ML17054B406

Forwards Draft SER Table of Contents & Chapter 1 for Feb 1984 ACRS Meeting

ML17054B406
Person / Time
Site:	Nine Mile Point
Issue date:	01/29/1985
From:	Weinkam E Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 8502190251
Download: ML17054B406 (68)

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Docket No.. 50-410 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 JAN 28 198~

NOTE TO FILE FROM:

SUBJECT:

Edward J.

Meinkam III, Project Manager Licensing Branch No.

2 Division of Licensing NINE MILE POINT UNIT 2 SER I provided a copy of the draft Nine Mile Point Unit 2 SER Table of Contents and Chapter 1 to the applicant, Niagara Mohawk Power Corporation, for their use in preparing for the February ACRS meeting.

Enclosure:

As stat 6

Noted:

wencer 1

cc:

PDR LPDR Edward J.

le nkam III, Project Manager Licensing anch No.

2 Division of Licensing 8502190251 850129'oR oooo'oooosxo PDR

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Nine Hile Point 2 SER Jl.

TABLE OF CONTENTS Pacae ABSTRACT......

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1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT..

Introduction.

General Plant Description................

Shared Facilities and Equipment.

Comparison p/ith Similar Facility Designs Identification of Agents and Contractors.

Summary of Principal Review Matters Hodifications to the Facility During the the Staff Review...........

1.8 Outstanding Issues 1.9 Confirmatory Items

1. 10 License Condition Items 1.11 Unresolved Safety Issues 1.1 1.2 1.3 1.4 1.5 1.6 1.7 2

SITE CHARACTERISTICS....

Course of 1"1 1-3 1-6 1-7 1-7 1-8 1"9 1"9 1-9 1-9 1-9 2-1

2. 1 Geography and Demography 2-1 2.1.1 2.1.2 2.1.3

2. 1.4 Site Location and Description.

Exclusion Area Authority and Control.

Population Distribution.

Conclusion.

2-1 2-1 2-2 2-4 2.2.

Nearby Industrial, Transportation, and Military Facilities 2.2.1 2.2.2 2.2.3 Transportation Routes Nearby Facilities..

Conclusion 2-4 2-5 2-6 2.3 Meteorology.

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2-6 2.3. 1 2.3. 2 2.3.3

2. 3.4 2.3.5 Regional Climatology.

'ocal Meteorology.

Onsite Meteorological Measurements Program.

Short-Term (Accident) Diffusion Estimates..

Long-Term (Routine) Diffusion Estimates 2-6 2-7 2-8 2-9 2-9 2.4 Hydrologic Engineering.

2-10 2.4. 1 Hydrologic Description.

2.4.3 Probable Haximum Flood on Streams and Rivers ~

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2-10 2-11 2-12 NHP-2 SER v

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TABLE OF CONTENTS (Continued) 2.4.4 2.4. 5 2.4. 6 2.4.7 2.4 ~ 8 2:4. 9

2. 4. 10

2. 4. 11 2.4. 12 2.4. 13 2.4. 14 2.4. 15 Potential Dam Failures............

Probable Maximum Surge and Seiche Floodi Probable Maximum Tsunami Flooding.

Ice Effects Cooling-Water Canals and Reservoirs Channel Diversions Flood Protection Requirements.......

Cooling Water Supply Groundwater........

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Accidental Releases of Liquid Effluents Groundwater and Surface Waters Technical Specifications and Emergency Operation Requirements.....

Conclusions ng o

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2-12 2-12 2-13 2-13 2-14 2-14 2-15 2-16 2"17 2-18 2-19 2-19 2.5 Geology and Seismology.

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2-19 2.5.1 2.5.2 2.5.3 2.5

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Seismology..

Capable Faults Stability of Subsurface Materials and Foundations...

Stability of Slopes Embankments and Dams 2-20 2-32 2-35 2-35 2-46 2-46 3

DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS.........

3-1 3.1 3.2 General.............

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Classification of Structures,

Systems, and Components 3-1 3-1 3.3 3.2. 1 Seismic Classification.

3.2.2 System equality Group Classification.

V Wind and Tornado Criteria'nd Loadings 3-1 3-2 3-3 3.4 3.3.1 Wind Design Criteria...............

3. 3. 2 Tornado Des ign Criteri a.

Water Level (Flood) Design.

3-3 3-4 3-5 3.5 3.4. 1 Flood Protection 3.4.2 Water Level (Flood) Design Procedures.

Hlsslle Protection 3-5 3-6 3-7 3.5.1 Missile Selection and Description.....

3.5.2 Structures,

Systems, and Components To Be Protected from Externally Generated Hissil

3. 5. 3 Barrier Design Procedures es.....

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3-7 3-17 3-18 NHP-2 SER vl

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TABLE OF CONTENTS (Continued) 3.6 Protection Against Dynamic Effects Associated With the Postulated Rupture of Piping.......

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3-19 3.6.1 3.6.2 Plant Design for Protection Against Postulated Piping Failures in Fluid Systems Outside Containment..

Determination of Rupture Locations and Dynamic Effects Associated With the Postulated Rupture of Piping..

3-19 3-20 3.7 Seismic Design.

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3-23 3.7.1 3.7.2 3.7.3 3.7.4 Seismic Seismic Seismic Seismic Design Parameters System Analysis Subsystem Analysis Instrumentation System 3-23 3-23 3-24 3-25 3.8 Design of Seismic Category I Structures..

3-26 3.8. 1 3.8.2 3.8.3 3.8.4 3.8. 5 3.8.6 Concrete Containment...

Steel Containment...

Concrete and Structural Structures Other Seismic Category Foundations......

Structural Audit...

Steel Internal I Structures..

3-26 3-28 3-28 3-30 3-32 3-33 3.9 Mechanical Systems and Components..

3-34 3.9. 1 3.9.2 3.9.3 3.9.4 3.9.5 3.9.6 Special Topics for Mechanical Components......

Dynamic Testing and Analysis of Systems, Components, and Equipment.'SME Code Class 1, 2, and 3 Components, Component Supports; and Core-Support Structures...........

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Control Rod Drive Systems Reactor Pressure Vessel Internals Inservice Testing of Pumps and Valves 3-34 3-35 3-41 3-46 3-47 3-48 3.10 Seismic and Dynamic qualification of Seismic Category I Mechanical and Electrical Equipment..............

3-50 3.10.1

'ei'smic and Dynamic qualification..

3. 10. 2 Pump and Valve Operability Assurance........

3-50 3-52

3. 11 Environmental qualification of Electrical Equipment Important to Safety and Safety-Related Mechanical Equipment.

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3"54

3. 11. 1 Introduction.

3-54 NMP-2 SER vl 1

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TABLE OF CONTENTS (Continued)

3. 11. 2

3. 11. 3 Background......

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Completeness of the Environmental qualification Program....

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3-54 3"55 4

REACTOR..............

4-1 4.1 4.2 4.3 Summary Description...

Fuel System Design..........

Nuclear Design..................

4-1 4-1 4"4

4. 3. 1 Evaluation Description.

4.3.2 Evaluation Findings 4-4 4-5 4.4 Thermal and Hydraulic Design.

4"6 4.4.1 4.4.2 4.4.3 4.4.4 4.4.5 4.4.6 4.4. 7 4.4.8 4.4. 9 Thermal-Hydraulic Design Bases

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Thermal-Hydraulic Analysis Basis.

Thermal-Hydraulic Analysis Methods Thermal-Hydraulic Stability.

Crud Deposition...

Loose-Part Monitoring System.

NUREG-0737 Item II.F.2, Inadequate Cooling Detection System.

Thermal-Hydraulic Comparison..

Evaluation Findings Core 4-6 4-7 4-7 4-7 4-8 4-8 4-9 4"11 4-12 4.5 Reactor Materials 4-12 4 '

4.5. 1 Control Rod Drive Structural Materials 4.5.2 Reactor Internals and Core-Support Materials Functional Design of Control Rod Drive Systems.....

4-12 4-13 4-14 5

REACTOR COOLANT SYSTEMS 5-1 5.1 5.2 Summary Description.

Integrity of Reactor Coolant Pressure Boundary.....

5"1 5-1 5.2.1 5.2.2 5.2.3 5.2.4 5.2.5 Compliance With ASME Codes and Code Cases...

Overpressurization Protection.

Reactor Coolant Pressure Boundary Materials.

Reactor Coolant Pressure Boundary Inservice Inspection and Testing...;

Reactor Coolant Pressure Boundary Leakage Detection 5-1 5-2 5-5 5-7 5-9 5.3 Reactor Vessel.

5-10 5.3. 1 5.3.2 5.3.3 Reactor Vessel Materials.

Pressure-Temperature Limits...

Reactor Vessel Integrity..

5-10 5-15 5-16 NMP-2 SER V111

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TABLE OF CONTENTS (Continued)

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5.4 Component and Subsystem Design.....................

5-16 5.4.1 5.4.2

5. 4. 3" 5.4. 4*

5.4. 5" 5.4. 6 5.4.7 5.4.8 5 4 9*

5.4. 10" 5.4. 11 5.4. 12 Reactor Coolant Pumps Steam Generators.......

Reactor Core Isolation Cooling System......

Residual Heat Removal System.

Reactor Mater Cleanup System..

Pressurizer Relief Tank..

Reactor Coolant System High Point Vents'

5-16 5-16 5-17 5-19 5-22 5-24 5-24 6

ENGINEERED SAFETY FEATURES 6-1

6. 1 Materi al s 6-1 6.1.1 Engineered Safety Features Materials

6. 1.2 Protective Coating Systems (Paints)-

Organic Materials................

6-1 6-2 6.2 Containment Systems 6-2 6.2.1 6.2.2 6.2.3

6. 2.4 6.2.5 6.2.6 06.2.7 Containment Functional Design.....

Containment Heat Removal System.

Secondary Containment Containment Isolation System.

Combustible Gas Control System......

Containment Leakage Testing Program.

Fracture Prevention of Containment Press Boundary.

ure 6-3 6-18 6-20 6-22 6-24 6-27 6-30 6.3 Emergency Core Cooling System.

6-31 6 ~ 3.1 6.3 ~ 2 6.3.3 6.3.4 6.3.5 6.3.6 System Description.

Evaluation of Single Failures qualification of Emergency Core System.

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Testing.

Performance Evaluation.

Conclu'sions Cooling

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6-31 6-32 6-33 6-35 6-35 6-38 "The July 1981 revision of the Standard Review Plan (NUREG-0800) does not include sections addressing FSAR sections that consist of background or design data used in the review of other sections.

The section numbers have been retained in this report to provide continuity and to ensure a close correlation between subsequent SER sections and their associated SRP sections.

NMP-2 SER 1X

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TABLE OF CONTENTS (Continued) 6.4 6.5 Control Room Habitability SRP "Systems".....

Engineered-Safety-Feature Atmosphere-Cleanup System...

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6-39 6"39 6.5.1 6.5.2 6.5.3

6. 5.4 6.5.5

System Description

and Containment Spray As a System.

Fission-Product-Control "and Structures".......

Ice Condenser As a Fiss System.

Evaluation Findings....

Evaluation......

Fission-Product-Cleanup Systems SRP

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ion-Product-Cl eanup 6"39 6-41 6-41 6"41 6-41 6.6 Inservice Inspection of Class 2 and 3 Components...

6-42 6.6.1 6.6.2 6.6.3 6.6.4 Compliance Mith the Standard Review Plan....

Examination Requirements Evaluation of Compliance Mith 10 CFR 50.55a(

for NHP-2.....

Conclus>ons...

g) 6-42 6-43 6-43 6-44 6.7 Hain Steam Isolation Valve Leakage-Control System o

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6-44 7

INSTRUHENTATION AND CONTROLS 7-1 7.1 Introducti on.

7-1 7.1.1 7.1.2 7.1.3 7.1.4 7.1.5 Acceptance Criteria.

Hethod of Review.

General Findings..

Specific Findings...........

THI-2 Action Plan Items 7-1 7-1 7-1 7-2 7-3 7.2 Reactor Trip System.

7-4

7. 2. 1 System Descr iption..

7. 2. 2 Specific Findings.....

7.2.3 Evaluation Findings.........

7-4 7-6 7-13 7.3 Engineered Safety Features Systems.

7-14 7.4

7. 3. 1 System Descriptions..

7. 3. 2 Speci fic Findings

7. 3. 3 Evaluation Findings Systems Required for Safe Shutdown.

7-14 7-35 7-37 7-39

7. 4. 1 System Descriptions

7. 4. 2 Specific Findings..

7.4 ~ 3 Evaluation Findings 7-39 7-41 7-43 NHP-2 SER X

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TABLE OF CONTENTS (Continued)

7. 5 Safety-Related Display Information..........................

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7-44 7.5. 1 System Descriptions...................

7.5.2 'pecific Findings......................

7.5.3 Evaluation Findings..

7"44 7-47 7-49 7.6 Other Instrumentation Systems Important to Safety...........

7.6. 1 System Descriptions.

7.6.2 Specific Findings.....

7.6.3 Evaluation Findings.....

7-50 7-50 7-53 7-55 7.7 Control Systems...

7.7. 1 System Descriptions...

7.7.2 Specific Findings 7.7.3 Evaluation Findings 8

ELECTRIC POWER SYSTEMS 7-56 7-56 7-59 7-61 8-1 8.1 Acceptance Criteria.

8.2 Offsite Power System.

8.2.1 General Description.

8. 2 ~ 2 Grid Analysis.

8.2. 3 Conclusions 8-1 8-1 8-1 8-2 8-2 8.3 Onsite Emergency Power Systems 8.3. 1 AC Power System (On Site)...................

8.3.2 DC Power Systems (On Site).......

8.3.3 Conclusions 8-3 8"3 8-7 8-8 8.4 Other Electrical Features'nd Requirements for Safety.....

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8-8 8.4.1 8.4.2 8.4.3 8.4.4 8.4. 5 8.4.6 8.4.7 8.4.8 Adequacy of Station Electric Distribution System Voltages

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Containment Electrical Penetrations.....

Thermal Overload Protection Bypass Power Lockout to Manually Controlled, Electrically Operated Valves........

Physical Identification and Independence of Redundant Safety-Related Electrical Systems.

Non-Safety Loads on Emergency Sources Flooding of Electrical Equipment.........

Load-Sequencing Design.

8-8 8-10 8-11 8-11 8-12 8-14 8-15 8-15 NMP-2 SER X1

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TABLE OF CONTENTS (Continued)

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9 AUXILIARYSYSTEMS......................

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

9-1 9.2 9.1.1 9.1.2 9.1.3 9 ~ 1.4 9.1.5 Water New-Fuel Stoppage.....;....

Spent-Fuel Storage......

Spent-Fuel-Pool Cooling and Cleanup Light-Load Handling System (Related Refueling)..

Overhead Heavy-Load Handling System.

ystems...............

S Systemo

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to 9"1 9"3 9-5 9-8 9" 10 9-12 9.2.1 9.2.2 9.2.3 9.2.4 9.2.5 9.2.6 9.2.7 9.2.8 Station Service Mater System..

Reactor Auxiliary Cooling Mater System (Reactor Building Closed-Loop Cooling Water System).........

Demineralized Water Makeup System (Makeup Water Treatment System)...........

Potable and Sanitary Mater Systems (Domestic Mater, Sanitary Drains, and Disposal Systems).............

Ultimate Heat Sink.

Condensate Storage Facilities...

Turbine Building Closed"Loop Cooling Mater System..

Plant Chilled Water System..

9-12 9-13 9-14 9-15 9-15 9-17 9-18 9"18 9.3 Process Auxiliaries..

9-20 9.3.1 9.3

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Equipment and Floor Drainage System.

Chemical and Volume Control System...

Standby Liquid Control System.

9-20 9"21 9-27 9-28 9-28 Heating, Ventilation, and Air Conditioning Systems......

9-29'.5 9.4.1 9.4. 2 9.4. 3 9.4.4 9.4.5 Other 9.5. 1 9.5.2 9.5.3 9.5.4 Control Building and Normal Switchgear Building HVAC System.

Spent-Fuel Pool Area Ventilation System (Reactor Building HVAC System)..

Radwaste Building HVAC System.

Turbine Building Ventilation System..

Engineered Safety Feature Ventilation System.....

Auxiliary Systems Fire-Protection Progr am.

Communications Systems..

Lighting Systems..

Emergency Diesel Engine Fuel Oil Storage and Transfer System..

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9-29 9-31 9-33 9-33 9-34 9-36 9-36 9-47 9-49 9-51 NHP-2 SER xi 1

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TABLE OF CONTENTS (Continued)

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-~e 9.5.5 9.5.6

9. 5.7 9.5.8 Emergency Diesel Engine Cooling Water System...

Emergency Diesel Engine Starting System........

Emergency Diesel Engine Lubrication System.....

Emergency Diesel Engine Combustion-Air-Intake and Exhaust System.............

9-56 9-59 9-62 9-65 10 STEAM AND POWER CONVERSION SYSTEM..

10"1 10.1 Summary Description....

10.2 Turbine Generator.....................

10-1 10-1

10. 2. 1*

10.2.2" 10.2.3 Turbine Disk Integrity 10.3 Main Steam Supply System.........

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10-3 10-3 10.3. 1 Hain Steam Supply System (Up To and Hain Steam Isolation Valves)......

10.3.2*

10. 3. 3"

10. 3. 4"

10. 3. 5" 10.3.6 Steam and Feedwater System Materials Including the 10-4 10-5 10.4 Other Features......................

10-6 10.4. 1 10.4. 2 10.4. 3

10. 4. 4

10. 4. 5

10. 4. 6

10. 4. 7

10. 4. 8

10. 4. 9 Hain Condenser.

Hain Condenser Evacuation System..

Turbine Gland Sealing System..

Turbine Bypass System.

Circulating Water System..

Condensate Demineralizer System..

Condensate and Feedwater System...

Steam Generator Blowdown System (PWR Auxiliary Feedwater System (PWR).

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10-6 10-7 10-8 10-9 10-10 10-10 10-11 10-12 10-12 11 RADIOACTIVE WASTE HANAGEMENT 11.1 Liquid and Gaseous Effluent Source Terms 11.2 Liquid Waste Management Radwaste System ll"1 11-2 ll.2.1 System Description.

11. 2. 2 Evaluation Findings 11-2 11-3 "The July 1981 revision of the Standard Review Plan (NUREG-0800) does not include sections addressing FSAR sections that consist of background or design data used in the review of other sections.

The section numbers have been retained in this report to provide continuity and to ensure a close correlation between subsequent SER sections and their associated SRP sections.

NMP-2 SER xi 1 1

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TABLE OF CONTENTS (Continued) ll.3 Gaseous Waste Hanagement Systems...................'...

11-5 ll.3. 1 ll.3. 2 System Description......

Evaluation Findings....

11-5 11-6 ll.4 Solid Waste Hanagement System........

11-7 Il.5 11.4. 1 System Description.

11.4.2 Evaluation Findings..........

Process and Effluent Radiological Monitoring and Sampling Systems...................

11-7 11-8 11"9 ll.5. 1

11. 5. 2 ll.5. 3 System Description...

NUREG-0737 Item II.F.l P Accident Monitoring Instrumentation.

Evaluation Findings 11-9 11"ll ll"12 12 RADIATION PROTECTION.

12-1

12. 1 Ensuring That Occupational Radiation Exposures Are As As Is Reasonably Achievable Low 12-1

12. l. 1 Policy Considerations..

12. 1. 2 Design Considerations

12. 1.3 Operational Considerations 12-1 12-2 12-2

- 12.2

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

12. 3 ~ 1

12. 3. 2

12. 3. 3

12. 3. 4 Facility Design Features.....

Shielding.

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Ventilation...

Area Radiation and Airborne Radioactivity Honitoring Instrumentation....................

12"4 12-5 12-6 12-7

12. 4

12. 5 Dose Assessment............

Operational Radiation Protection Program.

12"8 12-9

12. 5. 1

12. 5. 2

12. 5. 3 Organi zati on.

Equipment, Instrumentation, and Facilities....

procedures................

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12-9 12-10 12-11 13 CONDUCT OF OPERATIONS 13-1

13. 1 Organizational Structure and Operations

13. l. 1 Management and Technical Support Organization.

13. 1.2 Operating Organization.

13-1 13-1 13"2 NMP-2 SER xiv

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13. 2 13.3 T

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ra>ning...........................

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13.2. 1 Reactor Operator Training Program................

13.2.2 Training for Nonlicensed Plant Staff....

Emergency Planning.............

13.3. 1 Introduction...................

13.3.2 Evaluation of the Emergency Plan.....

13.3.3 Federal Emergency Management Agency (FEMA) Findin on Offsite Emergency Plans and Preparedness......

13.3.4 Conclusions gs

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13-4 13-4 13-12 13-12 13-12 13-13 13-27 13-27

13. 4 Operational Review 13-28

13. 4. 1 Review and Audit.......

13.4.2 TMI Action Plan Items 13-28 13-28

13. 5 Plant Procedures 13-28

13. 6 13.5. 1 Administrative Procedures 13.5.2 Operating and Maintenance Procedures Physical Security Plan.......

13-28 13-31 13-36

13. 6. 1 Introduction..

13.6.2 Staff Evaluation.

13.6.3

- Physical Barriers 13-36 13-36 13-36 14 INITIALTEST PROGRAM.

14-1 15 ACCIDENT ANALYSES 15-1

15. 1 15.2

15. 3

15. 4 Anticipated Operational Occurrences Decrease in Heat Removal.......

Accidents Reactivity and Power Distribution Anomalies........

15-2'5"7 15-7 15-7 15.4. 1 Continuous Rod Withdrawal During Reactor Startup.

15.4.2 Rod-Withdrawal Error at Power 15.4.3 Control Rod Misoperation.

15.4.4/15.4.5 Startup of a Recirculation Loop at an

'nc'orrect Temperature.'5.4.6 Chemical and Volume Control System Malfunction That Results in a Decrease in the Boron Concentration in the Reactor Coolant (PWR).......

15.4.7 Operation of a Fuel Assembly in an Improper Position - Fuel-Misloading Event.

15.4.8 Spectrum of Rod-Ejection Accidents (PWR).........

15.4u9 Spectrum of Rod-Drop Accidents...................

15-7 15"8 15-10 15-10 15-10 15-10 15-11 15-11 NMP"2 SER xv

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TABLE OF CONTENTS (Continued) 15.5 Increase in Reactor Coolant System Inventory.

15.6 Decrease in Reactor Coolant Inventory..

15. 6. 1

15. 6. 2

15. 6. 3

15. 6.4

15. 6. 5 Inadvertent Opening of a BWR Relief Valve...........

Radiological Consequences of the Failure of Small Lines Carrying Primary Coolant Outside Containment..

Radiological Consequences of Steam Generator Tube Failure (PWR)

Radiological Consequences of Main Steamline Break Outside Secondary Containment.................

Radiological Consequences of Loss-of-Coolant A

A ccsdent....

15.7 Radioactive Releases From a Subsystem or Component..........

15. 7. 1

15. 7. 2

15. 7. 3

15. 7.4

15. 7. 5 Waste Gas System Failure".

Radioactive Liquid Waste System Leak or Failure (Release to Atmosphere)*...........

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Postulated Radioactive Releases Resulting From Liquid-Containing-Tank Failures Radiological Consequences of Fuel-Handling Accidents Spent Fuel Cask Drop Accidents 15.8 Anticipated Transients Without Scram.

15.9 THI Action Plan Requirements 15.9.1 NUREG-0737 Item II.B.l, Reactor Coolant System Vents

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15. 9. 2 NUREG-0737 Item II.K.l. 1, IE Bulletins on Measures To Mitigate Small-Break LOCAs and Loss-of-Feedwater Accidents

15. 9. 3 NUREG-0737 Item II.K.3, Final Recommendations of Bulletins and Orders Task Force 15.9.4 NUREG-0737 Item III.D.1, Primary Coolant Outside Containment..

16 TECHNICAL SPECIFICATIONS 17 EQUALITY ASSURANCE 17.1 General..

17. 2 Organization....

17.3 guality A'ssurance Program..

17.4 Conclusions 18 HUMAN FACTORS ENGINEERING.

18. 1 Detailed Control Room Design Review.

18.2 Safety Parameter Display System.

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15"14 15-14 15"14 15-14 15-14 15-14 15-15 15-17 15-17 15"17 15-17 15-18 15"19 15"19 15-19 15-19 15-20 15-22 15-26 16-1 17" 1 17-1 17-1 17-1 17"2 18-1 18-1 18-3 "Deleted from the July 1981 revision of the Standard Review Plan (NUREG-0800).

NHP-2 SER XV1

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TABLE OF CONTENTS (Continued)

'fi;rI 19 20 21 22 REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS.

COMMON DEFENSE AND SECURITY............

FINANCIAL QUALIFICATIONS...............

FINANCIAL PROTECTION AND INDEHNITY REQUIREMENTS

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19"1 20-1 21-1 22-1 2 ~ 1 General.....................'.........

2 22.2 Preoperational Storage of Nuclear Fuel...........

22.3 Operating License.............

22-1 22-1 22"1 23 CONCLUSIONS 23"1 APPENDICES Chronology of NRC Staff Radiological Review of Nine Hile Point Nuclear Station, Unit 2 References Nuclear Regulatory Commission Unresolved Safety Issues Acronyms and Initialisms NRC Staff Contributors and Consultants Hark II Chugging Load Specification Effects of Desynchronization Control of Heavy Loads at Nuclear Power Plants LIST OF FIGURES 2.1 2.2 2.3 2.4 2.5 0

2.6 2.7 Exclusion area, Nine Mile Point Nuclear Station.

Site plan and plant facilities, Nine Mile Point Nuclear Station..

Area within 20 km of site of Nine Mile Point Nuclear Station.....

Area within 80 km of site of Nine Mile Point Nuclear Station.....

Revetment-ditch typical cross-section, Nine Mile Point Nuclear Station, Unit 2 General

layout, Nine Mile Point Nuclear Station, Unit 2..........

Typical stratigraphy at Nine Nile Point Nuclear Station, nest 2..................

U 2-49 2"50 2-51 2-52 2"53 2-54 2-55 6.1 6.2 General arrangement of reactor building section 2-2, Nine Mile Point Nuclear Station, Unit 2 Primary containment pressure recirculation pump suction line break with feedwater, case C, Nine Mile Point Nuclear Station, 6-46 6.3 6.4 U t 4

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13. 3 Upper management nuclear organization of Niagara Mohawk Power Corporation for Nine Mile Point Nuclear Station, Unit 2'........

Nuclear and security organization of Niagara Hohawk Power Corporation for Nine Mile Point Nuclear Station, Unit 2..

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Nuclear engineering organization of Niagara Mohawk Power Corporation for Nine Mile Point Nuclear Station, Unit 2..........

13-40 13-41 13-42 NHP-2 SER Xvi 1

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TABLE OF CONTENTS (Continued)

LIST OF FIGURES (Continued)

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13. 6

13. 7 17.1 Nuclear generation organization of Niagara Hohawk Power Corporation for Nine Mile Point Nuclear Station, Unit 2..........

Production organization of Niagara Mohawk Power Corporation for Nine Mile Point Nuclear Station, Unit 2..........

Technical organization of Niagara Hohawk Power Corporation for Nine Hile Point Nuclear Station, Unit 2....

Chemistry and radiation protection organization...

Corporate and gA organization of Niagara Mohawk Power Corporation for Nine Hile Point Nuclear Station, Unit 2..........

13-43 13-44 13-45 13-46 17-4 LIST OF TABLES 1.1 1.2 1.3 1.4 Cross-reference table Outstanding issues Confirmatory issues License conditions for THI Action Plan items........

l-ll 1-14 1-14 1"14 2.1 2.2 Design parameters of rock and soil - design of seismic

'Category I structures Seismic Category I structures

- foundation data......

2-57 2-58 3.1 4.1 5.1 5.2 6.1 6.2 6.3 6.4 9.1 Reliability criteria.

Thermal-hydraulic design comparison.

Sa fety/re 1 ief va1 ve compari son.............................

Pressurization events resulting in pressure-relief actuation..

Comparison of BMR containment designs..

Conformance of Nine Hile Point Nuclear Station, Unit.2, design to NRC acceptance criteria.

Host%limiting combinations for Nine Mile Point Nuclear Station Unit 2

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Demonstration of compliance with ECCS criteria....

Applicant@ conformance to recommendations made in NUREG/CR-0660.

3"56 4-17 5"25 5-25 6-50 6-51 6-58 6-58 9-67 ll.1 ll.2 ll.3 Principal parameters'used in calculating radioactive effluents from Nine Mile Point Nuclear Station, Unit 2..

Design parameters of principal components considered in the evaluation of liquid and gaseous radioactive waste treatment systems for Nine Mile Point Nuclear Station, Unit 2...........

Calculated releases of radioactive materials in liquid effluents from Nine Hile Point Nuclear Station, Unit 2.....

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LIST OF TABLES (Continued)

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11.4 Calculated releases of radioactive materials in gaseous e fluents from Nine Hile Point Nuclear Station, Unit 2,

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11-16 11-17 14.1 Open and confirmatory items, initial test program......

15. 1 Radiological consequences of selected design-basis accidents 15.2 Assumptions used to evaluate the steamline break outside secondary containment doses.

15.3 Assumptions used for the rod-drop accident 15.4 Assumptions used to calculate LOCA doses 15.5 Assumptions used to evaluate fuel-handling accidents........

14"5 15-28 15-29 15"30 15"31 15-31

16. 1 Issues to be included in the Technical Specifications for Nine Hile Point Nuclear Station, Unit 2...

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16" 2 17.1 Regulatory guidance applicable to evaluation of a quality assurance program.

17-5 NHP"2 SER Xix

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1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT

1. I Introducti on This is a Safety Evaluation Report (SER) on the application for an operating license (OL) for the Nine Mile Point Nuclear Station, Unit 2 (NMP-2).

The Niagara Mohawk Power Corporation tendered an application with the Atomic Energy Commission (AEC) for a construction permit (CP) on March 8, 1972.

Inasmuch as more information was needed for the initial filing, the application for a CP was not officially docketed for the safety review until June 15, 1972.

The AEC (now the Nuclear Regulatory Commission (NRC)) reported the results of its preconstruction review in an SER dated June 15, 1973.

Following a public hearing before an Atomic Safety and Licensing Board (ASLB), the Construction Permit for NMP-2 was issued on June 24, 1974.

On October 27, 1978, the CP was amended to add Central Hudson Gas and Electric Corporation, New York State Electric and Gas Corporation, Long Island Lighting

Company, and Rochester Gas and Electric Corporation as co-owners and co-applicants.

On February 28, 1983, the CP was extended to December 31, 1987.

The Niagara Mohawk Power Corporation (hereinafter referred to as the applicant),

acting as agent and representative for the owners, tendered an application for an operating license for NMP-2 by letter dated January 31, 1983.

When the NRC staff (staff) acceptance review was completed, the Final Safety Analysis Report (FSAR) for NMP-2 was docketed on April 12, 1983.

Before issuing an OL for a nuclear power plant, the staff is required to conduct a review of the effects of the plant on public health and safety.

The staff safety review of NMP-2 has been based on NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Reactors, LWR Edition" (SRP)

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An audit review of each of the areas listed in the Areas of Review section of'he SRP was performed according to the guidelines provided in the Review Procedures portion of the SRP.

Exceptions to this practice are noted in the applicable sections of this report.

This SER summarizes the results of the staff's radiological safety review of NMP-2 and delineates the scope of the technical details considered in evalu-ating the radiological safety aspects of its proposed operation.

The design of the station was reviewed against the federal regulations, CP criteria, and the

SRP, except where noted otherwise.

The SRP covers a variety of site condi-tions and plant designs.

Each section is written to provide the complete proce-dure and all acceptance criteria for all of the areas of review pertinent to the section.

However, 'for any given application, the staff may select and emphasize particular aspects of each SRP section as appropriate for the appli-cation.

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 feature may be sufficiently similar to that of a previous plant so that a de novo review of the feature is not needed.

NMP-2 SER

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During the course of its review, the staff held a number of meetings with representatives of the applicant to discuss the design, construction, and pro-posed operation of the plant.

The staff requested additional information, which the applicant provided in amendments to the FSAR.

This information is available to the public for review at the NRC Public Document Room at 1717 H Street, NW, Washington, D. C.,

and at the local Public Document Room at the Penfield Library, State University College,

Oswego, N.Y.

13126.

Following the THI-2 accident, the Commission paused in its licensing activities to assess the impact of the accident.

During this pause, the recommendations of several groups established to investigate the lessons learned from THI-2 became available.

All available recommendations were correlated and assimilated into a "THI Action Plan,"

now published as NUREG-0600, entitled "NRC Action Plan Developed As a Result of the THI-2 Accident."

Additional guidance relating to implementation of the Action Plan is in NUREG-0737, "Clarification of THI Action Plan Requirements,"

and in Supplement 1 to NUREG-0737.

Licensing require-ments based on the lessons learned from the THI-2 accident have been established to provide additional safety margins.

These have been incorporated into the design and operation of NMP-2.

Table l.l provides a cross-reference

.relating the THI items to the sections in this report where they are discussed.

Sections 2 through 22 of this report contain the NRC review and evaluation of both the non-THI-and THI-related issues.

Section 23 presents the staff's conclusions.

Appendix A is a chronology of NRC's principal actions related to the safety (or radiological} review of the application.

Appendix B is a compilation of the printed material (references) used during the course of the review.

Availa-bility of all material cited in this report is described on the inside front cover of this report.

Sections of Title 10 of the Code of Federal Re ulations (10 CFR) (including the General Design Criteria (GDC) in Appendix A to 10 CFR 50),

NRC regulatory guides (RGs),

and sections of the SRP, including Branch Technical Positions (BTPs) will be identified as appropriate; they are not included in Appendix B.

Appendix C is a discussion of how various Unresolved Safety Issues (USIs) relate to the application.

Appendix 0 is.a list of abbreviations used in this report.

Appendix E is a list 'of principal contributors.

Appendix F

is "Hark II Chugging Load Specification Effects of Desynchronization."

Appen-dix G is "Control of Heavy Loads at Nuclear Power Plants, Nine Hile Point Unit 2 (Phase 1)."

In accordance with the provisions of the National Environmental Policy Act (NEPA) of 1969, a Draft Environmental Statement (OES) that set forth the environmental considerations related to the proposed construction and operation of NHP-2 was prepared by the staff and was published in July 1984.

The Final Environmental Statement (FES) is scheduled to be published in February 1985 and wi 11 include a consideration of public comments received on the OES.

The review and evaluation of NHP-2 for an operating license is only one of many stages at which the staff reviews the design, construction, and operating features of the facility.

The facility design was extensively reviewed before the applicant was granted a construction permit for the facility.

Construction of the facility has been monitored in accordance with a detailed monitoring and inspection program at the operating license (OL) stage.

The staff has NMP-2 SER I"2

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reviewed the final design of the facility to determine that the C

regulations have been met. If an operating license is granted, NMP t

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

tions, and the facility will be subject to the staff s continuing inspection program.

In addition to the staff review, the Advisory Committee on Reactor Safeguards (ACRS) will review the application and will meet with both the applicant and the staff to discuss the final design and proposed operation of the plant.

The Committee's report to the Chairman of the NRC will be included in a supple-ment to this SER.

The NRC Project Manager assigned to the OL application for NHP-2 is Hs.

Mary F.

Haughey.

Hs.

Haughey may be contacted by calling (301) 492-7897 or by writing Hs. Hary F.

Haughey Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C.

20555

1. 2 General Plant Descri tion NMP-2 is located on a 364-ha (900-acre) site owned by Niagara Mohawk Power Corporation (NMPC), and is situated on the southeast shore of Lake Ontario,

Scriba, New York, approximately 10 km (6.2 mi) northeast of the. city of Oswego.

NMP-2 and support facilities occupy about 18.2 ha (45 acres) and share the site, with Nine Mile Point Nuclear Station, Unit 1 (NMP-1) (Docket No, 50-220), which has been in commercial operation since 1969.

The Nine Mile Point site is ad-jacent to the James A. FitzPatrick Nuclear Power Plant owned by the Power Authority of the State of New York (PASNY).

NMP-2 is located 274 m (900 ft) east of NHP-1 and about 716 m (2,350 ft) west of the James A. FitzPatrick Plant.

J NMP-2 employs an NSSS consisting of a single cycle, forced-circulation BWR that produces steam for direct use in the steam turbine.

The plant-rated core thermal power level is 3,323 HWt, corresponding to a net electrical output of 1,080

HWe, and design thermal power of 3,463 HWt, corresponding to a gross electrical output of 1,202 MWe.

The thermal power used for the plant transient and loss-of-coolant (LOCA) analyses is 3,463 MWt.

All safety systems have been designed for a thermal power of 3,489 MWt.

The NSSS supplier is General Electric Com-pany - Nuclear Energy Operations (GE-NEO).

The balance of the plant is designed and constructed by Stone and Webster Engineering Corporation (SWEC).

The principal structures of the NMP-2 facility are the primary containment structure, the reactor building and auxiliary bays, the radwaste building, the control building, the diesel generator building, the screenwell building, the intake and discharge

tunnels, the main stack, the normal switchgear building, NMP-2 SER 1-3

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the auxiliary boiler building, the standby gas treatment build>ng, the natural-draft cooling tower, the auxiliary service building, the decontamination

area, and the hydrogen storage area.

The primary containment structure houses the reactor pressure

vessel, reactor recirculation pumps and motors, drywell cooling system unit coolers, safety/

relief valves, accumulators, and other equipment.

The containment design employs the BMR Hark II concept of over-under pressure suppression with multiple downcomers connecting the reactor drywell to the water-filled pressure suppres-sion chamber.

The primary containment is a steel-lined, reinforced-concrete enclosure housing the reactor and the suppression pool.

The reactor building and auxiliary bays enclose the primary containment struc-ture.

These structures provide secondary containment when the primary contain-ment is closed and in service, and provide/ primary containment when the primary containment is open, as during refueling.

These structures house the remaining portions of the NSSS, refueling and fuel storage equipment for the reactor water cleanup (RWCU) system, equipment for the standby liquid control system equipment for the reactor building closed-loop cooling water system, and other equipment.

The outer wall of the reactor building is reinforced concrete up to the crane rail level above the refueling floor.

Above the crane rail level, the super-structure is a steel frame using metal wall panels with sealed joints.

Access to the building is through air locks.

The reactor vessel contains the core and supporting structures; the steam sepa-rators and dryers; the jet pumps; the control rod guide tubes; the distribution lines for the feedwater, core sprays, and standby liquid control; the incore instrumentation; and other components.

The main connections to the vessel include the steam lines, coolant recirculation lines, feedwater lines, control rod drive and incore nuclear instrument housings, core spray lines, residual heat removal lines, standby liquid control line, core differential pressure line, jet pump pressure sensing lines, and water level instrumentation.

The reactor vessel is designed for a pressure of 1,250 psig.

The nominal opera-ting pressure in the steam space above the separat'mrs is 1,020 psia.

The vessel is fabricated of low-alloy steel and is clad internally with stainless steel (except for the top head nozzles and nozzle weld zones which have cladding).

The reactor core is cooled by demineralized water that enters the lower portion of the core and boils as it flows upward around the fuel rods.

The steam leaving the core is dried by steam separators and dryers located in the upper portion of the reactor pressure vessel (RPY).

The steam is then directed to the turbine through the main steamlines.

Each steamline has two isolation valves in series, one on either side of the primary containment barrier.

Four emergency core. cooling systems (ECCSs) are provided to limit fuel cladding temperature to 2,20Q4F in the event of a breach in the reactor coolant pressure boundary that results in a loss of reactor coolant.

These systems are the high-pressure core spray (HPCS), the automatic depressurization system (AOS), the low-pressure core spray (LPCS),

and the low-pressure coolant injection (LPCI).

Operation of the ECCS is initiated automatically, when required, regardless of the availability of offsite power.

NHP-2 SER

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Fuel for the reactor core consists of slightly enriched uranium dioxide (UOz) pellets sealed in Zircaloy-2 tubes.

These tubes (or fuel rods) are assembled into individual fuel assemblies.

Gross control of the core is achieved by movable, bottom-entry, control rods.

The control rods are cruciform in shape and are dispersed throughout the lattice of fuel assemblies.

The control rods are positioned by individual control rod drive lines.

Each fuel assembly has several fuel rods with burnable poison mixed in solid solution with the (UOz).

The burnable poison diminishes the reactivity of the fuel. It is depleted as the fuel reaches the end of the fuel cycle.

The reactor recirculation system consists of two recirculation pump loops external to the RPY.

These loops provide the piping path for the driving flow of water to the RPV jet pumps.

Each external loop contains one high-capacity motor-driven recirculation

pump, two motor-operated maintenance

valves, and one hydraulically operated flow control valve.

The variable position hydraulic flow control valve operates in conjunction with a low frequency motor generator set to control reactor power level through the effects of coolant flow rate on moderator void content.

The jet pumps are RPY internals.

They provide a continuous internal circula-r tion path for the major portion of the co're coolant flow.

The jet pumps are located in the annular region between the core shroud and the vessel inne~ wall.

Any recirculation line break still allows core flooding to approximately two-thirds of the core height, the level of the inlet of the jet pumps.

The power generation complex includes several contiguous buildings:

the reactor building with two auxiliary bays, the control building, the turbine building, and the radwaste building.

Other buildings, such as the security facility, also are located in the general station area.

A screenwell for the circulating and service water systems is located approximately 107 m (350 ft) northwest of the centerline of the reactor building.

Condenser cooling for HMP-2 is provided from a counterflow, natural-draft, hyperbolic, concrete. cooling tower located approximately 330 m (1,003 ft) south of the centerline of the reactor building.

The ultimate heat sink for emergency core cooling is Lake Ontario.

Below grade and north of the screenwell

building, there are two concrete tunnels that convey the service water intake, service water discharge, and cooling tower blowdown.

A safety-related intake pipe is enclosed in each tunnel.

The intake pipes extend from the intake'haft approxi-mately 396 m (1,300 ft) northward under Lake Ontario to the submerged intake structures.

One tunnel also contains the discharge pipe, which extends approxi-mately 550 m (1,800 ft) to the discharge diffuser.

Radionuclides are emitted to the atmosphere from two locations at HMP-2.

These are the stack and the combined vent for the radwaste and reactor buildings.

Liquid radwaste is stored for decay or concentrated to a solid waste for con-trolled disposal at regulated storage sites.

Provisions are made for control of the components of nuclear safety systems from the control room.

In the event that the control room becomes inaccessible, the reactor is designed to be brought down from power range operation to cold HMP-2 SER 1-5

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shutdown conditions by use of necessary controls located in the remote shutdown room.

Essential safety actions are designed to be carried out by equipment of suffi-cient redundance and independence that no single failure can prevent the required action.

A standby power supply system is provided for the operation of emergency systems and engineered safety features during and following the shutdown of the reactor when the preferred power supply is not available.

The standby power supply system consists of three standby diesel generators.

One generator is dedicated to each of the three divisions of the safety-related electric power distribu-tion system feeding each Class lE load group.

Any two of the three standby diesel generators have sufficient capacity to start and supply all needed engi-neered safety features and emergency shutdown loads in case of a LOCA and/or loss of offsite power (LOOP).

The standby diesel generator fuel oil storage tanks are sized to hold a 7-day supply of fuel oil based on the engine running continuously at full load.

A LOCA and/or LOOP signal initiates start of the standby diesel generators and the generators pick up the loads in a programmed

.sequence.

Standby diesel generators are independent and feed separate load groups through separate, physically and electrically isolated distribution systems.

Failure of any one unit will not jeopardize the capability of the remaining standby diesel generators to start and run the required shutdown system and engineered safety feature loads.

A 125-V emergency dc power system feeds all safety-related dc protection, con-trol and instrumentation

loads, and safety-related dc motors under normal opera-tion of the plant as well as during emergency conditions.

The system is divided into three redundant divisions each consisting of its own battery, primary and backup battery chargers, switchgears/motor control centers, and distribution panels.

Each division feeds dc loads associated with corresponding divisions of the safety-related electric power distribution system.

Batteries and battery chargers are redundant and feed separate load groups through separate and iso-lated distribution systems, and failure of any one unit will not jeopardize the capability of remaining units to feed associated loads.

0 1.3 Shared Facilities and E ui ment The applicant has indicated in the FSAR that NMP-2 is a single unit and does not share structures,

systems, or components important to safety.

A number of non-safety-related facilities and equipment at NMP-2 will be shared with NMP-1.

These include (1) the Technical Support Center (2) security (3) the railroad spur (4) the instrument calibration and repair facility (5) the meteorological tower (6) the respirator repair facility (7) the whole-body counter (8) the portal monitor NNP-2 SER

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(9) laundry facilities (10) respiratory equipment repair,

assembly, and testing facilities (ll) emergency operating facility (12) fire main loops (interconnected)

The sharing of Lake Ontario as the ultimate heat sink for NMP-1, NHP-2, and FitzPatrick is discussed in Section 9.2.5 of this report.

1.4 Com arison With Similar Facilit Desi ns NHP-2 has design features that are similar to those that the staff has evalua-ted and approved previously for other nuclear power plants now under construc-.

tion or in operation.

Examples of plants with similar design features are Washington Public Power Supply System (WPPSS),

Unit 2; Zimmer, Unit 1; and LaSalle County Station, Units 1 and 2.

To the extent feasible and appropriate, the staff has made use of previous evaluations of these plants in conducting the review of NHP-2.

Where this has been

done, the appropriate sections of this report identify the other facilities involved.

The staff safety evalua-tions for these facilities have been published and are available for public inspection at the NRC Public Document Room.

1.5 Identification of A ents and Contractors Niagara Mohawk Power Corporation (NHPC) owns 41K of NMP-2 and acts as agent for the following co-owners:

Utilities Percent Central Hudson Gas and Electric Corporation Long Island Lighting Company*

New York State Electric and Gas Corporation

-Rochester Gas and Electric Corporation 9

18 18 14 NHPC has responsibility for licensing,

design, procurement, construction, opera-tion and all related operations with respect to NMP-2.

0 NMPC has retained GE (l) to design, fabricate, and deliver the direct-cycle boiling water nuclear steam supply system (NSSS),

(2) to fabricate the first core of nuclear fuel, and (3) to provide technical direction of installation and startup of this equipment.

The balance of the plant is designed and con" structed by Stone and Webster Engineering Corporation (SWEC), the architect-engineer.

.The turbine generator is manufactured by GE.

The applicant utilizes consultants in specialized areas.

For example, Dames and Moore was used in consultation for hydrology, geology, and seismology; "NMPC is currently paying for Long Island Lighting Company's current financial commitment for the cost of NHP-2.

NMP-2 SER 1-7

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Heteorological Evaluation Services, Inc.

(HES) assisted in the area of meteor-ology; and Lawler, Hatusky, and Skelly Consulting Engineers (LHS) assisted in the areas of aquatic biology, physical hydrology, lake circulation, temperature, and discharge evaluation.

1.6 Summar of Princi al Review Mattters The staff technical review and evaluation of the information submitted by the applicant considered, or will consider, the principal matters summarized below.

The population density and land-use characteristics of the site environs and the physical characteristics of the site (including seismology, meteor-ology, geology, and hydrology) to establish that (a) these characteristics have been determined adequately and have been given appropriate considera-tion in the plant design, and (b) the site characteristics are in accord-ance with the Commission siting criteria in 10 CFR 100, taking into consi-deration the design of the facility, including the engineered safety features provided.

'2)

The design, fabrication, construction, and testing criteria, and the expec-ted performance characteristics of the plant structures,

systems, and com-ponents important to safety to determine that (a) they are in accord with the GOC, quality assurance criteria, regulatory guides, and other appro-priate rules,

codes, and standards, and (b) any departures from these criteria, codes, and standards have been identified and justified.

(5)

The expected response of the facility to various anticipated operating transients and to a broad spectrum of postulated accidents.

On the basis of this evaluation, the staff determined that the potential calculated consequences of a few highly unlikely postulated accidents (design-basis accidents) would exceed those of all other accidents considered.

The staff perfor'med conservative analyses of these design-basis accidents to deter-mine that the calculated potential offsite radiation doses that might re-sult--in the very unlikely event of their occurrence would not exceed the Commission guidelines forqsite acceptability given in 10 CFR 100.

The applicant'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 plans for emergency actions to be taken in the unlikely event of an accident that might affect the gen-eral public to determine that the applicant is technically qualified to operate the facility safely.

The design of the systems provided for control of radiological effluents from the facility to determine (a) that these systems are capable of con-trolling the release of radioactive wastes from the facility within the limits of the Commission regulations in 10 CFR 20 and (b) that the appli-cant is capable of operating the equipment provided so that radioactive releases are reduced to levels that are as low as is reasonably achievable (ALARA) within the context of the Commission regulations in 10 CFR 50 and to meet the dose-design objectives of Appendix I to 10 CFR 50.

HHP"2 SER 1-8

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The applicant's quality assurance program for the operation of the facili-ties to ensure (a) that the program complies with the Commission regula-tions in 10 CFR 50 and (b) that the applicant will have proper controls over the facility operations so that there is reasonable assurance that the facility can be operated safely and reliably.

1.7 Modifications to the Facilit Durin the Course of the Staff Review During the review, the staff met (see Appendix A to this report) with the appli-cant's representatives, contr actors, and consultants to discuss various techni-cal matters related to the facility.

Also, the staff made a number of visits to the site to assess specific safety matters related to the station.

The applicant has made changes to the facility design or to the FSAR as a result of the staff review.

The staff has included these design changes in its review.

Details concerning these changes are included in FSAR amendments and in appro-priate subsections of this report.

1.8 Outstandin Issues The staff has identified certain outstanding issues in its review that have not been resolved with the applicant at the time this report was issued.

The staff will complete its review of these items before the operating license is issued.

The outstanding items are identified in Table 1.2.

The staff will discuss the resolution of these items in a supplement to this report.

1.9 Confirmator Items At this stage in the review, there are some items that have essentially been resolved to the staff's satisfaction, but for which certain confirmatory infor-mation has not been provided by the applicant (Table 1.3).

The applicant will be asked to provide the confirmatory information in the near future. If the staff review of the information provided for an item does not confirm prelimin-ary conclusions, that item will be treated as open.

The staff will discuss the resolution of the confirmatory items in a supplement to this report.

1. 10 License Condition Items There are certain issues for which a license condition may be desirable to ensure that staff requirements are met by a specified date (Table 1.4).

These condi-tions will be in the form of a condition in the body of the operating license.

l. 11 Unresolved Safet 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 HMP-2 SER 1"9

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shall be submitted to the Congress on or before January 1, 1978, and progress reports shall be included in the annual report to the Con-gress thereafter.

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

Congress, NUREG-0410, in January 1978, which describes the generic issues pro-gram of the Office of Nuclear Reactor Regulation (NRR) that had been implemen-ted early in 1977.

The NRR program described in NUREG-0410 provides for the identification of generic issues, the assignment of priorities, the develop-ment 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 continually 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 staff 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 staff require-ments should be modified.

These issues being studied are sometimes called Ge-neric 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 re-solving these generic issues is provided in Appendix C to this report.

NMP-2 SER 1-10

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Table 1.1 Cross-reference table for THI-2 Action Plan items TMI Item Shortened title SER Section I.A.1.1 I.A.1.2 I.A.1.3 I.A.2.1 I.A.2.3 I.A.3.1 I.B.1.2 I. C.1 I.C.2 I. C.3 I. C.4 I. C.5 I.C.6 I.C.7 Shift Technical Advisor Shift Supervisor responsibilities Shift staffing Immediate upgrade of RO and SRO training and qualification Administration of training program Revised scope and criteria for licensing exams Independent Safety Engineering Group Short-term accident/procedure review Shi ft/relief turnover procedures Shift Supervisor responsibilities Control room access Feedback of operating experience Verification of correct performance of operator activities NSSS vendor review. of procedures

13. l. 2

13. 5. 1

13. 1. 2 13.2.1.3

13. 2. 1. 3

13. 2

13. 4

13. 5. 1

13. 5. 1

13. 5.1

13. 5. 1

13. 5. 1

13. 5. 1

13. 5. 2. 3 I. C.8 Pilot monitoring of selected engineering procedures for NTOLs 13.5.2.3 I.D.1 I.D.2 I.G.1 II.B. 1 II.B.2 II. B. 3 II.B.4 II.D. 1 Control room design review Safety parameter display system Training during low-power testing Reactor coolant system vents Plant shielding Postaccident sampling Training for mitigating core damage Relief and safety valve position indication 18 18 18

15. 9. 1 12.3. 2 9.3.2

13. 2. l. 3

3. 10. 2 NHP-2 SER 1-11

I

Table 1.1 (Continued)

TMI Item II.D. 3 II.E.4.1 II.E.4.2 II.F.1.1 II.F.1.2 II.F.1.3 II.F.1.4 II.F.1. 5 II.F.1.6 II.F.2 II.K.1.5 II. K. 1. 10 II.K. 1. 22 II. K. 1. 23 II.K.3. 3 II. K. 3. 13 II.K.3.14 II. K. 3. 15 II. K. 3. 16 II. K. 3. 17 II. K. 3. 18 II. K. 3. 21 II.K. 3. 22 Shortened title Direct indication of relief/safety valve position Dedicated hydrogen penetrations Containment isolation dependability Noble gas monitor Iodine/particulate sampling Containment high-range monitor Containment pressure Containment water level Containment hydrogen Instrumentation for detection of inadequate core cooling Review ESF valves Operability status Auxiliary heat removal system procedures RV level procedures Reporting SRV failures Separation of HPCI/RCIC systems Isolation ot isolation condensers on high radiation Preclude spurious isolation of HPCI and RCIC Reduction of challenges and failures of SRVs Report of ECCS outage Modify ADS logic Restart of core spray/LPCI Automatic switch of RCIC system suction SER Section 7.5.2.2 6.2.5 6.2.4.3

11. 5

11. 5 12.3.4 7.5.2.3 7.5,2.3 7.5.2.3 4.4.7

15. 9

15. 9. 2

15. 9. 2

15. 9. 2

15. 9. 3

15. 9. 3 15.9. 3

15. 9. 3

15. 9. 3

15. 9. 3

15. 9. 3 7.3.2 7.4.2 NMP-2 SER 1-12

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Table 1.1 (Continued)

THI Item II.K.3.24 II.K.3. 25 II.K.3. 27 II.K.3. 28 II. K. 3. 30 II. K. 3. 31 II.K. 3.44 II.K.3.45 II. K. 3. 46 III.A. 1. 2 III.A.2 III. D. 1. 1 III. D. 3. 3 III. D. 3. 4 Shortened title Space cooling HPCI/RCIC Loss of power to pump seal coolers Common reference level gualification of ADS accumulators Revised small-break LOCA Plant-specific calculations per 10 CFR 50.46 Evaluate transients with single failure Hanual depressurization

Response

to Hichelson's concerns Upgrade emergency support facilities Emergency preparedness Integrity of systems outside containment Inplant radiation monitoring Control room habitability SER Section 9.4. 5

15. 9. 3

18. 1

3. 10. 2

15. 9. 3 15.9. 3

15. 9. 3

15. 9. 3

15. 9. 3

13. 3

13. 3

15. 9.4

12. 5. 2 6 ~ 4 NHP"2 SER 1-13

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Table 1.2 Outstanding Issues Issue (1)

Snow loads (2)

Break analysis of reactor water cleanup line (3)

Preservice and inservice inspection plan SER Section

2. 3.2 3.6.2 3.9.6, 5.2.4, (4)

Equipment qualification (5)

Steam bypass of the suppression pool (6)

Secondary containment bypass leakage (7)

Containment isolation (8)

Containment leak testing (9)

Containment fracture toughness (GDC 51)

(10) Postaccident monitoring instrumentation (ll) Separation criteria (12) Safe and alternate shutdown (13) Essential lighting (14) Air start system (15) Operations management 3.10, 3.11

6. 2.1.8 6.2.3.1, 15.6

6. 2.4 6.2.6 6.2.7 7.5.2.2 8.4.5 9.5.1.4 9.5.3 9.5.4, 9.5.6 13.1, 13.4, (16) Procedures generation package (17) Preoperational and startup test abstracts (18)

SPDS P'CRDR

13. 5. 2

14. 0 18.1, 18.2 Nl1P-2 SER 1-14

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Table 1.3 Confirmatory issues Issue SER Section Design of parapet scuppers on the roofs of safety-related buildings 2.4.2.2 (5)

(6)

Ver tica 1 floor flexibi lity SRV/pool dynamic loads on containment interior structure (7)

Analytical results for the reactor internals for LOCA and SSE (2)

Construction quality control tests on revetment ditch (3)

Feedwater check valves (4)

Pipe break criteria

2. 5. 6. 2. 4 3.6.2 3.6.2 3.7.2, 3.7.3 3.8.3 3.9.2.4 (8)

Results of Mark II hydrodynamic loads for NSSS piping, components, and equipment 3.9.3

~ 1 (9)

Leak rate test program (10) Confirmation of number of ADS SRVs needed to achieve a

rapid depressurization during a small-break LOCA based on a plant-specific ECCS analysis.

3.9.6 5.2.2 (11)

(12)

(43)

(14)

Lead factors Plant-specific LOCA analysis Verification of CONTEMPT LT/028 computer code Pool dynamics 5.3.1.2 6.3, 15.9.3 6.2.1.3 6.2.1.7.3 (a)

Pool swell loads (b)

Loads on submerged boundaries (c)

Hultivent, lateral load (d)

CO and chugging loads inside the pedestal (e)

Steam condensation submerged drag loads (f)

Bulk-to-local temperature differences (g)

Single-failure analysis (h) quencher air clearing load (i)

SRV submerged structure load (j)

SRV inplant test (k)

Wetwell-drywell vacuum breakers (l)

Hark III containment concerns (25) Reverse flow testing 6.2.6 NHP-2 SER 1-15

0(

C C4'P t.tP

Tabl e 1. 3 (Continued)

Issue (16) Haximum hydrogen generation from the chemical reaction of the cladding with water or steam (17) Instrument setpoints (18) Anticipated transients without scram - mitigation system SER Section 6.3.5

\\

7.2.2.3

7. 2. 2.4 (19) Hinimum number of channels required to initiate protection actions 7.2.2.6 (20} Isolation of circuits (21) Separation of Class 1E equipment and circuits (22) Testing of protection systems instrumentation (23) Hanual initiation of RCIC (24) Capability for safe shutdown following loss of electrical power to instrumentation and controls 7.2.2.8

7. 2. 2. 10 7.3.2.5 7.4.2.2 7.4.2.4 (25)

LPCI and LPCS injection valves interlocks (26) Multiple control system failures (27) High-energy-line breaks and consequential control systems failures 7.6.2.1

7. 7.2.1 7.7.2.2 (28) Adequacy of station electric distribution system voltage (29) Supporting analysis required to confirm adequacy of LFHG 0

motor circuit breaker as backup overcurrent protection for recirculation pump motor electrical penetration 8.4.1 8.4.2 (30) Site visit confirmation that the 15-ft color-marking interval for cables is sufficient to verify their correct separation (31) Verification of the implementation of the electrical separation design criteria during site visit (32) Review of analysis or design changes related to quali-fication of electrical equipment for flooding 8.4.5 8.4.5 8.4.7 (33)

P or tab 1 e radio communications demons trati on (34} Emergency lighting (35) Procedures for filling'uel oil storage tanks 9.5.2 9.5.3

9. 5.4. 1 NMP-2 SER 1"16

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I

Table 1.3 (Continued)

Issue (36) Details of 1-in. vent line (37) Division III diesel generator operation - severe conditions (38) Fuel oil storage and transfer system -

P8 IDs (39) Procedures for maintaining diesel generator jacket water temperature (40) Diesel generator interface on P8 ID (41) Procedures for minimum loading of diesel generators (42) Divisions I, I.I, and III diesel generator air-start systems (43) Division III air dryer - installation and performance monitoring (44) Fire damper control of combustion products (45) Concrete dust control (46) Solid radioactive waste process control program and a

compliance program to meet the requirements of 10 CFR 61 for land disposal of radioactive waste (47) Alert and notification of the public within 15 minutes (48)

EOF staffing (49) Basis for recommendations for protective measures (50) Compliance with ATMS rule (10 CFR 50.62)

SER Section 9.5.4.1 9.5.4.1 9.5.4.2 9.5.5 9.5.5, 9.5.6 9.5.5 9.5.6 9.5.6

9. 5.8

~

9. 5.8 ll.4. 2

13. 3. 2. 5

13. 3. 2. 8 0

13. 3. 2. 10

15. 8 (51) IE Bulletin 79-08 item 6 (II. K. 1. 5) and item 8 (II. K. l. 10)

15. 9. 2 (52) Installation of equipment for the automatic restart of RCIC on low water level (53) Modification of ADS logic (II.K.3.18)

(54) Installation of modification to RCIC pipe break detection

.circuitry (II.K. 3. 15)

(55) Integrity of systems outside containment likely to contain radioactive material (III.0. l. 1)

15. 9. 3 15.9.3

15. 9. 3

15. 9. 4 NMP-2 SER 1-17

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1

Table 1.4 License conditions issue (1)

Turbine system maintenance program (2)

Thermal hydraulic stability analysis beyond Cycle 1 (3)

Fire protection (4)

Operability of PASS system (5)

Operation with partial feedwater SER Section 3.5.1.3 4.4.4 9.5

~ 1.9

9. 3.2

15. 1 NMP-2 SER 1" 18

/