmed his commitment, the staff will report on the supervision of the

control function circuits in a subsequent safety evaluation report.

9.5.1.3 Other Items Relating to the Station Fire Protection Program

Fire Barriers

Three-hour fire-rated barriers are provided between the reactor building and

auxiliary building, control building and auxiliary building, service building,

and auxiliary building, and control building and turbine building. All floors,

walls, and ceiling enclosing the control room and the cable spreading room are

rated at a minimum 3-hour-fire rating. Equivalent 3-hour-fire separation will

Watts Bar SER 9-.31
be maintained between adjacent diesel generator units within the diesel generator building. The main control room area contains peripheral rooms which are

located within the main control room 3-hour-fire barrier. These peripheral

rooms are provided with automatic sprinklers, detectors, and 11⁄2-hour-fire-rated

barriers. A minimum 3-hour-fire-rated coating of Pyrocrete 102 is applied to

all exposed structural steel required to maintain structural integrity of the

cable spreading room. Other fire areas, having low or minimal fire loadings,

are provided with 11⁄2-hour-fire-rated barriers. The staff has reviewed the fire

hazard analysis, including the fire loading, fire detection system, and fire

suppression system in these areas, and found that a postulated fire in these

areas would not be sufficient to breach the fire-barrier integrity. The staff,

therefore, concludes that this is an acceptable alternative.

The applicant has proposed by letter, dated September 9, 1980, to provide a

1⁄2-hour-fire-rated barrier for conduit and/or cable trays of one of the two

redundant divisions that are within 20 ft of each other. On February 19, 1981,

the Commission requirements for the fire protection of safe shutdown systems

became effective (Appendix R to 10 CFR 50). The staff will require that the

applicant meet the provisions of Section III.G.2 of Appendix R to 10 CFR 50 by

providing a 1-hour-fire-rated barrier (will maintain circuit integrity for

1 hour) for one of the two redundant divisions that are within 20 ft of each

other. The cable tray supports shall be protected.to achieve the'same fire

rating as the cable tray or conduit itself to ensure the structural integrity

of the cable tray for 1 hour. The 1-hour-fire-rated barriers are required to

be provided for the cables and/or cable trays as listed in the applicant's

letter dated September 9, 1980.

At elevation 741 ft of the intake pumping station, the redundant essential raw

cooling water pumps are separated by a 3-hour-fire barrier. These pumps are

also separated from the traveling screen pumps by a 3-hour barrier; however,

this barrier has an open scupper at the base of the wall in each ERCW pump

room.

The scupper openings penetrating the fire wall between the ERCW pump rooms and

traveling screen rooms are provided for drainage of rain water from the open

pump rooms to the pump well. By letter dated August 28, 1981, the applicant

provided the following design-analysis:

The slope of the floor deck at elevation 741 ft is such that an oil spill from

any one train of ERCW pumps does not have a direct route to the other train of

pumps. The deck is sloped tG the openings in the south wall (ERCW pump room

side) so that a postulated oil spill will flow to the scuppers pass-through,

and immediately drop into the noncritical traveling screen wells in the traveling screen and screen wash pump room. The wall separating the ERCW pump rooms

and traveling screen rooms is intended to provide radiant heat protection from

an exposure fire. The roof of the intake pumping station deck is constructed

of wide flange beams provided for missile protection. However, the roof design

permits free air flow between the, beams so that, in the event of fire, heat

will not stratify or bank down from the ceiling, thereby minimizing the temperature rise within the room.

The applicant found that this wall separating the ERCW pumps from the adjacent

traveling screens and screen wash pump room is adequate to prevent the spread

Watts Bar SER 9-32
of fire. Thus more than one ERCW train would not be jeopardized in a fire.

The staff concurs with the applicant's finding in this regard.

The fire-rated assemblies provided are adequate with the above modifications

for the areas affected; meet'the guidelines in Section D.i.j. of Appendix A to

BTP 9.5-1 and Section III.G of Appendix R to 10 CFR 50; and are, therefore,

acceptable.

Fire Doors, Dampers, and Fire Barrier Penetrations

The staff has also reviewed the placement of the fire doors to ensure that

fire doors of proper fire rating have been provided.

Doors separating the control building from the turbine building are normally

closed. Heavy equipment doors are locked and operated by card readers.

Operation of these doors is alarmed in the main control room. Strict administrative procedures will be used to ensure that the doors are not left open or

propped open during maintenance or plant operation.

In the diesel generator building at elevation 742 ft, the lube oil storage room

is enclosed in 3-hour-fire-rated construction; however, the 3-hour fire doors

are in the open position and close only upon melting of the thermal link above

the door or by discharge of the C02 system for the room. The staff is concerned

that fire may spread to adjacent areas before the fire doors are closed, or

that some obstruction in the doorway may keep the fire door from closing completely. To meet the guidelines of Section 6-6.3.2 of NFPA 101 (1976) as well

as of Section 4-4.1.2 of NFPA 30, "Flammable and Combustible Liquids Code,"

these doors should be self-closing. In addition to the guidelines of NFPA

Codes, Section N of Appendix R to 10 CFR 50 requires the doors to be selfclosing. The staff require the applicant to keep these doors closed and locked

or alarmed, with alarm and annunciation in the control room.

At elevation 760 ft of the diesel generator building, each switchgear room is

separated from the other by 3-hour-fire-rated construction. The doors separating these rooms are not labeled. By letter dated November 10, 1981, the applicant has stated that these doors are 3-hour fire doors (including their frames

and hardware) that have been tested and approved by a nationally recognized

laboratory or to replace them with 3-hour fire doors.

Fire doors in most of the fire-cell and fire-area boundaries are UL labeled.

The special purpose doors in the auxiliary building, such as flood doors and

pressure doors, are not UL labeled. These doors are designed to ASME Standard

and are of heavy-welded-steel construction. The applicant has evaluated these

doors and determined that they will provide a fire rating commensurate to the

fire loading in the areas or cells they separate. The security doors in the

main control room are not UL labeled. They are made of bullet-resistant,

heavy-gauge steel, and the door manufacturer has certified that the doors are

equivalent to UL-tested fire doors rated for 3 hours. The applicant considers

these nontested doors equivalent. Similar doors were found acceptable for the

Sequoyah nuclear plant. Therefore, the staff finds these acceptable.

Penetrations, including electrical penetration seals through rated barriers,

are sealed to provide fire resistance equivalent to the barrier itself.

Watts Bar SER 9-33
Ventilation penetrations through fire-rated barriers are protected by.standard

fire doors dampers. All of the fire doors/dampers are UL listed.

Some of the 11⁄2-hour-fire-rated dampers are not installed according to the

manufacturer's instruction or NFPA 80, "Fire Doors and Windows." By letter

dated August 28, 1981, the applicant agreed to modify these installations in

the same way they were modified for the Sequoyah nuclear plaht.

By letter dated March 1, 1982, the applicant has stated and provided supporting

justification that the Watts Bar electrical fire barrier penetrations meet the

requirements of ASTM E-119. The staff will compare the applicant's test

methods to the requirements of ASTM E-119 to confirm this statement, and will

report its findings in a supplement.to the SER.

The staff has reviewed the fire doors and dampers. During the July 20-24, 1981

site visit, the applicant agreed to modify the design and installation to be

identical with those accepted at the Sequoyah plant as meeting the guidelines

of Appendix A to BTP ASB 9.5-1. The design of the doors and dampers is, therefore, acceptable, pending confirmation that these modifications are complete.

9.5.1.4 Emergency Lighting

By letter dated August 28, 1981, the applicant stated that the 11⁄2-hour fixed

emergency seal-beam lighting units are modified throughout the plant with small

capacity bulbs, so that a lighting duration of approximately 8 hours is provided.

The staff finds this arrangement unacceptable because the amount of light in a

given area is reduced below acceptable levels.

The staff will require the applicant to provide adequate illumination of not

less than 1.0 ft-candle measured at the floor at all points of equipment operation needed for shutdown, as well as at all points on the floor, including

angles and intersections of corridors, passageways, stairways to valves, and

for routes to and from all of the areas. The staff will report on this item in

a supplement to this SER.

9.5.1.5 Fire Protection for Specific Areas

Component Cooling Water Pump Area (El 713)

On elevation 713 ft of the auxiliary building, all five (two from each unit

and one swing) component cooling water (CCW) pumps are located together.

Adjacent to these safety-related pumps are the two motor-driven auxiliary feedwater pumps (both trains) of Unit 1, which are also safety related. Both

Unit 2 auxiliary feedwater pumps are located approximately 125 ft away, down

the corridor. Power-operated control valves for the CCW pumps are located

immediately above the CCW pumps on an .open-grating mezzanine. Various safetyrelated cable trays are also located in the area. A preaction sprinkler system

is proposed for the ceiling level only; this would not offer adequate protection

against an exposure fire because of the many obstructions between the ceilinglevel sprinkler and the floor below.

At the request of the staff, the applicant verbally has agreed to provide

Watts Bar SER 9-34
(1) automatic sprinkler coverage under the pipe-break barrier for the Units 1

and 2 motor-driven auxiliary feedwater pumps

(2) automatic sprinkler coverage under the mezzanine for all five CCW pumps

(3) a 1-hour-fire-rated barrier separating the train A CCW pump from the

train B CCW pumps, so that the barrier will extend approximately 3 ft

above the highest point of each pump

(4) additional smoke detectors to actuate the proposed sprinkler systems and

to ensure early warning of a fire

The above will be documented in a future FSAR amendment.

The staff has reviewed the applicant's fire hazards analysis and fire protection provided for the CCW pump area and motor-driven auxiliary feedwater pump

area. The fire protection to be provided is the same as that previously

accepted by the staff at the Sequoyah nuclear plant. The staff concludes that

appropriate fire protection has been provided and, after the modifications are

implemented, will meet the guidelines of Appendix A to BTP ASB 9.5-1. It-is,

therefore, acceptable.

Cable SpreadingRoom

The cable spreading room is shared by both units. The walls, floors, and

ceiling are designed to have a fire rating of 3 hours. An automatic preaction

sprinkler system has been provided. The system has two horizontal levels in

the cable spreading room: (1) an upper level near the ceiling, and (2) an

intermediate level approximately halfway between the floor and ceiling. The

sprinklers in the intermediate level are staggered horizontally between the

upper level sprinkler grid. A manual total flooding C02 system has been

provided as a backup system. Hose stations are provided. A cross-zoned

ionization smoke detection system also is installed in this area.

The exposed cable in the room has been coated with a flame retardant to minimize

fire propagation. In the event of a fire in the control room or cable spreading

room, plant shutdown capability can be maintained from the auxiliary control

room, which is completely separate and independent of these areas.

The staff has reviewed the applicant's fire hazards analysis and fire protection provided for the cable spreading room and considers that fire protection

and emergency shutdown procedures have been provided and conform to the provisions of Appendix A to BTP 9.5-1. Thus, they are acceptable.

Fire Protection Inside Containment

The major fire hazard within the. containment is the reactor coolant pump (RCP)

lube oil system. To prevent a fire as a result of oil leakage, the applicant

has provided an oil collection system for each RCP. The oil collection system

includes-enclosing the RCP oil lift pump and providing an oil collection basin

at the access platform elevation of each pump to collect and drain away any

combustible liquid and/or suppression system discharge. Any discharge is

drained from the collection basin into the containment floor and equipment

drain sump located inside primary containment. Each RCP is provided with a

Watts Bar SER 9-35
heat collection hood which also will act as a heat collector to reduce the

response time of the thermal detectors and the thermal-actuated-water-spray

nozzles installed inside the housing. The fixed automatic water spray system

is designed in accordance with NFPA 15. The oil collection system meets the

requirements of Section III.0 of Appendix R to 10 CFR 50 and is, therefore,

acceptable.

An automatic fixed-water-spray system has been provided for the charcoal HEPA

filters in the lower containment air cleaning units. The water spray system

is designed according to NFPA 15.

Areas of divisional interaction within the annulus area will be protected by

an automatic-fixed-water-spray system designed according to NFPA 15 with the

exception that conventional sprinkler heads will be used. In addition, all

exposed cables within this area will be coated with a flame retardant. The

staff also will require the applicant to provide a 1-hour-fire-rated barrier

for all divisional interactions involving redundant functions of equipment or

circuits necessary to achieve safe shutdown in the event of a fire.

A standpipe and hose system, designed according to NFPA 14, has been provided

to complement the fixed-water-suppression system in the reactor building. The

standpipe system within its containment will be normally dry and arranged to

admit water to the system through manual operation of remote control devices

located at each hose station.

The fire detection system is designed according to NFPA 72D with Class A supervision. Thermal detectors are provided for the charcoal filters and HEPA

filters. Thermal-rate-compensated and flame detectors are provided for the RCP

motors. Smoke, photoelectric, and/or thermal-rate-compensated detectors are

provided for divisional cable interaction areas.

Photoelectric smoke-duct detectors are provided for each lower containment

cooling unit and each upper compartment cooling unit. In addition, photoelectric smoke-duct detectors are provided for the exhaust ducts serving the

containment-purge-air-exhaust systems and the emergency gas treatment system.

In the annulus area, heat and smoke collectors are provided for fire detection

so that a quick response can be obtained.

The staff has reviewed the applicant's fire hazard analysis and fire protection

provided for the area inside containment. The staff concludes that appropriate

fire protection is provided for this area and meets the guidelines of Appendix A

to BTP ASB 9.5-1 Appendix R to 10 CFR 50. It is, therefore, acceptable.

Control Room

The control room complex is separated from other areas of the plant by 3-hourrated fire-resistant construction. Doors between the control room and turbine

building and auxiliary building are 3-hour-rated-fire doors. All other doors

in the complex are 11⁄2-hour fire rated. Three-hour fire dampers are installed

in ducts that penetrate the wall from the control building to the auxiliary

building. -y

Fire detection, fire extinguishers, standpipe hose station, automatic preaction

sprinkler system (at selected locations), emergency lighting and a manual

smoke-venting system are provided for the control room complex.

Watts Bar SER 9-36
Both carpeting and a dropped suspended ceiling with a vinyl dust cover are to

be installed in the control room. By letter dated April 30, 1981, the applicant stated that the dropped ceiling panels in the main control room will be a

UL-listed material having a flame spread classification of 15 and a smoke

density of 101-200 when tested in accordance with ASTM E-84. The carpeting in

the control room will be material which has been tested in accordance with the

preceding standard. This material has a flame spread classification of 25, a

fuel contribution classification of 30, and a smoke density classification of

130. The staff has reviewed the ceiling tile and carpeting and concludes that

they meet the staff guidelines. Therefore, they are acceptable.

There are no smoke detectors installed above the suspended ceiling. At this

time, the concealed space is devoid of combustible material. A future modification or addition of any combustible material would be evaluated at the time of

modification.

Below the main control room consoles, a 3 ft X 4 ft access Walkway extends down

into the cable spreading room approximately 4 ft. This walkway is separated

from the cable spreading room by a 3-hour-fire-rated barrier. By letter dated

November 10, 1981, the applicant stated that all safety-related cabling that

passes through the enclosed walkway from the spreading room to the termination

strips on the main control room cabinets are enclosed in metal cable gutters

to a point just above the main control room floor where the cable gutters

interface with cable risers in the control room cabinets. The cabling enters

the metal gutters from the spreading room cable tray system at the bottom of

the enclosed raceway passing through 3-hour-fire-rated penetration seals. It

was the staff's concern that, because of a large exposed concentration of

cables, it would be difficult to locate and extinguish a fire should it develop

in the walkway. Because the metal gutters enclose the cables to a point just

above the room floor elevation, the cables are not in a fire propagating

configuration. Therefore, an automatic gas suppression system is not needed.

The existing manual fire extinguishers will provide adequate fire protection

for this area. The staff finds that the fire protection for this area meets

the guidelines of BTP ASB 9.5-1, Section F.2, and is, therefore, acceptable.

Other Plant Areas

The applicant's Fire Hazards Analysis addresses other plant areas not specifically

discussed in this report. The applicant has committed to install additional

detectors, fire barriers, and automatic sprinklers. The staff finds that the

fire protection for these areas with the commitment made by the applicant to

be in accordance with the guidelines of Appendix A to BTP ASB 9.5-1 and is,

therefore, acceptable.

Hydrogen Piping

A hydrogen piping system is routed through safety-related areas and, as such,

a fire or explosion may affect safety-related systems or equipment in the area.

At meetings held August 27, 1981 and February 18, 1982, the applicant verbally

stated that no hydrogen line is routed through safety-related areas needed for

shutdown. At those meetings, the staff requested the applicant to provide a

commitment that hydrogen lines that are routed through safety-related areas

Watts Bar SER 9-37
will be either seismically qualified or that flow limiters will be provided.

The staff will report on this matter in a supplement to this SER.

Auxiliary Control Room

To enhance plant shutdown capability, the applicant has provided an auxiliary

control room (ACR) which is completely separated and independent of the main

control room. In the event of a damaging fire in the main control room or

cable spreading room, plant shutdown capability can be maintained from the ACR.

The staff was concerned that a-spilled flammable liquid fire could affect the

auxiliary control room and the adjacent auxiliary control instrument rooms at

elevation 757 ft of the auxiliary building. By letter dated August 28, 1981,

the applicant agreed to install a curb on all four auxiliary control instrument

room openings to prevent such an incident.

Each instrument room at elevation 757 ft of the auxiliary building also has a

curtain-type fire damper opening located near the floor surface. The staff

was concerned that any material stored adjacent to the damper may inhibit its

motion and prevent its function. By letter dated August 28, 1981, the applicant

agreed to provide'a fixed wire mesh screen on the inside face of the fire damper

in each instrument room to ensure its proper function.

The staff finds the fire protection for this area to be in accordance with

Appendix A to BTP ASB 9.5-1 and is, therefore, acceptable.

Askarel-insulated Transformers

Askarel-insulated tranformers are located in various areas of the plant without

being located in a separate cut-off room. Also, in the immediate area of these

transformers are various redundant safety-related cable trays and/or conduit.

The staff was concerned that an explosion of an Askarel-insulated transformer

would damage these cable trays and conduits necessary for safe shutdown.

The following locations contain these transformers:

(1) ERCW intake structure, elevation 711 ft

(2) auxiliary building, elevation 692 ft, east and west ends

(3) auxiliary building, elevation 772 ft, room A-10, mechanical equipment room

(4) auxiliary building, elevation 737 ft, coordinates and A3

The applicant verbally agreed on August 27, 1981, to provide an analysis which

shows that the Askarel-insulated transformers cannot explode. This will be

documented in a future FSAR amendment, and the staff will report on this item

in a supplement to this SER.

9.5.1.6 Fire Protection for'Safe Shutdown Capability

By letter dated September 9, 1980, the applicant provided a comparison of his

fire protection for safe shutdown capability with the staff guidelines and a

separation analysis. The staff reviewed this information and finds that the

applicant's fire protection for safe shutdown does not meet the guidelines

because 1⁄2-hour-fire-rated barriers rather than 1-hour-fire-rated barriers are

provided where redundant shutdown systems are within 20 ft of each other in thi

Watts Bar SER 9-38
same fire area. The staff will require the applicant to install 1-hour fire

barriers and fixed'suppression systems as described in Sections 9.5.1.2 and

9.5.1.3 of this report to conform to meet the guidelines given in Appendix'R

to 10 CFR 50.

By letter dated August 28, 1981, the applicant provided verification that these

areas were analyzed in the fire protection interaction analysis. The applicant

has identified areas where additional protection and/or separation is required

to ensure a safe shutdown condition. As a result, the applicant has committed

to relocate one division to maintain a minimum of 20-ft separation between

divisions for the circuits as listed in his letter dated September 9, 1980.

The applicant has installed an alternate shutdown system for the control room,

cable spreading room, and auxiliary instrument rooms. The alternate shutdown

system is divided into five independent components (a central area and four

transfer switch rooms) that are separated from each other by 11⁄2-hour-fire-rated

barriers. An automatic suppression system is provided for the central area.

The controls and instrumentation are provided with transfer capability so that

they are isolated from both the cable spreading room and the control room. A

fire in either the control room or cable spreading room would not jeopardize

operation of the alternate shutdown panels, nor would a fire in the alternate

shutdown system cause functions in the control room or cable spreading room to

be lost. The entire alternate shutdown system is under review,.and the results

will be addressed in a supplement to this report.

9.5.1.7 Administrative Controls and Fire Brigade

The administrative controls for fire protection consists of the fire protection

organization, the fire brigade training, the controls over cumbustibles and

ignition source, the prefire plans and procedures for fighting fires, and

quality assurance. The fire brigade will be composed of five members per

shift. To have proper coverage during all phases of operation, members of each

shift crew will be trained in fire protection in accordance with NRC guidance,

including Regulatory Guide 1.101, "Emergency Planning for Nuclear Power Plants."

The applicant has agreed, by letter dated September 9, 1980, to implement the

fire protection program contained in the staff supplemental guidance "Nuclear

Plant Fire Protection Functional Responsibilities, Administrative Controls and

Quality Assurance," dated August 29, 1977, including (1) fire brigade training,

(2) control of combustibles, (3) control of ignition sources, (4) fire fighting

procedures, and (5) quality assurance.

The applicant will implement the plant administrative controls and procedures

before fuel loading.

The staff concludes that, with the commitments, the size of the fire brigade,

the necessary equipment, and the adequacy of the training, the administrative

controls will conform to the recommendations of the NFPA, to Appendix A to BTP

ASB 9.5-1, and to the supplemental staff guidelines. Therefore, it is

acceptable.

9.5.1.8 Appendix R Statement

On February 19, 1981, the Commission approved a rule concerning fire protection.

The staff uses the requirements of Appendix R as guidelines in the licensing

Watts Bar SER 9-39
of plants after January 1, 1979. On April 27, 1981, the Commission required

that operating licenses issued after January 1, 1979 contain a condition requiring compliance with commitments made by an applicant and agreed to by the staff

after differences between the applicant's program and the guidelines set forth

in Appendix A to BTP 9.5-1 and Appendix R to 10 CFR 50 have been identified and

evaluated. The applicant has not committed to meet the technical requirements

of Appendix R to 10 CFR 50 or provide equivalent protection. The operating

license will be conditioned to ensure that the applicant meets the guidelines

of Appendix R to 10 CFR 50 or to provide equivalent protection.

9.5.1.9 Conclusions

The following items are related to the guidelines of 10 CFR 50, Appendix R,

and are expected to be addressed by the applicant in July 1982:

(1) providing 1-hour fire barriers for redundant cables

(2) providing 8-hour emergency lighting

(3) maintaining closure of fire-rated doors to diesel generator building lube

oil storage area

(4) alternate shutdown system

For the two other open items, the staff requires further information from the

applicant. These are

(1) documentation on Askarel-insulated transformers

(2) a commitment that hydrogen lines that are routed through safety-related

areas will be either seismically qualified or will have flow limits

installed

The applicant has been informed of the necessity of resolution of their open

items so that all fire protection features can be implemented before fuel

load.

Once the applicant submits his response, the staff will review these items and

report its findings in a supplement to the SER.

In addition, the staff will complete its confirmatory review regarding compliance to the requirements of ASTM E-119 and will report its findings in a

supplement to the SER.

9.5.2 Communication Systems

The communication system (discussed below) is designed to provide reliable

intraplant and interplant (plant-to-offsite) communications under both normal

plant operation and accident conditions.

Watts Bar SER 9-40
9.5.2.1 Intraplant System

Telephone Systems

(1) Private Automatic Exchange (PAX) with Code Call

A 200-line PAX is installed to provide primary two-way communications

throughout the Watts Bar nuclear plant. This PAX is equipped with provisions for: regular two-way telephone conversation, executive right-ofway, code call and answer, fire alarm, revertive call switch, single-digit

access to electrical control room manual telephone switchboard, paging

call and answer over PAX instruments, and single-digit access to system

direct dialing circuits.

The power for the PAX is supplied from a 48-V dc source consisting of two

100-amp battery chargers, a regulating power board, and a battery. The

chargers are powered by dual ac voltage sources through train "A" and

train "B" circuits from diesel-backed boards.

The battery is sized to carry the load at full capacity for 8 hours

without the chargers.

The main switching equipment for this system will be located in the

communications room which is in a seismic Category 1 building.

Code call uses the same end devices as the paging and alarm systems that

are installed throughout the plant area. The PAX is equipped with a

special code-call switch to permit dialing of code numbers and answering

by dialing a special answer number. The conversation is limited to 60 sec

to prevent tie-up of the equipment.

(2) Manual Telephone Switchboard

A 20-line, cordless, pushbutton switchboard is installed on the electrical

control area desk between Units 1 and 2 control rooms. This board is

equipped as follows: I key - executive right-of-way trunk; 13 keys -

future; 2 keys - 2-way trunks to PAX; 1 key - fire alarm trunk; and 3 keys

- touch tone to carrier and microwave. Provision is made for remote

answering of the turret from the Unit 1 control desk,, Unit 2 control desk,

and the shift engineer's office.

(3) Administrative Telephone System

These multiple-line telephones are installed in various offices in the

plant. The administrative telephone system switching equipment will be

located in the office building and will function without the PAX for those

trunks which are not PAX trunks.

Two of the lines on each instrument are used for intercom. Power is

supplied from a local 120-V ac convenience outlet.

Watts Bar SER 9-41
Intercom Systems

(I) Administrative Intercom System

This intercom system is an integral part of the administrative telephone

system; it consists of two separate intercom lines and is common between

all the instruments in the administrative telephone system. Power is supplied from the administrative telephone system power supply.

(2) Division of Power System Operations (DPO) Engineers' Intercom

This intercom system consists of six station master intercoms and/or staff

intercoms in the supervisor's office, DPO engineers' shop, relay room,

and communication room. Power is supplied from a local 120-V ac source.

(3) Power Stores Intercom and Paging System

This intercom system consists of 12 station master intercoms, remote

speakers, intercom call back pushbuttons, and staff intercoms in the

storekeeper's office and storeroom. Power is supplied from a local 120-V

ac source.

(4) Operational Intercom

This intercom system consists of 12 station master intercoms located in

the Units 1 and 2 control rooms, the electrical control room, and shift

engineer's office. Power is supplied from a local 120-V ac convenience

outlet.

Sound-Powered Telephone Systems

(1) Plant Operation

The following separate systems are provided for each of the two units (a

total of 12 systems):

SPI - sound-powered jack system for turbine control and auxiliary

power system

SP2 - sound-powered jack system for feedwater, steam, and condensate

system

SP3 - sound-powered jack system for reactor control

SP4 - sound-powered jack system for reactor coolant and auxiliary

steam

SP5 - sound-powered jack system for engineered safeguards and

auxiliaries

SP6 - sound-powered jack system for reactor refueling system

Watts Bar SER 9-42
(2) Backup Control Center System

This system consists of two completely redundant systems. Each subsystem

is-wired direct and independent of any other communications systems. .Wiring routes avoid the spreading room, unit control rooms, and auxiliary

instrument room. Sound-powered equipment and circuits are provided in the

diesel generator building, 480-V ac shutdown board room, 6.9-kV ac shutdown

board rooms, auxiliary building, and auxiliary control room.

(3) Health Physics System

A direct circuit is provided between the health physics office and the

Units I and 2 control rooms (physically on the electrical control area

desk). Two pairs are provided between these locations--one for magneto

signaling and one for talking over sound-powered handsets.

(4) Diesel Building to Main Control 'Room

A direct circuit is provided between the shielded waiting room in the

diesel generator building and the main control room at the diesel generator

control panel. A four-conductor waterproof cable is provided between

these locations--one twisted pair for magneto signaling and oneltwisted

pair for talking over sound-powered handsets.

The sound-powered systems do not require any power supply. All required energy

is generated by the speaker.

Closed-Circuit Television

(1) Radwaste

Two cameras are provided for remotely viewing radwaste packaging operations. One camera is mounted on a trolley atop the shield wall, and the

other is mounted on the overhead crane. Two 14-in. monitors and the

necessary camera control equipment are located in a control console behind

the shield wall.

(2) Underwater

One portable underwater camera is provided for remote viewing and permanent

documentation via videotape of all underwater operation. One 14-in.

monitor, two videotape recorders, and the necessary camera control equipment and cable are mounted on a wheeled cabinet to provide a portable

system.

(3) Reactor Containment and Control Room

Two portable cameras are provided for remotely viewing refueling operations. Permanent wiring for the cameras is terminated in plug receptacles

on the refueling floor and lower compartment of each unit and at the

common spent fuel pit area. A 14-in. monitor and video switcher are provided in the electrical control room. This system is to be used primarily

by operations but is also useful in providing training information.

Watts Bar SER 9-43
Codes, Alarms, and Paqing System

This system is designed to utilize a solid-state amplifier speaker end device

which is capable of 110 db at 10 ft and is shared between the code call,

paging, evacuation, all clear, and'fire alarm systems. The order of priority

is fixed by relay logic as follows:

(1) evacuation alarm

(2) all clear

(3) fire alarm

(4) code call

(5) paging

The solid-state tone generators are plug-in devices and may be easily replaced.

Eight different tones are available, of which four are used.

There are two sources of 24-V dc power distributed to the speaker-amplifiers,

and approximately half in each area of the plant are supplied from each source.

Each source is quite reliable because it is supplied from chargers that are

backed up by batteries capable of supplying the load for 8 hours.

The paging system shares the code call and alarms system speakers that are.

installed throughout the entire plant area. Paging handsets are provided on

both unit control desks, the electrical control desk, and on the shift engineer's desk, and in the auxiliary control room. In addition to the paging

handset locations, this equipment may be accessed from any PAX telephone.

Paging and code call are equal in priority except that paging can be temporarily advanced in priority to second so that the person actuating the evacuation signal may give oral instructions. Single or multiple open circuits or

amplifier failures in individual units will not prevent the remaining equipment

from functioning. The failure of the PAX equipment will not impair the use of

the paging equipment from the local paging stations at the unit operator's

desk, the electrical control room operator's desk, the shift engineer's desk,

and the auxiliary control room.

9.5.2.2 Interplant (Plant-to-Offsite) Communication Systems

Microwave Radio

Two microwave circuits provide access to the power system control center

(PSCC). Forty-two channels are provided to the Watts Bar Hydro Plant repeater

terminal and 18 channels to the Montlake Repeater Station. The circuit to

Watts Bar Hydro employs frequency diversity radio frequency (RF) equipment; and

the Montlake circuit operates in the hot standby mode.

The terminals are housed in the communications room and are connected to

roof-mounted parabolic antennas via waveguide. These antennas are aligned

with tower-mounted passive reflectors to direct the signal to the proper

repeater station.

A 24-V battery and charger-power board system is installed for the exclusive

use of both these microwave circuits. It is fed from two separate ac sources,

and the battery is capable of operating the system for a minimum of 8 hours

without chargers.

Watts Bar SER 9-44
Insulated Shield Wire Carrier (ISWC) Circuits

Two ISWC circuits are provided. The major electronic portions of the system

is housed in the communications room. One is routed to the Roane 500-kV

Substation and one to the Sequoyah nuclear plant. The circuit to the Roane

500-kV Substation is cross-connected to the microwave system at Watts Bar for

extension to the PSCC. The circuit to Sequoyah is cross-connected to the

microwave system at Sequoyah for extension to the PSCC.

The equipment is solid-state and is powered by the 48-V dc battery chargerpower board system for the PAX telephone system.

Commercial Telephone Service

Public telephone service is provided for various offices throughout the plant.

The phones for plant superintendent and assistant, the administrative officer,

and the receptionist have access to the Chattanooga TVA PAX. The commerical

telephone lines are terminated on protectors located in the communications

room and extended to instruments located throughout the plant. The operation

of these lines is not dependent upon onsite power.

VHF Radio Systems

There are three VHF radio systems at Watts Bar, and none of the three systems

is located in the communications room. Thus, complete loss of the communications room would not cause the loss of these links. Each of these radio

systems has a separate location so that it is unlikely that all three would be

out of service simultaneously. The three radio systems are power system operations VHF radio system, public safety service VHF systems, and radiological

VHF radio system.

(1) Power System Operations VHF Radio

This system consists of a base station located in the relay board room

with control stations in the control room and the DPO Engineers' Office.

It is intended for daily use by the DPO engineers and will not be used in

the control room except in emergencies. The system is powered by redundant

power supplies fed from two separate ac sources..

(2) Public Safety Service (PSS) VHF Radio

This system consists of VHF radio mobile equipment in PSS vehicles,

portables in the possession of PSS officers, and fixed equipment in the

gatehouse and in the visitor's reception area of the plant. This system

is not intended for use by plant operating personnel, but it could be

used in an emergency.

(3) Radiological VHF Radio

This system consists of a base station and a control station located in

the meteorological building with additional control stations in the control room and in the health-physics laboratory. Also included with this

system are 10 pocket-type portable receivers assigned to key operations

supervisors for after-hours offsite emergency contact.

Watts Bar SER 9-45
The applicant was asked to identify all working stations in the plant where it

may be necessary for plant personnel to communicate with the control room during and/or following transients and/or accidents in order to mitigate the consequences of the event and to attain a safe cold plant shutdown. He was also

requested to provide for each of these areas the maximum sound levels that

could exist, the types of communication available, ability of the systems to

provide adequate communication with maximum background noise, and performance

requirements and tests needed to ensure adequate communications. The applicant

has provided a list of communication systems in the control room, auxiliary

control room, and nearby working stations. By letter dated February 26, 1982,

the applicant provided all requested information except a commitment to test

the communication systems in accordance with Regulatory Guide 1.68. The staff

requires that the communication systems be preoperationally tested in accordance

with Regulatory Guide 1.68 to demonstrate that the systems will function

properly and provide adequate communication with the maximum potential background noise levels. With the imposition of this requirement as a condition of

the license, the staff considers this item resolved.

The scope of the review of the Watts Bar communication system included assessment of the number and types of communication systems provided, assessment and

adequacy of the power sources, and verification of functional capability of the

communication systems under all conditions of operation.

The basis for acceptance in the staff review was conformance of the design criteria and bases and design of the installed communication systems to the acceptance criteria in Section II of the SRP 9.5.2. Other bases for acceptance were

conformance to industry standards and the ability of the systems to provide

effective communications from diverse means within the Watts Bar plant during

normal and emergency conditions under maximum potential noise levels.

Based on its review, the staff concludes that the communication systems for

Watts Bar are in conformance with the above cited standards, criteria and

design bases, can perform their design functions, and, therefore, are acceptable.

9.5.3 Lighting System

The lighting system for Watts Bar is designed to provide adequate lighting in

all areas of the station and consists of normal and standby ac lighting systems

and an emergency dc lighting system. The design is based on illumination

levels that equal or exceed those recommended by the Illuminating Engineering

Society for central stations.

The normal lighting system is for general lighting of the plant. The major

power supply is through two alternate feeders from the 6.9-kV common boards A

and B to selective and interrupter switch and 3-phase 6900-120/208-V ac transformers, feeding a lighting board. These lighting boards are located in the

turbine and auxiliary building of the main plant. Other lighting boards in

the service building, office building, gatehouse, and so forth, are fed from

480-V boards through 3-phase 480-120/208-V ac transformers. These lighting

boards feed the normal lighting cabinets distributed throughout the main plant.

In the power boards and control rooms, alternate rows of fixtures or alternate

fixtures are fed from different lighting boards to prevent total blackout in a

particular area in case of failure of one or the other lighting boards or

cabinets.

Watts Bar SER 9-46
The standby lighting system forms a part of the normal lighting requirements

and is normally energized at all times. This system is fed from 480-V control

and auxiliary building vent boards to 3-phase 480-120-V ac transformers to each

standby lighting cabinet. The control and auxiliary building vent boards have

a normal and alternate ac power supply and, in case of the failure of these

supplies, the vent boards are fed from the diesel generators. The cable

feeders to the standby cabinets located in the seismic Category 1 structure are

routed in redundant raceways, and the fixtures are dispersed among the normal

lighting fixtures.

The third lighting system is the emergency system. The feeder to this system

is electrically held in the off position until a power failure occurs on the

ac systems. Then the emergency lighting cabinets are automatically energized

from the 125-V dc vital battery boards. 'This system is an essential supporting

auxiliary system for the ESF, and the cable feeders to the lighting cabinets

are routed in the redundant ESF cable tray system or in conduit. The fixtures

are incandescent type and are dispersed among the normal and standby fixtures

with alternate emergency fixtures being fed from redundant power trained lighting cabinets. Each diesel generator unit has a lighting cabinet which supplies

the normal lighting for that unit. The diesel generator building lighting cabinets are fed through 480-208/120-V 3-phase local lighting transformers, which

in turn are fed from the associated diesel generator 480-V shutdown boards.

The diesel generator will start within the prescribed time to provide the

480-V ac power requirements to the 480-V shutdown boards, thus supplying power

again to the diesel generator building lighting transformers.- Low-level lighting required for maintenance or operating procedures and ingress/ egress in

the event of loss of normal lighting is supplied from fixtures with a selfcontained battery and inverter charger.

The plant lighting systems are designed so that a single failure cannot degrade

the essential lighting below a safe level. The plant lighting systems are

tested at installation and provisions are installed for testing the dc emergency

system.

The scope of the review of the lighting system for Watts Bar included assessment

of all components necessary to provide adequate lighting during both normal

and emergency operating conditions, the adequacy of the power sources for the

normal and emergency lighting systems, and verification of functional capability

of the lighting system under all conditions of operation.

The basis for acceptance in the staff review was conformance of the design

bases and criteria and the design of the lighting systems and necessary auxiliary supporting systems to the acceptance criteria in Section II of SRP

Section 9.5.3.

Other bases for acceptance were conformance to industry standards and.the

ability to provide effective lighting in all areas of the Watts Bar plant under

all conditions of operations.

Based on its review, the staff concludes that the various lighting systems provided at the Watts Bar plant are in conformance with the above-cited standards,

criteria and design bases, can perform their design function, and are, therefore, acceptable.

Watts Bar SER' 9-47
9.5.4 *Emergency Diesel Engine Fuel Oil Storage and Transfer System

9.5.4.1 Emergency Diesel Engine Auxiliary. Support Systems (General)

There are four emergency diesel generators for the plant (two per unit), and

each diesel engine has the following auxiliary systems, which are addressed in

detail in the SER sections indicated:

(1) fuel oil storage and transfer system (Section 9.5.4.2)

.(2) cooling water system (Section 9.5.5)

(3) starting system (Section 9.5.6)

(4) lubrication system (Section 9.5.7)

(5) combustion intake and exhaust system (Section 9.5.8)

This section of the SER applies to all of the above systems.

Except for the diesel generator exhaust stacks and the fuel oil fill and the

vent lines, the diesel generator and its auxiliary support systems, including

the fuel oil storage tanks that are buried in the building foundation, are

housed in a seismic Category I diesel generator building which provides substantial (but not total) protection from the effects of tornadoes, tornado

missiles, and floods. The design of the diesel generator building roof contained openings that provide potential pathways'for tornado missiles such that

safety-related diesel generator equipment could be damaged by the missiles.

The applicant has committed to install tornado-missile protection for the

*equipment. Tornado-missile protection for the exhaust stacks is discussed in

Sections 3.5.2 and 9.5.8. Protection from the effects of tornadoes, tornado

missiles, and floods is evaluated in Section 3 of this report. Tornadomissile protection of the diesel generator fuel oil fill and vent lines is

discussed in Section 9.5.4.2.

As discussed in the following sections, the requirements of GDC 2 and 4, and

the recommendations and guidance of Regulatory Guides 1.115 ("Protection

Against Low Trajectory Turbine Missiles") and 1.117 are not fully met.

The staff does not have sufficient information to determine if the diesel

generator auxiliary support systems can perform their design safety functions

under all conditions. Upon receipt of theorequested information, the staff

will report its findings in a supplement to the SER.

Each diesel generator and its auxiliary systems is independent of the other.

The only item that is shared is the auxiliary fuel oil fill connection to the

yard storage tank. This portion of the system is nonseismic and properly isolated from the balance of the systems. Therefore, the requirements of GDC 5

are met.

The diesel engine and its engine-mounted and separately skid-mounted portions

of the auxi.liary support systems piping and components normally furnished with

the diesel generator package are designed to seismic Category I requirements

and follow the guidelines of the Diesel Engine Manufacturers Association (DEMA)

standards. The diesel engine and its mounted auxiliary support systems piping

and components conform to the requirements of IEEE Standard 387-1977, which

endorses the DEMA standard and the guidelines of Regulatory Guide 1.9. The

Watts Bar SER 9-48
diesel engine and its auxiliary support systems meet the quality control requirements of 10 CFR 50, Appendix B. The quality assurance program is evaluated in

Section 17.0 of this report.

The applicant has defined the engine-mounted piping as all piping on the engine

and all piping on the engine auxiliary skid up to the first welded, screwed,

or flanged connection on the skid. The staff finds this unacceptable. The staff

defines the engine-mounted piping as that piping from the engine block to the

engine interface. This interface is the first welded, screwed, or flanged connection off the engine block. The applicant has been informed of this definition,

and all quality classification boundaries discussed in Sections 9.5.4.1 through

9.5.8 are based on this definition.

Accumulation of dust--including dust generated from concrete floors and walls--

on the electrical equipment associated with starting of the diesel generators

(such as, auxiliary relay contacts and control switches) is limited by the diesel

generator building ventilation system design and operation, plant design and

administrative procedures.

The applicant will perform preoperational and startup tests of the diesel

engine auxiliary support systems in accordance with recommendations and guidelines of Regulatory Guide 1.68. The adequacy of the test program is evaluated

in Section 14 of this report.

The design of the diesel engine auxiliary support systems is evaluated with

respect to the recommendations and guidelines of BTP ASB 3-1, "Protection

Against Postulated Piping Failures in Fluid System Piping Outside Containment,"

and MEB 3-1, "Postulated Break and Leakage Locations in Fluid System Piping

Outside Containment." Evaluation of protection against dynamic effects

associated with the postulated pipe system failures is covered in Section 3.6

of this report.

The adequacy of the fire protection for the emergency diesel generator and

associated auxiliary support.systems with respect to the recommendations and

guidelines of BTP CMEB 9.5-1, "Guidelines for Fire Protection for Nuclear Power

Plants," is evaluated in Section 9.5.1 of this report.

The designs of the diesel generator auxiliary support systems also have been

evaluated with respect to the recommendations of NUREG/CR-0660, "Enhancement of

Onsite Emergency Diesel Generator Reliability." This report made specific

recommendations on increasing the reliability of nuclear power plant emergency

diesel generators. Information requests concerning these recommendations were

transmitted to the applicant during the review process. The applicant responded

in the amendments to the FSAR stating how he meets or will meet the recommendations of NUREG/CR-0660.

The staff has reviewed these responses and determined that conformance to the

recommendations is as listed in Table 9.1.

On the basis of its review, the staff concludes that there i.s sufficient

assurance of diesel generator reliability to warrant unrestricted plant'operation through the first refueling period. However, to ensure long-term reliability of the diesel generator installations, the staff requires that the following

Watts Bar SER 9-49
Table 9.1 Conformance with recommendations

of NUREG/CR-0660

Recommendation Conformance SER Section

(1) Moisture in air starting system Partial 9.5.6

(2) Dust and dirt in diesel generator room Yes 9.5.8

(3) Turbocharger gear drive problem Partial 9.5.4.1

(4) Personnel training Yes 9.5.4.1

(5) Automatic prelube No 9.5.7

(6) Testing, test loading, and preventive Partial 9.5.4.1

maintenance

(7) Improve the identification Yes 9.5.4.1

of root cause of failures

(8) Diesel generator ventilation and Yes 9.5.8

combustion air systems

(9) Fuel storage and handling No 9.5.4.2

(10) High temperature insulation * 9.5.4.1

for generator

(11) Engine cooling water Yes 9.5.5

(12) Concrete dust control Yes 9.5.4.1

(13) Vibration of instruments Yes 9.5.4.1

*Explicit conformance is considered unnecessary by the staff in view of the

equivalent reliability provided by the design, margin, and qualification testing requirements that are normally applied to emergency standby diesel

generators.

design and procedural modifications be implemented prior to startup following

the first refueling:

(1) Moisture in the Air-Starting System

This item is discussed in Section 9.5.6 of this report.

(2) Turbocharger Gear Drive Problem

The diesel generators at Watts Bar have a turbocharger mechanical drive

gear assembly whose gear ratio is 18.1. This drive gear assembly has not

Watts Bar SER 9-50
been designed to operate at no-load or light-load conditions and fullrated speed for prolonged periods. The manufacturer, Electro-Motive

Division (EMD) of General Motors Corporation, has developed or has under

development heavy duty turbocharger drive gear assemblies that meet the

recommendation of NUREG/CR-0660. To improve the reliability and availability of the diesel generators on demand, the staff requires the installation of a heavy duty turbocharger drive gear assembly as recommended by

NUREG/CR-0660. The applicant, in a letter dated October 9, 1981, has

agreed to comply with this position.

(3) Automatic Prelube

This item is discussed in Section 9.5.7 of this report.

(4) Test Loading, No-Load and Low-Load Operation

The applicant had stated that operating instructions are written to minimize no-load and low-load operation and will contain a precautionary

statement against these conditions. The staff found this unacceptable and

so informed the applicant. In letters dated October 9, 1981, and

January 19, 1982, the applicant committed to the following: "After four

hours of operations at less than 30 percent load, the diesel generator is

run at a minimum of 50 percent load for at least 30 minutes. After an

accident situation when the diesel generator has run for an extended

period of time at low or no load, the load is gradually increased until

the exhaust smoke is approximately twice as dense as normal. The increasing load is then stopped until the smoke clears. This procedure is

repeated until full load can be carried with a clear exhaust." By letter

dated March 3, 1982, the applicant defined "extended period of time" to be

8 hours. The staff finds the above to be acceptable, provided it is

incorporated in the plant operating procedures. The staff will verify

this before fuel load.

(5) Fuel Storage and Handling

This item is discussed in Section 9.5.4.2 of this report.

The present diesel generator design meets the requirements of GDC 17, 18, and

21. Upon completion of the above changes and modifications, the design of the

diesel generator and its auxiliary systems will also be in conformance with

recommendations of NUREG/CR-0660 for enhancement of diesel generator reliability and the related NRC guidelines and criteria. The staff, therefore, concludes that this will provide reasonable assurance of diesel generator reliability through the design life of the plant.

9.5.4.2 Emergency Diesel Engine Fuel Oil Storage and Transfer System

The design functions of the emergency diesel engine fuel oil storage and

transfer system are to provide a separate and independent fuel oil supply train

for each diesel generator and to permit operation of the diesel generator at

engineered safety feature load requirements for a minimum of 7 days without

replenishment of fuel. The system is to be designed to meet the requirements

of GDC 2, 4, 5 and 17. The meeting of the requirements of GDC 2, 4, and 5 is

discussed in Section 9.5.4.1 above.

Watts Bar SER 9-51
There are four emergency diesel generators for the plant (two per unit). The

diesel generators are of the tandem type with a pair of diesel engines connected

to a generator. Each diesel generator fuel oil storage and transfer system

consists of two 550-gal fuel oil day tanks (one per diesel engine) sufficient

to power the diesel engine at rated load for approximately 2 hours; four

embedded tanks with a total capacity of 68,000 gal of fuel storage sufficient

to power the diesel engine at maximum Continuous load conditions for 7 days;

two ac motor-driven transfer pumps powered from the associated diesel; filters;

strainers; and the associated piping, valves, instrumentation, and controls.

In addition there are two 71,000-gal nonseismic yard fuel oil storage tanks

from which the diesel generator 7-day storage tanks can be refilled.

Each diesel engine fuel oil storage and transfer system is independent and

physically separated from the other systems supplying the other diesel generators. A single failure within any one of the systems will affect only the

associated diesel generator. Therefore, the requirements for GDC 17, as

related to the capability of the fuel oil system to meet independence and

redundancy criteria, are met.

Except for the diesel oil storage tank fill and vent lines and the auxiliary

skid-mounted piping and components from the day tank to the diesel engine

interface, the diesel engine fuel oil storage and transfer system piping and

components up to the diesel engine day tank are designed to seismic Category I,

ASME Section III, Class 3 (Quality Group C) requirements. They also conform to

the guidelines of Regulatory Guides 1.26 and 1.29.

The engine-mounted piping and components, from the engine block to the engine

interface, are considered part of the engine assembly and are seismically

qualified to seismic Category I requirements as part of the diesel engine

package. This piping and the associated components--such as valves, fabricated

headers, fabricated special fittings, and the like--are designed, manufactured,

and inspected in accordance with the guidelines and requirements of ANSI

Standard B31.1, "Code for Pressure Piping"; ANSI N45.2, "Quality Assurance

Program Requirements for Nuclear Power Plants"; and 10 CFR 50 Appendix B. The

engine-mounted fuel oil piping and associated components are intentionally over

designed (subjected to low working stresses) for the application, thereby

resulting in high operational reliability. The design of the engine-mounted

fuel oil piping and components to the cited design philosophy and standards is

considered equivalent to a system designed to ASME Section III Class 3 requirements with regard to system functional operability and inservice reliability.

The diesel oil storage tank fill line and the auxiliary skid-mounted piping and

components from the day tank to the diesel engine interface are designed as

seismic Category I. The system description and diagrams indicate that this

piping is designed to ANSI B31.1 and is Quality Group D. This is not acceptable.

The staff requires the above piping and components be designed to ASME Section III, Class 3 (Quality Group C) requirements and conform to the guidelines

of Regulatory Guide 1.26. The applicant has been informed of this position.

The staff will pursue this issue with the applicant, and report the results in

a supplement to the SER.

The design of the emergency diesel engine fuel oil storage and transfer system

does not conform to ANSI-N195, "Fuel Oil Systems for Diesel Generators." In

addition, the system does not completely conform to the fuel oil quality and

Watts Bar SER 9-52
tests and guidelines of Regulatory Guide 1.137, "Fuel Oil Systems for Standby

Diesel Generators," Positions C.2.a through C.2.g.

The applicant was requested to provide a more detailed description of the fuel

oil storage and transfer system. Sufficient information had not been provided

to determine acceptability of the system with regards to tornado-missile

protection of the fuel oil storage fill and vent lines. The applicant, in a

letter dated October 9, 1981, provided the justification why the fill and

vent lines should not be protected. The staff found this justification not

acceptable. By letter dated March 17, 1982, the applicant committed to provide

missile protection for the fuel oil storage tank-vent lines. The staff finds

this acceptable pending verification. During a June 21, 1982 telecon, TVA

stated that in the event all four fuel and storage tank fill lines are rendered

inoperable by tornado missiles, the tanks can be filled through the tank manway

openings located in the hallway area of the diesel generator building.

The hallway area is separated from the four diesel generator rooms. The

alternate mode of filling the storage tanks will require routing a hose (qualified for this service) from the delivery vehicle through the C02 tank storage

room, the lube oil storage room, and into the hallway. The staff finds this

alternate mode of filling the fuel oil storage tanks acceptable provided that

fire watches are stationed in these areas during the tank filling period. This

requirement will be included in the Technical Specifications.

The scope of review of the.diesel engine fuel oil storage and transfer system

included layout drawings, piping and instrumentation diagrams, and descriptive

information in Section 9.5.4 of the FSAR for the system and auxiliary support

systems essential to its operation.

The bases for acceptance in the staff review was conformance of the design

criteria and bases and design of the diesel engine fuel oil storage and transfer system to the requirements of GDC 17 with respect to redundancy and physical

independence, the guidance of the cited Regulatory Guides, the recommendations

of NUREG/CR-0660, and industry codes and standards.

Based on its review, the staff concludes that the emergency diesel engine fuel

oil storage and transfer system meets, except as stated above, the requirements

of GDC 2, 4, 5, and 17, the guidance of the cited Regulatory Guides, and the

recommendations of NUREG/CR-0660 and industry codes and standards. The staff

will report the results of its review of the requested additional information

in a supplement to this SER.

9.5.5 Emergency Diesel Engine Cooling Water System

The design function of the emergency diesel engine cooling water system is to

maintain the temperature of the diesel engine within a safe operating' range

under all load conditions and to maintain the engine coolant preheated during

standby conditions to improve starting reliability. The system is to be

designed to meet the requirements of GDC 2, 4, 5, 17, 44, 45, and 46. The

meeting of the requirements of GDC 2, 4, and 5 is discussed in Section 9.5.4.1

above.

The emergency diesel engine cooling water system is a closed loop system and

cools the cylinder liners, cylinder heads, and turbocharger combustion aftercoolers. The major components of this system for each diesel engine include

Watts Bar SER 9-53
turbocharger aftercoolers, jacket water cooler, two engine driven jacket water

coolant pumps, an expansion tank, a lube oil cooler, an electric emersion

heater, a thermostatic three-way valve, required instrumentation, controls and

alarms, and the associated piping and valves to connect the equipment. When

the diesel engine is operating, the heat generated is rejected to the essential

raw cooling water system by means of the jacket water cooler.

During operation of the diesel engine, temperature regulation of the diesel

engine coolant is accomplished automatically through the action of a temperature sensing three-way thermostatic valve.. When the engine is idle, the engine

coolant is heated to a temperature of 115'F to 125°F by an electric heater and

continuously circulated by natural circulation through the lube oil cooler.

The heated lube oil is returned to the engine sump and provides the heat necessary to preheat the cooling water in the engine. This is accomplished by a

combination of radiation, convection, and conduction of heat through the metal

parts of the engine and engine block to the cooling water. Natural circulation

of the cooling water through the engine block and the balance of the system,

preheats the cooling water system. The temperature is controlled by a thermostat to keep the engine warm and ready to accept Toads within the prescribed

time interval. Alarms have been provided to enable the control room operator

to monitor the diesel generator cooling while the unit is in the standby mode

or in operation.

The diesel generator room supply ventilation air does not appear to be preheated

nor does the room seem to be heated in any manner. This plant is located in an

area where the temperature can drop to below freezing levels. Therefore, the

diesel generator room temperature could conceivably approach outside ambient

temperature. With the diesel generator room at or below freezing level, the

means of preheating the entire engine cooling water volume may not be adequate

to keep the engine sufficiently preheated to ensure a Successful fast start and

load-accepting capability in an emergency. Improper preheating of the diesel

engine units may prevent performance of their required safety function and may

degrade availability of the diesel generators to an unacceptable level. The

staff finds this unacceptable. The staff has requested additional information

from the applicant, and will report on its findings upon receipt of the

requested information.

There are two emergency diesel generators for each unit of the Watts Bar plant,

and each has a physically separate and independent cooling water system.

Therefore, the requirements of GDC 17 and 44 as related to redundancy and

single-failure criteria are met.

The diesel engine cooling water system piping and components up to the diesel

engine interface, including auxiliary skid-mounted piping, are designed to

seismic Category I requirements and meet the recommendations of Regulatory

Guide 1.29.

The diesel engine cooling water system piping and components for all engines up

to the engine interface including auxiliary skid-mounted piping is designed and

classified as Quality Group D. This is unacceptable. The staff requires that

this piping be designed to ASME Section III Class 3 (Quality Group C) requirements and meet the recommendations of Regulatory Guide 1.26. The applicant has

been informed of this position. The staff will pursue this issue with the

applicant, and report the results in a supplement to the SER.

Watts Bar SER 9-54
The engine-mounted piping and components, from the engine block to the engine

interface, are considered part of the engine assembly and are seismically

qualified to seismic Category I requirements as part of the diesel engine

package. This piping and the associated components--such as valves, fabricated

headers, fabricated special fittings, and the like--are designed, manufactured,

and inspected in accordance with the guidelines and requirements of ANSI

Standard B31.1, ANSI N45.2, 10 CFR 50 Appendix B. The engine-mounted cooling

water piping and' associated components are intentionally over designed (subjected to low working stresses) for the application, thereby resulting in high

operational reliability. The design of the engine-mounted cooling water piping

and components to the cited design philosophy and standards is considered

equivalent to a system designed to ASME Section III, Class 3 requirements with

regard to system functional operability and inservice reliability.

The diesel engine cooling water system conforms with Regulatory Guide 1.9,

Position C7, as it relates to engine cooling water protective interlocks. The

diesel generator system protective interlocks are discussed in Section 8.3 of

this report.

The diesel engine cooling water system has provisions to permit periodic

inspection and functional testing during standby and normal modeslof power

plant operation as required by GDC 45 and 46.

The applicant was requested to provide a more detailed description of the

cooling water system. Sufficient information has not been provided in the

following areas to determine acceptability of the system. These areas are

(1) The time period the diesel generator is capable of operating fully loaded

without secondary cooling has not been addressed. Sufficient water is

needed in the engine and expansion tank to absorb the heat generated during this period. The time period the diesel generator can operate fully

loaded without secondary cooling should be in excess of the time needed to

restore essential raw cooling water to the diesels in the event of a loss

of offsite power.

(2) The expansion tank is not located at the highest point in the system;

thus, there is the possibility of the system to become air bound with

potential damage from waterhammer. This has been addressed by the applicant, but the response is unacceptable.

The scope of review of the emergency diesel engine cooling water system included

layout drawings, piping and instrumentation diagrams, and descriptive information in Section 9.5.5 of the FSAR for the system and auxiliary support systems

essential to its operation.

The basis for the acceptance in the staff review was conformance of the design

criteria and bases and design of the diesel engine cooling water system to the

GDC 17 and 44 with respect to the redundancy and physical independence and

GDC 45 and 46 with respect to inspection and testability of the system; to the

guidance of the cited Regulatory Guides; and to the recommendations of

NUREG/CR-0660 and industry codes and standards; and the ability of the system

to maintain stable diesel engine cooling water temperature under all load

conditions.

Watts Bar SER 9-55
Based on its review, the staff concludes that the emergency diesel engine

cooling water system meets, except as stated above, the requirements of GDC 2,

4, 5, 17, 44, 45, and 46 and meets the recommendations of NUREG/CR-0660 and

industry codes and standards. However, the staff concludes that the system

does not fully conform with the requirements of GDC 2 and 4 and the guidance of

the cited Regulatory Guides, since the staff does not have sufficient information to determine if the system can perform its design safety function under

all conditions, and, therefore, it is not acceptable. Upon receipt of the

requested information, the staff will report its findings in a supplement to

this SER.

9.5.6 Emergency Diesel Engine Starting Systems

The design function of the emergency diesel engine starting system is to provide a reliable method for automatically starting each diesel generator so that

the rated frequency and voltage is achieved and the unit is ready to accept

required loads within 10 seconds. The system is to be designed to meet the

requirements of GDC 2, 4, 5, and 17. The meeting of the requirements of GDC 2,

4, and 5 is discussed in Section 9.5.4.1 of this SER.

There are four emergency tandem diesel generators (two per generator) for the

Watts Bar plant. Each emergency diesel engine has an independent and redundant

air-starting system consisting of two separate full-capacity air-starting subsystems each with sufficient air capacity to provide a minimum of five consecutive cold engine starts. Redundancy in starting system is provided so that a

malfunction or failure in one system does not impair the ability of the other

system to start its diesel engine. This meets the requirements of GDC 17.

Each subsystem of each engine-starting system includes an air compressor, a

receiver tank, intake air filters, two air-starting motors and starting valves,

equipment. Alarms annunciate on the local panel and in the main control room

to enable the operators to monitor the air pressure of the diesel generator

air-starting system.

The air-starting system for the diesel generators has no air dryer installed

downstream of the air compressors. The applicant relies on periodic removal of

excess water from the starting air receivers. Operating experience has shown

that accumulation of water inithe air-starting system has been one of the most

frequent causes of diesel engine failure to start on demand. Thus, the staff

requires that the engine-starting air be dried. Air at the outlet of the

receivers is saturated to the system air pressure and temperature. Air flow

from this point accompanied with reduction in system pressure and/or temperature

will result in water condensation. Because periodic draining of the air

receivers will not provide dry diesel engine starting air, the staff requires

that air dryers be installed upstream of the air receivers. The applicant in

Amendment 44 to the FSAR has committed to install air dryers in the diesel

air-starting system. In a letter dated October 9, 1981, the applicant committed to install the air dryers before fuel load. The staff finds this commitment acceptable, and will verify installation before fuel load.

The diesel engine air-starting system piping and components from the isolation

valves before the receivers to the diesel engine interface, including auxiliary

skid-mounted piping, are designed to seismic Category I requirements and the

recommendations of Regulatory Guide 1.29. The compressors and air dryers,

which are not required during the starting cycle of the diesel generator, are

Watts Bar SER 9-56
designed to ASME Section VIII requirements. The diesel engine air-starting

system piping and components for all engines up to the engine interface-including auxiliary skid-mounted piping are designed and classified as Quality

Group D. This is unacceptable. The staff requires that this piping be designed

to ASME Section III Class 3 (Quality Group C) requirements and meet the recommendations of Regulatory Guide 1.26. The applicant has been informed of this

position. The staff will pursue this issue with TVA, and report the results in

a supplement to the SER.

The engine-mounted piping and components, from the engine block to the engine

interface, are considered part of the engine assembly and are seismically

qualified to Category I requirements as part of the diesel engine package.

This piping and the associated components, such as valves, fabricated headers,

fabricated special fittings, and the like, are designed, manufactured, and

inspected in accordance with the guidelines and requirements of ANSI B31.1,

ANSI N45.2, and 10 CFR 50, Appendix B. The engine-mounted air-starting piping

and associated components are intentionally over designed (subjected to low

working stresses) for the application, thereby resulting in high operational

reliability. The design of the engine-mounted air-starting piping and components to the cited design philosophy and standards is considered equivalent to

a system designed to ASME Section III Class 3 requirements with regard to

system functional operability and inservice reliability.

The scope of review of the emergency diesel engine starting system included

layout drawings, piping and instrumentation diagrams, and descriptive information in Section 9.5.6 of the FSAR for the system and auxiliary support systems

essential to its operation.

The bases for acceptance in the staff review were conformance of the design

criteria and bases and design of the diesel engine air-starting system to the

requirements of GDC 17 with respect to redundancy and physical independence,

the guidance of the cited Regulatory Guides, the additional guidance in Section II of SRP 9.5.6; the recommendations of NUREG/CR-0660 and industry codes

and standards; and the ability of the system to start the diesel generator

within a specified time period.

Based on its review, the staff concludes that the emergency diesel engine airstarting system meets, except as stated above, the requirements of GDC 2, 4, 5

and 17, and meets the recommendations of NUREG/CR-0660 and industry codes and

standards. However, the staff concluded that the system does not fully conform

with requirements of GDC 2 and 4 or the guidance of the cited Regulatory Guides

and SRP Section 9.5.6 since the staff cannot determine that the system can

perform its design safety function under all conditions, and therefore, it. is

not acceptable. The above issues have been discussed with the applicant. Upon

receipt of the requested information, the staff will report its findings in a

supplement to this SER.

9.5.7 Emergency Diesel Engine Lubricating Oil System

The design safety function of the emergency diesel engine lubricating oil

system is to provide a supply of filtered lubrication oil to the various moving

parts of the diesel engine including pistons and bearings. The system is to be

designed to meet the requirements of GDC 2, 4, 5, and 17. The meeting of the

requirements of GDC 2, 4, and 5 is discussed in Section 9.5.4.1 of this SER.

Watts Bar SER 9-57
Major components of the emergency diesel engine lubricating oil system include

engine driven lube, oil piston cooling and scavenging oil pumps, a lube oil

collection sump, strainers and filters, an ac motor-driven circulating oil

pump, a lube oil cooler, instrumentation, controls and alarms, and associated

piping and valves to connect the equipment. A crankcase pressure detector

assembly is provided for protection from crankcase overpressure. Alarms and

protective devices are provided to enable the control room operator to monitor

the diesel generator lube oil system during standby, startup, or operation.

The emergency diesel engine lubrication oil system is an integral part of the

diesel engine and thus subject to the requirements of GDC 17, with regard to

system independence and single-failure criterion. The system has three subsystems that circulate lube oil through the engine for lubrication and cooling

when the engine is operating. The engine heat is rejected from the lube oil

cooler to the jacket water which in turn uses the ERCW system/as a heat sink.

The three subsystems are: (1) the main engine lube oil system, (2) the piston

cooling lube oil system, and (3) the scavenging lube oil system. The main

engine lube oil system supplies oil to the various moving parts of the diesel

engine including all main bearings and the camshaft bearings. The piston lube

oil system supplies lube oil for piston cooling and lubrication of the piston

pin bearing surfaces. The scavenging oil system supplies the other systems

with cooled and filtered oil. There is also an auxiliary keep-warm lube oil

system for use when the diesel engine is on standby, at which time the lube oil

is heated by the jacket cooling water system. This is done to improve the

first-try starting reliability.

The auxiliary keep-warm lubrication system for the diesel engine is composed of

a continuously operating oil circulating ac motor-driven pump which prelubricates the turbocharger bearings only and circulates oil to the lube oil cooler

for preheating. The other wearing parts of the engine do not receive any

lubrication until after the engine starts, and the engine driven lube oil pumps

reach full speed. This is not acceptable and the applicant was so informed.

The applicant, in a letter dated October 9, 1981, submitted a proposed modification to the system to alleviate the problem of dry starting of the engine.

The proposed modification continuously prelubricates all wearing parts of the

engine except the rocker arm assembly and the upper portions of the engine.

This is unacceptable. The staff requires lubrication to these parts during

standby, either on a continuous or intermittent basis, or justification that

dry starting will not cause excessive wear on these parts. This has been

discussed with the applicant. The staff will pursue this issue with TVA, and

report the results in a supplement to the SER.

The standby diesel engine lubricating oil system piping and components up to

the diesel engine interface, including auxiliary skid-mounted piping, are

designed to seismic Category I requirements and to recommendations of Regulatory Guide 1.29.

The standby diesel engine lube oil system piping and components up to the

engine interface including auxiliary skid-mounted piping is designed and

classified as Quality Group D. This is unacceptable. The staff requires that

this piping be designed to ASME Section III, Class 3 (Quality Group C) requirements and meet the recommendations of Regulatory Guide 1.26. The applicant has

been informed of this position. The staff will pursue this issue with TVA, and

report the results in a supplement to the SER.

Watts Bar SER 9-58
The engine-mounted piping and components, from the engine block to the engine

interface, are considered part of the engine assembly and are seismically

qualified to Category I requirements as part of the diesel engine package.

This piping and the associated components--such as valves, fabricated headers,

fabricated special fittings, and the like--are designed, manufactured, and

inspected in accordance with the guidelines and requirements of ANSI B31.1,

ANSI N45.2, and 10 CFR 50 Appendix B. The engine-mounted lubricating oil

piping and associated components are intentionally over designed (subjected to

low working stresses) for the application, thereby resulting in high operational

reliability. The design of the engine-mounted lubricating oil piping and components to the cited design philosophy and standards is considered equivalent

to a system designed to ASME Section III Class 3 requirements with regard to

system functional operability and inservice reliability.

The diesel generator lubricating oil system conforms with Regulatory Guide 1.9

position C.7, as it relates to diesel engine lubrication system protective

interlocks. The diesel generator system protective interlocks are discussed in

Section 8.3 of this SER.

The scope of review of the diesel generator lubricating oil system included

piping and instrumentation diagrams, as well as the descriptive information in

Section 9.5.7 of the FSAR for the system and auxiliary support systems essential

to its operation.

The basis for acceptance in the staff's review was conformance of the design

criteria and bases and design of the diesel engine lubricating oil system to

the requirement of GDC 17 with respect to redundancy and physical independence,

the guidance of the cited Regulatory Guides, the additional guidance in Section II of SRP 9.5.7, and the recommendations of NUREG/CR-0660 and industry

codes and standards.

Based on its review, the staff concludes that the emergency diesel engine

lubricating oil system meets, except as stated above, the requirements of

GDC 2, 4, 5, and 17 and meets industry codes and standards. However, the staff

has concluded that the system does not fully conform with the requirements of

GDC 2 and 4, the guidance of the cited Regulatory Guides and SRP 9.5.7, or the

recommendations of NUREG/CR-0660 since the staff does not have sufficient

information to determine if the system can perform its design safety function

under all conditions, and therefore, it is not acceptable. The applicant has

been asked to provide a more detailed description of the lubricating oil

system, and a description of the diesel engine crankcase explosion protective

features. After receipt of the requestedinformation, the staff will report

its findings in a supplement to this SER.

9.5.8 Emergency Diesel Engine Combustion Air Intake and Exhaust System

The design function of the emergency diesel engine combustion air intake and

exhaust system is to supply filtered air for combustion to the engine and to

dispose of the engine exhaust to atmosphere. The system is to be designed to

meet the requirements of GDC 2, 4, 5, and 17. The meeting of the requirements

of GDC 2, 4, and 5 is discussed in Section 9.5.4.1 of this SER.

A separate source of combustion air for each diesel engine.is taken from the

diesel generator building air intakes through an air filter, intake silencer,

Watts Bar SER 9-59
turbocharger compressor, and aftercoolers. The path of the exhaust gas discharge is through the turbocharger, exhaust silencer, and exhaust ducting to

the outside of the building. This meets the requirement of GDC 17, with regard

to system independence, redundancy, and single-failure criterion.

The exhaust system is separate from the air intake system to reduce the possibility of contamination of the intake air with recirculated exhaust gases. The

location of the air intake structure and design precludes the intake of fire

extinguishing agents, other noxious gases stored on site, and other deleterious

material (such as combustion products) that would affect diesel generator

operation. In addition, the intake and exhaust structures are designed to

prevent clogging of these structures from rain, snow, ice, and freezing rain.

The standby diesel engine combustion air intake and exhaust system piping and

components up to the diesel engine interface are designed to seismic Category I

requirements and to the recommendations of Regulatory Guide 1.29.

The standby diesel engine combustion air intake and exhaust system piping and

components up to the engine interface is designed and classified as Quality

Group D. This is unacceptable. The staff requires that this piping be designed

to ASME Section III Class 3 (Quality Group C) requirements and the recommendations of Regulatory Guide 1.26. The applicant has been informed of this

position. The staff will pursue this issue with TVA, and report the results in

a supplement to the SER.

The engine-mounted piping and components, from the engine block to the engine

interface, are considered part of the engine assembly and are seismically

qualified to Category I requirements as part of the diesel engine package.

This piping and the associated components--such as valves,.fabricated headers,

fabricated special fittings--are designed, manufactured, and inspected in

accordance with the guidelines and requirements of ANSI B31.1, ANSI N45.2, and

10 CFR 50 Appendix B. The engine-mounted air intake and exhaust piping and

associated components are intentionally over designed (subjected to low working

stresses) for the application, therebyresulting in high operational reliability.

The design of the engine-mounted air intake and exhaust piping and components

to the cited design philosophy and standards is considered equivalent to a

system designed to ASME Section XIII Class 3 requirements with regard to system

functional operability and inservice reliability.

The exhaust piping and silencers for the standby diesel generators are located

inside the diesel generator building to provide protection from tornado missiles.

The standby diesel generator exhaust pipes extend approximately 2 ft above the

diesel building roof (elevation 1.773.5 ft), are approximately 29 ft apart, and

are centered on a line parallel to the north wall. This portion of the exhaust

piping is not designed to withstand tornado missiles. A 2-1/2 to 3-ft parapet

has been provided on the building roof, which partially shields the exposed

portion of the exhaust pipes from horizontal tornado missiles. Because the

diesel generator building is separated by a substantial distance from the rest

of the Watts Bar plant, the parapet does not provide total protection from tornado

missiles. Damage to the diesel exhaust pipes from tornado missiles would result

in either deformation or severing of the exhaust pipe. Severing the diesel

exhaust pipe would not affect the operation of the diesels; however, deformation could result in a decrease in performance of the corresponding diesel

generator. This is unacceptable and does not meet the requirements of GDC 2

Watts Bar SER 9-60
and 4. By letter dated June 4, 1982, the applicant committed to modify the

diesel generator exhaust stacks to ensure that a tornado missile striking them

would sever the exhaust stack completely or provide missile protection for the

exhaust stacks. As discussed in Section 3.5.2, the staff finds this commitment acceptable pending review of the details of the modification.

The scope of review of the diesel generator intake and exhaust system included

layout drawings, piping and instrumentation diagrams, and descriptive information in FSAR Section 9.5.8 for the system and auxiliary support systems essential to its operation.

The basis for the acceptance in the staff review was conformance of the design

criteria and bases and design of the diesel engine air intake and exhaust

system to GDC 17 with respect to redundancy and physical independence; the

guidance of the cited Regulatory Guides; the additional guidance in Section II

of SRP 9.5.8; the recommendations of NUREG/CR-0660 and industry codes and

standards; and the ability of the system to provide sufficient combustion air

and release of exhaust gases to enable the emergency diesel generator to

perform on demand.

Based on its review, the staff concludes that the emergency diesel engine

intake and exhaust system meets, except as noted above, the requirements of

GDC 2, 4, 5, and 17 and meets the recommendation of NUREG/CR-0660 and industry

codes and standards. However, the staff has concluded that the system does not

fully conform with the requirements of GDC 2 and 4 or the guidance of the cited

Regulatory Guides and SRP Section Plan 9.5.8 since the staff cannot determine

that the system can perform its design safety function under all conditions

and, therefore, it is not acceptable. After receipt of the requested information, the staff will report its findings in a supplement to this SER.

Watts Bar SER 9-61

10 STEAM AND POWER CONVERSION SYSTEM

10.1 Summary Description

The steam and power conversion system is designed to remove heat energy from

the primary reactor coolant loop via four steam generators and to generate

electric power in the turbine generator. After the steam passes through the

high and low pressure turbines, the main condensers deaerate the condensate

and transfer the rejected heat to the open cycle circulating water system.

This system uses a cooling tower for each unit to dissipate the rejected heat

to the atmosphere by evaporation, radiation, and convection. The condensate

is reheated and returned as feedwater to the steam generators. The entire

system is designed for the maximum expected energy from the nuclear steam

supply system.

A turbine steam bypass system is provided to discharge directly to the condenser

up to 40 percent of the main steam flow around the turbine during transient

conditions. This bypass capacity, together with a 10-percent reactor automatic

step load reduction capability, is sufficient to withstand a 50-percent generator load loss without tripping the turbine or tripping the reactor.

10.2 Turbine Generator

10.2.1 Turbine Generator Design

The turbine generator converts steam power into electrical power and has a

turbine control and overspeed protection system. The design function of the

turbine control and overspeed protection system is to control turbine action

under all normal or abnormal conditions and to ensure that a full load turbine

trip will not cause the turbine to overspeed beyond acceptable limits and to

minimize the probability of generation of turbine missiles in accordance with

the requirements of GDC 4. The turbine control and overspeed protection system

is therefore essential to the overall safe operation of the plant.

The turbine generator is manufactured by the Westinghouse Company and is a

tandem-compound type (single shaft) with one double-flow high pressure turbine

and three double-flow low pressure turbines. The rotational speed is 1800 rpm

and is designed for a gross generator output of 1218 MWe at a nominal plant

exhaust pressure of 2.0 in. mercury (absolute).

The turbine generator is equipped with an electrohydraulic control (EHC)

system. The EHC system consists of an electronic governor using solid-state

control techniques in combination with a high pressure hydraulic actuating

system. The system includes electrical control circuits for steam pressure

control, speed control, load control, and steam control valve positioning.

Overspeed protection is accomplished by three independent systems (normal speed

governor, mechanical overspeed, and electric backup overspeed control systems).

The normal speed governor modulates the turbine control valves to maintain

Watts Bar SER 10-1
desired speed load characteristics, and it will close the intercept valves and

control valves at 103 percent of rated speed. The mechanical overspeed sensor

trips the turbine stop, control, and combined intermediate valves by deenergizing the hydraulic fluid systems when 111 percent of rated speed is reached.

The turbine steam valves close in 0.15 seconds, and the extraction valves close

in less than a second after an overspeed condition is detected. These valves

are designed to fail closed on loss of hydraulic system pressures. The electrical backup overspeed sensor will trip these same valves when 111 percent of

rated speed is reached by independently deenergizing the hydraulic fluid

system. Both of these actions independently trip the energizing trip fluid

system. The overspeed trip systems can be tested while the unit is on line.

Therefore the requirements of GDC 4 are met.

To protect the turbine generator, the following signals will shut down the

turbine: (1) low bearing oil pressure, (2) low vacuum trip, (3) high thrust

bearing temperature, (4) high turbogenerator vibration, (5) low differential

water pressure across generator starter coils, (6) high stator coil outlet

water temperature, (7) low EHC fluid tank level, (8) low lube oil tank pressure, (9) low EHC fluid pressure, (10) low auto stop oil pressure, (11) turbine

overspeed at 111 percent of rated speed (electrical trip), (12) turbine overspeed at 111 percent of rated speed (mechanical trip), (13) reactor trip,

(14) safety injection signal, (15) EHC dc power failure, (16) loss of both main

feedwater.,turbines, (17) steam generator high-high level, (18) generator

differential current trip, (19) generator neutral overvoltage trip, (20) generator time overcurrent (voltage supervised) trip, (21) generator negative

sequence trip, (22) generator backup and main transformer feeder differential

trip, (23) generator reverse power trip, (24) unit station service transformer

A or B overcurrent trips, (25) main transformer sudden pressure trip, (26) main

and unit station service transformers differential trip, (27) 500-kV bus 2

station 3 breaker failure trip, (28) 500-kV bus 2 station 3 differential set 1

or set 2 trips, and (29) manual trip.

An inservice inspection program for the main steam stop and control valves and

reheat valves is provided by the applicant; it covers only the exercising and

observing of the main steam stop, control, reheat stop and intercept valves at

least once a week. This is unacceptable. The staff requires the following

inservice inspection programi (1) dismantling and inspection of all turbine

steam valves, at approximately 3-1/3-yr intervals during refueling or maintenance shutdowns coinciding with the inservice inspection'schedule, and

(2) exercising and observing of the main steam stop and control, reheat stop,

and intercept valves at least once a week. The applicant has not committed to

provide the above inservice inspection program. The staff requires the- total

program and will include this requirement in-the Technical Specifications.

The applicant will include preoperational and startup tests of the turbine

generator in accordance with Regulatory Guide 1.68, "Initial Test Programs for

Water Cooled Power Plants." The adequacy of the test program is evaluated in

Section 14 of this report.

The turbine generator overspeed protection system has been evaluated with

regard to the recommendations of BTPs ASB 3-1 and MEB 3-1. Evaluation of

protection against dynamic effects associated with the postulated pipe system

failure is covered in Section 3.6 of this report.

Watts Bar SER 10-2
The scope of review of the turbine generator included descriptive information

in FSAR Section 10.2, flow charts, and diagrams. The bases for acceptance

in the staff review were conformance of the design criteria and bases and

design of the turbine generator system to GDC 4 with respect to the prevention

of the generation of turbine missiles, the additional guidance in Section II of

Standard Review Plan 10.2 and industry codes and standards.

Based on its review, the staff concludes that the turbine generator overspeed

protection system meets the requirements of GDC 4 and the guidance of SRP Section 10.2, can perform its designed safety functions, and is, therefore,

acceptable.

10.2.2 Turbine Disc Integrity

The staff has reviewed FSAR Section 10.2.3 and concludes that the integrity of

the turbine will be adequate and that reasonable assurance is provided that

the applicable parts of the GDC will be met.

The turbine discs and rotors are forged from vacuum-degassed steel by processes

that minimize flaws and provide adequate fracture toughness. These materials

have the lowest fracture appearance transition temperatures and highest Charpy

V-notch energies obtainable on a consistent basis. The maximum tangential

stress in discs and rotors resulting from centrifugal forces, interference fit,

and thermal gradients does not exceed 0.75 of the yield strength of the materials at 115 percent of the rated speed.

The preservice inspection program calls for 100 percent ultrasonic inspection

(UT) of each rotor and disc forging before finish machining and magnetic

particle (MT) after finish machining. No MT flaw indications are permissible

in bores, holes, keyways, and other highly stressed regions.

Since 1979 the staff has known of the stress corrosion problems in low pressure

rotor discs in Westinghouse turbines. The manufacturer has conducted a program

to establish the methods and intervals for ISI of the bore and keyway areas in

these discs. Westinghouse has developed a method for predicting crack growth

rate based on all of the cracks found to date in Westinghouse turbines, past

history of similar turbine disc cracking, and results of laboratory tests.

This prediction method takes into account two main parameters, the yield

strength of the disc, and the temperature of the disc at the bore area where

the cracks of concern are occurring. The higher the yield strength of the

material and the higher the temperature, the faster the crack growth rate will

be.

In July 1981, NRC advised all licensees and applicants by letter that they were

requested to continue their current programs in cooperation with the turbine

suppliers to monitor properly the condition of the disc bore and keyway areas.

The staff considers that these programs meet the intent of current staff

guidelines.

The turbine meets staff criteria regarding the use of materials with acceptable

fracture toughness and adequate design. Preservice and inservice inspection

criteria are in accordance with current staff guidelines. The materials,

Watts Bar SER 10-3
processes, and'designs used by the applicant are therefore considered acceptable. The staff concludes that these provisions provide reasonable assurante

that the probability of disc failure with missile generation is low during

normal operation, including transients up to design overspeed.

10.3 Main Steam Supply System

The function of the main steam supply system is to supply steam from the steam

generators to the high-pressure turbine and other auxiliary equipment for power

generation. Section 10.3.1 evaluates the safety-related portion of the main

steam system including the main steam isolation valves (MSIVS). Section 10.3.2

evaluates the nonsafety-related portion of the main steam system downstream of

the MSIVs up to and including the turbine stop valves.

10.3.1 Main Steam Supply System (up to and Including the Main Steam Isolation

Valves)

The main steam supply system (MSSS) routes the steam generated in each of the

four steam generators to the high pressure, low pressure, and main feedwater

pump turbines, and to other auxiliary equipment for power generation through

four main steamlines. Each main.steamline contains one bidirectional MSIV,

five safety valves, and one air-operated atmospheric dump valve. The portions

of the main steamlines from the steam generators through the containment and up

to and including the main steam isolation valves (including the main steam

safety valves and atmospheric dump valves) are located in flood- and tornadoprotected structures (see to Sections 3.4.1 and 3.5.2 of this SER) and meet

Quality Group B and seismic Category I requirements, thus conforming to the

requirements of GDC 2 and the guidelines of Regulatory Guides 1.26 and 1.29.

Unlike the MSSS at Sequoyah, the system at Watts Bar utilizes one bidirectional

MSIV in each main steamline to prevent the backflow of steam, rather than a

check valve in conjunction with a unidirectional MSIV aligned in series. These

valves are capable of closing within 5 seconds after receipt of a closure

signal on "high-high" containment pressure signal or high steam flow coincident

with low steam generator pressure or low reactor coolant T . The valves are

designed to stop steam flow from either direction. Failuri"gf one MSIV to

close coincident with a steamline break will not result in the uncontrolled.

blowdown of more than one steam generator. In the event of a steamline break

upstream of an MSIV and a failure of an MSIV to close on one of the unaffected

steam generators, blowdown of the unaffected steam generators is prevented by

the closure of the nonseismic Category I turbine stop valves and turbine bypass

valves., which serves as an acceptable backup for this accident.

Safety valves and power-operated relief (atmospheric dump) valves are provided

for each steam generator immediately outside the containment structure, upstream

of the main steam isolation valves. The PORVs (one per steam generator) are

air operated. Normal air supply is provided by the station control and service

air system. If the normal compressed air supply fails to supply adequate air

pressure, the auxiliary compressed air system automatically provides the

required air supply to those components whose operation is required to maintain

plant safety.(see to Section 9.3.1 of this SER). This backup air supply

ensures valve operability during all plant conditions. Thus, the requirements

of GDC 34, "Residual Heat Removal," are satisfied.

Watts Bar SER 10-4
All equipment required to function to ensure main steam isolation when called

upon is protected against the effects of high-energy pipe breaks. This equipment is located in seismic Category I structures, protected from the effects

of flooding, tornado missiles, and arranged so that it is not affected by

internally generated missiles. Thus, it satisfies the requirements of GDC 2

and 4 and the guidelines of Regulatory Guide 1.117 and BTP ASB 3-1.

The applicant has provided a comparison of the system to that which was designed for Sequoyah. The staff has reviewed the information provided in the comparison

and concurs with the applicant's findings. Thus, the staff reaffirms that the

conclusions stated in the Sequoyah SER (NUREG-O011) are applicable to Watts

Bar.

The staff has further reviewed the system.for compliance with the applicable

GDC, Regulatory Guides, and BTPs. The staff concludes that the system design conforms to the requirements of GDC 2 and 4 with respect to protection against

natural phenomena and environmental effects and missiles; the requirements of

GDC 34 with respect to residual heat removal capability and single-failure

criterion; and the guidelines of Regulatory Guides 1.26, 1.29, 1.117 and BTP

ASB 3-1 regarding quality group and seismic classification, the effects of

tornado missiles, high energy pipe breaks, and the MSIV closure time position.

It is, therefore, acceptable.

10.3.2 Main Steam Supply System

This portion of the main steam system (downstream of main steam isolation

valves) is not required to effect or support safe shutdown of the reactor.

The main steam system is designed to deliver steam from the steam generators

to the high-pressure turbine. The main steam and turbine steam systems provide

steam to the feedwater pump turbines, emergency feedwater pump turbine,- turbine

gland seal system, turbine bypass system, and steam supply to the moisture

separator reheaters. The main steam system from the MSIV to the turbine stop valves and all branch lines are designed to the requirements of ANSI B31.1 and

are nonseismic.

The scope of review of the main steam supply system (between the main steam

isolation valve and up to and including the turbine stop valves) included descriptive information in Section 10.3 of the FSAR and flow charts and diagrams.

The basis for acceptance in the staff review was conformance of the design

criteria and bases and design of main steam supply system to the acceptance

criteria in Section II of SRP 10.3.

Based on its review, the staff concludes the main steam supply system between

the main steam isolation valves and up to and including the turbine stop valves

is in conformance with the above-cited criteria and design bases, can perform

its designed functions, and is, therefore, acceptable.

10.3.3 Steam and Feedwater Systems Materials

GDC 1 requires, in part, that systems important to safety be designed to

quality standards commensurate with the importance to safety of the functions

Watts Bar SER 10-5
to be performed. The materials utilized in the steam and feedwater systems

are reviewed for compliance with GDC 1.

The mechanical properties of materials selected for Class 2 components of the

steam and feedwater systems satisfy Appendix I of Section III of the ASME

Boiler and Pressure Vessel Code, and Parts A, B, or C of Section II of the

Code.

The fracture toughness properties of ferritic materials satisfy the requirements

of the Code. These fracture toughness tests and mechanical properties required

by the Code provide reasonable assurance that ferritic materials will have adequate safety margins against the possibility of nonductile behavior or rapidly

propagating fracture.

The onsite cleaning and cleanliness controls during fabrication satisfy the

positions given in Regulatory Guide 1.37 and the requirements of ANSI N45.2.1-

1973.

Low alloy steel is not utilized for main steam or feedwater systems components.

Therefore, Regulatory Guide 1.50, "Control of Preheat Temperature for Welding

of Low Alloy Steel," is not applicable.

Tubular products are nondestructively examined in accordance with the Code.

Conformance with the codes, standards, and.Regulatory Guides mentioned constitutes an acceptable basis for assuring the integrity of steam and feedwater

systems, and for meeting, in part, the requirements of GDC 1.

10.3.4 Secondary Water Chemistry

In late 1975, the staff incorporated provisions into the Standard Technical

Specifications that required limiting conditions for operation and surveillance

requirements for secondary water chemistry parameters. The Technical Specifications for all pressurized water reactor plants that have been issued an operating license since 1974 contain either these provisions or a requirement to

establish these provisions after baseline chemistry conditions have been

determined. The intent of the provisions was to provide added assurance that

the operators of newly licensed plants would properly monitor and control

secondary water chemistry to limit corrosion of steam generator components such

as. tubes and tube support plates.

In a number of instances, the Technical Specifications have significantly

restricted the operational flexibility of plants with little or no benefit

with regard to limiting degradation of steam generator tubes and the tube support plates. Based on this experience and the knowledge gained in recent

years, the staff has concluded that Technical Specification limits are not the

most effective way of assuring that steam generator degradation will be

minimized.

Because of the complexity of the corrosion phenomena involved and the state of

the art as it exists today, the staff is of the opinion that, in lieu of specifying limiting conditions in the Technical Specifications, a more effective

Watts Bar SER 10-6
approach would be to institute a license condition that required the implementation of a secondary water chemistry monitoring and control program containing

appropriate procedures and administrative controls.

The required program and procedures are to be developed by applicants with

input from their reactor vendors, or other consultants, to account for site and

plant-specific factors that affect chemistry conditions in the steam generators.

In the view of the staff, plant operation following such procedures would

provide assurance that licensees would devote proper attention to controlling

secondary water chemistry, while also providing the needed flexibility to allow

them to deal effectively with an offnormal condition that might arise.

The applicant has stated that the Watts Bar and Sequoyah Plant secondary water

chemistry monitoring and control program and implementing systems are identical.

The Sequoyah secondary water chemistry program has been previously evaluated

by the staff and found acceptable. The staff evaluation of the Sequoyah

secondary water chemistry program confirmed that the program addressed the six

program criteria of the staff position and is based on the steam generator

water chemistry program recommended by the NSSS vendor (Westinghouse).

On the basis of the above evaluation, the staff concludes that the proposed

secondary water chemistry monitoring and control program for Watts Bar meets

(1) the requirements of GDC 14 insofar as secondary water chemistry control

ensures primary boundary material integrity; (2) Acceptance Criterion 3 of SRP

Section 5.4.2.1 (NUREG-0800); (3) Positions 2 and 3 of BTP MTEB 5-3, Revision 2;

and (4) the program criteria in the staff's position. Therefore, it is acceptable. The staff will condition the operating license to require that the

proposed secondary water chemistry monitoring and control program be carried

out.

10.4 Other Features

10.4.1 Main Condenser

The main condenser is designed to function as a heat sink for the turbine

exhaust steam, turbine bypass steam, and other turbine cycle flows, and to

receive and collect condensate flows for return to the steam generators. The

main condenser transfers heat to the circulating water systems, which use

cooling towers to dissipate the rejected heat.

The main condenser is not required to effect or support safe shutdown of the

reactor or to perform in the operation of reactor safety features. The main

condenser has three shells and is designed to produce a turbine back pressure

of 1.5, 2.15, and 3.065 in. mercury absolute for the three zones when operating at rated turbine output. The main condenser design includes provisions

for condensate deaeration and hotwell surge storage of condensate for approximately 31⁄2-minutes supply at design conditions. Offgas from the main condenser'

is processed in the main condenser evacuation system, which is described and

evaluated in Section 10.4.2 of this report.

The main condenser is designed to accept full load exhaust steam from the main

turbine and reactor feedwater pump turbines, up to 40 percent of the main steam

Watts Bar SER 10-7
flow from the turbine bypass system, and other cycle steam flows. The main

condenser also is designed to deaerate the condensate to the required water

quality. Copper-nickel tubes have been used to minimize corrosion and erosion

of condenser tubes. Condenser tube leakage could result in degradation of the

feedwater quality with potential for corrosion of secondary system components. The applicant monitors the condensate by means of a sampling system to give an

indication of tube leakage. The applicant, in response to a request for additional information, provided details on the detection, control, and correction

of condenser cooling water leakage into the condensate.

The applicant will include preoperational and startup tests of the main condenser in accordance with recommendations of Regulatory Guide 1.68. The

adequacy of the test program is evaluated in Section 14 of this report.

The scope of review of the main condenser included layout drawings and descrip- tive information of the condenser in FSAR Section 10.4.1.

The basis for acceptance in the staff review was conformance of the design criteria and bases and design of the condenser to the acceptance criteria in

Section II of SRP Section 10.4.1 and industry standards.

Based on its review, the staff concludes that the main condenser is in conformance with the above cited criteria and design bases, can perform its designed function, and is, therefore, acceptable.

10.4.2 Main Condenser Evacuation System

The main condenser evacuation system, one for each unit, is designed to establish and maintain main condenser vacuum by removing noncondensable gases from

the condenser and discharging the gases through a turbine building roof ventilator. The system is designed to Quality Group D and to a nonseismic design classification. Each main condenser evacuation system consists of three

mechanical vacuum pumps, an electrical heating coil, a HEPA filter, and a

carbon adsorber. Air and noncondensables from the filtered vacuum pump exhaust

are continuously monitored by a radiation monitor before release to the

environment.

The scope of the staff review included the system's capability to process radioactive gases and the design provisions incorporated to monitor and control

releases of radioactive materials in gaseous effluents in accordance with

GDC 60 and 64. Based on its evaluation, the staff finds the main condenser

evacuation system acceptable. The basis for staff acceptance has been conformance of the applicant's design, design criteria, and design bases for the main

condenser evacuation system to the applicable regulations.

10.4.3 Turbine Gland Sealing System

The turbine gland sealing system (TGSS) is designed to control radioactive

steam leakage from, and air inleakage into, the turbine. The components of the

system are designed to Quality Group D and to a nonseismic design classification. The TGSS consists of labyrinth seals' a steam supply system, a gland steam condenser, and an exhaust fan. Steam is supplied to the labyrinth seals

from the auxiliary boiler system during startup and from the main steam system

Watts Bar SER 10-8
during load operations. The exhaust fan maintains a slight vacuum in the

system and exhausts the noncondensables to a vent on the outside of the turbine

building.

The staff has reviewed the applicant's system description and design criteria

for the components of the TGSS and found them consistent with the criteria

given in Regulatory Guide 1.26.

The basis for acceptance in the staff review has been conformance of the applicant's design, design criteria, and design bases for the turbine gland sealing

system to the applicable guidelines referenced above. Based on the evaluation,

the staff finds-the proposed TGSS acceptable.

10.4.4 Turbine Bypass System

The turbine bypass system is designed to bypass up to 40 percent of main steam

flow to the main condenser. This capacity together with a 10 percent reactor

automatic step load capability, is sufficient to withstand a 50 percent generator load loss without tripping the turbine or causing control rod movement.

The turbine bypass system is used to control coolant temperature as follows:

(1) during reactor heatup to rated pressure; (2) while the turbine generator

is being brought up to speed and synchronized; (3) during power operation when

the reactor steam generator exceeds the transient turbine steam requirements;

and (4) during reactor cooldown. This system is not required to perform during

accident conditions.

The bypass system is composed of the following: (1) 12 air-operated valves,

(2) associated instruments'and controls, and (3) piping. Each valve is rated

for a capacity of 532,170 lb/hr'of steam flow at 900 psia pressure. The

12 turbine bypass valves are mounted between the main steam isolation valves

and turbine stop valves on manifolds which are piped to the condenser. The

bypass flow is distributed to the three pressure zones of the condenser by 12

10-in, perforated pipes designed to ensure that no high velocity steam jet can

impinge on the condenser tubes. The turbine bypass system is not required for

plant control following an accident and is not a safety-related system. The

turbine bypass valves are designed to fail closed upon loss of air to the air

operated valves. The 12 turbine bypass valves are designed to close on loss

of main condenser vacuum.

The applicant will include preoperational and startup tests of the turbine

bypass system in accordance with recommendations of Regulatory Guide 1.68. The

adequacy of the test program is evaluated in Section 14.1 of this report. The

turbine bypass system can be tested while the uni.t is on line. The applicant's

proposed inservice inspection program will provide for stroking the valves

every fueling outage and a visual external inspection every refueling outage.

The staff finds this unacceptable. The staff requires, as a maximum, the

stroking of the valve on a periodic basis (at least once a quarter). This

requirement will be included in the Technical Specifications.

The turbine bypass system has been evaluated with regards to the recommendations of BTP ASB 3-1 and MEB 3-1. Evaluation of protection against dynamic

effects associated with the postulated pipe system failures is covered in

Section 3.6 of this report.

Watts Bar SER 10-9
The scope of review of the turbine bypass system included drawings, piping,

and instrumentation diagrams and descriptive information of the system in

Section 10.4.4 of the FSAR.

The basis for acceptance in the staff review was conformance of the design

criteria and bases and design of the turbine bypass system to the acceptance

criteria in Section II of SRP Section 10.4.4 and industry standards.

Based on the review, the staff concludes that the turbine bypass system is in

conformance with the above-cited criteria and design bases, can perform its

designed function, and is, therefore, acceptable.

10.4.5 Condenser Circulating Water System

The condenser cooling water (CCW) system is designed to remove the heat rejected

from the main condenser to the environment via two natural draft cooling towers.

The nonsafety-related system (Quality Group D, nonseismic Category I) is not

required to maintain the reactor in a safe shutdown condition or mitigate the

consequences of accidents. Watts Bar utilizes "closed mode" cooling, as opposed to the "closed mode," "open mode," or combination mode cooling that can

be employed at Sequoyah. Also at Watts Bar, the CCW pumping station is an

independent structure located in the yard between the cooling towers and the

turbine building, rather than being housed in the intake structure, as is the

case of Sequoyah. Thus, the system meets the guidelines of Regulatory

Guide 1.26.

The CCW system is not a shared system. All components of the system are unitized and are therefore independent from the other unit. Each natural draft

cooling tower is designed to reject the full heat load waste of a single unit

main condenser, and the associated unit's heat load from the essential raw

cooling water (ERCW) and raw cooling water (RCW) systems. Makeup for the CCW

system is provided by the ERCW and RCW system discharges.

The cooling towers, condenser cooling water pumping station, pumps and motors,

and conduits are not required to be designed to seismic Category I, tornadic

wind, or maximum flood requirements, because they are not required for the safe

shutdown of the plant. The cooling towers are located so that the tower's

structural failure as a result of a seismic event, tornado, or any other

natural phenomenon could not damage any safety-related structure, system, or

component. Thus, the system design is consistent with the requirements of

GDC 2 and the guidelines of Regulatory Guide 1.28, "Quality Assurance Program

Requirements (Design and Construction)."

The applicant has verified that in the event of a CCW pipe rupture in the turbine building assuming continuous makeup to the CCW system, flooding of the

turbine or service buildings will not endanger any safety-related equipment,

including essential electrical systems. All penetrations (doorways, mechanical

piping, and electrical penetrations) from the turbine or service buildings to

the auxiliary or control buildings are sealed up to elevation 729 ft msl (1 ft

above grade level). Piping of any system which conveys flow (makeup) to the

heat rejection system has provisions to prevent backflow of the condenser

circulating water into any area where flooding of safety-related components

would result from a failure of the system providing the flow.

Watts Bar SER 10-10
The applicant has provided a comparison of this system to that which was

designed for Sequoyah. Because of the difference in component locations

between the two facilities, the major differences have been identified as

follows:

(1) Watts Bar utilizes closed mode cooling.

(2) The Watts Bar system does not require a main condenser vacuum priming

system for establishing and maintaining flow paths full of water.

(3) The CCW system is unitized for each unit.

(4) Makeup is not from the ultimate heat sink.

(5) Watts Bar has a separate CCW pumping station.

(6) Watts Bar has provisions for chlorination into the CCW.

The staff has reviewed the information provided in the comparison and concurs

with the applicant's findings. Because the CCW system is a nonsafety-related

system, these differences do not affect the maintenance of the reactor in a

safe shutdown condition or the mitigation of consequences of the accident.

The staff, therefore, reaffirms that the conclusions stated in the Sequoyah

SER (NUREG-O011) are applicable to Watts Bar.

The staff has further reviewed the system for compliance with the applicable

GDC, Regulatory Guides, and BTPs and concludes that the CCW system conforms to

the requirements of GDC 2 and 4 with respect to protection against natural

phenomena and environmental effects due to pipe breaks and to the guidelines

of Regulatory Guides 1.26 and 1.29, Position C.2, regarding the quality group

classification and the protection of safety-related systems from failures in

nonsafety-related systems. It is, therefore, acceptable.

10.4.6 Condensate Cleanup System

The condensate cleanup system (CCS) is designed to maintain the purity of feedwater for the steam generators by filtration to remove solid corrosion products

and by ion exchange to remove dissolved impurities. The CCS operates in conjunction with the steam generator blowdown system and the feedwater chemical

addition system to control the secondary side chemistry. The CCS has six

demineralizers capable of handling full condensate flow.

The staff has reviewed the system in accordance with SRP Section 10.4.6 of

NUREG-0800 and finds:

(1) The CCS is capable of removing contaminants from the full condensate flow

during normal operation and anticipated operational occurrences to produce feedwater purity in accordance with BTP MTEB 5-3. Therefore, the

CCS meets the requirements of GDC 14 as it relates to controlling secondary water chemistry to reduce corrosion so that the integrity of the

primary coolant boundary (steam generator tubes) will be maintained.

Watts Bar SER 10-II
(2) The instrumentation and sampling equipment provided is adequate to monitor

and control process parameters in accordance with BTP MTEB 5-3.

Based on its evaluation performed in accordance with SRP Section 10.4.6

(NUREG-0800), the staff concludes that the proposed condensate cleanup system

is acceptable.

10.4.7 Condensate and Feedwater Systems

The condensate-feedwater system is designed to supply a sufficient quantity of

feedwater to the steam generator secondary side inlet during all normal operating conditions and to guarantee that feedwater will not be delivered to the

steam generators when feedwater isolation is required. The system includes

all components and equipment from the condenser outlet through the containment

isolation valves to the steam generators and to the heater drain system. The

system serves no safety function and is therefore classified as nonsafety

related (Quality Group D, nonseismic Category I), except for the portion of

the system between the check valves located outside the containment, including

the containment isolation valves and up to the steam generators. This portion

is classified as safety related and is designed to seismic Category I, Quality

Group B requirements in order to ensure feedwater system isolation in accident

situations. This portion of the system is located in seismic Category I,

flood- and tornado-protected structures. The structures provide protection

against tornado missiles. The essential equipment is separated from the

effects of internally generated missiles and is not affected by failures in

high energy piping, thus satisfying the requirements of GDC 2 and 4 and the

guidelines of Regulatory Guide 1.17. Complete isolation of the main feedwater

system is provided when it is required to mitigate the consequences of a steam

or feedwater line break. The main feedwater regulator valves and the main

feedwater isolation valves, on receipt of the feedwater isolation signal

(high-high steam generator level, ESF actuation signal, or reactor trip) will

close in 5.0 seconds and 6.5 seconds, respectively. Feedwater is isolated from

any one steam generator on indication of high level in that particular steam

generator.

The use of the standby feedwater pump is the normal means for starting up and

shutting down the plant. This pump is also automatically activated in the

event of the loss of one main feedwater pump. This is accompanied by an

automatic turbine runback to 85 percent of load if the power level is above

80 percent of full power. Should main feedwater flow continue to decrease, the

auxiliary feedwater system will automatically activate when the low-low steam

generator level is reached. The auxiliary feedwater system (see Section 10.4.9)

automatically provides flow to the steam generators for decay heat removal

upon the loss of normal feedwater supply.

The applicant has stated that the feedwater system has been designed to prevent

waterhammer induced by the piping system, and has committed to perform a test

utilizing standard plant procedures which demonstrates that unacceptable damage

will not occur (see Section 10.4.9 of this SER).

The applicant provided a comparison of the system to that which was designed

for Sequoyah. The comparison stated that the design and design philosophy of

the systems are similar. The comparison also identified differences in system

Watts Bar SER 10-12
capacity, component operating parameters, component differences, and piping

differences because the Watts Bar steam generators are different from those

used at Sequoyah. The staff has reviewed the information provided in the

comparison and concurs with the applicant's findings.

The staff has further reviewed the system for compliance with the applicable

GDC, Regulatory Guides, and BTPs and concludes that the condensate and feedwater system conforms to the requirements of GDC 2 and 4 with respect to

protection against natural phenomena and missiles, and meets the guidelines of

Regulatory Guides 1.26, 1.29, and 1.117 regarding its quality group and seismic

classification and tornado missile protection. It is, therefore, acceptable.

10.4.8 Steam Generator Blowdown System

The steam generator blowdown system (SGBS) is designed to assist in the control

of the secondary side water chemistry by removing chemical and radioactive

impurities which accumulate in the steam generator water. The staff has

reviewed this system in accordance with SRP Section 10.4.8 of NUREG-0800.

The portion of the SGBS from the steam generator up to and including the first

isolation valve outside containment isconstructed according to Quality Group B

and seismic Category I requirements, conforming to the specifications of

Regulatory Guide 1.29 and GDC 2 with respect to seismic considerations.

The components of the SGBS downstream from the outer isolation valves are considered to be nonsafety-related and are constructed according to the quality

standards of Quality Group D, as augmented by Regulatory Position C.1.1 of

Regulatory Guide 1.143 ("Design Guidance for Radioactive Waste Management

Systems, Structures, and Components Installed in Light-Water-Cooled Nuclear

Power Plants"), thereby meeting the quality standards requirements of GDC 1.

The SGBS is adequately designed to be capable of removing chemical impurities

and radioactive contamination from steam generator secondary water during

normal operation and anticipated occupational occurrences. The SGBS design

therefore meets the primary boundary material integrity requirements of GDC 14

as it related to maintaining acceptable secondary water chemistry control by

reducing the corrosion rate.

Instrumentation and automatic controls are provided to monitor and control the

operation of the blowdown system, with provision for sampling of the blowdown

in conformance with the guidelines of BTP MTEB 5-3.*

Based on the foregoing evaluation performed in accordance with SRP Section

10.4.8 of NUREG-0800, the staff concludes that the SGBS is acceptable.

10.4.9 Auxiliary Feedwater System

The auxiliary feedwater system (AFWS) supplies, in the event of a loss of the

main feedwater supply, sufficient feedwater to the steam generators to remove

primary system stored and residual core energy.

The system is designed to start automatically in the event of a loss of offsite

electrical power, a trip of both main feedwater pumps, a safety injection

Watts Bar SER 10-13
signal, or a low-low steam generator water level, any of which will result in,

may be coincident with, or may be caused by a reactor trip. It has adequate

capacity to maintain the reactor at hot standby for 2 hours and then cool the

reactor coolant system (RCS) to the temperature at which the residual heat

removal (RHR) system may be placed in operation. The system may also be manually initiated, thus conforming with the guidelines of Regulatory Guide 1.62,

"Manual Initiation of Protective Actions."

The two reactor units have separate AFW systems. Major components of each

system are two 470-gpm electric motor-driven pumps and one 940-gpm turbinedriven pump. The AFWS is designed to Quality Group B, seismic Category I

requirements for the portion between the main feedwater line connection up to

and including the first check valve downstream of the flow control/isolation

valve. The remainder of the system is designed to Quality Group C, seismic

Category I requirements with the exception of the portion between the condensate

storage tank up to but not including the check valves on the suction side of

the AFW pumps. This portion of the system which provides the condensate supply

is designed to Quality Group D, nonseismic Category I requirements. Thus, the

system design meets the guidelines of Regulatory Guides 1.26 and 1.29.

Each AFWS consists of two trains, each train utilizing one electric motordriven pump to supply two steam generators. One pump is capable of providing

sufficient flow to its steam generators to prevent the release of reactor

coolant via the pressurizer safety valves, thus satisfying the requirements of

GDC 44. The turbine-driven pump provides AFW flow to all four steam generators

via redundant trains.

The preferred water sources for all AFW pumps are the two nonseismic, Quality

Group D condensate storage tanks (CST). An unlimited backup water source, the

essential raw cooling water (ERCW) system can provide the AFW pumps with the

required supply in the event the CST is unavailable. The ERCW supply is automatically (or remote-manual) initiated when low pressure is sensed in any one

of the three AFW pump suction lines.

The applicant has provided verification (through analysis) that the AFW pumps

can survive the transition to the backup water source in the event the preferred

source is unavailable. The staff will require that the vendor of the pumps,ý

concur with the results of the analysis that verify pump survivability or that

the applicant perform a suitable test (such as preoperational) which demonstrates that the pumps can survive the transfer.

The electric motor-driven pumps and all associated controls, valves, and other

supporting systems can be powered from onsite ac sources. The turbine-driven

pump and all associated valves, controls, and other supporting systems are

powered by steam, compressed air, and, if necessary, dc power. The turbine-

.driven pump is designed to be independent from all ac power for 2 hours.

The AFW system is housed within seismic Category I, tornado-missile-protected

structures. The system is not protected against flooding. During flood operations, the fire protection system will directly supply the steam generator with

water. Thus, the system conforms to the requirements of GDC 2 and the guidelines of Regulatory Guide 1.117. Protection from internal flooding due to pipe

rupture is discussed in Section 9.3.3 of this SER.

Watts Bar SER 10-14
The system design has provisions which permit functional testing and inservice

inspection of the system and components during plant operating conditions, thus

satisfying the requirements of GDC 45 and 46.

The applicant has committed to perform an AFWS test'which demonstrates that

unacceptable damage because of waterhammer will not occur when the AFWS is

activated.

The applicant has provided a comparison of the system to that which was designed

for Sequoyah. The staff has reviewed the information provided in the comparison

and concur with his findings.

Auxiliary Feedwater System Evaluation (II.E.1.1)

The staff has reviewed the Watts Bar AFWS against the requirements of the

March 10, 1980 letter, which corresponds to Item.II.E.1.1 of NUREG-0660 and

NUREG-0737. These requirements are

(1) Provide a deterministic evaluation of the AFWS as compared to the requirements in SRP Section 10.4.9 and BTP ASB 10-1.

(2) Provide a reliability analysis of the AFWS similar in method to that

described in Enclosure 1.

(3) Incorporate the short-term and long-term recommendations into the design

of the AFWS.

(4) Provide a response to Enclosure 2 regarding the design basis for the AFWS

flow requirements.

The staff has reviewed the applicant's response to the requirements of the

staff's March 10, 1980 letter in conjunction with the Sequoyah/Watts Bar -

comparison and concludes that the design of the AFW system is in conformance

with the guidelines of SRP Section 10.4.9 BTP ASB 10-1. The staff further

concludes that sufficient AFW flow (470 gpm) can be delivered to two steam

generators by the AFW system to meet the minimum heat removal requirements

following any design-basis transient or accident assuming the worst-case single

active failure.

The staff has reviewed the applicant's response regarding the AFW reliability

analysis of Watts Bar in conjunction with the Sequoyah AFWS reliability analysis.

The staff concurs with the applicant that the differences between the Watts Bar

AFWS and the Sequoyah AFWS design do not necessitate the modification of the

Sequoyah input reliability data for use in the Watts Bar analysis. Thus, the

staff concludes that the Watts Bar AFWS reliability analysis is acceptable.

The staff has reviewed the applicant's response with regard to the implementation of the short- and long-termýrecommendations and finds them acceptable.

The applicant has committed to incorporate all short- and long-term recommendations of the March 10, 1980 letter before receiving an operating license.

The staffhas further reviewed the system for compliance with the applicable

GDC, Regulatory Guides, and BTPs and concludes that the AFWS meets the requireWatts Bar SER 10-15
ments of GDC 2, 44, 45, and 46 and BTP-ASB 10-1 with respect to protection against natural phenomena, the ability to transfer heat under accident conditions, decay heat removal capability, and to provide isolation to ensure system function, periodic inspection and testing, and meets the guidelines of Regula- tory Guides 1.26, 1.29, and 1.117 regarding quality group and seismic classification and tornado-missile protection, and is, therefore, acceptable.

Sequential Auxiliary Feedwater Flow Analysis (II.K.2.19)

NUREG-0737 requires the applicant to provide a benchmark analysis of sequential auxiliary feedwater flow to the steam generators following a loss of main feedwater.

The staff has completed a generic review on this subject and has determined

that no further action is necessary for the Watts Bar review. These concerns

are not applicable to nuclear steam supply systems with inverted U-tube steam

generators such as the Westinghouse-designed steam generator of the Watts Bar

plant.

Watts Bar SER 10-16
U

11 RADIOACTIVE WASTE MANAGEMENT

11.1 Summary Description

The radioactive waste management systems are designed to provide for the controlled handling and treatment of liquid, gaseous, and solid wastes. The

liquid radioactive waste system processes wastes from equipment and floor

drains, sample wastes, decontamination and laboratory wastes, regenerant

chemical wastes, and laundryand shower wastes. The gaseous radioactive waste

system provides holdup capacity to allow decay of short-lived noble gases

stripped from the primary coolant and treatment of ventilation exhausts through

HEPA and charcoal adsorbers as necessary to reduce releases of radioactive

materials to as low as is reasonably achievable (ALARA) levels in accordance

with 10 CFR 20 and 50.34a. The solid radioactive waste system provides for the

solidification, packaging, and storage of radioactive wastes generated during

station operation before shipment offsite to a licensed facility for burial.

In its evaluation of the liquid and gaseous radioactive waste systems, the

staff has considered: (1) the capability of the systems for keeping the levels

of radioactivity in effluents ALARA based on expected radwaste inputs over the

life of the plant, (2) the capability of the systems to maintain releases below

the limits in 10 CFR 20 during periods of fission product leakage at design

levels from the fuel, (3) the capability of the systems to meet the processing

demands of the station during anticipated operational occurrences, (4) the

quality group and seismic design classification applied to the equipment and

components and structures housing these systems, (5) the design features that.

will be incorporated to control the releases of radioactive materials in

accordance with GDC 60, and (6) the potential for gaseous release as a result

of hydrogen explosions in the gaseous radwaste system.

In its evaluation of the solid radioactive waste treatment system, the staff

has considered: (1) system design objectives in terms of expected types,

volumes, and activities-of waste processed for offsite shipment, (2) waste

packaging and conformance to applicable Federal packaging regulations and

provisions for controlling potentially radioactive.airborne dusts during baling

operation, and (3) provisions for onsite storage before shipping.

In the evaluation of the process and effluent radiological monitoring and sampling systems, the staff has considered the system's capability: (1) to monitor all normal and potential pathways for release of radioactive materials to

the environment, (2) to control the release of radioactive materials to the

environment, and (3) to monitor performance of process equipment and detect

radioactive material leakage between systems.

The staff has also evaluated the quantities of radioactive materials that will

be .released in liquid and gaseous effluents and the quantity of radioactive

waste that will be shipped offsite to a licensed burial facility. In making

these determinations, the staff has considered waste flows, activity levels,

and equipment performance, consistent with expected normal plant operation,

Watts Bar SER 11-1
including anticipated operational occurrences, over the projected 30-yr

operating life of the plant.

The estimated releases of radioactive materials in liquid and gaseous effluents

were calculated using the PWR GALE Code described in NUREG-0017, "Calculation

of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors (PWR GALE Code)," April 1976. The principal parameters

used in these calculations are given in Table 3.2 of the FES for Watts Bar

(NUREG-0498, December 1978). The liquid and gaseous source terms are given in

Tables 3.3 and 3.4, respectively, of the FES.

Based on the evaluation in the FES, the staff concluded that the liquid and

gaseous radioactive waste treatment systems for Watts Bar are capable of maintaining releases of radioactive materials in liquid and gaseous effluents ALARA

levels in accordance with 10 CFR 50.34a, and with Sections II.A, II.B, II.C,

and II.D of Appendix I to 10 CFR 50. Since that time, the staff conclusion has

not changed.

The staff has. reviewed the applicant's design, design-criteria, and design

bases for the radioactive waste management systems at Watts Bar. The acceptance criteria used as a basis for the staff evaluation are in SRP Sections 11.1,

11.2, 11.3, 11.4, and 11.5 of NUREG-0800 (formerly NUREG-75/087). These acceptance criteria include applicable GDC, 10 CFR 20.106, and Appendix I to 10 CFR 50.

Guidelines for implementation of the requirements of the acceptance criteria

are provided in the ANSI standards, Regulatory Guides, and other documents identified in NUREG-0800. There has been no major change in the acceptance criteria

between NUREG-75/087 and NUREG-0800. Conformance to the acceptance criteria

provides a basis for concluding that the radioactive waste management systems

meet the requirements of 10 CFR 20 and 50.

Based on its evaluation, as described below, the staff finds the design of the

liquid, gaseous, and solid radioactive waste systems and associated process and

effluent radiological monitoring and sampling systems to be acceptable.

11.2 Liquid Waste Management

The liquid waste processing system (LWPS) for Watts Bar is shared between Units 1

and 2. The LWPS consists of process equipment and instrumentation necessary to

collect, process, monitor, and recycle or dispose of radioactive liquid wastes.

The liquid radwaste system is designed to collect and process wastes based on

the origin of the waste in the plant and the expected levels of radioactivity.

All liquid waste is processed on a batch basis to permit optimum control of

releases., Before being released, samples are analyzed to determine the types

and amounts of radioactivity present. Based on the results of the analyses,

the waste is recycled for eventual reuse in the plant, retained for further

processing, or released under controlled conditions to the environment.

A radiation monitor in the discharge line automatically terminates liquid waste

discharges if radiation measurements exceed a predetermined level. The liquid

radioactive waste processing system consists of the tritiated and nontritiated

waste subsystems and a condensate regenerant waste subsystem. In addition, the

Watts Bar SER 11"2
chemical and volume control system (CVCS) processes letdown from the primary

system to control boron concentration and reactor water purity.

In its evaluation model, the staff assumed that a portion of the CVCS flow will

be released through the LWPS for tritium control. A deep-bed regenerable

demineralizer system is provided for treatment of turbine condensate. Steam

generator blowdown will be cooled and normally sent directly to the condensate

cleanup system for processing and reuse in the plant. Laundry, hot shower, and

decontamination wastes are normally released without treatment; the floor drain

(dirty waste) subsystem is used to treat effluents from these sources when

radioactivity concentrations are in excess of pre-established limits.

The LWPS consists of the boron recycle system, the tritiated waste system,. the

floor drain (dirty waste) system that handles nontritiated waste and condensate

demineralizer regeneration solutions, and the laundry and shower drain system.

These systems are described in Section 3.2.3.1 of the FES, which considered

contributions of liquid waste from the turbine building drains and the steam

generator blowdown system.

The LWPS for Watts Bar is essentially identical with that evaluated in the SER

for Sequoyah, NUREG-O011, March 1979.

Conformance with Federal Regulations and Branch Technical Positions

The LWPS is located in the auxiliary building which is designed to seismic

Category I criteria. The seismic design and quality group classification and

capacities of principal components considered in the liquid waste processing

system evaluation are listed in Table 11.1. The staff finds the applicant's

liquid-radioactive waste treatment system.design to be acceptable in accordance

with the guidelines of Regulatory Guide 1.143. The system design includes

measures intended to control the release of radioactive materials as a result

of potential overflows from indoor and outdoor storage tanks. Tank levels are

monitored either locally or in the control room and high level alarms are

activated should preset levels be exceeded. Overflow provisions such as sumps,

dikes, and overflow lines permit the collection and subsequent processing of

tank overflow. The staff considers these provisions to be capable of controlling the release of radioactive materials to the environment.

The staff concludes that the LWPS is capable of reducing the releases of radioactive materials in effluents ALARA in accordance with 10 CFR 50.34a and,

therefore, is acceptable.

The staff has determined that the LWPS is capable of reducing the release of

radioactive materials in liquid effluents to concentrations below the limits in

10 CFR 20 during periods of fission product leakage from the fuel at design

levels.

11.3 Gaseous Waste Management

The gaseous radioactive waste processing and plant ventilation systems are

designed to collect, store, process, monitor, recycle, and/or discharge potentially radioactive gaseous wastes which are generated during normal operation

Watts Bar SER 11-3
Table 11.1 Design parameters of principal components considered in the

evaluation of liquid and gaseous radioactive waste treatment

systems

Component Number Capacity each

LIQUID SYSTEMS*

Tritiated Waste Processing System

Tritiated Drain Collector..Tank

Waste Evaporator Package

Condensate Demineralizer

Nontritiated Waste Processing System

1

1

1

24,700

2

100

gal

gpm

gpm

Floor Drain Collector Tank

Auxiliary Waste Evaporator Package

Waste Condensate Tanks

Laundry, Hot Shower, Chemical Waste and

Decontamination Waste Processing System

Laundry and Hot Shower Collector Tanks

Cask Decontamination Tank

Chemical Drain Tank

Condensate Demineralizer Regenerant Waste

Processing System

Nonreclaimable Waste Collection Tank

Condensate Demineralizer Waste Evaporator

Package

GASEOUS SYSTEMS*

Gaseous Waste Processing System

1

1

2

1

22,800

15

1,500

2,000

gal

gpm

gal

gal

2

1

1

600

15,000

600

gal

gal

gal

1 11,000 gal

1 30 gpm

Compressors

Decay Tanks

2

9

40 scfm

600 ft 3 .

*Quality Group and seismic design in accordance with Regulatory Guide 1.143.

of the plant. The systems consist of equipment and instrumentation necessary

to reduce releases of radioactive gases and particulates to the environment.

The principal sources, of gaseous waste are the effluents from the gaseous waste

processing system, condenser vacuum pumps, and ventilation exhausts from the

auxiliary building, reactor containment, and turbine building.

Watts Bar SER 11-4
The gaseous waste processing system (GWPS) for Watts Bar is shared between

Units 1 and 2. The GWPS collects and stores the hydrogenated fission product

gases stripped from the primary coolant letdown, the volume control tanks, and

the reactor drain tanks, by compressing through a recombiner into one of nine

gas decay tanks. Releases from the gas decay tanks are processed through HEPA

filters and charcoal adsorbers installed in the reactor building vent before

release to the environment. Ventilation exhaust air from the auxiliary building and turbine building is released to the environment without treatment.

Condenser vacuum pump exhaust is processed through HEPA filters and charcoal

adsorbers before release to the environment.

The GWPS and the plant ventilation systems are described in Section 3.2.3.2 of

the FES. The systems for Watts Bar are essentially identical with those evaluated in the Sequoyah SER.

Conformance with Federal Regulations and Branch Technical Positions

The seismic design and quality group classification and capacities of the prin- cipal equipment in the GWPS are listed in Table 11.1. The staff finds the

applicant's design criteria for the gaseous waste processing system and structure housing this system to be in conformance with the guidelines given in

Regulatory Guide 1.143 and, therefore, acceptable.-

The GWPS provides for monitoring hydrogen and oxygen upstream of the waste gas

compressor and the contents of the gas storage tanks. If the oxygen content

exceeds a predetermined level, an alarm will sound in the reactor control room,

alerting the operator to the condition. Flow will then be diverted from the

gas decay tank being filled to a standby tank and a nitrogen dilutent introduced into the system to reduce the potential for a hydrogen explosion. The

hydrogen and oxygen monitoring system does not meet the staff's current acceptance criteria because redundant monitors are not provided and because the

system is not designed to automatically intiate action to mitigate the potential for explosion in the event of a high oxygen content; therefore, the staff

will provide a Technical Specification which will require sampling and analysis

every 4 hours during gas monitor outages and will require that the reactor be

shut down if the gas monitor outage exceeds 7 days. With the inclusion of the

Technical Specification described above, the staff finds the system acceptable.

The staff concludes that the GWPS and the plant ventilation systems are capable of reducing radioactive materials in effluents ALARA in accordance with

10 CFR 50.34a and, therefore, are acceptable.

The staff has determined that the GWPS is capable of reducing the release of

radioactive materials in gaseous effluents to concentrations below the limits

in 10 CFR 20, during periods of fission product leakage from the fuel at design

levels.

11.4 Solid Waste Management System

The solid waste system (SWS) for Watts Bar is shared between Units 1 and 2, and

is designed to process two general types of solid wastes: wet solid wastes

which requires solidification before shipment and dry solid wastes which

Watts Bar SER 11-5
requires packaging and, in some cases, compaction before shipment to a licensed

burial facility.

Dry solid wastes--consisting mainly of ventilation air filters, contaminated

clothing, paper, laboratory glassware, and tools--are compacted in 55-gal drums

by a waste baler. The baler is equipped with a shroud to prevent the escape of

radioactive materials during the compaction process. During the baling operation, the air flow in the vicinity of the baler is exhausted by a fan through a

HEPA filter to the auxiliary building ventilation system to reduce the potential

for airborne radioactive dusts. The staff estimates the dry solid waste total

to be 4100 ft/yr per reactor with a total activity content of less than 5 Ci.

The staff finds the design for the dry SWS acceptable.

Wet solid wastes--consisting of waste evaporator bottoms, chemical drain tank

effluents, and spent resin--will be treated by a solidification system at Watts

Bar that differs from that described in the Sequoyah SER. The applicant is

evaluating the use of a mobile solidification and dewatering system. Before

operation, the applicant must submit the methodology for dewatered-packaging

of wet radioactive waste with a process control program to ensure complete

solidification of all wet solid waste in conformance with the guidelines of

Section 11.4 of NUREG-0800. By letter dated March 29, 1982, TVA committed to

provide its process control program for the Watts Bar facility, which will be

reviewed and included in the technical specifications. The staff finds this

commitment acceptable.

11.5 Process and Effluent Radiological Monitoring and Sampling Systems

The process and effluent radiological monitoring systems are designed to provide

information concerning radioactivity levels in systems throughout the plant,

indicate radioactive leakage between systems, monitor equipment performance,

and monitor and control radioactivity levels in plant discharges to the

environs.

Table 11.2 provides the locations of continuous monitors. Monitors on certain

effluent release lines automatically terminate discharges should radiation

levels exceed a predetermined value. Systems which are not amenable to-continuous monitoring or for which detailed isotopic analyses are required are periodically sampled and analyzed in the plant laboratory.

The staff has reviewed the locations and types of effluent and process monitoring provided. Based on the plant design and on continuous monitoring locations and intermittent sampling locations, the staff has concluded that all

normal and potential release pathways are monitored. The staff has also determined that the sampling and monitoring provisions are adequate for detecting

radioactive material leakage to normally uncontaminated systems and for monitoring plant processes which could affect radioactivity releases. On this

basis the staff considers the monitoring and sampling provisions to meet the

requirements of GDC 60, 63, and 64 and the guidelines of Regulatory Guide 1.21

("Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Release

of Radioactivity in Liquid and Gaseous Effluents from Light Water-Cooled Nuclear

Power Plants"), and, therefore, are acceptable.

Watts Bar SER 11-6
The Technical Specifications for the process effluent radiological monitoring

systems, instrumentation, controls, and the sampling and analysis programs for

Watts Bar will be similar to those at Sequoyah.

11.6 Evaluation Findings

In its evaluation, the staff has calculated releases of radioactive materials

in liquid and gaseous effluents for normal operation including anticipated

operational occurrences based on expected radwaste inputs over the life of the

plant.

In the evaluation, the staff has determined that the applicant's design of the

liquid and gaseous waste treatment systems satisfies the design objectives of

Appendix I to 10 CFR 50.

The staff concludes that the liquid and gaseous radwaste treatment systems will

reduce radioactive materials in effluents ALARA in accordance with

10 CFR 50.34a and, therefore, are acceptable.

The staff has considered the potential consequences resulting from reactor

operation with a 1-percent operating power fission product source term and

determined that, under these conditions, the concentrations of radioactive

materials in liquid and gaseous effluents in unrestricted areas will be a small

fraction of the limits specified in 10 CFR 20.

The staff has considered the capabilities of the radwaste systems to meet the

anticipated demands of the plant as a result of anticipated operational occurrences and has concluded that the liquid, gaseous, and solid waste system capacities and design flexibilities are adequate to meet the anticipated needs of

the plant.

The staff has reviewed the applicant's quality assurance provisions for the

radwaste systems, the quality group classifications used for system components,

the seismic design applied to the gaseous waste processing system, and the

seismic classification applied to the design of structures housing the radwaste

systems. The design of the radwaste systems and structures housing these

systems meet the guidelines set forth in Regulatory Guide 1.143.

The staff has reviewed the provisions incorporated in the applicant's design to

control the releases of radioactive materials in liquids as a result of

inadvertent tank overflows and concludes that the measures proposed by the

applicant are consistent with our acceptance criteria as set forth in

Regulatory Guide 1.143.

The staff review of the radiological process and effluent monitoring system

included the provisions for sampling and monitoring all normal and potential

effluent discharge paths in conformance with GDC 64, for providing automatic

termination of effluent releases and assuring control over releases of radioactive materials in effluents in conformance with GDC 60 and Regulatory

Guide 1.21, for sampling and monitoring plant waste process streams for process

control in conformance with GDC 63, for conducting sampling and analytical

programs in conformance with the guidelines in Regulatory Guide 1.21, and for

Watts Bar SER 11-7
Table 11.2 Process and effluent radiation monitoring system

Monitor

Monitor sensitivity Stream monitored* Number classification PC/ml

LIQUID

Component Cooling Water 3/plant Gamma-Scintillation 9.6 x 10-7 (Co-60) Service Water 4/plant Gamma-Scintillation 9.6 x 10-7 (Co-60) Steam Generator Blowdown 4/plant Gamma-Scintillation 9.6 x 10-7 (Co-60)

(Process)

Waste Disposal System** l/plant Gamma-Scintillation 9.6 x 10-7 (Co-60) Steam-Generator Blowdown 2/plant Gamma-Scintillation 9.6 x 10-7 (Co-60) Liquid Discharge**

GASEOUS

Auxiliary Building Exhaust

Vent**

Particulate l/plant Beta-Scintillation 4.7 x 10-11 (Co-60) 1-131 1/plant Gamma-Scintillation 1.9 x 10-9 (0-131) Noble Gas l/plant Beta-Scintillation 4 x 10-7 (Kr-85) Gaseous Waste Process

System. Discharge** l/plant Beta-Scintillation ý5.5 x lo-4 (Xe-133) Condenser Vacuum Pump, Low 2/plant Beta-Scintillation 2.3 x 10-7 (Xe-133)

Range Exhaust

Condenser Vacuum Pump, High 2/plant Beta-Scintillation 1.5 x 10-5 (Xe-133)

Range Exhaust

Containment Purge Exhaust*" 4/plant Beta-Scintillation 1L1 x 10-s (Xe-133) Shield Building Vent Exhaust 2/plant Beta-Scintillation 4.7 x 10-11 (Co-60)

*All monitors have instrument malfunction and high radiation visual and'audible

alarms in the main control room. "*These monitors terminate the release when the radiation level exceeds a predetermined level.

monitoring process and effluent streams during postulated accidents. The

review included piping and instrument diagrams and process flow diagrams for

the liquid and gaseous radwaste systems and ventilation systems, and the

location of monitoring points relative to effluent release points. The staff

concludes that the applicant's radiological process and effluent monitoring

systems are acceptable.

Based on the foregoing evaluation, the staff concludes that.the radwaste treatment and monitoring systems are acceptable. The basis for acceptance has been

conformance of the applicant's designs, design criteria, and design bases for

the radwaste treatment and monitoring systems to the applicable regulations and guides referenced above, as well as to staff technical positions and industry standards.

Watts Bar SER 11-8
The applicant has committed to submit the methodology for dewatering wet

radioactive wastes and a process control program for the solidification of

radwastes. The staff finds this commitment acceptable and will ensure that the

approved process control program is included in the Technical Specifications.

11.7 NUREG-0737 Items

11.7.1 Wide Range Noble Gas, Iodine, and Particulate Effluent Monitors

(II.F.l(1) and II.F.l(2))

NUREG-0737 requires that noble gas monitoring instrumentation and equipment for

iodine and particulate sampling, for the detection and measurement of radionuclides in plant effluents be installed by January 1, 1982, or 4 months before

an operating license is issued. The'applicant has proposed that the installation date be modified so that the required equipment would be in place before

fuel loading and has provided interim procedures based on plans developed for

Sequoyah. The interim measures are not applicable for the Watts Bar nuclear

plant, as indicated in Enclosure 2 of NUREG-0737. Therefore, the staff

requires the applicant to provide the information to meet the guidelines of

Regulatory Guide 1.97 (Revision 2) and comply with NUREG-0737 4 months before

fuel loading, so that the staff can complete its review which will be reported

in a supplement to the SER.

11.7.2 Primary Coolant Outside Containment (III.D.I.1)

NUREG-0737 requires that a-program to reduce leakage from systems outside containment be implemented before issuance of a full-power license. The applicant

submitted'his response to this item on September 14, 1981 and June 11, 1982.

The submittal included a general description of the program and committed to

provide the initial results of the leak reduction program by fuel loading of

Unit 1. TVA has agreed to submit the complete program in the near future.

When this information is received, the staff will review it and report its

findings in a supplement to the SER.

Watts Bar SER II-_9

12 RADIATION PROTECTION

12.1 General

The staff has evaluated the proposed radiation protection program presented in

Chapter 12 of the Watts Bar FSAR in accordance with the criteria set forth in

the SRP, NUREG-75/087 (Revision 1), Chapter 12, and NUREG-0737, "Clarification

of TMI Action Plan Requirements," Items II.B.2, II.F.1(3), III.D.3.3, I.B.1.2,

and II.B.3. The radiation protection measures incorporated at Watts Bar are

intended to "ensure that internal and external radiation exposures to station

personnel, contractor, and the general population due to station conditions,

including anticipated operational occurrences, will be within applicable

limits, and furthermore, will be as low as is reasonably achievable."

The basis for staff acceptance of the Watts Bar Radiation Protection Program is

that doses to personnel will be maintained within the limits of 10 CFR 20 and

that the applicant's radiation protection designs and program features are also

consistent with the guidelines of Regulatory Guide 8.8, "Information Relevant

to Ensuring that Occupational Radiation Exposures at Nuclear Power Stations

Will Be As Low As Is Reasonably Achievable," Revision 3. Shielding is provided

to reduce levels of radiation. Ventilation is arranged to control the flow of

potentially contaminated air. Radiation monitoring systems, are employed to

measure levels of radiation in potentially occupied areas and to measure airborne radioactivity throughout the plant. A health physics program is proVided for plant personnel and visitors during reactor operation, maintenance,

refueling, radwaste handling and inservice inspection.

These and other radiation protection features will help ensure that occupational

radiation exposures are maintained as low as is reasonably achievable (ALARA)

during plant operation and during decommissioning.

On the basis of its review of the Watts Bar FSAR, the staff concludes that the

radiation protection measures incorporated in the design will provide reasonable assurance that occupational doses will be maintained ALARA and below the

limits of 10 CFR 20. These radiation design features are consistent with the

guidelines of Regulatory Guide,8.8.

12.2 Ensuring that Occupational Radiation Exposures Are As Low As Reasonably

Achievable

The applicant provides a management commitment that Watts Bar will be designed,

constructed, and operated in a manner consistent with the-guidelines of Regulatory Guides 8.8; 8.10, "Operating Philosophy for Maintaining Occupational

Exposures As Low As is Practicable"; and 1.8, "Personnel Selection and Training,"!

(Revision 1) and 10 CFR 20. The ALARA philosophy was applied during the

initial design of the plant. During the ensuing design and construction

phases, the applicant has continued to review, update, and modify the plant

design as a result of ALARA considerations.

Watts Bar SER 12-1
The plant health physicist has the responsibility to ensure that radiation

exposures are maintained ALARA. The health physicist is responsible for

developing the radiation protection training program, the radiation surveillance

program, and health physics procedures to ensure that exposures of all personnel

are kept within the limits of 10 CFR 20 and ALARA. The health physicist is

supported by and under the supervision of the Chief of the Radiological Hygiene

Branch (an offsite organization). These procedures and programs are developed

incorporating Regulatory Guides 8.2 ("Guide for Administrative Practice in

Radiation Monitoring"), 1.8, 8.8, and 8.10 and 10 CFR 20. Therefore, the

policy considerations are acceptable.

To reduce radiation exposures the applicant has incorporated general considerations into the design to reduce: (1) the need to enter radiation fields,

(2) the time of exposure when entry is necessary, and (3) the dose rate during

exposure. These general considerations are implemented by specific radiation

protection design guidelines. Also, information gained from the applicant's

study of operating experience from power reactors is factored into the design.

Finally, design reviews are performed by radiation protection personnel to

ensure that occupational radiation exposures will be ALARA. These design

considerations are consistent with the guidance of Regulatory Guide 8.8 and,

therefore,- are acceptable.

The applicant has incorporated the following facility and equipment design

features at Watts Bar to satisfy the design objectives and to maintain radiation

exposure ALARA:

(1) Instruments and components that require frequent maintenance or calibration are located in the lowest practical radiation fields.

(2) Vents and drains on piping are located to minimize potential crud traps.

(3) The ventilation system is designed to ensure control of airborne contaminants and for easy access and service to keep doses ALARA during alterations,

maintenance, decontamination, and filter changes.

(4) To minimize the spread of contamination and to facilitate decontamination

in the event spillage occurs, floor drains are provided in equipment

cubicles to prevent the spread of radioactive liquids; tight fitting doors

with seals are utilized in access ways to cubicles housing radioactive gas

handling equipment or equipment that processes high temperature fluids.

If the doors are louvered, they are equipped with back draft dampers.

Walls and floors in these areas are coated with special materials that are

easily cleaned in case excessive leakage of contaminants has occurred.

(5) Area and airborne radioactivity monitoring equipment with local readout is

included in areas where personnel normally have access. Portable instrumentation supplements the stationary equipment as appropriate.

(6) Portable shielding is available for situations in which permanent shielding

is needed but not practical.

(7) Remote handling equipment is used to package and .transfer liquid and solid

waste products and to transfer new fuel and spent fuel assemblies.

Watts Bar SER 12-2
The applicant has committed to include in the operational procedures measures

for reducing exposure and the criteria for implementation of those measures

consistent with the guidance of Regulatory Guide 8.8. The staff concludes that

the operating and maintenance personnel at Watts Bar will follow specific plans and procedures to ensure that ALARA goals are achieved in the operation of the

nuclear station and, therefore, the applicant's ALARA program is acceptable.

12.3 Radiation Sources

Section 12.2 of the Watts Bar FSAR describes the sources of contained and airborne radioactivity used as inputs for the dose assessment and for the design

of the shielding and ventilation systems. The methods and bases used by the

applicant to estimate the source terms are also described.

The location and strengths of the contained radiation sources that must be

shielded or included in the dose assessment are provided. The bases for the

source terms meet the staff's acceptance criteria as described below. The

fission product source terms are based on a failed fuel fraction of 0.01. The

coolant and corrosion activation product source terms are based on measurements

at operating pressurized water reactors, and they are consistent with American

National Institute Standard N237-1976. Neutron and prompt gamma source terms

are based on reactor core physics calculations and operating reactor experience.

The contained radiation source terms presented are comparable to estimates by

other applicants with pressurized water reactor designs and are acceptable.

The reactor core is the primary source of radiation in the containment, emitting

neutrons and gamma rays. The reactor coolant system is the next highest source

of radiation in the containment. The reactor coolant contains fission products from fuel clad defects and activation and corrosion products. Of these radiation sources, nitrogen-16 is the predominant activity in the reactor coolant

pumps, steam generators, and the reactor coolant piping. In buildings other

than the containment, the primary sources of personnel exposure are fission

products, activation and corrosion products, and spent fuel assemblies (in the

fuel building). The shielding used to protect personnel from these sources is

based on fission source terms for full-power operation with 1-percent fuel

cladding defects. Other parameters used, as well as a complete description of

source term development, are in Chapter 11. The source terms presented are

comparable to estimates by other applicants with similar designs and are

acceptable.

The sources of airborne radioactivity and concentrations of airborne radioactivity in various parts of the station are provided, the methods used to

calculate those airborne concentration estimates are consistent with commonly

accepted methods and, therefore, are acceptable. The airborne radioactivity

source terms presented are comparable to estimates by other applicants with

similar designs and are acceptable. The staff concludes that the radiation

source information for Watts Bar is consistent with SRP Section 12.2 criteria.

12.4 Radiation Protection Design Features

FSAR Section 12.3 describes the radiation protection design features of the

plant that are intended to maintain occupational exposures ALARA. Separate

descriptions are presented for the categories of facility design features,

Watts Bar SER 12-3
shielding, ventilation, and area radiation and airborne monitoring instrumentation.

The applicant has provided evidence that the dose-causing functions performed

by workers have been considered in the plant design. Features have been

included in the design to help maintain exposure ALARA in the performance of

those functions. These features will facilitate access to work areas, reduce

or allow the reduction of source intensity, reduce the time required in radiation

fields, and provide for portable shielding and remote handling tools. The

applicant's facility design features are consistent with the guidance of

Regulatory Guide 8.8. Therefore, the staff concludes that the facility design

features are acceptable. The applicant has provided five radiation zones as a

basis for classifying occupancy and restrictions on access to various areas.

On this basis, maximum design dose rates are established for each zone and used

as input for design of shielding for the respective zone. The areas that will

have to be occupied on a predictable basis during normal operations and anticipated occurrences are zoned so that exposures will be below the limits 10 CFR 20

and will be ALARA, taking into account expected use and occupancy of the space.

The zoning system and access control features also meet the posting and entry

requirements of 10 CFR 20.203. The staff concludes that the radiation zone

system is acceptable.

The applicant has included many features in the design and layout of equipment

and facility shielding to maintain radiation exposures as low as practicable.

Among these are the location of major sources in shielded cubicles with access

through labyrinths to eliminate direct shine to accessible areas, the use 6f

shielded valve stations, and provisions for adequate cleaning of tanks.

Shielding and access control around the spent fuel transfer tube is acceptable.

The applicant has provided information on: (1) the types of structural barriers

and shielding provided to ensure acceptable radiation levels in adjacent

occupied areas, (2) a plan and elevation layout drawing of the areas near the

spent fuel transfer tubes, and (3) information on positive access control and

radiation monitoring to the areas near the spent fuel transfer tube. The staff

concludes that the radiation protection measures incorporated in the Watts Bar

design will provide reasonable assurance .that occupational radiation exposure

will be ALARA during spent fuel transfers and are acceptable.

Several features are included in the plant design and operational program to

minimize the buildup of activated corrosion products--a major contributor to

occupational dose. The use of high-cobalt, hard-facing wear materials in the

primary system has been limited to those places where it is necessary. Materials

with low cobalt impurity are specified for primary coolant system alloys. The

process of bright annealing reduces the corrosion tendency and also minimizes

the crud retention properties of the steam generator tubes. Valve and pipe

connections have been designed to control crud buildup. Control of chemistry

in the primary system will also control this buildup. The radiation fields

associated with major equipment in the plant are to be frequently reduced by

decontamination of components and systems containing buildups of cobalt.

Therefore, the staff concludes that the applicant has given acceptable consideration to the inclusion of design features to minimize the buildup of activated

corrosion products.

Watts Bar SER 12-4
In response to a staff question, the applicant has included features in the

radiation protection design specifically for the purpose of maintaining doses

ALARA during decommissioning. However, many of the features included in the

design to reduce doses during operation will also help reduce doses during

decommissioning. Specification and limitation on cobalt content in equipment

components will serve to limit radiation doses from crud buildup during both

operation and subsequent decommissioning. The staff estimates that the collective occupational dose because of decommissioning will be of the same order of

magnitude as average annual doses as a result of operation. Therefore, the

staff concludes that the applicant has given acceptable consideration to the

issue of personnel exposure during decommissioning.

The shielding was designed to meet the requirements of the radiation dose rate

zone system discussed above. The applicant's shielding design methods--including

the use of source terms, cross section data, shield and source geometries, and

radiation transport calculational schemes--are consistent with generally

accepted practice. Also, the applicant commits that the shield design and

construction will be consistent with the guidance of Regulatory Guides 8.8 and

1.69, "Concrete Radiation Shields for Nuclear Power Plants."

The shielding design review conducted for Sequoyah June 16, 1980, is applicable

to Watts Bar. This design review, which is discussed in Sequoyah SER Supplement 2, dated August 1980, meets the staff positions in NUREGs-0578 and -0737

and is acceptable for fuel loading and full power operation of Watts Bar.

The ventilation system is designed to ensure that airflow will be from areas of

low potential for airborne radioactivity to areas of higher'potential and then

to filters or vents. Also, the system will maintain concentrations of airborne

radioactivity in normally occupied areas within the limits of 10 CFR 20 and 50.

The ventilation filter trains are designed to allow exposures to be maintained

ALARA during servicing, consistent with the guidance of Regulatory Guide 8.8.

Therefore, the staff concludes that the ventilation system radiation protection

design features-are acceptable.

Detectors for the area radiation monitoring system will be located in normally

occupied areas which have the potential for radiation fields in excess of the

maximum design radiation dose rate. The detectors are designed to cover the

expected and maximum design dose rates and dose rates as a result of anticipated

operational occurrences. The monitors will have readout and annunciation in

the control room. They will also have variable alarm set points and local

audible alarms. The detectors will be calibrated quarterly. Therefore, the

staff concludes that the area radiation monitoring system design is acceptable.

The applicant has provided area radiation monitors around the fuel storage

areas to meet the requirements of 10 CFR 70.24 and to be consistent with the

guidance of Regulatory Guide 8.12, "Criticality Accident Alarm Systems."

The-applicant will rely on the area radiation monitoring system and portable

radiation monitoring instruments to assess the radiation hazard to personnel in

areas that may be accessed during the course of an accident. The area monitors

will receive backup power from the diesel generators. The portable instruments

will be placed to be readily accessible to-personnel responding to an emergency.

Watts Bar SER 12-5
The portable instruments will be designed with a sufficient instrument range

for use in the event of an accident.

The airborne radioactivity monitoring system is designed (1) to provide a clear

indication locally and to operations personnel when abnormal amounts of airborne radioactive material occur and (2) to provide information so that action

can be taken to ensure that inhalation of airborne particulates and iodine is

within the limits of 10 CFR 20. These airborne radioactivity monitors have the

capability to detect 10 maximum-possible-concentration hours (mpc-hr) of

particulate and iodine radioactivity in any compartment that has a possibility

of containing airborne radioactivity and that may be occupied by personnel.

The applicant's portable airborne radioactivity monitoring systems will monitor

air in areas not provided with fixed airborne radioactivity monitors. The

objectives and location criteria for these monitors are in conformance with

10 CFR 20 and 50 and Regulatory Guides 8.2 and 8.8. The staff concludes that

the radiation protection design features for Watts Bar are acceptable with the

criteria of SRP Section 12.3 and Items III.D.3.3 of NUREG-0737.

12.5 Dose Assessment

The estimates of annual person-rem exposure are based on conservatively assumed

radiation sources, design shielding, calculated design dose rates, and personnel

level--taking into account expected functions and occupancy times, and a

working year of 2000 hours. Based on expected dose rates and occupancy times,

expected airborne radioactivity concentrations, and estimates of the time and

personnel necessary to perform the various tasks involved in plant operation,

the applicant estimates that nonmaintenance personnel will receive 120 personrems per year. He estimates that maintenance personnel will receive 70-140

person-rems per year in the early plant life; as the plant ages the dose which

maintenance personnel will receive will increase to 210-280 person-rem per

year.

The dose assessment discussed above includes a breakdown of the estimated

annual person-rem doses that workers will receive during power operation within

the containment. This descriptive information meets the minimum information

needs set forth in Regulatory Guide 8.19, "Occupational Radiation Dose Assessment in Light Water Reactor Power Plants--Design Man-Rem Estimates." Therefore,

the staff finds the bases for the Watts Bar exposure estimates acceptable and

consistent with the acceptance criteria in the SRP Section 12.4. The applicant's

dose assessment fulfills these acceptance criteria. The estimates are reasonable in light of the operating experience reported in NUREG-0109, "Occupational

Radiation Exposure at Light Water Cooled Power Reactors 1969-1975."

12.6 Health Physics Program

FSAR Section 12.5 describes the applicant's health physics program. The

description includes the radiation protection organization, equipment, instrumentation and facilities, and procedures for radiation protection.

The applicant's organization will include health physics professionals and

technicians. The health physicist will have the responsibility for implementing

the health physics program and maintaining exposures ALARA. The organizational

aspects of the program are consistent with the guidance of Regulatory Guide 8.8.

Watts Bar SER 12-6
The applicant's radiation protection facilities will include portable instrument storage areas, personnel and equipment decontamination areas, change room,

access control point, health physics station, counting room, radiochemical

laboratory. A mobile facility will be used for whole body counting. A variety

of personnel monitoring devices and personnel protection equipment will also be

provided. Protective clothing, respiratory protection devices, and personnel

dosimeters will be included in the available equipment. The health physics

equipment, instrumentation, and facilities are consistent with the guidance of

Regulatory Guide 8.8, and are, therefore, acceptable.

12.7 NUREG-0737 Items

12.7.1 Plant Shielding (II.B.2)

NUREG-0737 requires assurance of adequate access to vital areas and protection

of safety equipment following an accident resulting in a degraded core. By

submittals dated June 16, 1980, and September 14, 1981, the applicant has pro- vided documentation of actions to be taken at Watts Bar to implement Item II.B.2

of NUREG-0737.

The shielding design review conducted for Sequoyah (dated June 16, 1980) is

applicable to Watts Bar. This design review, which is discussed in Sequoyah

SER Supplement 2, dated August 1980, meets the positions in NUREGs-0578 and

-0737 and is acceptable for fuel loading and full power operation of Watts Bar.

12.7.2 High Range Incontainment Monitor (II.F.1(3))

NUREG-0737 requires continuous indication in the control room from a high range

(107 rad/hr) incontainment monitor.

By submittals dated October 29 and July 2, 1981, and July 18, 1980, the applicant

provided documentation of actions being taken to implement Item II.F.1.(3).

The applicant has committed to install redundant high range radiation monitors

in containment, in both the upper and lower containments by fuel loading. The

locations are adequate to provide representative postaccident readings. The

monitors are General Atomic gamma monitors with a range of 1 to 107R/hr and are

powered from a vital bus with remote indication and recording in the control

room. Sensitivity is adequate to measure all fission product gamma radiation,

including XE'33 energies. Environmental and sesmic qualifications will meet

the criteria of Regulatory Guides 1.100 ("Seismic Qualification of Electric

Equipment for Nuclear Power Plants"), 1.89 ("Qualification of Clas.s 1E Equipment for Nuclear Power Plants"), and 1.97, Revision 2 ("Instrumentation for LightWater-Cooled Nuclear Power Plants To Assess Plant and Environs Conditions

During and Following an Accident"). Calibration will be performed in accordance

with the vendor's methods at each refueling outage. These actions meet the

staff positions in NUREG-0737 and are acceptable for fuel loading. The installation of these monitors will be verified by the staff.

12.7.3 Inplant Radioiodine Monitor (III.D.3.3)

NUREG-0737 requires assurances that the radioiodine concentration can be

accurately measured in areas within the plant where plant personnel may be

present during an accident.

Watts Bar SER 12-7
By submittal dated September 14, 1981, the applicant has provided documentation

of actions to be taken at Watts Bar to implement Item III.D.3.3 of NUREG-0737.

The Watts Bar Units have portable, low-volume air sampling equipment with both

charcoal and silver zeolite filters available for radioiodine sampling.

Analyses can be performed in the radiochemical laboratory using a Nuclear Data

6620 system with 3 Ge(Li) detectors.

Backup capability can be provided by an Eberline SAM-2/NaI detection system.

Alternate counting facilities, which provide a capability to promptly and

accurately analyze samples under low background conditions, are located in the

Sequoyah plant training center and the Sequoyah plant radiochemical laboratory.

These facilities each have Nuclear Data 6620 systems with 2 Ge(Li) detectors.

Procedures and training for postaccident sampling and analysis are provided

under the Watts Bar Health Physics Instruction Manual, Radiation Control

Instruction Manual, and inplant Training Program.

The postaccident radioiodine sampling and analysis provisions described for the

Watts Bar facility satisfactorily meet the staff positions for fuel loading and

full-power operations as outlined in NUREG-0737 and are acceptable.

Watts Bar SER 12-8
13 CONDUCT OF OPERATIONS

13.1 Organizational Structure of Applicant

13.1.1 Management and Technical Support Organization

General

The staff reviewed the organizational structure of TVA in connection with the

application for an OL for the Sequoyah nuclear plant. The Watts Bar review,

therefore, is an update of that review. As part of the Watts Bar review, a

review team visited the corporate office of TVA in Chattanooga, Tennessee, and

the Watts Bar site near Spring City, Tennessee, from October 27-29, 1981. The

team reviewed the proposed organization for operation of Watts Bar from the

level of the Senior Corporate Officer, who will be in overall charge of nuclear

operations for TVA, down to and including the proposed operating staff at the

plant. The review covered the organizational structure for the operation and

for support of the plant staff, the levels of staffing, the experience level of

principal individuals of both corporate and plant staffs, the interfaces between

the plant staff and its corporate support structure, and TMI Action Plan items.

The review team consisted of five persons, one of whom is the resident inspector

for Watts Bar.

Beginning with a meeting at TVA's corporate office in Chattanooga, the applicant

provided the review team with a thorough briefing on his organization for the

operation of Watts Bar. This briefing was followed by discussions with corporate

personnel so that the review team could gain a feeling for the responsibilities

and attitudes of the individuals, and an understanding of how they fit into the

organization and how they interface with other onsite and offsite organizational

units. Discussions with plant personnel also were held at the site to obtain

the same type of information. At the conclusion of the corporate and plant

management meetings, the team presented the audit results to TVA management.

The following description and evaluation are based on information contained in

the FSAR, the applicant's response to the NRC Action Plan Items (NUREG-0660),

and supplemental information obtained during the visit. The staff used the

guidelines of SRP Section 13.1.1 of NUREG-0800 (July 1981) for its review of

this section.

13.1.2 Corporate Organization and Technical Support

The corporate organization related to the construction and operation of Watts

Bar is shown in Figures 13.1, 13.2, and 13.3.

The applicant acts as his own architect/engineer. Engineering, design, and

construction are performed by TVA personnel with the assistance of contractors.

The Division of Engineering Design serves as the plant architect/engineer, and

the*Division of Construction is responsible for constructing the plant.

Watts Bar SER 13-1
rt

Cd.

m

ORGANIZATION OF THE TENNESSEE VALLEY AUTHORITY

I BOARD OF DIRIECTORS

OFFICE OF THE GENERAL MANAGER

Plaming Staff Budget Staff

Information Office Washington Office

Equal Employment Opportunity Staff

Office of the

General Counsel

I-h

C1

rP3

Office of Agricultural

and Chemical

Developmenit

l)ivisons:

Agricultural

Development

Chemical

Development

Chemical

Operations

Office of Power.

Divisions:

Nuclear Power

Power System

Operations

Transmission

Planning and

Engineering

Power

Construction

Fuels

Power Utilization

.Energy Conservatlo

and Rates

Energy Research

Office of Engineering

Design and

Construction

Divisions:

Engineering Design

Construction

X)fflce of Natural

Resources

Divisions:

Land and Forest

Resources

Water Resources

Natural Resources

Services

Land Between the Lakes

I

Office of Community

Development

Regional Studies Staff

Divisions:

Economic

Development

Community

Services

Uffice of Management

• Services

Union-Management

RelationsStaff

Divisions:

Property nnd Services

Finance

Purchasing

Personnel

Management Systems

lealthand Safety

- Medical Services

-Occupational I.ealth

and Safety

- Environmental

Compliance

Figure 13.1 Organization of the Tennessee Valley Authority

Source: FSAR Figure 13.1-1
C+

cn

OFFICE OF POWER

MANAGER OF POWER DEPUTY MANAGER

ASSISTANT MANAGER ASSISTANT MANAGER (ADMINISTRATIVE)

(FINANCING AND PLANNING)

POWER FINANCING STAFF

BUDGET AND COST CONTROL STAFF

POWER PLANNING STAFF

NUCLEAR REGULATION AND

SAFETY

QUALITY I AUDIT ASSURANCE AND STAFF

REGULATORY STAFF

NUCLEAR SAFETY STAFF

MANAGEMENT SERVICES STAFF

PERSONNEL MANAGEMENT

STAFF

LABOR RELATIONS

SYSTEMS PLANNING AND

DEVELOPMENT STAFF

POWER INFORMATION-STAFF

i,.,.

(,,,

Y

1ANAGER OF

POWER OPERATIONS

DIVISIONS

NUCLEAR POWER

FOSSIL AND HYDRO

POWER

POWER SYSTEMS

OPERATIONS

I

MANAGER OF ENERGY

CONSERVATION AND UTILIZATION

DI.VISIONS;

ENERGY CONSERVATION AND

RATES

ENERGY USE AND DISTRIBUTOR

RELATIONS

ENERGY DEMONSTRATIONS

AND TECHNOLOGY

MANAGER OF

POWER ENGINEERING

POWER PROJECTS. STAFF

DIVISIONS

TRANSMISSION PLANNING

AND ENGINEERING

FUELS

POWER CONSTRUCTION

Figure 13.2 Office of Power organizational chart
cr-I

CU

PERSONNEL. MANAGEMENT

SERVICES SERVICES STAFF STAFF

L. 0. HOLLiMAN G. J. ODELL

I

MANAGER

NUCLEAR

PRODUCT ION

T. G. CAMBELL

MANAGER

TECHN ICAL

SUPPORT 3. M. BALLENTINE*

i

BROWNS FERRY SEQAUOYA WATTS BAR NUCLEAR PLANT NUCLEAR PLANT NUCLEAR PLANT SUPERINTENDENT SUPERINTENDENT SUPERINTENDENT

G. T. JONES C. C. MASON W. T. COTTLE

BELLEFONTE

NUCLEAR PLANT

SUPERINTENDENT

A. M. QUALLS

FIELD

SERVICES

BRANCH

C. R. BRIMER

NUCLEAR POWER ORGANIZATION

THE DIRECTOR OF THE NUCLEAR POWER DIVISION WILL IMPLEMENT HIS RESPONSIBILITY THROUGH

A PRODUCTION MANAGER AND A TECHNICAL SUPPORT MANAGER. HE IS SUPPORTED IN THIS

POSITION BY AN ASSISTANT DIRECTOR. THE PRODUCTION MANAGER WILL ACT FOR THE

DIVISION DIRECTOR IN SUPPORTING AND DIRECTING THE POWER PLANT SUPERINTENDENT. THE

TECHNICAL SUPPORT MANAGER WILL PROVIDE THE PLANTS WITH THE REQUIRED TECHNICAL

SERVICES.

*NAME APPEARS TWICE

MAINTENANCE QUALITY EMERGENCY

AND ASSURANCE AND PREPAREDNESS TRAINING ENGINEERING COMPLIANCE AND PROTECTION BRANCH

MANAGER BRANCH BRANCH R. J. JOHNSON J. M. BALLBNTINE* R. C. PARKER J. W. MUPMAN

INSTRUMENT MECHANICAL REACTOR TCHNRICAL

AND CONTROLS BRANCH ENGINEERING SERVICES

BRANC1 T. BRANCH BRANCH H. B. BOUNDS T. D. KNIGHT R. A. SESSOMS

Figure 13.3 Division of Nuclear Power organizational chart
The Division of Nuclear Power is responsible for preoperational and startup

testing and operation and maintenance of the plant. The Division of Nuclear

Power is part of the Office of Power, which is responsible for the Operation

of all TVA power plants.

The Division of Nuclear Power is headed by the Director of Nuclear Power, who

reports to the Manager of Power (Office of Power). The Manager of Power in

turn reports through the Office of the General Manager to the TVA Board of

Directors (see Figures 13.1 and 13.2).

The Director of Nuclear Power, who administers the work of the division, is

assisted by an Assistant Director, a Nuclear Production Manager, a Technical

Support Manager, Personnel Services Staff, and the Management Services Staff

(see Figure 13.3).

The Nuclear Production Manager is responsible for nuclear production activities

including preoperational testing, power generation, and field services at TVA

nuclear generating plants. This includes ensuring that nuclear production

activities comply with the NRC operating license and established division

standards and requirements. The Nuclear Production Manager is assisted by the

Assistant Nuclear Production Manager.

The Nuclear Field Services Branch under the Nuclear Production Manager is

responsible for scheduling, planning, and implementing refueling outages,

major maintenance, and maintenance-related modifications, and for coordinating

the Installation of major modifications-at nuclear plants. The branch provides

supervision, administration, engineering, craft manpower, materials, and equipment required to carry out its field services responsibilities.

The Technical Support Manager is responsible for developing and maintaining

written programs and providing direct technical support to the nuclear plants

in the areas of maintenance and engineering support, quality assurance and

compliance, emergency preparedness and protection, industrial safety, and

training. The Technical Support Manager supervises the Maintenance and Engineering Manager, the Quality Assurance and Compliance Branch Chief, the

Emergency Preparedness and Protection Branch Chief, and the Training Branch

Chief.

The Maintenance and Engineering Manager is responsible for administering the

division's maintenance and engineering programs including the development and

maintenance of technical standards for nuclear production activities that are

consistent with regulatory requirements and TVA policy. The Maintenance and

Engineering Manager supervises the Electrical and Instrument and Controls; the

Mechanical, the Reactor Engineering, and the Technical Services Branches.

These branches provide direct technical support to the nuclear plant and field

service organizations.

The Quality Assurance and Compliance Branch is responsible for developing,

maintaining, and administering a quality assurance and commitment management

program including quality assurance (QA) program standards, vessel and piping

baseline and inservice programs, division audit and surveillance activities,

and commitment identification, tracking, and followup. The branch administratively supervises the Plant QA Sections including providing criteria, standards,

and technical guidance for plant QA activities.

Watts Bar SER 13-5
The Emergency Preparedness and Protection Branch is responsible for developing

and maintaining division protection programs including industrial safety, fire

protection, environmental protection, radiation protection, physical security,

and emergency preparedness. The Emergency Preparedness and.Protection OrganizatiQns provide direct technical support in the implementation of these programs. In carrying out these responsibilities, these organizations verify that

division activities comply with regulatory and TVA requirements and policy.

The Nuclear Training Branch is responsible for administering the nuclear operator

training program (NOTP), nuclear operator licensing training, and other technical

training programs. The objective of the NOTP is to provide qualified personnel

for operating positions in TVA's nuclear power plants. The program will culminate in a 2-year Associate Degree in Mechanical Engineering Technology. The

branch is responsible for the operation and maintenance of the Power Operations

Training Center, located near the Sequoyah nuclear power plant.

Other TVA Organizations

Other organizations within TVA which supplement the Watts Bar plant staff and

Division of Nuclear Power staff are

.The Division of Power Systems Operations

Power Systems Operations (PSO) provides the services of its central

laboratory and technical staff. In addition, field test engineers are

provided for chemical and laboratory tests and for solution of special

technical problems.

o The Division of Transmission Planning and-Engineering

Advice and assistance regarding the engineering and design of the electrical

transmission lines, substations, and communication facilities are available

from the Division of Transmission Planning and.Engineering.

o The Division of Fuels

The Division of Fuels plans the supply of fuel resources to serve TVA's

power demand forecasts. It defines requirements for nuclear fuels and

fuel-related components, and participates with the Division of Purchasing

in preparing and negotiating contracts for the procurement and processing

of nuclear fuel materials and fuel-related components (that is, all the

components of the fuel and control rod assemblies and the fuel channels).

The Division of Engineering Design

The Division of Engineering Design is responsible for designing all TVA

power plants and other major facilities. The services of this organization are available for advice and consultation on all engineering problems

and for carrying out the design of the major plant changes or additions

that may be required. After Watts Bar is operating, all plant changes or

additions must be approved by the Division of Nuclear Power.:

Watts Bar SER 13-6
0 .Regulatory Staff (Office of.Power)

One of the most important functions of the Regulatory Staff is to act as

liaison with the NRC and coordinate representation in meetings with the

NRC, licensing hearings, public meetings, and other discussions of proposed

licensing actions with other organizations participating as required.

The Division of Construction

The services of the DiVision of Construction are available for any major

plant construction that may be required-at the site.

The Division of Chemical Development

The Division of Chemical Development can furnish technical advice and

services on chemical and chemical engineering problems and provides

chemical engineering development-services in certain fields.

o The Division of Occupational Health and Safety (OCH&S)

The Division of Occupational Health and Safety is responsible for furnishing special services in the fields of industrial hygiene, environmental

hygiene, radiological protection,. pollution control, and other related

areas,. The Radiological-Hygiene Branch within OCH&S'provides administrative'supervision for the health physics unit at the plant.

Corporate Support Personnel

The number of professional support personnel (engineers) in the Division of

Nuclear Power is approximately as follows:

'Management 12

Mechanical Branch 31

Electrical & Instrument and Controls Branch 45

Technical Services Branch 41

QA and Compliance Branch 37

Emergency Preparedness Branch' 17

Reactor Engineering Branch 67

F•ield Services Branch 50

Total 300

In addition to the professional support personnel in the Division of Nuclear

Power, TVA has available many professionals from other organizations within TVA

that can supplement the Watts Bar and division of nuclear power staff when

needed. The Division of Engineering Design is responsible for the design of

Watts Bar and, therefore, has many professionals familiar with Watts Bar and

its systems.

Evaluation of Nuclear Corporate Organization and Staffing

TVA has recently reorganized its nuclear corporate organization by creating

some new branches, (such as the Mechanical, Technical Services, and Emergency

Preparedness and Protection Branches) and eliminating the outage management and

control and test branches. TVA also has moved experienced operations people

Watts Bar SER 13-7•
from the nuclear plants staffs into key corporate positions. The staff believes

that these changes will result in a stronger, more responsible corporate support

organization for the operating nuclear plants. All key management positions,

except the Manager of Maintenance and Engineering, have been filled (see

Figure 13.3).

The staff has reviewed the resumes of some of the individuals filling the

principal corporate positions related to nuclear operatons, such as the Director

of Nuclear Power (25 years nuclear experience, including 14years operations

experience), Assistant Director (20 years nuclear experience, including 6 years

of operations experience), Manager of Nuclear Production (10 years nuclear

experience, including 8 years operations experience), Manager of Technical

Support (20 years nuclear experience, including 10 years operations experience),

Chief of Mechanical Branch (10 years nuclear experience, including 2 years

operations experience), Chief of Technical Services Branch (12 years nuclear

experience, including 4 years operations experience), Chief of Reactor Engineering Branch (15 years nuclear experience, including 2 years operations experience),

Chief of Nuclear Training Branch (15 years nuclear experience), and Supervisor

of Preoperational Test Staff (15 years nuclear experience, including 8 years

operating experience). The staff concludes that the applicant meets the acceptance criteria of SRP Section 13.1.1 of NUREG-0800.

TVA has the advantage over some other utilities because TVA personnel have

gained considerable knowledge as a result of doing the detailed design,

engineering, and construction of Watts Bar, rather than employing an outside

architect/engineer to do this work. Moreover, TVA's experience operating

nuclear power plants and the recent transfer of people with extensive operating

experience into the corporate office or to Watts Bar should provide valuable

knowledge to management in solving operational problems common to any nuclear

power plant.

13.1.3 Plant Staff Organization

In the FSAR Amendment 44 and in supplemental information obtained during the

site visit, the applicant described the proposed plant operating organization

and the qualification requirements for key operating personnel.

The plant organizational chart is shown in Figure 13.4.

The Plant Superintendent

Onsite operational activities are conducted-under the supervision of the Plant

Superintendent. He reports to the Manager of Nuclear Production, who in turn.

reports to the Director, Division of,Nuclear Power of the Office of Power. The

Plant Superintendent is responsible for the operation of the plant. The plant

staff consists of approximately 500 full-time positions including organizational

units responsible for power plant engineering (approximately 57 people); power

plant operations (approximately 131 people); power plant mechanical maintenance

(approximately 71 people); power plant electrical maintenance (approximately

36 people); power plant instrument maintenance (approximately 87 people);

health physics; quality assurance; and other nontechnical supporting services.

Watts Bar SER 13-8
•-----

W I DIVISION OF NUCLEAR POWER

QUALITY

ASSURANCE I ER •AND COMPLIANCE

m TECHNICAL L -

SUPPORT"

OFFSITE CHIEF,

FIELD QUALITY

I ASSURANCE I

ONSITE g STAFF QUAL ITT,

L - - - - JASSURANCE INOEPENDENT MANAGER SAFETY ENGINEUV]NE

SUPERVISOR

- - QUALITY -

ASSURANCE

SECTION

L QUALITY I L• S PJU EI COORDINATORDVSINO - U I 

QUALITY NUCLEAR POWER

ASSURACE FIELD SERVICES URANCE BANCH

ENGINEERS I EA C

ASSISTANT ASSISTANTASSISTANT SUPEINTEDENTPLANT SUPERINTENDENT SUPERINTENDENT I SUPERINTENDENT FIELD ENGINEERING HEALTH AND POE MAINTENANCESERICES AND SAFETY SERVICES SYSTEMS I OPERATIONS L OPERATIONS

OPERATIONS P LIIIC

1 S AL SECURI SAFETY INSTRUMENT ELECTRICAL

ENIC 0I NTENCACE MAINTENANCE ANENAC

SUPERVISOR SUPERIO SERVICES KAOI' NTNK UK ANEA ANTENANCE

SERVICES -. SECTION SECTION SECTION ASSISTANT 0 "'S"M -OFi ]g - - SUPERVISOR SUPERVISOR SUPERVISON ''1'! OPERATIO INS OCCUPATIOVAL P00EeAS N SIRDIVISN OF TECHNICAL EN•IECRS TECGNICAL ENGINEERS TECHNICAL ENGINEERS

'STORESOR TO P C FOREMEN FOREREN FOREMEN S R SUPERVISOR SAFETY L SEAVICES _CRAFTSMEN CRAFTSMEN CRAFTSMEN

SRI L - -

ENGINEERING &SHIFT SHIFT TECHNICAL r DI OSON OF

TEST UNIT , ENGINEER - ADVISER T NUCLEAR POWER JANITORIAL L SUPERVSOR SRO PERSONNEL SERIS

L R ESSTAFF LABORER SERVICES

-Responsible for fire protectimo

Figure 13.4 Plant organizational chart
The principal groups that function directly under the supervision of the Plant

Superintendent and Assistant Plant Superintendents are the Operations Section,

Engineering Section, Mechanical Maintenance Section, Instrument.Maintenance

Section, and Electrical Maintenance Section. Staff services are provided by

the Management Services staff, Quality Assurance staff, and the Health Physics

Unit of the Industrial and Radiological Hygiene Branch. The last is under the

administrative supervision of the Division of Occupational Health and Safety.

The Watts Bar plant organization follows the pattern developed through experience and in use at all TVA conventional and nuclear generating plants.

Plant employees are selected primarily from existing TVA conventional plant

staffs and TVA's central office and must have significant conventional power

plant experience plus varying degrees of nuclear experience. Personnel qualifications will meet the criteria set forth in the ANSI N18.1-1971, "Standard

for Selection and Training of Personnel for Nuclear Power Plants."

Assistant Plant Superintendent

There are three assistant plant superintendents: Engineering and Operations

Assistant Plant Superintendent, Health and Safety Services Assistant Plant

Superintendent, and the Maintenance Assistant Plant Superintendent. These

superintendents assist the Plant Superintendent in planning, coordinating, and

directing the plant activities under their control. In the absence of the

Plant Superintendent, an assistant plant superintendent will be responsible for

management of the plant activities.

Reporting to the Engineering and Operations Assistant Plant Superintendent are

the Engineering Section Supervisior and the Operations Section Supervisor.

Reporting to the Health and Safety.Services Assistant Plant Superintendent are

the Safety Engineering Supervisor, who is responsible for fire protection; the

Plant Compliance Supervisor; the Management Services Supervisor; and functionally the Medical Supervisor, the Health Physics Unit, the Security Supervisor,

the Power Stores Supervisor, and the Training Officer. Reporting to the

Maintenance Assistant Superintendent are the Mechanical Maintenance Section

Supervisor, the Instrument Maintenance Section Supervisor, the Electrical

Maintenance Section Supervisor, and the Janitorial and Laborer Services

Supervisor.

Operations Section

The Operations Section is under the direction of the Operations Supervisor, who

reports to the Assistant Plant Superintendent--Engineering and Operations and

is responsible for all plant operations. It provides operating personnel for

the preoperational testing, fuel loading, startup, and operational testing. It

is responsible for coordinating and scheduling the training program for all

operations personnel. It also provides the nucleus of emergency teams such as

the plant rescue and fire-fighting organizations.

Engineering Section

The Engineering Section is under the direction of the Engineering Section

Supervisor, who reports to the Assistant Plant Superintendent--Engineering and

Operations. The Engineering Section is responsible for providing technical

Watts Bar SER 13-10
direction and staff assistance in the areas of nuclear, mechanical, and chemical

engineering. Responsibilities of this section include plant and equipment

performance tests, inplant fuel management, waste management, and chemistry

control.

Mechanical, Electrical, and Instrument Maintenance Sections

The Mechnical, Electrical, and Instrument Maintenance Sections are under the

direction of section supervisors who report to the Assistant Plant Superintendent--Maintenance, and are responsible for all maintenance work and inspections in the plant. This includes scheduling and conducting periodic tests on

the systems assigned to the sections associated with the reactor and engineered

safeguards as required by the Technical Specifications and licenses. The

sections develop and carry out preventive maintenance programs that ensure that

the repair and replacement of parts are consistent with the intent of applicable codes and basic requirements of the original equipment.

Health Physics Unit

The Health Physics Unit of the Radiological Hygiene Branch is responsible for

health physics activitites at this plant. It develops and applies radiation

standards and procedures; reviews proposed methods of plant operation; participates in development of plant documents; and assists in the plant training

program, providing specialized training in radiation protection. It conducts

comprehensive onsite environmental monitoring-before, during, and after plant

startup and provides radiological health coverage for all operations including

maintenance, fuel handling, waste disposal, and decontamination. It is responsible for personnel and inplant radiation monitoring and maintains continuing

records of personnel exposures, plant radiation, and contamination levels. It

advises on potentially harmful factors in the working environment other than

radiation.

This unit is under the administrative supervision of the Chief, Industrial and

Radiological Hygiene Branch, in the Division of Occupational Health and Safety

and reports functionally to the Health and Safety Services*Assistant Plant

Superintendent.

Management Services Staff

The Management Services staff, under the supervision of the Management Services

Supervisor, performs management service functions, administrative, and clerical

services for the plant.

Plant Quality Assurance Staff

The Plant Quality Assurance (QA) staff is under the direction of the plant QA

Staff Supervisor and is responsible for developing, planning, initiating, and

directing a comprehensive nuclear power quality assurance/quality control

program in the plant. The Plant QA Supervisor reports to the Quality Assurance

and Compliance Branch Chief located in the corporate technical support

organization.

Watts Bar SER 13-11
Shift Staffing and Shift Technical Advisor

In the FSAR Amendment 44, the applicant has committed to minimum shift staffing

and overtime limitations that conform to the guidance given in NUREG-0737. In

TVA Administrative Instructions.AI-2.4 (shift staffing) and in information

received during its audit visit in October 1981, and in a letter dated

January 29, 1982, the applicant outlined the-minimum shift crew and overtime

limitations.

The shift crew composition for one unit for normal routine operations will

consist of one Shift Engineer (SRO), who is.the .Station Shift Supervisor; one

Assistant Shift Engineer (SRO); two unit operators (ROs); and four assistant

unit operators. Three health physics technicians also will be assigned to each

shift, one for each unit and one for the fire brigade. One additional Assistant Shift Engineer (SRO) and one additional unit operator (RO) will be required

for two-unit operation. Besides the above shift staffing, each shift also will

have a Shift Technical Advisor (STA) and a communicator (Clerk). In the event

that the Assistant Shift Engineer needs to leave the control room, the Shift

Engineer or another SRO will relieve the Assistant Shift Engineer.

The applicant states that the Shift Technical Advisor will have the following

qualifications: (1) training in basic engineering principles (degree in

engineering or science), (2) extensive training in plant transient and accident

response, (3) Technical Specifications training with emphasis on the bases for

limiting conditions for operation, and (4) significant reactor training on

systems and operating procedures. The 32 weeks of STA training will meet the

NRC's 60-hour college credit recommendation.

The staff has reviewed the TVA Administrative Instructions AI-2.4-on shift

staffing and overtime and found the instructions acceptable, since they meet

the guidance of NUREG-0737. The staff also has reviewed the STA training and

qualifications as outlined in a TVA letter dated January 29, 1982 and found

that the requirements outlined in the October 30, 1979 letter from H. R. Denton

with regard to the 60-credit hour college level course work and STA experience

have been met.

Operator License Candidates in "Pipeline"

Based on a five-shift crew operation, a minimum of 10 senior reactor operators

(SROs) and 10 reactor operators (ROs) is required. For 1982, there are 32 cold

license candidates for shift staffing, 23 of these are SROs and 9 are ROs. Ten

of these candidates were previously licensed on Sequoyah, which is practically

identical to Watts Bar in design.

If one assumes an average examination failure rate of about 30 percent and an

attrition rate of about 10 percent, these candidates normally would produce

just enough licensed operators for five-shift operation at fuel load. However,

because of the number previously licensed on Sequoyah, a failure rate of less

than 30-percent can be expected. Thus, the staff finds that the present plans

should provide an adequate number of licensed operators before fuel loading.

The applicant also has planned to have 19 hot license reactor operator candidates for. shift staffing shortly after plant startup. The addition of these

candidates should provide the necessary margin for-proper staffing.

Watts Bar SER 13-12
Fire Protection Organization and Fire Brigade Members

The Safety Engineering Supervisor is responsible for fire protection (see

Figure 13.4). The Watts Bar Fire Brigade consists of one Assistant Shift

Engineer, who is the Fire Brigade Leader, and three Assistant Unit Operators,

plus a chemical laboratory technician. The staff finds the proposed fire

protection organization and the fire brigade membership acceptable, provided

plant administrative procedures clearly state that when the Assistant Shift

Engineer assumes his duties as Fire Brigade Leader, his control room duties are

temporarily assumed by the Shift Supervisor (Shift Engineer), or by another

SRO, if one is available (see Section 13.5.1). The staff finds the responsibilities for fire protection and the composition of the fire brigade are in

accordance with BTP-CMEB.9.5-1.

Plant Staff and Qualifications

The staff has reviewed the number of people on the plant staff (professionals

and nonprofessionals) and found it acceptable, as compared to other similar

plants recently reviewed and licensed. The staff has also reviewed the qualification, education, and experience requirements for the plant personnel, as

well as the resumes of key staff members filling these positions, and found

them acceptable. They meet the requirements of ANSI N18.1-1971.

The applicant has stated that each operating shift will have at least one

person with experience in PWR startup operations at a comparable facility.

After having interviewed several of the plant staff members and reviewed the

resumes of key plant personnel, the staff finds that many have obtained startup

and operating experience at Sequoyah. The staff will condition the license to

require that there will be at least one person who has commercial PWR startup

experience on each operating shift (at least until a nominal 100 percent power

operation is obtained). This condition can be satisfied by using either

appropriately qualified TVA personnel or contractor personnel.

Preoperational and Startup Testing Organization

The applicant has described the startup organization in FSAR Section 14.2. The

Division of Nuclear Power has the responsibility for administration of the

preoperational and startup test programs. The Preoperational Test Section

consists of approximately 20 systems-oriented engineers functioning under the

direction of the section supervisor who acts as the Preoperational Test Program

Coordinator.

The Power Plant Superintendent is the onsite representative responsible for the

overall conduct and coordination of the preoperational and startup test programs.

He directs onsite implementation of the preoperational test program through the

Preoperational Test Section Supervisor, who acts as preoperational test program

coordinator. The Plant Engineering Section Supervisor is responsible to the

Power Plant Superintendent for coordination of the startup test program. These

two supervisors assure that the detailed test instructions are responsive to

the specific designs of the Watts Bar systems. In addition, they are responsible for the conduct of the tests, evaluation of test results, documentation

of results, and final disposition of test records.

Watts Bar SER 13-13
The staff has reviewed the resumes of key personnel in the preoperational and

testing organization and compared them with the requirements given in ANSI

N18.1-1971 for positions the staff considers equivalent to the functions to

which the applicant has assigned the personnel. The staff considers these

qualifications acceptable.

Summary and Conclusions

Based on its review of information given in the FSAR Amendment 44 and in the

review meeting with,the applicant in October 1981, the staff concludes

(1) The corporate and plant organizational structures for operation and the

total number of personnel planned to staff the organizations are acceptable.

(2) Engineering, design, construction,'and startup are performed by TVA

personnel with some assistance from Westinghouse; TVA has, therefore,

gained considerable knowledge about the Watts Bar plant..

(3) TVA's experience from operating other nuclear power plants has provided

experience to managment in solving operational problems common to any

nuclear power plant.

(4) The qualification, education, and experience requirements for the corporate

technical staff supporting the nuclear plant operations and the qualifications of key personnel filling these positions are acceptable.

(5) The shift staff will be augmented as necessary with individuals who have

prior experience with large PWR operations.

(6) As a result of the recent reorganization it appears that there are clear

management controls and effective lines of authority and communications

between the organizational units involved in the management, operation,

and technical support for the operation of the Watts Bar plant.

These findings contribute to the staff's conclusion that the applicant complies

with the requirements of 10 CFR 50.40(b) and that the applicant will have the

necessary managerial and technical resources to provide assistance to the plant

staff in the event of an emergency, as specified in SRP Section 13.1.2 of

NUREG-0800.

13.2 Training

13.2.1 Licensed Operator.Training Program

The overall conduct and administration of the plant training program is the

responsibility of the Plant Superintendent. He has designated the Training

Supervisor as responsible for administering the training program and monitoring

the program effectiveness. The applicant states that the training program will

provide reasonable assurance that decisions and actions during all plant conditions will be made consistent with plant safety procedures and operational

limits established to protect.the public health and safety. The program for

formal education and training of the reactor operator has been designated to

meet the individual needs of the participants, depending upon their background,

Watts Bar SER 13-14
previous training, and expected job assignment. The program conforms to the

requirements set forth in ANSI N18.1-1971, 10 CFR 55, and Item I.A.2.1 of

NUREG-0737.

The training program for personnel, who will be licensed, consists of the

following discrete segments: courses in nuclear power plant steam and mechanical fundamentals; power plant electric fundamentals and mathematics and

general physics review; PWR technology, systems description, heat transfer,

fluid fl~ow and thermodynamics; mitigating core damage; reactor system simulator

training; research reactor training; and refresher training.

The simulator training program, which consists of a 12-week program, is cQnducted

at the TVA training center utilizing the facilities builtto simulate the

Sequoyah nuclear plant. Because the Sequoyah and Watts Bar plants are essentially identical, this simulator faithfully reproduces, for training purposes,

all the Watts Bar transients. Instructors at the simulator are TVA employees

experienced at Browns Ferry or Sequoyah nuclear plants. Each student in the

simulator training course will receive at least 100 hours of control room time.

At the time of licensing'of Sequoyah Units 1 and 2, the applicant submitted

requalification programs and a replacement training program, together with

following TMI Action Plan Items I.A.2.1, I.A.2.3, I.A.3.1 and II.B.4, which are

applicable to Watts Bar, as well as to Sequoyah. These requalification programs

and replacement training program, together with the TMI Items mentioned above,

were approved acceptable for the Sequoyah units. Therefore, they are applicable

to and acceptable for the Watts Bar plant.

On the basis of its review, the staff concludes that the training programs and

schedule for the applicant's staff are acceptable for the preoperational test

program, for operating license examinations, and for fuel loading.

The basis for this review was conformance to positions on SRP Section 13.2.1 of

NUREG-0800, Regulatory Guide 1.8, NUREG-0737, and 10 CFR 55.20 through 55.23

and 55.25.

13.2.2 Training for Nonlicensed Personnel

The applicant has presented his training program for nonlicensed personnel in

the FSAR. The training is under the direction of the Chief--Nuclear Training

Branch, who is responsible for the development, implementation, and administration of all training activities in the Division of Nuclear Power. The Chief--

Nuclear Training reports to the Manager--Technical Support (see Figure 13.3).

Supervisors, engineers, technicians, and craftsmen are provided initial training,

retraining, and specialized courses to satisfy the applicable requirements of

their particular positions. This section was reviewed against the guidelines

of SRP Section 13.2.2 of NUREG-0800 (July 1981).

General Employee Training

All persons employed at the Watts Bar nuclear plant, including temporary

personnel, will be trained in the following areas commensurate with their job

duties:

.Watts Bar SER 13-15
(1) general description of plant and facilities

(2) job-related procedures and instructions

(3) radiological health and safety program

(4) station emergency plans

(5) industrial safety program

(6) fire protection program

(7) security program

(8) quality assurance program

Specialized Courses

Some of the specialized courses include

(1) IBM Selectric input/output devices course

(2) STA training

(3) PWR instrumentation and control

(4) basic nuclear instrumentation program

(5) computer programming and review

(6) Prodac computer maintenance

(7) electro-hydraulic control system course

(8) shift technical advisor training

(9) mitigating core damage training

(10) fire protection training

Mitigating Core Damage Training

The mitigating core damage training program was developed to ensure that all

Watts Bar nuclear plant operating personnel are trained in the use of installed

plant systems to control or mitigate an accident in which the core is severely

damaged.

The training program includes the following areas:

(1) incore instrumentation

(2) excore instrumentation

(3) vital instrumentation

(4) primary chemistry

(5) radiation monitoring

(6) gas generation

The program consists of classroom training in the above-listed areas with a

final examination. The staff finds the program acceptable.

Training for New Engineers

The division training organization has engineering training courses for new

engineers coming into the division, as well as simulator short courses on-both

simulators. These courses provide engineers, new to the field of nuclear

power, an overall understanding of the basics of nuclear power production and

the design and purpose of support systems. In addition, openings are available

on a limited basis for corporate support personnel to attend existing operator

and STA training classes.

Watts Bar SER 13-16
Fire Brigade Training

The training of the fire brigade meets Appendix R, 10 CFR 50 requirements. The

fire brigade refresher training is given quarterly to ensure all brigade

members receive the training during a 2-year period. The required subjects are

covered. Fire brigades are drilled quarterly on announced drills and annually

on an unannounced drill.

Each Shift Engineer and Assistant Shift Engineer receives a formal fire brigade

leadership training course within I year of his appointment and once every

3 years thereafter. A firefighting and fire prevention course is included

in the student generating plant operator training program to ensure all operators

receive basic fire training. Additionally, operators in all classifications

and other plant employees as deemed necessary by the Plant Superintendent

receive the fire brigade member refresher training course at least once every

year. All new employees receive instructions on proper actions in the event of

a fire emergency within the first 2 weeks of employment.

Full-scale fire drills will be conducted on an annual basis. In addition,

drills for plant fire brigades will be conducted once each quarter for each

shift brigade. Once each quarter, a review of fire emergency procedures will

be included in the craft safety meetings.

The staff has reviewed the applicant's fire brigade training program, found

that it meets the requirements of Appendix R to 10 CFR 50, and is, therefore,

acceptable.

STA Training

The applicant states the Shift Technical Advisor will have the following qualifications: (1) training in basic engineering principles (engineering degree),

(2) extensive training in plant transient and accident response, (3) technical

specification training with emphasis on the bases for limiting conditions for

operation, and (4) significant reactor training on systems and operating

procedures.

The STA training program consists of 32 weeks of training covering the following

topics: plant systems, plant familiarization and walk-through, applied fluid

flow, thermodynamics, simulator training, applied reactor physics, radiochemistry,

simulator transient and accident analysis, and management and supervisory

training.

The staff has reviewed the STA training and qualifications as outlined in the

January 29, 1982 submittal from the applicant and finds the requirements

outlined in the October 30, 1979 letter from H. R. Denton with regard to the

requirements of 60-credit hours of college-level course work and STA experience

have been met.

Review of Training

During a review meeting with the applicant in October 1981, the staff discussed

the Watts Bar training program and the training organization as described in

the FSAR and by the applicant at that meeting. The staff also visited the TVA

Watts Bar SER 13-17
Power Operations Training Center, which is located near the Sequoyah nuclear

station. The review team found that the training organization is well staffed

and that the training program for nonlicensed personnel meets the requirements

of 10 CFR 50.34 (a and b) and the acceptance criteria of SRP Section 13.2.2 of

NUREG-0800.

13.3 Emergency Preparedness Evaluation

13.3.1 Introduction.

The applicant submitted a Watts Bar plant radiological emergency plan on

January 27, 1982.

The plan was reviewed against the requirements in 10 CFR 50.33 and 50.47, therequirements in Appendix E to 10 CFR 50, and the guidance evaluation criteria

in NUREG-0654/FEMA-REP-1, Revision 1, entitled "Criteria for Preparation and

Evaluation of Radiological Emergency Response Plans and Preparedness in Support

of Nuclear Power Plants," November 1980, which is endorsed by Revision 2 to

Regulatory Guide 1.101.

The review was done in accordance with NUREG-0800, Revision 2, July 1981, SRP

Section 13.3, "Emergency Planning."

This evaluation report follows the format of Part II of NUREG-0654 in that each

of the planning standards is listed and followed by a summary of the applicable

portions of the plan that relate to that specific standard. The final section

of this report provides the staff's conclusions.

13.3.2 Evaluation of the Emergency Plan

Assignment of Responsibility (Organizational Control)

The planning standard of NUREG-0654 requires that primary responsibilities for

emergency response by the nuclear facility'licensee, and by state and local

organizations within the emergency planning zones have been assigned, the

emergency responsibilities of the various supporting organizations have been

specifically established, and each principal response organization has staff

to respond to and to augment its initial response on a continuous basis.

The shift engineer on duty at the Watts Bar plant is designated as the Site

Emergency Director. When an abnormal condition arises, it is his responsibility to determine if the abnormality meets any of the emergency classifications specified in the plan and to implement the plan, if necessary. There

is a 24 hour-a-day communication capability between the station and Federal,

state, and local response organizations to ensure rapid transmittal of accurate

notification information and emergency assessment data.

Responsibility for overall performance of the emergency response organization

is vested in the Site Emergency Director, who is responsible for the overall

direction of the plant emergency organization. Qualified members of the

station staff, who report directly, have been assigned specific responsibilities for the major elements of emergency response.

Watts Bar SER 13-18
Updated written agreements with appropriate agencies and organizations are

maintained by the Division of Occupational Health and Safety and include police

departments, ambulance services,:and hospitals near Watts Bar and the U.S. DOE.

Onsite Emergency Organization

The planning standard of NUREG-0654 requires that the onshift facility licensee

responsibilities for emergency response are unambiguously defined, adequate

staffing to provide initial facility accident response in key functional areas

is maintained at all times, timely augmentation of response capabilities is

available, and the interfaces among various onsite response activities and

offsite support and response activities are specified.

The shift engineer assumes the responsibilities of the Site Emergency Director

until relieved by the Plant Superintendent or his designated alternate. The

authorities and responsibilities of the Site 'Emergency Director have been

clearly specified, including those that cannot be delegated. The Site Emergency

Director can immediately and unilaterally declare an emergency and make offsite

notifications.

Station staff emergency assignments have been made and the relationship between

the emergency organization and normal staff complement are shown in the-plan.

Positions and/or titles and qualifications of shift and plant staff personnel

both on and off the site, who are assigned major emergency functional duties,

are listed. Table 1 of the Watts Bar plan, as revised by letter dated April 22,

1982, describes minimum onshift staffing levels that conform to the guidance of

Table B-1 of NUREG-0654.

TVA management personnel will supply support services utilizing any necessary

manpower and equipment. A long-term emergency organization framework is in

place, headed by the Central Emergency Control Center Director. Interfaces

between and among the TVA staff, station staff, governmental and private sector

organizations and technical and/or engineering contractor groups have been

specified along with services to be provided.

Emergency Response Support and Resources

The planning standard of NUREG-0654 requires that arrangements for requesting

and effectively using assistance resources have been made, arrangements to

accommodate state and local staff at the licensee's near-site emergency operations facility have been made, and other organizations capable of augmenting

the planned response have been identified.

TVA established its central emergency control center (CECC) in.Chattanooga,

Tennessee. The stated purpose of the CECC and its staff is to provide the

facilities and manpower for evaluating, coordinating, and directing the overall

activities involved in coping with a radiological emergency at any TVA nuclear

facility.

When the emergency plan is activated, the Shift Engineer on duty at the reactor

control room notifies the Operations Duty Specialist in Chattanooga, who notifies Tennessee state authorities, the TVA Information Office and the Directors

of four TVA emergency response centers, and in the case of a general emergency,

also makes a direct parallel notification to local county emergency operations

centers.

Watts Bar SER 13-19
The four TVA emergency'response centers are

(1) The CECC in Chattanooga

(2) The division of nuclear power emergency center (DNPEC) located immediately

adjacent to the CECC. The DNPEC provides organizational, material,

personnel, and technical support from Nuclear Maintenance, Reactor Engineering, Controls and Test, and Outage Branches.

(3) The Knoxville emergency control center provides the technical support of

the Division of Engineering Design personnel and the personnel and equipment. support of the Division of Construction.

(4) The Muscle Shoals emergency control center provides staff and equipment

for performing environmental radiological monitoring and dose assessments

and for recommending protective actions for the public.

The TVA radiological emergency plan also has identified facilities and staff at

Oak Ridge at the George C. Marshall Space Flight Center, Tennessee state and

local police organizations--all available for assistance in an emergency.

The conceptual design of the emergency response facilities has not been received

for Watts Bar, as required by'Generic Letter 81-10 from Darrell G. Eisenhut,

Director Division of Licensing, NRR, dated February 18, 1981 (Subject: Post-TMI

Requirements for the emergency operations facility). This concept will be

reviewed separately and wi~ll be addressed in a supplement to the SER.

Emergency Classification System

The planning standard of NUREG-0654 requires that a standard emergency classification and action level scheme, the bases of which include facility system and

effluent parameters, is in use by the nuclear facility licensee, and state and

local response plans call for accurate information to be provided by facility

licensees for determining the minimum initial offsite response measures.

TVA utilizes the following emergency classifications:

(1) Notification of unusual event

(2) Alert

(3) Site emergency

(4) General emergency

This system of classification is consistent with the systems used by state and

local emergency organizations. The initiating conditions used for recognizing

and declaring the emergency class are based on specific measurable values or

observable conditions defined as emergency action levels (EALs). The specific

instrument readings and parameters required for determination of these EALs are

detailed in the inplant procedures and will be used as thresholds for determining the emergency classifications.

Notification Methods and Procedures

The planning standard of NUREG-0654 requires that procedures be established for

notification, by the licensee, of state and local response organizations and

Watts Bar SER 13-20
for notification of emergency personnel by all response organizations; the

content of initial and 'followup messages to response organizations and to the

public be established; and means to provide early notification and clear

instruction to the populace.within plume exposure pathway emergency planning

zone (EPZ) be established.

Procedures have been established for notification of state and local response

organizations in case of emergency. The Site Emergency Director has been given

authority and responsibility to initiate prompt notificaton to these agencies.

TVA notifies the local authorities through the permanently manned position of

the operations duty specialist (ODS). The ODS, upon notification from the

affected plant of a general emergency, is responsibile for notifying the local

Civil Defense Agency at once, before any other action is taken. Notification

of the appropriate state agency is the ODS's next action.

In a letter dated August 20, 1981, the applicant committed to a prompt notification system having the design objective capability to essentially complete the

initial notification of the public within the plume exposure pathway EPZ within

about 15 minutes. TVA has committed to expedite procurement of a prompt

notification system to be installed and operational in accordance with the

following estimated dates (per TVA letter dated April 22, 1982):

Complete system conceptual design December 1, 1981

Complete equipment procurement August 2, 1982

Complete land acquisition July 15, 1982

Begin construction August 16, 1982

Complete construction October 29, 1982

Complete system checkout November 30, 1982

Turn system control over to

State of Tennessee July 1, 1983

The system at Watts Bar is described in a letter dated January 19, 1982 and

consists of three components: (1) fixed sirens, (2) moble sirens, and (3) tone

alert radios.

(1) Fixed sirens

Nineteen 125-dB electronic sirens will be used for warning areas where

there are high concentrations of transient people. Such areas would be

Watts Bar Lake, boat launches, marinas, and campgrounds.

Eight 125-dB, ten 115-dB, and twenty-nine local coverage electromechanical

sirens will be used to warn permanent residents in the more densely

populated areas.

(2) Mobile sirens

State and local emergency personnel will patrol predetermined routes with

mobile sirens. These routes will be predominantly in the sparsely populated

areas of the 5- to 10-mi EPZ.

(3) Tone alert radios

These radios will be placed 'in selected institutions, both public and

private, within the 10-mi EPZ.,

Watts Bar SER 13-21
The adequacy of this system will be evaluated during actual system tests--that

is during joint exercises of the facility with offsite authorities--and will be

upgraded if necessary in response to any identified deficiencies.

Emergency Communications

The planning standard of NUREG-0654 requires that provisions exist for prompt

communications among principal response organizations to emergency personnel

and to the public.

The station communications system is designed to provide secure, redundant, and

diverse communications to all essential onsite.and offsite locations during

normal operations and under accident conditions. Within-station systems areý

comprised of a public address system, two-way radio systems, a private automatic

(PAX) exchange, and a sound-powered telephone system. Offsite systems are

comprised of both commercial and leased telephone lines, a microwave system,

and two-way radio systems. A Bell Telephone ring-down system is dedicated as

the primary means of communication between plant and offsite emergency control

centers.

These telephones plus other systems located in plant areas are manned 24 hours

a day. The Site Emergency Director will in emergency situations communicate

directly with the TVA Operations Duty Specialist who is responsible for providing initial notification to the appropriate state emergency organizations. In

the event of a general emergency, he is required to notify the appropriate

local response agency. These offices are manned 24 hours a day.

Communications between the control room, the technical support center, and

operations support center (such as PAX telephone) are available.

Tests of the radiological emergency communications network are held weekly.

Public Education and Information

The planning standard of NUREG-0654 requires that information is made available

to the public on a periodic basis on how they will be notified and what their

initial actions should be in an emergency (listening to a local broadcast

station and remaining indoors), the principal points of contact with the news

media for dissemination of information during an emergency (including the

physical location or locations) are established in advance, and procedures for

coordinated dissemination of information to the public are established.

TVA, in coordination with appropriate state agencies, annually provides all

residents within 16.1 km of each TVA nuclear plant a pamphlet containing educational information on radiation, respiratory protection, sheltering, and

evacuation routes. The pamphlet also contains contacts for additional information, including local Civil Defense offices and TVA's toll-free Citizen Action

Line. A copy of each mailer is forwarded to the Division of Occupational

Health and Safety.

TVA holds news media training sessions at least annually to acquaint reporters

with the basics of reactor operations, the health effects of radiation, emergency

planning, points of contact for information during an emergency, nuclear waste

management, operator training, and other pertinent subjects.

Watts Bar SER 13-22
Following an emergency, the Information Office will assess the need for a

postemergency followup on the effectiveness of the emergency information

program.

In an emergency, the Manager of Power serves as the chief spokesperson for TVA

at press briefings. The CECC Information Officer is responsible for coordinating TVA'news releases with the TVA centers and the state as required.

TVA has identified the primary near-site media center.(NSMC) in a letter from

L. M. Mills to H. R. Denton dated January 27, 1982 as being in the Quality Inn

at the intersection of Interstate 75 and State Highway 68 in Sweetwater,

Tennessee, and the alternate NSMC as being in the Ramada Inn at the same

intersection. Written agreements with the Quality Inn and Ramada Inn are not

yet signed, but oral agreement between the Quality Inn and TVA has been obtained.

By letter dated April 22, 1982, TVA has stated that the written agreements will

be signed by January 1, 1983. The staff finds this acceptable and will confirm

that these agreements are signed in accordance with this schedule.

Emergency Facilities and Equipment

The planning standard of NUREG-0654 requires that adequate emergency facilities

and equipment to support the emergency response be provided and maintained.

Emergency facilities needed to support an emergency response have been provided,

including a technical support center (TSC) and an operations support center

(OSC). Each will be activated for an alert or higher emergency classifications.

The TSC has been established in the relay room in the power house control bay

inside the protected area. The TSC contains a complete set of functional plant

drawings and necessary technical information.

The OSC (assembly area) is located in the station's locker and lunch room and

will be the assembly point for assigned response personnel. Equipment and

supplies--respiratory protective devices, protective clothing, portable

lighting--are available, if needed.

Stored equipment is inspected and inventoried and replaced when calibration or

repair is needed. Sufficient equipment exists to ensure a minimum inventory

in case of replacement delay. Portable monitoring instruments are stored in

the health physics area. Calibration of equipment is carried out at intervals

recommended by the supplier of the equipment or quarterly.

The concept of the TVA response facilities, which includes the NRC nearsite

facility, has not been received as required by Generic Letter 81-10 dated

February 18, 1981, for Watts Bar facility. Once this concept is received,

it will be reviewed separately and addressed in a supplement to the SER.

The meteorology equipment at the site meets the criteria of Regulatory

Guide 1.23, "Onsite Meteorological Programs," February 17, 1972. An alternate

source of data has been identified to which TVA may gain-access. Other information per Appendix 2 of NUREG-0654 has been received and is under staff review.

The meteorological program will be reviewed separately and will be addressed in

a supplement to the SER.

Watts Bar SER 13-23
Accident Assessment

The planning standard of NUREG-0654 requires that adequate methods, systems,

and equipment for assessing and monitoring actual or potential offsite consequences of a radiological emergency condition be in use.

Onsite capability and resources to provide initial and continuing assessment

throughout the course of an accident include process, effluent, and area

monitors that read out in the control room; postaccident sampling capability;

and containment monitoring. The applicant has provided information on these

capabilities in response to the criteria of NUREG-0737.

The applicant is prepared to assess the consequences of potential or actual

releases off site. TVA has transmitted to the staff a copy of the Muscle

Shoals Emergency Control Center (MSECC) Implementing Procedures. These procedures include emergency actual and predictive dose assessment procedures for

atmospheric and liquid releases of radioactivity from the Watts Bar plant.

Watts Bar has a van equipped to monitor the terrestrial environment for radioactivity. The vehicle will be dispatched for any emergency situation of the

alert, site, or general class within 30 minutes.

An additional sampling teamcan be at the plant within 2 hours of notification.

A third team will be dispatched from the Muscle Shoals facility. The MSECC

teams may be transported by helicopter or fixed-wing aircraft.

A dose calculational methodology for identifying the potential scope of radiological consequences of emergency situations, including capabilities for dose

projection using real-time meteorological information and for dispatch of

radiological monitoring teams is under continuing review as more information

becomes available from the applicant. This will be addressed in a supplement

to the SER.

Protective Response

The planning standard of NUREG-0654 requires that a range of protective actions

be developed for the plume exposure pathway EPZ for emergency workers and the

public. Guidelines for the choice of protective actions during an emergency,

consistent with Federal guidance, are developed and in place, and protective

actions for the ingestion exposure pathway EPZ appropriate to the locale have

been developed.

The applicant's radiological emergency plan (REP) includes a table of recommended

protective actions, including shelter and evacuation, with special consideration

for children and pregnant women to reduce whole-body and thyroid doses from

exposure to a gaseous plume, that are consistent with both the State of Tennessee

and U.S-. Environmental Protection Agency (EPA) guidelines. The recommended

actions include the.statement that officials may implement low-impact protective actions at lower values in keeping with the principle of maintaining

radiation exposures ALARA.

The applicant has committed to make recommendations for the ingestion exposure

pathway to state and local agencies, but these agencies are responsible for the

Watts Bar SER 13-24
decision to act upon such recommendations. The State of Tennessee Radiological

Emergency Response Plan states that protective actions will be based on the

protective action guides developed by the U.S. EPA and the U.S. Food and Drug

Administration.

An evacuation times assessment study within the plume exposure pathway EPZ in

accordance with the criteria set forth in Appendix 4 to NUREG-0654 has been

provided. This item will be reviewed separately and will be addressed in a

supplement to the SER.

Radiological Exposure Control

The planning standard of NUREG-0654 requires that means for controlling radiological exposures, in an emergency, be established for emergency workers. The

means for controlling radiological exposures shall include exposure guidelines

consistent with EPA Emergency Worker and Lifesaving Activity Protective Action

Guides.

Implementing procedures have been developed to prevent or minimize exposure to

radiation for onsite individuals. These procedures include evacuation, accountability, radiological monitoring, and decontamination of nonessential personnel.

Respiratory protective equipment and protective clothing is provided for essential plant personnel who would remain on site. The applicant's REP includes

emergency guidelines for doses during an extreme emergency that are consistent

with EPA Emergency Worker and Lifesaving Activity Protective Action Guides.

The REP states that personnel must be made aware of possible consequences of

such exposures and must be selected on a voluntary basis unless they are members

of an emergency team and have previously consented to receive the exposure.

The Site Emergency Director can authorize exposures in excess of 10 CFR 20

limits.

The State of Tennessee and local agencies are responsible for implementing

action to protect the health and safety of the public off site. The applicant

will recommend protective actions, but the state and local governments are

responsible for deciding if any actions are needed and what they should be.

Potential choices for dose reduction for the public are shelter; evacuation;

closing of public water supplies; confiscation of crops, food, and dairy

products; placing milk animals on uncontaminated stored feed; and use of

potassium iodide. TVA has potassium iodide available for public use at state

direction, according to specific agreements with the various states.

Medical and Public Health Support

The planning standard of NUREG-0654 requires that arrangements be made for

medical services for contaminated injured individuals.

The applicant has made arrangements wth Athens General Hospital and Rhea County

Hospital of Athens and Dayton, Tennessee, respectively, to provide medical

assistance to site personnel injured or exposed to radiation and/or radioactive

material. These facilities have emergency plans, staff training programs, and

adequate equipment and supplies for receiving radiologically contaminated

patients. The Oak Ridge Hospital in Oak Ridge, Tennessee, is available for the

care and treatment of radiation accident victims from TVA and is used as a

backup hospital. Based on the quality of the facilities at the Athens General

Hospital and Oak Ridge Hospital, the staff finds the arrangement acceptable.

Watts Bar SER 13-25
Emergency medical equipment, such as stretchers and respirators, is strategically located throughout the plant, and approved trauma kits and other

specified equipment are readily available for use by the medical emergency

response teams.

The applicant has a health station that contains the normal complement of first

aid supplies and equipment necessary to treat injuries not involving hospitalization or medical services.

An ambulance is available at the site to transport injured workers to hospitals.

An agreement is maintained with Rhea County and McMinn County Ambulance Service

to transport injured workers when necessary.

The Division of Medical Services provides training to selected onsite personnel

to qualify them in first aid and emergency medical care. These personnel then

serve as members of the plant medical emergency response teams.

Recovery and Reentry Planning and Postaccident Operations

The planning standard of NUREG-0654 requires that general plans for recovery

and reentry be developed.

The Site Emergency Director and the site emergency organization will direct the

recovery and reentry operations following an incident. All other TVA resources

plus other governmental and vendor support will be available through the TVA

corporate organization to aid the Site Emergency Director in developing,

evaluating, and implementing specific recovery and reentry operations. All

major postincident recovery measures shall be performed in accordance with

written procedures.

The decision to downgrade an incident will be made by the Site Emergency

Director after consultation with his plant technical and operations staffs and

will be coordinated with the CECC Director.

Coordination of recovery activities will be handled through the CECC, Chattanooga,

Tennessee.

Exercises and Drills

The planning standard of NUREG-0654 requires that periodic exercises be conducted

to evaluate major portions of emergency response capabilities, periodi'c drills

be conducted to develop and maintain key skills, and deficiencies identified as

a result of exercises or drills be corrected.

The purpose of an exercise is to test the integrated capability of TVA, Tennessee

state and local emergency response organizations. It is designed to test a

major portion of the basic elements existing within emergency preparedness .

plans and organizations. The Division of Occupational Health and Safety

evaluates deficiencies disclosed in the critique and coordinates corrective

actions through TVA and follows up to ensure completion.

A combined exercise involving TVA, state, and local personnel will be held

annually. The scenario for the exercise will.be mutually agreed to and rotated

Watts Bar SER 13-26
each year to ensure that all major elements of the emergency plan are tested

over a 5-year period. At least once every 6 years, an exercise will be scheduled

for each of the off shifts.

Drills based on emergency conditions will be held at least annually for response

components--such as fire, medical, health physics, and communications--to

ensure maximum effectiveness of the plan.

Each scenario is forwarded to the radiological emergency planning group,

Radiological Hygiene Branch, not later than 2 weeks before conducting a drill.

This group shall approve the scenario and forward copies to the Quality Assurance

(QA) Staff, Radiological Hygiene Branch, and the QA and Audit Staff, Office of

Power. Each drill will be conducted and critiqued by an independent TVA

organization.

Radiological Emergency Response Training

The planning standard of NUREG-0654 requires that radiological emergency

response training be provided to those who may be called on to assist in an

emergency.

TVA provides training in emergency procedures to all permanent plant personnel

and all nonplant personnel expected to be on site for longer than 1 week. The

training enables each employee to have a working knowledge of the emergency

plan and his/her responsibility and action.if an emergency is declared.

Training consists of initial training classes and annual retraining, drills,

and activation of the alarms to maintain familiarity with the features of the

emergency plan. Training and annual retraining is provided to those offsite

agencies who may be involved during an emergency and will include procedures

for notification, basic radiation protection, their expected roles, and site

access procedures, as applicable.

Offsite agencies, which potentially will be called upon to participate in the .plan, have concurred with the responsibilities assigned their agency in the

plan by executing a Letter of Agreement with TVA.

Emergency medical technicians are certified by TVA under a recognized system to

provide emergency care and related services to victims of illness or injury.

Responsibility for the Planning Effort: Development,* Periodic Review, and

Distribution of Emergency Plans

The planning standard of NUREG-0654 requires that responsibilities for plan

development and review and for distribution of emergency plans be established,

and planners be properly trained.

Emergency Plan

The Emergency Plan and Emergency.Plan Implementing Procedures are formally

reviewed annually for adequacy and applicability by the Division of Occupational

Health and Safety (OCHS), noting any required changes. OCHS issues controlled

revisions and ensures that all holders receive all changes.

Watts Bar SER 13-27
The qualifications of TVA personnel responsible for radiological emergency

planning include academic training-in engineering or science. Several employees

have graduate training in nuclear engineering or health physics, and all have

at least 2 years' experience in engineering, health physics, or emergency'

planning.

The QA Staff, Radiological Hygiene Branch, and the QA and Audit Staff, Office

of Power, audit the plan yearly for compliance with existing regulations and

its own internal requirements. These QA organizations are responsible for

offering recommendations on overall plan improvement. The results of audits

are documented, reported to appropriate organizational management, and retained

in the respective QA files for a period of 5 years.

TVA has agreements with outside organizations for radiological emergency

support to furnish specific services. Copies of the letters documenting these

agreements are forwarded to the Division of Occupational Health and Safety.

These letters are updated every 2 years by the TVA organizations requiring

these services.

13.3.3 Conclusions

On the basis of its review against the criteria in "Criteria for Preparation

and Evaluation of Radiological Emergency Response Plans and Preparedness in

Support of Nuclear Power Plants," NUREG-0654, Revision 1, November 1980, the

staff concludes that the Watts Bar Nuclear Plant Radiological Emergency Plan,

upon satisfactory correction of the items identified in the foregoing evaluation

and listed below, will provide an adequate planning basis for an acceptable

state of emergency preparedness and will meet the requirements of 10 CFR 50 and

Appendix E thereto.

(1) Sufficient information on the following item identified in the foregoing

Watts Bar evaluation has not been received.

- Standard C - Near-Site Emergency Operations Facility for Watts Bar

(2) Information on the following items identified in the foregoing Watts Barevaluation has been received and is under staff review to determine if the

standards in 10 CFR 50.47(b) are met.

- Standard H - Meteorological Program

- Standard I - Dose Calculational Methodology

- Standard J - Evacuation Times Assessment Study

(3) Acceptable review of the FEMA findings and determinations as to whether

the state and local emergency response plans are adequate.

(4) Acceptable completion of the joint exercise held in accordance with the

requirements of Section IV.F.1.b of Appendix E to 10 CFR 50 to demonstrate

the dynamic capability to implement the onsite and offsite emergency

response plans.

Once the above items are completed, the staff's evaluation on the overall state

of emergency preparedness for the Watts plant will be presented in a supplement

to this report.

Watts Bar SER 13-28
13.4 Review and Audit

The applicant has established a review and audit program to ensure that the

operation of the plant is in conformance with established operating procedures

and license provisions and to review and approve design changes, changes in

operating procedures, or changes in the Technical Specifications that may

constitute an unreviewed nuclear safety question, as defined in 10 CFR 50.59.

This section was reviewed against the guidelines of SRP Section 13.4 of NUREG0800.

Onsite Review

A continuing review of operational activities is a normal function of the plant

staff. Two plant organizations, the Plant Operations Review Committee (PORC) and the Plant QA Staff, have specific responsibilities in the area of review of

plant operation. A description of these two operations, including their

responsibility and authority, is in Chapter 17 of the FSAR. In addition, the

quorum requirements, meeting frequency, responsibility, authority, and record

requirements of the Plant Operations Review Committee are in the Technical

Specifications.

The PORC serves in an advisory capacity to the-plant superintendent in matters

related to safety in plant operation. It meets at regular monthly intervals

and for special meetings called by the superintendent. Membership includes the

Plant Superintendent (chairman), Assistant Plant Superintendent (alternate

chairman), Engineering Section Supervisor, Operations Section Supervisor,

Safety Engineering Supervisor, Plant Compliance Supervisor, Mechanical Maintenance Section Supervisor, Instrument Maintenance Section Supervisor, Electrical

Maintenance Section Supervisor, Quality Assurance Supervisor, and the Health

Physicist. The minutes of each meeting are recorded. Responsibilities of the

PORC include the review of activities affecting plant safety. These activities

areldescribed in the administrative section of the appropriate Technical

Specifications. In addition, the PORC reviews the adequacy of the plant QA

program and recommends appropriate changes.

Besides the PORC and Plant QA Staff, theapplicant will establish an Independent

Safety Engineering Group (ISEG). The ISEG will consist of three permanently

assigned engineers on site working in conjunction with a central office core

staff. In addition to the onsite ISEG, a group of five engineers assigned to

the central office will act as a part-time support to the plant ISEG in any

problems or investigations that can be dealt with off site. This will include

investigating any potential problems that can be of a generic nature or that

require information not readily available to the onsite ISEG. Both the onsite

and offsite ISEG staffs will be extensively involved with nuclear experience

review, which will aid in getting pertinent information to responsible personnel

within the plant and in bringing potential problem information to the ISEG

members. This will include reviewing plant industry licensee event reports

(LERs), Institute of Nuclear Power Operation significant experience reports,

NRC Power Reactor Events, and so forth.

The proposed ISEG approach for Watts Bar is a departure from that implemented

at Sequoyah. The applicant believes the Watts Bar approach may be an improvement on the existing position. TVA proposes to implement the new approach at

Watts Bar SER 13-29
Watts Bar only and evaluate the effectiveness of the two approaches at a later

date.

At the audit meeting in October 1981, the staff requested more detailed information on ISEG, including the experience requirements of the ISEG members and to

whom the ISEG supervisor reports. In a letter dated January 29, 1982, the

applicant provided additional information on ISEG.

The applicant states that an ISEG member must have a degree in engineering, or

equivalent, and at least 4-1/2 years experience in the operation, design, or

licensing of nuclear power plants. A safety problem identified by the ISEG

will be first brought to the attention of the responsible plant employee. If

the problem is not corrected at this level, the compliance supervisor will

bring it to the attention of the plant superintendent. If the compliance

supervisor is not satisfied with the resolution, the communication channel to

the offsite Technical Support Manager will be used to bring the problem to the

attention of the central office technical support organization. The ISEG

supervisor reports to the Manager -*Technical Support located off site.

Independent Review

Independent review of operational activities is the function of the TVA Nuclear

Safety Review Board (NSRB). The NSRB functions during the planning, design,

construction, and operation phases of all TVA nuclear plants to advise the

Manager of Power on the adequacy and implementation of TVA nuclear safety

policies and programs.

The NSRB is a standing committee responsible for the independent review program

and is composed of a chairman and five or more members. The chairman of the

NSRB is the Executive Officer, Nuclear Safety Staff (NSS). The membership

includes at least one management-grade member each from the Division of Nuclear

Power, the Division of Fuels, the Licensing Staff, the Nuclear Safety Staff,

and the Office of Power Quality Assurance and Audit Staff. Members are selected

from among those regularly involved as supervisors or staff specialists in

nuclear facility engineering, operation, safety analysis, and related activities, and may serve on a part-time basis. No more than one-third of the

members are from any one organization. Qualified alternates are permitted;

however, no more than two alternates can participate as voting members in NSRB

activities at any one time. In the event of the legitimate absence of principal

members, alternate members are selected by the chairman as required to accomplish

the review and audit functions of the NSRB.

NSRB members shall have experience in one or more of the areas listed below to

be capable of (1) recognizing and evaluating nuclear safety problems and their

effects and (2) verifying the adequacy of corrective actions. To ensure that

appropriate expertise is brougnt to bear, the NSRB, through the use of members,

alternate members, and consultants, shall collectively have the experience and

competence required to provide independent review and audit and to assess the

adequacy of review and audit by others of designated activities in the following

areas:

(1) nuclear power plant operations

(2) nuclear engineering

(3) chemistry and radiochemistry

Watts Bar SER 13-30
(4) metallurgy

(5) instrumentation and control

(6) radiological safety

(7) mechanical and electrical engineering

(8) administrative controls and quality assurance

If the NSRB requires additional expertise, the NSRB shall have access to TVA

personnel and outside consultants in specialty fields associated with the

unique characteristics of the plant involved and shall be provided assistance

from the TVA line organizations, as necessary, to accomplish its responsibilities

in a timely manner.

Individual members of the NSRB, alternate members, and consultants shall have a

baccalaureate or advanced degree in engineering or the sciences or the equivalent.

Individuals providing expertise in one of the disciplines listed above shall

have at least 5 years of professional.experience in the discipline or specialty

area represented.

The NSRB is responsible for reviewing nuclear safety-related activities, programs, and events to evaluate the safety of TVA nuclear plants and to ascertain

that nuclear safety aspects are. being adequately considered. It shall be responsible for identifying problems and verifying the adequacy of corrective actions

and conducting reviews, audits, and investigations of any subject of nuclear

safety concern and may make recommendations, including recommended solutions,

to appropriate management within the Office of Power.. The NSRB shall be responsible for obtaining the appropriate technical assistance, as necessary, to

accomplish its responsibilities.

At the audit meeting in October 1981, the staff asked the applicant to provide

further information on the Nuclear Safety Review Board (NSRB), including responsibilities, authority and qualifications of its members. In a letter dated

November 9, 1981, the applicant submitted copies of the Nuclear Safety Review

Board Charter and the Nuclear Safety Review Procedures (Power - NSRP 3.1) dated

January 1981.

The staff reviewed the above documents and found them acceptable as they meet

the requirements of ANSI N18.7-1976, "Administrative Control and Quality

Assurance for the Operational Phase of Nuclear Power Plants," as endorsed by

Regulatory Guide 1.33, "Quality Assurance Program Requirements."

Summary and Conclusion

The applicant's program for the review of plant operations is in conformance

with staff guidance described in Regulatory Guide 1.33 and applicable industry

standards, (ANSI N18.7) and is acceptable. The applicant has described the program for the review of operational activities. The program includes reviews by

the plant staff organization, reviews of safety-related activities independent

of the operating organization, and independent reviews and assessments of plant

activities by an independent group located on site. The staff has reviewed the

provisions for these reviews in respect to organizational provisions, qualification requirements of those performing the review, and the subject matter to be

reviewed. The staff finds that the applicant's program for the review of

operational activities is in conformance with staff guidance described in

Regulatory Guide 1.33 and the applicable industry-standard (ANSI-N18.7); the

Watts Bar SER 13-31
qualification levels for plant staff personnel performing reviews meet the

guidelines of Regulatory Guide 1.8 and applicable industry standard ANSI N18.1,

Section 4.4; the provisions for an independent review meet the guidelines of

Regulatory Guide 1.33 and applicable industry standard ANSI N18.7 (ANSI 3.2),

Section 4.3 and ANSI/ANS 3.1, Section 4.7; and the applicant's Safety Engineering Group meets the guidelines of Section I.B.1.2 of NUREG-0737, and meets the

acceptance criteria of Section 13.4 of NUREG-0800 (July 1981).

13.5 Plant Procedures

13.5.1 Administrative Procedures

The Plant Superintendent issues, in the form of standard practices, instructions

governing employee actions and established standards for plant operation. The

applicant has stated~in the FSAR that these instructions are to be written in

accordance with standard ANSI 18.7-1976, Revision 2, (ANS 3.2) and contain

administrative restrictions and station requirements established to ensure safe

operation of the plant within the limits set by the facility licenses and

Technical Specifications.

A formalized system of written instructions conforming to the requirements of

the Operation QA Program Plan is employed in support of the standard practices.

Plant instructions are prepared by the appropriate section supervisors or other

staff personnel. They are issued only after reivew by the PORC and approval by

the Plant Superintendent. Instructions covering plant operations, maintenance

work, tests, equipment changes, and other activities which might adversely

affect safety are put into effect only after being reviewed by the PORC, and

with written authorization of'the Plant Superintendent. It is the Plant

Superintendent's responsibility to ensure that required reviews and approvals

are completed before authorizations are issued.

The PORC is responsible for reviewing proposed changes to plant instructions.

On the basis of the recommendations received from this group, the power plant

superintendent is responsible for determining whether further review is required

before approving a change.

Administrative Instructions prescribe the methods whereby plant instructions

can be temporarily changed without delay when the need arises. Temporary

change to plant instructions shall be made in accordance with the Technical

Specifications. Administrative Instructions define the authority and responsibility of the ROs and SROs.

This section was reviewed against the guidelines of SRP Section 13.5.1 of

NUREG-0800.

Shift Supervisor Responsibilities (Administrative Duties)

In the FSAR, Administrative Instruction AI-2.1, and in a January 29, 1982

letter, the applicant has stated that his administrative procedures, shift

supervisor job descriptions, and training programs emphasize the primary

management responsibility of the Shift Engineer. In addition, periodic

retraining acts to reinforce the Shift Engineer's command responsibilities and

the applicant will periodically issue a management directive which emphasizes

this assignment of responsibility.

Watts Bar SER 13-32
The applicant also has stated that the plant administrative procedures will be

reviewed to ensure that they clearly define the authority and responsibilities

of each position on shift. The duties of the Station Shift Supervisor are

performed by the Shift Engineer (SRO). The duties of the control room SRO are

performed by the Assistant Shift Engineer on each unit.

The Assistant Shift Engineer's normal work station is in the control room. In

the event that the Assistant Shift Engineer needs to leave the control room,

the shift engineer or another SRO will relieve the Assistant Shift Engineer in

modes 1, 2, 3, and 4.

The administrative duties of the Shift Supervisor will be reviewed by the

senior officer of TVA responsible for plant operations. Administrative

functions that detract from or are subordinate to ensuring safe operation of

the plant will be assigned to other employees. The following actions have

already been taken:

(1) A clerk will be assigned to the Shift Engineer's office on each shift to

provide administrative support to the Shift Engineer.

(2) Routine nonmanagement duties formerly assigned to Assistant Shift Engineers

have been assigned to other employees.

As noted in Section 13.1, the staff requires that the plant administrative

procedures clearly define that when the Assistant Shift Engineer assumes the

role of Fire Brigade Leader, his control room duties are temporarily assumed by

the Shift Supervisor, or by another SRO, if one is available.

Shift and Relief Turnover Procedures

The applicant has stated in the FSAR that he has developed and-will implement

shift and relief turnover procedures for Watts Bar Units that provide assurance

that the oncoming shift has adequate knowledge of critical plant status information and system availability. A checklist will be completed and signed by

offgoing and oncoming shifts at each shift turnover.

This checklist includes critical plant parameters and allowable limits, availability and proper alignment of safety systems, and a listing of safety system

components in a degraded mode along with the length of time in that mode. All

shift personnel responsible for the status of critical equipment have relief

checklists for oncoming and offgoing shifts that will include any core cooling

equipment under maintenance or test that could degrade a safety system. In

addition, a system will be established to evaluate the effectiveness of the

turnover procedures.

Control Room Access

The applicant stated in the FSAR that he had developed and will implement

plant-specific administrative procedures that establish specific individual

authority and responsibility, as well as delineating various system or equipment functions related to controlling personnel access during normal and

accident conditions. A control room access plan has been developed to provide

direction to all members of the plant staff to ensure that those persons

responsible for safe operation of the plant are able to perform effectively.

Watts Bar SER 13-33
In addition, TVA has developed and will implement procedures that establish a

clear line of authority and responsibility in the control room in the event of

an emergency. These procedures clearly define the lines of communication and

authority for plant management personnel and ensure that the Shift Supervisor,

his assistant, or senior licensed management personnel are the only plant

personnel who have the authority to direct licensed activities of licensed

reactor operators.

Procedures for Feedback of Operating Experience to Plant Staff

The applicant's procedure implements a program to meet the specific requirements of NUREG-0737, which deals with the review of operating experience

reports. It• establishes a system to ensure the review of operating experience

reports to document their applicability to TVA plants, to provide required

written responses, and to ensure proper disposition of all applicable items.

This review can detect potential problems before they occur at a TVA plant and

find new ideas for solving existing problems and improving plant operation.

The sources of information and experience review listed below shall be distributed to responsible organizations and plant staffs to determine applicability'

to the TVA nuclear program. Where possible, this information will be condensed

to prevent extraneous and unimportant material from burdening the plant staff.

Vendor information letters

Inhouse-LERs

NRC bulletins

NRC circulars

NRC information notices

EPRI Reports - Technical Summaries

INPO Summary (SEE-IN) or other industry review of LERs

Division of Engineering Design Experience Review

TVA's Reactor Engineering Branch (REB) is responsible for (1) coordinating the

review (2) the dissemination of operating experience information to organizations that could be affected by this experience,• and (3) ensuring that final

resolution of recommendations made by the reviewing organization is completed.

Upon receipt by Nuclear Power of any of the defined sources of review material,

(with the exception of inhouse LERs and any industry review of LERs), REB will

determine the appropriate branch or TVA's staff to be assigned primary responsibility. Once this is determined, representatives of TVA's REB, the responsible

branch or TVA's staff, and the plant compliance staff will discuss the extent

of each group's efforts to resolve the issue and to ensure that duplication of

effort is minimized. The assigned branch or staff will be responsible forreviewing the assigned material and, in conjunction with the plant compliance

staff, will make recommendations for changes in equipment and operating

practices.

TVA's REB will use information from the inhouse LERs and the industry review of

LERs to establish a data base for the purpose of identifying significant trends

in equipment failure or operating errors. Should a significant trend develop,

the appropriate staff will be assigned primary responsibility. Representatives

of TVA's REB and the plant compliance staff will discuss the extent of each

group's efforts to resolve the issue. The assigned organization will be

Watts Bar SER 13-34
responsible for making recommendations for changes in equipment and operating

practices.

The Plant. Superintendent shall

,(1) Evaluate recommendations and incorporate operating changes, initiate

equipment changes, modify or institute inhouse training, or make information available.

(2) Develop plant instructions that ensure that the information and/or action

is reviewed and understood by the appropriate people in the plant. The

plant instructions shall be explicit as to the personnel who must review

the different types of experience review.

Changes in operating instructions and practices and their bases

should be reviewed as a minimum by shift operating personnel and

upper plant management affected by the change. Certain changes will

affect other plant sections. Any other affected sections should be

identified to ensure that the information is reviewed by all applicable plant personnel.

The bases for design and equipment changes made as a result of an

experience review should be understood by upper plant management,

operating personnel, and other applicable sections.

Changes in the training program as a result of an experience review

shall be routed by applicable personnel as deemed appropriate. A

means will be provided to ensure that affected personnel become aware

of and understand that information of sufficient importance should

not-wait for emphasis through routine training and retraining programs.

Guidance on Procedures for Verifying Correct Performance of Operating

Activities

The applicant has stated in the FSAR that-the current plant administrative

procedures require that:

(1) The alignment of all systems and components important to safety be verified

prior to unit startup

(2) Changes in the alignment of any system important to safety be recorded on

the applicable status sheet of the configuration control system

(3) Shift personnel being relieved communicate information on any abnormal

plant condition including temporary conditions

(4) System operability be demonstrated before a system is returned to service

(5) Approval by the Shift Supervisor or his representative be received before

the performance of any activity on any systems important to safety or any

activity that may affect systems important to safety. The Shift Supervisor

or his representative is notified when any activity authorized to be per- formed on a system important to safety is completed or a change occurs in

the scope of the activity.

Watts Bar SER 13-35
Plant operating instructions require completion of a startup checklist before

unit startup. This checklist is used to verify correct alignment of all systems

important to safety. In addition, alignment of systems important to safety is

reviewed each shift. Any time a critical component is changed from its normal

position or condition, an entry is made on the applicable status sheet of the

configuration control system. Panel checklists are reviewed each shift to

verify proper panel alignment exists for all systems important to safety.

It is the applicant's opinion that this verification function can be performed

adequately by an assistant unit operator (AUO),and that the use of licensed

unit operators is not necessary. The AUO has sufficient training and familiarity with plant systems to ensure correct system alignment, and this policy

will allow the licensed operator to remain in the control room. Second verification of system status will be required for systems important to safety,

including but not limited to those systems listed below.

(1) auxiliary feedwater

(2) emergency core cooling

(3) emergency gas treatment'

(4) essential raw cooling water

(5) reactor coolant system

(6) component cooling water

(7) containment spray

(8) residual heat removal

(9) emergency diesel generators

(10) upper head injection

(11) spent fuel pit cooling

In addition, all calibrations and loop checks of the reactor protection system

instrumentation require double verification signoffs before returning the

instrument to service. The applicant will include additional systems for

double verification as necessary.

In a letter dated January 29, 1982, the applicant provided additional information on how the AUO will be qualified to perform the verification function.

All AUOs presently assigned to the Watts Bar facility shall be systematically

placed in a break-in period of training for each work station that they are not

presently qualified on. The break-in period may vary from 1 day to 5 days

depending on the complexity of the work station. The Operations Supervisor

shall schedule the AUOs for this break-in training. The break-in period training for each AUO work station shall be conducted by an experienced AUO. The

training, as a minimum, shall consist of such things as

(1) routine duties of work station such as readings and inspections

(2) familiarization of work area such as layout of floor plan, equipment in

area, fire protection, and telephones

(3) operations of equipment such as startup, normal operation, and shutdown

(4) type of work in area when unit is starting up or shutting down

(5) hazards within the work area

(6) -cautions/precautions relative to work area

Following the break-in training period, an oral certification examination shall

be given to establish that the training has been adequate to qualify the trainee

Watts Bar SER 13-36
to competently and safely man the work station. The examination shall be administered by a qualified operator other than the operator serving as instructor.

The Shift Engineer of each group shall select the examiner from his group of

qualified operators. The examination shall be documented and attested to by

the Shift Engineer. If the results of the examination are unsatisfactory, the

trainee will be put back into training for whatever additional period of time

is required, as decided by the Shift Engineer, to qualify him for the position.

This will be decided through subsequent oral examination or he will be restricted

from assuming responsibility for the particular work location.

Summary and Conclusion

The staff concludes that the Administrative Procedures are acceptable and con-.

tribute to meeting the requirements of 10 CFR 50.40(b), and 50.54(1). This

conclusion is based on the following:

The applicant has described the program and procedures that provide administrative controls over activities important to safety. These include the activities

of the preparation, review, and approval of plant operating and maintenance

procedures; the responsibility and duties of shift personnel; shift relief and

turnover procedures; access to the control room; limitations on working hours;

the feedback of operating information to plant personnel; the procedure for

verifying the correct performance of operating activities; and the administrative provisions for the control of the initial plant test program. The staff

has reviewed these provisions and finds they meet the staff guidance described

in Section 5.2 of ANSI/ANS 3.2 and in Regulatory Guide 1.33 and the applicable

parts of the Task Action Plan Items I.A.1.2, I.A.I.3, I.C.2, I.C.3, I.C.4, I.C.4,

I.C.5, and I.C.6. The applicant also has met the guidelines of ANSI B30.2 and

Section 4.4.6 of ANS 3.1 and, therefore, meets the acceptance criteria of SRP

Section 13.5.1 of NUREG-0800.

13.5.2 Operating and Maintenance Procedures

The staff has reviewed the applicant's description of his plan for development

and implementation of operating and maintenance procedures. The review consisted of (1) a comparison of the plans for Watts Bar to the plans for Sequoyah,

as described in Section 13.5 of the FSAR and (2) a comparison of selected Watts

Bar emergency operating procedures with the Sequoyah procedures.

The review of Section 13.5 of the FSAR disclosed that the applicant's plans

for the-preparation, review, approval, and use of written procedures at Watts

Bar are essentially identical to those for Sequoyah. Therefore, the discussion

and conclusions regarding operating and maintenance procedures in Section,13.5

of the Sequoyah SER are applicable to Watts Bar.

The comparative review of emergency operating procedures included Abnormal

Operating Instruction, AOI-16, Loss of Normal Feedwater, and the following

Emergency Operating Instructions:

EOI-O Immediate Actions and Diagnostics

EOI-1 Loss of Reactor Coolant (LOCA)

EOI-3 Steam Generator Tube Rupture

EOI-13 Anticipated Transients Without Scram

Watts Bar SER 13-37
In each case, the Watts Bar procedure was similar to the Sequoyah procedure. Some necessary differences were noted in the procedures due to differences in

the design of .the plants. Those differences were related to setpoints for

steam generator level and pressure, reactor coolant system temperatures, diesel

generator maximum loading, auxiliary feedwater flow, and containment spray pump

suction arrangement.

Other differences, which did not appear necessary, were also noted between the

procedures for the two plants. These differences are as follows:

AOI-16 Loss of Normal Feedwater

(1) Nomenclature for auxiliary feedwater pumps is not consistent.

(2) The caution in the Sequoyah procedure that warns the operator that the

pressurizer PORVs may open if only one auxiliary feedwater pump is available, and instructs him to verify it has reseated, is missing from the

Watts Bar procedure.

(3) The Watts Bar procedure does not verify reactor trip breaker P-4 contact

position after reactor trip.

EOI-O Immediate Actions and Diagnostics

(1) The Watts Bar procedure (which is to be used for Units 1 and 2) addresses

only the Unit 1 controlled hydrogen ignition system.

(2) The Watts Bar procedure does not instruct the operator to open breakers on

the ice condenser air handling units.

(3) Following instructions for resetting safety injection and securing emergency core cooling pumps, the same caution statement is listed twice.

These items were discussed with the applicant. The applicant stated that these

differences will be eliminated. Subject to confirmation that this action is

completed, the discussion and conclusions on emergency operating procedures contained in the Sequoyah Safety Evaluation Report Supplement 2 (Section 22.2, Item I.C.1) are applicable to Watts Bar nuclear plant with the following

exceptions:

(1) The Watts Bar power-ascension test procedures and emergency operating procedures have not yet been reviewed by the NSSS vendor (Westinghouse

Electric Corporation) as required by NUREG-0737, Item I.C.7. The applicant has stated in his September 14, 1981 response to the TMI Task Action

Plan that this will be accomplished. Subject to confirmation that this

review has been completed, the staff concludes that the requirements of

NUREG-0737, Item I.C.7,'have been met.

(2) Because of the similarity of the Watts Bar and Sequoyah plants and proce- dures, the pilot monitoring of the selected emergency operating procedures for Watts Bar did not include simulator exercises or control room walkthroughs of the procedures. Subject to confirmation that the Watts.Bar

procedures have been revised to make them consistent with the Sequoyah

Watts Bar SER 13-38
procedures, the staff concludes that the requirements of NUREG-0737,

Item I.C.8, have been met.

The staff review of EOI-13, Anticipated Transient Without Scram, disclosed

that it was essentially identical to that of Sequoyah. Therefore, the staff

concludes that the ATWS procedure is acceptable pending the outcome of the

Commission's rulemaking. (See Sequoyah SER Supplement 2, Section 15.2, for

further discussion.)

13.5.3 NUREG-0737 Items

Reporting Safety Valve and Relief Valve Failures and Challanges (II.K.3.3)

In a submittal dated September 14, 1981, the applicant committed to promptly

report failures in primary system safety valves and relief valves and he has

committed to report all challenges to these valves in his annual report. These

schedules are consistent with the requirements of NUREG-0737, and the staff

finds them acceptable.

Report on Outages of Emergency Core Cooling System (II.K.3.17)

In a submittal dated September 14, 1981, the applicant committed to develop and

implement before fuel loading a plan for collecting emergency core cooling

system (ECCS) outage information, including: (1) outage dates and duration of

outage, (2) cause of the outage, (3) ECC systems or components involved in the

outage, and (4) corrective action taken. The staff requires that this information be available for its review. Based on the review of this information, the

applicant should propose changes to improve the availability of ECCS equipment,

if needed. The applicant's proposed plan includes all items required by

NUREG-0737 and its implementation schedule is consistent with the requirements

of NUREG-0737; therefore, the staff finds this acceptable.

13.6 Physical Security Plan

The applicant has submitted documents entitled "Watts Bar Nuclear Plant Physical

Security Plan," "Watts Bar Nuclear Plant Safeguards Contingency Plan," and

"Watts Bar Nuclear Plant Security Personnel-Training and Qualification Plan"

for protection against radiological sabotage. The plans were reviewed in

accordance with Section 13.6, "Physical Security," of the July 1981 edition of

the staff's SRP (NUREG-0800).

As a result of our evaluation, certain portions of these plans were identified

as requiring additional information and upgrading to satisfy the requirements

of Section 73.55 and Appendices B and C of 10 CFR 73.

The applicant has filed revisions to the contingency'plan and the guard training

and qualification plan which satisfied these requirements. The revised plans

are considered to comply with the Commission's regulations in 10CFR 50 and 70

and are, therefore, acceptable. However, the applicant has failed to file

appropriate revisions to the physical security plan. This issue remains an

open item. Based on discussions with the applicant on this matter, the staff

concludes that the physical security plan, when formally modified to incorporate

the required changes, will be acceptable.

Watts Bar SER 13-39
An ongoing.review of the progress of the implementation of these plans will be

performed by the staff to ensure conformance with the performance requirements

of 10 CFR 73.

The identification of vital areas and measures used to control access to these

areas, as described in the physical security plan, may be subject to amendments

in the future.

The applicant's physical security and safeguards contingency plans for the

Watts Bar nuclear plant are being protected from unauthorized disclosure in

accordance with 10 CFR 73.21.

Watts Bar SER 13-40
14 INITIAL TEST PROGRAM

The Watts Bar test program is divided into 10 parts which extend from the point

at which construction progress is sufficient to allow individual component and

system tests through the plant acceptance test at rated power. The listing

'below illustrates the grouping of the various test program parts into categories

of construction tests, preoperational tests, and startup tests.

Test program parts

(1) Installed equipment inspection

program

(2) Cleaning and flushing program

(3) Integrity test program

(4) Equipment checkout, operation,

and adjustment

(5) System functional test program

(6) Core-loading program

(7) Initial criticality and lowpower physics program

(8) At-power physics and core

performance measurements program

(9) Transient test program

(10) Plant acceptance test program

Test program categories

Construction test program

Construction test program

Construction test program

Construction test program

Preoperational test program

Startup test program

Startup test.program

Startup test program

Startup test program

Startup test program

Construction Tests

The adequacy of installation and preliminary operation of systems and components is verified by a construction test program consisting of inspections and

electrical/mechanical tests performed in accordance with written procedures and

instructions. This test program is completed before the preoperational testing

phase and encompasses parts 1, 2, 3, and 4 of the initial test program.

Preoperational Test Program

The preoperational test program ensures that equipment and systems important

to the safety of the plant, employees, and general public perform in accordance

with design criteria to ensure that initial fuel loading, initial criticality,

and subsequent power operation can be safely undertaken. As the installation

of individual components and systems is completed, the necessary tests, adjustWatts Bar SER 14-1
ments, calibrations, and system operations are conducted to verify proper

system performance. The tests are conducted, where practical, following plant

operating and emergency procedures, and thus, serve to verify the correctness

of the procedures.

Whenever possible, these tests are performed under the same conditions as

experienced under subsequent station operations. During system tests for which

unit parameters are not available and cannot be simulated because of existing

plant conditions, the systems are operationally tested as far as possible

without these parameters. The remainder of the tests are performed under plant

conditions when the parameters are available. Abnormal unit conditions are

simulated during testing when required and when such conditions do not endanger

personnel or equipment, or contaminate systems whose cleanliness has been

established.

The preoperational test program includes both safety-related and nonsafetyrelated preoperational tests. The safety-related preoperational tests demonstrate the capability of safety-related structures, systems, and components to

meet performance requirements and to satisfy design criteria. The nonsafetyrelated preoperational tests are conducted on nonsafety-related systems and

components to satisfy reliability and availability. Preoperational tests are

conducted on those systems that

(1) are relied on for safe shutdown and cooldown of the reactor under normal

plant conditions and for maintaining the reactor in a safe condition for

an extended shutdown period

(2) are relied on for safe shutdown and cooldown of the reactor under transient

and postulated accident conditions and for maintaining the reactor in a

safe condition for an extended shutdown period following such conditions

(3) are relied on for ,establishing conformance with safety limits or limiting

conditions for operations that are included in the Technical Specifications

(4) are classified as engineered safety features actuation systems or are

relied on to support or ensure operation of engineered safety features

actuation systems within.design limits

(5) are assumed to function during an accident or for which credit is taken

in the accident analysis

(6) are used to process, store, control, or limit the release of radioactive

materials

Startup Test Program

The startup test program is designed to ensure that the plant actually performs

in accordance with design requirements. This program begins with initial core

loading, includes postloading tests, low-power testing, and power escalation

tests. In addition to confirming design bases, the startup test program will

demonstrate, where practical, that the plant is capable of withstanding the

anticipated transients and postulated accidents. The staff's review of all

phases of the test programs concentrated on the administration of the test

Watts Bar SER 14-2
program and the completeness of the preoperational and startup tests. It

should be noted that the applicant modified his initial Chapter 14 program to

c-losely resemble the testing program previously approved and used for the

Sequoyah Nuclear Station. Differences between the two plants (both technical

and testing) were identified and explained by the applicant, and evaluated

during the staff's review.

The staff's review of Chapter 14 of the FSAR consisted of the following parts.

The Commission's Safety Evaluation Report that was issued at completion of the

construction-permit review was reexamined to determine the principal design

criteria for the plant and to identify any specific concerns or unique design

features that would warrant special test consideration. Chapters 1 through 12

of the FSAR were reviewed for familiarization with the facility design and

nomenclature. Chapter 15 was reviewed to identify assumptions pertaining to

performance characteristics that should be verified by testing and to identify

all structures, systems, components, and design features that were assumed to

function (either explicitly or implicitly) in the accident analysis. Licensee

event report summaries for operating reactors of similar design were reviewed

to identify potentially serious events and chronic or generic problems that

might warrant special test consideration. Standard Technical Specifications

for Westinghouse pressurized water reactors were reviewed to identify all

structures, systems, and components that would be relied on for establishing

conformance with safety limits or limiting conditions for operations. And

finally, the startup test reports for other pressurized water reactor plants

were reviewed to identify problem areas that should be emphasized in the Watts

Bar initial test program.

The object of the staff review of FSAR Chapter 14 is to determine whether the

acceptance criteria stated in the SRP (NUREG-0800) are met. The staff's review

covered several aspects of the initial test program including the following

major considerations:

(1) Test procedures have been developed using input from the NSSS vendor, the

architect-engineer, the applicant's engineering staff, and other equipment

suppliers and contractors as needed. Operating experiences at similar

plants were factored into the development of the test procedures.

(2) Tests are being conducted using approved test procedures and administrative

controls which cover (a) the completion of test prerequisites, (b) the

completion of necessary data sheets and other documentation, and (c) the

review and approval of modifications to test procedures. The applicant

stated that'administrative procedures also cover implementation of modifications or repair requirements identified as being required by the te-sts

and any necessary retesting.

(3) The results of each test are reviewed for technical adequacy and completeness by qualified personnel including the NSSS vendor and-the architectengineer as appropriate. Preoperational test results;are reviewed before

fuel loading, and the startup test results from each activity or power

level will be reviewed before proceeding to the next activity or power

level.

Watts Bar SER 14-3
(4) Normal plant operating and emergency procedures are used in performing

the initial test program, thereby verifying the correctness of the procedures to the extent practical.

(5) The applicant's schedule for conducting the initial test program allowed

adequate time to conduct all preoperational tests and startup tests. The

sequential schedule for performing the startup tests established that

systems required to prevent, limit, or mitigate the consequences of

postulated accidents will be. tested before 30 percent of rated power is

exceeded and that the safety of the plant will not depend on the performance of untested systems structures, and components. Preoperational test

procedures will be available for Regional Administrator review at least

60 days before the expected performance of the test, and startup test

procedures will be available at least 60 days before fuel loading.

(6) Each test procedure abstract presented in Chapter 14 of the Watts Bar FSAR

was reviewed. The staff verified that there are test abstracts for those

structures, systems, components, and design features that correspond to

the criteria for preoperational and startup tests stated previously in

this SER.

(7) The test objectives, prerequisites, test methods, and acceptance criteria

for each test abstract were reviewed in sufficient detail to establish

that the functional adequacy of the structures, systems, components, and

design features will be demonstrated.

(8) The test program's conformance with applicable Regulatory Guides was

reviewed. In addition to Regulatory Guide 1.68 (Revision 0), "Preoperational and Initial Startup Test Programs for Water-Cooled Power Reactors,"

the review included

(a) Regulatory Guide 1.20 (Revision 2), "Comprehensive Vibration Assessment Program for Reactor Internals During Preoperational and Initial

Startup Testing."

(b) Regulatory Guide 1.30 (Revision 0), "Quality Assurance Requirements

for the Installation, Inspection, and Testing of Instrumentation and

Electric Equipment (Safety Guide 30)."

(c) Regulatory Guide 1.37 (Revision 0), "Quality Assurance Requirements

for Cleaning of Fluid Systems and Associated Components of WaterCooled Nuclear Power Plants."

.(d) Regulatory Guide 1.41 (Revision 0), "Preoperational Testing of

Redundant Onsite Electric Power Systems To Verify Proper Load Group

Assignments."

(e) Regulatory Guide 1.52 (Revision 2), "Design, Testing, and Maintenance

Criteria for Engineered-Safety-Feature Atmosphere Cleanup System Air

Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power

Plants."

Watts Bar SER 14-4
(f) Regulatory Guide 1.79 (Revision 1), "Preoperational Testing of

Emergency Core Cooling Systems for Pressurized Water Reactors."

(g) Regulatory Guide 1.80 (Revision 0), "Preoperational Testing of

.Instrument Air Systems.".

(h). Regulatory Guide 1.95 (Revision 1), "Protection of. Nuclear Power

Plant Control Room Operators Against an Accidental Chlorine Release."

(i) Regulatory Guide 1.108 (Revision 1), "Periodic Testing of Diesel

Generators Used as Onsite Electric Power Systems at Nuclear Power

Plants."

(j) Regulatory Guide 1.116 (Revision O-R), "Quality Assurance Requirements'

for Installation, Inspection, and Testing of Mechanical Equipment

and Systems."

(k) Regulatory Guide 1.128 (Revision•0), "Installation Design and Installation of Large Lead Storage Batteries for Nuclear Power Plants."

(1) Regulatory Guide-1.139, "Guidance for Residual Heat Removal (For

Comment)."

(m) Regulatory Guide 1.140, "Design, Testing, and Maintenance Criteria

for Normal Ventilation Exhaust System Air Filtration and Adsorption

Units of Light-Water-Cooled Nuclear Power Plants (For Comment)."

The applicant made a number of changes to the initial test program because of

staff comments. Examples of these changes are as follows:

(1) Testing was added to address the requirements of Task Action Plan

Item I.G.1, "Training During Low Power Testing."

(2) The applicant modified the initial draft test program to nearly replicate

the test program used at the Sequoyah Nuclear Plant. Differences between

the specific Watts Bar and Sequoyah tests were provided by letter from

L. M. Mills to E. Adensam dated November 14, 1981.

(3) For those tests where differences did exist, the applicant modified the

test abstracts to address the following requests:

(a) Verify operation of air relief valves'at design set points.

(b) Include piping vibration testing of the auxiliary air control system.

(c) Provide additional testing of polar crane hoist brakes.-

(d) Supply technical justification for conducting the dynamic automatic

steam dump control test at particular power levels.

(e) Perform rod control tests at hot shutdown instead of cold shutdown

conditions.

Watts Bar SER 14-5
(f) Supply technical justification for integrated leak rate test duration

being 8 hours instead-of 24 hours.

(g) Revise description of criteria used to assess minimum and maximum

relief valve capacity measurement.

Based on its review of the above considerations, the staff has concluded that with the exception of the items noted below, the initial plant test program is

acceptable and meets the requirements of 10 CFR 30.53 with regard to initial

testing of radiation detection and monitoring instruments; 10 CFR 50.34(b) (6)(iii), which requires inclusion of plans for preoperational testing and

initial operations in the FSAR; 10 CFR 50, Appendix B, Section XI, which

requires a test program to ensure that all testing required to demonstrate that

structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures that

incorporate the requirements and acceptance limits contained in applicable ,design documents; and 10 CFR 50, Appendix J, Section III.A.4, which requires a preoperational measurement of the overall integrated. leaktightness of the primary reactor containment building under specified pressure conditions.

(1) FSAR Section 14.2.11.2 currently requires that test procedures be available for Regional Administrator review only 1 month before the test

(preoperational tests) or fuel load (startup tests). By letter dated

February 26, 1982, the applicant committed to revise Section 14.2.11.2 to

state that procedures will be available 60 days before the test (preopera- tional tests) or fuel load (startup tests). The staff considers this item

resolved, pending confirmation that the Watts Bar FSAR is appropriately

amended.

(2) FSAR Table 14.2-3 does not currently reflect the conformance of preopera- tional tests with the followilng applicable Regulatory Guides:

(a) Regulatory Guide 1.20, Revision 2

(b) Regulatory Guide 1.52, Revision 2

(c) Regulatory Guide 1.79, Revision 1

By letter dated June 11, 1982, the applicant committed to revise Section

6.3.4.1 and Table 14.2-3 of the FSAR to reflect compliance with the above

revisions to these Regulatory Guides. The staff considers this item

resolved, pending confirmation that the FSAR is appropriately amended.

(3) The FSAR does not currently include the following systems as part of the

Watts Bar initial test program. These are systems which current NRC staff

positions have determined are important to the safe operation of the

plant.

(a) Condensate system

(b) Condenser circulating water (included in Sequoyah test program)

(c) Hotwell level control system

Watts Bar SER 14-6
(d) Condensate storage tank auxiliaries, including systems used for temperature control of tanks and suction lines and indication and alarm

functions

(e) 48-V dc system

(f) Failed fuel detection system

(g) Chemical addition systems for the secondary plant

(h) Turbine gland sealing system and gland seal water system

(i) Standby lighting system

(j) Ventilation systems for the intake pumping station

(k) Leakage tests of ECC systems and testing of leak detection and

pumping systems provide to control leakage from ECC systems

(1) Turbine building area ventilation system

(m) Raw cooling water system (included in Sequoyah test program)

By letters dated April 21 and May 20, 1982, the applicant committed to include

the following systems in the-preoperational test program:

* failed fuel detection system

* standby lighting system

* ventilation system for the intake pumping station

* leakage tests of ECC systems and testing of leak detection and pumping

systems provided to control leakage from ECC system

The staff concurs with this commitment and will review these test abstracts,

when they are received, and will report its findings in a supplement to the

SER. However, the staff has concluded that the condensate storage tank

auxiliaries and the 48 V dc system also should be included in the initial plant

test program. The staff requires that appropriate abstracts of the test be

included in the FSAR (see letter to Parris dated June 9, 1982). The staff

will review the test abstracts when they are received, and will report its

findings in a supplement to the SER. The remaining seven systems will be

tested under the noncritical Systems Test Program, and will be reviewed for

adequacy by the staff.

Conclusions

The staff has concluded that when the above three items are satisfactorily

resolved, the initial test program described in the application will meet the

acceptance criteria of Section 14.2 of the SRP (NUREG-0800, July 1981) and that

the successful completion of the program will demonstrate the functional adequacy

of plant structures, systems, and components.

This review and evaluation were performed with the assistance of Battelle Pacific

Northwest Laboratories' personnel.

Watts-Bar SER 14-7

15 ACCIDENT ANALYSIS

15.1 General Discussion

The applicant evaluated the ability of Watts-Bar to withstand normal and abnormal transients and a broad spectrum of postulated accidents without undue hazard

to the health and safety of the public. The results of these analyses are used

to show conformance with GDC 10 and 15.

During its review of the transients and accidents analyses of Section 15, the

staff has considered GDC 21, 26, and 28 and Regulatory Guides 1.53 and 1.105 as

they apply to the events analyzed to ensure that the applicable requirements

have been met.

For each event analyzed the worst operating conditions were assumed, and credit

was taken for minimum engineered safeguards response. Parameters specific to

individual events were conservatively selected.

Event scenarios were slected consistent with the guidance provided by BTPs

ASB 3-1 and MEB 3-1. Performance of the auxiliary feedwater system was assumed

to be consistent with the requirements of Items II.E.1.1 and II.E.1.2 of

NUREG-0737 (see Sections 10.4.9 and 7.8.2, respectively).

Two types of events were analyzed:

(1) those incidents which might be expected to occur during the lifetime of

the reactor (anticipated transients)

(2) those incidents not expected to occur which have potential to result in

significant radioactive material release (accidents)

15.2 Normal Operation and Anticipated Transients

The applicant has analyzed several events expected to occur one or more times

in the life of the plant. The acceptance criteria are the acceptable fuel

design limit (DNBR remains greater than 1.30 using the W-3 correlation) and the

reactor coolant pressure stays below 110 percent of design.

The transients analyzed are protected by the following reactor trips in accordance with GDC 20.

(1) power range high neutron flux

(2) high pressure

(3) low pressure

(4) overpower WT

(5) overtemperature WT

Watts Bar SER 15-1
(6) low coolant flow

(7) pump undervoltage/underfrequency

(8) low steam generator level

(9) high steam generator water level

Time delays to trip, calculated for each trip signal, are included in the

analyses.

The nuclear feedback coefficients were conservatively chosen to produce the

most adverse core response. :The reactivity insertion curve used to represent

the control insertion accounts for a stuck rod'in accordance with GDC 26.

The flow coastdown code PHOENIX has been found satisfactory for evaluating

transients and accidents in the Watts Bar plant. Staff reviews of the FACTRAN,

BLKOUT, MARVEL, and LOFTRAN codes have progressed to the point that there is

reasonable assurance that analyses results dependent on the codes will not be

appreciably altered by any-method revisions that may be required by the staff.

Transients and accidents were analyzed for Watts Bar using a procedure for

evaluating fuel performance,jwhich conservatively bounds the consequences of

the event by accounting for fabrication and operating uncertainties directly in

the calculations. DNBRs were calculated using the W-3 correlation, with a

minimum DNBR of 1.3 used as a design limit.

The applicant accounts for errors in initial conditions by making the following

assumptions as appropriate for the event being considered:

(1) core power, 3411, +2% MW

(2) average reactor coolant system temperature (TAV), 572.2, +6.50 F

(3) pressure (at pressurizer), 2250, +30 psi

These assumptions for initial condition are acceptable because they are conservatively applied to produce the most adverse effects.

Analyses of abnormal transients have been submitted and show that the integrity

of the RCS boundary is maintained and that the minimum DNBR is above the applicable limit. These transients can be classified as reactivity insertions,

loss-of-cooling, increased cooling, and change in coolant inventory.

15.2.1 Loss-of-Cooling Transients

The applicant has analyzed the following events which result in loss of cooling:

(1) loss of reactor coolant flow

(2) loss of external load, turbine trip

(3) loss of normal feedwater flow

(4) loss of nonemergency power to the station auxiliaries

The transients that produce the most severe effects are the complete loss of

flow and loss of external load; the former has the greatest effect on departure

from nucleate boiling ratio (MDNBR = 1.31) and the latter on system pressure

(highest system pressure approximately 2570 psia). The staff has reviewed the

Watts Bar SER .15-2
results of these analyses and finds them acceptable because the fuel design

limits and primary systems pressure limits are not violated in any of these

transients.

15.2.2 Increased Cooling Transients

The applicant has analyzed the following events which produced increased

primary system cooling:

(1) decrease in feedwater temperature

(2) increase in feedwater flow

(3) excessive load increase

(4) opening of steam generator relief valve

Although none of these transients is limiting, the most sever( i terms of DNBR

and system pressure are the excessive load increase events. Giiy slight changes

in primary system pressure were calculated, and the DNBR falls to about 1.33.

The staff finds these results acceptable because they do not violate the appropriate limits.

15.2.3 Change in Coolant Inventory Transients

The applicant has analyzed the following events which result in changes in the

primary system inventory.

(1) opening of pressurizer safety/relief valve

(2) actuation of emergency core cooling system

The results calculated for opening of a safety/relief valve indicate that the

system pressure declines slowly until overtemperature AT trip occurs. Departure from nucleate boiling ratio for this event remains above 1.30 and is not

limiting. These results-are consistent with those generically reported in

WCAP-9600 in which core uncovery does not occur.

Emergency core cooling system (ECCS) operation could be initiated by a spurious

signal or operator error. Two cases were examined--one in which reactor trip

occurs simultaneously with the safety injection signal--the other in which

reactor trip occurs later in the transient. The reactor pressure never exceeds

initial values and the DNBR never drops below its initial value for this

transient.

The staff finds the results of these transients acceptable because the fuel

damage limits and the primary system pressure limits are not violated.

15.2.4 Reactivity and Power Distribution Anomalies

15.2.4.1 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal from Zero

Power Conditions

The consequences of an uncontrolled rod cluster control assembly bank withdrawal at zero power have been analyzed. Such a transient can be caused by a

failure of the reactor control or rod control systems. The analysis assumes a

conservatively small (in absolute magnitude) negative Doppler coefficient and a

Watts Bar SER 15-3
positive moderator coefficient. Further, hot zero power initial conditions

with the reactor just critical are chosen because they are known to maximize

the calculated consequences. The reactivity insertion rate is assumed to be

equivalent to the simultaneous withdrawal of the two highest worth banks at

maximum speed (45 in. per min).

Reactor trip is assumed to occur on the low setting of the power range neutron

flux channel at 35 percent of full power (a 10 percent uncertainty has been

added to the set point value)'. The maximum heat flux is less than the full

power value and average fuel temperature' rises only to about 8500 F, much less

than that at full power operation.

The staff has reviewed the reactivity worths and reactivity coefficients used

in the analysis and concludes that conservative values have been used. The

staff also has reviewed the calculated consequences of this design transient

and conclude that they are acceptable.

15.2.4.2 Uncontrolled Rod Cluster Control Assembly Bank WithL awal at Power

The consequences of uncontrolled withdrawal of a rod bank in the power operating range have been analyzed. The effect of such an event is an increase in

coolant temperature (as a result of the core-turbine power mismatch) which must

be terminated before the fuel design limits are exceeded.

The analysis is performed as a function of reactivity insertion rates, reactivity feedback coefficients, and core power level. Protection is provided

by the high neutron flux trip and the overtemperature-W temperature trip.

In no case does the DNBR fall below 1.30. Adequate fuel cooling is therefore

maintained. The maximum power reached including uncertainties is 118 percent

of full power, thus precluding fuel centerline melting.

On the bases that the analytical methods are acceptable and input data are

reasonably conservative and that established fuel damage limits are not exceeded, the staff concludes that the analysis and calculated consequences of this

design transient are acceptable.

15.2.4.3- Rod Cluster Control Assembly Malfunctions

Rod cluster control assembly misalignment incidents including a dropped full

length assembly, a dropped full-length bank, a misaligned full length assembly

and the withdrawal of a single assembly during power operation power have been

analyzed by the applicant. Misaligned rods are detectable by: (1) asymmetric

power distributions sensed by excore nuclear instrumentation or core exit

thermocouples, (2) rod deviation alarm, and (3) rod position indicators. A

deviation of a rod from its bank by about 15 in., or twice the resolution of

the rod position indicator, will not cause power distributions to exceed design

limits. Additional surveillance will be required to ensure rod alignment if

one or more rod position channels are out of service.

In the event of a dropped assembly, the automatic controller may return the

reactor to full power. Analysis indicates that DNB will not occur during this

event. For the case of dropped groups, the reactor is tripped by the power

Watts Bar SER 15-4
range negative neutron flux rate trip and the reactor is shut down without

violating fuel damage limits.

For. cases where a group is inserted to its insertion limit with a single rod

in the group stuck in the fully withdrawn position, analysis indicates that

departure from nucleate-boiling will not occur. The staff has reviewed the

calculated estimates of the reactivity and power distribution changes that may

accompany postulated misalignments of representative assemblies. The staff

has concluded that the values used in this analysis conservatively bound the

expected values, including calculational uncertainties.

The inadvertent withdrawal of a single assembly requires multiple failures in

the control system, multiple operator errors or deliberate operator actions

combined with a single failure of the control system. As a result, the single

assembly withdrawal is classified as an infrequent occurrence. The resulting

transient is similar to that resulting from bank withdrawal, but the increased

peaking factor may cause DNB to occur in the region surrounding the withdrawn

assembly. Less than 5 percent of the rods in the core experience DNB for such

a transient. The staff concludes that the analysis and calculated consequences

of this event are acceptable.

15.2.4.4 Inadvertent Boron Dilution

Pressurized water reactors use soluble boron compounds to control core reactivity. The staff is concerned that accidental dilution during power operation,

refueling, startup, or shutdown could result in an uncontrolled increase in

the core reactivity. The applicant has identified alarms that would alert the

operator to a boron dilution event in time to allow the operator appropriate

response times per SRP Section 15.4.6 for all modes of operation. A particular

concern was a postulated boron dilution event while the plant was being shut

down. During this period, and while refueling, valves are administratively

locked closed to isolate all known sources of unborated water. At this time

the applicant relies on an alarm from the source range neutron flux detector.

The applicant provided justification for maintaining the alarm set points within 1/2 decade of the source flux level. Based on this margin and on the maximum

possible rates of dilution, the applicant's analysis showed that the event would

be detected and announced by the high flux at the shutdown alarm within a time

period that left sufficient margin for the operator to correct the situation

before criticality occurred. Fifteen minutes is the required minimum time

margin at these conditions in accordance with the SRP.

The applicant has committed to a schedule for setting and monitoring the gap

between the high flux at the shutdown alarm level and the shutdown source flux

level that is consistent with the analysis presented. The setting is to. be no

higher than 1/2 decade above the count rate, and the margin is to be verified

(or reset if necessary) every 30 minutes for the first 2 hours, every 2 hours

for the next 6 hours, and once per shift thereafter until the flux level has

stablilized. The required procedures and schedule for verification of the set

point are to be incorporated in the operator's surveillance instructions. The

staff finds that the procedures and analyses provided by the applicant acceptably justify the use of a high flux at the shutdown alarm to alert the operator

to a boron dilution event during shutdown and refueling on a timely basis.

However, the staff concludes that there is insufficient information for it to

Watts Bar SER 15-5
judge that the criteria for single failure are met. The staff requires the

applicant to resolve this issue before fuel load by providing an evaluation of

the boron dilution event that shows how the single-failure criterion is met.,

The staff finds that in theanalysis, the reactivity changes in the boron

dilution event are accounted for satisfactorily. The applicant's analysis defines a region of reactor conditions for the event that are considered safe, according to NRC criteria as described in SRP Section 15.4.6, and, with resolution of the staff's concern about meeting the single-failure criterion, the

applicant's boron dilution event analysis can be acceptable.

15.2.4.5 Inadvertent Loading of a Fuel Assembly into the Improper Position

Strict administrative controls in the form of previously approved established

procedures and startup testing are followed during fuel loadings to prevent operation with a fuel assembly in an improper location or a misloaded burnable

poison assembly. Nevertheless, an analysis of the consequences of a loading

error has been performed.

Comparisons of power distributions calculated for the nominal fuel loading pattern and those calculated for five loadings with misplaced fuel assemblies

or burnable poison assemblies are presented by the applicant. The selected

nonnormal loadings represent the spectrum of potential inadvertent fuel misplacement. Calculations included, in particular, thelpower in assemblies that

contain provisions for monitoring with incore detectors.

As part of.the required startup testing, the incore detector system is used to detect misloaded fuel before operating at power. The analysis described above

shows that all but one of the above misloading events would be detected by this

test. In the excepted case--an interchange of region 1 and 2 assemblies near

the center of the core--the increase in the power peaking is approximately equal to the uncertainty in the measurement of this quantity (approximately 8

percent). This uncertainty is allowed for in analyses so that this misloading

event does not result in unacceptable consequences.

Based on its review of the analyses described above, the staff concludes that

an improperly loaded fuel assembly or burnable poison assembly that could cause

a significant safety problem would be detected by the operator with the instrumentation provided.

15.2.4.6 Rod Cluster Control Assembly .Ejection

The mechanical failure of a control rod mechanism pressure housing would result

in the ejection of a rod cluster control assembly. For assemblies initially' inserted, the consequences would be a rapid reactivity insertion together with

an adverse core power distribution, possibly leading to localized fuel rod

damage. Although mechanical provisions have been made to make this accident

extremely unlikely, the applicant has analyzed the consequences of such an

event.

Methods used in the analysis are reported in WCAP-7588, Revision 1, "An Evaluation of the Rod Ejection Accident in Westinghouse Reactors Using Spatial

Watts Bar SER 15-6
Kinetics Methods," which has been reviewed and accepted by the staff.* This

report demonstrates that the model used in the accident analysis is conservative relative to a three-dimensional kinetics calculation.

The applicant's criteria for gross damage of fuel are a maximum clad temperature of 2700'F and-an energy deposition of 200 calories per gram in the hottest pellet. These criteria are more conservative** than those proposed in

Regulatory Guide 1.77. Therefore they are acceptable.

Four cases were analyzed: beginning-of-cycle at 102 percent and zero power

and end-of-cycle at 102 percent and zero power.' The highest clad temperature,

2653°F, was reached in the zero power end-of-cycle case and the highest fuel

enthalpy, 160 calories per gram, was reached in the end-of-cycle full power

case. The analysis also shows that less than 10 percent of the fuel experiences DNB and less than 10 percent of the hot pellet melts. Analyses have

been performed to show that the pressure surge produced by the rod ejection

is mild and will not approach the reactor coolant system emergency limits.

Further analyses have shown that a cascade effect (that is, the ejection of

a further rod as a result of the ejection of the fi.rst one) is not credible.

The ejected rod worths and reactivity coefficients used in the analysis have

been reviewed and have been judged to be conservative. Also the assumptions

and methods of analysis used by the applicant are in accordance with or are

more conservative than those recommended in Regulatory Guide 1.77. Therefore,

the staff concludes that this analysis is acceptable. The analysis of the

doseconsequences for this design-basis accident is described in Section 15.4

of this SER.

15.2.5 Conclusions

Based on its review of the transients discussed above, the staff concludes

that the analyses demonstrate the operation of the plant will not result in

any violation of fuel design or reactor coolant pressure design limits, conform with GDC 10 and 15, and are acceptable with the exception of the issue

regarding an inadvertent boron dilution event. This will be discussed in a

supplement to the SER.

15.3 Limiting Accidents

There are a number of events and postulated accidents which are not expected

during the life of the plant. Within this group are events which form the

design basis for the various barriers and safety systems. The acceptance

criteria for accidents are provided in 10 CFR 50 and 10 CFR 100.

In the analysis of these events,.the applicant must investigate a broad spectrum of related events to determine the bounding case, including the worst

*Acceptance letter dated August 28, 1973 (See NUREG-0390).

**Regulatory Guide 1.77 has an acceptance criterion of 280 calories per gram

energy deposition and no criterion for clad temperature other than that'

implicit in requirements for fuel and pressure vessel damage.

Watts Bar SER 15-7
single active failure. Sensitivity studies are performed to identify para- meters for initial conditions and appropriate credit for systems and their

performance during the limiting events in terms of protection of various

barriers.

15.3.1 Loss-of-Coolant Accident

The applicant has provided LOCA analyses to demonstrate conformance with the

requirements of 10 CFR 50.46 for (ECCSs). These analyses identify the limiting location, type, discharge coefficient, and size break for Watts Bar Units 1 and

2. Additional analyses confirm conservatism in inputs and demonstrate sensitivity of calculated peak cladding temperature to uncertainity in input parameters.

The applicant has cited spectrum analyses performed for the Floating Nuclear

Plant and McGuire to show the most limiting break location to be in the pump discharge line. The Floating Nuclear Plant and McGuire analyses include the

effect of upper head injection (UHI). The staff finds these analyses to be an

acceptable reference to determine worst-break location.

The applicant performed a break spectrum study for large ruptures in reactor

coolant pump discharge piping. An evaluation model that accounts for the UHI

(WCAP-8749) and an appropriate metal/water reaction model were used in conformance with 10 CFR 50, Appendix K. The applicant presented its input values

for a number of primary system initial conditions and ECCS parameters. These

parameters were indicated as maximum, minimum, or nominal values expected during plant operation. The net effect of these input parameters were stated

to be conservative for the LOCA analyses.

The analyses were performed with an assumed containment backpressure which has

been reviewed and found acceptable. This is discussed further in Section 6.2.1.

The study identified large guillotine breaks to be more limiting than large split breaks with perfect mixing and showed that the most limiting break is a

double-ended coefficient guillotine rupture in pump discharge piping with a discharge of 0.6 (DECLG, Cd=O.6). Guillotine ruptures were analyzed with

both perfect and imperfect mixing as required by the staff in its approval of

the model.

The applicant has submitted analyses for a spectrum of small break LOCA analyses (4 in., 6 in., 8 in.). These analyses show that the 6-in. break is the

limiting small break, the calculated peak cladding temperature is 1435'F, the

local metal water reaction is 0.392 percent, and the core wide oxidation is

less than 0.3 percent. The calculated peak cladding temperatures for these

small breaks are far below those for large breaks and clearly are not limiting.

The break spectrum study has indicated that for the Watts Bar reactor the worst

break is the double-ended cold leg guillotine rupture, calculated by assuming perfect mixing and a coefficient of discharge equal to 0.6 (DECLG, Cd=O.6,

perfect mixing). The calculated peak cladding temperature for this case,

2095°F; the calculated maximum local metal/water reaction, 4.7 percent; and the

calculated total core-wide metal/water reaction of less than 0.3 percent are

Watts Bar SER 15-8
below the allowable limits specified in 10 CFR 50.46 (b) of 2200*F, 17 percent,

and I percent, respectively. The analyses were performed based on a total

peaking factor of 2.32 (a peak linear power of 12.63 kilowatts per foot) at 102

percent of the rated core power level of 3411 megawatts thermal (MWt).

The applicant has provided additional analyses and information to satisfy the

plant-specific conditions specified in the staff approval of the UHI evaluation

model.

The minimum UHI accumulator pressure is higher than the saturation pressure

based on the maximum allowable upper head temperature. Therefore, UHI will

occur before flashing in the upper head as required by the staff's acceptance

of the evaluation model.

In its analyses, the applicant assumed that the initial upper head temperature

was equal to the cold leg temperature. Westinghouse has performed scale model

tests simulating a reactor upper head region with UHI hardware. These tests

have shown that 4 percent bypass blow into the upper head is sufficient to

maintain the temperature in this region at the cold leg temperature. Verificational data taken at McGuire provide additional assurance that cold leg temperature will be achieved and, therefore, the staff finds this upper head temperature assumption acceptable. The applicant has referenced a sensitivity analysis to determine the effect of cold leg accumulator uncertainties on calculated

peak cladding temperatures assuming imperfect mixing and accounting for cold

leg accumulator uncertainities that are bounded by the worst-break calculation

(DECLG, Cd=O.6, perfect mixing). The applicant has assumed that the most

severe active single failure for the large break LOCA analyses is the loss of a

residual heat removal pump. Potential consequences of losing an ESF train were

assessed. It was concluded that the loss of an ESF train was less limiting

because of the benefits of increased containment pressure with only one containment spray train available.

A sensitivity study performed for the Sequoyah Nuclear Plant (see Sequoyah

FSAR) demonstrates the conservatism in the assumption of loss-of-offsite power.

Based on applicable plant design similarities and the large calculated margin,

the staff concludes that the assumption of loss-of-offsite power is similarly

conservative for the Watts Bar LOCA analyses.

The applicant has provisions for maintaining long-term cooling of the core.

The LOCA analyses presented show that the peak clad temperatures do not exceed

the allowable limit and that clad temperature trends, which include effects of

rod ballooning, and available long-term cooling demonstrate that a coolable

core geometry will be maintained as required by 10 CFR 50.46(b).

To assess the impact of cladding swelling and rupture, the applicant submitted,

in a September 22, 1981 letter, an analysis performed in accordance with the

interim model requirements described in NUREG-0630. This method is acceptable

for Watts Bar to meet 10 CFR 50.46(b) and 10 CFR 50, Appendix K, requirements.

The analysis determined a revised peaking factor (FQ) of 2.31 to account for

the impact of cladding swelling and rupture.

Based on this review and other SER sections describing the staff review of the

ECCS for Watts Bar the staff concludes that the ECCS performance conforms to

Watts Bar SER 15-9
the acceptance criteria of 10 CFR 50.46, NUREG-0630, NUREG-0297, and Appendix K

to 10 CFR 50.

15.3.2 Steamline Break

To demonstrate sensitivities to certain parameters, to show the influence of

UHI, and to show compliance with performance criteria, the applicant has

provided analyses of various steamline break events for Watts Bar.

The system code used to perform these analyses is MARVEL (discussed in

Section 15.2 of this SER).

The applicant's analyses indicate that the worst steamline break event is a

steamline rupture inside the main steam isolation valve of the affected steam

generator.

For this worst break, onsite power was assumed available and the worst single

failure was identified to be loss of one train of ECCS. Calculations were

performed using the stuck rod assumption considering core properties associated

with the sector nearest the affected steam generator. Credit was taken for

operator action at 10 min to throttle or terminate injection flow and to

isolate the affected steam generator. Results for this depressurization event

show no violation of pressure criteria. Although a return to criticality

occurs in the core (maximum linear heat generation rate less than 14kW/ft),

there is no core uncovery or fuel damage (DNBR greater than 1.3 using the W-3

correlation).

The identification of worst case conditions and trends in results calculated

for Watts Bar are consistent with those previously reported generically

(WCAP-9226) and under staff review. Based on its review, the staff finds the

steamline break analyses for Watts Bar and the results acceptable, because the

criteria identified in Section 15.3 are met.

In response to the staff's questions on vessel integrity following Section 15

events, the applicant has referenced the generic work being performed by the

Westinghouse Owners Group in addressing Generic Task A-49, Pressurized Thermal

Shock. The staff requires that the applicant participate (by providing analyses, commitments, and procedures) in the resolution of Task A-49, consistent

with the requirements and schedules of that generic task. Task A-49 is discussed in Appendix C. Subject to this generic resolution, the staff finds the

applicant's response acceptable.

15.3.3 Feedwater System Pipe Break

The applicant has provided a feedwater line break analysis for Watts Bar using

assumptions that would minimize secondary system heat removal capability,

maximize heat addition to the primary system coolant, and maximize the calculated primary system pressure.

The system code used to perform these analyses is MARVEL (discussed in SER

Section 15.2). The analysis assumed the most restrictive single failure of the

AFWS, a loss of the motor-driven AFW pump feeding the two intact steam generators. Operator action is assumed at 10 min to isolate the affected steam

Watts Bar SER 15-10
generator and direct AFWtflow to the intact steam generator. The acceptability

of this action is reviewed as part of NUREG-0737 Item II.E.1.1.

Water relief from primary system safety valves was calculated to occur for 1200

seconds, beginning at 825 seconds. The water relief flowrate was calculated

using the conservative homogeneous equilibrium flow model for saturated water.

No fuel damage was calculated to occur (core remained covered) and the peak

calculated RCS pressure was about 2570 psia, which is below the 110 percent

ASME Code limit of 2750 psia. The staff finds these results acceptable.

15.3.4 Reactor Coolant Pump Rotor Seizure

For the reactor coolant pump rotor seizure event, two cases were analyzed:

four loops operating, one locked rotor; and three loops operating, one locked

rotor. Codes used to perform these analyses are LOFTRAN, FACTRAN, and PHOENIX,

which are discussed in Section 15.2. From these analyses, the applicant has

determin ed that less than 10 percent of the fuel rods would fail. The maximum

calculat ed system pressure (2669 psia) was for the three loop case; the

maximum calculated clad temperature (1781*F) occurred in the case in which all

four loops were operating. The applicant has addressed the consequences of a

locked rotor event assuming a loss of offsite power by reanalyzing the event

assuming a 2-sec-delay from the time of reactor trip to the loss of offsite

power. This delay assumption has been reviewed by the staff and found acceptable. The results of the reanalysis indicate that the assumption of a delay in

the loss of offsite power (3.05 sec after event initiation) would not influence

the results discussed above. The calculational results meet the requirements

(applicable parts of GDC 27, 28, and 31, and 10 CFR 100) stated in NUREG-0800,

Section 15.3.3, and are acceptable.:

15.3.5 Reactor Coolant Pump Shaft Break

The applicant stated that the results of this accident are almost identical to

those for a rotor seizure. Our experience in other Westinghouse design reviews

indicates that the calculated consequences can be slightly worse. The applicant has reported that a shaft break event for Watts Bar would result in a 20

psi higher pressure (2689 psia) and 100 F higher temperature (1791°F) than a

locked rotor event, and the number of failed fuel rods would be less than 10

percent. The applicant also has indicated that the shaft break event would not

be influenced by a loss of offsite power~for the:same reasons as the locked

rotor event. The staff finds the shaft break analysis for Watts Bar acceptable

since the results meet the criteria (applicable parts of GDC 27, 28, and 31,

and 10 CFR 100) stated in NUREG-0800, Section 15.3.4.

15.3.6 Anticipated Transients Without Scram

A number of plant transients can be affected by a failure of the scram system

to function. For a PWR, the most important transients affected include loss of

normal feedwater, loss of electrical load, inadvertent control rod withdrawal,

and loss of normal' electrical power. In September 1973, the staff issued

WASH-1270, "Technical Report on Anticipated Transients Without Scram for

Water-Cooled Power Reactors," establishing acceptance criteria for anticipated

transients without scram (ATWS). In conformance with the requirements of

Watts Bar SER 15-11
Appendix A to WASH-1270, Westinghouse submitted an ATWS evaluation a in Topical

Report WCAP-8330, "Westinghouse Anticipated Transients Without Trip Analysis."

A reevaluation of the potential risks from ATWS has been published in

NUREG-0460, Volumes 1 through 4. In March 1980, the staff issued Volume 4 of

NUREG-0460, and in September 1980, the staff presented its recommendation on

ATWS to the Commission. If rulemaking is initiated by the Commission, the

applicant would then be required to modify Watts Bar to conform with ATWS

criteria and schedule requirements provided in the rule or as adopted by the

Commission. The following discussion presents the bases for operation of Watts

Bar before the adoption of a rule.

NUREG-0460 Volume 3 states: "The staff has maintained since 1973 (for example,

see pages 69 and 70 of WASH-1270) and reaffirms today that the present likelihood of severe consequences arising from an ATWS event is acceptably small and

presently there is no undue risk to the public from ATWS. This conclusion is

based on engineering judgment in view of: (a) the estimated arrival-rate of

anticipated transients with potentially severe consequences in the event of

scram failure; (b) the favorable operating experience with current scram

systems; and (c) the limited number of operating factors." In view of these

considerations and the staff's expectation that the necessary plant modifications will be implemented in 1 to 4 years following a Commission decision on

ATWS, the staff has generally concluded that PWR plants can continue to operate

because the risk from ATWS in this period is acceptably small. To further

reduce the risk from ATWS during the period before completion of the plant

modification determined to be necessary by the Commission, the staff has

required that applicant take the following steps:

(1) Develop emergency procedures to train operators to recognize ATWS events,

including consideration of scram indicators, rod position indicators, flux

valve and safety indicators, and any other alarms annunicated in the,

control room, with-emphasis on alarms not processed through the electrical

portion of the reactor scram system. The acceptability of these proce-.

dures is discussed in Section 13.

(2) Train operators to take actions in the event of an ATWS, including con-.

sideration of manually scramming the reactor by using the manual scram

button, prompt actuation of the auxiliary feedwater system to ensure

delivery to the full capacity of this system, and initiation of boration

by actuating the high-pressure,.safety-injection system to bring the

facility to a safe-shutdown condition. The acceptability of this training

is discussed in Section 13.

15.3.7 Conclusion

The applicant has presented results for various accidents which meet the staff

acceptance criteria as detailed in SRP Section 15. Therefore, the staff concludes that the applicant has provided adequate protection systems to mitigate

accidents in compliance with GDC 10, 15, and 20 and 10 CFR 50 and 100.

Watts Bar SER 15-12
15.4 Radiological Consequences of Accidents

The postulated design basis accidents analyzed by the applicant to determine

the offsite radiological consequences are the same as those analyzed for previously licensed PWRs. To evaluate the effectiveness of the engineered safety

features proposed for Watts Bar and to ensure that the radiological consequences of these accidents meet the applicable dose criteria, the staff has

analyzed the LOCA, the steamline break accident, the steam generator tube

rupture accident, the control rod ejection accident, the fuel-handling accident

and small-line failures. The calculated doses for these accidents are shown in

Table 15.1.

The short-term diffusion estimates (x/Q) used in the analyses are those presented and discussed in SER Section 2.3.4. The meteorological models described

in the Regulatory Guides referenced in these analyses are modified by those

presented in Regulatory Guide 1.145. See Section 2.3.4 for further discussion.

The review for this section was performed using the criteria set forth in,

NUREG-75/087, Revision 2. However, the acceptance criteria of the SRP,

NUREG-0800, for this section has not significantly changed; therefore, the

findings in Section 15.4 are in accordance with the acceptance criteria of the

SRP of NUREG-0800.

15.4.1 Loss-of-Coolant Accident

The applicant has selected and analyzed a hypothetical design basis LOCA and

has shown that the distances to the exclusion area boundaries and to the low

population zone boundaries are sufficient to provide reasonable assurance that

the radiological consequences of such an accident are within guidelines in

10 CFR 100.1 (a)(1) and (2). The analysis has included the following sources

and radioactivity transport paths to the atmosphere:

(1) contribution from containment leakage

(2) contribution from post-LOCA leakage from ESF systems outside containment

(3) contribution from containment purge

The staff's review confirms the applicant's findings based on the following:

(1) The applicant's provisions for a design of the containment system, the

emergency gas treatment system (EGTS) and the auxiliary building gas

treatment system (ABGTS) are acceptable as identified in Section 6 of this

report.

(2) The staff's independent analysis of the radiological consequences of a

hypothetical design-basis LOCA as described below.

Staff'Evaluation

(1). Containment Leakage Contribution

The Watts Bar plant includes a double containment design to collect and

filter the leakage of fission products from a postulated design-basis

Watts Bar SER 15-13
Table 15.1 Radiological consequences of design-basis accidents

Exclusion area Low

boundary, rems population zone, rems

Postulated accident thyroid whole body thyroid whole body

Loss of coolant

Containment leakage

0-2 hr 3.1 1.8 0.43 0.25

2-8 hr 0.56 0.22

8-24 hr 0.41 0.18

24-96 hr 1.92 0.15

96-720 hr 1.44 0.06

Total containment leakage 3.1 1.8 4.8 .86

ECCS component leakage 5.4 0.01 9.1 0.01

TOTAL LOCA 8.5 1.8 13.9 0.9

Steamline break outside

secondary containment

Long-term operation case

(Case 2) 7.1 (0.1 5.6 <0.1

Short-term operation case

(Case 3) 8.9 <0.1 3.8 <0.1

Control rod ejection

Containment leakage pathway 35.0 <0.9 42.0 0.2

Secondary system release pathway 12.0 <1.0 3.0 <1.0

Fuel-handling accident

In fuel-handling area 1.0 0.3 0.1 <0.1

Inside primary containment 5.1 0.3 0.7 <0.1

Small line break outside

.containment 17.0 <0.1 2.3 <0.1

Steam generator tube rupture

Case 1 (DEI-131 at 60 pCi/gram) 73 <0.1 12.0 <0.1

Case 2 (DEI-131 at 1 pCi/gram) 13 <0.1 3.0 <0.1

Watts Bar SER 15-14
LOCA. The double containment consists of a free-standing steel primary containment vessel surrounded by a reinforced concrete shield building. The reinforced concrete auxiliary building is also a part of the secondary containment barrier. Leakage which enters the secondary containment is

treated by either the EGTS or the ABGTS before its release to the atmosphere. Both of these systems are engineered safety features (ESF) (see SER Section 6.5.1). Another ESF is the ice condenser with sodium tetraborate added to the ice to enhance the removal of iodine in the containment

following a LOCA (see SER Section 6.5.4). The principal assumptions employed in the staff analysis are listed in Table 15.2. The dose model

and dose conversion parameters are consistent with those given in Regulatory Guide 1.4, "Assumptions Used for Evaluating the Potential Radiologi- cal Consequences of a Loss-of-Coolant Accident for Pressurized Water

Reactors."

In the analysis of the design basis LOCA, the primary containment was

assumed to leak at the design leak rate of 0.25 percent per day for the

first 24 hours following the accident and at 0.125 percent per day thereafter. The applicant established to the satisfaction of the staff that

that the shield building annulus and auxiliary building would not experience a period of positive pressure and, hence, no exfiltration. Therefore, no leakage was assumed to bypass the gas treatment systems throughout the course of the accident.

Ten percent of the leakage from the primary containment enters the auxiliary building following the accident, and the staff assumed that this

leakage was exhausted to the atmosphere through the ABGTS without credit

for holdup or mixing in the auxiliary building. Ninety percent of the

leakage from the primary containment enters the shield building annulus

where the staff assumed that it went directly to the intake of, the shield building annulus recirculation/exhaust system. Following passage through the emergency gas treatment system filters, a fraction of this leakage was

assumed to be exhausted to the atmosphere, with the remainder recirculated

to the shield building annulus where credit was given for mixing in

50 percent of the annulus free volume. The split between the exhaust and

recirculation fractions was assumed to be proportional to the flow rates

in the exhaust and recirculation paths of the systems.

(2) Post-LOCA Leakage from ESF System Outside Containment

As part of the LOCA, the staff has also evaluated the consequences of

leakage of containment sump water that is circulated by the ECCS after

that postulated accident. During the recirculation mode of operation the

sump water is circulated outside containment to the .auxiliary building. If a leak should develop, such as a pump'seal failure, a fraction of the

iodine in the water could become airborne in the auxiliary building and

exit to the atmosphere. For Watts Bar, the ECCS area in the auxiliary building is served by an ESF air filtration system (the ABGTS). Therefore, doses from passive failures were not considered as specified in SRP

Section 15.6.5, Appendix B.

Watts Bar SER 15-15
Table 15.2 Assumptions used in the calculation of the

radiolgical consequences following a

postulated loss-of-coolant accident

Power level (MWt) 3592

Operating time (yr) 3

Fractions of-core inventory available for leakage

(%)

Iodines 25

Noble gases 100

Initial iodine composition in containment (%)

Elemental 91

Organic 4

Particulate 5

Primary containment volumes (ft3 )

Upper compartment 6.51 E5

Lower compartment (including ice condenser) 5.85 E5

Shield building annulus volume (ft 3 ) 3.75 E5

Mixing fraction in annulus (%) 50

Annulus ventilation flow distribution (ft 3 )

Time step Recirculation flow Exhaust flow

ft 3 /min ft 3 /min

0-30 sec 0 0

30-180 sec 770 3230

180-360 sec 1700 2300

360-600 sec 2500 1500

600-1200 sec 3340. 660

1200-1800 sec 3810 190

1800 sec-30 days 3900 ý100

Filter efficiencies (%)

Elemental iodine 99

Organic iodine 95

Particulate iodine 99

Ice condenser removal efficiency (%)

Elemental iodine 30

Flowrate through ice condenser (ft 3 ) 40,000

Period of ice condenser effectiveness (min) 10-60

Primary containment leak rates (%)

0-24 hr 0.25

24-hr-30 days

Bypass leakage fraction (%) 0

Minimum excl'usion area boundary distance (m) 1200

Low population zone distance (m) 4828

Atmospheric diffusion (x/Q) values (sec/m 3 )

0-2 hr 3.6 E-4

0-8 hr 5.0 E-5

8-24 hr 3.3 E-5

1-4 days 1.3 E-5

4-30 days 3.7 E-6

Watts Bar SER 15-16
The staff evaluated the potential radiological consequences from normal

ECCS component leakage by assuming a total ECCS component leak rate of

1 gpm. The resulting radiological consequences were only 5 rems to the

thyroid at the exclusion area boundary and 9 rems to the thyroid at the

LPZ.

(3) Hydrogen Purge

The applicant has provided redundant hydrogen recombiners for the purpose of controlling any accumulation of hydrogen within the primary containment

following a LOCA. In the event of failure of both recombiners, the

applicant has provided a backup system. The purged containment effluent

would flow to the shield building annulus where it would be subsequently

discharged to the atmosphere through the EGTS filters. The staff finds

the combination of redundant recombiners plus a backup purge capability to

be an acceptable method for controlling the potential contribution to the

offsite doses from hydrogen purging following a LOCA.

Staff Conclusions

The staff's calculated thyroid and whole body doses from the hypothetical LOCA

are listed in Table 15.1. The staff concludes that the distances to the

exclusion area and to the low population zone boundaries of the Watts Bar site,

in conjunction with the ESF of the Watts Bar design, are sufficient to provide

reasonable assurance that the total radiological consequences of a postulated

LOCA will be within the exposure guidelines set forth in 10 CFR 100.

This conclusion is based upon the staff review of the applicant's analyses and

on an independent analysis performed by the staff to verify that the total

calculated doses are within the guidelines.

15.4.2 Main Steamline Break Outside Containment

Both the staff and the applicant have evaluated the radiological consequences

of a postulated steamline break accident occurring outside containment and

upstream of the main steam isolation valve. Although the contents of the

secondary side of the affected steam generator would be vented initially to the

atmosphere as an elevated release, the staff has conservatively assumed that

the entire release throughout the course of the accident is released under

ground level conditions. During the course of the accident, the shell side of

the affected steam generator was assumed to stay dry because auxiliary feedwater flow to the affected steam generator would be blocked off under the

conditions of this accident. Because of the dryout condition in the affected

steam generator, all iodine transported to the secondary side by leakage

(1 gpm) was assumed available for release to the atmosphere with no reduction

as a result of holdup or attenuation.

As part of its review, the staff investigated three scenarios. For Case 1, the

most reactive control rod is assumed to be stuck in the fully withdrawn position. The applicant has indicated and the staff agrees that no departure from

nucleate boiling is expected to occur and, therefore, no fuel cladding failure

is to be assumed in the calculations. With no fuel failures present, Case 1

becomes identical to Case 2, and no radiological consequences are presented for

Case 1.

Watts Bar SER 15-17
For Case 2, the staff assumed that an iodine spike occurred as a result of the

power and pressure transient caused by the accident. Prior to the accident,

the Watts Bar plant was assumed to be operating at the equilibrium primary

coolant Technical Specification limit of 1 PCi/gram dose equivalent iodine-131

(DEI-131). The iodine spike generated during the accident is assumed to

increase the release rate of iodine from the fuel by a factor of 500. This

increase in the release rate results in an increasing iodine concentration in

the primary coolant during the course of the accident. The radiological

consequences for this case have been calculated using assumptions found in

Table 15.3, and the consequence values are reported in Table 15.1.

For Case 3, the staff assumed that previous reactor operation has resulted in a

primary coolant concentration equal to the maximum transient full power Technical Specification limit (60 pCi/gram DEI-131). As in Case 2, the radiological

consequences were calculated using assumptions found in Table 15.3, and the

consequence values are reported in Table 15.1.

Table 15.3 Assumptions used to evaluate the radiological

consequences following a postulated main

steamline break accident outside containment

Power 3592 MWt

Preaccident dose-equivalent 1-131 in primary coolant 1.0 pCi/g

(Case 2)

Preaccident dose-equivalent 1-131 in primary coolant 60.0 pCi/g

(Case 3)

Primary-to-secondary leak rate, as limited by

Technical Specifications 1 gpm

All of the 1gpm leak occurs in the affected steam

generator

All the iodine transported to the shell side of the

steam generator by the leakage is lost to the

environment without decay

Iodine release rate from.fuel increases by a factor,

of 500 as a result of the accident (Case 2) x/Q values

0-2 hr at 1200 m 3.6 x 10-4 sec/m- 3

0.8-hr at 4828 m 5.0 x iO-5 sec/m- 3

Based on its findings, the staff concludes that the distances to the exclusion

area and low population zone outer boundaries for the Watts Bar site are sufficient to provide reasonable assurance that the calculated radiological consequences of a postulated main steamline failure outside the Watts Bar containment do not exceed (1) the exposure guidelines as set forth in 10 CFR

100.11 for the case in which failure occurs with a primary coolant iodine

concentration corresponding to a pre-accident iodine spike; and (2) 10 percent

Watts Bar SER 15-18
of these exposure guidelines for the case that the failure occurs with a

primary coolant activity corresponding to the maximum equilibrium'concentration for continued full power operation as stated in the Watts Bar plant-specific

Technical Specifications. The. staff has reviewed the Watts Bar plant-specific

Technical.Specifications to ensure that the assumptions used in the analyses identified above are not exceeded. The staff also concludes that the proposed

design and operation of the Watts Bar nuclear station is effective in controlling the release of fission products following a postulated main steamline

break accident.

These of conclusions are based on (1) the staff review of the applicant's analysis the radiological consequences, (2) the independent.dose calculation by the

staff using appropriate regulatory assumptions and conservative atmospheric

diffusion factors as discussed in Chapter 2 of this report, and (3) the Watts

Bar plant-specific Technical Specifications for the iodine concentration in the

reactor coolant consisting of a maximum allowable limit and a limit for the

equilibrium concentration for continued plant operation and the Watts Bar

Technical Specification limit on primary to secondary.leakage in the steam

generators.

15.4.3 Steam Generator Tube Rupture

A nonmechanistic guillotine break of a steam generator tube is postulated to

occur when the reactor is at power. Because of the rupture, the primary coolant boundary is breached. The leak rate is larger than the maximum capacity of

the charging pumps to maintain inventory and of the pressurizer heaters to

maintain pressure. Thus, both pressure and level in the pressurizer would

decrease. At about 15 min after the accident, either.the low pressurizer

pressure trip or the, low pressurizer level set point is reached. The resultant

reactor and turbine trip immediately terminate power output to the grid. This

disturbance to the grid is assumed to cause loss of offsite power to the plant.

With loss of offsite power, plant cooldown is effected by a combination of the

operation of automatic safety valves and manual atmospheric relief valves.

Diagnosis of the accident is achieved by observing annunciations of condenser

high radiation alarm, steam generator feedwater/steam flow mismatch, decreasing

pressurizer pressure and level, and increasing level in one steam~generator.

The staff assumed, conservatively, that it would take the operator 30 min from

the onset of tube rupture to diagnose the accident and, isolate the affected

steam generator.

Before reactor trip and steam venting, some iodine is removed from.the condensate by the air-ejector and released to the environment. However, radioactivity released via this route is small because of preferential retention of

iodine in.the condensate. At 15 min after the accident, reactor trip and loss

of offsite power terminate availability of the condenser.and steam is vented

through the safety/relief valves:as mentioned earlier... Some of the leakage occurring through the ruptured tube is assumed, to flash to steam in the affected

steam generator. The staff has have estimated the flash fraction using the

methodology presented in NUREG-0409, "Iodine Behavior in a PWR Cooling System Following a Postulated Steam Generator Tube Rupture Accident," and conservatively

assumed that all of this fraction is immediately released from the steam. generator to the environment. The staff analyzed two cases for the Watts Bar. In

Watts Bar SER 15-19
the first case, the staff assumed that the primary coolant concentration was at

the Technical Specification limit of 60 pCi/gram dose equivalent Iodine-131

(DEI-131).

In the second case, the staff assumed that an iodine spike occurs as a result

of the power and pressure transient caused by the accident. Before the accident, the primary coolant iodine concentration was assumed to be 1.0 pCi/ gram

DEI-131. The iodine release rate from the fuel is assumed to increase by a

factor of 500. The increase in the release rate results in an increasing

primary coolant iodine concentration during the accident.

For both cases, the steam generator tube failure accident has been evaluated

with a concurrent loss of offsite power. The assumptions used in the staff

analysis are listed in Table 15.4. The calculated doses are presented in

Table 15.1.

Table 15.4 Assumptions used for the calculation of the

radiological consequences following a

postulated steam generator tube rupture

accident

Power 3592 MWt

Preaccident dose equivalent 1-131 in primary coolant 1 pCi/gm and

60 pCi/gm (two cases

analyzed)

Initial secondary coolant activity 0.1 pCi/gm DE 1-131

Primary-to-secondary leak rate 1 gpm

(to unaffected steam generator)

Isolation of affected steam generator 30 min

Iodine release rate from fuel increases by a

factor of 500 at reactor trip for iodine spiking case

x/Q value

0-2 hr at 1200 m 3.6 x 10-4 sec/m3

0-8 hr at 4828'm 5.0 x 10-5 sec/m 3

The staff concludes that the distances to the exclusion area and to the low

population zone outer boundaries for the Watts Bar site, in conjunction with

the operation of the dose mitigating ESF systems, are sufficient to provide

reasonable assurance that the calculated radiological consequences of a postulated steam generator tube failure accident do not exceed: (1) the exposure

guidelines as set forth in 10 CFR 100.11 for the accident with an assumed

pre-accident iodine spike or with the highest worth control rod stuck out of

the core, and (2) 10 percent of these exposure guidelines for the accident with

an equilibrium iodine concentration in combination with an assumed accident

generated iodine spike.

Watts Bar SER 15-20
The staff conclusion is based on (1) the staff review of the applicant's analysis

of the radiological consequences, (2) the independent dose calculation by the

staff using conservative assumptions including atmospheric dispersion factors

as discussed in Chapter 2 of this report, and (3) the Watts Bar Technical Specifications for the iodine concentration in the primary and secondary coolant

system, and for the primary to secondary leakage in the steam generators.

15.4.4 Control Rod Ejection Accident

A nonmechanistic rupture of a control rod drive housing is postulated. Because

of the resultant opening in the pressure vessel, primary coolant is lost to the

containment, with concurrent rapid depressurization of the reactor pressure

vessel. Reactor trip, initiated by one of several trip signals, occurs rapidly.

Ejection of a control rod results in rapid reactivity insertion. The applicant

has calculated and conservatively assumed that 10 percent of the fuel elements

will experience cladding failure, releasing all their gap radioactivity. In

addition, 0.25 percent of the fuel rods are conservatively calculated to experience fuel melting. The released radioactivity is mixed immediately with the

primary coolant. Activity release to the environment may occur via each of two

pathways. The first pathway involves a release of activity to the primary containment which is then assumed to leak to the atmosphere as in the design basis

LOCA, except that the containment pressure is considered insufficient to effect

iodine removal from the ice condensers. In the second pathway, activity is

transferred from the primary to the secondary coolant via an assumed 1 gpm

primary-to-secondary leak rate. With loss of offsite power and subsequent

steam venting, some of the iodine transferred to the shell side is available

for leakage to the environment.

In calculating the consequences of this postulated event, the staff calculated

the doses as if all the activity was released via each of the above pathways.

The staff would expect the actual consequences to be some combination of these

pathways. The assumptions used in calculating the radiological consequences

are presented in Table 15.5 and the resultant doses for each pathway are contained in Table 15.1 of this SER.

The staff has reviewed the applicant's analysis for the radiological consequences following a postulated control rod ejection accident. The staff

concludes that the distances to the exclusion area and to the low population

zone boundaries for Watts Bar, in conjunction with the operation of the dose

mitigating ESF systems, are sufficient to provide reasonable assurance that the

calculated radiological consequences are well within the exposure guidelines as

set forth in 10 CFR 100.11.

The staff's conclusion is based on (1) the staff review of the applicant's

analysis of the radiological consequences, (2) the staff's independent dose

calculation using the recommendations of Appendix B of Regulatory Guide 1.77

and the atmospheric dispersion factors as discussed in Chapter 2 of this report,

and (3) the Watts Bar Technical Specifications for the primary-to-secondary

leakage in the steam generators. The staff has reviewed the proposed Watts Bar

Technical Specifications to ensure that the dose guidelines stated above would

not be exceeded.

Watts Bar SER 15-21
Table 15.5 Assumptions used forestimating the radiological consequences following a postulated control rod ejection accident

Power 3592 MWt

Volume of primary coolant 11790 ft 3

Primary-to-secondary leak rate 1 gpm

as limited by TechnicalSpecifications

10% of the fuel rods experience cladding failure,

releasing all their gap radioactivity. The

released activity is mixed immediately with

the primary coolant

0.25% of the fuel rods experience fuel melting and

all released activity is mixed immediately with'

the primary coolant

A fraction of the iodine transported to the shell side

of steam generators is lost to the environment

10% of the iodine transported to and mixed with the

secondary coolant is lost during the course.of the

accident

Primary system depressurized in about 1000 secs,

terminating primary-to-secondary leak

For the containment pathway, 50% of the iodine

released into the containment is plated out

instantaneously

Primary containment leak rate 0.25%

per day (containment leakage pathway)

With loss of offsite power, the steam releases from

the loss of ac power analysis is used to supplement the

primary to secondary leakage values

Iodine concentration in the secondary coolant 0.1 pCi/g DEI-131

x/Q values

0-2 hr at 1200 m 3.6 x 10-4 sec/m

.0-8 hr at 4828 m 5.0 x 10-5 sec/m

15.4.5 Fuel-Handling Accident

For the analysis of the fuel-handling accident in the fuel pool, the staff

assumed that a fuel assembly was dropped in the fuel pool.during refueling operations and that all of the fuel rods in the dropped assembly were damaged, thereby releasing the volatile fission gases from the fuel rod gaps into the

pool. The radiation monitors located above the pool would rapidly detect the

release of activity from the pool and initiate the ESF-grade ABGTS. The radioactive material that escaped from the fuel pool was assumed to be released to

the environment over a 2-hour period,*with the iodine activity reduced by

Watts Bar SER 15-22
filtration through the ABGTS. The radiological consequences following the

postulated accident are shown in Table 15.1 and the assumptions and parameters

used in the analysis are shown in Table 15.6. The dose model and dose conversion factors employed in the analysis were the same-as those given in Regulatory Guide 1.25, "Assumptions Used for Evaluating the Potential Radiological

Consequences of a Fuel Handling Accident in the Fuel Handling and Storage

Facility for Boiling and Pressurized Water Reactors."

Table 15.6 Assumptions used for estimating the radiological

consequences following a postulated fuel-handling

accident

Power level (mwt) 3592

Number of fuel rods damaged 264

Total number of fuel rods in core 50,952

Radial peaking factor of damaged rods 1.65

Shutdown time (hr) 100

Inventory released from damaged rods

(iodines and noble gases) (%) 10

Pool decontamination factors

Iodines 100

Noble gases 1

Iodine fractions released from pool (%)

Elemental 75

Organic 25

Iodine removal efficiencies for ABGTS (spent

fuel pool area) (%)

Elemental 99

Organic 99

Particulate 99

Iodine removal efficiencies for reactor

building purge system (%)

Elemental 95

Organic 95

Particulate 99

0-2 hr x/Q value at 1200 m 3.6 x 10-4 sec/m3

0-8 hr x/Q value of 4828 in 5.0 x 16-5 sec/m 3

The staff has also evaluated the consequences of a fuel-handling accident

inside primary containment. The applicant states that at all times during

refueling operations the containment will either be isolated or ventilated to

the atmosphere through the reactor building purge Ventilation system (RBPVS).

The RBPVS is an ESF system with filter efficiencies similar to those in the

Watts Bar SER 15-23
ABGTS. The assumptions regarding the reactor shutdown time, the number of

fuel rods damaged, the iodine decontamination efficiency of the water in the

refueling-cavity, and.the atmospheric dispersion factors for the fuel-handling accident inside containment are the same as those assumed for the fuel-handling

accident in the fuel pool area in the auxiliary building. The resultant radiological consequences following a postulated fuel-handling accident inside containment also are in Table 15.1 and the appropriate assumptions are in

Table 15.6.

The staff finds that the applicant has provided an adequate system to mitigate the radiological consequences of a postulated fuel-handling accident inside the

containment and in the spent fuel pool area. Thelstaff concludes that the fuelhandling area ventilation system meets the relevant requirements of GDC 61. The

staff further concludes that the distance to the exclusion area and to the low

population zone boundaries for Watts Bar, in conjunction with the operation of

dose mitigating ESF and implementation of plant procedures, are sufficient to

.provide reasonable'assurance that the calculated offsite radiological consequences of a postulated fuel-handling accident are well within the 10 CFR 100 exposure guideli6es.

The staff's conclusion is based on (1) the staff's determination that the design

features and plant procedures at Watts Bar meet the requirements of GDC 61 with

respect to radioactivity control; (2) the staff review of the applicant's assump- tions and analyses of the radiological consequences from the fuel handling accident; (3) the staff's independent analyses using the assumptions in Regulatory Guide 1.25, Sections C.i.a through C.1.k, and (4) the Watts Bar Technical Speci- fications relating to fuel-handling and ventilation system operation.

15.4.6 Failure of Small Lines Carrying Coo-lant Outside Containment

The staff's evaluation of the applicant's FSAR and supporting documentation

indicates that the most severe rupture of the small lines carrying primary

coolant outside containment would be the double-ended break of the 4-in.

chemical volume control system (CVCS) excess letdown and seal water return

line just outside containment.

In the staff's analysis, the primary coolant maximum equilibrium fission product concentration was taken to be at 1 pCi/gm, dose equivalent 1-131, followed by a

500-fold increase in the release rate of iodine from the fuel during the accident.

The staff has assumed that the CVCS excess letdown and seal water return line

will be isolated in 30 min following the break and that 62,883 pounds of primary coolant will be released. The iodine in the released coolant, which would flash

to steam, was assumed to be released directly to the environment. No credit for

holdup, filtering, or plateout was assumed for the auxiliary building. The atmosphere dispersion factors used in the staff's analysis are presented in Table 15.2.

The resultant 2-hr thyroid and whole body doses for this postulated accident at

the exclusion area boundary are found to be 17 rems and <0.1 rem, respectively.

The 30-day doses at the low population zone are lower by a factor of 7.4. These

doses are a small fraction (no more than 1%) of the 10 CFR 100 exposure guide- lines, are in accordance with Regulatory Guide 1.11, and, therefore, meet the

Watts Bar SER 15-24
acceptance criteria of SRP Section 15.6.2. The staff finds that the applicant's

design for mitigating the radiological consequences of the failure of small line

carrying primary coolant outside containment acceptable.

15.4.7 Postulated Radioactive Releases as a Result of Liquid Tank Failures

The consequences of tank failures for tanks located outside the reactor containment, which could result in releases of liquids containing radioactive

materials to the environs, were evaluated. Considered in the staff evaluation

were (1) the radionuclide inventory in each tank assuming a 1 percent operating

power fission product source term, (2) a tank liquid inventory equal to

80 percent of its design capacity, (3) the mitigating effects of plant design

including overflow lines and the location of indoor and outdoor storage tanks

in curbed areas designed to retain spillage, and (4) the effects of site

geology and hydrology.

The applicant has incorporated provisions in the design to retain releases from

liquid overflows as discussed in Section 11.2.1 of this SER. In the event of a

spill, the staff postulated liquid flow directly into the underdrain system

where it would be pumped to the holding pond as discussed in Section 2.4 of

this SER. For the evaluation, the staff assumed failure of an 11,000-gal boric

acid bottoms tank entering the groundwater by seepage through the floor and, if

it were not collected by the underdrain system, could move in the direction of

the Chickamauga Reservoir. The staff calculated the liquid transit time for

radwaste leakage to the reservoir to be greater than 267 days.

The combined dilution factor resulting from mixing with the groundwater and

reservoir water was calculated to be 3.4 x 107.

Based on its evaluation, the staff concludes that a failure of the tank will

give a concentration at the nearest surface water in an unrestricted area of

less than the limits of 10 CFR 20, Appendix B, Table II, Col. 2. Based on the

foregoing evaluation, the staff concludes that the provisions incorporated in

the applicant's design to mitigate the effects of component failures involving

contaminated liquids are acceptable.

15.5 NUREG-0737 Items

15.5.1 Thermal Mechanical Report (II.K.2.13)

NUREG-0737 requires the performance of a detailed analysis of the thermal

mechanical conditions in the reactor vessel during recovery from small breaks

with an extended loss of all feedwater.

In a submittal dated September 14, 1981, the applicant committed to the Westinghouse Owners Group generic resolution of this issue. The owners group schedule

is consistent with the schedule specified in NUREG-0737, and is acceptable. If

a need is indicated by the review of the Owners Group submittal, the applicant

will be required to supply plant-specific analysis based on the generic study.

15.5.2 Voiding in the Reactor Coolant System During Transients (II.K.2.17)

NUREG-0737 requires the applicant to analyze the potential for voiding in the

reactor coolant system during anticipated transients.

Watts Bar SER 15-25
In a submittal dated September 14, 1981, the applicant referenced a Westing- house Owners Group generic study (scheduled to be submitted in October 1981). to address the issue of voiding in the reactor coolant system. The submittal

is on a schedule consistent with NUREG-0737 requirements. The reference is

acceptable and, if a need is indicated by the review of the Westinghouse Owners Group study, the staff will require modifications to Watts Bar.

15.5.3 Installation and Testing of Automatic Power-Operated Relief Valve

Isolation System (II.K.3.1) and Report on Overall Safety Effect of

Power-Operated Relief Valve Isolation System (II.K.3.2)

These items require all PWR licensees to provide a system that uses the PORV

block valve to protect against a small-break [OCA. The system would automatically close the block valve when the reactor coolant system pressure decays after the PORV opens. The staff requirements provide, however, that such a

control system is not required if studies provided in response to Item II.K.3.2

show that the probability of the PORV sticking open is sufficiently small.

In a submittal dated September 14, 1981, the applicant referenced Westinghouse topical report WCAP-9804 to address this issue. The applicant states that he

agrees with the Westinghouse determination that an additional block valve

closure system would add little protection against a PORV LOCA.

The report was submitted on March 15, 1981, and is under staff review. If

modifications are indicated by the staff review of WCAP-9804, the staff will

require any necessary modifications to the Watts Bar system.

15.5.4 Automatic Trip of Reactor Coolant Pumps (II.K.3.5)

NUREG-0737 requires that the reactor coolant pumps be tripped automatically in

case of a small break LOCA. The applicantwas asked to consider other solutions to the small-break LOCA problem.

In a submittal dated September 14, 1981, the applicant has referenced the Westinghouse Owners Group generic resolution of this issue, which is progress- ing on a schedule consistent with the intent of NUREG-0737 requirements. Based

on the satisfactory progress of this issue and the applicant's reference, the

staff finds this item acceptable and will require modifications if *indicated by the review and resolution of this issue.

15.5.5 Small-Break LOCA Methods (II.K.3.30) and Plant-Specific Calculations

(II.K.3.31)

NUREG-0737 requires the NSSS vendor to revise its analysis methods for smallbreak LOCA analysis to show compliance with Appendix K of 10 CFR 50. The

applicant was then required to submit plant-specific calculations using the

above NRC-approved models.

In a submittal dated September 14, 1981, the applicant referenced Westinghouse Small-Break Evaluation Model modifications being performed for the Westinghouse Owners Group and scheduled for completion by January 1, 1982. The applicant has committed to provide new analyses using the approved Westinghouse model if

so required by NRC review of the model. This schedule is consistent with the

Watts Bar SER 15-26
requirements of NUREG-0737, and the staff finds the applicant's reference and

commitment acceptable.

15.6 Relative Risk of Low Power Operation

Initial licensing action for the Watts Bar facility is anticipated to be for

operation at power levels up to 5 percent of the full-rated power. The staff

has considered the risks attendant with this operating mode and has concluded

that there is a substantial reduction in risk for low power testing as compared

to continuous full power operation. Although a detailed quantitative review of

the Watts Bar facility has not been performed, five previous reviews, including

all light water reactor vendor designs, two of which were Westinghouse plants

(one being the Sequoyah facility), have been performed. In all these cases,

the reduction in risk that was calculated was much greater than two orders of

magnitude.

In general, there are three major factors that contribute to this reduction in

risk during low power operation

(1) There is additional time available for the operators to correct the loss

of important safety systems needed to mitigate relatively high risk

events, or to take alternate courses of action.

(2) The fission product inventory during this time would be very much less

than during full power operation.

(3) There is a reduction in required capacity for mitigation systems at low

power.

Based on the low fission product inventory, the staff has estimated a reduction

in risk of at least two orders of magnitude at 5 percent power versus full

power, without further review.

Watts Bar SER 15-27

16 TECHNICAL SPECIFICATIONS

The Technical Specifications in a license define certain features, characteristics, and conditions governing operation of a facility that cannot be changed

without prior approval of the staff. The finally approved Technical Specifications will be made a part of the Operating License. Included will be sections

covering definitions, safety limits, limiting safety system settings, limiting

conditions for operations, surveillance requirements, design features, and

administrative controls.

The Technical Specifications for this facility will be based on "Standard

Technical Specifications for Westinghouse Pressurized Water Reactors" (NUREG-0452,

Revision 4). This document has been updated from earlier revisions as a result

of continued discussion with Westinghouse and other licensees with Westinghouse

PWRs.

The staff is working with the applicant to prepare a draft of the Technical

Specifications for the Watts Bar facility. On the basis of its review to date,

the staff concludes that normal plant operation within the limits of the

Technical Specifications will not result in offsite exposure in excess of the

10 CFR 20 limits. Furthermore, the limiting conditions for operation and

surveillance requirements will ensure that necessary engineered safety features

will be available in the event of malfunctions within the facility.

During its review of the Watts Bar application, the staff identified certain

issues that must be included in the Technical Specifications as a condition of

staff acceptance. These issues are listed below and are discussed further in

the sections of this report as indicated.

(1) emergency flooding (2.4.3)

(2) test requirements for pressure isolation valves (3.9.6)

(3) surveillance requirements on leak rate testing (3.9.6)

(4) fuel rod bowing (4.2.3)

(5) fuel clad swelling (4.2.3)

(6) surveillance requirements on crud deposition (4.4.4.2)

(7) limiting conditions and surveillance requirements for loose parts monitoring

system (4.4.5)

(8) limitation on reactor coolant loop operation (4.4.7)

(9) administrative procedures for overpressure protection (5.2.2)

(10) requirements for water inventory balance (5.2.4)

Watts Bar SER 16-1
(11) statistical sampling on Weight of ice baskets (6.2.1.1)

(12) operating limits for ice weighing program (6.2.1.1)

(13) limits for containment leakage (6.2.6)

(14) safety system set points (7.1.3.1)

(15) surveillance requirements for reactor trip breaker shunt coils (7.2.3)

(16) set points for overpressure protection system (7.6.5)

(17) parameters for the dc power system monitoring annunciation (8.3.2.2)

(18) surveillance requirements for nonsafety loads powered from the dc distribution system and vital inverters (8.3.2.5)

(19) surveillance requirements for the submerged electrical equipment (8.3.3.1.1)

(20) surveillance requirements for shared dc power systems (8.3.3.2.1)

(21) surveillance requirements for shared ac power systems (8.3.3.2.2)

(22) surveillance requirements for protective devices on Class 1E power systems and their distribution circuits (8.3.3.3)

(23) periodic measurement of the fuses and terminal connection resistance

(8.3.3.6)

(24) surveillance requirements for valve positions in the fire protection system (9.5.1.2)

(25) Appendix R to 10 CFR 50 requirements or equivalent protection (9.5.1.8)

(26) inservice inspection requirements for main steam, stop, and control valves

and reheat valves (10.2.1)

(27) requirements for fill of diesel generator storage tanks through manways (9.5.4.2)

(28) limits for steam generator (10.3.4)

(29) inservice inspection program for valves in the turbine bypass (10.4.4)

(30) requirements for the solid waste management system (11.4)

Watts Bar SER 16-2
17 QUALITY ASSURANCE

17.1 General

The description of the quality assurance (QA) program for the operations phase

of the Watts Bar plant is contained in Section 17.2 of the FSAR through Amendment 46. Section 17.2 references the TVA Topical Report TVA-TR75-1A, "Quality

Assurance Program Description for Design, Construction, and Operation for Nuclear

Power Plants." The staff evaluation of this QA program is based on a detailed

review of this information and discussions with the applicant to determine how

the QA program for the operations phase complies with the requirements of

Appendix B to 10 CFR 50, "Quality Assurance Criteria for Nuclear Power Plants

and Fuel Reprocessing Plants"; with the applicable Regulatory Guides and ANSI

Standard listed in Table 17.1; and with Revision I of the SRP Section 17.2,

November 1965, "Quality Assurance During the Operations Phase." Since this

review effort, the SRP was updated to Revision 2 and issued in NUREG-0800, July 1,

1982. An additional review, conducted to determine the extent to which the

TVA QA program meets Revision 2, shows that the QA program complies with Revision 2 of the SRP except for the following controls to which no clear commitment is provided.

The program does not specify:

(1) Criteria for determining the size of the QA organization, including the

inspection staff (SRP Section 17.1, Item 1A5).

(2) That designated QA individuals are involved in day-to-day plant activities

important to safety (for example, the QA organization routinely attends

and participates in daily plant work schedule and status meetings to ensure

they are kept abreast of day-to-day work assignments throughout the plant

and that there is adequate QA coverage relative to procedural and inspection

controls, acceptance criteria, and QA staffing and qualification of personnel

to carry out QA assignments) (SRP Section 17.1, Item 1B6).

(3) That the QA organization reviews and documents concurrence with qualityrelated procedures (SRP Section 17.1, Item 2Blb).

(4) That the QA organization participates early in the QA program definition

stage to determine and identify the extent QA controls are to be applied

to specific structures, systems, and components. This effort involves

applying a defined graded approach to certain structures, systems, and

components in accordance with their importance to safety. This affects

such disciplines as design, procurement, document control, inspection tests,

special processes, records, audits, and others described in 10 CFR 50,

Appendix B (SRP Section 17.1, Item 2B3 partial).

(5)' That procedures are established and described requiring a documented check

to verify the dimensional accuracy and completeness of design drawings

and specifications (SRP Section 17.1, Item 3E1).

Watts Bar SER 17-1
TAble 17.1 Regulatory guidance applicable

to the quality assurance programs

1. Regulatory Guide 1.8 (Revision 1 - September 1975), "Personnel Selection

and Training"

2. Regulatory Guide 1.28 (Revision 1 - March 1978), "Quality Assurance

Program Requirements (Design and Construction)"

3. Regulatory Guide 1.30 (Revision 0 - August 1972),"Quality Assurance

Requirements for Installation, Inspection, and Testing of Instrumentation

and Electric Equipment"

4. Regulatory Guide 1.33 (Revision 2 - February 1978), "Quality Assurance

Program Requirements (Operation)"

5. Regulatory Guide 1.37 (Revision 0 - March 1973), "Quality Assurance

Requirements for Cleaning of Fluid Systems and Associated Components of

Water-Cooled Nuclear Power Plants"

6. Regulatory Guide 1.38 (Revision 2 - May 1977), "Quality Assurance

Requirements for Packaging, Shipping, Receiving, Storage, and Handling of

Items for Water-Cooled. Nuclear Power Plants"

7. Regulatory Guide 1.39 (Revision 2 - September 1977), "Housekeeping

Requirements for Water-Cooled Nuclear Power Plants"

8. Regulatory Guide 1.58 (Revision 1 - September 1980), "Qualification of

Nuclear Power Plant Inspection, Examination, and Testing Personnel"

9. Regulatory Guide 1.64 (Revision 2 - June 1976), "Quality Assurance

Requirements for the Design of Nuclear Power Plants"

10. Regulatory Guide 1.74 (Revision 0 - February 1974), "Quality Assurance

Terms and Definitions"

11. Regulatory Guide 1.88 (Revision 2 - October 1976), "Collection, Storage, and Maintenance of Nuclear Power Plant Quality Assurance Records"

12. Regulatory Guide 1.94 (Revision 1-- April 1965), "Quality Assurance

Requirements for Installation, Inspection, and Testing of Structural

Concrete and Structural Steel During the Construction Phase of Nuclear,

Power Plants"

13. Regulatory Guide 1.116 (Revision O-R -. May 1977), "Quality Assurance

Requirements for Installation, Inspection, and Testing of Mechanical

Equipment and Systems"

14. Regulatory Guide 1.123 (Revision 1 - July 1977), "Quality Assurance

Requirements for Control of Procurement of Items and Services for Nuclear

Power Plants"

15. ANSI N45.2.12 (Draft 3, Revision 4 - February 1974), "Requirements for

Auditing of Quality Assurance Programs for Nuclear Power Plants"

16. Regulatory Guide 1.146 (Revision 0 - August 1980), "Qualification of

Quality Assurance Program Audit Personnel for Nuclear Power Plants"

Watts Bar SER 17-2
(6) That procedures are established and described requiring that design drawings

and specifications be reviewed by the QA organization to ensure that the

documents are prepared, reviewed, and approved in accordance with company

procedures and that the documents contain the necessary QA requirements

such as inspection and test requirements, acceptance requirements, and

the extent of documenting inspection and test results (Ref. SRP Section 17.1, Item 3E2).

(7) That for commercial "off-the-shelf" items where specific quality assurance

controls appropriate for nuclear applications cannot be imposed in a practicable manner, special quality verification requirements shall be established

and described to provide the necessary assurance of an acceptable item by

the purchaser (SRP Section 17.1, Item 7B4).

(8) That program procedures provide criteria for determining the accuracy

requirements of inspection equipment and criteria for determining when

inspections are required or define how and when inspections are performed.

The QA organization participates in the above functions (SRP Section 17.1,

Item 1OA partial).

The above quality assurance controls generally provide for increased involvement

of the QA organization in certain quality,-related activities, and increased

emphasis on the need for procedures., documentation, and requirements for other

quality-related activities. As such, the staff believes that adequate quality

,assurance controls in these areas are already included in the QA program for

Watts Bar, and that the above listed items described in the SRP (NUREG-0800)

need not be mandatory at this time.

Revision 4 to the TVA Topical Report has been found acceptable by the staff,

and Revision 5 is currently under review. The information provided below is

based on those portions of Revision 5 that have been found acceptable.

17.2 Organization

The organization responsible for the operation of the Watts Bar nuclear plant

and for the establishment and execution of the operations phase QA program is

shown in Figure 17.1. During the operations phase, the Office of Engineering

Design and Construction provides design, procurement, construction, and installation support. This office includes a QA organization at the "Office" level

as shown in Figure 17.1.

The Office of Power has overall responsibility for the TVA power program, including operation of the Watts Bar nuclear plant; the Manager of Power has overall

responsibility for QA during startup and operation of the plant. The Manager

of Power also has responsibility for establishing quality policies, goals, and

objectives.

The QA program is implemented under the direction of the Deputy Manager of

Power (Energy Supply). Implementation is carried out through the Office of

Power Quality Assurance Manager and the directors of other involved divisions.

The Quality Assurance and'Audit Staff reports to the Office of Power Quality

Assurance Manager and is responsible to develop, coordinate, monitor, audit,

Watts Bar SER 17-3
I Ns'rRtJME NT", -ILI •IE.•iJAN IO I(:A COMU L I ICAT RNGN•ARCTO ISRIESH AI.

AND C:ONTRO I.S- BRANCH RNGINEERiING SRVNICES

LINI•S OF (.O\MMUNIC'ATION

I

Figure 17.1 TVA Quality Assurance Organization

Watts Bar SER 17-4
and evaluate the QA program to meet regulatory requirements and guidance as

well.as licensing commitments.

Within the Office of Power, there are full-time QA staffs at several organization levels. These QA personnel have the authority to identify quality problems,

to provide solutions, to verify implementation of solutions, and to stop an

activity through the appropriate channels in accordance with TVA procedures

when the work fails to comply with approved specifications and plans.

The Director of the Division of Nuclear Power is responsible for the operation

and maintenance of the Watts Bar nuclear plant during the operations phase.

He has delegated the responsibility of the day-to-day operation and maintenance

activities for the plant to the Nuclear Production Manager. The Watts Bar Plant

Superintendent, who reports to the Nuclear Production Manager,'has primary responsibility for operating and maintaining the Watts Bar Nuclear Plant in compliance

with the requirements of the Operating License and the plant Operational Quality

Assurance Manual. The resolution of any disputes on QA program requirements

arising between QA personnel and other department personnel that cannot be

resolved locally are referred to higher management for resolution, with eventual

resolution by the Manager of Power, if necessary.

The Field QA Section, under a supervisor who reports to the offsite QA and

Compliance Branch Chief, communicates directly with the Office of Power In-Plant

Quality Assurance Coordinator, who reports to the Quality Assurance Manager,

in matters relating to the policies and practices of the operational QA program.

This Supervisor and his field QA section perform QA functions relative to plant

operations and provide inspections and verification of those activities. They

review drawings, specifications, purchase requisitions, and plant instructions

and procedures covering activities such as-test, calibration, special processes,

maintenance, modification, and repair for compliance with the QA program requirements. They are responsible for developing and implementing the inspection

program covering operations, maintenance, repair, and test.

The Plant Operations Review Committee, chaired by the Plant Superintendent,

serves in an advisory capacity to the Plant Superintendent in operational matters

related to safety. This committee reviews all standard and emergency operation

and maintenance instructions and changes thereto; proposed tests and experiments,

proposed changes to technical specifications, and proposed modifications to

plant systems that affect nuclear safety. The Field QA Section Supervisor is

a member of this committee.

17.3 Quality Assurance.Program

The'QA program for the operation of the Watts Bar nuclear plant implements the

requirements of the TVA QA policies via the Office of Power Quality Assurance

Manual, the Operational Quality Assurance Manual, the Nuclear Fuel Quality

Assurance Manual, and other operating procedures and standard practices. These

documents control quality-related activities involving safety-related items so

that they comply with the requirements of Appendix B to 10 CFR 50 and other

commitments in the FSAR. The QA program requires that implementing documents

include detailed controls for (1) translating codes, standards, regulatory

requirements, technical specifications, engineering requirements, and process

Watts Bar SER 17-5
requirements into drawings, specifications, procedures, and instructions;

(2) developing, reviewing, and approving procurement documents, including changes;

(3) prescribing all quality-related activities in documented instructions, procedures, and drawings; (4) issuing and distributing approved documents; (5) purchasing items and services; (6) identifying materials, parts, and components;

(7) performing special processes; (8) inspecting and/or testing materials, equipment, processes, and services; (9) calibrating and maintaining measuring and

test equipment; (10) handling, storing, and shipping items; (11) identifying

the inspection, test, and operating status of items; (12) identifying and dispositioning nonconforming items; (13) correcting conditions adverse to quality;

(14) preparing and maintaining records; and (15) auditing activities which affect

quality.

An indoctrination and training program is established to ensure that personnel

performing activities affecting quality are knowledgeable in QA requirements,

implementing procedures, and instructions and that they have competence and

skill in the performance of their quality-related activities. The training

program also provides for retraining of personnel to ensure proficiency of performing activities affecting quality is maintained.

Quality is verified through checking, review, surveillance, inspection, testing,

and audit of quality-related activities. The QA program requires that quality

verification be performed by individuals who are not directly responsible for

performing the actual work activity. Inspections are performed in accordance

with procedures, instructions, and/or checklists approved by the Field QA Section,

the Plant Superintendent, and the Plant Operations Review Committee. Inspections

are performed by qualified personnel who are trained in accordance with TVA

training programs.

External audits of vendors and service contractors and internal audits of all

aspects of the QA program are conducted by TVA's QA organizations. Audits are

performed in accordance with pre-established written procedures by appropriately

trained personnel not having direct responsibilities in the areas being audited.

Audits, which are conducted at scheduled intervals and/or on a random unscheduled

basis, include an objective evaluation of (1) the effectiveness of implementation

of the QA program; (2) the adequacy of and compliance with QA policies, practices,

procedures, and instructions; (3) the adequacy of work areas, activities, processes, items, and records; and (4) product compliance with applicable engineering

drawings and specifications. The QA program requires documentation of audit

results and review by management having responsibility in the area audited to

determine and take any needed corrective action. Followup audits are performed

to determine that nonconformances are effectively corrected and that the corrective action precludes repetitive occurrences. Audit findings, which indicate

performance trends and the effectiveness of the QA program, are also reported

to responsible management for review and assessment.

In addition to audits, there is continual monitoring of onsite activities by

the Office of Power In-Plant Quality Assurance Coordinator, and there are annual

independent management reviews of parts of the QA program with the total program

being reviewed biennially.

Watts Bar SER ý17-6
17.4 Conclusion

The staff review of the Watts Bar nuclear plant QA program description for the

operations phase has verified that the criteria of Appendix B to 10 CFR 50 have

been adequately addressed in Section 17.2 of the FSAR through Amendment 46.

Based on its detailed review and evaluation, the staff concludes:

(1) The QA organizations of the TVA are provided sufficient independence from

cost and schedule (when opposed to safety considerations), have sufficient

authority to effectively carry out the operations QA program, and sufficient

access to management at a level necessary to perform their QA functions.

(2) The QA program description contains adequate QA requirements and a comprehensive system of planned and systematic controls that address each of

the criteria of Appendix B to 10 CFR 50 in an acceptable manner. This QA

program description, therefore, can serve as an adequate basis for the

development of specific policies and procedures to implement the QA responsibilities of TVA for the operation of the Watts Bar nuclear plant.

However, the list of items under the pertinent control'of the FSAR Chapter 17.0

QA program has not been approved by the staff. The staff will pursue this

item with the applicant and will report on the resolution in a supplement to

the SER.

Watts Bar SER 17-7

18 CONTROL ROOM DESIGN REVIEW

18.1 General

As part of the NRC task actions following the TMI-2 accident (Item I.D.1,

NUREG-0660, May 1980, and NUREG-0737, November 1980), the staff requires all

licensees and applicants for operating licenses to conduct a detailed control

room design review to identify and correct human engineering discrepancies.

These detailed control room design reviews will be initiated after issuance of

NUREG-0700,""Control Room Human Engineering Guidelines."

Those applicants for operating licenses who are unable to complete this review

before a license is issued were asked to make a preliminary design assessment.

of their control rooms to identify human engineering discrepancies and to

establish a schedule, approved by the staff, for correcting the deficiencies.

These applicants are required to complete the more detailed control room design

review, as are licensees with operating plants.

A staff team, assisted by human factors consultants from Lawrence Livermore

National Laboratory, BioTechnology, Inc., and the National Bureau of Standards,

conducted an onsite control room design review from October 6 through October 10,

1980.

The applicant performed a preliminary design assessment of the Watts Bar Unit 1

control room, and submitted his findings to the staff in a report dated

January 13, 1981. After review, the staff determined that the proposed corrective actions to all the human engineering discrepancies identified and reported

by the applicant were suitable.

Human engineering discrepancies, identified'as a result of the onsite control

room review and the applicant's preliminary design assessment, were documented

in the staff's draft control room design report and were transmitted to the

applicant. The report ranked the discrepancies according to their importance.

Observed discrepancies were given a priority rating of one, two, or three

(high, moderate, low), based on the increased potential. for operator error and

the possible Consequences of that error. The staff requires the applicant to

implement corrective measures for all priority 1and 2 deficiencies before an

operating license is issued because the correction of these items will significantly reduce the potential for operator error.

Appendix D to this report presents discussions and dispositions with regard to

the following: (1) priority 1 and 2 deficiencies; (2) priority 3 discrepancies;

(3) discrepancy priorities that have been changed and the rationale for making

such changes; and (4) discrepancies which have been determined to be invalid

and explanations for such determinations. No immediate actions are specified

for correcting priority 3 items because they will not significantly affect the

safe operation of the plant. However, the staff does require the applicant to

report on all priority 3 items as part of his detailed control room design

review and to determine the best solutions.

Watts Bar SER 18-1
At a meeting on June 16, 1981, identified deficiencies were discussed, measures

for correcting most of the deficiencies were resolved, and the schedule for

correcting them was established. All items were acceptably resolved as a

result of letters dated August 20 and October 27, 1981, and January 18,

March 26, and April 26, 1982, that contained the applicant's commitments.

18.2 Conclusions

The applicant's proposed corrections of the discrepancies documented in Appendix D

to this report are acceptable to the staff.

The staff will perform a confirmatory audit to-verify that actions have been

implemented to correct control room deficiencies and report the results in a

supplement to this report.

Based on its review of the applicant's submittals, the control room review, and

other clarifying information, the staff concludes that with the implementation

of the corrective actions specified in Appendix D (before an operating license

is issued) the potential for operator error leading to serious consequences as

a result of human factors considerations in the control room will be sufficiently low to permit safe startup and power operation of the Watts Bar nuclear plant.

The applicant must address all priority 3 items, as well as other deficiencies

that may be identified, in his detailed control room design review which will

be performed using NUREG-0700 for guidance, and final resolution of all

deficiencies must be on a schedule consistent with NUREG-0737.

Watts Bar SER 18-2
19 REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS

The Watts Bar application for an operating license is being reviewed by the

Advisory Committee on Reactor Safeguards. The NRC staff will issue a supplement to this Safety Evaluation Report after the Committee report to the

Commission is available. The supplement will append a copy of the Committee's

report, will address comments made by the Committee, and will describe steps

taken by the NRC staff to resolve any issues raised as a result of the

Committee's review.

Watts Bar SER 19-1

20 COMMON DEFENSE AND SECURITY

The application reflects that the activities to be conducted will be within the

jurisdiction of the United States and that all of the directors and principal

officers of the applicant are citizens of the United States. The applicant,

being an agency of the United States government, is not owned, dominated, or

controlled by an alien, a foreign corporation, or a foreign government. The

activities to be conducted do not involve any restricted data, but the applicant has agreed to safeguard any such data that might become involved in

accordance with the requirements of 10 CFR 50. The applicant will rely upon

obtaining fuel as it is needed from sources of supply available for-civilian

purposes, so that no diversion of special nuclear material for military purposes is involved. For these reasons, and in the absence of any information to

the contrary, the staff finds that the activities to be performed will not be

inimical to the common defense and security.

Watts Bar SER 20-1

21 FINANCIAL QUALIFICATIONS

On March 31, 1982, the NRC published in the Federal Register (47 FR 13750)

amendments to its regulations that entirely eliminate the review relating to

the financial qualifications of applicants for construction permits and

operating licenses. Because these amendments were effective immediately, there

will be no further review of the financial qualifications of the Tennessee Valley Authority.

Watts Bar SER 21-1

22 FINANCIAL PROTECTION AND INDEMNITY REQUIREMENTS

22.1 General

Pursuant to the financial protection and indemnification provisions of the

Atomic Energy Act of 1954, as amended (Section 170 and related sections), the

Commission has issued regulations in 10 CFR 140. These regulations set forth

the Commission's requirements with regard to proof of financial protection by,

and indemnification of, licenses for facilities such as power reactors under

10 CFR 50.

22.2 Preoperational Storage of Nuclear Fuel

The Commission's regulations in 10 CFR 140 require that each holder of a

construction permit under 10 CFR 50, who is also the holder of a license under

10 CFR 70 authorizing the ownership and possession for storage only of special

nuclear material at the reactor construction site for future use as fuel in the

reactor (after an operating license is issued under 10 CFR 50), shall, during

the interim storage period before licensed operation, have and maintain financial

protection in the amount of $1,000,000 and execute an indemnity agreement with

the Commission. Proof of financial protection is to be furnished prior to, and

the indemnity agreement executed as of, the effective date of the 10 CFR 70

license. Payment of an annual indemnity fee is required.

The applicant will furnish the Commission proof of financial protection in the

amount of $1,000,000 in the form of a Nuclear Energy Liability Insurance

Association Policy (Nuclear Energy Liability Policy, facility form NF-256).

Further, the applicant will execute an Indemnity Agreement with the Commission

effective as of the date of its preoperational fuel storage license. The

applicant will pay the annual indemnity fee applicable to preoperational fuel

storage.

22.3 Operating Licenses

Under the Commission's regulations, 10 CFR 140, a license authorizing the

operation of a reactor may not be issued until proof of financial protection in

the amount required for such operation has been furnished, and an indemnity

agreement covering such operation (as distinguished from preoperational fuel

storage only) has been executed. The amount of financial protection which must

be maintained for the Watts Bar plant (which have a rated capacity in excess of

100,000 electrical kilowatts), is the maximum amount available from private

sources (that is, the combined capacity of the two nuclear liability insurance

pools; this amount is currently $160 million).

Accordingly, licenses authorizing operation of the Watts Bar nuclear plant will

not be issued until proof of financial protection in the requisite amount has

been received and the requisite indemnity agreement executed.

Watts Bar SER 22-1
The staff expects that, in accordance with the usual procedure, the nuclear

liability insurance pools will provide, several days in advance of anticipated

issuance of the operating license document, evidence in writing, on behalf of

the applicant, that the present coverage has been appropriately amended so that

the policy limits have been increased, to meet the requirements of the Commission's

regulations for reactor operation. Similarly, operating licenses will not be

issued until an appropriate amendment to the present indemnity agreement has

been executed. The applicant will be required to pay an annual fee for operating

license indemnity as provided in our regulations.

On the basis of the above considerations, the staff concludes that the presently

applicable requirements of 10 CFR 140 have been satisfied and that, before an

operating license is issued, the applicant will be required to comply with the

provisions of 10 CFR 140 applicable to operating licenses, including those as

to proof of financial protection in the requisite amount and as to execution of

an appropriate indemnity agreement with the Commission.

Watts Bar SER 22-2
23 CONCLUSIONS

Based on its evaluation of the application as set forth above, the staff has

determined that, upon favorable resolution of the outstanding matters described

herein, it will be able to conclude that

(1) The application for facility licenses filed by the applicant, dated

October 4, 1976, complies with the requirements of the Atomic Energy Act

of 1954, as amended (Act), and the Commission's regulations set forth in

10 CFR Chapter I, except as duly exempted therefrom.

(2) Construction of the Watts Bar Nuclear Plant, Units 1 and 2 has proceeded,

and there is reasonable assurance that both will be substantially completed, in conformity with Construction Permits CPPR-91 and CPPR-92, the

application as amended, the provisions of the Act, and the rules and regulations of the Commission.

(3) The facilities will operate in conformity with the application as amended,

the provisions of the Act, and the rules and regulations of the Commission.

(4) There is reasonable assurance (a) that the activities authorized by the

Operating Licenses can be conducted without endangering the health and

safety of the public and (b) that such activities will be conducted in

compliance with regulations of the Commission set forth in 10 CFR

Chapter I.

(5) The applicant is technically and financially qualified to engage in the

activities authorized by the licenses, in accordance with the regulations

of the Commission set forth in 10 CFR Chapter I.

(6) The issuance of these licenses will not be inimical to the common defense

and security or to the health and safety of the public.

Before an Operating Licenses are issued to the applicant for operation of the

Watts Bar Nuclear Plant, Units 1 and 2, the units must be completed in conformity with the provisional Construction Permits, the application, the Act,

and the rules and regulations of the Commission. Such completeness of construction as is required for safe operation at the authorized power levels must

be verified by the NRC before the licenses are issued.

Furthermore, before Operating Licenses are issued, the applicant will be required to satisfy the applicable provisions of 10 CFR 140.

Watts Bar SER 23-1

APPENDIX A

CHRONOLOGY OF RADIOLOGICAL REVIEW

OF WATTS BAR NUCLEAR PLANT, UNITS 1 AND 2,

OPERATING LICENSE REVIEW

October 4, 1976

December 3, 1976

February 3, 1977

February

February

March 8,

4, 1977

18, 1977

1977

March

March

April

April

18, 1977

23-24, 1977

1, 1977

6, 1977

Final Safety Analysis Report for operating license for

'Watts Bar, Units 1 and 2, docketed. (Amendment 23)

Letter from applicant transmitting meteorological

information.

Meeting with applicant to discuss pipe restraints

(Summary issued March 15, 1977.)

Applicant submits Amendment 25 to FSAR.

Applicant submits Amendment 26 to FSAR.

Letter to applicant transmitting radiological safety review schedule.

Applicant submits Amendment 27 to FSAR.

Site visit.

ASLB issues Notice of Special Prehearing Conference.

Letter to applicant concerning potential consequences of a

fuel-handling accident.

Letter from applicant concerning fire protection program

re-evaluation.

Letter to applicant transmitting request for additional

information on a variety of subjects.

Letter to applicant transmitting request for additional

information on a variety of subjects.

Letter to applicant transmitting request for additional

information on a variety of subjects.

Applicant submits Amendment 28 to FSAR.

Letter to applicant concerning fire protection functional

responsibilities.

April 18, 1977

May 5, 1977

June 1, 1977

June 24, 1977

July 1, 1977

August 29, 1977

Watts Bar SER A-1
September 30, 1977

October 7, 1977

November 11, 1977

November 16, 1977

November 28, 1977

January 9, 1978

January 19, 1978

January 27,

January 27,

1978

1978

February 23, 1978

March 1, 1978

March 7, 1978

April 5, 1978

April 5, 1978

April.17, 1978

May 26, 1978

June 19, 1978

June 22, 1978

July 13, 1978

August 2, 1978

Letter to applicant requesting additional information on

meteorological data.

Applicant submits Amendment 29 to FSAR.

Applicant submits Amendment 30 to FSAR.

Letter to applicant requesting additional information

concerning quality assurance and conduct of operations.

Letter to applicant concerning requirements for physical

protection of licensed activities in nuclear power reactors

against industrial sabotage (generic).

Letter from applicant concerning schedule for construction

completion.

Letter to applicant requesting additional information on a

variety of subjects.

Applicant submits Amendment 31 to FSAR.

Letter to applicant forwarding letter from Corps of

Engineers (COE) and requesting response to COE concerns.

Letter to applicant concerning inservice testing program

for pumps and valves.

Meeting with applicant to discuss adequacy of seismic

design. (Summary issued March 28, 1978.)

Letter to applicant concerning implementation of Regulatory

Guide 1.97.

Meeting with applicant to discuss adequacy of seismic

design.

Applicant submits Amendment 32 to FSAR.

Applicant submits 1977 "Power Annual Report."

Applicant submits Amendment 33 to FSAR.

Letter to applicant requesting additional information on a

variety of subjects.

Meeting with applicant to discuss requests for information.

(Summary issued July 13, 1978.)

Letter to applicant requesting additional information on a

variety of subjects.

Letter from applicant responding to COE questions.

Watts Bar SER A7-2
August 11, 1978

August 15,1978

August 17, 1978

August 17, 1978

August 21, 1978

August 31, 1978

September 19, 1978

October 3, 1978

October

October

4,

5,

1978

1978

Letter to applicant concerning staffing requirements for

operating reactors.

Letter to applicant concerning standard format for

meteorological data on magnetic tape.

Letter to applicant requesting additional meteorological

data.

Meeting with applicant to discuss seismic design. (Summary

issued September 11, 1978.)

Letter from applicant concerning cavity annulus seal ring.

Applicant Submits Amendment 34 to FSAR.

Letter to applicant concerning environmental qualification

of equipment.

Letter to applicant concerning operation of dew point

sensor measurements.

Letters to applicant concerning seismic design.

Letter from applicant concerning impact of revised LOCA

and ,seismic input data.

Letter from applicant concerning seismic design criteria.

Meeting to discuss implementation of upgraded guard

qualification, training, and contingency planning

requirements.

Letter from applicant transmitting "Tritium Management

Considerations for PWRs."

Site visit by Instrumentation and Control Systems Branch.

Letter from applicant concerning scheduled fuel-load dates.

Applicant submits Amendment 35 to FSAR.

Letter from applicant concerning environmental and seismic

qualification.

Letter to applicant concerning justification of design

basis for tornados.

Applicant submits Amendment 36 to FSAR.

Letter to applicant concerning structural aspects of plant

design.

October 6, 1978

October 13, 1978

October 19, 1978

October

November

November

November

23-27, 1978

3, 1978

3, 1978

8, 1978

November 15, 1978

November 24, 1978

November 28, 1978

Watts Bar SER A-3
November 30, 1978

December 6, 1978

December 7, 1978

December 15, 1978

December 19, 1978

December 28, 1978

January 4, 1979

January 8, 1979.

January 15, 1979

January 24, 1979

February 6, 1979

February 9, 1979

February 16, 1979

Letter to applicant restating NRC views on seismic issues.

Letter from applicant concerning environmental qualification

of Class 1E equipment.

Letter from applicant concerning preservice baseline and

inservice program.

Letter to applicant concerning hydrologic aspects of

facility design.

Letter from applicant concerning seismic design.

Letter from applicant concerning use of Tefzel 200as

approved mechanical snubber seal material.

Letter from applicant concerning hydrologic analysis.

Letter to applicant concerning fire protection of facility

design.

Letter from applicant concerning check valves and leak

testing.

Applicant submits Amendment 37 to FSAR.

Letter to applicant concerning facility thermal-hydraulic

design.

Letter to. applicant concerning Instrumentation and Control

Systems Branch request for information regarding FSAR

solid state Ilogic testing and description of essential raw

cooling water system.

Letter to applicant informing that he is required to

provide an offsite dose calculation manual with proposed

Technical Specifications.

Letter from-applicant transmitting 19.78 Annual Report.

Letter from applicant transmitting safeguards contingency

plans.

Applicant submits Amendment 38 to FSAR.

Letter to applicant concerning geotechnical engineering

and reactor systems.

Letter to applicant concerning instrumentation and control

systems.

Letter from applicant concerning diesel generator relief

and drain valves.

March

March

March

March

March

April

16, 1979

21, 1979

26, 1979

26, 1979

27, 1979

6, 1979

Watts Bar SER A-4
April 25-26, 1979

April 27, 1979

May 1-2, 1979

May 16, 1979

May 17-18, 1979

May 21, 1979

May 23, 1979

May 29, 1979

June 13, 1979

June 14, 1979

June 28, 1979

July 10, 1979

July 10, 1979

July 25, 1979

August 1, 1979

August 1, 1979

August 1, 1979

August 9, .1979

August 9, 1979

Meeting with applicant to review fault trees developed for

Task A-17 and to obtain information on specific plant

components for use in development of fault trees at

component level..

Letter from applicant requesting extension of completion

dates specified inconstruction permits.

Forecast panel site.visit.

Letter to applicant concerning material integrity and

mechanical engineering information.

Meeting with applicant concerning buckling design criteria.

Letter.from applicant concerning safety-related piping.

Letter from applicant concerning containment sump.

Applicant submits Amendment 39 to FSAR.

Meeting with applicant to discuss policies

regarding review of CP and OL applications.

Meeting with applicant to discuss southern Appalachian

region tectonic study.

Letter from applicant transmitting "Southern Appalachian

Tectonic Study."

Letter to applicant notifying him that ASME Code

Class N-192 may be used in safety-related piping.

Meeting with.applicant concerning activities authorized

for all TVA facilities under construction.

Visit to Watts Bar Dam to obtain detailed information

needed to complete fault trees. (Summary issued July 26,

1979.)

Letter to applicant concerning proposed secondary water

chemistry program.

Letter from applicant transmitting "Report on Environmental

Qualification for Balance of Plant Equipment Class 1E."

Letter from applicant concerning seismic qualification

reports for transmitters mounted on local instrument

panels.

..Letter to applicant concerning mechanical engineering and

test program.

Applicant submits Amendment 40 to FSAR.

Watts Bar SER A-5
August 9, 1979

August 23, 1979

August 23,

September

1979

13, 1979

September 14,1979

September 27, 1979

October 10, 1979

.October 10, 1979

October 29, 1979

November 7, 1979

November 9, 1979

Letter from applicant transmitting "Earthquake Ground

Motion Study in Vicinity of Facility" and "Justification

of Seismic Design Criteria Used for Facilities."

Letter to applicant approving request to use Tefzel 200 as

mechanical snubber seal material.

Letter from applicant concerning revised fuel load schedule.

Letter to applicant concerning NRC emergency and operating

requirements resulting from NRC review following TMI

accident.

Meeting with applicant to collect detailed information on

location and operating characteristics of plant equipment

needed for evaluation of fault trees. (Summary issued

September 21, 1979.)

Letter to applicant concerning status and applicability of

TMI followup actions.

Letter to applicant discussing schedule for upgraded

emergency plans.

Letter from applicant concerning secondary water chemistry

control program.

Letter to applicant concerning environmental qualification

of reactor coolant temperature detectors and containment

pressure transmitters.

Letter from applicant concerning environmental qualification

of safety-related Class 1E process instrumentation.

Letter to applicant transmitting additional clarification

of TMI lessons-learned short-term requirements.

Letter to applicant concerning upgraded emergency plans.

Letter from applicant concerning alignment bed welds.

Letter from applicant concerning cable terminations in

junction boxes inside containment.

Letter from applicant concerning qualification of

resistance temperature detectors.

Letter to applicant requesting additional information

concerning evacuation times.

Letter from applicant upgraded emergency plans.

November

December

December

21, 1979

7, 1979

18, 1979

December 19, 1979

December 26, 1979

January 3, 1980

Watts Bar SER A-6
February 5, 1980

February 19, 1980

February 21, 1980

March 10, 1980

April 10, 1980

April 21, 1980

April 25, 1980

April 29, 1980

April 30, 1980

May 2, 1980

May 12, 1980

May 19, 1980

Letter to applicant regarding "Interim Staff Position on

Environmental Qulaification of Safety-Related Electrical

Equipment," NUREG-0588.

Letter' from applicant concerning containment vessel

buckling criteria.

Letter to applicant concerning qualification of

safety-related electrical equipment.

Letter to applicant concerning auxiliary feedwater system.

Letter to applicant forwarding comments on safeguards

contingency plan.

Letter to applicant concerning Category I masonry walls

(generic).

Letter to applicant concerning requirements for emergency

response facilities.

Letter from applicant concerning preservice baseline

inspection'and inservice inspection program.

Applicant submits Amendment 41 to FSAR.

Letter to applicant concrning plant-wide field audit of

installed electrical equipment and systems.

Letter from applicant concerning radiological emergency

plan.

Letter to applicant informing that ASME Code Cases N-234

and N-235 are acceptable for use in preservice and

inservice inspections of reactor pressure vessels.

Letter from applicant concerning preservice and inservice

inspection requirements.

Letter to applicant concerning potential for underclad

cracking in reactor vessel nozzles.

Letter to applicant concerning confirmatory piping

analysis.

Letter from applicant concerning seismic qualification of

Class 1E electrical equipment.

Letter to applicant concerning Commission decision on the

Union of Concerned Scientists (generic).

Letter to applicant concerning power reactor operating

licenses (generic).

June

June

June

June

June

June

9, 1980

18, 1980

23, 1980

23, 1980

25, 1980

26, 1980

Watts Bar SER A-7
June 30, 1980

July 1, 1980

July

July

2, 1980

11, 1980

July 31, 1980

July 31, 1980

August 13, 1980

August 13, 1980

August

August

18,

22,

1980

1980

Letter to applicant advising applicants and vendors of

regional meetings (generic).

Letter from applicant concerning ultrasonic examination of

welds.

Letter to applicant concerning evacuation times (generic).

Letter to applicant requesting up-to-date completion

schedules and fuel-load target dates (generic).

Letter to applicant requesting additional information

concerning power systems.

Letter to applicant concerning interim criteria for shift

staffing (generic).

Letter from applicant concerning cracking of reactor

vessel nozzles.

Letter from applicant concerning ultrasonic examination of

welds.

Meeting with applicant to discuss status of review.

Letter to applicant concerning ice condenser containment

and hydrogen control measures.

Letter from applicant transmitting 1979 Annual Report.

Letter to applicant concerning preliminary clarification

of TMI action plan requirements (generic).

Letter from applicant transmitting "Watts Bar Nuclear

Plant, Units 1 and 2, Fire Protection Submittal."

Letter from applicant concerning ATWS issues.

Letter to applicant transmitting errata sheets to

September 5, 1980, letter regarding preliminary clarification of TMI action plan requirements (generic).

Letter from applicant concerning functional criteria for

emergency response facilities.

Letter to applicant concerning environmental qualification

of safety-related equipment (generic).

Control'room design review.

Meeting with applicant concerning proposed waste-heat

utilization.

September

September

4,

5,

1980

1980

September 9, 1981

September

September

17,

19,

1980

1980

September 30, 1980

October 1, 1980

October 6-10, 1980

October 7, 1980

-Watts Bar SER A-8
October 30,

October 31,

October 31,

November 4,

1980

1980

1980

1980

November 12, 1980

November

November

November

November

November

November

November

December

December

December

December

December

December

13, 1980

14, 1980

14, 1980

17, 1980

21, 1980

21, 1980

26, 1980

4, 1980

5, 1980

8, 1980

8-10, 1980

9, 1980

9, 1980

Letter from applicant concerning extension of completion

dates in construction permits.

Letter from applicant concerning fuel storage physical

security plan.

Letter to applicant transmitting NUREG-0737 concerning TMI

action plan requirements (generic).

Letter to applicant concerning control room ceiling tiles.

Letter from applicant transmitting "Environmental

Radioactivity Levels Annual Report, 1979."

Letter to applicant concerning radiological emergency

response plan (generic).

Letter to applicant informing him that proposed deletion of

high-pressure blowdown tests is acceptable.

Letter to applicant concerning TMI-related requirements

for new operator licenses regarding training during lowpower testing.

Letter to applicant concerning inservice inspection of

pressure isolation valves.

Meeting with applicant concerning plans for emergency

management.

Letter to applicant concerning fuel storage physical

security plant.

Letter to applicant concerning environmental qualification

of safety-related electrical equipment.

Letter from applicant concerning electrical separation

field audit.

Letter to applicant concerning containment vessel building

analysis.

Meeting with applicant concerning safety and relief valve

piping computer model. (Summary issued January 15, 1981.)

Site meeting of forecast panel to determine NRC staffing

priorities. (Summary issued January 22, 1981.)

Letter from applicant concerning environmental

qualification testing of safety-related equipment.

Letter to applicant forwarding Revision 1 to NUREG-0654/

FEMA-REP-1 and summary of evacuation time estimate ratings

(generic).

Watts Bar SER A-9
December 18, 1980

January 13, 1981

Juanuary 15,

February 12,

February 17,

February 19,

February 23,

February 24,

February 25,

February 25,

February 26,

1981

1981

1981

1981

1981

1981

1981

1981

1981

Letter to applicant concerning ultimate capacity analysis

of ice condenser containments.

Letter from applicant concerning preliminary design review

of control rooms.

Letter to applicant concerning guard training and

qualification program.

Letter from applicant concerning Category I masonary

walls.

Letter to applicant concerning centralized emergency

operations program.

Letter from'applicant concerning containment vessel

building analysis.

Letter from applicant forwarding "Response to

Question 022.71, Short-Term Containment Response Analyses."

Letter from applicant concerning separation of electrical

equipment and system.

Letter from applicant concerning emergency preparedness.

exercises.

Letter to applicant concerning capability to mitigate

station blackout event and implement emergency procedures.

Letter from applicant concerning guard training and

qualifications plans.

Applicant submits Amendment 42 to FSAR.

Letter to applicant concerning environmental qualification

of safety-related electrical equipment.

Letter from applicant transmitting "Assessment of

Ultrasonic Reflectors in Watts Bar Unit 1 Reactor Vessel

Nozzle Bores."

Letter to applicant requesting additional information

concerning equipment qualification for seismic and

hydrodynamic loads.

Letter to applicant requesting additional information.

Letter from applicant concerning licensing schedule.

Letter from applicant concerning security personnel

training and safeguards contingency plans.

March

March

March

9, 1981

10, 1981

20, 1981

March 20, 1981

March

April

April

30, 1981

3, 1981

7, 1981

Watts Bar SER A-IO0
April 8, 1981

May I

May

May

May

May

May

May

May

May

May

June

June

I.,

7,

L9,

22,

25,

26S

27,

291

11

1981

1981

1981

1981

1981

1981

1981

1981

1981

1981

1981

1981

Letterfrom applicant concerning replacement bolting

material for.mounting of RHR pumps.

Letter to applicant concerning qualification of inspection,

examination and testing, and audit personnel.

Letter to applicant concerning natural circulation cooldown

(generic).

Letter to applicant requesting additional information.

Letter from applicant forwarding review comparing

Watts Bar to Sequoyah.

Meeting with applicant to discuss ICSB review process.

(Summary issued June,29, 1981.)

Meeting with applicant to discuss conceptual design for

central.proposed emergency response facility.

Meeting with applicant to resolve open SER issues.

Meeting with applicant to discuss compliance of materials

that constitute containment vessel boundary. (Summary

issued June 29, 1981.)

Letter to applicant transmitting control room review audit

report.

Letter from applicant concerning selection of break

locations/types as a result of piping reanalysis.

Letter from applicant concerning investigation of

nonreinforced-concrete masonry walls.

Letter to applicant concerning site visit to review

conceptual design for mitigating effects of potential core

melt accident.

Letter to applicant concerning meeting in Bethesda,

Maryland, to discuss requirements for environmental

qualification of safety-related electrical equipment

(generic).

Letter to applicant concerning Institute of Nuclear Power

Operation plant-specific evaluation reports (generic).

Letter from applicant concerning emergency procedures and

training for station blackout events.

Meeting with applicant to discuss proposed corrections to

deficiencies noted in control room design review/audit

report.

June 3, 1981

June

June

June

4, 1981

5, 1981

16, 1981

Watts Bar SER A-II1
June 26, 1981

June 29, 1981

July 8, 1981

July 14, 1981

July 20, 1981

July 21, 1981

July 23, 1981

July 24, 1981

July 28, 1981

July 31, 1981

August 3, 1981

August 3, 1981

August 10, 1981

August 17, 1981

August 20, 1981

August 20, 1981

August 25-27, 1981

August 27, 1981

Meeting with applicant and Westinghouse and Argonne National

Laboratory representatives to discuss *ICSB concerns

regarding balance-of-plant design.

Letter from applicant concerning ultimate capacity

analysis for facility containment.

Meeting with applicant, to discuss settlement monitoring

program and review foundations for Category I structures

and intake pumping station retaining walls. (Summary .issued August 7, 1981.),

Letter from applicant concerning-research project regarding

conceptual design for mitigation of effects of potential

core melt accidents.

Letter to applicant concerning adequacy of information

received to date.

Site visit to discuss instrument and control items.

Letter to applicant concerning installation of prompt

notification system (generic).

Applicant submits Amendment 43 to FSAR.

Meeting with applicant to discuss structural engineering,

review.

Letter to applicant concerning steam generator overfill

(generic).

Meeting with applicant to discuss facility seismic design.

Letter from applicant responding to Generic Letter 81701,

"Inspection, Examination, and.Testing and Audit Personnel."

Letter to applicant requesting additional information.

Letter from applicant transmitting final report concerning.

comparison of systems, components, and features between

Watts Bar and Sequoyah.

Letter from applicant concerning proposed corrective

actions for deficiencies identified in design review of

control room.

Letter from applicant concerning nonreinforced-concrete

masonry walls.

Site visit by Auxiliary Systems Branch.

Letter to applicant concerning the ICSB review.

Watts Bar SER A- 12
August 28, 1981

August 31, 1981

September

September

September

September

1981

September

September

September

September

'September

1981

September

September

September

September

September

September

September

September

2, 1981

3, 1981

3, 1981

9-11,

10, 1981

10, 1981

11, 1981

11, 1981

15-16,

15, 1981

16, 1981

16, 1981

21, 1981

Letter from applicant concerning the potential for liquefaction of soils at Watts Bar.

Letter to applicant concering long-term operability of deep

draft pumps.

Meeting concerning the fire protection review.

Meeting to discuss potential for liquefaction underneath

the ERCW pipelines and Class 1E electrical conduits.

Letter to applicant concerning Watts Bar review schedule

and requests for information.

Meeting to discuss management structure and technical

resources audit.

Meeting to discuss radiation protection procedures and

practices.

Letter from applicant advising the construction schedule

has been revised; fuel load dates are now August 1982 for

Unit 1 and June 1983 for Unit 2.

Letter from applicant concerning unresolved safety issues.

Letter from applicant concerning the comparison of the

Watts Bar-and Sequoyah containment systems.

Meeting to discuss Power Systems Branch concerns.

Letter from applicant responding to questions regarding

geology seismology, and chemical engineering.

Applicant submits Amendment 44 to FSAR.

Letter from applicant concerning the licensing review

schedule.

Letter from applicant forwarding W report "Seismic Testing

of the Nuclear Instrumentation System Power Range Detector

for the Watts Bar Plant."

Letter from applicant forwarding, revised critical structures,

systems, and components list.

Letter to applicant requesting additional information.

Letter from applicant concerning reactor fuel.

Meeting to discuss the licensing schedule.

21,

22,

22,

24,

1981

1981

1981

1981

Watts Bar SER A- 13
September

September

September

September

25,

25,

29,

29,

1981

1981

1981

1981

October 1, 2, 1981

October 5, 1981

October

October

October

October

October

October

October

October

October

October

October

5,

6,

7,

9,

14,

14,

16,

16,

20,

21,

21,

1981

1981

1981

1981

1981

1981

1981

1981

1981

1981

1981

Letter from applicant responding to request for information.

Letter from applicant responding to requests for additional

information.

Letter from applicant responding to request for additional

information.

Letter to applicant about the revised schedule for

completion of TMI Action Plan Item II.D.1, Relief and

Safety Valve Testing (Generic 81-36).

Meeting to discuss potential for liquefaction beneath the

ERCW pipelines and Class 1E electrical conduit. (Summary

issued November 4, 1981.)

Letter to applicant concerning results of scheduling

meeting. SER publication has been rescheduled to

December, 1, 1981.

Letter from applicant forwarding report containing

documentation of the TRIPE computer program.

Letter from applicant concerning Preservice Inspection

Program.

Letter from applicant providing information concerning

simulator examinations for Browns Ferry, Sequoyah, and

Watts Bar.

Letter from applicant responding to concerns of the Power

Systems Branch.

Letter to applicant requesting additional information

regarding comparison of containment systems of the

Sequoyah and Watts Bar plants.

Letter from applicant concerning schedule for submittal of

outstanding information on facility.

Letter from applicant concerning additional Power Systems

Branch open items.

Letter from applicant concerning compliance with GDC 51.

Letter from applicant concerning qualifications of key

plant personnel.

Letter to applicant requesting additional information.

Letter from applicant responding to Structural Engineering

Branch questions.

Watts Bar SER A- 14
October. 23, 1981

October 23, 1981

October 26, 1981

October 27, 1981

October 27-29,

1981

October 28, 1981

October 29; 1981

October 30, 1981

October 30, 1981

October 30, 1981

October-30, 1981

November 3, 1981

November 4, 1981

November 4, 1981

November 9, 1981

Letter to applicant concerning seismic qualification of

equipment important for safety.

Letter from applicant concerning containment functional

design and heat removal system.

Letter from applicant providing responses to request for

information.

Letter from applicant providing responses to request for

information concerning Regulatory Guide 1.97.

Meeting with applicant to discuss how his management

structure compares to the applicable guidelines of

NUREGs 0731 and 0737.

Letter form applicant responding to Auxiliary System

Branch questions.

Letter from applicant providing revised responses to

NUREG-0737 items.

Letter to applicant concerning Appendix R of 10 CFR 50,

Fire Protection Rule.

Letter from applicant forwarding inservice pump and valve

test program.

Letter from applicant concerning inservice inspection of

pressure isolation valves.

Letter from applicant providing responses to request for

information.

Letter from applicant concerning long-term operability of

deep draft pumps.

Letter to applicant providing final SEB criteria for

safety-related masonry wall evaluation.

Letter from applicant forwarding Revision 1 to Preservice

Inspection Program.

Letter from applicant providing responses to concerns

identified during site visit.

Letter to applicant concerning storage of low-level

radioactive wastes (Generic 81-38).

Meeting with applicant'to discuss the Preservice Inspection

Program.

Letter from applicant responding to request for additional

information.

November

November

November

10,

16,

16,

1981

1981

1981

Watts Bar SER A-15
November

November

November

November

December

18, 1981

19, 1981

20, 1981

30, 1981

3, 1981

December 3, 1981

December

December

December

December

December

December

December

December

December

December

December

3,

3,

7,

,7,

7,

9,

10,

11,

15,

15,

15,

1981

1981

1981

1981

1981

1981

1981

1981

1981

1981

1981

Letter from applicant forwarding Revisions 2 and 3 to

Preservice Inspection Program.

Meeting with applicant to discuss status of work regarding

liquefaction potential beneath the ERCW pipelines and

Class 1E electrical conduit.

Applicant submits Amendment 45 to FSAR.

Letter to applicant concerning NRC volume reduction policy

(Generic 81-39).

Letter to applicant concerning approval of the safeguards

contingency plan and guard training and qualification

plan.

Letter from applicant concerning new or additional

training and procedures to avoid or mitigate voiding in

the reactor vessel head during natural circulation

cooldown.

Letter from applicant concerning monitoring of the diesel

generator engine cooling system.

Letter from applicant concerning hydrodynamic

instability.

Letter to applicant requesting information.

Letter to applicant concerning radiological emergency plan

and procedures.

Letter from applicant concerning steel containment

buckling.

Letter from applicant forwarding report "Evaluation of

Structures, Piping', and Equipment for the 84th Percentile

Earthquake."

Letter from applicant responding to Reactor Systems Branch

questions.

Letter from applicant responding to request for

information.

Site visit to observe boring operations along the ERCW

pipelines.

Letter from applicant responding to questions from

Auxiliary Systems Branch.

Letter from applicant responding to questions.from

Instrumentation and Control Systems Branch.

Watts Bar SER A- 16
December 16, 1981

December 16,

December 16,

December 23,

December 23,

December 24,

.December 28,

December 29,

December 30,

1981

1981

1981

1981

1981

1981

1981

1981

January 7, 1982

Letter to applicant concerning qualifications of reactor

operators - license examinations (Generic 81-40).

Letter from applicant concerning guard training and

qualification plan.

Letter from applicant responding to questions from Hydrologic Engineering Branch.

Letter from applicant concerning applicability of

Westinghouse Topical Report WCAP-7769.

Letter from applicant concerning compliance with GDC 51.

Letter from applicant concerning carbon steel piping

corrosion.

Letter from applicant concerning compliance with 10 CFR 20,

50, and 100.

Letter from applicant responding to questions from

Material Engineering Branch.

Letter to applicant concerning issuance date for NRC

Safety Evaluation Report.

Letter from applicant responsing to questions from Power

Systems Branch.

Letter to applicant concerning new applications survey

(Generic 82-01).

Letter from applicant regarding ASME Section III, NCA 1140

to use ASME Section III, NCA-8240 for stamping of

nameplates.

Letter from applicant concerning acceptance of valves with

nameplate/data reports which have valve ANSI pressure

class rating.

Letter to applicant concerning plans for use of

Westinghouse Model D steam generators.

Letter to applicant requesting additional information.

Letter from applicant concerning masonry walls.

Letter from applicant providing revised response to NRC

question.

Letter from applicant forwarding Physical Security Plan.

Letter from applicant concerning hydrogen measures

required for ice condenser plants.

January

January

12,

19,

1982

1982

January 19, 1982

January

January

January

January

January

January

22,

22,

22,

27,

27,

27,

1982

1982

1982

1982

1982

1982

Watts Bar SER A- 17
January 28, 1982

January 29, 1982

February 1, 1982

February 1, 1982

February 3, 1982

February 3, 1982

February 3, 1982

February 4, 1982

February 5, 1982

February 8, 1982

February 9, 1982

February 12, 1982

February 12, 1982

February 16, 1982

February 17, 1982

February 17, 1982

February 18, 1982

February 22, 1982

February 26, 1982

Letter from applicant responding to request for additional

information.

Letter from applicant responding to concerns in draft SER

Chapter 13.

Letter from applicant forwarding information concerning

sheet pile walls.

Letter from applicant concerning equipment

qualification.

Letter from applicant concerning material and geometric

damping.,

Letter from applicant concerning the proposed waste heat

park.

Letter from applicant concerning the Technical

Specifications-for the flood protection plan.

Applicant submits Amendment 46 to the FSAR.

Letter to applicant forwarding review comments for the

Physical Security Plan.

Letter to applicant concerning nuclear power plant staff

working hours (Generic 82-02).

Letter from applicant forwarding report "Liquefaction of

the ERCW Pipeline Route."

Letter from applicant concerning low temperature

overpressure protection and BOP setpoint methodology.

Letter from applicant responding to questions concerning

reactor systems.

Meeting with applicant to discuss concerns regarding the

steam generator leg.

Meeting with applicant to discuss concerns raised by the

Power Systems Branch.

Letter from applicant concerning control of heavy loads.

Meeting with applicant to discuss the fire protection

program.

Letter from applicant responding to questions concerning

hydrologic engineering.

Letter from applicant concerning borings along Class 1E

conduit.

Watts Bar SER A-18
February 26, 1982 Letter from applicant responding

the initial test program.

to questions concerning

March 1, 1982

March

March

March

March

March

March

March

March

March

March

March

March

March

March

March

March

1, 1982

2, 1982

3, 1982

5, 1982

5, 1982

5, 1982

8, 1982

10, 1982

10, 1982

10, 1982

11, 1982

11, 1982

11, 1982

17, 1982

17, .1982

19, 1982

Letter from applicant concerning the Westinghouse

identified preheater problem for Model D steam

generators.

Letter from applicant concerning applicability of IE

Bulletin 79-27.

Letter from applicant concerning potential liquefaction.

Letter from applicant concerning power systems.

Letter from applicant concerning differential settlement.

Letter from applicant providing comments on draft Safety

Evaluation Report.

Letter from applicant concerning a change in the percent

hydrogen generation to 1.5 percent.

Letter to applicant concerning the Physical Security Plan,

Parts I and 2.

Letter from:applicant concerning the Auxiliary Feedwater

Analysis.

Letter from applicant concerning electronic compensation

for steam generator leg heatup error.

Letter from applicant concerning the inservice inspection

program for the turbine main steam valves and turbine

bypass system.,

Letter to applicant concerning use in INPO SEE-IN Program

(Generic 82-04).

Letter from applicant providing responses to concerns

specified in NUREG-0737.

Letter from applicant concerning applicability of IE

Bulletin 80-06.

Letter from applicant concerning the surveillance

requirements in the high flux shutdown alarm after a

reactor trip.

Letter from applicant concerning compliance with ANSI

N-195-1976 and Reg Guide 1.137 position C.2.

Letter from applicant forwarding report "Liquefaction

Evaluation of the ERCW Pipline Route."

Watts Bar SER A-19
March

March

29, 1982

29, 1982

March 30, 1982

April 2, 1982

April

April

2,

8,

1982

1982

April 9, 1982

April 9, 1982

April

April

April

April

13,

15,

19,

20,

1982

1982

1982

1982

Letter from applicant forwarding revisions to FSAR.

Letter from applicant concerning the fuel handling machine

and reactor vessel head lifting machine.

Letter from applicant concerning the Seismic Qualification

Review Team audit.

Letter to applicant concerning delay of issuance of Safety Evaluation Report.

Letter from applicant concerning differential settlement.

Letter from applicant concerning the prompt notification

system to be used for radiological emergencies.

Letter from applicant informing that fuel load dates have

been changed to August 1983 for Unit 1 and August 1984 for

Unit 2.

Letter from applicant concerning inadvertent containment

spray or air return fan actuation.

Site visit to review physical security.

Letter from applicant concerning compliance with General

.Design Criteria 17 and 18.

Letter to applicant, requesting additional information.

Letter to applicant concerning environmental qualification of safety-related electrical equipment (Generic 82-09).

Letter from applicant concerning seismic response

spectra.

Letter from applicant concerning the NRC Seismic.

Qualification Review Team audit of the plant.

Letter from applicant providing additional information on

the initial test program.

Letter from applicant concerning schedule for finalization

of Technical Specifications.

Meeting with applicant to discuss the Quality Assurance

Program. (Summary issued May 13, 1982).

Meeting with applicant to review the seismic qualification

of equipment.

Letter from applicant concerning safety factors for

sheetpile walls.

April 21, 1982

April 21, 1982

April 21, 1982

April 26, 1982

April 26, 1982

April 27-30, 1982

May 27, 1982

Watts Bar SER A-20
April 30, 1982

May 7, 1982

May 12, 1982

May 13, 1982

May 14, 1982

May 14, 1982

May 19, 1982

May 19, 1982

May 20, 1982

May 21, 1982

June 4, 1982

June 4, 1982

Letter from applicant concerning the availability of a

schedule for the Process Control Program.

Letter to applicant requesting additional information

concerning liquefaction.

Letter from applicant concerning transformer failure at

Sequoyah Nuclear Plant.

Letter from applicant commenting on report "Evaluation

the Buckling Stress Criteria for the Steel Containment

the-Watts Bar Nuclear Reactor.

of

of

Meeting with applicant to discuss potential for

liquefaction beneath the ERCW pipelines and Class 1E

electrical conduit. (Summary issued June 9, 1982.)

Letter from applicant concerning containment purging

during normal plant operation.

Letter from applicant forwarding Revision to the ASME

Section XI Preservice Inspection Program Technical

Instruction TI-50A.

Letter from applicant concerning Power Systems Branch

portion of the questions on the FSAR.

Letter from applicant concerning the initial test

program.

Letter to applicant requesting additional informtion

concerning diesel generator exhaust missile protection.

Letter from applicant concerning the analysis for the

effect of inadvertent containment spray and air return fan

actuation on containment pressure.

Letter from applicant forwarding Revision 2 to ASME

Section XI Preservice Inspection Program Technical

Instruction TI-50B.

June 4, 1982 Letter from applicant concerning evaluation

consequences of a tornado-generated missile

diesel generator exhaust.

of the

impacting the

June 9, 1982

June 9, 1982

June 10, 1982

Letter to applicant concerning initial test program.

Letter to applicant transmitting NUREG-0916 relative

to the restart of R. E. Ginna Nuclear Power Plant

(Generic 82-11).

Letter from applicant forwarding revisions to the Physical

Security Plan.

Watts Bar SER A-21
June

June

June

June

June

II,

11,

11,

11,

14,

1982

1982

1982

1982

1982

Letter from applicant concerning the conceptual design for

the emergency response facilities.

Letter from applicant concerning compliance with General

Design Criteria 17 and 18.

Letter from applicant providing updated response to

NUREG-0737.

Letter from applicant concerning compliance with

Regulatory Guides 1.20, 1,52, and 1.79.

Letter to applicant concerning results of May 14, 1982,

meeting to discuss the liquefaction potential beneath the

ERCW pipelines.

Letter from applicant summarizing the testing performed on

the Tayco Igniter for use in the Permanent Hydrogen

Mitigation System.

Letter from applicant concerning resolution of boron

dilution event concern.

Letter to applicant concerning schedule for review of

technical specifications.

June 14, 1982

June

June

18,

22,

1982

1982

Watts Bar SER A-22
APPENDIX B

BIBLIOGRAPHY

Agbabian Associates, "Correlation of Ground Response Spectra with Modified

Mercalli Site IntensitY," Topical Report SAN/1011-113, 1977.

Atomic Industrial Forum, NESP-0095 Report.

Bollinger, G. A., "The Giles County, Virginia, Seismic Zone Configuration and

Hazard Assessment," Conference on Earthquake and Earthquake Engineering in

the Eastern United States, September 14-16, 1981, Knoxville, TN, 1981.

Boore, D. M., W. B. 'Joyner, A. A. Oliver III, and R. A. Page, "Estimation of

Ground Parameters," Circular 795, U.S. Geological Survey, 1978.

Chinnery, M. A., "A Comparison of the Seismicity of Three Regions of the

Eastern U.S., " Bulletin of the Seismological Society of America, I,

Vol. 69, pp. 757-772, 1979.

Eardley, A. J., Structural Geology of North America, Harper and Rowe, New York,

1962.

Gupta, I. N., and 0. W. Nuttli, "Spatial Attenuation of Intensities for Central

U.S. Earthquakes," Bulletin of the Seismological Society of America,

Vol. 66, pp. 743-752, 1976.

Hadley, J. B. and J. F. Devine, "Seismotectonic Map of the Eastern United

States," U.S. Geological Survey, Department of the Interior Publication

MF-620, 1974.

Joyner, W. B., D. M. Boore, and R. L. Porcella, "Peak Horizontal Acceleration

and Velocity from Strong-Motion Record& Including Records from the 1979

Imperial Valley, California, Earthquake," U.S. Geological Survey, Open-File

Report 81-365, 1981.

King, P. B., "Tectonic Map of North America, "U.S. Geological Survey," Department

of the Interior, 1969.

Nuclear Safety Review Board (NSRB) Charter and the Nuclear Safety Review

Procedures (Power-NSRP 3.1), January 1981.

Nuttli, 0. W., "The Relationship of Sustained Maximum Ground Motion Acceleration

and Velocity to Earthquake Intensity and Magnitude: Miscellaneous Reports

S-73-1," Report 16, U.S. Army Corps of Engineers, 1979.

Watts Bar SER B-1
Nuttli, 0. W., "Similarities and Differences Between Western and Eastern United

States Earthquakes, and Their Consequences for Earthquake Engineering,"

Conference on Earthquakes and Earthquake Engineering in the Eastern United

States, September 14-16, 1981 Knoxville, TN.

Nuttli, 0. W., G. A. Bollinger, and D. W. Griffiths, "On the Relation Between

Modified Mercalli Intensity and Body-Wave Magnitude," Bulletin of the

Seismological Society of America, Vol. 69, 1979.

Tennessee Valley Authority "Justification of the Seismic Design Criteria Used

for the Sequoyah, Watts Bar, and Bellefonte Nuclear Power Plants, Phase II,"

Civil Engineering Branch, Division of Engineering Design, Tennessee.Valley

Authority, 1979.

Topical Report TVA-TR75-IA, "Quality Assurance Program Description of

Design, Construction, and Operation for Nuclear Power Plants."

U.S. Atomic Energy Commission (AEC), "Technical Report on Densification of

Westinghouse PWR Fuel," May 1974.

U.S. Department of Commerce, NOAA, "Climates of the United States," (J. L.

Baldwin) Washington, DC, 1974.

U.S. Nuclear Regulatory Commission, "Interim Safety Evaluation Report on the

Effects of Fuel Rod Bowing on Thermal Margin Calculations for Light Water

Reactors," December 1976.

U.S. General Services Administration, Office of the Federal Register National

Archives and Records Service, Code of Federal Regulations, Title 10, "Energy"

(including General Design Criteria), U.S. Government Printing Office,

Washington, DC, January 1981.

U.S. Nuclear Regulatory Commission, Memorandum and Order CLI-80-21, In the

Matter of "Petition for Emergency and Remedial Action," Washington, DC,

May 23, 1980.

--- , NUREG-75/087 Standard Review Plan for the Review of Safety Analysis

Reports for Nuclear Power Plants - LWR Edition, December 1975.

--- , NUREG-75/087, "Standard Review Plan for the Review of Safety Analysis for

Nuclear Power Plants, "Revision 2, January 1981.

--- , NUREG-0011, "Safety Evaluation Related to Operation of Sequoyah Nuclear

Plants, Units 1 and 2," March 1979.

NUREG-0017, "Calculations of Releases of Radioactive Materials in Gaseous

and Liquid Effluents for Pressurized Water Reactors (PWR-GALE Code),"

April 1976.

Watts Bar SER B-2
NUREG-0085, "The Analysis of Fuel Densification," January 1976.

NUREG-0109, "Occupational Radiation Exposure-at Light Water Cooled Power

Reactor 1969-1975," August 1976.

--- , NUREG-0297, "Safety Evaluation Report on.Westinghouse Electric Company

ECCS Evaluation Model for Plants Equipped with Upper, Head Injection;

April 1978.

--- , NUREG-0303 "Evaluation of.the Behavior of. Waterlogged Fuel Rod Failures in

LWRs," March 1978.

NUREG-0324, "Program for the Meteorolgical Evaluation of Routine Effluent

Releases at Nuclear Power Stations XOQDOQ," J. Goll, Draft Report,

September 1977.

--- , NUREG-0390, "Topical Report Review Status," July 1979.

--- , NUREG-0402, "Analysis of a Worldwide Strong Motion Data Sample to Develop

an Improved Correlation Between Peak Acceleration, Seismic Intensity and

Other Physical Parameters," J. R. Murphy and.L. J. O'Brien, Computer

Sciences Corporation, 1978.

--- , NUREG-0409, "Iodine Behavior in a PWR Cooling System-Following a Postulated

Steam Generator Tube Rupture Accident," January 1978.

--- , NUREG-0452, "Standard Technical Specifications for Westinghouse Pressurized

Water Reactors," Revision of May 1978 and June 1978, Revision 2, July

1980, and Revision 4, November 1981.

--- , NUREG-0460, "Anticipated Transients Without Scram for Light Water Reactors,"

Vol. 3, December 1978,.and Vol.. 4, March-1980..

--- , NUREG-0484, "Methodology for Combining Dynamic Responses," Revision 1,

May 1980.

--- , NUREG-O498,.".Final Environmental Statement, Watts.Bar Nuclear Plant, Units

land 2," December 1978

--- , NUREG-0577, "Potential for Low Fracture Toughness, and. Lamellar Tearing on

PWR Steam Generator and Reactor Coolant Pipe Supports," October 1979.

--- , NUREG-0578, "Analysis of Comments on 10 CFR Part 72," November 1980.

--- , NUREG-0588, "Interim Staff Position on Environmental Qualification of

Safety-Related Electrical Equipment," November 1979.,

--- , NUREG-0609, "Asymmetric Blowdown Loads on PWR Primary Systems," January

1981.

--- , NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants, Resolution of

Generic Technical Activity A-36,"'July 1980.

Watts Bar SER B-3
--- , NUREG-0630, "Cladding Swelling and Rupture Models for LOCA Analysis,"

Draft April 1980.

--- , NUREG-0654, "Criteria for Preparation and Evaluation of Radiological

Emergency Response Plans'and Preparedness in Support of Nuclear Power

Plants," February 1980 (Revision 1, November 1980; Revision 2, October

1981).

--- , NUREG-0660, "NRC Action Plan Developed as a Result of the TMI-2 Accident,"

Vol. 1, May 1980.

--- , NUREG-0696, "Functional Criteria for Emergency Response Facilities," July

1980.

--- , NUREG-0700, "Guidelines for Control Room Design Reviews," September 1981.

--- , NUREG-0731, "Draft Guidelines for Utility Management Structure and

Technical Resources," September 1980.

--- , NUREG-0737, "Clarification of TMI-Action Plan Requirements," November

1980.

--- , NUREG-0800 (formerly NUREG-75/087), "Standard Review Plan for Review of

Safety Analysis Reports for Nuclear Power Plants-,-LWR Edition," (includes

Branch Technical Positions) July 1981.

--- , NUREG/CR-0098, "Development of Criteria for Seismic Review of Selected

Nuclear Power Plants," Consulting Engineering Services, June 1978.

--- , NUREG/CR-0660, "Enhancement of Onsite Emergency Diesel Generator

Reliability," University of Dayton Research Institute, February 1979.

--- , NUREG/CR-1457, "Regional Relationships Among Earthquake Magnitude Scales,"

D. H. and D. L. Bernreuter, Lawrence Livermore National Laboratory,.

September 1980.

--- , NUREG/CR-2165, "An Investigation of Buckling of Steel Cylinders with

Circular Cutouts Reinforced in Accordance with ASME Rules," Los Alamos

Scientific Laboratory, July 1981.

WASH-1270, "Technical Report on Anticipated Transients Without Scram for

Water-Cooled Power Reactors," 'September 1973.

Westinghouse, Topical Report WCAP-7588, Revision 1, "An Evaluation of the Rod

Ejection Accident in Westinghouse Reactors Using Spatial Kinetics Methods,"

December 1971.

--- , Topical Report WCAP-7769, "Overpressure Protection for Westinghouse

Pressurized Reactors," Revision 1, October 8,1971.

--- , Topical Report WCAP-7956, "THINC-IV An Improved Program for ThermalHydraulic Analysis of Rod Bundle Cores," June 1973.

Watts Bar SER B-4
Topical Report WCAP-7982, "Final Densification Penalty Model," October

1972.

Topical Report WCAP-8054, "Application of the THINC-IV Program to PWR

Design," September 1973 (proprietary).

Topical Report WCAP-8078, "Ice Condenser Containment Pressure Transient

Analysis Methods."

Topical Report WCAP-8082-P-A, "Selection of Pipe Breaks for Loss-of-Coolant

Analysis," July 19, 1973 (proprietary).

Topical Report WCAP-8172-A, "Pipe Breaks for the LOCA Analysis of the

Westinghouse Primary Coolant Loops," July 1973.

Topical Report WCAP-8219," Fuel Densification Experimental Results and

Model for Reactor Application," October 1973.

Topical Report WCAP-8279, "Hydraulic Flow Test of 17 x 17 Final Assembly,"

February 1974.

Topical Report WCAP-8288, "Safety Analysis of 17 x 17 Final Assembly for

Combined Seismic and LOCA," December 1973.

Topical Report WCAP-8330, "Westinghouse Anticipated Transients Without

Trip Analysis," August 1974.

Topical Report WCAP-8312A, "Westinghouse Mass and Energy Release

Data for Containment Design," Revision 2, August 1975.

Topical Report WCAP-8346, "Evaluation of Fuel Rod Boring," May 1974.

Topical Report WCAP-8354, "Long Term Ice Condenser Containment Code, LOTIC

Code," April 1976 and Supplement 2.

Topical Report WCAP-8377, "Revised Clad Flattening Model," July 1974.

Topical Report WCAP-8479, "Westinghouse Emergency Core Cooling System

Evaluation Model Application to Plants Equipped With Upper Head Injection,"

October 1979 (proprietary).

Topical Report WCAP-8577, "The Application of Preheat Temperatures After

Welding of Pressure Vessel Steels," February 2, 1976.

Topical Report WCAP-8584, "Failure Mode and Effects Analysis of the

Engineered Safety Features Actuation System."

Topical Report WCAP-8692, "Fuel Rod Bow Evaluation," July 1979.

Topical Report WCAP-8785, "Improved Analytical Models Used in

Westinghouse Final Rod Design Computations, Application for Transient

Analyses, "September 1979.

Watts Bar SER B-5
Topical Report WCAP78963-A, "Safety Analyses for Revised Fuel Rod

Internal Pressure Design Basis, "August 1978 (proprietary).

--- , Topical Report WCAP-9220, "Westinghouse ECCS Evaluation Model," February

1978.

--- , Topical Report WCAP-9226, "Reactor Core Response to Excessive Secondary

Steam Releases, Revision 1, February 1, 1978.

--- , Topical Report WCAP-9292, "Dynamic Fracture Toughness of ASME SA-508

Class 2a and ASME SA-533 Grade A Class 2 Base and Heat-Affected Zone

Material and Applicable Weld Metals."

--- , Topical Report WCAP-9600, "Report on Small Break Accidents for Westinghouse

NSSS System," June 1979.

--- , Topical Report WCAP-9804, "Probabilistic Analysis and Operational Data in

Response to NUREG-0737 Item II.K.3.2 for Westinghouse Plants," February

1981.

Industry Codes and Standards

American Concrete Institute (ACI), Standards 318-63 and 318-71, "Building Code

Requirements for Reinforced Concrete."

--- , Standard 348, "Code Requirement for Nuclear Safety-Related Concrete

Structures."

--- , Standard 531-79.

American Institute of Steel Construction (AISC), "Specifications for the

Design, Fabrication, and Erection of Structural Steel for Building,"

Sixth Edition, 1969.

American National Standards Institute (ANSI), A58-1-1972 "American National

Standard Building Code Requirements for Minimum Design Loads in Buildings

and Other Structures."

--- , B31.1, "Code for Pressure Piping."

--- , B31.1.0, "Power Piping," 1967.

--- , N18.1-1971, "Selection and Training of Nuclear Power Plant Personnel."

--- , N18.7-1971 and N18.7-1972, "Administrative Control and Quality Assurance

for the Operational Phase of Nuclear Power Plants."

--- , N45.2-1971, "Quality Assurance Program Requirements for Nuclear Power

Plants."

--- , N45.2.1-1973, "Cleaning of Fluid Systems and Associated Components During

Construction Phase of Nuclear Power Plants."

Watts Bar SER B-6
--- , N45.2.12, "Requirements for Auditing of Quality Assurance Programs for

Nuclear Power Plants," Draft 3, Revision 4, February 1974.

--- , N101.2-1972, "Protective Coatings (Paints for Light Water Nuclear Reactor

Containment Facilities," 1972.

--- , N101.4, "Quality Assurance for Protective Coatings Applied to Nuclear

Facilities," 1972.

--- , N195, "Fuel oil Systems for Standby Diesel Generators."

--- , N237, "Standard Source Term Specification," March 9, 1976.

American Nuclear Society (ANS), 18.2, "Nuclear Safety Criteria for the Design

of Stationary Pressurized Water Reactor Plants," Draft, August 1970.

American Nuclear Society/American National Standards Institute (ANS/ANSI),

ANS-3.2/ANSI-N18.7-1976, "Administrative Controls and Quality Assurance

for the Operational Phase of Nuclear Power Plants."

American Petroleum Institute (API), 620, "Recommended Rules for Design and

Construction of Large, Welded Low-Pressure Storage Tank," Fourth Edition,

.February 1970.

--- , 650, "Welded Steel Tanks for.Oil Storage," Third Edition, July 1966.

American Society of Civil Engineers (ASCE), Paper No. 3269, "Wind Forces on

Structure," Transcripts of the GSCE, Vol. 126, Part III, 1961.

American Society of Mechanical Engineers (ASME), "Boiler and Pressure Vessel

Code."

--- , Section II, Class 2, 1971 Edition Through Summer of 1973.

--- , Section II, Parts A, B, and C.

--- , Section III, "Nuclear Power Plant Components," 1971 Edition, including

Addenda through Summer of 1972.

--- , Section III, Appendix G, "Protection Against Non-Ductile Failure," 1971

Edition, including Addenda through Summer of 1972.

--- , Section III, Appendix I.

--- , Section III, Class 1.

--- , Section III, Division 1, Class 1, 2,"and 3 Components.

--- , Section III, NB-2160.

--- , Section III, NB-2322.

--- , Section III, NB-2330.

Watts Bar SER B-7
--- , Section III, NB-3120.

-- , Section III, NB-3222.

--- , Section III, NG-2000.

Section III, Summer 1977 Addenda.

"--, Section VIII, Division 1, "Pressure Vessels,"

--- , Section XI, 1974 Edition, including Addenda through Summer of 1975.

--- , Section XI, IWA-1500.

--- , Section XI, IWB-3500.

American Society of Testing Materials (ASTM), E-84, "Surface Burning

Characteristics of Building Materials," 1976.

--- , E-119, "Fire Test of Building Construction and Materials," 1976.

--- , E-185, "Standard Recommended Practice for Surveillance Tests for Nuclear

Reactor Vessels," 1973.

Diesel Engine Manufacturers Association (DEMA), Standards.

Electric Power Research Institute (EPRI), NP-121, Vol. II, April 1976.

NP-933, December 1978.

Institute of Electric and Electronics Engineers (IEEE), 308, 1971 and 1974.

--- , 338, "IEEE Standard Criteria for the Periodic Testing of Nuclear Power

,Generating Station Safety Systems" 1975.

--- , 344, "Recommended Practices for Seismic Qualification of Class 1E

Equipment for Nuclear Power GeneratingStations," 1971, 1974, and 1975.

--- , 384, "IEEE Trial - Use Standard Criteria for Separation of Class 1E

Equipment and Circuits," 1974.

--- , 387, "Standard Criteria fr Diesel Generator Units Applied as Standby Power

Supplies for Nuclear Power Generating Stations," 1977.

National Fire Protection Association (NFPA), 12-1973, "Carbon Dioxide

Extinguishing Systems."

--- , 14-1974, "Standpipe and Hose Systems."

--- , 15-1973, "Water Spray Fixed Systems."

--- , 30-1973, "Flammable and Combustible Liquid Codes."

Watts Bar SER B-8
--- , 72D-1975, "Proprietary Protective Signaling Systems."

--- , 80-1979, "Fire Doors and Windows."

101-1976, "Code for Safety to Life from Fire in Buildings and Structures."

Welding Reserach Council, Bulletin 217.

Watts Bar SER B-9

APPENDIX C

NUCLEAR REGULATORY COMMISSION

UNRESOLVED SAFETY ISSUES

C.1 Unresolved'Safety Issues

The NRC staff continuously evaluates the safety requirements used in its

reviews against new information as it becomes available. Information related

to the safety of nuclear power plants comes from a variety of sources including

experience from operating reactors; research results; NRC staff and Advisory

Committee on Reactor Safeguards (ACRS) safety reviews; and vendor, architect/

engineer, and utility design reviews. Each time a new concern or safety issue

is identified from one or more of these sources, the need for immediate action

to assure safe operation is assessed. This assessment includes consideration

of the generic implications of the issue.

In some cases, immediate action is taken to ensure 'safety, (such as the derating

of boiling water reactors as a result of the channel box wear problems in 1975).

In other cases, interim measures, such as modifications to operating procedures,

may be sufficient to allow further study of the issue before licensing decisions

are made. In most cases, however, the initial assessment indicates that immediate licensing actions or changes in licensing criteria are not necessary. In

any event, further study may be deemed appropriate before the staff can make

judgments as to whether existing NRC staff requirements should be modified to

address the issue for new plants or if backfitting is appropriate for the longterm operation of plants already under construction or in operation.

These issues are sometimes called "generic safety issues" because they are

related to a particular class or type of nuclear facility rather than a specific plant. These issues have also been referred to as "unresolved safety

issues" (NUREG-0410, "NRC Program for the Resolution of Generic Issues Related

to Nuclear Power Plants," dated January 1, 1978). However, as discussed above,

such issues are considered on a generic basis only after the staff has made an

initial determination that the safety significance of the issue does not

prohibit continued operation or require licensing actions while the longer term

generic review is underway.

C.2 ALAB-444 Requirements

These longer term generic studies were the subject of a Decision by the Atomic

Safety and Licensing Appeal Board of the Nuclear Regulatory Commission. The

Decision was issued on November 23, 1977 (ALAB-444) in connection with the

Appeal Board's consideration of the Gulf States Utility Company application for

the River Bend Station Units 1 and 2.

Watts Bar Ser C-1
In the view of the Appeal Board,

In short, the board (and the public as well) should be in a position

to ascertain from the SER itself--without the need to resort to extrensic documents--the staff's perception of the nature and extent

of the relationship between each significant unresolved generic safety question and the eventual operation of the reactor under scrutiny.

Once again, this assessment might well have a direct bearing upon the

ability of the licensing board to make the safety findings required

of it on the construction permit level even though the generic answer

to the question remains in the offing. Among other thi'ngs, the furnished information would likely shed light on such alternatively

important considerations as whether: (1) the problem has already

been resolved for the reactor under study; (2) there is a reasonable

basis for concluding that a satisfactory solution will be obtained

before the reactor is put in operation; or (3) the problem would have

no safety implications until after several years of reactor operation

and, should it not be resolved by then, alternative means will be

available to ensure that continued operation (if permitted at all)

would not pose an undue risk to the public.

This appendix is specifically included to respond to the decision of the Atomic

Safety and Licensing Appeal Board as enunciated in ALAB-444, and as applied to

an operating license proceeding Virginia Electric and Power Company (North Anna

Nuclear Power Station, Unit Nos I and 2), ALAB-491, NRC 245 (1978).

C.3 Unresolved Safety Issues

In a related matter, as a result of Congressional action on the NRC budget for

Fiscal Year 1978, the Energy Reorganization Act of 1974 was amended (PL 95-209)

on December 13, 1977 to include, among other things, a new Section 210 as

follows:

UNRESOLVED SAFETY ISSUES PLAN

SEC. 210. The Commission shall develop a plan providing for specification and analysis of unresolved safety issues relating to nuclear

reactors and shall take such action as may be necessary to implement

corrective measures with respect to such issues. Such plan shall be

submitted to the Congress on or before January 1, 1978 and progress

reports shall be included in the annual report of the Commission

thereafter.

The Joint Explanatory Statement of the House-Senate Conference Committee for

the Fiscal Year 1978 Appropriations Bill (Bill S.1131) provided the following

additional information regarding the Committee's deliberations on this portion

of the bill:

SECTION 3 - UNRESOLVED SAFETY ISSUES

The House amendment required development of a plan to resolve

generic safety, issues. The conferees agreedto a requirement

that the plan be submitted to the Congress on or before

Watts Bar Ser C-2
January 1, 1978. The conferees also expressed the intent that

this plan should identify and describe those safety issues,

relating to nuclear power reactors, which are unresolved on

the date of enactment. It should set forth: (1) Commission

actions taken directly or indirectly to develop and implement

corrective measures; (2) further actions planned concerning such

measures; and (3) timetables andcost estimates of such actions.

The Commission should indicate the priority it has assigned to

each issue, and the basis on which priorities have been assigned.

In response to the reporting requirements of the new Section 210, the NRC staff

submitted to Congress on January 1, 1978, a report (NUREG-0410), "NRC Program

for the Resolution of Generic Issues Related to Nuclear Power Plants," that

describing the NRC generic issues program. The NRC program was already in

place when PL 95-209 was enacted and is of considerably broader scope than the

"Unresolved Safety Issues Plan" required by Section 210. In the letter

transmitting NUREG-0410 to the Congress on December 30, 1977, the Commission

indicated that "The progress reports, which are required by Section 210 to be

included in future NRC annual reports, may be more useful to Congress if they

focus on the specific Section 210 safety items."

It is the NRC's view that the.intent of Section 210 was to ensure that plans

were developed and implemented on issues with potentially significant public

safety implications. In 1978, the NRC undertook a review of more than 130

generic issues addressed in the NRC program to determine which issues fit this

description and qualify as "Unresolved Safety Issues" for reporting to the

Congress.

The NRC review included the development of proposals by the NRC staff and

review and final approval by the NRC Commissioners.

This review is described in NUREG-0510, "Identification of Unresolved Safety

Issues Relating to Nuclear Power Plants - A Report to Congress," dated January

1979. The report provides the following definition of an "Unresolved Safety

Issue":

An Unresolved Safety Issue is a matter affecting a number

of nuclear power plants that poses important questions concerning the adequacy of existing safety requirements for

which a final resolution has not yet been developed and

that involves conditions not likely to be acceptable over

the lifetime of the plants it affects.

Further, the report indicates that in applying this definition, matters that

pose "important questions concerning the adequacy of existing safety requirements" were judged to be those for which resolution is necessary to (1) compensate

for a possible major reduction in the degree of protection of the public health

and safety or (2) provide a potentially significant decrease in the risk to

the public health and safety. Quite simply, an "Unresolved Safety Issue" is

potentially significant from a public safety standpoint, and its resolution is

likely to result in NRC action on the affected plants.

Watts Bar Ser C-3
All of the issues addressed in the NRC program were systematically evaluated

against this definition as described in NUREG-0510. As a result, 17 "Unresolved Safety Issues" addressed by 22 tasks in the NRC program were identified.

The issues are listed below. Progress on these numbers was first discussed in

the 1979 NRC Annual Report. The number(s) of the generic task(s) (for example,

A-i) in the NRC program addressing each issue is indicated in parentheses

following the title.

UNRESOLVED SAFETY ISSUES (APPLICABLE TASK NOS.)

1. Waterhammer (A-i)

2. Asymmetric Blowdown Loads on the Reactor Coolant System (A-2)

3. Pressurized Water Reactor Steam Generator Tube Integrity (A-3, A-4, A-5)

4. BWR Mark I and Mark II Pressure Suppression Containments (A-6, A-7, A-8,

A-39)

5. Anticipated Transients Without Scram (A-9)

6. BWR Nozzle Cracking (A-10)

7. Reactor Vessel Materials Toughness (A-11)

8. Fracture Toughness of Steam Generator and Reactor Coolant Pump

Supports (A-12)

9. Systems Interaction in Nuclear Power Plants (A-17)

10. Environmental Qualification of Safety-Related Electrical Equipment (A-24)

11. Reactor Vessel Pressure Transient Protection (A-26)

12. Residual Heat Removal Requirements (A-31)

13. Control of Heavy Loads Near Spent Fuel (A-36)

14. Seismic Design Criteria (A-40)

15. Pipe Cracks at Boiling Water Reactors (A-42)

16. Containment Emergency Sump Reliability (A-43)

17. Station Blackout (A-44)

In the view of the staff, the "Unresolved Safety Issues" listed above are the

substantive safety issues referred to by the Appeal Board in ALAB-444 when it

spoke of "...those generic problems under continuing study which have...

potentially significant public safety implications." Of the 22 tasks identified

with the "Unresolved Safety Issues," 8 are not applicable to Watts Bar Units 1

and 2, and 6 of these 8 tasks (A-6, A-7, A-8, A-39, A-lO, and A-42) are peculiar

to boiling water reactors. Tasks A-4 and A-5 address steam generator tube problems in Combustion Engineering and Babcock and Wilcox plants. With regard to

the remaining 14 tasks that are-applicable to this facility, the NRC staff has

issued NUREG reports providing its proposed resolution of 5 of these issues.

Each of these has been addressed in this Safety Evaluation Report or will be

addressed in a future supplement. The table below lists those issues and the

section of this Safety Evaluation Report in which they are discussed.

Safety Evaluation

Task Number NUREG Report and Title Report Section

A-2 NUREG-0609, "Asymmetric Blowdown Loads 3.9.2

on PWR Primary Systems" 4.2.2

Watts Bar Ser C-4
A-9 NUREG-0460, Vol. 4, "Anticipated Transients 15.3.6

Without Scram for Light Water Reactors"

A-24 NUREG-0588, "Interim Staff Position on 3.11

Environmental Qualification of SafetyRelated Electrical Equipment"

A-26 NUREG-0224, "Reactor Vessel Pressure Transient 5.2.2

Protection for Pressurized Water Reactors" and

RSB BTP 5-2

A-31 NUREG-0800 (SRP 5.4.7 and BTP 5-1, "Residual 5.4.3

Heat Removal Systems" incorporate requirements

of USI A-31.)

A-36 NUREG-0612, "Control of Heavy Loads at 9.1.4

Nuclear Power Plants"

The remaining issues applicable to this facility are listed below.

GENERIC TASKS ADDRESSING UNRESOLVED SAFETY ISSUES

THAT ARE APPLICABLE TO WATTS BAR UNITS 1 AND 2

1. A-1 Waterhammer

2. A-3 Westinghouse Steam Generator Tube Integrity

3. A-11 Reactor Vessel Materials Toughness

4. A-12 Potential for Low Fracture Toughness and Lamellar Tearing

on PWR Steam Generator and Reactor Coolant Pump Supports

5. A-17 Systems Interactions in Nuclear Power Plants

6. A-40 Seismic Design Criteria

7. A-43 Containment Emergency Sump Reliability

8. A-44 Station Blackout

With the exception of Tasks A-43, and A-44, Task Action Plans for the generic

tasks above are included in NUREG-0649, "Task Action Plans for Unresolved

Safety Issues Related to Nuclear Power Plants." The Task Action Plan for Task

A-43 was issued in January 1981, and the Task Action Plan for A-44 was issued

in July 1980. Draft NUREG-0577, which represents staff resolution of USI A-12,

was issued for comment in November 1979. The Draft NUREG contained the Task

Action Plan for A-12. The information provided in NUREG-0649 meets most of the

informational requirements of ALAB-444. Each Task Action Plan provides a

description of the problem; the staff's approaches to its resolution; a general

discussion of the bases upon which continued plant licensing or operation can

proceed pending completion of the task; the technical organizations involved in

the task and estimates of the personnel required; a description of the interactions with other NRC offices, the Advisory Committee on Reactor Safeguards,

and outside organizations; estimates of funding required for contractorsupplied technical assistance; prospective dates for completing the task; and a

description of potential problems that could alter the planned approach on

schedule.

In addition to the Task Action Plans, the staff issues the "Office of Nuclear

Reactor Regulation Unresolved Safety Issues Summary, Aqua Book" (NUREG-0606) on

Watts Bar Ser C-5
a quarterly basis, which provides current schedule information for each of the

"Unresolved Safety Issues." It also includes information relative to the

implementation status of each "Unresolved Safety Issue" for which technical

resolution is complete.

The staff has reviewed the eight "Unresolved Safety Issues" listed above as

they relate to Watts Bar. Discussion of each of these issues, including

references to related discussions in the SER is provided in Section C.5. Based

on its review of these items, the staff concluded, for the reasons set forth in

Section C.5, that there is reasonable assurance that this facility can be

operated before the ultimate resolution of these generic issues without

endangering the health and safety of the public.

C.4 New "Unresolved Safety Issues"

An indepth and systematic review of generic safety concerns identified since

January 1979 has been performed by the staff to determine if any of these

issues should be designated as new "Unresolved Safety Issues." The candidate

issues originated from concerns identified in NUREG-0660, "NRC Action Plan as a

Result of the TMI-2 Accident"; ACRS recommendations; abnormal occurrence

reports; and other operating experience. The staff's proposed list was

reviewed and commented on by the ACRS, the Office of Analysis and Evaluation of

Operational Data (AEOD), and the Office of Policy Evaluation. The ACRS and

AEOD also proposed that several additional "Unresolved Safety Issues" be considered by the Commission. The Commission considered the above information

and approved the following four new "Unresolved Safety Issues":

A-45 Shutdown Decay Heat Removal Requirements

A-46 Seismic Qualfication of Equipment in Operating Plants

A-47 Safety Implications of Control Systems

A-48 Hydrogen Control Measures and Effects of Hydrogen Burns on SafetyEquipment

In addition to the four issues identified above, the Commission approved A-49,

"Pressurized Thermal Shock," as another USI in December 1981.

A description of the review process for candidate issues, together with: a list

of the issues considered is presented in NUREG-0705, "Identification of New

Unresolved Safety Issues Relating to Nuclear Power Plants, Special Report to

Congress," dated March 1981. An expanded discussion of each of the new "Unresolved Safety Issues" is also contained in NUREG-0705.

The applicability and bases for licensing prior to ultimate resolution of-the

five new USIs for Watts Bar are discussed in Section C.5.

C.5 Discussion of Tasks as They Relate to Watts Bar Units 1 and 2

This section provides the NRC staff's evaluation of the Watts Bar facilities

for each of the applicable "Unresolved Safety Issues." This includes the staff

bases for licensing prior to ultimate resolution of these issues with the

exception of USIs A-3 and A-46. The open issues in these areas are currently

Watts Bar Ser C-6
under review and will be addressed in a supplement to the SER. The staff

conclusions are based in part on information provided by the applicant in his

letter of September 11, 1981.

A-i Waterhammer

Waterhammer events are intense pressure pulses in fluid systems caused by any

one of a number of mechanisms and'systems conditions. Since 1971 there have

been over 100 incidents involving waterhammer in pressurized water reactors and

boiling water reactors. The waterhammers have involved steam generator feed-

.rings and piping, decay heat removal systems, emergency core cooling systems,

containment spray lines, service water lines, feedwater lines, and steamlines.

However, the systems most frequently affected by waterhammer effects are the

feedwater systems. The most serious waterhammer events have occurred in the

steam generator feedrings of pressurized water reactors. These types of waterhammer events are addressed in Sections 10.4.7 and 10.4.9 of this SER. System

design changes and testing requirements necessary to prevent this type of waterhammer have been implemented by TVA in accordance with recommendations provided

in the Westinghouse report "Minimization of Series D2/D3 Steam Generator Preheating Pressure Transient," dated June 30, 1977.

Under Generic Task A-i, the potential for waterhammer in various systems is

being evaluated and appropriate requirements and systematic review procedures

are being developed to ensure that waterhammer is given appropriate consideration in all areas of licensing review. A technical report, NUREG-0582,

"Waterhammer in Nuclear Power Plants" (July 1979), providing the results of an

NRC staff review of waterhammer events in nuclear power plants and stating

current staff licensing positions, completes a major subtask of Generic Task

A-i.

With regard to protection against other potential waterhammer events currently

provided in plants, piping design codes require consideration of impact loads.

Approaches used at the design stage include: (1) increased valve closure

times, (2) piping layout precluding water slugs in steam lines and vapor

formation in water lines, (3). use of snubbers and pipe hangers, and (4) use of

vents and drains. In addition, the staff requires that the applicant conduct a

preoperational vibration dynamic effects test program in accordance with

Section III of the ASME Code for all ASME Class 1 and Class 2 piping systems

and piping restraints during startup and initial operation. These tests will

provide adequate assurance that the piping and piping restraints have been

designed to withstand dynamic effects due to valve closures, pump trips, and

other operating modes associated with the design operational transients.

Nonetheless, in the unlikely event that a large pipe break did result from a

severe waterhammer event, core cooling is ensured by the emergency core cooling

systems, and protection against the dynamic effects of such pipe breaks inside

and outside of containment is provided.

In the June 10, 1981 briefing on the McGuire plant concerning the steam void

developing under the vessel head during normal cooldown, the concern was raised

regarding the potential for developing damaging waterhammer if the operator

were to turn on the UHI system after the void was established in the upper

head. It was speculated that after the void was formed the UHI-pipes would

Watts Bar Ser C-7
drain and fill with steam. On turning on the UHI the supposition was that the

cold-water/steam interaction could potentially break the UHI pipes.

Under the conditions of the supposition, the pressure differential across the

UHI system after the formation of the steam bubble is several hundred psi.

Under these conditions, there is no evident mechanism for the formation of an

opposing water slug. The injection water forms a single, high velocity, slug

sweeping and condensing the steam ahead of it. While there are no experimental

data or detailed analyses to prove or disprove the staff's judgment, the'staff

concluded that this situation would generate reaction forces estimated to be at

least an order of magnitude lower than waterhammer forces.

On reviewing the conditions of the normal cooldown, a somewhat artificial

scenario can be postulated in which the UHI pipes are submerged again, after

draining and filling with steam, so that a water seal is formed at the lower

end of the pipe. Upon actuation of the UHI, a steam bubble can be trapped and

condensed in the pipe allowing a siphon action to pull up water from the water

seal to form an opposing slug. This slug can impact the incoming UHI jet to

create a classical waterhammer. This scenario, however, implies a smaller

bubble than usual in the vessel head for which there is virtually no incentive

for the operator to attempt bubble condensation. Moreover, administrative

procedures will preclude this type of operator action. The review of the

applicant's natural circulation procedures does not include acceptance of UHI

actuation as a viable procedural method to collapse the void.

Moreover, Westinghouse provides their licensees with guidelines for dealing

with the vessel head steam void that are based on methods other than the use of

the UHI system.

Notwithstanding the above considerations, in the remote event that the UHI

pipes are ruptured, the event represents a small to intermediate hotleg break

that is well within the spectrum of LOCA analyses required for all plants and

which represents no unreviewed safety issue.

Therefore, the waterhammer scenario postulated as a possible result of the

operation of the UHI during normal cooldown and depressurization appears not to

be a significant safety issue.

If Task A-1 identifies some other potentially significant waterhammer scenarios

that have not explicitly been accounted for in the design and operation of

Watts Bar, corrective measures will be required at that time. The task has not

yet identified the need for requiring any additional measures beyond those

already implemented.

Based on the foregoing, the staff has concluded that Watts Bar can be operated

before ultimate resolution of this generic issue without undue risk to the

health and safety of the public.

A-3 Westinghouse Steam Generator Tube Integrity

The primary concern is the capability of steam generator tubes to maintain

their integrity during normal operation and postulated accident conditions. In

addition, the requirements for increased steam generator tube inspections and

Watts Bar Ser C-8
repairs have resulted in significant increases in occupational exposures to

workers. Corrosion resulting in steam generator tube wall thinning (wastage)

has been observed in several Westinghouse plants for a number of years. Plants

operating exclusively with an all volatile secondary water treatment process

have not experienced this form of degradation to date. The Watts Bar units

will operate on an all-volatile secondary water treatment process. Another

major corrosion-related phenomenon has also been observed in a number of plants

in recent years, resulting from a buildup of support plate corrosion products

in the annulus between the tubes and the support plates. This buildup eventually causes a diametral reduction of the tubes, called "denting," and deformation

of the tube support plates. This phenomenon has led to other problems, including stress corrosion cracking, leaks at the tube/support plate intersections,

and U-bend section cracking of tubes which were highly stressed because of

support plate deformation.

Specific measures such as steam generator design features and a secondary water

chemistry control and monitoring program, that the applicant has employed to

minimize the onset of steam generator tube problems are described in Section 10.2

of this SER. In addition, Section 5.4.2 discusses the inservice inspection

requirements. As described in Section 10.2, the applicant has met all current

requirements regarding steam generator tube integrity. The Technical Specifications will include requirements for actions to be taken in the event that steam

generator tube leakage occurs during plant operation.

The Watts Bar units use the Model D-3 steam generat6r design. As discussed in

Section 5.4.2.2 of this SER, a problem concerning the potential for tube

degradation caused by flow-induced vibration in the preheater section of

Model D-3 steam generators for power levels greater than 50% has been identified in nondomestic facilities. A program for resolution of this problem is

underway by Westinghouse. The program includes the use of scale model testing,

analytical studies, wear testing, periodic eddy current testing of operating

units, and the use of diagnostic internal and external instrumentation in

selected operating units. However, since the Watts Bar facility is scheduled

for startup no earlier than August 1983, it is anticipated that information

will be available from the Westinghouse program that can be factored into

procedures, controls, and necessary hardware modifications.

TVA is a member of the Electric Power Research Institute (EPRI) and the Steam

Generator Owners Group.(SGOG). TVA representatives regularly participate in

these meetings to maintain awareness of industry experience and obtain updates

on research in progress. In addition, experience will be constantly monitored,

along with the results of the numerous DOE, NRC, EPRI, and SGOG research programs now underway.

Special attention during preoperational testing will be given to tube vibration

and potential wear as a result of movement in the tube support sheets and

antivibration bars.

Task A-3 is expected to result in improvements in current staff requirements

for inservice inspection of steam generator tubes. These improvements will

include a better statistical basis for inservice inspection program requirements and consideration of the cost/benefit of increased inspection. Pending

completion of Task A-3, the measures taken at this facility should minimize the

Watts Bar Ser C-9
steam generator tube problems encountered. Further, the inservice inspection

and Technical Specification requirements will ensure that the applicant and the

staff are alerted to tube degradation should it occur.

Appropriate actions such as tube plugging, increased and more frequent

inspections, and power derating could be taken if necessary. Because the

improvements that will result from Task A-3 will be procedural (that is an

improved inservice inspection program), they can be implemented by the

applicant after operation of this facility begins, if necessary.

Based on the foregoing, the staff has concluded that Watts Bar can be operated

before ultimate resolution of this generic issue without undue risk to the

health and safety of the public, with the exception of the concern regarding

tube degradation potential in the preheater section of the Model D-3 steam

generator for power levels greater than 50%. The staff will report on the

results of its evaluation in a supplement to this SER.

A-11 Reactor Vessel Materials Toughness

Resistance to brittle fracture, a rapidly propagating catastrophic failure mode

for a component containing flaws, is described quantitatively by a material

property generally denoted as "fracture toughness." Fracture toughness has

different values and characteristics depending upon the material being considered. For steels used in a nuclear reactor pressure vessel, three considerations are important; first, fracture toughness increases with increasing

temperature; second, fracture toughness decreases with increasing load rates;

and third, fracture toughness decreases with neutron irradiation. In recognition of these considerations, power reactors are operated within restrictions

imposed by the Technical Specifications on the pressure during heatup and

cooldown operations. These restrictions ensure that the reactor vessel will

not be subjected to a combination of pressure and temperature that could cause

brittle fracture of the vessel if there were significant flaws in the vessel

materials. The effect of neutron radiation on the fracture toughness of the

vessel material is accounted for in developing and revising these Technical

Specification limitations.

For the service times and operating conditions typical of current operating

plants, reactor vessel fracture toughness for most plants provides adequate

margins of safety against vessel failure under operating, testing, maintenance,

and anticipated transient conditions, and accident conditions over the life of

the plant. However, results from a reactor vessel surveillance program and

analyses performed for up to 20 older operating pressurized water reactors and

those for some more recent vintage plants will have marginal toughness, relative to required margins at normal full power after comparatively short period

of operation. In addition, results from analyses performed by pressurized

water reactor manufacturers indicate that the integrity of some reactor vessels

may not be maintained in the event that a main steam line break or a loss-ofcoolant accident occurs after approximately 20 years of operation. The principal objective of Task A-11 is to develop an improved engineering method and

safety criteria to allow a more precise assessment of the safety margins that

are available during normal operation and transients in older reactor vessels

with marginal fracture toughness and of the safety margins available during

accident conditions for all plants.

Watts Bar Ser C-10
The staff's review of the fracture toughness of the Watts Bar Unit 1 reactor

vessel is incomplete at this time. However, based upon its evaluation to date

of the reactor vessel materials toughness, the staff has concluded that this

unit will have adequate safety margins for initial operation against brittle

failure during operating, testing, maintenance, and anticipated transient

conditions. The Unit 2 reactor vessel meets the fracture toughness requirements of Appendix G to 10 CFR 50. Therefore, it is expected to meet the

specified safety margins through its life. When Task Action Plan A-11 is

completed and explicit fracture evaluation criteria for accident conditions are

defined, all vessels will be reevaluated for acceptability over their design

lives. Since Task A-11 is projected to be completed well in advance of the

Watts Bar reactor vessel reaching a level of marginal fracture resistance,

acceptable vessel integrity for the postulated accident conditions will be

assured at least until the reactor vessel is reevaluated for long-term'acceptability. The staff's evaluation for long-term acceptability will be provided in

a supplement to the Safety Evaluation Report.

In addition, the surveillance program required by 10 CFR 50, Appendix H will

afford an opportunity to reevaluate the fracture toughness periodically during

the first half of design life.

Therefore, based upon the foregoing, the staff has concluded that Watts Bar can

be operated prior to resolution of this generic issue without undue risk to the

health and safety of the public.

A-12 Potential for Low Fracture Toughness and Lamellar Tearing on PWR

Steam Generator and Reactor Coolant Pump Supports

During the course of the licensing action for North Anna Power Station Units 1

and 2, a number of questions were raised as to the potential for lamellar

tearing and low fracture toughness of the steam generator and reactor coolant

pump support materials for those facilities. Two different steel specifications (ASTM A36-70a and ASTM A572-70a) covered most of the material used for

these supports. Toughness tests, not originally specified and not in the

relevant ASTM specifications, were made on those heats for which excess material

was available. The toughness of the A36 steel was found to be adequate, but

the toughness of the A572 steel was relatively poor at an operating temperature

of 80'F.

Because similar materials and designs have been used on other nuclear plants,

the concerns regarding the supports for the North Anna facilities are

applicable to other PWR plants. It was therefore necessary to reassess the

fracture toughness of the steam generator and reactor coolant pump support

materials for all operating PWR plants and those in CP and OL review.

NUREG-0577, "Potential for Low Fracture Toughness and Lamellar Tearing on PWR

Steam Generator and Reactor Coolant Pump Supports," was issued for comment in

November 1979. This report summarizes work performed by the NRC staff and its

contractor, Sandia Laboratories, in the resolution of this generic activity.

The report describes the technical issues, the technical studies performed by

Sandia Laboratories, the NRC staff's technical positions based on these studies,

and the NRC staff's plan for implementing its technical positions. As a part

Watts Bar Ser C-11
of initiating the implementation of the findings in this report, letters were

sent to all applicants and licensees on May 19 and 20, 1980. In these letters,

a revised proposed implementation plan was presented and specific criteria for

material qualifications were defined.

Many comments on both the draft of NUREG-0577 and the letters of May 19 and 20

have been received by the NRC staff and detailed consideration is presently

being given to these comments. After completing its review and analysis of the

comments provided, the staff will issue the final revision of NUREG-0577, which

will include a full discussion and resolution of the comments and a final plan

for implementation.

The staff estimates that its implementation review will require approximately 2

years. Because many factors (initiating event, low fracture toughness in a

critical support member in tension, low operating temperature, large flaw) must

be simultaneously present for failure of the support system, the staff has

determined that licensing for pressurized water reactors should continue during

the implementation phase. The staff conclusions regarding licensing and subsequent operation are not sensitive to the estimated length of time required

for this work.

TVA is currently reviewing the materials used in all primary reactor coolant

system component (reactor vessel, steam generators, reactor coolant pumps, and

pressurizer) supports to determine the degree of agreement with the proposed

criteria contained in NUREG-0577. Also, TVA is participating in the recently

formed Atomic Industrial Forum (AIF) Subcommittee on Material Requirements and

with.the Metals Property Council to develop an industry response to the generic

issue. A test program has been initiated at Sequoyah to monitor actual support

temperatures. This information will be extrapolated to the Watts Bar supports

and will be utilized in the evaluation of the support.

In the interim, until the fracture toughness of the support materials at Watts

Bar is evaluated and corrective measures implemented, if needed, operation is

justified based on the following.

Support failures from inadequate fracture toughness are not expected to occur

except under the unlikely combination of

(1) the initiating event (that is,a large pipe break) that has a low

probability of occurrence

(2) the existence of nonredundant and critical support structural members(s)

with low fracture toughness (many supports contain redundant members)'

(3) the existence of support structural members at operating temperatures low

enough that the, fracture toughness of the support material is reduced to a

level at which brittle failure could occur if a large flaw existed

(4) the existence of a flaw of such size that the stresses imparted during the

initiating event could cause the flaw to rapidly propagate, resulting in

brittle failure of the member(s)

Watts Bar Ser C- 12
With regard to the lamellar tearing issue, the results of an extensive

literature survey by Sandia revealed that, although lamellar tearing is a

common occurrence in structural steel construction, virtually no documentation

exists describing inservice failures as a result of lamellar tearing.

Nonetheless, additional research is recommended to provide a more definitive

and complete evaluation of the importance of lamellar tearing to the structural

integrity of nuclear power plant support systems.

Based on its review, the staff has concluded that there is reasonable assurance

that Watts Bar can be operated before ultimate resolution of this generic issue

without endangering the health and safety of the public.

A-17 Systems Interaction in Nuclear Power Plants

The staff's systems interaction program was initiated in May 1978 with the

definition of USI A-17 (Systems Interaction in Nuclear Power Plants) and was

intensified by TAP (NUREG-0660) Item II.C.3 (Systems Interaction). The concern

arises because the design, analysis, and installation of systems are frequently

the responsibility of teams of engineers with functional specialties--such as

civil, electrical, mechanical, or nuclear. Experience at operating plants has

led to questions of whether the work of these functional specialists is sufficiently integrated to'enable them to minimize adverse interactions among

systems. Some adverse events that occurred in the past might have been prevented if the teams had ensured that there was necessary independence of safety

systems under all conditions of operation.

TVA has not'described a comprehensive program that separately evaluates all

structures, systems, and components important to safety for the three

categories of adverse systems interactions (spatially coupled, functionally

coupled, and humanly coupled)., However, there is assurance that Watts Bar can

be operated without endangering the health and safety of the public. The plant

has been evaluated against current licensing requirements that are founded on

the principle of defense in depth. Adherence to this principle results in

requirements such as physical separation and independence of redundant safety

systems and protection against hazards such as high energy line ruptures,

missiles, high winds, flooding, seismic events, fires, human factors, and

sabotage. These design provisions are subject to review against the SRP which

requires interdisciplinary reviews of safety-grade equipment and addresses

different types of potential systems interactions. Also, the quality assurance

program which is followed during the design, construction, and operational

phases for each plant contributes to the prevention of introducing adverse

systems interactions. Thus, the current licensing requirements and procedures

provide an adequate degree of plant safety.

The NRC staff's current procedures assign primary responsibility for review of

various technical areas to specific organizational units and assign secondary

responsibility to other units, where there is a functional interface. Designers

follow somewhat similar procedures and provide the analyses of systems and

interface reviews. Task A-17 has been developing methods that could identify

adverse systems interactions which were not considered by current review procedures. The first phase of this study began in May 1978 and was completed in

February 1980 by Sandia Laboratories under contact to the NRC staff.

Watts Bar Ser C-ý13
The Phase I investigation was structured to identify areas where interactions

are possible between systems and have the potential of negating or seriously

degrading the performance of safety functions. The study concentrated on

commonly caused failures among systems that would violate a safety function.

The investigation was to then identify where NRC review procedures may not have

properly accounted for these interactions.

Sandia Laboratories used fault-tree analysis on the Watts Bar design to identify component failure combinations (cut-sets) that could result in loss of

a safety function. The cut-sets were further reduced by incorporating six

linking systems failures into the analysis. The results of the Sandia effort

indicated a few potentially adverse systems interactions within the limited

scope of the study. The staff reviewed the interactions for safety significance and generic implications. The staff concluded that no corrective

measures needed to be implemented immediately, except for the potential interaction between the PORV and its block valve. This interaction had been separately identified by the evaluations of the TMI-2 accident while Sandia was

studying Watts Bar. Because corrective measures were already being implemented, no separate measures were needed under USI A-17.

NUREG-0660 provides for a systems interaction follow-on study in Section II.C.3,

"Systems Interactions." Since April 1980, the NRR has intensified the effort

both by broadening the study of methods to identify potential systems interactions and by preparing guidance for audit reviews of selected plants for

systems interactions. Recent experience provides a basis from which the staff

is developing a more efficient review process for potential system interactions.

The process will provide for a resolution of USI A-17, assimilate operating

reactor experience, and rank identified systems interactions by their relative

importance to safety.

It is expected that the development of systematic ways to identify, rank, and

evaluate systems interactions will further reduce the likelihood of intersystem

failures that result in the loss of plant safety functions. A comprehensive

program is expected to employ analytical methods, visual inspections, experience

feedback, and simulator dependencies experiments. The LWR industry's current

experience with systems interaction reviews is fragmented. Experience like

that gained by the Phase I study is an essential ingredient to the staff's

considerations of a comprehensive systems interaction program. After the

resolution of USI A-17, the staff will determine whether TVA must perform

further evaluations for adverse systems interaction.

A-40 Seismic Design Criteria -- Short-Term Program

NRC regulations require that nuclear power structures, systems, and components

important to safety be designed to withstand the effects of natural phenomena

such as earthquakes. Detailed requirements and guidance regarding the seismic

design of nuclear plants are provided in the NRC regulations and in Regulatory

Guides. However, there are a number of plants with construction permits and

operating licenses issued before the NRC's current regulations and Regulatory

Guidance were in place. For this reason, rereviews of the seismic design of

various plants are being undertaken to ensure that these plants do not present

an undue risk to the public. Task A-40 is, in effect, a compendium of shortterm efforts to support such reevaluation efforts of the NRC staff, especially

Watts Bar Ser C-1.4
those related to older operating plants. In addition, some revisions to the

SRP sections and Regulatory Guides to bring them more in line with the stateof-the-art will result.

As discussed in Section 3.7 of this SER, the seismic design basis and seismic

design of the facility have been evaluated at the operating license stage, and

have been found acceptable. The staff does not expect the results of Task A-40

to affect these conclusions, because the techniques under consideration are

essentially those utilized in the review of this facility. Should the resolution of Task A-40 indicate a change is needed in licensing requirements, all

operating reactors, including Watts Bar, will be re-evaluated on a case-bycase basis. Accordingly, the staff has concluded that this facility can be operated before the ultimate resolution of this generic issue without endanger- ing the health and safety of the public.

A-43 Containment Emergency Sump Reliability

Following a postulated loss-of-coolant accident (a break in the reactor coolant

system piping), the water flowing from the break would be collected in-the

emergency sump at the low point in the containment. This water would be recirculated through the reactor system by the emergency core cooling pumps to

maintain core cooling. This water would also be circulated through the containment spray system to remove heat and fission products from the containment.

Loss of the ability to draw water from the emergency sump could disable the

emergency core cooling and containment spray systems.

One postulated means of losing the ability to draw water from the emergency

sump could be blockage by debris. A principal source of such debris could be

the thermal insulation on the reactor coolant system piping. In the event of a

piping break, the subsequent violent release of the high pressure water in the

reactor coolant system could rip off the insulation in the area of the break.

This debris could then be swept into the sump, potentially causing blockage.

Currently, regulatory positions regarding sump design are presented in Regulatory Guide 1.82, "Sumps for Emergency Core Cooling and Containment Spray Systems," which address debris (insulation). Regulatory Guide 1.82 recommends, in addition to providing redundant separated sumps, that two protective screens

be provided. A low approach velocity in the vicinity of the sump is required

to allow insulation to settle out before reaching the sump screening, and it is

required that the sump remain functional assuming that one-half of the screen

surface-area is blocked. The staff review of the sump design against Regulatory Guide 1.82 is found in SER Section 6.3.

A second postulated means of losing the ability to draw water from the emergency sump could be abnormal conditions in the sump or at the pump inlet such

as air entrainment, vortices, or excessive pressure drops. These conditions

could result in pump cavitation, reduced flow, and possible damage to the

pumps.

Currently, regulatory positions regarding sump testing are contained in Regula- tory Guide 1.79, "Preoperational Testing of Emergency Core Cooling Systems for

Pressurized Water Reactors," which addresses the testing of the recirculation

function. As indicated in Section 6.3 of this SER, the applicant will perform

Watts Bar Ser C-15
inplant tests to demonstrate that circulation through the sump can be reliably

accomplished. In addition, TVA has performed out-of-plant scale model tests of

the Watts Bar containment sump design (according to J. E. Gilleland's letter to

S. A. Varga, dated May 2,3, 1979). The tests-were performed utilizing several

design modifications identified during similar sump model tests for the Sequoyah

nuclear plant. These modifications have been incorporated into the Watts Bar

design. Therefore, based on the above, the staff has concluded that there, is

reasonable assurance that the Watts Bar facility can be operated before the

ultimate resolution of this generic issue without endangering the health and

safety of the public.

A-44 Station Blackout

Electrical power for safety systems at nuclear power plants must be supplied by

at least two redundant and independent divisions. The systems used to remove

decay heat to cool the reactor core following a reactor shutdown are included

among the safety systems that must meet these requirements. Each electrical

division for safety systems includes an offsite power connection, a standby

emergency diesel generator ac power supply, and dc sources.

Task A-44 involves a study of whether or not nuclear power plants should be

designed to accommodate a complete loss of all ac power (that is, loss of both

the offsite and the emergency diesel generator ac power supplies). This issue

arose because of operating experience regarding the reliability of ac power

supplies. A number of operating plants have experienced a total loss of offsite electrical power, and more occurrences are expected.

During each of these loss of offsite power events, the onsite emergency ac

power supplies were available to supply the power needed by vital safety

equipment. However, in some instances, one of the redundant emergency power

supplies has been unavailable. In addition, there have been numerous reports

of emergency diesel generators failing to start and run in operating plants

during periodic surveillance tests.

A loss of all ac power was not a design-basis event for Watts Bar. Nonetheless, a combination of design, operation, and testing requirements that

have been imposed on the applicant will ensure that these units will have

substantial resistance to a loss of all alternating current and that, even if a

loss of all alternating current should occur, there is reasonable assurance

that the core will be cooled. These are discussed below.

A loss of offsite ac power involves a loss of both the preferred and backup

sources of offsite power. The staff review and basis for acceptance of the

design, inspection, and testing provisions for the offsite power system aredescribed in SER Section 8.2. Also, a preliminary grid stability analysis was

performed. This also is discussed in Section 8.2.

If offsite power is lost, two independent and redundant diesel generators and

their associated distribution systems will deliver emergency power to safetyrelated equipment. The staff review of the design, testing, surveillance, and

maintenance provisions for the onsite emergency diesels is described in Section 8.3.1 of the SER. Staff requirements include preoperational testing to

ensure the reliability of the installed diesel generators in accordance with

Watts Bar Ser C-16
the requirements discussed in the SER. In addition, the applicant has been

asked to implement a program for enhancement of diesel generator reliability to better ensure the long-term reliability of the diesel generators. This program resulted from recommendations of NUREG/CR-0660, "Enhancement of Onsite Emergency

Generator Reliability."

Even if both offsite and onsite ac power is lost, cooling water from the

condensate storage tank can still be provided to the steam generators by the

auxiliary feedwater system by employing a steam-turbine-driven pump that does

not rely on ac power for operation.

The staff review of the auxiliary feedwater system design and operation is

described in SER Section 10.4.9.

The issue of station blackout was also considered by the Atomic Safety and

Licensing Appeal Board (ALAB-603) for the St. Lucie Unit 2 facility. In

addition, in view of the completion schedule for Task A-44 (October 1982), the

Appeal Board recommended that the Commission take expeditious action to ensure

that other plants and their operators are equipped to accommodate a station

blackout event. The Commission has reviewedthis recommendation and determined

that some interim measures should be taken at all facilities, including Watts

Bar, while Task A-44 is being conducted. Consequently, interim emergency procedures and operator training for safe operation of the facility and restoration of ac power will be required. The staff notifed the applicant of

these requirements in a letter from D. Eisenhut, NRC, dated February 25, 1981.

The staff will condition the OL for Watts Bar to indicate that these procedures and training must be completed by fuel load.

In addition, as part of the Sequoyah Unit 1 low-power test program, TVA per- formed a "blackout" test, which simulated loss of all offsite and onsite ac

power. The test results showed that during the blackout period, the turbinedriven auxiliary feedwater pumps delivered flow" to the steam generators as

required. Therefore, although the designs for the ac power and auxiliary feedwater systems at Sequoyah and Watts Bar are not identical, sufficientsimila'rities exist in the designs so that the test results (from a system operation standpoint) could be considered applicable to Watts Bar. Additional

information on this test and other concerns (operator training, procedures, and

so forth) related to this issue is provided in Supplement 2 to the Sequoyah SER

and in TVA's response (L.M. Mills' June 5, 1981 letter to E. Adensam) to NRC

generic letter 81-04, "Emergency Procedures for Training for Station Blackout

Events."

Based on the above, the staff has concluded that there is reasonable assurance

that Watts Bar can be operated before the ultimate resolution of this generic issue without endangering the health and safety of the public.

A-45 Shutdown Decay Heat Removal Requirements

Under normal operating conditions, power generated within a reactor is removed

as steam to produce electricity via a turbine generator. Following a reactor

shutdown, a reactor produces insufficient power to operate the turbine; however, the radioactive decay of fission products continues to produce heat

(so-called "decay heat"). Therefore, when reactor shutdown occurs, other

Watts Bar Ser C-17
measures must be available to remove decay heat from the reactor to ensure that

high temperatures and pressures do not develop which could jeopardize the

reactor and the reactor coolant system. It is evident, therefore, that all

light water reactors (LWRs) share two common decay heat removal functional

requirements: (1) to provide a means of transferring decay heat from the

reactor coolant system to an ultimate heat sink and (2) to maintain sufficient

water inventory inside the reactor vessel to ensure adequate cooling of the

reactor fuel. The reliability of a particular power plant to perform these

functions depends on the frequency of initiating events that require or

jeopardize decay heat removal operations and the probability that required

systems will respond to remove the decay heat.

This Unresolved Safety Issue will evaluate the benefit of providing alternate

means of decay heat removal which could substantially increase the plant's

capability to handle a broader spectrum of transients and accidents. The study

will consist of a generic system evaluation and will result in recommendations

regarding the desirability of and possible design requirements for improvements

in existing systems or an alternative decay heat removal method, if the

improvements or alternative can significantly reduce the overall risk to the

public.

The primary method for removal of decay heat from pressurized water reactors is

via the steam generators to the secondary system. This energy is transferred

on the secondary side to either the main feedwater or auxiliary feedwater

systems, and it is rejected to either the turbine condenser orthe atmosphere

via the steamline safety/relief valves. Following the TMI-2 accident, the

importance of the auxiliary feedwater system was highlighted, and a number of

steps were taken to improve the reliability of the auxiliary feedwater system.

The staff's review of these items is in SER Section 10.4.9. It was also

stipulated that plants must be capable of providing the required auxiliary

feedwater flow for at least 2 hours from one auxiliary feedwater pump train,

independent of any ac power source (that is, if both offsite and onsite ac

power sources are lost).

Pressurized water reactors also have alternate means of removing decay heat if

an extended loss of feedwater is postulated. This method is known as "feed and

bleed" and uses the high pressure injection system to add water coolant (feed)

at high pressure to the primary system. The decay heat increases the system

pressure and energy is removed through the power-operated relief valves and/or

the safety valves (bleed), if necessary.

At low primary system pressure (below about 200 psi), the long-term decay heat

is removed by the residual heat removal system to achieve cold shutdown

conditions.

Based on the above, the staff has concluded that Watts Bar can be operated

before ultimate resolution of this generic issue without endangering the health

and safety of the public.

A-46 Seismic Qualification of Equipment in Operating Plants

The design criteria and methods for the seismic qualification of mechanical and

electrical equipment in nuclear power plants have undergone significant change

Watts Bar Ser C-18
during the course of the commercial nuclear power program. Consequently, the

margins of safety provided in existing equipment to resist seismically induced

loads and perform the intended safety functions may vary considerably. The

seismic qualification of the equipment in operating plants must, therefore, be

reassessed to ensure the ability to bring the plant to a safe shutdown condition when it is subject to a seismic event. The objective of this USI is to

establish an explicit set of guidelines that could be used to judge the adequacy of the seismic qualification of mechanical and electrical equipment at

all operating plants in lieu of attempting to backfit current design criteria

for new plants. This guidance will concern equipment required to safely shut

down the plant, as well as equipment whose function is not required for safe

shutdown, but whose failure could result in adverse conditions which might

impair shutdown functions.

The seismic qualification review team has performed the seismic qualification

audit. The information obtained is under evaluation and the staff will report

on the results of its audit in a supplement to this SER.

A-47 Safety Implications of Control Systems

This issue concerns the potential for transients or accidents being made more

severe as a result of control system failures or malfunctions. These failures

or malfunctions may occur independently or as a result of the accident or transient under consideration. One concern is-the potential that a single failure

such as a loss of a power supply, short circuit, open circuit, or sensor failure could cause simultaneous malfunction of several control features. Such an

occurrence could conceivably result in a transient more severe than those transients analyzed as anticipated operational occurrences. A second concern is

that a postulated accident could cause control system failures which could

make the accident more severe than analyzed. Accidents could conceivably cause

control system failures by creating a harsh environment in the area of the control equipment or by physically damaging the control equipment. It is generally believed by the staff that such control system failures would not lead to

serious events or result in conditions that safety systems cannot safely handle.

Systematic evaluations have not been rigorously performed to verify this belief.

The potential for an accident that could affect a particular control system,

and effects of the control system failures, may differ from plant to plant.

Therefore, it is not possible to develop generic answers to these concerns, but

rather plant-specific evaluations are-required. The purpose of this USI is to

define generic criteria that will be used for plant-specific evaluations.

The Watts Bar control and safety systems have been designed with the goal of

ensuring that control system failures will not prevent automatic or manualinitiation and operation of any safety system equipment required to trip the

plant or to maintain the plant in a safe shutdown condition following any

"anticipated operational occurrence" or "accident." This has been accomplished

either by providing independence between safety and nonsafety systems or by

providing isolating devices between safety and nonsafety systems. These

devices preclude the propagation of nonsafety system equipment faults to the

protection system. Also, to ensure that the operation of safety system equipment is not impaired, the single-failure criterion has been applied in the

Watts Bar Ser C-19
plant design. Design reviews are being performed (IE Bulletin 79-27) to ensure

that the loading on certain Class 1E power boards maintains the separation and

independence of plant systems as designed.

A systematic evaluation of the control system design, as contemplated for this

USI, has not been performed to determine-whether postulated accidents could

cause significant control system failures which would make the accident consequences more severe than presently analyzed. However, a wide range of bounding

transients and accidents is analyzed to ensure that the postulated events such

as steam generator overfill and overcooling events would be adequately mitigated by the safety systems. In addition, systematic reviews of safety systems

have been performed with the goal of ensuring that control system failures

(single or multiple) will not defeat safety system action. These reviews are

part of an ongoing evaluation program to qualify Class 1E plant equipment to

function for all postulated service conditions to which it is subjected.

Based on the above, the staff has concluded that there is reasonable assurance

that Watts Bar can be operated before ultimate resolution of this generic issue

without endangering the health and safety of the public.

A-48 Hydrogen Control Measures and Effects of Hydrogen Burns on

Safety Equipment

Following a loss-of-coolant accident in a light water reactor plant, combustible gases, principally hydrogen, may accumulate inside the primary reactor

containment as a result of: (1) metal-water reaction involving the fuel

element cladding; (2) the radiolytic decomposition of the water in the reactor

core and the containment sump; (3) the corrosion of certain construction

materials by the spray solution; and (4) any synergistic chemical, thermal, and

radiolytic effects of postaccident environmental conditions on containment

protective coating systems and electric cable insulation.

Because of the potential for significant hydrogen generation as the result of

an accident, 10 CFR 50.44, "Standards for Combustible Gas Control System in

Light Water Cooled Power Reactors," and GDC 41, "Containment Atmosphere

Cleanup," require that systems be provided to control hydrogen concentrations

in the containment atmosphere:following a postulated accident to ensure that

containment integrity is maintained.

10 CFR 50.44 requires that the combustible gas control system provided be

capable of handling the hydrogen generated as a result of degradation of the

emergency core cooling system so that the hydrogen release is five times the

amount calculated in demonstrating compliance with 10 CFR 50.46 or the amount

corresponding to reaction of the cladding to a depth of 0.00023 in. whichever

amount is greater.

The accident at TMI-2 resulted in hydrogen generation well in excess of the

amounts specified in 10 CFR 50.44. As a result, it became apparent to NRC that

specific design measures are needed for handling larger hydrogen releases,

particularly for smaller low pressure containments. As a result, the Commission determined that a rulemaking proceeding should be undertaken to define

the manner and extent to which hydrogen evolution and other effects of a

degraded core need to be taken into account in plant design. An advance notice

Watts Bar Ser C-20
of this rulemaking proceeding on degraded core issues was published in the

Federal Register on October 2, 1980.

Recognizing that a number of years may be required to complete this rulemaking proceeding, a set of short-term or interim actions relative to hydrogen control

requirements were developed and implemented. These interim measures were issued

as a proposed rule.on December 23, 1981, in the Federal Register (46 FR 62281). For plants with ice condenser containments, such as Watts Bar, the proposed rule would require installation of improved hydrogen control systems and

demonstration of the capability of certain important safety systems to function

during and following hydrogen burning.

The applicant has stated that the research and development work being performed by TVA for the Sequoyah nuclear plant is directly applicable to Watts.Bar.

However, the staff has noted two differences in the designs of Watts Bar and

Sequoyah which may be significant to the degraded core hydrogen analysis: (1) Watts Bar has a containment spray flow of 4000 gpm per spray train as

opposed to a value of 4750 gpm for Sequoyah; and (2) Watts Bar has a containment design pressure of 15.0 psig, compared to a design pressure of 12.0 psig for Sequoyah. The applicant has agreed to provide additional justification for

his contention that the work performed on this issue for Sequoyah is directly applicable to Watts Bar. The staff will consider this additional justification

in its continuing review of this issue.

TVA is conducting research into hydrogen combustion phenomena and mitigation

measures in cooperation with Duke Power, American.Electric Power,, and the

Electric Power Research Institute, as well as on its own.

The staff will require as a licensing condition that the Watts Bar plant shall

be provided with an acceptable hydrogen control system justified by suitable

programs of experiment and analysis. The hydrogen control system must be

capable of handling an amount of hydrogen equivalent to that generated from the

reaction of 75 percent of the fuel cladding surrounding the active fuel region (excluding the cladding surrounding the plenum volume) with water, without loss

of containment structural integrity. The Watts Bar plant also shall be pro- vided with systems necessary to establishand maintain safe cold shutdown

and maintain containment integrity that are capable of performing their functions during and after being exposed to the environmental conditions created by the burning (or local detonation) of hydrogen.

The license condition will state that these requirements will have to be

satisfied before full-power operation.

In the interim, the applicant has committed to install a system referred to as

the permanent hydrogen mitigation system (PHMS) in.the Watts bar containments

before fuel loading. The PHMS consists.of igniters distributed throughout the

containment. This system is designed to improve control of hydrogen ignition to prevent containment overpressure failure. This system is the same as the

PHMS that has been installed at Sequoyah and that the staff has approved for

use at Sequoyah during the interim until the final safety review has been

completed by the staff.

Watts Bar Ser C-21
Based on'the above, the staff concludes that the Watts Bar facility can be

operated before complete resolution of this issue and completion of the

proposed rulemaking without undue risk to the health and safety of the public.

A-49 Pressurized Thermal Shock

Severe reactor-system overcooling events in a pressurized water reactor (PWR)

that could be followed by repressurization of the reactor vessel can result

from a variety of causes. These include instrumentation and control system

malfunctions and postulated accidents such as small break loss-of-coolant

accidents (LOCAs), main steamline breaks, or feedwater pipe breaks. Rapid

cooling of the reactor vessel internal surface causes a temperature gradient

across the reactor vessel wall. This temperature gradient results in thermal

stress, with a maximum tensile stress at the inside surface of the vessel. The

magnitude of the thermal stress depends on the temperature differences across

the reactor vessel wall. Effects of this thermal stress are compounded by the

hoop stress if the vessel is repressurized.

As long as the fracture resistance of the reactor vessel material remains high,

such transients will not cause failure. After the fracture toughness of the

vessel is reduced by neutron irradiation, severe thermal transients could cause

fairly small flaws near the inner surface to initiate and result in significant

cracking. The vessels of most concern are those with high radiation exposure,

which are made of material that has a relatively high sensitivity to radiation

damage (such as those made with welds of high copper content).

For failure of the RPV to occur, a number of contributing factors must be

present. These factors are: (1) a reactor vessel flaw of sufficient size to

initiate and propagate; (2) a level of irradition (fluence) and material

properties and composition sufficient to cause significant embrittlement (the

exact fluence is dependent upon materials present, for example, high copper

content causes embrittlement to occur more rapidly); (3) a severe overcooling

transient with repressurization; and (4) the crack resulting from the

propagation of initial cracks must be of such size and location that the vessel

fails.

The staff's preliminary review of overcooling events and their probabilities

included a study on overcooling events at Babcock and Wilcox (B&W) plants; a

survey of operating experience on Westinghouse (W) and Combustion Engineering

(CE) plants; a review of available accident analysis in Final Safety Analysis

Reports and in vendor topical reports; and a preliminary probabilistic analysis.

The preliminary results of these evaluations indicate that there is a probability of about 10-3 per reactor year that a B&W-designed plant will expe--

rience a severe overcooling transient similar to or worse than that experienced at Rancho Seco on March 20, 1978. The Rancho Seco transient is the

most severe overcooling transient experienced by any PWR in the United States.

The staff estimates that the probability of such an overcooling event in CE or

W-designed reactors is lower, perhaps by an order of magnitude, than for

B&W-designed reactors. This difference is based on design differences and on

operating experience.

In the 1978 Rancho Seco transient, reactor pressure was maintained at a fairly

high level (1500 psig to 2100 psig) throughout the cooldown. The minimum

temperature of the reactor coolant (280'F) during the transient was high enough

Watts Bar Ser C-22
so that material toughness of the reactor vessel was not affected. This

evaluation leads the staff to believe that if this transient were to be

repeated at Rancho Seco or any other B&W-designed facility within the next few

years, the reactor vessel failure would still be unlikely. Nonetheless,_the

possibility of vessel failure as a result of an overcooling event cannot be

completely ruled out. If an overcooling event such as that at Rancho Seco were

to occur, even for the vessel with the most limiting material properties in

existence today, the staff would not expect a failure.

The staff conclusion is supported by ORNL analyses of the Rancho Seco event,

which indicate that the threshold irradation level (neutron fluence) for crack

initiation (that is, small cracks growing to larger ones assuming conservative

initial material properties such as RTNDT =400 F and copper content of

0.35 percent) would be in the range of 1019 neutrons/cm2 . The highest neutron

fluence to date in a B&W-,designed facility is less than half the minimum value

listed above. It would, therefore, be several years before any B&W-designed facility reached its threshold irradiation level.

Some reactor vessels in CE and W facilities have somewhat higher fluences;

however, other mitigating factors--such as lower values of initial RTNDT--

provide a significant margin to failure should an overcooling event similar

to that at Rancho Seco occur.

As a result of its evaluations to date, the staff has concluded that the

probability of a severe overcooling transient (similar in magnitude to the

Rancho Seco event) is relatively low. For B&W-designed reactors this proba- bility is estimated to be about 10-3 per reactor per year, and for W- and

CE-designed reactors, it is lower, perhaps by an order of magnitude- Furthermore, the staff anticipates that this issue will be resolved before the

irradiation history at the Watts Bar facility is large enough to cause a

significant pressurized thermal shock concern.

Therefore, based on the foregoing, the staff concludes that the Watts Bar

facility can be operated before resolution of this issue, without undue risk to

the public.

Watts Bar Ser C-23

APPENDIX D

EVALUATION OF THE APPLICANT'S

CONTROL ROOM DESIGN

INTRODUCTION

The control room design review conducted by the Human Factors Engineering Branch (HFEB) included an evaluation of the control room layout, the adequacy of the information provided, the arrangement and identification of important controls and instrumentation displays, the usefulness of the audio and visual

alarm systems, the information recording and recall capability, lighting, and

other considerations of human factors that have an effect on operator performance. The review was performed by an inspection of the control panels, interviews with operators, and observation and videotaping of operators as they walked through selected emergency procedures.

DISCUSSION

Although the staff review identified human engineering discrepancies, the

staff found that overall the control room was designed to promote effective

and efficient operator actions. Annunciator panels are used to indicate

required operator actions. Most controls are well laid out, within easy reach

of most operators, and follow accepted design conventions for position and

direction of movement. Visual displays are adequate with respect to design, location, and illumination. Most controls and displays are well organized and

show consideration for functional and sequential arrangement. Control room

layout and physical design of the control panels and consoles reflect consideration of human anthropometry. Color has been used effectively, and the

mimics that were used on the panels and consoles enhance operator effectiveness

in interfacing with controls and displays.

HUMAN FACTORS DEFICIENCIES IDENTIFIED

The review team identified human engineering discrepancies that were documented

in the control room design review report transmitted to the applicant. The

report ranked the discrepancies according to their importance. The deficiencies

were given a priority rating of 1, 2, or 3 (high, moderate, low), based on the

increased potential for operator error and the possibility consequences of.that

error.

Deficiencies identifed as priority rating 1 or 2, and those involving simple corrective actions, must be corrected before an operating license (OL) is issued. These deficiencies are listed in the following section of this report, along with descriptions of the applicant's commitments to correct them.

Deficiencies that were given priority 3 ratings will be addressed by the

applicant in the performance of long-term studies associated with the detailed

control room design review to determine the best solutions. Those priority 3

Watts Bar SER D-1
items were identified in the control room design review report, dated May 8,

1981, and are not included herein. However, based on further evaluation, the

priority rating of some deficiencies was changed from a higher priority rating

to priority rating 3. These changes are discussed in this report.

Also included is a description of those discrepancies identified in the staff

draft control room design review report that were subsequently determined to

be invalid. Descriptions of those discrepancies and the rationale for their

deletion are presented.

HUMAN FACTORS DEFICIENCIES TO BE CORRECTED BEFORE ISSUANCE OF AN OPERATING

LICENSE

This section contains a description of those deficiencies that must be corrected

before an operating license is issued. Following each deficiency (or following

several related deficiencies) is a statement of the TVA corrective action

commitments.

1.0 CONTROL ROOM WORKSPACE

1.1 Several problems exist with regard to emergency equipment protective

clothing:

Donning is difficult.

Changing air tanks is a two-person operation.

With the 5-min warning bell, there is perhaps too little time to change.

The mask virtually prohibits the wearer from speaking to anyone.

Too many steps are necessary to don gear. (6.1.A)*

Operators will undergo periodic drills and training to maintain proficiency

in donning emergency gear. Protective clothing sizes will be clearly

identified.

An effort is presently under way to determine the best system for improving

communication when wearing the emergency breathing apparatus. The possible

modification to the emergency breathing apparatus face mask involves

equipping the mask with a standard telephone transmitter-receiver unit

mounted so that integrity of the face seal is not jeopardized. Each

device would have a transmitter transfer unit which would permit the

operator to switch between the telephone and paging systems.

The results of this evaluation will be submitted for staff review at least

120 days before an OL is issued. As part of the long-term review, TVA

will evaluate and select a "hands-free" mask.

*Throughout the report the use of parentheses, such as (6.1.A), refers to the

section and finding number used in the HFEB draft control room design review

dated May 8, 1981.

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1.2 There are no labels on the six area sound-powered phone jacks to identify

the area to which they are hard wired. (6.1.B.1)

Labels will be installed on sound-powered jacks SP1-SP6. Operators will

be instructed as to the, designated areas to which the jacks are wired.

1.3 No headsets are located on any of the back cabinets. (6.1.B.2)

Headsets will be installed on the back cabinets to improve control room

communications.

1.4 Distance and noise interfere with voice communication between the Unit 1

control room operator at the panels and the operator at the common panels.

The problem is particularly acute when a breathing apparatus is worn by

both. (6.1.B.3)

To help reduce the background noise level, carpet will be installed before

fuel load. Noisy equipment that interferes with voice communications will

be identified and modified to reduce its noise output (for example, the

radiation monitoring cabinet has been moved to a location outside the

horseshoe).

Additional noise measurements will be made after the above improvements

have been made, and further corrective action will be taken if necessary.

Results will be submitted to the NRC 120 days before an OL is issued.

2.0 WORKSPACE ENVIRONMENT

2.1 The control room lighting design appears to be for general illumination

purposes and not specifically for effective man/machine interface.

(6.2.A.1)

2.2 The plastic prismatic luminous ceiling diffusers minimize indirect glare

throughout the control room, but they contribute to direct glare on panel

meters, records, printers, and polished surfaces. (6.2.A.2)

2.3 The ac emergency lighting illumination level is too low for optimum

reading of displays. (6.2.A.5)

2.4 The ratio of luminosity (background luminous ceiling and the green painted

metal reading surface of panels/consoles), 140-ft lamberts/18-ft lamberts,

is 8/1. This ratio is 3 times the recommended standard. (6.2.A.7)

TVA control room lighting is divided into three categories: normal,

emergency, and standby. The normal lighting system is powered by offsite

ac power. The present normal control room lighting consists of a luminous

ceiling that provides general illumination to the control panels, walkways,

and so forth.

The emergency incandescent lighting system is battery powered and functions

when the main lighting power has failed. It activates before operator

action is required and operates until the diesel generators'are activated

Watts Bar SER D-3
and the standby lighting system goes in service. (See response to

Item 2.7 below for further information.)

Based on the results of a preliminary control room review, lighting glare is not a problem. This concern has been adequately resolved by the

painting (where practical) of reflective surfaces and by adjusting the

overhead lighting by the operators.

A luminosity ratio of below 8/1 is recommended in a TVA control room.

This ratio is acceptable for luminaire versus the surfaces adjacent to

them, according to the NRC guidance provided in NUREG-0700, "Guidelines

for Control Room Design Reviews," September 1981. However, it is not an

acceptable ratio for.a task area versus adjacent lighter surroundings. This issue will be further evaluated during the NUREG-0700 review.

Present plans call for the replacement of the luminous ceiling as the

result of an NRC concern over flame spread and smoke density characteristics (according to a letter from L. M. Millsto A. Schwencer dated

April 30, 1981 (NEB 810505 564)). Photographs and photometric measurements will be taken at Watts Bar following installation of the new ceiling and will be submitted to the NRC 120 days before an operating OL is issued

(assuming ceiling replacement has been completed by this time).

2.7 DC emergency lighting incandescent lamps are not located for optimum

reading of displays. (6.2.A.6)

The dc emergency lighting system has been tested and found to provide 1 footcandle of light from one train of emergency lighting and 3 footcandles of light from both trains of emergency lighting in designated

areas of the control.room. The system is designed to provide the above

light levels for a period of 2 hours as the bital battery voltage decays

to 105 volts. The light level readings were taken parallel to the face of

the board or working surface. Recently several light fixtures were added

to the .system, and the wattage of some of the incandescent bulbs was

increased and to ensure that the 1- and 3-footcandle requirement could be

met as specified by the preoperational test scoping document (preopera- tional test TVA 35). Therefore, the present system is adequately designed. However, present plans• call for the replacement of this ceiling as a

r esult of NRC concern.over flame spread and smoke density characteristics.

The new lighting levels will be measured and the results given to the NRC

for review 120 days before an operating license is issued.

3.0 ANNUNCIATORS

3.1 The auditory alarms that were functional were barely detectable above the

ambient background noise. (6.3.C)

Because the alarms have been partially disabled during construction, it

is possible that they are at times difficult to hear during construction

activities. However, all annunciator alarms will be operational before

fuel load. Also efforts are underway to reduce the background noise

level (such as. installing carpet, and relocating the radiation monitoring

equipment) before fuel load. The alarm and noise levels will be measured

Watts Bar SER D-4
and the results given to the NRC for review 120 days before an operating

license is issued.

3.2 The poor organization of many annunciator tiles makes them difficult to

locate in the standard operating instruction (SOI). No cross references

are provided. (6.3.D)

A window numbering system will be used to help the operator refer to

specific annunciator information in the SO.

3.3 Alarms are not prioritized. (6.3.E)

The alarms will be prioritized by implementing a priority alarm system

that is being developed. Each priority will have a unique color for

identification purposes. Annunciator windows will be color coded according

to their priority, with red used for emergency alarms and white for

nonemergency alarms. The applicant is investigating using an additional

color coding technique for the annunciator windows.

3.4 The contrast between illuminated and nonilluminated tiles is too low.

(6.3.G)

Contrast will be improved by using a reflector surface with high

reflectivity.

3.5 The poor relationship between the acknowledge-reset controls and their

associated annunciator displays could result in the incorrect acknowledgment of an alarm. (6.3.H)

This has been corrected. The acknowledge-reset controls, XS-55-4A and

XS-55-4D, have been relocated to positions 100 and 134 of panel 1-M-4,

respectively (per Engineering Change Notice (ECN) 2710).

3.6 Location of acknowledge-reset controls is not standardized on each panel.

(6.3.1)

Switches. have been shape-coded and highlighted with yellow striping to

aid the operator in locating the control. Also, an effort has been made

to locate the switches midway up toward the ends of the panels. In some

instances, however, this would upset the layout of other switches and

controls. In these cases, the annunciator switches were located as close

to these positions as possible.

3.7 Replacement of annunciator bulbs frequently subject the operator to a

shock. (6.3.J)

A shield will be placed over the leading and bottom edges of the power

busses.

4.0 CONTROLS

4.1 The location of some controls disregards the capabilities of the fifth

percentile person. (6.4.A)

Watts Bar SER D-5
The present system can accommodate a fifth percentile female even though

some operator movement may be required to accomplish tasks involving

controls located on the vertical portion of the panels. The addition of

a guard rail along the edge of the horizontal portion of the control

board will preclude accidental actuation of controls.

4.2 Labeling on many valve controllers does not consistently or clearly

associate direction of movement with resulting action. A clockwise

movement does not always result in opening the valve. (6.4.B)

Direction arrows will be applied before fuel load, and shape coding of the

controller knobs will be implemented if it is practical to do so.

4.3 Controls located near the front edges of sloping consoles create a

potential for inadvertent actuation, if they are not guarded. (6.4.C)

Panel guard rails will be installed.

4.4 Direction of control movement on "Trip" controllers is not standardized.

(6.4.D)

The switch on panel 1-M-4 follows the convention "Trip" to right, "Reset"

to left. The controller on panel 1-M-6 "Trips" to the right; movement to

the left is not used and is blocked from movement. However, the labeling

on the switch on panel 1-M-6 is confusing. It functions only as a trip

control but is labeled both trip and reset. This will be corrected and

will eliminate the confusion over apparent nonstandardization to trip

control movement.

4.5 The reactor trip-reset controller on Panel 1-M-6 is incorrectly labeled.

It functions only as a trip control. (6.4.E)

This will be corrected by relabeling the switch to indicate its function

as trip only.

4.6 Specific control handles were not easily distinguishable when they were

in a bank of controls (such as a string of five with one different).

(6.4.F)

This will be corrected by the addition of functional name tags and

demarcation lines (per ECN 2604).

4.7 Valve control handles which must be held for many seconds are difficult

to hold. (6.4.H)

Most of the valve control circuits have seal-in features which automatically drive the valves to the desired position after momentary actuation

of the control switch. Some valve control circuits intentionally do not

have such a feature, but in such cases complete valve travel is generally

accomplished within 5 seconds. This feature will be noted by additional

coding.

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4.8 The function (that is, use of the pull out/in feature) of some J-handle

switches is not clearly labeled. (6.4.1)

4.9 The meaning of "Pull to --- " labels associated with some "J-handles" is

not always clear with regard to their rotary position. (6.8.A.9)

Presently, labels exist. However, switches that have been identified by

the operators as confusing will be relabeled.

4.10 There are no protective guards over the emergency "Start" and "Stop"

pushbuttons on the diesel generator panel. (6.4.J)

Protective guards are not required because the emergency "Start" and

"Stop" switches have raised bezels to prevent inadvertent actuation.

4.11 The vertical cabinets along the walkway between Units 1 and 2 contain

numerous protruding controls which could be inadvertently actuated.

(6.4. K)

A red carpet will be installed at the base of the vertical panels to

designate an off-limit area to personnel not performing a required task.

5.0 DISPLAYS /

5.1 Scale divisions and ranges are not always immediately obvious. (6.5.A.1)

The scales of safety-related indicators will be color coded to aid the

operator in identifying normal and abnormal operating ranges. Each

safety-related parameter will be evaluated to ensure that the coding

technique does not confuse the operator as a result of possible operating

range changes that may be dependent upon plant mode. The selected safetyrelated indicators will be temporarily marked on the lens cover to indicate normal and abnormal operating range before fuel loading. When the

indicator is calibrated, a permanent color code will be applied to the

scale so as to eliminate the parallax problem. TVA agrees that calibration of these indicators after fuel loading is acceptable.

5.2 Some meters required for postaccident monitoring are intermixed with

other meters. (6.5.A.3)

5.3 All of the postaccident monitor (PAM) displays are labeled with temporary

red tape. (6.5.A.4)

PAM instrumentation in the main control room (MCR) will be uniquely

identified to facilitate recognition by the operators. The PAM identification method will use black tags with white lettering. Each tag will

contain in the upper right-hand corner the PAM symbol, which is a square

with an inscribed P1 or P2. A technology will be used so that the tag

surface is even or almost even, thus eliminating the possibility of dirt

accumulation within the lettering that would reduce contrast and legibility.

5.4 Displays indicating sequential information are not always located sequentially or grouped together for ease in visual scanning. (6.5.A.6)

Watts Bar SER D-7
This will be corrected by the addition of functional name tags and demarcation lines (per ECN 2604).

5.5 Meters indicating parameters that must be compared are not grouped for

ease in distinguishing similar or different values. (6.5.A.6)

This will be corrected by the addition of functional name tags and demarcation lines (per ECN 2604).

5.6 Meters that indicate different parameters look alike, have similar scales,

and are similarly labeled. Reading errors-could result. (6.5.A.7)

This will be corrected by the addition of functional name tags and demarcation lines (per ECN 2604).

5.7 Identification of a specific meter is not always obvious among a string of

meters. (6.5.A.8)

This will be corrected by the addition of functional name tags and demarcation.lines (per ECN 2604).

5.8 Meters labeled "B" left and "A" right violate stereotypical convention.

(6.5.A.9)

This will be corrected (per ECN 2710) by moving meters so that "A" is on

the left and "B" is on the right.

5.9 The subcooling margin monitor is not installed or on site. Readouts and

displays could not be observed.

A Class 1E vertical pressure indicator will be scaled to read-both saturation temperature and reactor coolant system (RCS) pressure. The operator

will be able to read the saturation temperature for the RCS pressure

directly from the indicator. By comparing this reading with the RCS

temperature, the operator will know the subcooling margin at any time

for all operating conditions without referring to the steam tables.

5.10 Control position indicator bulbs have single-failure filaments and no

lamp-test capabilities. (6.4.B.1)

5.11 A lamp-test capability is provided only for the overhead annunciator

panels and status panels. (6.7.L)

The control boards are under constant surveillance by the operators and

burned out bulbs are replaced upon detection. The shift change procedures

will be revised to instruct the operators to check for burned-out lamps at

shift change.

5.12 The valve position indicator on the EH Fluid operator's subpanel (1-M-2)

is not clear. A green lens indicates "Valve Open." (6.5.B.3)

This has been corrected by replacing the green lens with a red lens.

Watts Bar SER D-8
5.13 In the matrix of indicator lights, failure to achieve a proper system

status is often given by a light "off." No check is made during EP for

failed lights. This could lead the operator to assume that an operating

system had failed. (6.5.B.6)

This has been corrected by revising control room procedures to instruct

operators to use the lamp test feature on the status monitoring panels to

verify that a lamp is not burned out.

5.14 In the matrix of lights, spares are intermixed with operating lights.

Because the operator's task is to determine "all on," "spares off" creates

a requirement for the operator to review all lights to ensure that only

the spares are off. (6.5.B.7)

Spare windows will be blacked out.

5.15 Recorder scales and paper scales are not always compatible. (6.5.C.4)

This will be corrected by instructing the applicant to order the correct

paper as required.

5.16 Labels indicating the color code of pens are not always compatible with

the color of the pen trace indicating that parameter. (6.5.C.5)

This will be corrected by the addition or correction of labels.

5.17 Some recorders do not identify the parameter units indicated on both the

scales and paper. (6.5.C.6)

The computer-addressable recorders utilize many ranges making the identification of a particular range on the scale impractical. However, the

operator labels the recorder with stick-on labels to indicate the range

being used. The scales for the remaining recorders will be reviewed and

revised.

5.18 In some instances the test and calibration labels cover meter faces or

paper-scale indication. (6.5.C.7)

Tests and calibration labels will be removed from meter faces and paper

scales.

5.19 Some dual function recorders do not have nameplates indicating the parameters being monitored. (6.5.C.8)

This concern will be corrected with the addition of nameplates (per

ECN 2604).

5.20 Regarding steam generator strip charts, labels on chart windows contradict

those under chart units. It appears that recorders were reinstalled

incorrectly after maintenance. (6.5.C.9)

Chart recorders are presently in their correct locations.

Watts Bar SER D-9
6.0 PANEL LAYOUT

6.1 Controls are not arranged in logical order, (that is, by function or

sequentially). (6.6.A)

Controls are grouped by system. Within this system, controls are grouped

by function. Approximately 62 devices are being moved to conform to this

arrangement (per ECN 2710). Also demarcation and functional name tags

will be added (per ECN 2604).

6.2 Demarcation between Units 1 and 2 and common controls and displays is not

clearly indicated. (6.6.B)

Common panels containing controls and displays from both units will be

modified by adding color coding and specific unit numbering to provide

unique identification of each control and display.

6.3 Large string/matrices of switches are located at several places on panels,

specifically for component cooling water, water service systems, essential

raw cooling water, and ventilation. (6.6.C)

This will be corrected by the addition of functional name tags and demarcation lines (per ECN 2604).

6.4 The feedwater and condensate system had a sequence of valves (left to

right) CBABA. (6.6.D)

Devices violating the order of convention will be corrected (per ECN 2710).

6.5 There are several long strings (greater than 4) of vertical meters.

Mounting in strings increases the likelihood that the operator will read

the wrong display if it is near the middle of the string. (6.6.E)

This will be corrected by the addition of demarcation lines and functional

name tags (per ECN 2604).

7.0 CONTROL/DISPLAY (C/D) ARRANGEMENT

7.1 Some displays are not located directly above the controls that relate to

them. Some are even located on different panels. Effective control/

display relationships do not exist. (6.7.A)

7.2 There are many instances where controls and displays are arranged "B"'on

left and "A" on right (not typical sterotype). (6.7.C)

Approximately 62 devices are being moved to correct violations of conventional arrangements (per ECN 2710). In addition, demarcation lines and

functional name tags will be added.

7.3 Maintenance and test C/D are located on some panels that contain important

operating functions. (6.7.B)

Watts Bar SER D-10
Demarcation lines and functional name tags will be added for these controls

and displays to set them apart from the C/D for operating functions. An

effort will be made to relocate the test devices to a panel outside the

horseshoe.

7.4 Some functional groups of valves are not arranged according to operating

sequence. (6.7.D)

Valves are grouped by subfunctions within their representative system (per

ECN 2710). The grouping of valves by function within their respective

system is preferable to operating sequences that can vary for different

plant operating modes. Also demarcation lines and functional name tags

will be added (per ECN 2604).

7.5 Pattern recognition requirements on statalarm panels are too complex for

rapid verification. (6.7.F)

This will be corrected by the addition of functional name tags and demarcation lines (per ECN 2604).

7.6 Although most control switches are color coded by system, associated

displays are not color coded to aid in establishing control/display

relationship. (6.7.G)

This will be corrected by the addition of demarcation lines and functional

name tags (per ECN 2604).

7.7 Some annunciator windows for panel 1-M-9 are located above-panels 1-M-3

and 1-M-5. (6.7.1)

Because panel 1-M-9 is outside the horseshoe and is not continuously

monitored by the operator, the ventilation annunciators have been moved

inside the horseshoe where they are readily observable by the operator.

7.8 Amps display for feedwater pump B is mounted over a vertical string of

pump A-related switches. (6.7.M)

This has been corrected. On panel 1-M-4, Pump A-A motor current indicator

(EI-3-119A) is now located above pump A-A motor control (1-5-3-118A).

Pump B-B motor current indicator (EI-3-129A) is located above pump B-B

motor control (HS-3-128A).

8.0 LABELING AND CODING

8.1 Some control panels lack system, subsystem, and functional labels.

(6.8.A.1)

This will be corrected by the addition of demarcation lines and functional

name tags before fuel load. Labeling will be hierarchial, that is by

system and function. For example, the feedwater and condensate system

will be labeled by system, subsystem (main and auxiliary feedwater,

Watts Bar SER D-11
feedwater control, main feed pump turbine, and so forth), and device

(MFPT A trip reset, drain valve, and so forth).

8.2 No hierarchical system of group labeling is used on controls/displays.

(6.8.A.2)

Labeling will be arranged by system and function. Additionally, demarcation lines will be added (per ECN 2604).

8.3 Many panels have temporary labels. (6.8.A.3)

This will be corrected by the addition of permanent functional name tags

and demarcation lines. Temporary labels have been removed (per ECN.2604).

8.4 Control labels are placed below the control. (6.8.A.4)

Function labels are placed over the device (per ECN 2604). Labels containing the TVA-unique ID number and a brief description are placed below

the device. When name plate size does not allow the device name tag to be

placed on the device, the tag is placed below the device. This is the

case for controllers. Demarcation lines will be added where appropriate.

8.5 Label legibility at normal viewing distances is marginal. Lettering is

not sharp and contrast is low, especially for older labels where lettering is smaller and contrast is lower. (6.8.A.5)

This is being corrected by changing the coloring of the name tags to black

lettering on white background to improve contrast and legibility (per ECN 2604). A larger type font is also being used.

8.6 Some groups of controls/displays have no function identification labels.

(6.8.A.7)

This will be corrected by the addition of functional name tags (per ECN

2604).

8.7 Annunciator panels are not functionally labeled, and individual tiles are

not matrix number codedto assist i'n location. (6.8.A.10)

Functional name tags will be added to the annunciator panels. Windows

will be numbered by placing an engraved tag outside the upper left-hand

corner of the window. The numbers will correspond to the numbering system

already in use in the SOIs and on the drawings.

8.8 Some labels contain erroneous information. (6.8.A.12)

A continuing effort exists to correct erroneous labeling as it is

identified.

8.9 The type font used on labelling was too small, contrast was poor, and

placed on lines inconsistently. (6.8.A.13)

Watts Bar SER D- 12
Name tags have been changed to black lettering on white background to

improve contrast and legibility (per ECN 2604). Font size has also been

increased. Density of information does not permit line standardization.

8.10 No demarcation is used to delineate systems, subsystems, or functional

grouping on controls and displays. (6.8.B)

Demarcation lines will be added (per ECN 2604).

8.11 Most mimics are temporarily affixed to panels and are sometimes confusing.

(6.8.C.1)

The temporary mimics on panel 1-M-1 have been removed or replaced with

permanent mimics.

8.12 The number of colors used for switch backplates is excessive, and shading

differences are sometimes difficult to discriminate. (6.8.D.1)

Color coding is applied to the backplates on a system basis, hence the

large number of colors. To aid in distinguishing between systems, functional name tags and demarcation lines will be added (per ECN 2604).

8.13 A color coding scheme has not been applied to displays. (6.8.D.2)

Functional name tags and demarcation lines will be added (per ECN 2604).

9.0 PROCESS COMPUTER, KEYBOARDS, AND CRT

9.1 To initiate a computer function requires a series of complex operator

actions. (6.9.A)

9.2 Poor input/output indexing requires the operator to search through the

complete document list to find specific data point address information.

(6.9.B)

Operator actions and documentation required to initiate a function via the

computer interface console are described below. TVA feels that the

procedures will not be complex after the operator has been trained and

becomes familiar with the system.

The operator will

(1) reference the computer systems operator's manual for desired function

(2) key in the function and computer point ID via the operator' s console

The operator will have the reference listed below to determine the appropriate computer point ID. These will be available before an operating

license is issued.

a system index which alphabetically lists the systems and the computer

input and output associated with each system

Watts Bar SER D- 13
* a computer point ID alphabetical index that lists the computer inputs

and outputs

a character description alphabetical index that lists the computer

inputs and outputs of the 24-character description for each computer

point

9.3 The process computer and the IBM Selectric typewriters are very limited in

their ability to present pertinent real-time information. (6.9.C)

The IBM Selectric typers (15 characters per second) will be replaced with

General Electric Terminet Model 1200 typers. These typers are faster (120

characters per second). Additionally, there is a Mohawk line printer

capable of printing 300 lines per minute in the main control room. It has

a noise-suppressing hood. The Terminet typers will be mounted in consoles

which are designed for noise suppression.

9.4 The filing of process computer documents is not systematic or clearly

identified (6.9.D)

A systematic filing system will be developed and process computer documents will be clearly identified.

PRIORITY RATING 3 ITEMS

Items identified in the staff control room design review report as priority

rating 3 are not individually addressed in this SER. The staff requires the

applicant to address these items as well as other deficiencies that may be

identified in the detailed control room design review to be conducted

(NUREG-0700), and will expect final resolution of all deficiencies on a schedule consistent with NUREG-0737 ("Clarification of TMI Action Plan Requirements,"

November 1981).

The following list contains those findings that were changed to priority 3

after information furnished by TVA convinced the review team that these were

long-term review items. The information supporting the change is listed

following each finding (or following several related findings).

1.0 DISPLAYS

1.1 Meters mounted high on panels are difficult to read by fifth percentile

persons and introduce parallax problems. (6.5.A.10)

Parallax is an intrinsic problem of analog indication meters. However,

because operator actions are generally precipitated by significant departures from a standard operating band -- not by departures as small as

those indicated by parallax -- this is not a problem.

1.2 Some recorders do not indicate in real time. (6.5.C.2)

Chart paper will be provided that is graduated consistent with the normal

speed of the recorder. Date and time are logged on the leading edge of

the chart paper when it is loaded and when shifts change. When the speed

Watts Bar SER D- 14
is changed, the speed is marked on the chart paper. When chart speed is

known, the time of a particular event can be determined. Chart paper

which is graduated in real time is adjusted accordingly when it is loaded.

2.0 CONTROL/DISPLAY (C/D) ARRANGEMENT

2.1 Some annunciator tiles refer to specific controls or displays which are

either not labeled or which have different label information, including

inconsistent abbreviations. (6,7.E)

2.2 Some annunciator labeling is ambiguous. (6.7.J)

Functional name tags that provide information similar to that provided on

the annunciator tiles will be installed for all controls and displays (per

ECN 2604). Name tag size limitations limit the amount of information.

which can be presented as compared to the annunciator windows. Abbreviations conform to TVA standards. Abbreviations will be reviewed for

consistency in conjunction with the NUREG-0700 review.

2.3 Some annunciator windows on panel 1A are associated with controls located

below panel XC and vice versa (no standardization). (6.7.H)

The above windows are on the same panel and can be read from the acknowledge-reset switch. The quantity of annunciator windows does not permit

their always being located over their related controls. In these

instances, they are located as close to their related controls as

possible.

3.0 LABELING AND CODING

3.1, The control position markings for controls mounted high on panels are

difficult to view. (6.8.A.6)

The present system can accommodate a fifth percentile female even though

some body movements may be required to accomplish tasks involving controls

located in the vertical portion of the panels. In addition, the controls

in question are not safety related.

3.2 Position markings for some J-handle switches are not clear. Some twoposition switches have only one-position labels. (6.8.A.8)

This is not a problem because operator training is such that the operators

understand the use of this function. Also, many switches are momentary

contact where only one position actually initiates a specific action.

Spring-loaded, single-position switch labels will be evaluated as part of

the NUREG-000 review.

3.3 Some controls have more than two status lights associated with them, with

no labeling to indicate the purpose or meaning of the lights. (6.8.A.11)

3.4 Turbine pump indicators have two red lights but no label for either.

(6.8.A.14)

Watts Bar SER D-15
Status lights are color coded per TVA standards. Red denotes the active

or inservice state of a circuit, green denotes the inactive or out-ofservice state of a circuit, and white denotes an abnormal or disagreement

condition. Switches which have two or more identically colored indicator

lights (for example, turbine pump indicators) have been identified .(see

list below) and description labels will'be installed.

Panel 1-M-2

HS-47-61A - seal oil backup pump and turning gear oil pump

HS-47-111A - turning gear and bearing lift pumps

Panel 1-M-3

HS-3-116A - HPFP HDRA isolation valve

HS-3-126A - HPFP HDRB isolation valve

HS-3-136 - HPFP HDRA isolation valve

HS-3-179A - HPFP HDRB isolation valve

Panel-l-M-5

HS-68-8A - RC pump 1

HS-68-31A - RC pump 2 normal feeder breaker and oil pump

HS-68-50A - RC pump 3 normal feeder breaker and oil pump

HS-68-73A - RC pump 4 normal feeder breaker and oil pump

HS-62-108A - Charging pump A-A motor control

HS"62-104A - Charging pump B-B motor control

Panel 1-M-15

HS-27-9A - CCW pump 1A motor and valve control

HS-27-19A - CCW pump 1B motor and valve control

HS-27-29A - CCW pump 1C motor and valve control

HS-27-39A - CCW pump iD motor and valve control

Panel O-M-27A

HS-67-146A - CC HTXA DISCH control valve

HS-67-151A - CC HTXC DISCH valve to HRD A

HS-67-152A - CC HTXC DISCH valve to HRD B

Panel O-M-27B

HS-65-81 - shield building vent and containment annulus isolation

HS-65-83 - shield building vent and containment annulus isolation

Panel 1-M-9

HS-30-lA - purge air supply unit A

HS-30-4A - purge air supply unit B

HS-30-11A - instrument room supply and exhaust fans

Watts Bar SER D- 16
Some manufacturer-supplied equipment utilizes a color code (different from

the TVA color code. Through operator training and labeling (where pos- sible), this problem is minimized.

The grouping of lights over the three turbine oil pump switches indicates

oil pump ACB position and the presence or absence of oil pressure. Operators are aware of this through training and experience. There

presently is inadequate room to label each lamp. The pump switch position markings will be evaluated as part of NUREG-0700 review.

4.0 PROCESS COMPUTER, KEYBOARDS, AND CRT

4.1 The status monitoring system was not functional. (5.9.G)

Difficulties with hardware and software procurement preclude the operation of the status monitoring system before an operating license is issued.

TVA is evaluating the possibility of a replacement system on the comple- tion of the existing system.

CONTROL ROOM DESIGN REVIEW FINDINGS THAT WERE DETERMINED TO BE INVALID BASED

ON ADDITIONAL INFORMATION PROVIDED BY THE APPLICANT

The following section contains those findings that were deleted after the

control room design review report was issued, because of additional information

furnished by TVA; The information supporting deletion is listed following each

finding (or following several related findings).

1.0 ANNUNCIATORS

1.1 There is no autoclear capability when alarmed parameters are returned

within tolerance limits. Each alarmed tile requires operator manual

action to clear. (5.3.A)

1.2 There is no capability for silencing an alarm while preserving the flashing tiles (for diagnostic purposes). (6.3.8)

The present system requires manual action to clear the alarms. If the

alarmed condition has returned to normal, the window light goes off when

reset. If the alarmed condition has not returned to normal, the window

light goes from a flashing to a steady state when acknowledged. The

present manual-action-to-clear feature is acceptable because it requires the operator to give specific attention to the annunicator window and

hence the parameter being alarmed.

The duration of the horn is adjustable and will be adjusted from 30 to

approximately 5 seconds. The window will continue to flash until acknowledged.

1.3 The contrast between steady-state and flashing tiles is poor. (6.3.F)

The present flash rate of two to three flashes per second is adequate according to the guidance provided in NUREG-0700.

Watts Bar SER D- 17
2.0 CONTROLS

2.1 Handles of star switches obscure some of the labels. (6.4.G)

Many star-handled switches are located on vertical surfaces where label

viewing is not a problem. For switches located elsewhere, this is not a

problem because operators are able to move their heads if they are unable

to view a tag from a particular angle. Additionally, these switches are

shape coded to aid in the identification of the "selector" function of the

switch.

3.0 DISPLAYS

3.1 The status of the MFPT A turbine trip/reset switch is confusing. It

indicates "Red" when in the reset position and "Green" when tripped.

(6.5.B.2)

The colors used to denote switch status conform to TVA standards. Red

denotes the active or inservice state of a device, and green denotes the

inactive or out-of-service state of a device. This switch conforms to

this standard.

3.2 There is no recorder for indicating cold-leg temperature. (6.5.C.1)

There are presently two recorders for indicating cold-leg temperature.

The cold leg 1 and 2 temperature reactor is on panel 1-M-5 (cutout 66,

device TR-68-18). The cold-leg-3 and -4 temperature recorder also is on

panel 1-M-5 (cutout 69, device TR-68-60).

3.3 The use of the CT flux differential meter data requires the use of the

Technical Specification limit curve. (6.5.C.3)

The CT meter is not an identifiable item; however, the delta I flux meter

is. The curve for this device is provided for operator convenience and is

not required because the plant computer utilizes this curve and automatically signals the operator if there are any deviations from it. The curve

will not be mounted on the panel but will be located in the curve block.

The curve is used because the delta I flux changes throughout the life of

the core.

4.0 LABELING AND CODING

4.1 The electrical distribution panel mimic is incomplete. (6.8.C.2)

The auxiliary power mimics on panel I-M-1 are complete as designed. The

switchyard panel is not part of the reactor operator's area of responsibility.

5.0 PROCESS COMPUTER, KEYBOARDS, AND CRTs

5.1 No procedures could be found covering the necessary actions to take in the

event the computer failed. (6.9.E)

Watts Bar SER D- 18
The status of the plant process computer does not affect either normal

operating or emergency operating procedures, because these procedures

assume that no computer is available.

The primary purpose of the plant computer system is to present information

to the operator to assist in the operation of the nuclear steam supply

system and provide information on any offnormal conditions.

The computer system performs the following functions: obtains data by

scanning analog and digital sensors, performs limit supervision, alarms

offnormal conditions, logs various information on hard copy devices,

performs trending on analog recorders, monitors reactor control and

protection system operation, and monitors nuclear calculations.

Watts Bar SER D- 19

APPENDIX E

PRINCIPAL CONTRIBUTORS

NRC Personnel

J. Rajan

E. Hemminger

S. Kim

N. Romney

R. Rothman

J. Alterman

G. Staley

D. Gupta

K. Dempsie

F. Litton

J. Halapatz

B. Elliot

.M. Hum

D. Huang

D. Smith

J. Gleim

C. Sellers

J. Spraul

G. Bagchi

M. Haughey

J. Kennedy

J. Wing

B. Turovlin

P. Sears

D. Kubicki

C. Ferrell

M. Karlowicz

I. Dinitz

T. Chan

D. Wilson

H. Li

J. Knox

R. Giardina

F. Orr

J. Fairobent

F. Akstulewicz

K. Dempsie

J. Boegli

R. Serbu

J. Pulsipher

J. Vogelwede

M. Dunenfeld

Branch

Mechanical Engineering

Mechanical Engineering

Structural Engineering

Structural Engineering

Seismology

Geology

Hydrology

Geophysical Engineering

Accident Evaluation

Materials Engineering

Materials Engineering

Materials Engineerimg

Materials Engineering

Materials Engineering

Materials Engineering

Materials Engineering

Materials Engineering

Quality Assurance

Seismic Qualification

Seismic Qualification

Environmental Qualification

Chemical Engineering

Chemical Engineering

Fire Protection

Fire Protection

Siting Analysis

Financial Qualification

Indemnity Requirements

Auxiliary Systems

Instrumentation and Control

Instrumentation and Control

Power Systems

Power Systems

Reactor Systems

Meterology

Accident Evaluation

Accident Evaluation

Effluent Treatment

Radiological Assessment

Containment Systems

Core Performance

Core Performance

Watts Bar SER )E-1
J.

S.

B.

F.

D.

B.

M.

C.

M.

M.

C.

A.

R.

Holonich

Salah

Pedersen

Leiderbach

Ficher

Clayton

Goodman

Anderson

Virgilio

Gaitanis

Gaskin

Ramey-Smith

Ramirez

Core Performance

Operator Licensing

Licensee Qualifications

Procedures and Test Review

Procedures and Test Review

Procedures and Test Review

Procedures and Test Review

Generic Issues

Technical Specifications

Emergency Planning

Security

Human Factors Issues

Human Factors Issues

Contractor Branch

Argonne National Laboratories

Gage, Babcock, and Associates, Inc.

Pacific Northwest Laboratories

Lawrence Livermore National Laboratory

BioTechnology, Incorporated

National Bureau of Standards

Mechanical Engineering

Instrument. and Controls

Chemical Engineering (Fire Protection)

Procedures and-Test Review

Control Room Design Review

Control Room Design Review

Control Room Design Review

Watts Bar SER E-2
NRC77 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER (Assigned by DDC) (7,77) US ULA EUAOYCMISO

BIBLIOGRAPHIC DATA SHEET NTU.G-0847

4. TITLE AND SUBTITLE (Add Volume No., if appropriate) 2. (Leave blank)

Safety Evaluation Report related to operation of the

Watts Bar Nuclear Plant, Units 1 and 2 3. RECIPIENT'S ACCESSION NO.

7. AUTHOR(S) 5. DATE REPORT COMPLETED

MONTH YEAR

June 1982

9. PERFORMING ORGANIZATION NAME AND MAILING ADDRESS (Include Zip Code) DATE REPORT ISSUED

U. S. Nuclear Regulatory Commission MONTH -YEAR

Office of Nuclear Reactor Regulation June 1982

Washington, D. C. 20555

8. (Leave blank)

12. SPONSORING ORGANIZATION NAME AND MAILING ADDRESS (Include Zip Code) 10. PROJECT/TASK/WORK UNIT NO.

Same as 9 above

11. CONTRACT NO.

13. TYPE OF REPORT PERIOD COVERED (Inclusive dates)

15. SUPPLEMENTARY NOTES 14. (Leave blank)

Docket Nos. 50-390 and 50-391

16. ABSTRACT (200 words or less) .

The Safety Evaluation Report for the application filed by the Tennessee Valley Authority,

as applicant and owner, for a license to operate the Watts Bar Nuclear Plant, Units 1 and

2 (Docket Nos. 50-390 and 50-391), has been prepared by the Office of Nuclear Reactor

Regulation of the U.S. Nuclear Regulatory Commission. The facility is located in Rhea

County, Tennessee, near the Watts Bar Dam on the Tennessee River. Subject to favorable

resolution of the items discussed in this report, the staff concludes that the facility

can be operated by the applicant without endangering the health and safety of the public.

17. KEY WORDS AND DOCUMENT ANALYSIS 17a. DESCRIPTORS

17b. IDENTIFIERS/OPEN-ENDED TERMS

18. AVAILABILITY STATEMENT i AS (This report) 21. NO. OF PAGES

Unlimited 20. EC.RITY.C._ASS (This page) 22. PRICE

unclassiried S

NRC FORM 335 (7.77)