ML20093D505

Updated Final Safety Analysis Report, Chapter 3 Tables

ML20093D505
Person / Time
Site:	McGuire, Mcguire
Issue date:	04/11/2019
From:	Duke Energy Carolinas
To:	Office of Nuclear Reactor Regulation
Croon G, NRR/DORL/LPL2-1, 415-1023
Shared Package
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References
Download: ML20093D505 (121)

Text

McGuire Nuclear Station UFSAR Appendix 3A. Tables Appendix 3A. Tables

McGuire Nuclear Station UFSAR Table 3-1 (Page 1 of 4)

(13 OCT 2018)

Table 3-1. Summary of Criteria - Structures Loading Seismic Tornado Remarks Including Any Environ-mental Requirements Structure Q.A.

Reqd Category Normal Wind Dead and Equipment Live Containment Accident Pressure OBE SSE Wind Missile² Containment X

I X

X X

X X

Thermal Stresses and Partial Vacuum in Annulus. Equipment Missile Protected Containment and Reactor Building Foundation Slab X

I X

X X

X X

X X

Thermal Stresses Containment Interior Concrete X

I X

X X

X Differential Accident Pressure; Pipe Rupture Loads Thermal Stresses.

Equipment Missile Protected Containment Penetrations X

I X

X X

X X

Checked for Potential Pipe Whipping Containment Structural Steel X

I X

X X

X Thermal Stresses Nuclear Service Water Pipe X

I X

X X

X X

X X

Soil and Water Pressures on Buried Portion. Hydraulic Pressures.

Moving Equipment Loads. Static Seismic Analysis. Spec. By Mech.

Section.

Nuclear Service Water Intake/Discharge/

Overflow Structures X

I X

X X

X X

X X

Soil and Water Pressures.

Standby Nuclear Service Water Pond Dam X

I X

X X

X X

X X

Newmark's Method for Seismic Analysis of Dams; Hurricane Winds Auxiliary Building, Including Diesel Bldg Fuel Pool (Concrete Walls, Liner Plate)

X I

X X

X X

X X

X Soil and Water Pressures on Substructure; Tornado Pressure Drop; D/G Flywheel Reactor Missile Thermal Stresses and Cask Drop New Fuel Storage Vault X

I X

X X

X X

X3 Fuel Storage Racks X

I X

X X

X Thermal Stresses

McGuire Nuclear Station UFSAR Table 3-1 (Page 2 of 4)

(13 OCT 2018)

Loading Seismic Tornado Remarks Including Any Environ-mental Requirements Structure Q.A.

Reqd Category Normal Wind Dead and Equipment Live Containment Accident Pressure OBE SSE Wind Missile² Main Steam and Feedwater Supports, Through Isolation Valve and First Support Outside Reactor Building X

I X

X X

X X

X X

Pipe Loads, Pipe Rupture Loads Reactor Building X

I X

X X

X X

X X

Tornado Pressure Drop. Soil and Water Pressure on Substructure Station Vent (Note

1)

I X

X X

X X

X Refueling Water Storage Tank Foundations X

I X

X X

X X

X Refueling Water Storage Tank Pipe Trench I

X X

X X

X X

X Reactor Refueling Water Storage Tank Missile Wall X

I X

X X

X X

X X

Secondary Containment for Tank Relay House III X

X X

230 Kv Switch Station Steel and Fdts.

III X

X X

Step-Up and Auxiliary Transformer Foundations III X

X Access Railroad, Including Structures III X

X Administration Building III X

X Heating Boiler Vent III X

X Condenser Cooling Water Pipe III X

X Soil and Water Pressures. Moving Equipment Loads Condenser Cooling Water Intake Structure III X

X Soil and Water Pressures. Moving Equipment Loads

McGuire Nuclear Station UFSAR Table 3-1 (Page 3 of 4)

(13 OCT 2018)

Loading Seismic Tornado Remarks Including Any Environ-mental Requirements Structure Q.A.

Reqd Category Normal Wind Dead and Equipment Live Containment Accident Pressure OBE SSE Wind Missile² Condenser Discharge Structure III X

X Soil and Water Pressures. Moving Equipment Loads Discharge Canal, Dike and Riprap III X

X Hydrogen and Nitrogen Houses III X

X Main Steam Line

Supports, Excluding First Support Outside Reactor Building III X

X Service Building III X

X Turbine Building Equipment Supports III X

X Turbine Building Substructure III X

X X

Turbine Building Superstructure III X

X X

Soil and Water Pressures Turbine-Generator Foundation III X

X Yard Drainage III X

X Per Manufacturer's Recommendations Soll and Water Pressures, Moving Equipment Loads Hot Machine Shop, Decontamination Rooms, Shipping and Receiving Areas III X

X X

Soil and Water Pressures, Moving Equipment Loads Waste Water Collection Basin Dam (Non-(Nuclear Waste)

III X

X X

Units 1 and 2 Makeup Water Storage Tank Foundation X

I X

X X

X X

X

McGuire Nuclear Station UFSAR Table 3-1 (Page 4 of 4)

(13 OCT 2018)

Loading Seismic Tornado Remarks Including Any Environ-mental Requirements Structure Q.A.

Reqd Category Normal Wind Dead and Equipment Live Containment Accident Pressure OBE SSE Wind Missile² Cowans Ford Auxiliary Intake Structure III X

X X

525 Kv Switch Station Steel and Foundations III X

X X

Radwaste Building X

III X

X X

Meets the require-ments of Reg.

Guide 1.143 Symbols:

OBE

= Operating Basis Earthquake SSE

= Safe Shutdown Earthquake X

= Designed For

= Not Designed For Note 1:

The Station Vent is not designed for tornado misiles.

Note 2:

Turbine missiles are included per section 3.5.1.2 Note 3:

Reference Table 3-63

McGuire Nuclear Station UFSAR Table 3-2 (Page 1 of 2)

(09 OCT 2015)

Table 3-2. Summary of Criteria - Equipment Equipment Scope Quality Assurance Required Category3 Code Location Rad2 Source Seismic Tornado Remarks Including Any Environmental Requirements OBE SSE Wind Missile4 Containment Polar Crane D

X 11

ITTT, NEMA, NEC C

X X

Containment Accident Pressure, Dead and Equipment, Live Loads, Hold Down Device1 Fuel Handling Bridge Crane D

X 11

ITTT, NEMA, NEC AB X

X Dead and Equipment, Live Loads, Hold Down Device1 Symbols:

AB

= Auxiliary Building C

= Containment D

= Duke X

= Designed For

= Not Designed For OBE

= Operating Basis Earthquake SSE

= Safe Shutdown Earthquake

McGuire Nuclear Station UFSAR Table 3-2 (Page 2 of 2)

(09 OCT 2015)

Equipment Scope Quality Assurance Required Category3 Code Location Rad2 Source Seismic Tornado Remarks Including Any Environmental Requirements OBE SSE Wind Missile4 Notes:

1. Polar crane and cask crane designed for seismic loads in unloaded condition only.

2. X = Source of radiation

- = No source of radiation P = Possible source of radiation

3. Category II & III Structures are not safety related.

4. Turbine missiles are included per section 3.5.1.2

McGuire Nuclear Station UFSAR Table 3-3 (Page 1 of 1)

(14 OCT 2000)

Table 3-3. Summary of Codes and Standards for Components of Water-Cooled Nuclear Power Units By NRC Quality Group Class and ANS Safety Class Quality Group Class A ANS Safety Class 1, SC-1 Quality Group Class B ANS Safety Class 2, SC-2 Quality Group Class C ANS Safety Class 3, SC-3 Quality Group Class D ANS Non Nuclear Safety, NNS Component Codes and Standards Pressure Vessels ASME Boiler & Pressure Vessel Code,Section III, Class 1 ASME Boiler & Pressure Vessel Code,Section III, Class 2 ASME Boiler & Pressure Vessel Code,Section III, Class 3 ASME Boiler & Pressure Vessel Code,Section VIII, Division 1 0-15 psig Storage Tanks ASME Boiler & Pressure Vessel Code,Section III, Class 2 ASME Boiler & Pressure Vessel Code,Section III, Class 3 API-620 Atmospheric Storage Tanks ASME Boiler & Pressure Vessel Code,Section III, Class 2 ASME Boiler & Pressure Vessel Code,Section III, Class 3 API-650, AWWA D 100 or ANSI B96.1 Piping ASME Boiler & Pressure Vessel Code,Section III, Class 1 ASME Boiler & Pressure Vessel Code,Section III, Class 2 ASME Boiler & Pressure Vessel Code,Section III, Class 3 ANSI B31.1.0 Power Piping Pumps and Valves ASME Boiler & Pressure Vessel Code,Section III, Class 1 ASME Boiler & Pressure Vessel Code,Section III, Class 2 ASME Boiler & Pressure Vessel Code,Section III, Class 3 Valves - ANSI B31.1.0 Pumps

- Manufacturer's Standards

McGuire Nuclear Station UFSAR Table 3-4 (Page 1 of 23)

(13 OCT 2018)

Table 3-4. Summary of Criteria - Mechanical System Components Component Scope (2)

Safety Class (3)

Code (4)

QA Req'd (5)

Location (6)

Rad Source (7)

Seismic (1)

Wind (9)

Tornado Missile (22)

OBE (8)

SSE (8)

Reactor Coolant System Reactor Vessel W

1 III-1 X

C X

X X

X X

Reactor Coolant Pumps W

1 III-1 X

C X

X X

X X

Steam Generators (Tube)

B 1

III-1 X

C X

X X

X X

(Shell)

B 2

III-2 X

C X

X X

X X

Pressurizer W

1 III-1 X

C X

X X

X X

Pressurizer Relief Valves D

1 III-1 X

C X

X X

X X

Pressurizer Safety Valves W

1 III-1 X

C X

X X

X X

Pressurizer Relief Tank W

NNS VIII X

C X

X X

X X

RC Pump Motor Drain Tanks D

NNS API-620 C

P X

X RC Pump Motor Drain Tank Pump D

NNS C

P X

X Valves D

1,2,3, NNS III-1,-2,-3 B31.1.0 X

C,AB X

X X

X X

Safety Injection System Safety Injection Pumps W

2 P&V-II X

AB P

X X

X X

Accumulators W

2 III-C X

C P

X X

X X

Valves D&W 1,2,3 III-1,2,3 X

C,AB P

X X

X X

UHI Water Accumulators W

2 III-2 X

AB P

X X

X X

UHI Nitrogen Accumulator W

2 P&V-II X

AB P

X X

X X

UHI Surge Tank W

2 P&V-II AB P

X X

X X

Gas/Water Inter. Membrane W

2 P&V-II X

AB P

X X

X X

Residual Heat Removal System RHR Pumps W

2 P&V-II X

AB P

X X

X X

McGuire Nuclear Station UFSAR Table 3-4 (Page 2 of 23)

(13 OCT 2018)

Component Scope (2)

Safety Class (3)

Code (4)

QA Req'd (5)

Location (6)

Rad Source (7)

Seismic (1)

Wind (9)

Tornado Missile (22)

OBE (8)

SSE (8)

RHR Heat Exchangers (Tube)

W 2

III-C X

AB P

X X

X X

(Shell)

W 3

VIII X

AB P

X X

X X

Valves D

1,2 III-1,III-2 X

C,AB P

X X

X X

Containment Spray System CS Pumps D

2 III-2 X

AB X

X X

X X

CS Heat Exchangers (Tube)20 D

2 III--C X

AB X

X X

X X

(Shell)20 D

3 VIII X

AB P

X X

X X

CS Nozzles D

2 X

C P

X X

X X

Valves D

2 III-2 X

C P

X X

X X

Chemical & Volume Control System Pumps Charging, Reciprocating W

2 P&V-II X

AB X

X X

X X

Charging, Centrifugal W

2 P&V-II X

AB X

X X

X X

Boric Acid Transfer W

3 P&V-III X

AB X

X X

X X

Heat Exchangers Regenerative W

2 III-C X

C X

X X

X X

Letdown (Tube)

W 2

III-C X

AB X

X X

X X

(Shell)

W 3

VIII X

AB P

X X

X X

Excess Letdown (Tube)

W 2

III-C X

C X

X X

X X

(Shell)

W 2

III-C X

C P

X X

X X

Seal Water (Tube)

W 2

III-C X

AB X

X X

X X

(Shell)

W 3

VIII X

AB X

X X

X X

Tanks

McGuire Nuclear Station UFSAR Table 3-4 (Page 3 of 23)

(13 OCT 2018)

Component Scope (2)

Safety Class (3)

Code (4)

QA Req'd (5)

Location (6)

Rad Source (7)

Seismic (1)

Wind (9)

Tornado Missile (22)

OBE (8)

SSE (8)

Volume Control W

2 III-C X

AB X

X X

X X

Boric Acid D

3 VIII X

AB P

X X

X X

Boric Acid Batching W

NA VIII AB X

X Chemical Mixing W

NA VIII AB X

X Resin Fill D

NA VIII AB X

X Suction Stabilizer on Recip.

Charging Pump D

2 III-2 X

AB X

X X

X X

Demineralizers Mixed Bed W

3 VIII X

AB X

X X

X X

Cation W

3 VIII X

AB X

X X

X X

Filters Reactor Coolant W

2 III-C X

AB X

X X

X X

Seal Water Return W

2 III-C X

AB X

X X

X X

Seal Water Injection W

2 III-C X

AB X

X X

X X

Boric Acid W

3 III-C X

AB X

X X

X Miscellaneous Letdown Orifices W

2 III-2 X

C X

X X

X X

Boric Acid Blender W

3 III-3 X

AB X

X X

X Valves D

1,2,3 III-1,III-2 III-3 X

C,AB X

X X

X X

Recip. Charging Pump Accum.

D 2

III-2 X

AB X

X X

X X

Boron Recycle System Pumps Evaporator Feed W

NNS B31.1 AB X

X X

Reactor Makeup Water W

NNS AB X

X

McGuire Nuclear Station UFSAR Table 3-4 (Page 4 of 23)

(13 OCT 2018)

Component Scope (2)

Safety Class (3)

Code (4)

QA Req'd (5)

Location (6)

Rad Source (7)

Seismic (1)

Wind (9)

Tornado Missile (22)

OBE (8)

SSE (8)

Recycle Evap. Concen. Pump D

3 III-3 AB X

X X

X X

Seal Cooling Water Pump D

NNS AB P

X Tanks Holdup (11)

D 3

D100 X

AB X

X X

X X

Reactor Makeup Water Storage (11)

D NNS D100 O

X X

X Reagent W

NNS V111 AB X

X Seal Cooling Water Tank D

NNS AB P

X X

Demineralizers Evaporator Feed W

3 VIII X

AB X

X X

X X

Evaporator Condensate W

NNS VIII AB P

X X

Filters Evaporator Feed W

3 VIII X

AB X

X X

X X

Evaporator Condensate W

NNS VIII AB P

X X

Evaporator Concentrates W

NNS VIII AB P

X X

Seal Cooling Water Filter D

NNS VIII AB X

X X

X X

Miscellaneous Evaporator - Gas Stripper Pkg W

NNS B31.1 AB X

X X

Evaporator Condensate Return Unit D

NA AB Valves D

2,3 III-2,III-3 X

C,AB X

X X

X X

Seal Cooling Water HX (Tube)

D NNS VIII AB P

X X

(Shell)

D NNS VIII AB X

X Boron Thermal Regeneration System

McGuire Nuclear Station UFSAR Table 3-4 (Page 5 of 23)

(13 OCT 2018)

Component Scope (2)

Safety Class (3)

Code (4)

QA Req'd (5)

Location (6)

Rad Source (7)

Seismic (1)

Wind (9)

Tornado Missile (22)

OBE (8)

SSE (8)

Heat Exchangers Moderating (Tube)

W 3

VIII X

AB X

X X

X X

(Shell)

W 3

VIII X

AB X

X X

X X

Letdown Chiller (Tube)

W 3

VIII X

AB X

X X

X X

(Shell)

W NA VIII AB X

X X

X Letdown Reheat (Tube)

W 2

III-C X

AB X

X X

X X

(Shell)

W 3

VIII X

AB X

X X

X X

Miscellaneous Chiller Units W

NA AB X

X Valves D

2,3 III-C,III-3 X

AB X

X X

X X

Chiller Surge Tank W

NA VIII AB X

X Boron Chiller Pumps W

NA P&V-III AB X

X NR Demin.

W 3

III-3 X

AB X

X X

X X

Annulus Ventilation System Fans D

3 AMCA(12)

X AB P

X X

X X

Filters D

3

-(13)

X AB P

X X

X X

Moisture Separator D

3

-(13)

X AB P

X X

X X

Valves D

3 III-3 X

AB P

X X

X X

Ice Condenser Refrigeration System Ice Baskets W

2 As Applicable X

C P

X X

X X

Ice Bed Doors W

2 As Applicable X

C P

X X

X X

NF Refrigeration Units W

NA AB X

X

McGuire Nuclear Station UFSAR Table 3-4 (Page 6 of 23)

(13 OCT 2018)

Component Scope (2)

Safety Class (3)

Code (4)

QA Req'd (5)

Location (6)

Rad Source (7)

Seismic (1)

Wind (9)

Tornado Missile (22)

OBE (8)

SSE (8)

Ice Machine (Abandoned)

Ice Machine (New)

W D

NA NA AB O

X X

Ice Condenser Bridge Crane W

NA C

X X

Air Handling Units W

NA As Applicable C

X X

Valves D

2 III-2 X

C,AB P

X X

X X

NF Glycol Pumps W

NNS AB X

X NF Glycol Mixing & Storage Pump D

NNS AB X

X Ice Solution Pumps D

NNS O

X X

NF Floor Cooling Pumps W

NNS C

X X

X X

X NF Glycol Mixing & Storage Tank D

NNS VIII AB X

X Ice Solution Mixing Tanks D

NNS O

X X

NF Glycol Expansion Tank W

NNS C

X X

Ice Bin & Annex Cond. Units W

NA AB X

X Ice Bin W

NA AB X

X Ice Annex W

NA AB X

X Ice Bin & Annex Air Handlers W

NA AB X

X Ice Blower Package W

NA AB X

X Ice Cond. Cyclone Receiver W

NA C

X X

Rotary Valve Assembly W

NA AB X

X NF Floor Cooling Defrost Heater W

NA C

X X

NF Slab Cooling W

NA B31.1 C

X X

NF Glycol Bypass Strainer W

NA B31.1 AB X

X Lower Support Coolers W

NA C

X X

Ice Machine Cooling Towers D

NA O

X X

McGuire Nuclear Station UFSAR Table 3-4 (Page 7 of 23)

(13 OCT 2018)

Component Scope (2)

Safety Class (3)

Code (4)

QA Req'd (5)

Location (6)

Rad Source (7)

Seismic (1)

Wind (9)

Tornado Missile (22)

OBE (8)

SSE (8)

Ice Machine Cooling Tower Pumps D

NA O

X X

Containment Isolation System Valves D

2 III-2 X

C,AB X,P X

X X

X Containment Air Return &

Hydrogen-Skimmer System Air Return Fans D

2 AMCA(13)

X C

X X

X X

Hydrogen Skimmer Fans D

2 AMCA(13)

X C

X X

X X

Valves D

2 III-2 X

C X

X X

X Component Cooling System Pumps Cooling Water B

3 III-3 X

AB P

X X

X X

Drain Tank D

NA AB P

X X

Heat Exchangers (Tube)

D 3

VIII X

AB X

X X

X (Shell)

D 3

VIII X

AB P

X X

X X

Tanks Surge D

3 III-3 X

AB P

X X

X X

Drain D

NA AB P

X X

Valves D

2,3,NNS III-2 X

AB P

X X

X X

III-3 X

C P

X X

X X

NA AB P

X X

X X

Liquid Waste Disposal System Waste Collection Section

1.

Tanks Waste Evap. Feed D

NNS III-3 AB X

X X

McGuire Nuclear Station UFSAR Table 3-4 (Page 8 of 23)

(13 OCT 2018)

Component Scope (2)

Safety Class (3)

Code (4)

QA Req'd (5)

Location (6)

Rad Source (7)

Seismic (1)

Wind (9)

Tornado Missile (22)

OBE (8)

SSE (8)

Waste Drain D

NNS B31.1 AB P

X X

Laundry & Hot Shower D

NNS VIII AB P

X X

Reactor Coolant Drain W

NNS VIII C

X X

X Floor Drain D

NNS VIII AB X

X X

Aux. Floor Drain Tank D

NNS D100 RF X

X X

X X

Aux. Waste Evap. Feed Tank D

NNS D100 RF X

X X

X X

2.

Pumps Waste Evap. Feed Tank D

NNS VIII AB X

X X

Waste Drain Tank W&D NNS B31.1 AB X

X X

Laundry & Hot Shower Tank W

NNS P&V-III AB P

X X

Floor Drain Tank W

NNS P&V-III AB P

X X

Reactor Coolant Drain Tank W

NNS P&V-III C

X X

X RHR & CS Pump Room Sump D

NNS AB P

X X

Groundwater Drainage Sump D

3 III-3 X

AB X

X X

X Waste Evap. Feed Tank Sump D

NNS AB P

X X

Floor Drain Tank Sump D

NNS AB P

X X

Containment Floor &

Equipment Sump D

NNS C

P X

X X

X Incore Inst. Sump D

NNS C

P X

X X

X Aux. FDT Pump D

NNS VIII RF X

X X

Aux. WEFT Pump D

NNS VIII RF X

X X

IRF Sump Pump D

NNS RF P

X X

McGuire Nuclear Station UFSAR Table 3-4 (Page 9 of 23)

(13 OCT 2018)

Component Scope (2)

Safety Class (3)

Code (4)

QA Req'd (5)

Location (6)

Rad Source (7)

Seismic (1)

Wind (9)

Tornado Missile (22)

OBE (8)

SSE (8)

IRF Pipe Trench Sump Pumps D

NNS RF P

X X

Laundry Machine Sump Pumps D

NNS CW X

X X

Decon Equipment Sump Pumps D

NNS CW X

X X

3.

Strainers Laundry & Hot Shower Tank W

NNS VIII AB P

X X

Floor Drain Tank W

NNS VIII AB P

X X

4.

Filters Laundry & Hot Shower Tank Primary W

NNS VIII AB P

X X

Tank Secondary W

NNS VIII AB P

X X

Floor Drain Tank W

NNS VIII AB P

X X

5.

Miscellaneous Reactor Coolant Drain Tank Heat Exchanger (Tube)

W NNS VIII C

P X

X X

X (Shell)

W 3

VIII X

C P

X X

X X

Valves W

2,3 III-3 X

AB X

X X

X X

Gas Sample Vessel D

3 III-3 X

AB X

X X

X X

Waste Processing System

1.

Tanks Mixing and Settling D

NNS VIII AB P

X X

Mixing and Settling Tank Reagent Tank D

NA AB X

X

McGuire Nuclear Station UFSAR Table 3-4 (Page 10 of 23)

(13 OCT 2018)

Component Scope (2)

Safety Class (3)

Code (4)

QA Req'd (5)

Location (6)

Rad Source (7)

Seismic (1)

Wind (9)

Tornado Missile (22)

OBE (8)

SSE (8)

Evaporator Conc. Lines Flush D

NNS VIII AB X

X X

Seal Cooling Water Tank D

NA AB P

X X

2.

Pumps Mixing and Settling Tank D

NNS VIII AB P

X X

Mixing and Settling Tank Sludge D

NNS VIII AB P

X X

Waste Evapor. Concen.

Pump D

3 III-3 X

AB X

X X

X X

Seal Cooling Water Pump D

NNS AB P

X X

Seal Cooling Water Filter D

NNS VIII AB P

X X

Seal Cooling Water HX (Tube)

D NNS VIII AB P

X X

(Shell)

D NNS VIII AB P

X X

3.

Miscellaneous Waste Evap. Condensate Return D

NA AB X

X Waste Evaporator W

NNS B31.1 AB X

X X

Waste Evap. Cond. Demin.

W NNS VIII AB P

X X

Waste Evap. Cond. Filter W

NNS VIII AB P

X X

Waste Evap. Reagent Tank W

NNS VIII AB P

X X

Waste Evap. Feed Filter W&D NNS III-C,VIII AB X

X X

Laundry & Hot Shower Carbon Filter D

NNS VIII AB P

X X

Laundry & Hot Shower Demineralizer D

NNS VIII AB P

X X

Laundry & Hot Shower Demin. Filter D

NNS VIII AB P

X X

McGuire Nuclear Station UFSAR Table 3-4 (Page 11 of 23)

(13 OCT 2018)

Component Scope (2)

Safety Class (3)

Code (4)

QA Req'd (5)

Location (6)

Rad Source (7)

Seismic (1)

Wind (9)

Tornado Missile (22)

OBE (8)

SSE (8)

Valves D

3,NNS III-3,B31.1.0 X

AB X

X X

X X

Waste Monitor & Disp. Section

1.

Tanks Waste Monitor D

NNS VIII AB P

X X

Recycle Monitor Tank D

NNS VIII AB P

X X

Vent. Unit Cond. Dr. Tank D

NNS VIII AB P

X X

2.

Pumps Vent. Unit Cond. Dr. Tank D

NNS AB P

X X

Waste Monitor Tank W

NNS VIII AB P

X X

Recycle Monitor Tank W&D NNS VIII AB P

X X

3.

Miscellaneous Waste Monitor Demineralizer Filter D

NNS VIII AB P

X X

Valves D

NNS B31.1.0 AB X

X X

Gaseous Waste Disposal System Waste Gas Compressor Pkg W

NNS B31.1.0 AB X

X X

Waste Gas Tanks D

3 VIII X

AB X

X X

X X

Hydrogen Recombiners W

3 AB X

X X

X X

Auto Gas Analyzer D

NNS AB X

X X

X X

Gas Decay Tank Drain Pump W

NNS AB P

X X

Gas Trap W

NNS AB P

X X

Valves D

2,3 III-2,III-3 X

C,AB X

X X

X X

Solid Waste Disposal System

1.

Tanks

McGuire Nuclear Station UFSAR Table 3-4 (Page 12 of 23)

(13 OCT 2018)

Component Scope (2)

Safety Class (3)

Code (4)

QA Req'd (5)

Location (6)

Rad Source (7)

Seismic (1)

Wind (9)

Tornado Missile (22)

OBE (8)

SSE (8)

Spent Resin Storage Tank D

3 III-3 X

AB X

X X

X X

Chemical Drain Tank D

NNS VIII AB X

X X

Evaporator Concentrates Storage Tank D

NNS VIII AB X

X X

Resin Batching Tank D

NNS VIII AB X

X X

Evap. Concen. Batch Tank D

NNS VIII AB X

X X

Binder Storage Tank D

NA B

2.

Pumps Spent Resin Sluicing W

NNS AB X

X X

Chemical Drain Tank W

AB X

X X

Resin Dewatering Pump D

NNS AB X

X X

Binder Transfer Pump D

NA O

3.

Miscellaneous Hydraulic Compactor D

NNS AB P

X X

Spent Resin Sluicing Filter W

NNS VIII AB X

X X

X Valves D

3,NNS III-3,B31.1 X

AB X

X X

X X

Radwaste Pump Skid Unit D

NNS B31.1 AB X

X X

Fuel Pool Cooling & Cleanup System Cooling Pumps B

3 III-3 X

AB X

X X

X X

Cooling Heat Exchanger D

3 VIII X

AB X

X X

X X

Cooling Strainers D

3 VIII X

AB X

X X

X X

Demineralizer D

NNS VIII AB X

X X

Filter D

NNS VIII AB X

X X

Skimmer Pump D

NNS AB X

X X

McGuire Nuclear Station UFSAR Table 3-4 (Page 13 of 23)

(13 OCT 2018)

Component Scope (2)

Safety Class (3)

Code (4)

QA Req'd (5)

Location (6)

Rad Source (7)

Seismic (1)

Wind (9)

Tornado Missile (22)

OBE (8)

SSE (8)

Skimmer Filter D

NNS AB X

X X

Valves D

3,NNS III-3,B31.1.0 X,-

AB X

X X

X X

Diesel Generator Fuel Oil System Day Tanks D

3 (18)

X AB (18)

(18)

X X

Booster Pumps D

3 (18)

X AB (18)

(18)

X X

Drip Tanks, Pumps, & Filters D

NNS AB X

X Recirculation Pump D

NNS O

Recirculation Filter D

NNS O

Transfer Filters D

3 III-3 X

AB X

X X

X Transfer Pumps D

3 (18)

X AB (18)

(18)

X X

Fuel Oil Tanks (17)

D 3

VII X

B X

X X

X Relief Valves D

3,NNS (18)

X, -

AB, o (18), -

(18), -

X, -

X, -

Valves (In-line & Others)

D 3,NNS III-3,(18), -

X, -

DB,B X

X X

X Diesel Generator Cooling Water System Cooling Water Heat Exchangers D

3 VIII X

AB X

X X

X Intercooler Pumps D

3 (18)

X AB (18)

(18)

X X

Jacket Water Circulating Pumps D

3 (18)

X AB (18)

(18)

X X

Jacket Water Heaters D

3 (18)

X AB (18)

(18)

X X

Surge Tanks D

3 (18)

X AB (18)

(18)

X X

Temperature Regulating Valves D

3 (18)

X AB (18)

(18)

X X

Valves (In-line & Others)

D 3

III-3,(18)

X AB (18)

(18)

X X

Diesel Generator Lube Oil System

McGuire Nuclear Station UFSAR Table 3-4 (Page 14 of 23)

(13 OCT 2018)

Component Scope (2)

Safety Class (3)

Code (4)

QA Req'd (5)

Location (6)

Rad Source (7)

Seismic (1)

Wind (9)

Tornado Missile (22)

OBE (8)

SSE (8)

Dirty Lube Oil Tanks D

NNS TB Heater Pumps D

3 (18)

X AB (18)

(18)

X X

Oil Heaters D

3 (18)

X AB (18)

(18)

X X

Full Flow Filters D

3 VIII,(18)

X AB (18)

(18)

X X

Lube Oil Coolers D

3 (18)

X AB (18)

(18)

X X

B & A Lube Oil Pumps D

3 (18)

X AB (18)

(18)

X X

Intake Strainers D

3 VIII X

AB X

X X

X Temperature Regulating Valves D

3 (18)

X AB (18)

(18)

X X

Relief Valves D

3 (18)

X AB (18)

(18)

X X

Valves (In-line & Others)

D 3,NNS III-3,(18),-

X,-

AB,B X

X X

X Diesel Generator Starting Air System Air Compressors D

NNS X

AB (18)

(18)

X X

Starting Air Tanks D

3 VIII,(18)

X AB (18)

(18)

X X

Line Filters D

3 III-3 X

AB X

X X

X Control Air Filters D

3 (18)

X AB (18)

(18)

X X

Filter - Moisture Traps D

3 (18)

X AB (18)

(18)

X X

Relief Valves D

3 NNS III-3,(18)

X AB X,(18)

X,(18)

X X

Valves (In-line & Others)

D 3

III-3 X

AB X

X X

X Diesel Generator Intake and Exhaust System Intake Silencers D

3 (18)

X AB (18)

(18)

X X

Exhaust Silencers D

3 (18)

X AB (18)

(18)

X X

Valves D

3,NNS III-3,-

X AB X

X X

X

McGuire Nuclear Station UFSAR Table 3-4 (Page 15 of 23)

(13 OCT 2018)

Component Scope (2)

Safety Class (3)

Code (4)

QA Req'd (5)

Location (6)

Rad Source (7)

Seismic (1)

Wind (9)

Tornado Missile (22)

OBE (8)

SSE (8)

Diesel Generator Crankcase Vacuum System Vacuum Blowers D

NNS X

AB (18)

(18)

X X

Diesel Generator Room Sump Pump System Generator Pit Sump Pumps D

NNS AB X

X Generator Room High Capacity Sump Pumps D

3 (15)

X AB X

X X

X Generator Room Low Capacity Sump Pumps D

NNS AB X

X Valves D

3,NNS III-3,-

X,-

AB,AB X

X X

X Nuclear Service Water System Pumps D

3 III-3 X

AB X

X X

X Strainers D

3 VIII X

AB X

X X

X Valves D

2,3 III-2,III-3 X

AB,O X

X X

X Conventional Service Water System Pumps D

NA SB Strainers D

NA SB Valves D

NA SB Fuel Handling System Reactor Manipulator Crane W

NNS X

C X

X X

Fuel Transfer Tube W

2 III-2 X

C,AB X

X X

X Underwater Fuel Conveyor W

NNS X

C,AB X

X Spent Fuel Manipulator Crane W

NNS X

AB X

X X

Fuel Handling Tools W

NNS X

C,AB X

X

McGuire Nuclear Station UFSAR Table 3-4 (Page 16 of 23)

(13 OCT 2018)

Component Scope (2)

Safety Class (3)

Code (4)

QA Req'd (5)

Location (6)

Rad Source (7)

Seismic (1)

Wind (9)

Tornado Missile (22)

OBE (8)

SSE (8)

Sampling System Sample Heat Exchanger D

NNS VIII AB X

X X

Sample Vessel D

NNS VIII AB X

X X

Valves D

2,3 III-2,III-3 X

C,AB X

X X

X X

Refueling Water System Refueling Water Pump D

NNS AB P

X X

Refueling Water Recirc. Pumps D

NNS AB P

X X

Refueling Water Pump Strainer D

NNS AB P

X X

Storage Tank D

2 D100 X

O P

X X

X 19 FW Pipe Trench Sump Pump D

NA O

Valves D

2 III-2, B31.1.0 X

C,AB P

X X

X X

Equipment Decontamination System Pump D

NNS AB X

X Tank D

NNS D100 AB X

X Valves D

NNS (18)

X C,AB X

X X

X Fire Protection System Fire Pumps D

NA (14)

O Valves D

NA TB,SB,O B,AB Station Ventilation Systems Containment Ventilation

1.

Containment Purge Fans D

NNS AB P

X X

McGuire Nuclear Station UFSAR Table 3-4 (Page 17 of 23)

(13 OCT 2018)

Component Scope (2)

Safety Class (3)

Code (4)

QA Req'd (5)

Location (6)

Rad Source (7)

Seismic (1)

Wind (9)

Tornado Missile (22)

OBE (8)

SSE (8)

Filters D

NNS AB P

X X

2.

Other Systems(16)

Fan/Coil Units D

NNS C

P X

X Filters D

NNS C

P X

X CRDM Fans D

NNS C

P X

X

3.

Auxiliary Bldg. Ventilation Fan/Coil Units D

NA AB X

X Filters (Charcoal)

D 3

X AB P

X X

X X

ES Air Handling Units D

3 X

AB P

X X

X X

4.

Control Bldg. Ventilation Fan D

3 X

AB P

X X

X X

Filters D

3 X

AB P

X X

X X

Air Conditioning Unit D

3 X

AB X

X X

X

5.

Diesel Bldg. Ventilation Fans D

3 X

AB P

X X

X X

Filters D

3 X

AB P

X X

X X

Main Steam System Isolation Valves D

2 III-2 X

DH X

X X

X Feedwater System Isolation Valves D

2 III-2 X

DH X

X X

X Auxiliary Feedwater System Motor Driven Pumps D

3 III-3 X

AB X

X X

X Steam Turbine Driven Pumps D

3 III-3 X

AB X

X X

X Valves D

2,3 III-2,III-3 X

AB,DH X

X X

X

McGuire Nuclear Station UFSAR Table 3-4 (Page 18 of 23)

(13 OCT 2018)

Component Scope (2)

Safety Class (3)

Code (4)

QA Req'd (5)

Location (6)

Rad Source (7)

Seismic (1)

Wind (9)

Tornado Missile (22)

OBE (8)

SSE (8)

Storage Tank D

D-100 O

Steam Dump Systems Turbine Bypass D

NA TB Relief Valves D

2 III-2 X

DH X

X X

X23 Safety Valves D

2 III-2 X

DH X

X X

X23 Steam Generator Blowdown Recycle System Note: Steam Generator Blowdown Recycle System is no longer in service. Flow path has been permanently isolated per NSM MG-1/2-2430.

SG Blowdown HX D

NNS VIII AB P

X X

SG Blowdown Tank D

NNS VIII AB P

X X

SG Blowdown Pumps D

NNS AB P

X X

SG Blowdown Recycle Demin.

D NNS VIII AB P

X X

Blowoff Tank D

NNS VIII TB Blowoff Tank Pumps D

NNS TB Valves D

2 III-2 X

C,AB P

X X

X X

Condenser Circulating Water System Condenser Circulating Pumps D

NA O

Valves D

NA TB,O Instrument Air System Compressors D

NA SB,O After Coolers D

NA SB Air Tanks (1A, 1B, 2A, 2B)

D NA III-3 AB X

X X

X Air Receivers (A, B, C)

D NA VIII SB Air Dryers D

2 VIII SB

McGuire Nuclear Station UFSAR Table 3-4 (Page 19 of 23)

(13 OCT 2018)

Component Scope (2)

Safety Class (3)

Code (4)

QA Req'd (5)

Location (6)

Rad Source (7)

Seismic (1)

Wind (9)

Tornado Missile (22)

OBE (8)

SSE (8)

Valves D

2 III-2 X

C,AB P

X X

X X

Main Steam Isolation Valve Air Tanks D

3 III-3 DH X

X Hydrogen System Hydrogen Vessels D

NA VIII O

Valves D

NA AB Nitrogen System Nitrogen Vessels D

NA VIII O

Valves D

NA VIII AT,TB Containment Air Release and Addition System Filters D

NNS AB P

X X

Valves D

2 III-2 X

AB,C P

X X

X X

McGuire Nuclear Station UFSAR Table 3-4 (Page 20 of 23)

(13 OCT 2018)

Component Scop e

(2)

Safety Class (3)

Code (4)

QA Req' d

(5)

Location (6)

Rad Source (7)

Seismic (1)

Wind (9)

Tornado Missile (22)

OBE (8)

SSE (8)

Notes:

1. Equipment located in the Containment and Auxiliary Building not designed for seismic loading will be checked to verify that fault of such equipment will not result in the loss of function of safety class equipment.

2.

D

=

Duke B

=

Babcock & Wilcox W

=

Westinghouse

3.

1

=

Safety Class 1 2

=

Safety Class 2 3

=

Safety Clas 3 NNS

=

Non-Nuclear Safety NA

=

Not Applicable

4.

III-1

=

ASME Boiler and Pressure Vessel Code - Section III, Class 1 III-2

=

ASME Boiler and Pressure Vessel Code - Section III, Class 2 III-3

=

ASME Boiler and Pressure Vessel Code - Section III, Class 3 VIII

=

ASME Boiler and Pressure Vessel Code - Section VIII B31.1.0 =

ANSI B31.1.0 (1967)

D100

=

American Waterworks Association, Standard for Steel Tanks, Standpipes, Reservoirs, and Elevated Tanks for Water Storage, AWWA, D100

McGuire Nuclear Station UFSAR Table 3-4 (Page 21 of 23)

(13 OCT 2018)

Component Scop e

(2)

Safety Class (3)

Code (4)

QA Req' d

(5)

Location (6)

Rad Source (7)

Seismic (1)

Wind (9)

Tornado Missile (22)

OBE (8)

SSE (8)

API-620

=

American Petroleum Institute Recommended Rules for Design and Construction of Large Welded Low Pressure Storage Tanks ACI

=

American Concrete Institute AMCA

=

Air Moving and Conditioning Association NFUL

=

National Fire Underwrites Laboratory P&V-1

=

ASME Code for Pumps and Valves for Nuclear Power, Class I P&V-11 =

ASME Code for Pumps and Valves for Nuclear Power, Class II P&V-III

=

ASME Code for Pumps and Valves for Nuclear Power, Class III III-A

=

ASME Boiler and Pressure Vessel Code,Section III, Class A III-B

=

ASME Boiler and Pressure Vessel Code,Section III, Class B III-C

=

ASME Boiler and Pressure Vessel Code,Section III, Class C See Table 3-57 for HVAC design codes

5.

Safety related quality assurance required: X = Yes; - = No

6.

C

=

Containment RB

=

Reactor Building AB

=

Auxiliary Building TB

=

Turbine Building SB

=

Service Building DB

=

Diesel Building DH

=

Dog House

McGuire Nuclear Station UFSAR Table 3-4 (Page 22 of 23)

(13 OCT 2018)

Component Scop e

(2)

Safety Class (3)

Code (4)

QA Req' d

(5)

Location (6)

Rad Source (7)

Seismic (1)

Wind (9)

Tornado Missile (22)

OBE (8)

SSE (8)

O

=

Outdoors above ground B

=

Buried in ground CW

=

Contaminated warehouse RF

=

Radwaste facility

7.

X

=

Source of radiation

=

No source of radiation P

=

Possible source of radiaiton

8.

X

=

Designed for

=

Not designed for

9.

X

=

Protected by virtue of location in a structure designed for tornado wind and missiles

=

No protection required

10.

Redundant electrical heaters are supplied

11.

Tank is provided with diaphragm membrane for oxygen exclusion

12.

AMCA Class III and performance tested in accordance with AMCA Standard Test Code for air moving devices.

13.

Performance test required.

14.

National Fire Protection Association No. 20.

15.

Pump for this service designed and manufactured in accordance with Section III of ASME Boiler and Pressure Vessel Code is either unavailable or not as suitable for this service as conventional pump not in conformance to the ASME Code. Conventional, non-code pump will be furnished.

16.

This refers to Upper and Lower Containment Ventilation Systems.

17.

Tanks are designed for all external forces due to soil and water, inlcuding buoyancy.

18.

These components were not built to applicable codes and standards, but have been qualified seismically.

McGuire Nuclear Station UFSAR Table 3-4 (Page 23 of 23)

(13 OCT 2018)

Component Scop e

(2)

Safety Class (3)

Code (4)

QA Req' d

(5)

Location (6)

Rad Source (7)

Seismic (1)

Wind (9)

Tornado Missile (22)

OBE (8)

SSE (8)

19.

The storage tank, itself, is not designed to resist a tornado generated missile; instead, a wall has been installed around the tank to elevation 773+1 to resist any tornado generated missile. This wall assures that a minimum quantity of water is available to mitigate any postulated accident caused by a tornado.

20.

CS HX was replaced at EOC-8. The configuration was changed to reverse flow patterns; the shell has containment spray, III-B, and tubes have raw lake water, III-C.

21.

Lower Support Coolers - Function abandoned prior to start up. Now air-cooled due to radiation.

22.

Turbine missiles are included per section 3.5.1.2

23.

Reference Table 3-63

McGuire Nuclear Station UFSAR Table 3-5 (Page 1 of 1)

(05 APR 2011)

Table 3-5. System Piping Classification Duke System Piping Class NRC Quality Class ANS Safety Class Code Design Criteria Designed For Seismic Loading A

A 1

Class 1, ASME Section III, 1971 Yes B

B 2

Class 2, ASME Section III, 1971 Yes C

C 3

Class 3, ASME Section III, 19711 Yes D

B 2

Class 2, ASME Section III, 1971 No E

D NNS ANSI B31.1.0 (1967)

No F

D NNS ANSI B31.1.0 (1967)

Yes G

ANSI B31.1.0 (1967)

No H

Duke Power Company Specifications No Note:

Code Applicability: Due to the numerous code references located throughout this SAR, no attempt is made to revise these references as Codes are amended, superseded or substituted. The code references specified above are the basis for design and materials. Duke establishes an effective code date for the station in accordance with 10 CFR 50.55a, reviews and may elect to comply with portions of or all the latest versions of the above Codes unless material and/or design commitments have progressed to a stage of completion such that it is not practical to make a change. When only portions of Code Addenda are utilized, the appropriate engineering review of the entire addenda assures that the overall intent of the Code is still maintained.

1. The Nuclear Service Water System meets the requirements of ANSIB31.7, Class III, which was the required code at the time of initial design, procurement and installation.

McGuire Nuclear Station UFSAR Table 3-6 (Page 1 of 1)

(05 APR 2011)

Table 3-6. System Valve Classification Duke System Piping Class NRC Quality Class ANS Safety Class Code Design Criteria Designed For Seismic Loading A

A 1

Class 1, ASME Section III, 1971 Yes B

B 2

Class 2, ASME Section III, 1971 Yes C

C 3

Class 3, ASME Section III, 1971 Yes D

B 2

Class 2, ASME Section III, 1971 No E

D NNS ANSI B31.1.0 (1967)

No F

D NNS ANSI B31.1.0 (1967)

Yes G

ANSI B31.1.0 (1967)

No H

Duke Energy Specifications No

McGuire Nuclear Station UFSAR Table 3-7 (Page 1 of 7)

(09 OCT 2015)

Table 3-7. Electrical Systems and Components Summary of Criteria Equipment Safety Class QA Req'd Location Rad.

Environ Seismi c SSE Tornado Wind Missil e

Scope Air Conditioning System (Control, Equipment, and Cable Room)

Compressor Motors 1E X

AB X

X X

D Vent Fan Motors 1E X

AB X

X X

D Chill Water Pump Motors 1E X

AB X

X X

D Control Room Makeup Fan Motor 1E X

AB X

X X

D Equipment & Cable Room Makeup Fan Motor 1E X

AB X

X X

D Annulus Vent System Fan Motors 1E X

AB X

X X

X D

Moisture Separator Electric Heaters 1E X

AB X

D Air Operated Valves (Solenoids) 1E X

AB X

X X

D Auxiliary Building Vent System ESF Air Handling Unit Motors 1E X

AB X

X X

D Cabling Equipment (Essential Auxiliaries)

Cable Support System 1E X

AB,C,RB X(Inside Cont.)

X X

X D

Cable 1E X

AB,C,RB X(Inside Cont.)

X X

D Chemical and Volume Control System Chg. Pump Motor (Centrifugal) 1E X

AB X

X X

X W

McGuire Nuclear Station UFSAR Table 3-7 (Page 2 of 7)

(09 OCT 2015)

Equipment Safety Class QA Req'd Location Rad.

Environ Seismi c SSE Tornado Wind Missil e

Scope Boric Acid Transfer Pump Motor 1E X

AB X

X X

W Selected MO Valve Motors 1E X

AB,C X

X X

X D

Containment Air Return Systems Air Return Fan Motors 1E X

C X

X X

X D

H2 Skimmer Fan Motors & MO Valve Motors 1E X

C X

X X

X D

Containment Spray System Containment Spray Pump Motors 1E X

AB X

X X

X D

Selected MO Valve Motors 1E X

AB X

X X

X D

Containment Isolation Systems Containment Isolation Valve Motors 1E X

RB,C,AB X

X X

X D

Component Cooling Water System Component Cooling Pump Motors 1E X

AB X

X X

D Selected M.O. Valve Motors 1E X

AB X

X X

D Diesel Building Ventilation System Diesel Room Supply Fan Motors 1E X

AB X

X X

D Proportioning Dampers 1E X

AB X

X X

D Emergency Disel Auxiliary Systems Emergency Diesel Generator 1E X

AB X

X X

D Diesel Crank Case VAC Pump Motor 1E X

AB X

X X

D Diesel F.O. Transfer Pump Motor 1E X

AB X

X X

D

McGuire Nuclear Station UFSAR Table 3-7 (Page 3 of 7)

(09 OCT 2015)

Equipment Safety Class QA Req'd Location Rad.

Environ Seismi c SSE Tornado Wind Missil e

Scope Diesel Gen. F.O. Booster Pump Motors 1E X

AB X

X X

D Diesel L.O. Filter Pump Motor 1E X

AB X

X X

D Diesel 600/120V Panelboard 1E X

AB X

X X

D Diesel Lube Oil Pump Motors 1E X

AB X

X X

D Diesel Jacket & Intercooler Pump Mtr 1E X

AB X

X X

D Diesel Air Compressor Motors 1E X

AB X

X X

D Diesel Generator Battery & Charger 1E X

AB X

X X

D Auxiliary Feedwater System Auxiliary Feedwater Pump Motor 1E X

AB X

X X

D Spent Fuel Cooling System Cooling Pump Motors 1E X

AB X

D Deleted Per 2008 Update Nuclear Instrumentation System Detectors 1E X

C X

X X

X W/D Amplifier Assembly 1E X

C X

X X

X D

Nuclear Instrumentation Racks

1. Source Range Equipment 1E X

AB X

X X

D

2. Intermediate Range Equipment 1E X

AB X

X X

D

3. Power Range Equip.

1E X

AB X

X X

W Penetrations 1E X

C,RB X

X X

X D

McGuire Nuclear Station UFSAR Table 3-7 (Page 4 of 7)

(09 OCT 2015)

Equipment Safety Class QA Req'd Location Rad.

Environ Seismi c SSE Tornado Wind Missil e

Scope 4 KV ES Aux Power System 4 KV Metalclad Bus 1E X

AB X

X X

D 4 KV Metalclad ES Switchgear 1E X

AB X

X X

D 4160/600V Transformers 1E X

AB X

X X

D 600V Essential Aux Power System 600V ES Aux Power system 1E X

AB X

X X

D 600V ES Motor Control Centers 1E X

AB X

X X

D 600/208V Essential Transformers 1E X

AB X

X X

D 125V DC Vital Instrumentation & Control Power System 125VDC Battery Chargers 1E X

AB X

X X

D 125VDC Batteries 1E X

AB X

X X

D 125VDC Distribution Centers 1E X

AB X

X X

D 125VDC Panelboards 1E X

AB X

X X

D 120V AC Vital Power System 125V DC/120V AC Static Inverters 1E X

AB X

X X

D 120 AC Vital Power Panelboards 1E X

AB X

X X

D Process Inst. & Control System Process I & C Racks (Protection) 1E X

AB X

X X

W Pri Sys Detectors (Protective) 1E X

C,AB X

X X

X W

Sec Sys Detectors (Protective) 1E X

AB X

X X

X W

McGuire Nuclear Station UFSAR Table 3-7 (Page 5 of 7)

(09 OCT 2015)

Equipment Safety Class QA Req'd Location Rad.

Environ Seismi c SSE Tornado Wind Missil e

Scope Aux Sys Detectors (Protection) 1E X

C,AB X

X X

X W/D Control Panels (Class 1E Circuits)

Main Control Board 1E X

AB X

X X

D RCP Volt/Freq Monitor Panel 1E X

AB X

X X

D Aux Feedwater Panels 1E X

AB X

X X

D Aux Shutdown Panels 1E X

AB X

X X

D Diesel Generator Control Panels 1E X

AB X

X X

D HVAC Control Panels 1E X

AB X

X X

D Misc Termination Cabinets 1E X

AB X

X X

D Radioactive Waste Systems RHR & CS Pump Room Sump Pump Motor 1E X

AB X

X X

X D

Residual Heat Removal System RHR Pump Motors 1E X

AB X

X X

X W

Selected M.O. Valve Motors 1E X

C,AB X

X X

X D

Rod Control System Reactor Trip Switchgear 1E X

AB X

X X

W Safety Injection System Safety Injection Pump Motors 1E X

AB X

X X

X W

Selected Solenoid Valves 1E X

AB X

X X

X D

Selected M.O. Valve Motors 1E X

AB X

X X

X D

McGuire Nuclear Station UFSAR Table 3-7 (Page 6 of 7)

(09 OCT 2015)

Equipment Safety Class QA Req'd Location Rad.

Environ Seismi c SSE Tornado Wind Missil e

Scope Service Water System (Nuclear)

Service Water Pump Motors 1E X

AB X

X X

D Motor Operated Valve Motors 1E X

AB X

X X

D Reactor Protection System & Engineered Safety Feature Actuation System Solid State Protection Sys Racks 1E X

AB X

X X

W ESF Test Cabinet 1E X

AB X

X X

W Auxiliary Safeguard Cabinets 1E X

AB X

X X

W Steam Supply System Main Steam Isolation Valve Solenoids 1E X

AB X

X X

D Note:

The instrumentation described in Sections 7.2 through 7.6 is subject to the pertinent requirements of the Quality Assurance Program.

EXPLANATION OF SYMBOLS Table Headings Symbols Scope W

= Westinghouse D

= Duke Safety Class 1E

= Electrical Safety Class 1E (Ref. 3.2.4.1) 2E

= Electrical Safety Class 2E (Ref. 3.2.4.1) 3E

= Electrical Safety Class 3E (Ref. 3.2.4.1) 4E

= Electrical Safety Class 4E (Ref. 3.2.4.1)

McGuire Nuclear Station UFSAR Table 3-7 (Page 7 of 7)

(09 OCT 2015)

Equipment Safety Class QA Req'd Location Rad.

Environ Seismi c SSE Tornado Wind Missil e

Scope

= No Safety Class Q.A. Req'd X

= Yes

= No Location C

= Containment RB

= Reactor Building AB

= Aux. Building TB

= Turbine Building SB

= Service Building O

= Outdoors Above Ground Rad Environ X

= Designed for Rad. Environment

= Not Designed For Seismic S.S.E

= Safe Shutdown Earthquake X

= Designed For

= Not Designed For Tornado X

= Protection Required (Protected by virtue of location in a structure designed for tornado wind and missiles)

= No Protection Required Equipment ESF

= Engineered Safety Features

McGuire Nuclear Station UFSAR Table 3-8 (Page 1 of 1)

(22 APR 2017)

Table 3-8. Design Basis Tornado Generated Missiles Design Basis Tornado Missiles are as defined below:

1. A 2 in. by 4 in. by 12 ft board weighing 40 pounds per cubic foot, end on at a speed of 300 mph.

2. A cross-tie, 7 in. by 9 in. by 8.5 ft weighing 50 pounds per cubic foot at a speed of 300 mph.

3. An automobile weighing 2,000 pounds with an impact area of 20 sq. ft. at 20 ft. above grade at a speed of 100 mph.

4. A steel pipe 2 in. in diameter by 7 ft. long end on at 100 mph.

Note: Design basis tornado generated missile direction is horizontal only (i.e. with no vertical velocity component).

McGuire Nuclear Station UFSAR Table 3-9 (Page 1 of 1)

(14 OCT 2000)

Table 3-9. Category 1 Structures and Missiles Protected Against Structure Types of Missiles Protected Against Reactor Building Tornado Missiles and Turbine-Generator Missiles Containment Interior Structures Internal Missiles Diesel Generator Building Tornado Missiles and Turbine-Generator Missiles Block Dividing Wall D/G Flywheel Missiles Auxiliary Building (outside walls, roof, floors and some internal walls)

Tornado Missiles, Turbine-Generator Missiles and Selected Internal Missiles Spent Fuel Pool Tornado Missiles and Selected Internal Missiles, Turbine-Generator Missiles Control Room (outside walls and roof)

Tornado Missiles and Turbine-Generator Missiles

McGuire Nuclear Station UFSAR Table 3-10 (Page 1 of 1)

(14 OCT 2000)

Table 3-10. CRDM Housing Plug Ejection Plug Weight: 11 lbs Plug O.D.: 2.75 in.

Travel, x (ft)

Velocity, v (ft/sec)

Kinetic Energy (ft/lb) 1 240 9,750 2

335 19,900 3

370 23,300 4

415 29,200 5

440 33,000

McGuire Nuclear Station UFSAR Table 3-11 (Page 1 of 1)

(14 OCT 2000)

Table 3-11. Drive Shaft Ejection Diameter = 1.75 in.

Length = 300 in.

Weight = 120 lbs.

Drive Shaft Travel Outside Housing1 (ft)

Drive Shaft Velocity (ft/sec)

Drive Shaft Kinetic Energy (ft-lb)

Missile Shield2 Steel Plate Thickness (in.)

Missile Shield2 Additional Concrete Slab Thickness (in.)

1 151 42,900 1

2 162 49,000 1

2.5 3

171 55,000 1

10 4

179 60,200 1

16 5

189 66,500 1

25 10 269 134,700 Notes:

1. Distance from top of rod travel housing to bottom of missile shield.

2. These thicknesses are indicative only, and shield design can optimize between steel and concrete thicknesses.

McGuire Nuclear Station UFSAR Table 3-12 (Page 1 of 1)

(14 OCT 2000)

Table 3-12. Drive Shaft and Mechanism Ejection Missile Weight: 1500 lbs Impact O.D.: 3.75 in.

Travel, x (ft)

Velocity (ft/sec)

Kinetic Energy (ft-lb) 1 14.3 4,600 2

20.2 9,200 3

24.8 13,800 4

28.6 18,400 5

32.0 23,000

McGuire Nuclear Station UFSAR Table 3-13 (Page 1 of 1)

(14 OCT 2000)

Table 3-13. Temperature Element Assembly Characteristics

1. For a tear around the weld between the boss and the pipe:

Characteristics without well with well Flow Discharge Area 0.11 in.2 0.60 in.2 Thrust Area 7.1 in.2 9.6 in.2 Missile Weight 11.0 lb 15.2 lb Area of Impact 3.14 in.2 3.14 in.2 Missile Weight divided by Impact Area 3.5 psi 4.84 psi Velocity 20 ft/sec 120 ft/sec

2. For a tear at the junction between the temperature element assembly and the boss for the without well element and at the junction between the boss and the well for the with well element.

Flow Discharge Area 0.11 in2 0.60 in2 Thrust Area 3.14 in2 3.14 in2 Missile Weight 4.0 lb 6.1 lb Area of Impact 3.14 in2 3.14 in2 Missile Weight divided by Impact Area 1.27 psi 1.94 psi Velocity 72 ft/sec 120 ft/sec

McGuire Nuclear Station UFSAR Table 3-14 (Page 1 of 1)

(14 OCT 2000)

Table 3-14. Pressurizer Space Valve Missiles Weight Flow Discharge Area Thrust Area Impact Area Weight to Impact Area Ratio Velocity

1. Safety Relief Valve Bonnet (3" x 6")

350 lb 2.86 in.2 80 in.2 24 in.2 14.6 psi 110 fps

2. 3" Motor Operated Isolation Valve Bonnet (plus Motor and Stem) 400 lb 5.5 in.2 113 in.2 28.3 in.2 14.1 psi 135 fps

3. 2" Air Operated Relief Valve Bonnet (plus stem) 75 lb 1.8 in.2 20.7 in.2 20 in.2 3.75 psi 115 fps

4. 3" Air Operated Spray Valve Bonnet (plus stem) 120 lb 5.5 in.2 50.3 in.2 50 in.2 2.4 psi 190 fps

5. 4" Air Operated Spray Valve 200 lb 9.3 in.2 50.3 in.2 50 in.2 4 psi 190 fps

McGuire Nuclear Station UFSAR Table 3-15 (Page 1 of 1)

(14 OCT 2000)

Table 3-15. Properties of Credible Turbine Missiles Missile Weight (lb)

Initial Velocity fps Exit Velocity fps Energy (ft-lb) 106 Missile Area ft2(1)

A1 A2 A3 Disc No. 1 3521 Missile Contained Disc No. 2 3611 665 341 6.5 4.77 2.55 3.28 Disc No. 3 2741 Missile Contained Disc No. 4 3194 629 312 4.8 2.4 1.96 3.60 Disc No. 5 3961 574 345 7.3 3.03 2.52 4.00 Notes:

1. See Figure 3-4.

2. Disc No. 5 controls the thickness of the barrier.

3. or more details on the turbine missile properties, see Section 3.5, Reference No. 4.

McGuire Nuclear Station UFSAR Table 3-16 (Page 1 of 2)

(22 APR 2017)

Table 3-16. Internal Missiles. (1)(4)

Missile Penetration Depth "t" Inches Minimum Barrier Thickness Required, D=3t Inches CRDM Housing Plug Ejection (FSAR Table 3-

10) 2.1 6.3 Drive Shaft Ejection (FSAR Table 3-11) 1.1 3.3 Drive Shaft and Mechanism Ejection (FSAR Table 3-11) 1.112 3.336 Temperature Element Assembly Missiles (FSAR Table 3-13) 0.54 1.62 Safety Relief Valve Bonnet (Table 3-14) 1.96 5.88 3" Motor Operated Isolation Valve Bonnet Plus Motor and Stem (Table 3-14) 1.81 5.43 2" Air Operated Relief Valve Bonnet Plus Stem (Table 3-14) 0.38 1.14 3" Air Operated Spray Valve Bonnet plus Stem (Table 3-14) 0.65 1.95 4" Air Operated Spray Valve (Table 3-14) 1.04 3.12 Turbine Missiles (2) (4 )

Disc No. 5 of the Turbine 8.29 24.9 Diesel Generator Missiles(5 )

A 10# Section of the D/G Flywheel 1.7 5.1 Design Basis Tornado Generated Missiles (3) (4) (6)

A 2" x 4" wooden board @300 mph 3.50 10.50 A Cross Tie 6" x 9" x 8.5' @ 300 mph 3.13 9.39 An Automobile 2000 lbs at speed of 100 mph 0.109 0.327 A steel pipe 2" diameter, 7' long @ 100 mph 2.03 6.09

McGuire Nuclear Station UFSAR Table 3-16 (Page 2 of 2)

(22 APR 2017)

Missile Penetration Depth "t" Inches Minimum Barrier Thickness Required, D=3t Inches Notes:

1. Barriers for internal missiles include reinforced concrete walls, slabs and other components of the interior structure subjected to the paths of these missiles. Whenever a steel barrier is encountered by the missile, the required thickness of this barrier becomes 1/12 of the thickness required for a reinforced concrete barrier (see Reference 2).

2. Barriers are 1) Reactor Building dome and 2) Auxiliary Building roof.

3. Barriers are reinforced concrete slabs and walls, e.g.,: a) Reactor Building dome, b)

Reactor Building shell, and c) Auxiliary Building roof and appropriate side walls.

4. The penetration depths of the reinforced concrete missile barriers are based on a concrete strength of 5000 psi.

5. Barrier is the reinforced block wall separating Diesel Generator Rooms A from B.

6. Design basis tornado generated missile direction is horizontal only (i.e. with no vertical velocity component).

McGuire Nuclear Station UFSAR Table 3-17 (Page 1 of 1)

(09 OCT 2015)

Table 3-17. Minimum Barrier Thicknesses for all Category 1 Structures For Which Missile Protection Is Provided Structure Minimum Reinforced Concrete Missile Barriers Thickness Inches Reactor Building Cylindrical Shell 36 Reactor Building Dome 27 Reactor Building Interior Structure

1. Walls 24

2. Slabs 24 Auxiliary Building

1. Main Building Roof 28(1)

Walls 12

2. Diesel Generator Building Roof 28 Walls 36 Dividing Wall 12(2)

3. Spent Fuel Storage Area Roof 28 Walls 24

4. New Fuel Storage Vault Roof 30 Walls 24

5. RN Manway Missile Barrier 12 Notes:

1. The thickness provided is considered as one independent thickness or the accumulation of several successive slabs.

2. Thickness is for grout filled, reinforced concrete masonry wall.

McGuire Nuclear Station UFSAR Table 3-18 (Page 1 of 1)

(14 OCT 2000)

Table 3-18. High Energy Mechanical Piping Systems Analyzed for Consequences of Postulated Piping Breaks System or Portion Thereof Operating During Normal Reactor Operation System Identification Pipe Break Protection Method High Energy Safety Related Systems Steam Generator Blowdown Recycle System BB (a) (b)

Auxiliary Feedwater System (Motor Driven Pump Portion)

CA (a) (b)

Nitrogen System GN (a)(b)

Reactor Coolant System NC (a) (b)

Residual Heat Removal System ND (a) (b)

Safety Injection System NI (a) (b)

Nuclear Sampling System NM (a) (b)

Boron Thermal Regeneration System NR (a)

Chemical and Volume Control System (Letdown Portion and Sealwater Injection)

NV (a) (b)

Other High Energy Systems Feedwater System CF (a) (b) (c)

Main Steam Supply to Auxiliary Equipment SA (a)

Main Steam System SM (a) (b) (c)

Main Steam Vent to Atmosphere System SV (a)

Pipe Whip Protection Methods Legend:

(a)

Physical Separation (b)

Piping Restraints (c)

Enclosures, structural, guard pipes, etc., (designed specifically for pipe break).

Note:

1. High Energy Systems may contain moderate energy portions; however, for brevity, high energy systems are only listed in this table.

McGuire Nuclear Station UFSAR Table 3-19 (Page 1 of 2)

(09 OCT 2015)

Table 3-19. Moderate Energy Mechanical Piping Systems Analyzed for Consequences of Postulated Piping Breaks System or Portion Thereof Operating During Normal Reactor Operation System Identificatio n

Pipe Break Protection Method Moderate Energy Safety Related Systems Auxiliary Feedwater System (Turbine Driven Portion)

CA (a)

Diesel Fuel Oil System FD (a)

Refueling Water system FW (a)

Component Cooling System KC (a)

Diesel Generator Cooling Water System KD (a)

Spent Fuel Cooling System KF (a)

Diesel Generator Lube Oil System LD (a)

Boron Recycle System NB (a)

Residual Heat Removal System ND (a)

Containment Spray System NS (a)

Nuclear Service Water System RN (a)

Containment Ventilation Cooling Water System RV (a)

Main Steam Supply to Aux. Equipment SA (a)

FWP Turbine Exhaust TE (a)

Diesel Generator Starting Air System VG (a)

Gaseous Waste Recycle System WG (a)

Liquid Waste Recycle System WL (a)

Solid Waste Disposal System WS (a)

Deleted Per 2015 Update Other Moderate Energy Systems Auxiliary Steam System AS (a)

Recirculated Cooling Water System KR (a)

Ice Condenser Refrigeration System NF (a)

Fire Protection System RF (a)

Equipment Decontamination System WE (a)

Liquid Waste Monitor & Disposal System WM (a)

Chemical Addition System YA (a)

McGuire Nuclear Station UFSAR Table 3-19 (Page 2 of 2)

(09 OCT 2015)

System or Portion Thereof Operating During Normal Reactor Operation System Identificatio n

Pipe Break Protection Method Control Area Chilled Water System YC (a)

Plant Heating System YH (a)

Make-up Demineralizer System YM (a)

Pipe Whip Protection Methods Legends:

(a)

Physical Separation (b)

Piping Restraints (c)

Enclosures, structural, guard pipes, etc., (Designed specifically for pipe break)

McGuire Nuclear Station UFSAR Table 3-20 (Page 1 of 6)

(06 OCT 2003)

Table 3-20. Comparison of Duke Pipe Rupture Criteria And NRC Requirements Of Branch Technical Positions APCSB 3-1 (November 1975), MEB 3-1 (July 1981), and NRC Regulatory Guide 1.46 (May 1973)

NRC Criteria Duke Criteria APCSB 3-1, Section B.2.c SAR Section 3.6.2.2 Section B.2.c. requires that piping between containment isolation valves be provided with pipe whip restraints capable of resisting bending and torsional moments produced by a postulated failure either upstream or downstream of the valves. Also, the restraints should be designed to withstand the loadings from postulated failures so that neither isolation valve operability nor the leaktight integrity of the containment will be impaired.

Terminal ends should be considered to originate at a point adjacent to the required pipe whip restraints.

Duke criteria is roughly equivalent to NRC criteria as clarified below:

The containment structural integrity is provided for all postulated pipe ruptures. In addition, for any postulated rupture classified as a loss of coolant accident, the design leaktightness of the containment fission product barrier will be maintained.

Penetration design is discussed in SAR Section 3.9.2.8. This section also discussed penetration guard pipe design criteria.

Terminal ends are defined as piping originating at structure or component that act as rigid constraint to the piping thermal expansion.

APCSB 3-1, Section B.2.d

1. The protective measures, structures, and guard pipes should not prevent the access required to conduct inservice inspection examination.

2. For portions of piping between containment isolation valves, the extent of inservice examinations completed during each inspection interval should provide 100 percent volumetric examination of circumferential and longitudinal pipe welds.

3. Inspection ports should be provided in guard pipes to permit the required examination of circumferential welds. Inspection ports should not be located in that portion of guard pipe passing through the annulus.

4. The areas subject to examination should be defined in accordance with Examination Categories C-F and C-G for Class 2 piping welds in Tables IWC-2520.

SAR Section 5.2.8 Duke criteria is different than the NRC criteria due to the code effective date as described below:

ASME Class 2 piping welds will be inspected in accordance with Tables ISC-251 of Section XI (1971), through Winter 1971 Addenda, of the ASME Code, as accessibility permits. Inservice inspection program requirements are given in SAR Section 5.2.8.

McGuire Nuclear Station UFSAR Table 3-20 (Page 2 of 6)

(06 OCT 2003)

NRC Criteria Duke Criteria APCSB 3-1, Appendix A High Energy fluid systems are defined as those systems that, during normal plant conditions, are either in operation or maintained pressurized under conditions where either or both of the following are met:

1. maximum operating temperature exceeds 200°F, or

2. maximum operating pressure exceeds 275 psig.

SAR Section 3.6.1.2 Duke criteria is the same as NRC criteria with expansion of definition as clarified below:

1. Non-liquid systems with a maximum normal pressure less than 275 psig are not considered high energy regardless of the temperature. Such low pressure system (i.e., Auxiliary Steam, 50 psig, 340°F) do not contain sufficient sensible energy to develop sudden, catastrophic failures. Propagation of a crack to a full failure is extremely unlikely.

2. Exception to the 200°F threshold for high energy systems is taken for non-water systems such as ethylene glycol. Such systems that operate at less than their boiling temperature are considered moderate energy.

APCSB 3-1, Appendix A In piping runs which are maintained pressurized during normal plant conditions for only a portion of the run (i.e., up to the first normally shut valve) a terminal end of such runs is the piping connection to this closed valve.

SAR Section 3.6.2.2.1 Duke criteria is different from NRC criteria as described and justified below:

Terminal ends are considered at piping originating at structure or components that act as rigid constraint to the piping thermal expansion. Typically, the anchors assumed for the code stress analysis would be terminal ends. Stresses in the system either side of the closed valve will be about the same; therefore, terminal end classification based on constraint and high stresses are not applicable.

McGuire Nuclear Station UFSAR Table 3-20 (Page 3 of 6)

(06 OCT 2003)

NRC Criteria Duke Criteria MEB 3-1, Section B.1.b(6)

Section B.1.b(6) requires that guard pipe assemblies between containment isolation valves meet the following requirements:

1. The design pressure and temperature should not be less than the maximum operating temperature and pressure of the enclosed pipe under normal plant conditions.

2. The design stress limits of Paragraph NE-3131(c) should not be exceeded under the loading associated with design pressure and temperature in combination with the safe shutdown earthquakes.

3. Guard pipe assemblies should be subjected to a single pressure test at a pressure equal to design pressure.

SAR Section 3.9.2.8 Duke criteria is different from NRC criteria as described and justified below:

Guard pipes provided between containment isolation valves is designed in accordance with SAR Section 3.9.2.8. Guardpipe thicknesses were developed using the criteria of ASME Code Case 1606 and the appropriate loading combination and stress limits of Table 3-48. Guard pipes are subjected to a pressure test as required by the material specification before welding to the penetration assembly.

It is impractical to test guard pipes in the finished penetration assembly due to the configuration and potential damage to internal process pipe and associated insulation. Independent design analysis have been conducted to provide assurance that Duke penetration designs are acceptable. In addition, the extent of NDT conducted on guard pipes to flued head butt weld is such to assure integrity of design.

McGuire Nuclear Station UFSAR Table 3-20 (Page 4 of 6)

(06 OCT 2003)

NRC Criteria Duke Criteria MEB 3-1, Sections B.1.c(3)

Breaks in non-nuclear piping should be postulated at the following location:

1. Terminal ends,

2. At each intermediate pipe fitting, welded attachment, and valve.

Note: PER GENERIC LETTER 87-11, ARBITRARY INTERMEDIATE BREAKS ARE NOT REQUIRED TO BE POSTULATED.

SAR Section 3.6.2.2.1 Duke criteria is roughly equivalent to NRC criteria as described and justified below:

Breaks in Duke Class F piping (non-nuclear, seismic) are postulated at terminal ends and at intermediate locations based on the use of ASME Section III analysis techniques, the same as Duke Class B and C piping. Duke Class F piping is constructed in accordance with ANSI B31.1 and is dynamically analyzed and restrained for seismic loadings similar to ASME Section III piping. Materials are specified, procured, received, stored, and issued under Duke's QA program similar to ASME Section III materials except that certificate of compliance in lieu of mill test reports are acceptable on minor components, and construction documentation for erected materials is not uniquely maintained. Construction documentation for erected materials is generically maintained. MTR are required for the bulk of piping materials.

MEB 3-1, Section B.2.e Thru-wall cracks may be postulated instead of breaks in those fluid systems that qualify as high energy fluid systems for short operational periods. This operational period is defined as about 2 percent of the time that the system operates as a moderate energy fluid system.

SAR Section 3.6.1.2 Duke criteria is roughly equivalent to NRC criteria as clarified below:

The operational period that classifies such systems as moderate energy in either:

1. One percent of the normal operating lifespan of the plant, or

2. Two percent of the time period required to accomplish its system design function.

McGuire Nuclear Station UFSAR Table 3-20 (Page 5 of 6)

(06 OCT 2003)

NRC Criteria Duke Criteria Regulatory Guide 1.46 Longitudinal breaks are postulated in piping runs 4 inches nominal pipe size and larger. Circumferential breaks are postulated in piping runs exceeding 1 inch nominal pipe size.

SAR Section 3.6.2.2.1 Duke criteria is the same as NRC Branch Technical Position APCSB 3-1 and roughly equivalent to Regulatory Guide 1.46 with expansion of definition as described below:

Longitudinal breaks are postulated in piping runs 4 inches nominal pipe size and larger except that longitudinal breaks are not postulated at terminal ends where the piping has no longitudinal welds.

Regulatory Guide 1.46 A whipping pipe should be considered capable of rupturing an impacted pipe of smaller nominal pipe size and lighter wall thickness.

SAR Section 3.6.2.2 Item 10 Duke criteria is the same as NRC Branch Technical Position APCSB 3-1 and roughly equivalent to Regulatory Guide 1.46 with expansion of definition as described below:

The energy associated with a whipping pipe is considered capable of (a) rupturing impacted pipes of smaller nominal pipe sizes, and (b) developing thru-wall cracks in larger nominal pipe sizes with thinner wall thicknesses.

Regulatory Guide 1.46 Measures for restraint against pipe whipping need not be provided for piping where:

1. the design temperature is 200°F or less, and

2. the design pressure is 275 psig or less.

SAR Section 3.6.1.2 Duke criteria is roughly equivalent to NRC Branch Technical Position APCSB 3-1 and differs from Regulatory Guide 1.46 with expansion of definition as described below:

High energy piping is reviewed for pipe whipping and is defined as those systems that during normal plant conditions are either in operation or maintained pressurized under conditions where either or both of the following are met:

1. maximum temperature exceeds 200°F, or

2. maximum pressure exceeds 275 psig, except that (1) non-liquid piping system with a maximum pressure less than or equal to 275 psig are not considered high energy regardless of the temperature, and (2) for liquid systems other than water, the atmospheric boiling temperature can be applied.

McGuire Nuclear Station UFSAR Table 3-20 (Page 6 of 6)

(06 OCT 2003)

NRC Criteria Duke Criteria Systems are classified as moderate energy if the total time that either of the above conditions are met is less than either:

1. one (1) percent of the operating lifespan of the plant, or

2. two (2) percent of the time period required to accomplish its system design function.

Note:

1. Pipe breaks in the RCS primary loop are not considered in certain aspects of plant design, as defined in Reference 3 of Section 3.6.6.

McGuire Nuclear Station UFSAR Table 3-21 (Page 1 of 1)

(06 OCT 2003)

Table 3-21. Postulated Break Locations in Reactor Coolant Loops Location of Postulated Rupture Type

1. Reactor Vessel Inlet Nozzle1 Circumferential

2. Reactor Vessel Outlet Nozzle1 Circumferential

3. Steam Generator Inlet Nozzle1 Circumferential

4. Steam Generator Outlet Nozzle1 Circumferential

5. Reactor Coolant Pump Inlet Nozzle1 Circumferential

6. Reactor Coolant Pump Outlet Nozzle1 Circumferential

7. 50° Elbow on the Intrados1 Longitudinal

8. Loop Closure Weld in Crossover Leg1 Circumferential

9. Residual Heat Removal (RHR) Line/Primary Coolant Loop Connection Circumferential (Viewed from the RHR line)

10. Accumulator (ACC) Line/Primary Coolant Loop Connection Circumferential (Viewed from ACC line)

11. Pressurizer Surge (PS) Line/Primary Coolant Loop Connection Circumferential (Viewed from the PS line)

Note:

1. Reference 1 of Section 3.6.6 defines the original basis for postulating pipe breaks in the Reactor Coolant System Primary Loop. References 3 and 4 of Section 3.6.6 provide the basis for eliminating the previously postulated reactor coolant system pipe breaks with the exception of those breaks at branch connections from certain aspects of design considerations.

McGuire Nuclear Station UFSAR Table 3-22 (Page 1 of 1)

(14 OCT 2000)

Table 3-22. Comparison of Actual Moments To Reference Fatigue Analysis Moments Node No.1 Fatigue Analysis Loadings Used In Fatigue Analysis McGuire Loadings Cumulative Usage Factor SI (psi)2 SI/Sm1 Mi (in-lbs) (OBE Moments)

Mi (in-lbs)

(Thermal Expansion Moments)

Mi (in-lbs) (OBE Moments)

Mi (in-lbs)

(Thermal Expansion Moments) 404

.632 94,772 5.26

.341 x 108

.243 x 108

.0164 x 108

.199 x 108 413

.0093 65,509 3.64

.243 x 108

.163 x 108

.0113 x 108

.122 x 108 415

.2656 75,528 4.19

.243 x 108

.203 x 108

.0200 x 108

.153 x 108 438

.0382 70,744 3.94

.266 x 108

.536 x 107

.0538 x 108

.047 x 108 459

.000 33,406 1.85

.211 x 108

.837 x 108

.1363 x 107

.073 x 108 468

.0060 61,100 3.39

.29 x 108

.729 x 107

.0849 x 108

.050 x 108 484

.0959 81,820 4.54

.29 x 108

.811 x 107

.1150 x 108

.057 x 108 Notes:

1. The loop closure weld, RHR line connection, accumulator line connection and surge line connection locations have not been included in this table since selection of these locations for postulated breaks is independent of detailed stress and fatigue analyses. Also, node numbers are defined in WCAP-8172.

2. SI = maximum primary plus secondary stress intensity range computed using Equation 10 of paragraph NB3653 of ASME,Section III. When 3Sm, the limit of Equation 10, is exceeded, the requirements is NB3653.6 are used.

Specifically, Equations 12 and 13 are satisfied and a factor Ke is used in the fatigue analysis (Equation 14).

McGuire Nuclear Station UFSAR Table 3-23 (Page 1 of 1)

(27 MAR 2002)

Table 3-23. Deleted Per 2002 Update

McGuire Nuclear Station UFSAR Table 3-24 (Page 1 of 1)

(14 OCT 2000)

Table 3-24. Exceptions to Criteria. Presented in FSAR Section 3.6 for the Performance of the Pipe Break Analysis System Exception Reason for Exception Alternative to Criteria Safety Injection (10" accumulator lines)

Pipe break restraints are not provided for postulated breaks at eight locations in the 10" accumulator lines inside containment.

Construction of restraints would transfer large loads to the shield wall requiring a number of large anchor bolts. Since the shield wall is heavily reinforced, installation of high tension anchor bolts is impossible. Extreme congestion in the area further prohibits an adequate restraint design.

Inservice inspection with the addition of monitoring unidentified RCS Leakage utilizing the RCS Leakage Detection System is provided for the postulated break locations. These measures are considered equivalent to a rupture restraint, and therefore, adequate protection is provided for the postulated breaks.

McGuire Nuclear Station UFSAR Table 3-25 (Page 1 of 1)

(14 OCT 2000)

Table 3-25. Damping Values for Westinghouse Supplied Equipment(6)

Item Damping (Percent of Critical)

Normal / Upset Conditions Faulted Condition OBE SSE/DBA Primary Coolant Loop System - Components 2

4(3)

Primary Coolant Loop System - Large Piping1 2(7) 4(7)

Small Piping 1/2, 1(2) 2 Welded Steel Structures 2

4(5)

Bolted and/or Riveted Steel Structures 4

7(4)

Control Rod Drive, Mechanisms & Support System 5(5) 5(3)

Fuel Assemblies 7(5) 9(3)

Note:

1. Generally applicable to 12" or larger piping

2. For multiple hanger supported piping

3. Damping of 3% used for steam generator replacement analysis

4. Damping of 5% used for steam generator replacement analysis

5. Damping of 2% used for steam generator replacement analysis

6. Also includes the BWI supplied replacement steam generators

7. N-411 damping used for OBE & SSE steam generator replacement analyses

McGuire Nuclear Station UFSAR Table 3-26 (Page 1 of 1)

(14 OCT 2000)

Table 3-26. Reactor Building and Interior Structure Comparison of Peak Responses Mass Location Peak Acceleration on the Response Spectrum(G's)

Percent Increase Fixed Base Combined Interaction Reactor Vessel Support 0.498 0.493

-0.1 Steam Generator Support 1.724 1.692

-1.85 Penetration Support 1.074 1.176

+9.5

McGuire Nuclear Station UFSAR Table 3-27 (Page 1 of 1)

(14 OCT 2000)

Table 3-27. Comparison of Response Spectrum And Time-History Responses(1)(2)

Mass Point Accelerations (G's)

Moments (X104 Ft-K)

Shears (X103 Kips)

Response

Spectrum Time-History Response Spectrum Time-History Response Spectrum Time-History 1

.0467

.1174 27.26 30.97 4.72 6.28 2

.0638

.1278 23.21 25.86 4.47 5.60 3

.0872

.1434 18.63 20.21 4.35 5.31 4

.1125

.1601 14.43 15.16 4.16 4.94 5

.1344

.1766 11.81 12.13 3.79 4.32 6

.1407

.1811 10.94 11.14 3.70 4.18 7

.1818

.2138 7.51 7.57 2.91 3.00 8

.2247

.2458 4.60 4.64 2.43 2.45 9

.2645

.2750 2.35 2.37 1.88 1.89 10

.2785

.2848 1.73 1.75 1.38 1.39 11

.3028

.3002 0.91 0.92 1.13 1.14 12

.3284

.3153 0.32 0.32 0.74 0.75 13

.3528

.3301 0.0 0.0 0.37 0.37 Notes:

1. Refer to Figure 3-20 for mathematical model

2. Values for Response Spectrum Technique are based on the original analysis (MCC-1134.02-00-0003), before Steam Generator replacement and Ice Condenser mass adjustments.

McGuire Nuclear Station UFSAR Table 3-28 (Page 1 of 1)

(14 OCT 2000)

Table 3-28. Maximum Blowdown LOCA Load Resultants for the Containment Interior Structure Loading Maximum Value Break Location Time OVERTURNING MOMENT 280,950 FT-K ELEMENT NO. 1(1) 0.312 SEC.

HORIZONTAL SHEAR 3,977 KIPS ELEMENT NO. 1(1) 0.312 SEC.

UPLIFT 6,354 KIPS ELEMENT NO. 2(1) 0.222 SEC Note:

1. The blowdown history and structural geometry for Elements No. 1 and No. 6, No. 2 and No. 5 are the same, therefore, the maximums for No. 1 apply to No. 6 and the maximums for No. 2 apply to No. 5.

Refer to Table 6-1 through Table 6-6 for Containment Interior Structure blowdown model

McGuire Nuclear Station UFSAR Table 3-29 (Page 1 of 12)

(13 APR 2008)

Table 3-29 McGuire Nuclear Station - Containment Subcompartment Pressures Calculated Values of differential Pressure (psi)-Across Compartments MCGUIRE NUCLEAR STATION CONTAINMENT SUBCOMPARTMENT PRESSURES CALCULATED VALUES OF DIFFERENTIAL PRESSURE (PSI) - ACROSS COMPARTMENTS Design Diff Pressure

(= 1.4 x Diff Pressure)

COMP BETWEEN AND COMP MAXIMUM DIFF PRESSURE TIME (SEC)

ELEMENT MINIMUM DIFF PRESSURE TIME (SEC)

ELEMENT MAXIMUM MINIMUM 1

2 0.7158E+01 0.0190 H.L. 1

-0.5517E+01 0.0190 H.L. 2 0.1002E+02

-0.7723E+01 1

6 0.1099E+02 0.0640 H.L. 1

-0.1086E+02 0.0550 H.L. 6 0.1539E+02

-0.1520E+02 1

7 0.1040E+02 0.2700 C.L. 1

-0.2167E+00 0.0370 C.L. 6 0.1456E+02

-0.3033E+00 1

8 0.1168E+02 0.2340 C.L. 1

-0.2000E+00 0.0280 H.L. 6 0.1634E+02

-0.2800E+00 1

9 0.1064E+02 0.2340 C.L. 1

-0.1917E+00 0.0280 H.L. 6 0.1490E+02

-0.2683E+00 1

25 0.1245E+02 0.0730 H.L. 1

-0.1917E+00 0.0280 H.L. 6 0.1743E+02

-0.2683E+00 1

26 0.1237E+02 0.0730 H.L. 1

-0.4250E+00 2.5020 C.L. 6 0.1731E+02

-0.5950E+00 1

27 0.1113E+02 0.0640 H.L. 1

-01917E+00 0.0280 H.L. 6 0.1557E+02

-0.2683E+00 1

28 0.1236E+02 0.0730 H.L. 1

-0.4250E+00 2.5020 C.L. 6 0.1730E+02

-0.5950E+00 1

33 0.1119E+02 0.0550 H.L. 1

-0.2458E+01 0.1450 H.L. 6 0.1567E+02

-0.3442E+01 1

34 0.1238E+02 0.0730 H.L. 1

-0.4167E+00 2.6280 C.L. 6 0.1732E+02

-0.5833E+00 2

1 0.5517E+01 0.0190 H.L. 2

-0.7158E+01 0.0190 H.L. 1 0.7723E+01

-0.1002E+02 2

3 0.6350E+01 0.0640 H.L. 1

-0.4767E+01 0.0280 H.L. 3 0.8890E+01

-0.6673E+01 2

10 0.7783E+01 0.2880 C.L. 2 0.0 0.0100 C.L. 6 0.1090E+02 0.0 2

11 0.8867E+01 0.2340 C.L. 2 0.0 0.0100 C.L. 6 0.1241E+02 0.0 2

12 0.8117E+01 0.2340 C.L. 2 0.0 0.0100 C.L. 6 0.1136E+02 0.0 2

25 0.8775E+01 0.1270 C.L. 2 0.0 0.0100 C.L. 6 0.1228E+02 0.0 2

26 0.8583E+01 0.0910 H.L. 1

-0.4333E+00 2.4840 C.L. 6 0.1202E+02

-0.6067E+00 2

27 0.7192E+01 0.0460 H.L. 2

-0.1500E+00 2.4840 C.L. 6 0.1007E+02

-0.2100E+00 2

28 0.8567E+02 0.0910 H.L. 1

-0.4333E+00 2.4840 C.L. 6 0.1199E+02

-0.6067E+00 2

33 0.6975E+01 0.0370 H.L. 2

-0.1992E+01 0.0910 H.L. 5 0.9765E+02

-0.2788E+01

McGuire Nuclear Station UFSAR Table 3-29 (Page 2 of 12)

(13 APR 2008)

MCGUIRE NUCLEAR STATION CONTAINMENT SUBCOMPARTMENT PRESSURES CALCULATED VALUES OF DIFFERENTIAL PRESSURE (PSI) - ACROSS COMPARTMENTS Design Diff Pressure

(= 1.4 x Diff Pressure)

COMP BETWEEN AND COMP MAXIMUM DIFF PRESSURE TIME (SEC)

ELEMENT MINIMUM DIFF PRESSURE TIME (SEC)

ELEMENT MAXIMUM MINIMUM 3

2 0.4767E+01 0.0280 H.L. 3

-0.6350E+01 0.0640 H.L. 1 0.6673E+01

-0.8890E+01 3

4 0.4675E+01 0.0280 H.L. 3

-0.5725E+01 0.0190 H.L. 4 0.6545E+01

-0.8015E+01 3

13 0.5850E+01 0.2880 C.L. 3

-0.3333E-01 0.0010 C.L. 3 0.8190E+01

-0.4667E-01 3

14 0.7083E+01 0.2520 C.L. 3

-0.3333E-01 0.0010 C.L. 3 0.9917E+01

-0.4667E-01 3

15 0.6583E+01 0.2520 C.L. 3

-0.3333E-01 0.0010 C.L. 3 0.9217E+01

-0.4667E-01 3

25 0.6700E+01 0.0640 H.L. 3

-0.3333E-01 0.0010 C.L. 3 0.9380E+01

-0.4667E-01 3

27 0.6550E+01 0.0550 H.L. 3

-0.5583E+00 2.2140 C.L. 1 0.9170E+01

-0.7817E+00 3

28 0.6717E+01 0.2700 C.L. 3

-0.4083E+00 2.5380 C.L. 6 0.9403E+01

-0.5717E+00 3

29 0.7317E+01 0.2880 C.L. 3

-0.4333E+00 2.5380 C.L. 6 0.1024E+02

-0.6067E+00 3

30 0.6667E+01 0.0640 H.L. 3

-0.4167E+00 2.5560 C.L. 6 0.9333E+01

-0.5833E+00 3

33 0.5683E+01 0.0370 H.L. 3

-0.1183E+01 0.0820 H.L. 6 0.7957E+01

-0.1657E+01 3

35 0.6858E+01 0.2700 C.L. 3

-0.4000E+00 2.5380 C.L. 6 0.9602E+01

-0.5600E+00 4

3 0.5725E+01 0.0190 H.L. 4

-0.4675E+01 0.0280 H.L. 3 0.8015E+01

-0.6545E+01 4

5 0.5717E+01 0.0190 H.L. 4

-0.5558E+01 0.0190 H.L. 5 0.8003E+01

-0.7782E+01 4

16 0.6217E+01 0.2880 C.L. 4

-0.3917E+00 0.0550 C.L. 4 0.8703E+01

-0.5483E+00 4

17 0.7500E+01 0.2340 C.L. 4 0.0 0.0010 C.L. 6 0.1050E+02 0.0 4

18 0.6900E+01 0.2340 C.L. 4 0.0 0.0010 C.L. 6 0.9660E+01 0.0 4

25 0.6783E+01 0.0460 H.L. 4 0.0 0.0010 C.L. 6 0.9497E+01 0.0 4

29 0.7383E+01 0.3060 C.L. 4

-0.5083E+00 2.5200 C.L. 1 0.1034E+02

-0.7117E+00 4

30 0.6908E+01 0.2880 C.L. 4

-0.4917E+00 2.5200 C.L. 1 0.9672E+01

-0.6883E+00 4

31 0.6758E+01 0.0370 H.L. 4

-0.4417E+00 2.4120 C.L. 6 0.9462E+01

-0.6183E+00 4

33 0.6475E+01 0.0370 H.L. 4

-0.1317E+01 0.0730 H.L. 2 0.9065E+01

-0.1843E+01

McGuire Nuclear Station UFSAR Table 3-29 (Page 3 of 12)

(13 APR 2008)

MCGUIRE NUCLEAR STATION CONTAINMENT SUBCOMPARTMENT PRESSURES CALCULATED VALUES OF DIFFERENTIAL PRESSURE (PSI) - ACROSS COMPARTMENTS Design Diff Pressure

(= 1.4 x Diff Pressure)

COMP BETWEEN AND COMP MAXIMUM DIFF PRESSURE TIME (SEC)

ELEMENT MINIMUM DIFF PRESSURE TIME (SEC)

ELEMENT MAXIMUM MINIMUM 4

36 0.6950E+01 0.2520 C.L. 4

-0.4750E+00 2.5200 C.L. 1 0.9730E+01

-0.6650E+00 5

4 0.5558E+01 0.0190 H.L. 5

-0.5717E+01 0.0190 H.L. 4 0.7782E+01

-0.8003E+01 5

6 0.5367E+01 0.0190 H.L. 5

-0.7433E+01 0.0190 H.L. 6 0.7513E+01

-0.1041E+02 5

19 0.7458E+01 0.2880 C.L. 5 0.0 0.0010 C.L. 6 0.1044E+02 0.0 5

20 0.8567E+01 0.2340 C.L. 5 0.0 0.0010 C.L. 6 0.1199E+02 0.0 5

21 0.7817E+01 0.2160 C.L. 5 0.0 0.0010 C.L. 6 0.1094E+02 0.0 5

25 0.8617E+01 0.1180 C.L. 5 0.0 0.0010 C.L. 6 0.1206E+02 0.0 5

30 0.8417E+01 0.1000 H.L. 6

-0.5333E+00 2.3940 C.L. 1 0.1178E+02

-0.7467E+00 5

31 0.7025E+01 0.0460 H.L. 5

-0.1667E+00 2.3760 C.L. 1 0.9835E+01

-0.2333E+00 5

32 0.8417E+01 0.1000 H.L. 6

-0.5250E+00 2.4300 C.L. 1 0.1178E+02

-0.7350E+00 5

33 0.6758E+01 0.0370 H.L. 5

-0.2125E+01 0.0910 H.L. 2 0.9462E+01

-0.2975E+01 6

5 0.7433E+01 0.0190 H.L. 6

-0.5367E+01 0.0190 H.L. 5 0.1041E+02

-0.7513E+01 6

22 0.9958E+01 0.2700 C.L. 6

-0.2250E+00 0.0370 C.L. 1 0.1394E+02

-0.3150E+00 6

23 0.1521E+02 0.3040 C.L. 6

-0.2100E+00 0.0360 C.L. 1 0.2129E+02

-0.2940E+00 6

24 0.1023E+02 0.0460 C.L. 6

-0.1833E+00 0.0370 C.L. 1 0.1431E+02

-0.2567E+00 6

25 0.1210E+02 0.0730 H.L. 6

-0.1833E+00 0.0370 C.L. 1 0.1694E+02

-0.2567E+00 6

30 0.1201E+02 0.0730 H.L. 6

-0.5250E+00 2.5200 C.L. 1 0.1682E+02

-0.7350E+00 6

31 0.1091E+02 0.0550 H.L. 6

-0.1833E+00 0.0370 C.L. 1 0.1527E+02

-0.2567E+00 6

32 0.1201E+02 0.0730 H.L. 6

-0.5250E+00 2.4300 C.L. 1 0.1681E+02

-0.7350E+00 6

33 0.1090E+02 0.0550 H.L. 6

-0.2842E+01 0.1450 H.L. 1 0.1526+02

-0.3978E+01 6

37 0.1199E+02 0.0730 H.L. 6

-0.3833E+00 2.4660 C.L. 1 0.1679E+02

-0.5367E+00

McGuire Nuclear Station UFSAR Table 3-29 (Page 4 of 12)

(13 APR 2008)

MCGUIRE NUCLEAR STATION CONTAINMENT SUBCOMPARTMENT PRESSURES CALCULATED VALUES OF DIFFERENTIAL PRESSURE (PSI) - ACROSS COMPARTMENTS Design Diff Pressure

(= 1.4 x Diff Pressure)

COMP BETWEEN AND COMP MAXIMUM DIFF PRESSURE TIME (SEC)

ELEMENT MINIMUM DIFF PRESSURE TIME (SEC)

ELEMENT MAXIMUM MINIMUM 7

1 0.2167E+00 0.0370 C.L. 6

-0.1040E+02 0.2700 C.L. 1 0.3033E+00

-0.1456E+02 7

2 0.1550E+01 0.0280 H.L. 1

-0.7950E+01 0.2700 C.L. 1 0.2170E+01

-0.1113E+02 7

8 0.3275E+01 0.0460 H.L. 1

-0.9333E+00 0.4860 H.L. 1 0.4585E+01

-0.1307E+01 7

10 0.4592E+01 0.0460 H.L. 1

-0.3425E+01 0.0460 H.L. 2 0.6428E+01

-0.4795E+01 7

25 0.8242E+01 0.0820 H.L. 1

-0.2417E+01 0.2520 H.L. 1 0.1154E+02

-0.3383E+01 7

27 0.6550E+01 0.0730 H.L. 1

-0.5875E+01 0.2340 H.L. 1 0.9170E+01

-0.8225E+01 7

34 0.8125E+01 0.0820 H.L. 1

-0.3025E+01 2.1600 C.L. 2 0.1137E+02

-0.4235E+01 8

1 0.2000E+00 0.0280 H.L. 6

-0.1168E+02 0.2340 C.L. 1 0.2800E+00

-0.1634E+02 8

7 0.9333E+01 0.4860 H.L. 1

-0.3275E+01 0.0460 H.L. 1 0.1307E+01

-0.4585E+01 8

9 0.2408E+01 0.0640 H.L. 1

-0.1558E+01 0.6300 C.L. 1 0.3372E+01

-0.2182E+01 8

11 0.3500E+01 0.0640 H.L. 1

-0.2892E+01 0.1990 H.L. 1 0.4900E+01

-0.4048E+01 8

25 0.5850E+01 0.0820 H.L. 1

-0.2592E+01 0.2340 H.L. 1 0.8190E+01

-0.3628E+01 9

1 0.1917E+00 0.0280 H.L. 6

-0.1064E+02 0.2340 C.L. 1 0.2683E+00

-0.1490E+02 9

8 0.1558E+01 0.6300 C.L. 1

-0.2408E+01 0.0640 H.L. 1 0.2182E+01

-0.3372E+01 9

12 0.2450E+01 0.0730 H.L. 1

-0.1975E+01 0.1990 H.L. 1 0.3430E+01

-0.2765E+01 9

25 0.3608E+01 0.1270 H.L. 2

-0.1442E+01 0.2340 H.L. 1 0.5052E+01

-0.2018E+01 10 2

0.0 0.0100 C.L. 6

-0.7783E+01 0.2880 C.L. 2 0.0

-0.1090E+02 10 7

0.3425E+01 0.0460 H.L. 2

-0.4592E+01 0.0460 H.L. 1 0.4795E+01

-0.6428E+01 10 11 0.2567E+01 0.0460 H.L. 2

-0.6583E+00 0.4860 H.L. 2 0.3593E+01

-0.9217E+00 10 13 0.4583E+01 0.0910 H.L. 1

-0.3658E+01 0.0550 H.L. 3 0.6417E+01

-0.5122E+01

McGuire Nuclear Station UFSAR Table 3-29 (Page 5 of 12)

(13 APR 2008)

MCGUIRE NUCLEAR STATION CONTAINMENT SUBCOMPARTMENT PRESSURES CALCULATED VALUES OF DIFFERENTIAL PRESSURE (PSI) - ACROSS COMPARTMENTS Design Diff Pressure

(= 1.4 x Diff Pressure)

COMP BETWEEN AND COMP MAXIMUM DIFF PRESSURE TIME (SEC)

ELEMENT MINIMUM DIFF PRESSURE TIME (SEC)

ELEMENT MAXIMUM MINIMUM 10 25 0.6825E+01 0.1090 H.L. 1

-0.1525E+01 0.2520 H.L. 2 0.9555E+01

-0.2135E+01 10 27 0.4858E+01 0.0550 H.L. 2

-0.4492E+01 0.2880 C.L. 2 0.6802E+01

-0.6288E+01 11 2

0.0 0.0100 C.L. 6

-0.8867E+01 0.2340 C.L. 2 0.0

-0.1241E+02 11 8

0.2892E+01 0.1990 H.L. 1

-0.3500E+01 0.0640 H.L. 1 0.4048E+01

-0.4900E+01 11 10 0.6583E+00 0.4860 H.L. 2

-0.2567E+01 0.0460 H.L. 2 0.9217E+00

-0.3593E+01 11 12 0.2108E+01 0.1000 H.L. 1

-0.1383E+01 0.6840 C.L. 2 0.2952E+01

-0.1937E+01 11 14 0.3575E+01 0.1000 H.L. 1

-0.2900E+01 0.0730 H.L. 3 0.5005E+01

-0.4060E+01 11 25 0.5208E+01 0.1180 H.L. 1

-0.1783E+01 0.2340 H.L. 2 0.7292E+01

-0.2497E+01 12 2

0.0 0.0100 C.L. 6

-0.8117E+01 0.2340 C.L. 2 0.0

-0.1136e+02 12 9

0.1975e+01 0.1990 H.L. 1

-0.2450E+01 0.0730 H.L. 1 0.2765E+01

-0.3430E+01 12 11 0.1383E+01 0.6840 C.L. 2

-0.2108E+01 0.1000 H.L. 1 0.1937E+01

-0.2952E+01 12 15 0.2408E+01 0.1180 H.L. 1

-0.2017E+01 0.0820 H.L. 3 0.3372E+01

-0.2823E+01 12 25 0.3392E+01 0.1270 H.L. 1

-0.9667E+00 0.2520 H.L. 2 0.4748E+01

-0.1353E+01 13 3

0.3333E-01 0.0010 C.L. 3

-0.5850E+01 0.2880 C.L. 3 0.4667E-01

-0.8190E+01 13 10 0.3658E+01 0.0550 H.L. 3

-0.4583E+01 0.0910 H.L. 1 0.5122E+01

-0.6417E+01 13 14 0.2308E+01 0.0460 H.L. 3

-0.2417E+00 0.4500 H.L. 3 0.3232E+01

-0.3383E+00 13 16 0.3575E+01 0.0550 H.L. 3

-0.4200E+01 0.0550 H.L. 4 0.5005E+01

-0.5880E+01 13 25 0.5592E+01 0.0820 H.L. 3

-0.6417E+00 0.9720 H.L. 3 0.7828E+01

-0.8983E+00 13 27 0.4817E+01 0.0640 H.L. 3

-0.3725E+01 1.2600 C.L. 2 0.6743E+01

-0.5215E+01 13 29 0.5625E+01 0.0820 H.L. 3

-0.2508E+01 1.7460 C.L. 4 0.7875E+01

-0.3512E+01 13 35 0.5525E+01 0.0820 H.L. 3

-0.2583E+01 1.2780 C.L. 2 0.7735E+01

-0.3617E+01

McGuire Nuclear Station UFSAR Table 3-29 (Page 6 of 12)

(13 APR 2008)

MCGUIRE NUCLEAR STATION CONTAINMENT SUBCOMPARTMENT PRESSURES CALCULATED VALUES OF DIFFERENTIAL PRESSURE (PSI) - ACROSS COMPARTMENTS Design Diff Pressure

(= 1.4 x Diff Pressure)

COMP BETWEEN AND COMP MAXIMUM DIFF PRESSURE TIME (SEC)

ELEMENT MINIMUM DIFF PRESSURE TIME (SEC)

ELEMENT MAXIMUM MINIMUM 14 3

0.3333E-01 0.0010 C.L. 3

-0.7083E+01 0.2520 C.L. 3 0.4667E-01

-0.9917E+01 14 11 0.2900E+01 0.0730 H.L. 3

-0.3575E+01 0.1000 H.L. 1 0.4060E+01

-0.5005E+01 14 13 0.2417E+00 0.4500 H.L. 3

-0.2308E+01 0.0460 H.L. 3 0.3383E+00

-0.3232E+01 14 15 0.1783E+01 0.0730 H.L. 3

-0.1267E+01 0.7920 C.L. 3 0.2497E+01

-0.1773E+01 14 17 0.2833E+01 0.0730 H.L. 3

-0.3275E+01 0.0730 H.L. 4 0.3967E+01

-0.4585E+01 14 25 0.4333E+01 0.0910 H.L. 3

-0.1392E+01 0.7920 C.L. 3 0.6067E+01

-0.1948E+01 15 3

0.3333E-01 0.0010 C.L. 3

-0.6583E+01 0.2520 C.L. 3 0.4667E-01

-0.9217E+01 15 12 0.2017E+01 0.0820 H.L. 3

-0.2408E+01 0.1180 H.L. 1 0.2823E+01

-0.3372E+01 15 14 0.1267E+01 0.7920 C.L. 3

-0.1783E+01 0.0730 H.L. 3 0.1773E+01

-0.2497E+01 15 18 0.1975E+01 0.0820 H.L. 3

-0.2317E+01 0.0820 H.L. 4 0.2765E+01

-0.3243E+01 15 25 0.2817E+01 0.0910 H.L. 3

-0.6417E+00 0.2340 H.L. 3 0.3943E+01

-0.8983E+00 16 4

0.3917E+00 0.0550 C.L. 4

-0.6217E+01 0.2880 C.L. 4 0.5483E+00

-0.8703E+01 16 13 0.4200E+01 0.0550 H.L. 4

-0.3575E+01 0.0550 H.L. 3 0.5880E+01

-0.5005E+01 16 17 0.2558E+01 0.0370 H.L. 4

-0.3583E+00 0.4680 H.L. 4 0.3582E+01

-0.5017E+00 16 19 0.3800E+01 0.0460 H.L. 4

-0.3883E+01 0.0820 H.L. 6 0.5320E+01

-0.5437E+01 16 25 0.5733E+01 0.0730 H.L. 4

-0.7417E+00 0.8280 H.L. 4 0.8027E+01

-0.1038E+01 16 29 0.5750E+01 0.0730 H.L. 4

-0.2508E+01 1.7820 C.L. 3 0.8050E+01

-0.3512E+01 16 31 0.4900E+01 0.0550 H.L. 4

-0.3767E+01 1.0440 C.L. 5 0.6860E+01

-0.5273E+01 16 36 0.5667E+01 0.0730 H.L. 4

-0.2558E+01 1,2600 C.L. 6 0.7933E+01

-0.3582E+01 17 4

0.0 0.0010 C.L. 6

-0.7500E+01 0.2340 C.L. 4 0.0

-0.1050E+02

McGuire Nuclear Station UFSAR Table 3-29 (Page 7 of 12)

(13 APR 2008)

MCGUIRE NUCLEAR STATION CONTAINMENT SUBCOMPARTMENT PRESSURES CALCULATED VALUES OF DIFFERENTIAL PRESSURE (PSI) - ACROSS COMPARTMENTS Design Diff Pressure

(= 1.4 x Diff Pressure)

COMP BETWEEN AND COMP MAXIMUM DIFF PRESSURE TIME (SEC)

ELEMENT MINIMUM DIFF PRESSURE TIME (SEC)

ELEMENT MAXIMUM MINIMUM 17 14 0.3275E+01 0.0730 H.L. 4

-0.2833E+01 0.0730 H.L. 3 0.4585E+01

-0.3967E+01 17 16 0.3583E+00 0.4680 H.L. 4

-0.2558E+01 0.0370 H.L. 4 0.5017E+00

-0.3582E+01 17 18 0.1925E+01 0.0640 H.L. 4

-0.1342E+01 0.7020 C.L. 4 0.2695E+01

-0.1878E+01 17 20 0.2942E+01 0.0640 H.L. 4

-0.3042E+01 0.1000 H.L. 6 0.4118E+01

-0.4258E+01 17 25 0.4500E+01 0.0820 H.L. 4

-0.1458E+01 0.2160 H.L. 4 0.6300E+01

-0.2042E+01 18 4

0.0 0.0010 C.L. 6

-0.6900E+01 0.2340 C.L. 4 0.0

-0.9660E+01 18 15 0.2317E+01 0.0820 H.L. 4

-0.1975E+01 0.0820 H.L. 3 0.3243E+01

-0.2765E+01 18 17 0.1342E+01 0.7020 C.L. 4

-0.1925E+01 0.0640 H.L. 4 0.1878E+01

-0.2695E+01 18 21 0.2033E+01 0.3040 H.L. 6

-0.2240E+01 0.3760 C.L. 3 0.2814E+01

-0.3136E+01 18 25 0.2992E+01 0.0910 H.L 4

-0.7583E+00 0.2340 H.L. 4 0.4188E+01

-0.1062E+01 19 5

0.0 0.0010 C.L. 6

-0.7458E+01 0.2880 C.L. 5 0.0

-0.1044E+02 19 16 0.3883E+01 0.0820 H.L. 6

-0.3800E+01 0.0460 H.L. 4 0.5437E+01

-0.5320E+01 19 20 0.2533E+01 0.0460 H.L. 5

-0.6750E+00 0.3060 H.L. 5 0.3547E+01

-0.9450E+00 19 22 0.3300E+01 0.0460 H.L. 5

-0.4683E+01 0.0460 H.L. 6 0.4620E+01

-0.6557E+01 19 25 0.6708E+01 0.1090 H.L. 6

-0.1558E+01 0.2520 H.L. 5 0.9392E+01

-0.2182E+01 19 31 0.4733E+01 0.0550 H.L. 5

-0.4442E+01 0.2880 C.L. 5 0.6627E+01

-0.6218E+01 20 5

0.0 0.0010 C.L. 6

-0.8567E+01 0.2340 C.L. 5 0.0

-0.1199E+02 20 17 0.3042E+01 0.1000 H.L. 6

-0.2942E+01 0.0640 H.L. 4 0.4258E+01

-0.4118E+01 20 19 0.6750E+00 0.3060 H.L. 5

-0.2533E+01 0.0460 H.L. 5 0.9450E+00

-0.3547E+01 20 21 0.2050E+01 0.1000 H.L. 6

-0.1400E+01 0.6840 C.L. 5 0.2870E+01

-0.1960E+01 20 23 0.2983E+01 0.1900 H.L. 6

-0.3542E+01 0.0640 H.L. 6 0.4177E+01

-0.4958E+01

McGuire Nuclear Station UFSAR Table 3-29 (Page 8 of 12)

(13 APR 2008)

MCGUIRE NUCLEAR STATION CONTAINMENT SUBCOMPARTMENT PRESSURES CALCULATED VALUES OF DIFFERENTIAL PRESSURE (PSI) - ACROSS COMPARTMENTS Design Diff Pressure

(= 1.4 x Diff Pressure)

COMP BETWEEN AND COMP MAXIMUM DIFF PRESSURE TIME (SEC)

ELEMENT MINIMUM DIFF PRESSURE TIME (SEC)

ELEMENT MAXIMUM MINIMUM 20 25 0.5100E+01 0.1180 H.L. 6

-0.1850E+01 0.2340 H.L. 5 0.7140E+01

-0.2590E+01 21 5

0.0 0.0010 C.L. 6

-0.7817E+01 0.2160 C.L. 5 0.0

-0.1094E+02 21 18 0.2083E+01 0.1090 H.L. 6

-0.2033E+01 0.0730 H.L. 4 0.2917E+01

-0.2847E+01 21 20 0.1400E+01 0.6840 C.L. 5

-0.2050E+01 0.1000 H.L. 6 0.1960E+01

-0.2870E+01 21 24 0.2042E+01 0.1999 H.L. 6

-0.2492E+01 0.0730 H.L. 6 0.2858E+01

-0.3488E+01 21 25 0.3300E+01 0.1270 H.L. 6

-0.9750E+00 0.2340 H.L. 5 0.4620E+01

-0.1365E+01 22 6

0.2250E+00 0.0370 C.L. 1

-0.9958E+01 0.2700 C.L. 6 0.3150E+00

-0.1394E+02 22 19 0.4683E+01 0.0460 H.L. 6

-0.3300E+01 0.0460 H.L. 5 0.6557E+01

-0.4620E+01 22 23 0.3317E+01 0.0460 H.L. 6

-0.1008E+01 0.4860 C.L. 6 0.4643E+01

-0.1412E+01 22 25 0.8183E+01 0.0820 H.L. 6

-0.2492E+01 0.2340 H.L. 6 0.1146E+02

-0.3488E+01 22 31 0.6508E+01 0.0640 H.L. 6

-0.5850E+01 0.2340 H.L. 6 0.9112E+01

-0.8190E+01 22 37 0.8033E+01 0.0820 H.L. 6

-0.3217E+01 1.7280 C.L. 5 0.1125E+02

-0.4503E+01 23 6

0.2000E+00 0.0370 C.L. 1

-0.1095E+02 0.2160 C.L. 6 0.2800E+00

-0.1533E+02 23 20 0.3542E+01 0.0640 H.L. 6

-0.2983E+01 0.1900 H.L. 6 0.4958E+01

-0.4177E+01 23 22 0.1050E+01 0.2500 H.L. 6

-0.3450E+01 0.1080 H.L. 5 0.1470E+01

-0.4830E+01 23 24 0.2417E+01 0.0640 H.L. 6

-0.1533E+01 0.6120 C.L. 6 0.3383E+01

-0.2147E+01 23 25 0.5850E+01 0.0820 H.L. 6

-0.2567E+01 0.2340 H.L. 6 0.8190E+01

-0.3593E+01 24 6

0.1833E+00 0.0370 C.L. 1

-0.1023E+02 0.0460 H.L. 6 0.2567E+00

-0.1431E+02 24 21 0.2492E+01 0.0730 H.L. 6

-0.2042E+01 0.1990 H.L. 6 0.3488E+01

-0.2868E+01 24 23 0.1533E+01 0.6120 C.L. 6

-0.2417E+01 0.0640 H.L. 6 0.2147E+01

-0.3383E+01

McGuire Nuclear Station UFSAR Table 3-29 (Page 9 of 12)

(13 APR 2008)

MCGUIRE NUCLEAR STATION CONTAINMENT SUBCOMPARTMENT PRESSURES CALCULATED VALUES OF DIFFERENTIAL PRESSURE (PSI) - ACROSS COMPARTMENTS Design Diff Pressure

(= 1.4 x Diff Pressure)

COMP BETWEEN AND COMP MAXIMUM DIFF PRESSURE TIME (SEC)

ELEMENT MINIMUM DIFF PRESSURE TIME (SEC)

ELEMENT MAXIMUM MINIMUM 24 25 0.3583E+01 0.0910 H.L. 6

-0.1433E+01 0.2340 H.L. 6 0.5017E+01

-0.2007E+01 25 25 0.0 2.9520 C.L. 6 0.0 2.9520 C.L. 6 0.0 0.0 26 1

0.4250E+00 2.5020 C.L. 6

-0.1237E+02 0.0730 H.L. 1 0.5950E+00

-0.1731E+02 26 27 0.2917E+00 2.5920 C.L. 6

-0.4700E+01 0.2520 C.L. 1

-.4083E+00

-0.6580E+01 26 28 0.2083E+00 0.5040 C.L. 3

-0.2417E+00 0.3240 C.L. 2 0.2917E+00

-0.3383E+00 26 32 0.6417E+00 0.4860 C.L. 3

-0.5583E+00 0.2700 H.L. 2 0.8983E+00

-0.7817E+00 26 34 0.2000E+00 0.3060 H.L. 2

-0.6667E-01 0.9180 C.L. 5 0.2800E+00

-0.9333E-01 27 1

0.1917E+00 0.0280 H.L. 6

-0.1113E+02 0.0640 H.L. 1 0.2683E+00

-0.1557E+02 27 2

0.1500E+00 2.4840 C.L. 6

-0.7192E+01 0.0460 H.L. 2 0.2100E+00

-0.1007E+02 27 3

0.5583E+00 2.2140 C.L. 1

-0.6550E+01 0.0550 H.L. 3 0.7817E+00

-0.9170E+01 27 10 0.4492E+01 0.2880 C.L. 2

-0.4858E+01 0.0550 H.L. 2 0.6288E+01

-0.6802E+01 27 26 0.4700E+01 0.2520 C.L. 1

-0.2917E+00 2.5920 C.L. 6 0.6580E+01

-0.4083E+00 27 28 0.4648E+01 0.2340 C.L. 1

-0.2917E+00 2.5920 C.L. 6 0.6522E+01

-0.4083E+00 27 34 0.4792E+01 0.2700 C.L. 1

-0.2750E+00 2.6460 C.L. 6 0.6708E+01

-0.3850E+00 27 35 0.4475E+01 0.2700 C.L. 1

-0.2750E+00 2.6460 C.L. 6 0.6256E+01

-0.3850E+00 28 2

0.4333E+00 2.4840 C.L. 6

-0.8567E+01 0.0910 H.L. 1 0.6067E+00

-0.1199E+02 28 3

0.4083E+00 2.5380 C.L. 6

-0.6717E+01 0.2700 C.L. 3 0.5717E+00

-0.9403E+01 28 26 0.2417E+00 0.3240 C.L. 2

-0.2083E+00 0.5040 C.L. 3 0.3383E+00

-0.2917E+00 28 27 0.2917E+00 2.5920 C.L. 6

-0.4658E+01 0.2340 C.L. 1 0.4083E+00

-0.6522E+01 28 29 0.1833E+01 1.0260 C.L. 1

-0.3333E-01 1.8180 C.L. 4 0.2567E+01

-0.4667E-01

McGuire Nuclear Station UFSAR Table 3-29 (Page 10 of 12)

(13 APR 2008)

MCGUIRE NUCLEAR STATION CONTAINMENT SUBCOMPARTMENT PRESSURES CALCULATED VALUES OF DIFFERENTIAL PRESSURE (PSI) - ACROSS COMPARTMENTS Design Diff Pressure

(= 1.4 x Diff Pressure)

COMP BETWEEN AND COMP MAXIMUM DIFF PRESSURE TIME (SEC)

ELEMENT MINIMUM DIFF PRESSURE TIME (SEC)

ELEMENT MAXIMUM MINIMUM 28 30 0.4417E+00 0.2520 H.L. 5

-0.4833E+00 0.2700 H.L. 2 0.6183E+00

-0.6767E+00 28 35 0.5500E+00 0.5940 C.L. 6

-0.2417E+00 0.2340 H.L. 1 0.7700E+00

-0.3383E+00 29 3

0.4333E+00 2.5380 C.L. 6

-0.7317E+01 0.2880 C.L. 3 0.6067E+00

-0.1024E+02 29 4

0.5083E+00 2.5200 C.L. 1

-0.7383E+01 0.3060 C.L. 4 0.7117E+00

-0.1034E+02 29 13 0.2508E+01 1.7460 C.L. 4

-0.5625E+01 0.0820 H.L. 4 0.3512E+01

-0.8050E+01 29 16 0.2508E+01 1.7820 C.L. 3

-0.5750E+01 0.0730 H.L. 4 0.3512E+01

-0.8050E+01 29 28 0.3333E-01 1.8180 C.L. 4

-0.1833E+01 1.0260 C.L. 1 0.4667E-01

-0.2567E+01 29 30 0.3333E-01 2.0520 H.L. 4

-0.1850E+01 0.7560 C.L. 1 0.4667E-01

-0.2590E+01 29 35 0.3333E-01 2.5920 C.L. 6

-0.1542E+01 0.5940 C.L. 1 0.4667E-01

-0.2158E+01 29 36 0.4167E-01 2.9160 C.L. 1

-0.1483E+01 0.5580 C.L. 6 0.5833E-01

-0.2077E+01 30 3

0.4167E+00 2.5560 C.L. 6

-0.6667E+01 0.0640 H.L. 3 0.5833E+00

-0.9333E+01 30 4

0.4917E+00 2.5200 C.L. 1

-0.6908E+01 0.2880 C.L. 4 0.6883E+00

-0.9672E+01 30 5

0.5333E+00 2.3940 C.L. 1

-0.8417E+01 0.1000 H.L. 6 0.7467E+00

-0.1178E+02 30 28 0.4833E+00 0.2700 H.L. 2

-0.4417E+00 0.2520 H.L. 5 0.6767E+00

-0.6183E+00 30 29 0.1850E+01 0.7560 C.L. 1

-0.3333E-01 2.0520 H.L. 4 0.2590E+01

-0.4667E-01 30 31 0.3667E+00 2.5560 C.L. 1

-0.4642E+01 0.2340 C.L. 6 0.5133E+00

-0.6498E+01 30 32 0.9167E-01 0.6840 C.L. 1

-0.1000E+00 0.4860 C.L. 2 0.1283E+00

-0.1400E+00 30 36 0.5750E+00 0.5760 C.L. 1

-0.2667E+00 0.2340 C.L. 6 0.8050E+00

-0.3733E+00 31 4

0.4417E+00 2.4120 C.L. 6

-0.6758E+01 0.0370 H.L. 4 0.6183E+00

-0.9462E+01 31 5

0.1667E+00 2.3760 C.L. 1

-0.7025E+01 0.0460 H.L. 5 0.2333E+00

-0.9835E+01 31 6

0.1833E+00 0.0370 C.L. 1

-0.1091E+02 0.0550 H.L. 6 0.2567E+00

-0.1527E+02

McGuire Nuclear Station UFSAR Table 3-29 (Page 11 of 12)

(13 APR 2008)

MCGUIRE NUCLEAR STATION CONTAINMENT SUBCOMPARTMENT PRESSURES CALCULATED VALUES OF DIFFERENTIAL PRESSURE (PSI) - ACROSS COMPARTMENTS Design Diff Pressure

(= 1.4 x Diff Pressure)

COMP BETWEEN AND COMP MAXIMUM DIFF PRESSURE TIME (SEC)

ELEMENT MINIMUM DIFF PRESSURE TIME (SEC)

ELEMENT MAXIMUM MINIMUM 31 16 0.3767E+01 1.0440 C.L. 5

-0.4900E+01 0.0550 H.L. 4 0.5273E+01

-0.6860E+01 31 19 0.4442E+01 0.2880 C.L. 5

-0.4733E+01 0.0550 H.L. 5 0.6218E+01

-0.6627E+01 31 22 0.5850E+01 0.2340 H.L. 6

-0.6508E+01 0.0640 H.L. 6 0.8190E+01

-0.9112E+01 31 30 0.4642E+01 0.2340 C.L. 6

-0.3667E+00 2.5560 C.L. 1 0.6498E+01

-0.5133E+00 31 32 0.4592E+01 0.2340 C.L. 6

-0.3667E+00 2.4480 C.L. 1 0.6428E-01

-0.5133E+00 31 36 0.4425E+01 0.2160 C.L. 6

-0.3500E+00 2.5560 C.L. 1 0.6195E+01

-0.4900E+00 31 37 0.4408E+01 0.2340 C.L. 6

-0.2167E+00 2,7900 C.L. 1 0.6172E+01

-0.3033E+00 32 5

0.5250E+00 2.4300 C.L. 1

-0.8417E+01 0.1000 H.L. 6 0.7350E+00

-0.1178E+02 32 6

0.5250E+00 2.4300 C.L. 1

-0.1201E+02 0.0730 H.L. 6 0.7350E+00

-0.1681E+02 32 26 0.5583E+00 0.2700 H.L. 2

-0.6417E+00 0.4860 C.L. 3 0.7817E+00

-0.8983E+00 32 30 0.1000E+00 0.4860 C.L. 2

-0.9167E-01 0.6840 C.L. 1 0.1400E+00

-0.1283E+00 32 31 0.3667E+00 2.4480 C.L. 1

-0.4592E+01 0.2340 C.L. 6 0.5133E+00

-0.6428E+01 32 37 0.1500E+00 0.5760 C.L. 3

-0.4833E+00 0.6120 C.L. 4 0.2100E+00

-0.6767E+00 33 1

0.2458E+01 0.1450 H.L. 6

-0.1119E+02 0.0550 H.L. 1 0.3442E+01

-0.1567E+02 33 2

0.1992E+01 0.0910 H.L. 5

-0.6975E+01 0.0370 H.L. 2 0.2788E+01

-0.9765E+01 33 3

0.1183E+01 0.0820 H.L. 6

-0.5683E+01 0.0370 H.L. 3 0.1657E+01

-0.7957E+01 33 4

0.1317E+01 0.0730 H.L. 2

-0.6475E+01 0.0370 H.L. 4 0.1843E+01

-0.9065E+01 33 5

0.2125E+01 0.0910 H.L. 2

-0.6758E+01 0.0370 H.L. 6 0.2975E+01

-0.9462E+01 33 6

0.2842E+01 0.1450 H.L. 1

-0.1090E+02 0.0550 H.L. 6 0.3978E+01

-0.1526E+02 33 25 0.6075E+01 0.2340 C.L. 1

-0.2000E+00 0.0370 H.L. 6 0.8505E+01

-0.2800E+00 34 1

0.4167E+00 2.6280 C.L. 6

-0.1238E+02 0.0730 H.L. 1 0.5833E+00

-0.1732E+02

McGuire Nuclear Station UFSAR Table 3-29 (Page 12 of 12)

(13 APR 2008)

MCGUIRE NUCLEAR STATION CONTAINMENT SUBCOMPARTMENT PRESSURES CALCULATED VALUES OF DIFFERENTIAL PRESSURE (PSI) - ACROSS COMPARTMENTS Design Diff Pressure

(= 1.4 x Diff Pressure)

COMP BETWEEN AND COMP MAXIMUM DIFF PRESSURE TIME (SEC)

ELEMENT MINIMUM DIFF PRESSURE TIME (SEC)

ELEMENT MAXIMUM MINIMUM 34 7

0.3025E+01 2.1600 C.L. 2

-0.8125E+01 0.0820 H.L. 1 0.4235E+01

-0.1137E+02 34 25 0.2758E+01 1.7100 C.L. 3

-0.1058E+01 0.3060 H.L. 2 0.3862E+01

-0.1482E+01 34 26 0.6667E-01 0.9180 C.L. 5

-0.2000E+00 0.3060 H.L. 2 0.9333E-01

-0.2800E+00 34 27 0.2750E+00 2.6460 C.L. 6

-0.4792E+01 0.2700 C.L. 1 0.3850E+00

-0.6708E+01 35 3

0.4000E+00 2.5380 C.L. 6

-0.6858E+01 0.2700 C.L. 3 0.5600E+00

-0.9602E+01 35 13 0.2583E+01 1.2780 C.L. 2

-0.5525E+01 0.0820 H.L. 3 0.3617E+01

-0.7735E+01 35 27 0.2750E+00 2.6460 C.L. 6

-0.4475E+01 0.2700 C.L. 1 0.3850E+00

-0.6265E+01 35 28 0.2417E+00 0.2340 H.L. 1

-0.5500E+00 0.5940 C.L. 5 0.3383E+00

-0.7700E+00 35 29 0.1542E+01 0.5940 C.L. 1

-0.3333E-01 2.5920 C.L. 6 0.2158E+01

-0.4667E-01 36 4

0.4750E+00 2.5200 C.L. 1

-0.6950E+01 0.2520 C.L. 4 0.6650E+00

-0.9730E+01 36 16 0.2558E+01 1.2600 C.L. 6

-0.5667E+01 0.0730 H.L. 4 0.3582E+01

-0.7933E+01 36 29 0.1483E+01 0.5580 C.L. 6

-0.4167E-01 2.9160 C.L. 1 0.2077E+01

-0.5833E-01 36 30 0.2667E+00 0.2340 C.L. 6

-0.5750E+00 0.5760 C.L. 1 0.3733E+00

-0.8050E+00 36 31 0.3500E+00 2.5560 C.L. 1

-0.4425E+01 0.2160 C.L. 6 0.4900E+00

-0.6195E+01 37 6

0.3833E+00 2.4660 C.L. 1

-0.1199E+02 0.0730 H.L. 6 0.5367E+00

-0.1679E+02 37 22 0.3217E+01 1.7280 C.L. 5

-0.8033E+01 0.0820 H.L. 6 0.4503E+01

-0.1125E+02 37 25 0.2842E+01 1.6380 C.L. 4

-0.8750E+00 0.3060 H.L. 4 0.3978E+01

-0.1225E+01 37 31 0.2167E+00 2.7900 C.L. 1

-0.4408E+01 0.2340 C.L. 6 0.3033E+00

-0.6172E+01 37 32 0.4833E+00 0.6120 C.L. 4

-0.1500E+00 0.5760 C.L. 3 0.6767E+00

-0.2100E+00

McGuire Nuclear Station UFSAR Table 3-30 (Page 1 of 1)

(14 OCT 2000)

Table 3-30. Design Loading Conditions The operating condition categories are defined as follows from ASME III, NB 3113.

Normal Conditions - Normal conditions are any conditions in the course of system startup, operation in the design power range, hot standby and system shutdown, other than Upset, Emergency, Faulted or Testing Conditions.

Upset Conditions (Incidents of Moderate Frequency) - Any deviations from Normal Conditions anticipated to occur often enough that design should include a capability to withstand the conditions without operational impairment. The Upset Conditions include those transients which result from any single operator error or control malfunction, transients caused by a fault in a system component requiring its isolation from the system and transients due to loss of load or power. Upset Conditions include any abnormal incidents not resulting in a forced outage and also forced outages for which the corrective action does not include any repair of mechanical damage. The estimated duration of an Upset Condition shall be included in the Design Specifications.

Emergency Conditions (Infrequent Incidents) - Those deviations from Normal Conditions which require shutdown for correction of the conditions or repair of damage in the system. These conditions have a low probability of occurrence but are included to provide assurance that no gross loss of structural integrity will result as a concomitant effect of any damage developed in the system. The total number of postulated occurrences for such events shall not cause more than 25 stress cycles having an Sa value greater than that for 106 cycles from the applicable fatigue design curves of Figures 1-9.0.

Faulted Conditions (limiting Faults) - Those combinations of conditions associated with extremely-low-probability, postulated events whose consequences are such that the integrity and operability of the nuclear energy system may be impaired to the extent that considerations of public health and safety are involved. Such considerations require compliance with safety criteria as may be specified by jurisdictional authorities.

Note:

Definition of Terms from the ASME Boiler and Pressure Vessel Code,Section III, 1971

McGuire Nuclear Station UFSAR Table 3-31 (Page 1 of 1)

(14 OCT 2000)

Table 3-31. Codes and Specifications for Design of Category 1 Structures and Equipment Supports Structural Component Design Codes and Specifications Concrete ACI 318-63 ACI 307-69(1)

Regulatory Guide 1.15 Concrete Reinforcement ASTM A615, Grades 40 and 60 Cadwelds Regulatory Guide 1.10(2)

Structural Steel and Plates ASTM A-36 AISC, 7th Edition(3)

Containment Vessel Shell Subsection B Section III of the ASME Code 1968 Edition Including all the addenda through the Summer of 1970.

(The high strength embedded anchor bolts for the steam generator enclosure walls meet the requirements of Section A for normal operation and 1986 Code, Appendix 'F', for faulted condition.)

Steel Pipes USS T-1 Note:

1. For the design of the Reactor Building for Temperature effects.

2. Valid test results are used. A test is not considered valid if a failure occurs in the bar or near testing machine grips. Test samples for B Series Splices are sister splices only.

3. For visual inspection of structural welds, reference the Visual Weld Acceptance Criteria For Structural Welding at Nuclear Power Plants, NCIG-01, Rev. 2 dated 5/7/85.

McGuire Nuclear Station UFSAR Table 3-32 (Page 1 of 1)

(14 OCT 2000)

Table 3-32. Reactor Building Loading Combinations and Code Requirements Loading Combination Code or Stress Requirements

1. DL + CL

2. DL + OL + OBE

3. DL + OL + DBA

4. DL + OL + W WSD, stresses in reinforcement and concrete are in accordance with Chapter 10, ACI 318 - 1963 Code. AISC with allowable stresses of Fs.

5. DL + OL + SSE

6. DL + OL + Wt

7. DL + OL + SSE + DBA + Y ACI-318, USD with no factors applied to loadings. AISC with allowable stress of 1.5 Fs or 0.9 Fy whichever is less.

DL

= Dead Load, including weight of permanent equipments attached to the Reactor Building.

CL

= Construction Loads.

OL

= Normal Loads on the Reactor Building due to plant operation such as thermal loads and normal penetration loads due to pipe reactions on the building. OL also includes Live Loads during plant operation.

OBE

= Operating Basis Earthquake Load, inertia forces due to base excitation of 8 percent G.

SSE

= Safe Shutdown Earthquake Load, inertia forces due to base excitation of 15 percent G.

DBA

= Design Basis Accident, includes the differential pressures on the Reactor Building and the associated thermal loads.

W

= Normal wind loads.

Y

= Loading on structure due to pipe rupture.

Wt

= Tornado Loading as defined in Section 3.3.2.2.

ACI-318

= Building Code Requirements for Reinforced Concrete, June 1963.

AISC

= Specification for the Design, Fabrication and Erection of Structural Steel Buildings, 1969.

Fs

= Steel allowable stresses as specified in AISC, Part 1.

WSD

= Working Stress Design Method.

Fy

= Steel specified yield stress.

USD

= Ultimate Strength Design method.

As a minimum, the margin of safety as specified in the above codes and specifications is met in the design of the Reactor Building.

AISC is used for the design of structural steel embedments and pipe sleeves through the Reactor Building wall.

In addition to the Design conditions defined above, the roof has been investigated, spot checked, and found structurally adequate for the following loading condition:

SL OL DL U

+

+

=

where SL = Severe snow and ice loads and is defined in Section 2.4.10.

McGuire Nuclear Station UFSAR Table 3-33 (Page 1 of 1)

(14 OCT 2000)

Table 3-33. Containment Vessel Loading Combination and Code Requirements Loading Combination Code Reference DL + CL ASME - Normal Condition DL + OL + DBA ASME - Normal Condition DL + OL + OBE ASME - Normal Condition DL + OL + OBE + P' ASME - Normal Condition DL + OL + SSE ASME - Emergency Condition DL + OL + SSE + DBA ASME - Emergency Condition DL + OL + SSE + P' ASME - Emergency Condition ASME =

ASME Boiler and Pressure Vessel Code,Section III, Subsection B, 1968, including all addenda through Summer of 1970.

DL

=

Own weight of the Containment Vessel and all the permanent attachments to the Containment.

CL

=

Construction Loads.

DBA

=

Design Basis Accident which includes temperature and pressure effects.

OBE

=

Operating Basis Earthquake, 8 percent G.

SSE

=

Safe Shutdown Earthquake, 15 percent G.

OL

=

Normal Operating Loads of the Containment Vessel, including Live Loads, thermal loads and operating pipe reactions.

P'

=

External pressure due to the internal vacuum created by accidental trip of the Containment Spray System.

Stress limits of the Containment Vessel are as prescribed in Figure N-414 of the ASME,Section III, Nuclear Vessels, 1968, including all the addenda up to the Summer of 1970. Buckling is considered in all loading combinations.

McGuire Nuclear Station UFSAR Table 3-34 (Page 1 of 1)

(14 OCT 2000)

Table 3-34. Containment Materials Material Location Material Specification Base Liners SA-516, Grade 60 or 70 Base Liner Embedments SA-516, Grade 60 and/or ASTM A36 Knuckle Plate SA-516, Grade 60 or 70 Shell and Dome Plate SA-516, Grade 60 or 70 Penetrations (Piping and Electrical)

SA-106, Grade B and/or SA-516, Grade 60 Personnel Locks SA-516, Grade 60 and/or Grade 70 Stiffeners SA-516, Grade 60 or 70 Equipment Hatch SA-516, Grade 60 and/or Grade 70 Anchor Bolts SA 320-L43 Anchor Bolt Anchor Plates SA-516, Grade 60

McGuire Nuclear Station UFSAR Table 3-35 (Page 1 of 1)

(14 OCT 2000)

Table 3-35. Deleted Per 1998 Update

McGuire Nuclear Station UFSAR Table 3-36 (Page 1 of 1)

(13 APR 2008)

Table 3-36 Factors of Safety Against Buckling of Shell Panels for the McGuire Containment Vessel Point*

Axial kips/inch Shearing Load kips/inch Safety Factor Against Buckling Shell Panels Treated as Flat Plates Shell Panels Treated as Curved Plates 1

-5.70 5.28 6.00 3.27 2

-3.99 5.31 8.60 4.43 3

-3.02 4.47 11.42 5.93 4

-4.49 1.75 8.36 4.93 5

-4.53 1.08 8.33 4.97 6

-4.13 1.22 9.13 5.43

• See Figure 3-132

McGuire Nuclear Station UFSAR Table 3-37 (Page 1 of 1)

(14 OCT 2000)

Table 3-37. Comparison Between the Actual Design Differential Pressures and the Latest Design Differential Pressures For the Major Structural Barriers of the Interior Structure Structural Barrier Actual Design Diff.

Pressures (psi) (= 1.4 X Diff. Press.)

Latest Calculated Diff. Pressures (psi)

Pressure Values Used for Design Ice Condenser Floor 11.37 8.12 12.74 6.80 5.67 12.74 6.74 4.60 7.84 6.74 4.52 7.84 7.88 6.52 12.74 8.05 6.79 12.74 7.93 7.15 12.74 6.86 6.40 7.84 6.63 4.44 7.84 9.11 4.55 12.74 11.25 7.97 12.74 Operating Floor 17.43 14.13 15.51 12.28 11.35 15.51 9.38 9.19 15.51 9.49 9.53 15.51 12.06 11.25 15.51 16.94 13.57 15.51 Wingwalls 6.71 5.74 10.00 6.27 7.36 10.00 2.16 5.16 10.00 2.08 5.20 10.00 6.19 7.17 10.00 6.17 5.55 10.00 Crane Wall See Note 1.

Note:

1. The Crane Wall is designed to sustain several critical loading combinations. The differential pressures comprise part of these combinations, hence it is not practical to single out these differential pressures on the crane wall for comparison purposes.

McGuire Nuclear Station UFSAR Table 3-38 (Page 1 of 1)

(14 OCT 2000)

Table 3-38. Comparison Between Kalnin's Program Results and Finite Elements Results w (cps)

Mode Kalnin's Finite Elements Reactor Building N = 1 1

4.964768 4.9748 2

13.66 13.729 Containment Vessel N = 0 1

24.9259 24.97 N = 1 1

9.284 9.3234 2

25.69 25.660

McGuire Nuclear Station UFSAR Table 3-39 (Page 1 of 1)

(14 OCT 2000)

Table 3-39. Containment Interior Structures Loading Combinations and Code Requirements Loading Combination Code or Stress Requirements

1. DL + CL

2. DL + OL + Pa

3. Deleted

4. DL + OL + OBE WSD, stresses in reinforcement and concrete are in accordance with Chapter 10, ACI-318 - 1963 Code. AISC with allowable stresses of Fs.

5. DL + OL + SSE

6. DL + OL + SSE + Ta

7. DL + OL + SSE + Pa + Y ACI-318, USD with no factors applied to loadings, AISC with allowable stress of 1.5 Fs or 0.90 Fy whichever is less.

DL

=

Dead Load, including weight of permanent equipment.

CL

=

Construction Loads.

OL

=

Normal Operating Loads, these loads are associated with the plant operation, including Live Loads.

Pa

=

Differential Pressure across the individual Internal Structures due to a Loss-of-Coolant Accident.

Y

=

Loading on structure due to pipe rupture.

Ta

=

Thermal loads on the Internal Structures Components due to a Loss-of Coolant Accident.

OBE

=

Operating Basis Earthquake Load, inertia forces due to base excitation of 8 percent G.

SSE

=

Safe Shutdown Earthquake Load, inertia forces due to base excitation of 15 percent G.

ACI-318

=

Building Code Requirements for Reinforced Concrete, June 1963.

AISC

=

Specification for the Design, Fabrication and Erection of Structural Steel Buildings, 1969.

WSD

=

Working Stress Design Method.

USD

=

Ultimate Strength Design method.

Fy

=

Yield stress of steel.

Fs

=

Allowable stresses in steel as specified in AISC, Part 1.

Note:

Pa and Ta do not act simultaneously on the structure

McGuire Nuclear Station UFSAR Table 3-40 (Page 1 of 1)

(14 OCT 2000)

Table 3-40. Design Properties of Seals Durometer Tensile Elongation Compression Set

1. Membrane Seals:

Initial 65 (approx.)

2260 psi 470%

14.1%

2. Compressible Seals:

Initial 40 1400 psi 800%

12.88%

Note:

1. All tests are in accordance with ASTM D-2000-XX. XX = The current ASTM D-2000 revision at the time of procurement.

McGuire Nuclear Station UFSAR Table 3-41 (Page 1 of 1)

(14 OCT 2000)

Table 3-41. Divider Barrier Seals Minimum Acceptable Physical Properties Membrane Type Seals Tensile Strength Mk 10 39.7 lbs.

Mk 11 39.7 lbs.

Mk 12 10.6 lbs.

Mk 13 288 lbs.

Mk 20 42.4 lbs.

Compression Type Seals Tensile Strength Elongation Durometer 40 Duro 575 psi 350%

+5 pts. from initial 60 Duro 1333 psi 387%

+5 pts. from initial

McGuire Nuclear Station UFSAR Table 3-42 (Page 1 of 2)

(10 OCT 2009)

Table 3-42. Auxiliary Building Loading Conditions Area Loading Conditions Remarks Compartments at El. 695 A, G, H Soil and water pressure. See Note 1.

Compartments at El. 716 A, G, H Soil and water pressure. See Note 1.

Compartments at El. 733 and Equipment Rooms and Switchgear Rooms A, G, H Soil and water pressure. See Note 1.

Diesel Generator Rooms A, B, C, F, G, H Soil and water pressure.

Fuel Pool A, B, C, G, H.

• See Note 2.

Fuel Pool Racks A, G, H See Note 1.

Control Room A, B, C, F, G, H Hot Machine Shop A

Laboratory Area A

Personnel Decontamination A

Waste Shipping Area A

Fuel Shipping Area A

Isolation Valve Area A, B, C, F, G, H, I Soil and water pressure.

New Fuel Racks A, G, H

McGuire Nuclear Station UFSAR Table 3-42 (Page 2 of 2)

(10 OCT 2009)

Area Loading Conditions Remarks Note:

1. Enclosed by enveloping structure designed for wind, tornado wind, tornado missiles and turbine missile, as applicable.

2. Designed for thermal stresses and cask drop accident.

A = All normal dead, equipment, live and wind loads.

B = Normal dead and equipment loads plus tornado loadings C = Tornado missiles.

F = Turbine-Generator missile (or Diesel-Generator missile).

G = Normal dead and equipment loads plus operating seismic loads.

H = Normal dead and equipment loads plus design seismic loads.

I = Normal dead and equipment loads plus design seismic loads plus pipe rupture and pressure loads.

• = Soil and water pressure, pressure due to equipment or railroad ramp.

McGuire Nuclear Station UFSAR Table 3-43 (Page 1 of 2)

(14 OCT 2000)

Table 3-43. Auxiliary Building Loading Combinations Loading Conditions5 Load Combination A,G U = 1.5D + 1.8L U = 1.25 (D + L + W)

U = 0.9D + R +1.1E U = 1.25 (D + R + E)

B,H U = 1.25 (D + R) + E' U = 1.25D + Wt C,F Analyzed on bases of "Design of Protective Structures," Amirikian, A., Bureau of Yards and Docks, Department of the Navy, NAVDOCKS P-51, 1950.

I U = 1.25 (D + R) + P + E' + Y U

= Required Ultimate Load Capacity of Section D

= Dead Load (including equipment load and normal operating pipe loads)

L

= Operating Live Load W

= Wind Load Wt

= Tornado Loading E

= Operating Basis Earthquake Load E'

= Safe Shutdown Earthquake Load R

= Piping Seismic Reactions P

= Peak compartment pressure generated by a postulated pipe break Y

= Equivalent static load on structure from pipe break (including a dynamic load factor of 2.0)

Fs

= Allowable stresses in steel in accordance to AISC, Part 1.

Fy

= Allowable yield stress in steel

McGuire Nuclear Station UFSAR Table 3-43 (Page 2 of 2)

(14 OCT 2000)

Loading Conditions5 Load Combination

1. Design Strengths are in accordance with ACI 318-63, Sections 1504 and 1505 for concrete and reinforcing steel.

2. Concrete design is based on the Ultimate Strength Method in accordance with the June 1963 version of ACI 318. For design of those structural elements not covered in ACI 318-63, the ACI Standard 318-71 is used.

3. Structural Steel design is based on Part 1 of the February, 1969 version of the AISC Specification except that for those loading conditions including Wt, E', P and Y, the maximum allowable steel stress is permitted to be 1.5 Fs or 0.9 Fy, whichever is less.

4. In addition to the Design conditions as defined above, the roof has been investigated, spot checked, and found structurally adequate for the following loading condition:

SL L

D U

+

+

=

where SL = severe snow and ice loads and is defined in Section 2.4.10.

5. See UFSAR Table 3-42, (Page 2 of 2) for the explanation of the Symbols (A thru I) used in Loading Conditions.

McGuire Nuclear Station UFSAR Table 3-44 (Page 1 of 1)

(14 OCT 2000)

Table 3-44. New Fuel Storage Vault Loading Combinations Loading Conditions4 Load Combination A, G U=1.5D + 1.8L U=1.25 (D + L + W)

U=0.9D + R + 1.1E U=1.25 (D + R + E)

B, H U=1.25 (D + R) + E' U=1.25D + Wt C

Analyzed on bases of "Design of Protective Structures," Amirikian, A., Bureau of Yards and Docks, Depart of the Navy, NAVDOCKS P-51, 1950.

I U=1.25 (D + R) + P + E' + Y U

= Required Ultimate Load Capacity of Section D

= Dead Load (including equipment load and normal operating pipe loads)

L

= Live Load W

= Wind Load Wt

= Tornado Loading E

= Operating Basis Earthquake Load E'

= Safe Shutdown Earthquake Load R

= Piping Seismic Reactions P

= Peak compartment pressure generated by a postulated pipe break Y

= Equivalent static load on structure from pipe break (including a dynamic load factor of 2.0)

1. Design Strengths are in accordance with ACI 318-63, Sections 1504 and 1505 for concrete and reinforcing steel.

2. Concrete design is based on the Ultimate Strength Method in accordance with the June 1963 version of ACI 318. For design of those structural elements not covered in ACI 318-63, the ACI Standard 318-71 is used.

3. Structural Steel design is based on Part 1 of the February, 1969 version of the AISC Specification except that for those loading conditions including Wt, E', P and Y, the maximum allowable steel stress is permitted to be 1.5 Fs or 0.9 Fy.

4. See UFSAR Table 3-42 (Page 2 of 2) for the explanation of the Symbols (A through I) used in Loading Conditions.

McGuire Nuclear Station UFSAR Table 3-45 (Page 1 of 1)

(14 OCT 2000)

Table 3-45. Factors of Safety for Category 1 Structures Against Overturning and Sliding Structure Factor of Safety Overturning Sliding Reactor Building Complex 2.3 2.24 Main Auxiliary Building 9.23 2.82 Diesel Generator Building 8.2 3.7 Main Steam Line Isolation Valve Enclosure 1.10 1.22 New Fuel Vault 1.23 1.15 SNSW Intake 2.44 1.18 SNSW Discharge 2.72 1.24 SNSW Overflow Spillway 1.05 1.28

McGuire Nuclear Station UFSAR Table 3-46 (Page 1 of 2)

(14 OCT 2000)

Table 3-46. Piping Systems Included in Vibration Test Program System Reactor Coolant System Safety Injection System Residual Heat Removal System Containment Spray System Chemical and Volume Control System Boron Recycle System Boron Thermal Regeneration System Component Cooling System Liquid Waste Disposal System Fuel Pool Cooling and Cleanup System Diesel Generator Fuel Oil System Diesel Generator Cooling Water System

McGuire Nuclear Station UFSAR Table 3-46 (Page 2 of 2)

(14 OCT 2000)

System Diesel Generator Lube Oil System Nuclear Service Water System Refueling Water System Main Steam System Feedwater System Auxiliary Feedwater System Steam Dump System Containment Ventilation Cooling Water System Control Area Chilled Water System Steam Generator Blowdown Recycle System Recirculated Cooling Water System

McGuire Nuclear Station UFSAR Table 3-47 (Page 1 of 2)

(14 OCT 2000)

Table 3-47. Design Conditions, Load Combinations, and Code Compliance Criteria for Duke Classes B, C, and F Piping Condition Loads Code Compliance Criteria

1. Sustained Loads5 Pressure Weight Other Sustained Mechanical loads Primary stresses Sh 3, 7, 8

2. Thermal Expansion Thermal Expansion Thermal Anchor Movements Maximum Secondary Stress Envelope 3, 7, 8

3. Upset Loads Pressure Weight OBE (Inertia)

OBE (Anchor Movements)1 DFL 2 Wind4 (Primary Stresses) 1.2 Sh 7, 8

4.

a. Faulted Loads Pressure Weight SSE (Inertia)

DFL 2 Tornado 4 (Primary Stresses) 2.4 Sh 7, 8

b. Faulted Loads Pressure Weight Pipe Rupture 6 (Primary Stresses) 2.4 Sh 7, 8

McGuire Nuclear Station UFSAR Table 3-47 (Page 2 of 2)

(14 OCT 2000)

Condition Loads Code Compliance Criteria Notes:

1. Stresses due to seismic displacements such as anchor movements may alternatively be considered as secondary stresses and combined with thermal expansion.

2. Dynamic Internal Fluid Loads are occasional loads such as relief valve thrust, steamhammer, waterhammer or loads associated with Plant Upset or Faulted Condition where appropriate.

3. The allowable stress, SA, may be increased when primary stresses due to sustained loads are less than Sh per ASME Section III, Subsection NC-3611.1(b)4(a).

4. Wind as defined in UFSAR Section 3.3.1.1 is applicable to the Upset Condition, but not concurrent with seismic loads inertia or anchor movement loadings per ASME III 1971, Subsection NC-3622.

Tornado as defined in UFSAR Section 3.3.2.1 is applicable to the Faulted Condition, but not concurrent with seismic loads inertia or anchor movement loadings per ASME III 1971, Subsection NC-3622.

5. If, during operation, the system normally carries a medium other than water (air, gas, steam),

sustained loads should be checked for weight loads during hydrotest as well as normal operation weight loads.

6. Pipe rupture loadings include LOCA and MSLB as applicable.

7. ASME Code Case N-318-4, "Procedure for Evaluation of the Design of Rectangular Cross Section Attachments on Class 2 or 3 Piping,Section III, Division 1", may be used in case of pipe welded attachment qualification. It should be documented in appropriate calculations.

8. ASME Code Case N-392-1, "Procedure for Evaluation of the Design of Hollow Circular Cross Section Welded Attachments on Class 2 or 3 Piping,Section III, Division 1", may be used in case of pipe welded attachment qualification. It should be documented in appropriate calculations

McGuire Nuclear Station UFSAR Table 3-48 (Page 1 of 1)

(14 OCT 2000)

Table 3-48. Stress Criteria For Reactor Containment Mechanical Penetrations Duke Class B Condition Piping Loads Criteria

1. Normal Thermal Displacement Pressure Weight ASME III, Class 2

2. Upset Thermal Displacement OBE (Displacement)

Pressure Weight OBE (Inertia)

(Secondary Stresses) SA (Primary Stresses) 1.2 Sh

3. Faulted Thermal Displacement1 SSE (Displacement)1 Pressure Weight SSE (Inertia) Pipe Rupture (Primary Stresses) 2.4 Sh Note:

1. For the faulted condition, the displacement induced stresses are considered primary stresses.

McGuire Nuclear Station UFSAR Table 3-49 (Page 1 of 1)

(14 OCT 2000)

Table 3-49. Stress Criteria For Supports, Restraints, and Anchors Duke Class A Condition Piping Loads Criteria

1. Normal Thermal Displacement Pressure (As Applicable)

Weight ASME III, Class 1

2. Upset Thermal Displacement OBE (Displacement)

Pressure (As Applicable)

Weight OBE (Inertia)

ASME III, Class 1

3. Faulted Thermal Displacement SSE (Displacement) Pressure (As Applicable) Weight SSE (Inertia) Pipe Rupture (Primary Stresses) Yield Stress At Operating Temperature or ASME III, Class 11

4. Faulted Thermal Displacement SSE (Displacement) Pressure (As Applicable) Weight SSE (Inertia) Pipe Rupture (Primary Stresses) Yield Stress At Operating Temperature or ASME III, Class 11 Note:

1. ASME III, Class 1 allowables (i.e. Level D) may be used for standard vendor supplied pipe support components provided a certified Design Report Summary (DRS) or Load Capacity Data Sheet (LCDS) is available for the component.

McGuire Nuclear Station UFSAR Table 3-50 (Page 1 of 1)

(14 OCT 2000)

Table 3-50. Stress Criteria For Supports, Restraints, and Anchors Duke Classes B, C, and F Condition Piping Loads Criteria

1. Normal Thermal Displacement Pressure (As Applicable)

Weight ASME III, Class 2

2. Upset Thermal Displacement OBE (Displacement)

Pressure (As Applicable)

Weight OBE (Inertia) wind1 ASME III, Class 2

3. Faulted Thermal Displacement SSE (Displacement) Pressure (As Applicable) Weight SSE (Inertia) tornado1 (Primary Stresses) Yield Stress At Operating Temperature or ASME III, Class 22

4. Faulted Thermal Displacement SSE (Displacement) Pressure (As Applicable) Weight SSE (Inertia) Pipe Rupture Tornado1 (Primary Stresses) Yield Stress At Operating Temperature or ASME III, Class 22 Note:

1. Wind as defined in UFSAR Section 3.3.1.1 is applicable to the Upset Condition, but not concurrent with seismic loads inertia or anchor movement loadings per ASME III 1971, Subsection NC-3622.

Tornado as defined in UFSAR Section 3.3.2.1 is applicable to the Faulted Condition, but not concurrent with seismic loads inertia or anchor movement loadings per ASME III 1971, Subsection NC-3622.

2. ASME III, Class 2 allowables (i.e. Level D) may be used for standard vendor supplied components provided a certified Design Report Summary (DRS) or Load Capacity Data Sheet (LCDS) is available for the component.

McGuire Nuclear Station UFSAR Table 3-51 (Page 1 of 1)

(14 OCT 2000)

Table 3-51. Stress Criteria For Safety Class 2 and 3 Cylindrical Shell Type Equipment and Components And Their Supports Condition Loads Criteria

1. Normal & Upset (includes Normal Operating Effects Plus OBE Effects)

Nozzle Loads Pressure Weight Support Reactions ASME Section III Class 2 or 3 (See Table 3-4)

2. Faulted (includes Normal Operating Effects Plus SSE Effects)

Nozzle Loads Pressure Weight Support Reactions Pressure Boundary - ASME Section III, Class 2 and 3 and Par. NB-3225 Supports - (Primary Stresses) Yield Stress At Operating Temperature

McGuire Nuclear Station UFSAR Table 3-52 (Page 1 of 1)

(14 OCT 2000)

Table 3-52. Westinghouse Design Criteria for ASME Class 2 and 3 Components Condition Vessels/Tanks Pumps Normal S

0.1 Pm ASME Section III Subsection NC-3400 and ND-3400 S

0.3 P

P b

m

+

Upset S

0.1 Pm S

0.1 Pm S

0.3 P

P b

m

+

S 5.1 P

P b

m

+

Faulted S

5.1 Pm S

2.1 Pm S

0.3 P

P b

m

+

S 8.1 P

P b

m

S

= Allowable stress for the material at temperature as given in ASME Section III Pm

= Membrane stress, i.e., the component of normal stress which is uniformly distributed across the thickness of the solid section under consideration.

Pb

= Bending stress, i.e., the linearly variable component of normal stress across the thickness of the solid section under consideration.

ASME CLASS LOADING COMBINATION CONDITION CLASSIFICATION 2

1. Pressure + Deadweight +

Thermal (Nozzle loads only)

Normal

2. Pressure + Deadweight +

OBE Upset

3. Pressure + Deadweight +

SSE Faulted 3

1. Pressure + Deadweight +

Thermal (Nozzle loads only)

Normal

2. Pressure + Deadweight +

SSE Faulted

McGuire Nuclear Station UFSAR Table 3-53 (Page 1 of 1)

(14 OCT 2000)

Table 3-53. Guard Pipe Designs Relying on ASME Code Case 1606 Penetration Mk. No.

Process Pipe Size (NPS)

System Energy Classification of Process Pipe Ratio:

Design Pressure Pressure Rating Note 1 154 34" SM - Main Steam - A High Energy 1.22 261 34" SM - Main Steam - B High Energy 1.22 393 34" SM - Main Steam - C High Energy 1.22 441 34" SM - Main Steam - D High Energy 1.22 329 3"

NV - Charging Pumps to Regenerative Hx.

High Energy 1.55 339 2"

NV - Sealwater Injection High Energy 1.56 343 2"

NV - Sealwater Injection High Energy 1.56 344 2"

NV - Sealwater Injection High Energy 1.56 350 2"

NV - Sealwater Injection High Energy 1.56 316 4"

NI - Safety Injection -

A Moderate Energy 1.56 319 4"

NI - Safety Injection -

B Moderate Energy 1.56 348 2"

NI - UHI Check Valve Flush Moderate Energy 1.56 351 3"

NI - From Boron Injection Tank Moderate Energy 1.56 352 4"

NI - Safety Injection -

A & B Moderate Energy 1.56 N/A Note 2 34" SM - Main Steam High Energy 1.43 N/A Note 2 18" CF - Feedwater High Energy 1.39 Note:

1. Design Pressure refers to the maximum operating pressure during normal plant conditions. Pressure rating refers to allowable pressure utilizing Eq (3) of Paragraph NC-3641-1 of ASME Section III, 1971 Edition through Winter 1971 Add.

2. This guard pipe encloses process pipe in areas other than penetrations.

McGuire Nuclear Station UFSAR Table 3-54 (Page 1 of 1)

(14 OCT 2000)

Table 3-54. Comparison of Predicted PWHIP Response - Inelastic Pipe Element. (Example Problem:

Figure 3-117)

Response

Theoretical Prediction 1 PWHIP Prediction 2 Conclusion Maximum Displacement ymax @ tm 0.8041" @ 0.0668 sec.

0.8038" @ 0.067 sec.

Predicted ymax and tm are exact within three significant figures of accuracy.

Peak-to-Peak Elastic Displacement Range ymax - ymin 0.3669" 0.3679" Predicted response is within 0.3% of theoretical.

Elastic Natural Period of Free Vibration Tn 0.1107 sec.

0.112 sec.

Predicted response is within 1.2% of theoretical.

Note:

1. Theoretical response based on the solution by Biggs, Introduction to Structural Dynamics, McGraw-Hill, 1964, recomputed with intermediate calculations to 5 significant figures.

2. PWHIP response calculated using a numerical integration time step of 0.001 seconds and output at that time interval.

[HISTORICAL INFORMATION - NOT REQUIRED TO BE REVISED]

McGuire Nuclear Station UFSAR Table 3-55 (Page 1 of 1)

(14 OCT 2000)

Table 3-55. Comparison of Predicted PWHIP Response - Inelastic Yield Element. (Example Problem: Figure 3-118)

Response

Theoretical Prediction1 PWHIP Prediction2 Conclusion Maximum Displacement ymax @ tm 0.8041" @ 0.0668 sec.

0.8051" @ 0.067 sec. (0.8043" when corrected for initial gap3 PWHIP response exact within 3 significant figures of accuracy.

Peak-to-Peak Elastic Displacement Range ymax - ymin 0.3669" 0.3670" PWHIP response exact within 3 significant figures of accuracy.

Elastic Natural Period of Free Vibration Tn 0.1107 sec.

0.1105 sec.

PWHIP response exact within 3 significant figures of accuracy.

Note:

1. Theoretical response based on the solution by Biggs, Introduction to Structural Dynamics, McGraw-Hill, 1964, recomputed with intermediate calculations to 5 significant figures.

2. PWHIP response calculated using a numerical integration time step of 0.0005 seconds

3. PWHIP model included an initial gap of 0.0005". It is calculated that initial kinetic energy at the time of impact (0.0015 seconds) increases maximum displacement by 0.0003". Therefore, the initial gap resulted in a maximum displacement 0.0008" greater than that of a zero gap model.

[HISTORICAL INFORMATION - NOT REQUIRED TO BE REVISED]

McGuire Nuclear Station UFSAR Table 3-56 (Page 1 of 1)

(14 OCT 2000)

Table 3-56. Comparison of Predicted PWHIP Response - Inelastic Yield Element. (Example Problem: Figure 3-120)

Response

Theoretical Prediction 1 PWHIP Prediction 2 Conclusion Time of initial impact (y = h) to 0.07198 sec.

y =.9867" @ t = 0.072 sec.

y = 1.0145" @ t = 0.073 sec.

PWHIP response is within one integration time step of theoretical.

Velocity at initial impact yo 27.785 in/sec.

in/sec 27.70 sec 001

.9867" 0145

.1

=

PWHIP response is exact within three significant figures of accuracy.

Time to zero velocity tm 0.10037 sec.

0.101 sec.

PWHIP response is within one integration time step of theoretical.

Maximum displacement ym 1.5506 in.

1.5514 in.

PWHIP response is exact within three significant figures of accuracy.

Notes:

1. Theoretical response based on the solution by Thomason, Vibration Theory and Application, Prentice-Hall, 1965.

2. PWHIP response calculated using a numerical integration time step of 0.001 seconds.

[HISTORICAL INFORMATION - NOT REQUIRED TO BE REVISED]

McGuire Nuclear Station UFSAR Table 3-57 (Page 1 of 1)

(27 MAR 2002)

Table 3-57. HVAC Design Codes Component Design Code

1. Standard for Installation of Air Cond'g. & Ventilation Systems NFPA 90A

2. Central Station Air Handling Units ARI 430-66

3. Chillers:

ASHRAE 30-60 ARI 555-63 ARI 550-66

4. Coils:

ARI 410-64 ASHRAE 33-64

5. Fans:

AMCA 99 AMCA 210-67

6. Fire Dampers UL 555-1968

7. Electric Coils:

UL 499-1968 NEMA HE-1-1966

8. Carbon Filters ANSI/ASME N509-1976

9. Dampers Refer to system purchase specification

McGuire Nuclear Station UFSAR Table 3-58 (Page 1 of 1)

(14 OCT 2000)

Table 3-58. Maximum Deflections for Reactor Internals Under Blowdown and Seismic Excitation (1-Millisecond Double-Ended Break)

[HISTORICAL INFORMATION - NOT REQUIRED TO BE REVISED]

Component Blowdown Deflection (Inches)

Seismic Deflection, (Inches)

Direction Maximum Total Deflection, (Inches)

Allowable Deflection, (Inches)

Deflection for No Loss of

Function, (Inches)

Cold Leg Hot Leg Upper Barrel Radial Inward 0.0 0.057 0.002 Horizontal 0.059 4.1 8.2 Radial Outward 0.431 0.029 0.002 Horizontal 0.460 0.5 1.0 Upper Core Plate 0.016 0.015 0

Vertical 0.016 0.1001 0.150 Rod Cluster Control Guide Tubes (deflection as a beam)

(54)

<Allowable 0.010 Horizontal

<Allowable 1.0 1.60 to 1.75 (2)

<N.L.F

>Allowable 0.010 Horizontal

<N.L.F

>Allowable 1.0 1.60 to 1.75 (5)

>N.L.F.

0.010 Horizontal

>N.L.F 1.0 1.60 to 1.75 Fuel Assembly

~0

~0

~0 Horizontal

~0 0.036 0.072 Thimble (Cross section distortion)

Note:

1. Only to assure that the plate will not touch a guide tube.

McGuire Nuclear Station UFSAR Table 3-59 (Page 1 of 1)

(14 OCT 2000)

Table 3-59. Maximum Stress Intensities for Reactor Internals (1-Millisecond Pipe Break and Seismics) [HISTORICAL INFORMATION -

NOT REQUIRED TO BE REVISED]

Component BLOWDOWN STRESSES, PSI Hot Leg Break Cold Leg Break Maximum Membrane Maximum Total Maximum Membrane Maximum Total Direct Seismic1 Stress, psi Vertical/horiz Maximum Total Blowdown Plus Seismic psi Barrel (Grith Weld) 21,440 31,340 38,900 46,200 100/400 46,700 Barrel (Flange Weld) 19,820 29,720 18,430 47,400 410/600 48,710 Upper Support Columns (1) 6,200 39,200

---/---

39,200 (55) 6,200

<20,300

---/---

<20,300 Fuel Assembly - Top Nozzle Plate 28,700 8,000 0/0 28,700 Fuel Assembly - Bottom Nozzle Plate 38,700 40,800 400/---

41,200 Fuel Assembly Thimbles 6,600 6,600 2,300 2,300

---/---

6,600 Allowable Stress Maximum Membrane Pm + 2.4 Sm + 39,800 psi Maximum Total Pm + Pb = 3.0 Sm = 48,000 psi S (evaluated at 588°F) = 16,600 psi (per Winter Addenda ASME Section III Code)

Note:

1. Values are for high level seismic plant.

McGuire Nuclear Station UFSAR Table 3-60 (Page 1 of 3)

(14 OCT 2000)

Table 3-60. Electrical Systems & Components Seismic Criteria [HISTORICAL INFORMATION -

NOT REQUIRED TO BE REVISED]

Equipment4 Base Load Levels3 Freq. Range (Hz)1 Air Conditioning System (Control, Equipment & Cable Room)

Compressor Motors Note 2 Vent Fan Motors Note 2 Chill Water Pump Motors Note 2 Control Room Makeup Fan Motor Note 2 Equipment & Cable Room Makeup Fan Motor Note 2 Annulus Vent System Fan Motors Note 2 Air Operated Valves (Solenoids) 0.5g 1 to 33 Aux. Bldg. Vent System EFS Air Handling Unit Motors Note 2 Chemical & Volume Control System Selected M.O. Valve Motors 0.5g 5 to 200 Containment Air Return System Air Return Fan Motors Note 2 H2 Skimmer Fan Motors Note 2 Containment Spray System Containment Spray Pump Motors Note 2 Selected M.O. Valve Motors 0.5g 5 to 200 Containment Isolation System Containment Isolation Valve Motors 0.5g 5 to 200 Component Cooling Water System Comp. Cooling Pump Motors Note 2 Selected M.O. Valve Motors 0.5g 5 to 200 Deisel Bldg. Vent. System Diesel Room Supply Fan Motors Emergency diesel Aux. Systems Emerg. Diesel Generator Note 2 Dsl. Crank Case Vac. Pump Motor 0.25g 0.5 to 40

McGuire Nuclear Station UFSAR Table 3-60 (Page 2 of 3)

(14 OCT 2000)

Equipment4 Base Load Levels3 Freq. Range (Hz)1 Dsl. F.O. Transfer Pump Motor 0.25g 0.5 to 30 Dsl. L.O. Filter 0.25g 0.5 to 40 Dsl. 600/120V Pnl. Board 0.25g 0.5 to 40 Dsl. Lube Oil Pump Motors 0.25g 0.5 to 40 Dsl. Jacket & Intercoolant Pump Motor 0.25g 0.5 to 40 Dsl. Air Compressor Motors 0.25g 0.5 to 40 Auxiliary Feedwater System Aux. Feedwater Pump Motor Note 2 Penetrations 0.25, 2.1 or 3.75g depending on location 0.5 to 40 4KV ES Aux. Power System 4KV Metalclad ES Switchgear 0.8g 0.5 to 50 4160/600V Transformers 0.8g 0.5 to 50 600V Essential Aux. Power System 600V ES Motor Control Centers 1.2g 1 to 40 600/208V Essential Transformers 0.4g 0.5 to 40 125V DC Vital Inst. & Cont. Power System 125V DC Battery Chargers 0.4g 0.5 to 40 125V DC Batteries 0.4g 0.5 to 40 125V DC Distribution Centers 1.2g 1 to 40 125V DC Panelboards 1.2g 1 to 40 120V AC Vital Power System 125VDC/120V AC Static Inverters 0.4g 0.5 to 40 120VAC Vital Pwr Panelboards 1.2g 1 to 40 Radioactive Waste Systems RHR & CS Pump Room Sump Pump Motor Note 2 Residual Heat Removal System Selected M.O. Valve Motors 0.5g 5 to 200 Safety Injection System Motor Operated Valve Motors 0.5g 5 to 200 Steam Supply System

McGuire Nuclear Station UFSAR Table 3-60 (Page 3 of 3)

(14 OCT 2000)

Equipment4 Base Load Levels3 Freq. Range (Hz)1 Main Steam Iso. Valve Solenoids Note 2 Notes:

1. Hertz

2. Seismic qualification being accomplished by analyses

3. Meaning maximum horizontal ground or floor acceleration level for SSE

4. For other equipment see references sited in Section 3.1

McGuire Nuclear Station UFSAR Table 3-61 (Page 1 of 2)

(13 APR 2008)

Table 3-61. Post-Accident Equipment (Inside Containment) Operational Requirements Equipment Name Operating Mode Required Duration of Operation INSTRUMENTATION Pressurizer pressure channels (4)

Continuous SI Initiation Pressurizer level channels (3)

Continuous Two weeks Steam generator level channels N/R (16)

Continuous Four months Reactor Coolant system pressure W/R (2)1 Continuous Two weeks Reactor Coolant RTD N/R Continuous Reactor Trip Reactor Coolant System RTD W/R Continuous Two weeks Containment Sump Level1 Continuous Four months Containment Pressure1 Continuous Three months VALVES & OPERATORS SIS motor operated valves, high head injection line (4)

Continuous Five minutes Containment isolation valves, motor operated & solenoid operated Operate on signal Five minutes Containment air return & H2 skimmer system valves Open on signal Five minutes OTHER EQUIPMENT Deleted Per 2008 Update Safeguard equipment power, control and instrument cable Continuous 4 mo.

Containment Air Return and H2 Skimmer Fans and Motors Continuous 2 mo.

Penetrations As indicated for above listed equipment As indicated for above listed equipment Containment High Range Radiation Monitor Detections Continuous Two weeks Acoustical Valve Position Monitor Sensors & Amps Continuous Two weeks

McGuire Nuclear Station UFSAR Table 3-61 (Page 2 of 2)

(13 APR 2008)

Equipment Name Operating Mode Required Duration of Operation Limit Switches Continuous Five minutes Pressure Switches Continuous Five minutes Note:

1. Transmitters for these signals are located outside of the Containment.

2. See FSAR Figure 1-9 through Figure 1-16 for location of Equipment.

McGuire Nuclear Station UFSAR Table 3-62 (Page 1 of 1)

(30 NOV 2012)

Table 3-62. Control Complex Areas Ventilation Systems Analysis Results Control Complex Area Temperature1 Relative Humidity1 Control Room 75°F 45%

Cable Room 85°F 30%

Battery and Equipment Room 80°F 50%

Switchgear Rooms 85°F 30%

Motor Control Center Rooms 85°F 30%

Ventilation Equipment Rooms 100°F 30%

Electrical Penetration Rooms (EL 767) 85°F 30%

Note:

1. The temperature and relative humidity values are nominal and may vary by +/-5°F and +/-10% RH respectively.

McGuire Nuclear Station UFSAR Table 3-63 (Page 1 of 1)

(13 OCT 2018)

Table 3-63. Structures, Systems and Components Included in TORMIS Analysis Not Designed for Design Basis Tornado Generated Missiles3 Category 1 SSC Unit 1 & 2 Main Steam Safety Valves (MSSVs) Exhaust Piping and associated supports1 2 Unit 1 & 2 Steam Generator Power Operated Relief Valves (PORVs) and associated piping and supports1 2 Unit 1 & 2 Turbine Driven Auxiliary Feedwater (TD AFW) Exhaust (TE) Pipe2 Unit 1 & 2 Control Room Air Ventilation System (CRAVS) Intakes (VC/YC Intakes)2 Unit 1 & 2 Spent Fuel Building (north facing wall)

Notes:

1. The SSCs located in the Unit 1 Exterior Doghouse are not included as they have positive tornado missile protection.

2. Only the portion of the Structure that is not protected from the Design Basis tornado generated missiles are included. The Design Basis tornado generated missiles have a horizontal only projection.

3. Additional details and target areas can be found in Section 3.5.6 Reference 7.