ML20049H288

Chapter 13 to Gessar, Conduct of Operations.

ML20049H288
Person / Time
Site:	05000447
Issue date:	02/12/1982
From:	GENERAL ELECTRIC CO.
To:
References
NUDOCS 8202230070
Download: ML20049H288 (24)

Text

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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O SECTION 13.6

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CONTENTS Section Title Page 13.6 INDUSTRIAL SECURITY 13.6-1 13.6.1 Preliminary Planning 13.6-1 13.6.2 Security Plan 13.6-1 13.6.3 BOP Interface 13.6-1 13.6.3.1 Introduction 13.6-1 13.6.3.2 Design Bases 13.6-1 13.6.3.3 Vital Areas 13.6-2 13.6.3.4 Methods of Access Control 13.6-2 13.6.3.5 Card Reader Door Assignments and Locations 13.6-3 13.6.3.6 Access Control and Security Measures Through Exterior Doors to the Reactor

; Island 13.6-3 N/ 13.6.3.7 Eullet-Resisting Walls and Doors, Security Grills and Screens 13.6-6 13.6.3.8 Balance of Plant Interfaces 13.6-7 TABLES Table Title Page 13.6-1 Card Reader Door Assignments and Locations 13.6-9 h

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GESSAR II 22A7007 l 238 NUCLEAR ISLAND Rev. O t'N 13.6 INDUSTRIAL SECURITY 13.6.1 Preliminary Planning To be provided by Applicant 13.6.2 Security Plan To be provided by Applicant 13.6.3 BOP Interface 13.6.3.1 Introduction The physical protection system of standard plants in which special nuclear material is used is designed to protect against acts of industrial sabotage and theft. The industrial security design

("~N requirements are concerned mainly with separation, isolation, and

-' control of access for vital areas and the capability for detection of inoperability of vital equipment. The capability for detection of inoperability of vital equipment is discussed in Chapter 7.

This section concerns itself with the control of access to areas containing vital equipment.

For the purpose of this section, the " vital equipment" is defined as any equipment, system, device or material, failure, destruction or release of which could directly or indirectly endanger the public health and safety by exposure to radiation. Equipment or systems that would be required to function to protect public health and safety following such failure, destruction, or release are also considered to be vital.

13.6.3.2 Design Bases 3 Security functions described herein are incorporated into the over-all Reactor Island design such that the plant is in compliance with the requirements of 10CFR73.

13.6-1

GESSAR II 22A7007 238 NUCLEAR ISLAND Rc *r . 0 13.6.3.3 Vital Areas Design analysis of the standard Reactor Island plant reveals the following eight areas to be " vital areas":

(1) interior of the Reactor Building; (2) Control and Equipment Room in the Control Building; (3) areas in the Auxiliary Building containing core cooling systems; (4) remote shutdown panel in the Auxiliary Building; (5) Standby Gas Treatment System areas, including ducts; (6) spent fuel pool area in the Fuel Building; (7) Class lE cable tunnels; and (8) Division 1, 2, and 3 diesel generators.

Consideration is given to all essential components of systems in vital areas (e.g., pumps, piping, valves, electrical and instru-mentation systems). In many cases, (previously referenced sec-tions) separation and isolation requirements are consistent with safety requirements which have already been met. lience, access control is considered separately.

l 13.6.3.4 Methods of Access Control Access control of the previously defined vital areas is achieved by door controls, which are equipped with one or more of the fol-lowing security systems:

O 13.6-2

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0

} 13.6.3.4 Methods of Access Control (Continued) d (1) localized alarmed doors; (2) keyed cylinder lock doors; (3) card reader doors; and/or (4) bullet-resistant doors and walls, security grills, and screens.

Additionally, a door status indicator light (red / green) is provided at all secondary containment vestibules.

All doors at card reader locations have surface-mounted alarm switches at the head of the door.

/~'N The Card Reader Door System was designed and installed specifically for the purpose of industrial security.

13.6.3.5 Card Reader Door Assignments and Locations Assignments and locations are outlined in Table 13.6-1 13.6.3.6 Access Control and Security Measures Through Exterior Doors to the Reactor Island In order to prevent unauthorized access from exterior areas and UOP facilities to the Reactor Island, the design of the Reactor Island has provided the following control measures at these desig-nated openings.

O tv 13.6-3

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 13.6.3.6 Access Control and Security Measures Through Exterior Doors to the Reactor Island (Continued)

Reactor, Auxiliary and Fuel Buildings Opening No. Elevation Building fiethod of Control (1) F-25-33 (-) 9 ft 0 in. Fuel Secured from inside and alarmed (railway access from exterior).

(2) F-29-32 (-) 5 ft 3 in. Fuel Secured from inside, alarmed, with card reader (one-way emer-gency exit).

(3) A-16-25 (-) 6 ft 10 in. Auxiliary Secured from inside, localized alarm, with card reader (Central Service Facility to Auxiliary Building).

Control Building Opening No. Elevation Building Method of Control (1) C-19-44 (-) 6 ft 10 in. Control Secured from inside and alarmed (dock to Control Building access corridor).

(2) C-25-44 (-) 11 ft 0 in. Control Secured from inside, localized alarm, with card reader (Control Building to Turbine Building).

O 13.6-4

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0

(~'g 13.6.3.6 Access Control and Security Measures Through Exterior

( ,/ Doors to the Reactor Island (Continued)

Radwaste Building Opening No. Elevation Building Method of Control

(1) W-3-33 (-) 6 ft 10 in. Radwaste Secured from inside, localized alarm, with card reader (exterior to corridor of Rad-waste Building).

(2) W-37-40 (-) 6 ft 10 in. Radwaste Secured from inside, localized alarm, with card reader (exterior to corridor of Rad-waste Building).

(3) W-43-46 (-) 6 ft 10 in. Radwaste Secured from inside and alarmed (exterior to truck loading area).

[ ) (4) W-46-48 (-) 6 ft 10 in. Radwaste Secured from inside

\s / and alarmed (one-way emergency exit).

Diesel Generator Building Opening No. Elevation Building Method of Control (1) G-1-1 (-) 6 ft 10 in. Diesel Secured from inside, Generator localized alarm, with (Div. 1) card Jeader (dock to Div. i generator).

(2) G-9-1 (-) 6 ft 10 in. Diesel Secured from inside, Generator localized alarm, with (Div. 3) card reader (dock to Div. 3 generator).

(3) G-5-1 (-) 6 ft 10 in. Diesel Secured from inside, Generator localized alarm, with (Div. 2) card reader (dock to Div. 2 generator).

(4) G-5-3 (-) 6 ft 10 in. Diesel Secured from inside C,s) Generator and alarmed (one-way (Div. 2) emergency exit).

13.6-5

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 13.6.3.7 Bullet-Resisting Walls and Doors, Security Grills and Screens Design analysis of the Reactor Island has resulted in the Control Room, Control Room equipment area, Division 1 and 4 Cable Room area and Division 2 and 3 Cable Room area being designated as a particu-larly high security zone. Specific precautionary measures have been incorporated into the building design to minimize forceable access to this area. The measures are as follows:

(1) The corridor wall providing access to the four areas mentioned above is constructed of 1/2-in, steel plate supported on 6-in. WF columns, making the wall bullet resistant.

(2) The four access doors off the corridor which opens onto these areas are bullet resistant and able to withstand the impact of a 220-grain, soft-point bullet, with a muzzle velocity of 2410 ft/sec and a muzzle energy of 2830 ft lb ll fired from a 30-06 high-power rifle with a 24-in. barrel at a distance of 15 ft or less.

(3) The HVAC ducts on both the supply and return systems to these areas are equipped with 3/4-in. steel bars installed on 6-in. centers.

(4) The exterior air exhaust systems, located an the build-ing roofs or high on the building sidewalls, are equipped with stainless steel grills. The grills are fabricated of double crimp weave mesh, star wire screen astroloy and installed with one-way and nonremovable fasteners.

O 13.6-6

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O g-'S 13.6.3.8 Balance of Plant Interfaces The following interface information must agree with the applicant's comprehensive description of the physical security program for the plant site:

(1) the site security force; (2) definition and designation of high-security areas of the plant site, including physical barriers and peripheral fencing; (3) access controls to BOP buildings adjoining the Reactor Island; (4) locations of alarm stations; (5) Personnel Badge Control System; O

(6) Card Reader and Key Control System; (7) Security Communications Systems; (8) location of access control points to protected and vital areas; (9) location of parking lots relative to the clear areas adjacent to the physical barriers surrounding protected areas; (10) special features of the terrain that may present special vulnerability probleas; and (11) location of relevant law enforcement agencies and their geographical jurisdictions.

C 13.6-7

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 13.6.3.8 Balance of Plant Interfaces (Continued)

It is also the responsibility of the utility-applicant to prevent unauthorized entry to the Reactor Island portion of the steam tunnel from the Turbine Building.

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O O O Table 13.6-1 CARO READER DOOR ASSIGNMENTS AND LOCATIONS (Continued)

Card Opening Reader No. Elevation Building No.* Location 37 (-) 6 ft 10 in. Radwaste W-3-33 Outside to corridor 38 (-) 6 ft 10 in. Radwaste W-42-44 Truck Loading Area to maintenance area 39 (-) 6 ft 10 in. Radwaste W-37-40 Outside to corridor 40 Radwaste W-8-51 Concrete walk to stairwell w

(+) 9 ft 2 in.

41 (+) 9 ft 2 in. Radwaste W-9-52 Concrete walk to waste filter z

42 (+) 9 ft 2 in. Radwaste W-29-53 Concrete walk to waste filter co g

am y 43 (+) 9 ft 2 in. Radwaste W-39-55 Concrete walk to stairwell Es T 44 (+) 6 ft 10 in. Diesel G-5-1 Dock to Div. 2 diesel generator xx Generator ss

[ Div. 2 yH and 3 g D

45 (-) 6 ft 10 in. Diesel G-9-1 Dock to Div. 3 diesel generator Generator Div. 2 and 3 46 (-) 9 ft 0 in. Fuel F-ll-36 Passage to vestibule 47 (-) 6 ft 10 in. Fuel F-5-24 Hall to vestibule 48 ( + ', 28 ft 6 in. Fuel F-5-52 Hall to vestibule 49 (+) 28 ft 6 in. Fuel F-27-57 Vestibule to equipment area 50 (+) 28 ft 6 in. Auxiliary A-25-64 Fuel Building hall to Auxiliary Building corridor w ww 51 (+) 28 ft 6 in. Auxiliary A-1-51 Fuel Building vestibule to Auxiliary j>

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Table 13.6-1 CARD READER DOOR ASSIGNMENTS AND LOCATIONS (Continued)

Card Opening Reader No. Elevation Building No.* Location 52 (-) 6 ft 10 in. Auxiliary A-24-26 Corridor to cable tunnel 53 (+) 51 ft Fuel F-5-65 Elevator hall to hall 7-1/2 in.

54 (-) 17 ft 0 in. Fuel F-13-19 Access area to FPCC Equipment Room 55 (+) 11 ft 0 in. Control C-27-47 Corridor to Electrical Equipment $

Room 56 (+) 28 ft 6 in. Control C-12-53 Stairwell to equipment area EO s &m

." 57 (+) 28 ft 6 in. Control C-15-61 Stairwell to equipment area yy p xx w mm N *For door opening locations, see Figures 9.5-5 through 9.5-18 (A-101 through A-107, A-lll, mH A-ll2, A-ll5 through A-ll7, A-121 and A-122), k e

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O O O Table 13.6-1 CARD READER DOOR ASSIGNMENTS AND LOCATIONS (Continued)

Card Opening Reader No. Elevation Building No.* Location 37 (-) 6 ft 10 in. Radwaste W-3-33 Outside to corridor 38 (-) 6 ft 10 in. Radwaste W-42-44 Truck Loading Area to maintenance area 39 (-) 6 ft 10 in. Radwaste W-37-40 Outside to corridor W-8-51 Concrete walk to stairwell N 40 (+) 9 ft 2 in. Radwaste 41 (+) 9 ft 2 in. Radwaste W-9-52 Concrete walk to waste filter

=

42 (+) 9 ft 2 in. Radwaste W-29-53 Concrete walk to waste filter co om h 43 (+) 9 ft 2 in. Radwaste W-39-55 Concrete walk to stairwell Er i 44 (+) 6 ft 10 in. Diesel G-5-1 Dock to Div. 2 diesel generator xx Generator ss

[ Div. 2 MH and 3 h O

45 (-) 6 ft 10 in. Diesel G-9-1 Dock to Div. 3 diesel generator Generator Div. 2 and 3 46 (-) 9 ft 0 in. Fuel F-ll-36 Passage to vestibule 47 (-) 6 ft 10 in. Fuel F-5-24 Hall to vestibule 48 (+) 28 ft 6 in. Fuel F-5-52 Hall to vestibule 49 (+) 28 ft 6 in. Fuel F-27-57 Vestibule to equipment area 50 (+) 28 ft 6 in. Auxiliary A-25-64 Fuel Building hall to Auxiliary Building corridor w mw Auxiliary A-1-51 Fuel Building vestibule to Auxiliary O>

51 (+) 28 ft 6 in.

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Table 13.6-1 CARD READER DOOR ASSIGNMENTS AND LOCATIONS (Continued)

Card Opening Reader No. Elevation Building No.* Location 52 (-) 6 ft 10 in. Auxiliary A-24-26 Corridor to cable tunnel 53 (+) 31 ft Fuel F-5-65 Elevator hall to hall 7-1/2 in.

54 (-) 17 ft 0 in. Fuel F-13-19 Access area to FPCC Equipment Room 55 (+) 11 ft 0 in. Control C-27-47 Corridor to Electrical Equipment $

Room g 56 (+) 28 ft 6 in. Control C-12-53 Stairwell to equipment area SO w t~ zn w

, 57 (+) 28 ft 6 in. Control C-15-61 Stairwell to equipment area yy p Nx H HH M mH

• For door opening locations, see Figures 9.5-5 through 9.5-18 (A-101 through A-107, A-lll, A-112, A-ll5 through A-ll7, A-121 and A-122). k e

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GESSAR II 22A7007 238 NUCLEAR ISLAND , Rev. 0

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SUMMARY

TABLE OF CONTENTS Chapter /

Section Title Volume 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT

_ 'l.1 INTRODUCTION 1 1.1.1 Type of Lic,ense Required 1.1.2 Identification of Applicant 1.1.3 Number of Plant Units

, 1.1.4 Description of Location

~. 1.1.5 Type of Nuclear Steam Supply System 1.1.6 Type of Containment i

1.1.7 Core Thermal Power Levels 1.1.8 Scheduled completion and Operation Dates

1.2 GENERAL PLANT DESCRIPTION 1 1.2.1 Principal Design Criteria s

g 1.2.2 Plant Description 1.3 COMPARISON TABLES 1 1.3.1 ~ Comparisons with Similar Facility Designs 1.3.2 Comparisons of Final and Preliminary Information 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS 1 1.4.1 Applicant 1.4.2 Architect Engineer - Nuclear Island Design 1.4.3 Nuclear Steam Supply System Supplier 1.4.4 Turbine-Generator Vendor 1.4.5 Consultants 1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION 1 1.5.1 Current Levelopment Programs 1.5.2 PSAR Commitment Items 1.6 MATERIAL INCORPORATED BY REFERENCE 1 iii

GESSAR II 22A?007 23B NUCLEAR ISLAND ,

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SUMMARY

TABLE OF CONTENTS (C6ntinued)

Chapter / .

Section Title Volume 1.7 DRAWINGS AND GTHER DF,TAILCD INFORMATION 1 1.7.1 Electrica), Instrumentation, and r

Control Drawittgs ,

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1.7.2 Piping and Instrumentation Diagr;lms.

1.7.3 Abbreviations and Symbols 1.8 CONFORMANCE TO NRC REGULATORY GUIDES 1

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1.8.1 Compliance Assessment Method -

1.9 STANDARD DESIGNS 1 l.9.1 Interfaces ,

l.9.2 Exceptians _

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SUMMARY

TABLE OF CONTENTS (Continued)

Chapter /.

Section Title Volume

s. ,2 SITE CilARACTERISTICS

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SUMMARY

l 2.1 GEOGRAPHY AND DEMOGRAPHY l 2.1.1 Site Location and Description 2.1.2 Exclusion Area Authority and Control 2.1.3 Population Distribution s s ,2.2 NEARBY INDUSTRIAL, TRANSPORTATION AND MILITARY FACILITIES 1 s- 2.2.1 Location and Routes 2.2.2 Descriptions 2.2.3 Evaluation of Potential Accidents 2.3 METEOROLOGY l 2.3.1 Regional Climatology 2.3.2 Local Meteorology O)

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2.3.3 Onsite Meteorological Measurements Program 2.3.4 Short-Term Atmospheric Diffusion Estimates 2.3.5 Long-Term Atmospheric Diffusion Estimates 2.4 HYDROLOGIC ENGINEERING 1 2.4.1 Hydrologic Description 2.4.2 Floods 2.4.3 Probable Maximum Flood (PMF) on Streams and Rivers 2.4.4 Potential Dam Failures, Seismically Induced 2.4.5 Probable Maximum Surge and Seiche Flooding 2.4.6 Prcbable Maximum Tsunami Flooding 2.4.7 Ice Effects 2.4.8 Cooling Water Canals and Reservoirs 2.4.9 Channel Diversions

() 2.4.10 Flooding Protection Requirements v

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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O

SUMMARY

TABLE OF CONTENTS (Continued)

O Chapter /

Section Title Volume 2.4.11 Low Water Considerations 2.4.12 Dispersion, Dilution, and Travel Times of Accidental Releases of Liquid Effluents in Surface Waters 2.4.13 Ground Water 2.4.14 Technical Specifications and Emergency Operation Requirements 2.5 GEOLOGY, SEISMOLOGY, AND GEOTECHNICAL Ei1GINEERING 1 2.5.1 Basic Geologic and Seismic Information 2.5.2 Vibratory Ground Motion 2.5.3 Surface Faulting 2.5.4 Stability of Subsurface Materials and Foundations 2.5.5 Stability of Slopes 2.5.6 Embankments and Dams e

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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O

SUMMARY

TABLE OF CONTENTS (Continued)

Chapter /

Section Title Volume 3 DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS 3.1 CONFORMANCE WITH NRC GENERAL DESIGN CRITERIA 2 3.1.1 Summary Description 3.1.2 Criterion Conformance 3.2 CLASSIFICATION OF STRUCTURES, COMPONENTS, AND SYSTEMS 2 3.2.1 Seismic Classification 3.2.2 System Quality Group Classifications 3.2.3 System Safety Classifications 3.2.4 Quality Assurance 3.2.5 Correlation of Safety Classes with Industry Codes 3.3 WIND AND TORNADO LOADINGS 2 f-- 3.3.1 Wind Loadings

\m/ 3.3.2 Tornado Loadings 3.3.3 BOP Interface 3.3.4 References 3.4 WATER LEVEL (FLOOD) DESIGN 2 3.4.1 Flood Protection 3.4.2 Analytical and Test Procedures 3.4.3 BOP Interface 3.4.4 References 3.5 MISSILE PROTECTION 2 3.5.1 Missile Selection and Description 3.5.2 Structures, Systems, and Components to be Protected from Externally Generated Missiles 3.5.3 Barrier Design Procedures 3.5.4 BOP Interface 3.5.3 References O

Vii

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O

SUMMARY

TABLE OF CONTENTS (Continued)

Chapter /

Section Title Volume 3.6 PROTECTION AGAINST DYNAMIC EFFECTS ASSOCIATED WITH THE POSTULATED RUPTURE OF PIPING 2 3.6.1 Postulated Piping Failures in Fluid Systems Inside and Outside of Containment 3.6.2 Determination of Break Locations and Dynamic Effects Associated with the Postulated Rupture of Piping 3.6.3 References 3.7 SEISMIC DESIGN 3 3.7.1 Seismic Input 3.7.2 Seismic System Analysis 3.7.3 Seismic Subsystem Analysis 3.7.4 Ceismic Instrumentation 3.7.5 BOP Interface 3.7.6 References 3.8 DESIGN OF SEISMIC CATEGORY I STRUCTURES 3 3.8.1 Concrete Containment 3.8.2 Steel Containment 3.8.3 Concrete and Steel Internal Structures of Steel Containment 3.8.4 Other Seismic Category I Structures 3.8.5 Foundations 3.8.6 BOP Interface 3.9 MECHANICAL SYSTEMS AND COMPONENTS 4 3.9.1 Special Topics for Mechanical Components 3.9.2 Dynamic Tc. sting and Analysis 3.9.3 ASME Code Class 1, 2 and 3 Components, Component Supports, and Core Support Structures 3.9.4 Control Rod Drive System 3.9.5 Reactor Pressure Vessels Internals 3.9.6 Inservice Testing of Pumps and Valves 3.9.7 BOP Interface 3.9.8 References viii

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0

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SUMMARY

TABLE OF CONTENTS (Continued)

Chapter /

Section Title Volume 3.10 SEISMIC QUALIFICATIONS OF SEISMIC CATEGORY I INSTRUMENTATION AND ELECTRICAL EQUIPMENT (INCLUDING HYDRODYNAMIC EFFORTS) 5 3.10.1 Seismic Qualification Criteria (Including Hydrodynamic Loads) 3.10.2 Methods and Procedures for Qualifying Electrical Equipment and Instrumentation 3.10.3 Methods and Procedure of Seismic Analysis or Testing of Supports of Electrical Equipment and Instrumentation (Including Hydrodynamic Loads) 3.10.4 Seismic Qualification Tests and Analyses (Including Hydrodynamic Loads) 3.11 ENVIRONMENTAL DESIGN OF SAFETY-RELATED

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MECHANICAL AND ELECTRICAL EQUIPMENT 3.11.1 Equipment Identification and 5

Environmental Conditions 3.11.2 Qualification Tests and Analyses 3.11.3 Qualification Results 3.11.4 Lous of Ventilation 3.11.5 Estimated Chemical and Radiation Environment APPENDIX 3A SEISMIC SOIL-STRUCTURE INTERACTION ANALYSIS OF THE NUCLEAR ISLAND 5 APPENDIX 3B CONTAINMENT LOADS 6,7 APPENDIX 3C COMPUTER PROGRAMS USED IN THE DESIGN OF SEISMIC CATEGORY I STRUCTURES 8 APPENDIX 3D ANALYSIS OF RECIRCULATION MOTOR AND PUMP UNDER ACCIDENT CONDITIONS 8 APPENDIX 3E DESCRIPTION OF SAFETY RELATED MECHANICAL AND ELECTRICAL EQUIPMENT 8 APPENDIX 3F DYNAMIC BUCKLING CRITERIA FOR CONTAINMENT VESSEL 8 O

O ix

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GESSAR II 12A7007 238 NUCLEAR ISLAND Rev. O

SUMMARY

TABLE OF CONTENTS (Continued)

Chapter /

Section Title Volume APPENDIX 3G PIPE FAILURE ANALYSIS 8 APPENDIX 3H EFFECT OF CONCRETE ANNULUS BELOW ELEVATION (-) 5 FT., 3 IN. ON SEISMIC DESIGN LOADS AND BUILDING RESPONSES 8 9

O X

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0

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SUMMARY

TABLE OF CONTENTS (Continued)

Chapter /

,Soction Title Volume 4 REACTOR 4.1

SUMMARY

DESCRIPTION 9 4.1.1 Reactor Vessel 4.1.2 Reactor Internal Components 4.1.3 Reactivity Control Systems 4.1.4 Analysis Techniques 4.1.5 References 4.2 FUEL SYSTEM DESIGN 9 4.2.1 General and Detailed Design Base 4.2.2 General Design Description 4.2.3 Design Evaluations 4.2.4 Testing and Inspection 4.2.5 Operating and Developmental

- Experience ,

(,j 4.2.6 References 4.3 NUCLEAR DESIGN 9 4.3.1 Design Bases 4.3.2 Deccription 4.3.3 Analytical Methods 4.3.4 Changes 1.3.5 References 4.4 THERMAL - HYDRAULIC DESIGN 9 4.4.1 Design Basis 4.4.2 Description of Thermal Hydraulic Design of the Reactor Core 4.4.3 Description of the Thermal and Hydraulic Design of the Reactor Coolant System 4.4.4 Evaluation 4.4.5 Testing and Verification 4.4.6 Instrumentation Requirements 4.4.7 References xi

. -- . . - - . . - . - ~.

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TABLE OF CONTENTS (Con tint.ed ) ,

Chapter /

Section Title Volume 4.5 REACTOR MATERIALS 9 4.5.1 Control Rod System Structural Materials 4.5.2 Reactor Internal Materials 4.5.3 Control Rod Drive Housing Supports

! 4.6 PUNCTIONAL DESIGN OF REACTIVITY CONTROL SYSTEMS 9 4.6.1 Information for Control Rad Drive System (CRDs) 4.6.2 Evaluations of the CRDs 4.6.3 Testing and Verification of the CRDs 4.6.4 Information for Combined Performance of Reactivity Syste.as 4.6.5 Evaluation of Combined Performance 4.6.6 References APPENDIX 4A CONTROL ROD PATTERNS AND ASSOCIATED i POWER DISTRIBUTION FOR TYPICAL BWR 9 4A.1 Introduction 4A.2 Power Distribution Strategy i 4A.3 Results of Core Simulation Studies 4A.4 Glossary of Terms 4A.5 References O

xii

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{%

v Chapter /

Section Title Volume 5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 5.1 SUMMAMY DESCRIPTION 10 5.1.1 Schematic Flow Diagram 5.1.2 Piping and Instrumentation Diagram 5.1.3 Elevation Drawing 5.2 INTEGRITY OF REACTOR COOLANT PRESSURE BOUNDARY 10 5.2.1 Compliance with Codes and Code Cases 5.2.2 Overpressure Protection 5.2.3 Reactor Coolant Pressure Houndary Materials 5.2.4 Inservice Inspection and Testing of Reactor Coolant Pressure Boundary 5.2.5 Reactor Coolant Pressure Boundary and ECCS Systera Leakage Detection System

(_/ 5.2.6 Referenceu 5.3 REACTOR VESSEL 10 5.3.1 Reactor Vessel Materials 5.3.2 Pressure / Temperature Limits 5.3.3 Reactor Vessel Integrity 5.3.4 References 5.4 COMPONENT AND SUBSYSTEM DESIGN 10 5.4.1 Reactor Recirculation Pumps 5.4.2 Steam Generators (PWR) 5.4.3 Reactor Coolant Piping 5.4.4 Main Steam Line Flow Restrictors 5.4.5 Main Steam Line Isolation System 5.4.6 Reactor Core Isolation Cooling System (RCIC) 5.4.7 Residual Heat Removal System 5.4.8 Reactor Water Cleanup System 5.4.9 Main Steam Lines and Feedwater Piping

() 5.4.10 Pressurizer xiii

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O Chapter /

Scction Title Volume 5.4.11 Pressurizer Relief Valve Discharge System 5.4.12 Valves 5.4.13 Safety and Relief Valves 5.4.14 Component Supports 5.4.15 References O

i 1

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NJ Chapter /

Section Title Volume 6 ENGINEERED SAFETY FEATURES 6.0 GENERAL 11 6.1 ENGINEERED SAFETY FEATURE MATERIALS 11 6.1.1 Metallic Materials 6.1.2 Organic Materials 6.2 CONTAINMENT SYSTEMS 11 6.2.1 Containment Functional Design 6.2.2 Containment Heat Removal System 6.2.3 Secondary Containment Functional Design 6.2.4 Containment Isolation System 6.2.5 Combustible Gas Control in Containment 6.2.6 Containment Leakage Testing

() 6.2.7 6.2.8 Suppression Pool Makeup System References 6.3 EMERGENCY CORE COOLING SYSTEMS 11 6.3.1 Design Bases and Summary Description 6.3.2 System Design 6.3.3 ECCS Performance Evaluation 6.3.4 Tests and Inspections 6.3.5 Instrumentation Requirements 6.3.6 References 6.4 HABITABILITY SYSTEMS 11 6.4.1 Design Basis 6.4.2 System Design 6.4.3 Systems Operational Procedures 6.4.4 Design Evaluations 6.4.5 Testing and Inspection 6.4.6 Instrumentation Requirements 6.4.7 Nuclear Island / BOP Interface O

XV

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Chapter /

O Section Title Volume 6.5 FISSION PRODUCT REMOVAL AND CONTROL SYSTEMS 11 6.5.1 Standby Gas Treatment System (SGTS)

C.S.2 Containment Spray Systems 6.5.3 Fission Product Control Systems 6.5.4 Ice Condensers as a Fission Product Control System 6.5.5 References 6.6 INSERVICE INSPECTION OF CLASS 2 AND 3 COMPONENTS 11 6.6.1 Components Subject to Examination 6.6.2 Accessibility 6.6.3 Examination Techniques and Procedures 6.6.4 Inspection Intervals 6.6.5 Examination Categories and Requirements 6.6.6 Evaluation of Examination Results 6.6.7 System Pressure Tests 6.6.8 Augmented Inservice Inspection to Protect Against Postulated Piping Failures 6.7 MAIN STEAM POSITIVE LEAKAGE CONTROL SYSTEM (MSPLCS) 11 6.7.1 Design Bases 6.7.2 System Description 6.7.3 System Evaluation 6.7.4 Inspection and Testing 6.7.5 Instrumentation Requirements 6.8 PNEUMATIC SUPPLY SYSTEM ll 6.8.1 Design Bases 6.8.2 System Description 6.8.3 System Evaluation 6.8.4 Inspection and Testing Requirements 6.8.5 Instrumentation Requirements APPENDIX 6A IMPROVED DECAY HEAT CORRELATION FOR LOCA ANALYSIS ll xvi

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N )]

Chapter /

Section Title Volume 7 INSTRUMENTATION AND CONTROL SYSTEMS

7.1 INTRODUCTION

(All Systems) 12 7.1.1 Identification of Safety-Related Systems 7.1.2 Identification of Safety and Power Generation Criteria 7.2 REACTOR PROTECTION (TRIP) SYSTEM (RPS) 12 7.2.1 Description 7.2.2 Conformance Analysis 7.3 ENGINEERED SAFETY FEATURES SYSTEM, INSTRUMENTATION AND CONTROL 13 7.3.1 Description 7.3.2 Analysis

-HPCS -Shield Building

-w ^^"" "8 * "9

) -ADS x_ / -Secondary Contain-

-LPCS ment Isolation

-RHR/LPCI -Primary Containment c

-CRVICS Isolation LCS

-MSPLCS -Standby Power

-RHR/ Containment -D-G Support Systems Spray -Essential Service

-RHR/ Suppression Pool Water Cooling -ESF Area Cooling

-Suppression Pool -Pneumatic Supply Makeup

-CB Atmospheric

-Combustible Gas Control Control

-CB Chilled Water

-SGTS 7.4 SYSTEMS REQUIRED FOR SAFE SHUTDOWN 14 7.4.1 Description 7.4.2 Analysis

-RCIC -RHR/ Shutdown Cooling es -SLC -Remote Shutdown

's_)

xvii

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Section Title Volume 7.5 SAFETY-RELATED DISPLAY INSTRUMENTATION 14 7.5.1 Description 7.5.2 Analysis

-Nuclonet Control -BOP Benchboard Consol

-Supervisory Moni-

-Standby Information toring Console P"" 1

-Display Control

-Rx Core Cooling BB System 7.6 ALL OTIIER INSTRUMENTATION SYSTEMS REQUIRED FOR SAFETY 14 7.6.1 Description 7.6.2 Analysis 7.6.3 Additional Design Considerations Analyses

! 7.6.4 References

-Neutron Monitoring -FPCCS

-Process Radiation -DW/ Containment

! Monitoring Vacuum Relief i

-Refueling Interlocks -Vent & Pressure Control

-Leak Detection

- ^

-Rod Pattern Control

- uppression Pool

-HP/LP System U 'U Interlock EU'#

Monitoring

-Recirculation Pump Trip l

O XViii l

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O Chapter /

Section Title Volume 7.7 CONTROL SYSTEMS NOT REQUIRED FOR SAFETY 14 7.7.1 -Description 7.7.2 Analysis 7.7.3 References

-RPV tnstrumentation -Leak Detection

-Rod Control & -Rod Block Trip Information -Fire Protection

-Recirculation Flow -Drywell Chiller &

Control Cooling

-Feedwater Control -Plant Instrument Air

-Performance Moni- -Neutron Monitoring toring System

-Radwaste 7.8 NI/ BOP INTERFACES 14 7.8.1 Essential Service Water (Supply)

O- System Instrumentation and Controls 7.8.2 Diesel Generator Fuel Oil Transfer System APPENDIX 7A I&C ELEMENTARY DIAGRAMS 15 i

i O

XiX t

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O Chapter /

Section Title Volume 8 ELECTRIC POWER

8.1 INTRODUCTION

16 8.1.1 Utility Grid Description 8.1.2 Onsite Electric Power System 8.1.3 Design Bases 8.2 OFFSITE POWER SYSTEM 16 8.2.1 Description 8.2.2 Analysis 8.2.3 Nuclear Island - BOP Interface 8.3 ONSITE POWER SYSTEMS 16 8.3.1 AC Power Systems 8.3.2 DC Power Systems 8.3.3 Fire Protection of Cable Systems O

O XX

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Chapter /

Section Title Volume 9 AUXILIARY SYSTEMS 9.1 FUEL STORAGE AND HANDLING 17 9.1.1 New Fuel Storage (High Density) 9.1.2 Spent Fuel Storage (High Density) 9.1.3 Fuel Pool Cooling and Cleanup System 9.1.4 Fuel-Handling System 9.2 WATER SYSTEMS 17 9.2.1 Essential Service Water (ESW) System 9.2.2 Closed Cooling Water System 9.2.3 Demineralized Water Makeup System 9.2.4 Potable and Sanitary Water Systems 9.2.5 Ultimate Heat Sink 9.2.6 Condensate Storage Facilities and Distribution Syctem

(,j 9.2.7 Plant Chilled Water Systems 9.2.8 Heated Water Systems 9.2.9 Nuclear Island / BOP Interfaces 9.3 17 PR_OCESS AUXILIARIES 9.3.1 Compressed Air Systems 9.3.2 Process Sampling System 9.3.3 Floor and Equipment Drainage Systems 9.3.4 Chemical and Volume Control System (PWR) 9.3.5 Standby Liquid Control System I

9.4 AIR CONDITIONING, HEATING, COOLING AND VENTILATION SYSTEM 17 9.4.1 Control Room HVAC System 9.4.2 Fuel Building HVAC Systen l

l 9.4.3 Auxiliary Building HVAC Systems t

9.4.4 Turbine Building Area Ventilation System 9.4.5 Reactor Building HVAC System xxi l .

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Chapter /

Section Title Volume 9.4.6 Radwaste Building HVAC System 9.4.7 Diesel-Generator Buildings HVAC Systems 9.5 OTHER AUXILIARY SYSTEMS 17 9.5.1 Fire Protection System 9.5.2 Communications Systems 9.5.3 Lighting Systems 9.5.4 Diesel-Generator Fuel Oil Storage and Transfer System 9.5.5 Diesel-Generator Cooling Water System 9.5.6 Diesel-Generator Starting Air System 9.5.7 Diesel Engine Lubrication System 9.5.8 Diesel Generator Combustion Air Intake and Exhaust System 9.5.9 Suppression Pool Cleanup System 9.5.10 Nuclear Island - BOP Interface APPENDIX 9A FIRE HAZARD ANALYSIS 18 xxii

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Section Title Volume 10 STEAM AND POWER CONVERSION SYSTEM 10.1

SUMMARY

DESCRIPTION 19 10.2 TURBINE GENERATOR 19 10.2.1 Design Bases Functional Limitations by Design or Operational Charac-teristics of the Reactor Coolant System 10.2.2 System Description 10.2.3 Turbine Disk Integrity 10.2.4 Evaluation 10.3 MAIN STEAM SUPPLY 19 10.4 OTHER FEATURES OF STEAM AND POWER CONVERSION SYSTEM 19 10.4.1 Main Condensers 10.4.2 Condenser Air Removal System 7-s

\s_) 10.4.3 Main Condenser Evacuation System 10.4.4 Turbine Bypass System 10.4.5 Circulating Water System 10.4.6 Condensate Cleanup System 10.4.7 Condensate and Feedwater System 10.4.8 Steam Generator Blowdown System (PWR) 10.4.9 Auxiliary Feedwater System (PWR) o XXiii

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Chapter /

Section Title Volume 11 RADIOACTIVE WASTE MANAGEMENT 11.1 SOURCE TERMS 19 11.1.1 Fission Products 11.1.2 Activation Products 11.1.3 Tritium 11.1.4 Fuel Fission Production Inventory and Fuel Experience 11.1.5 Process Leakage Sources 11.1.6 Radwaste System 11.1.7 Radioactive Sources in the Gas Treatment System 11.1.8 Source Terms for Component Failures 11.1.9 Other Releases 11.1.10 References 11.2 LIQUID WASTE MANAGEMENT SYSTEM 19 11.2.1 Design Basis 11.2.2 System Descriptions 11.2.3 Estimated Releases 11.3 GASEOUS WASTE MANAGEMENT SYSTEMS 19 11.3.1 Design Bases 11.3.2 Main Condenser Steam Jet Air Ejector Low-Temp RECHAR System Description 11.3.3 RECHAR System Operating Procedure 11.3.4 Radioactive Releases 11.3.5 References 11.4 SOLID RADWASTE SYSTEM 19 11.4.1 Design Bases 11.4.2 System Description O

xxiv

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)

Chapter /

Section Title Volume 11.5 PROCESS AND EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING SYSTEMS 19 11.5.1 Design Bases 11.5.2 System Description 11.5.3 Effluent Monitoring and Sampling 11.5.4 Process Monitoring and Sampling 11.5.5 Calibration and Maintenance 11.6 OFFSITE RADIOLOGICAL MONITORING PROGRAM 19 (v

\

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Section Title Volume 12 RADIATION PROTECTION 12.1 ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS REASONABLY ACHIEVABLE (ALARA) 19 12.1.1 Policy Considerations 12.1.2 Design Considerations 12.1.3 Operational Considerations 12.2 RADIATION SOURCES 19 12.2.1 Contained Sources 12.2.2 Airborne Radioactive Material Sources 12.2.3 References 12.3 RADIATION PROTECTION DESIGN FEATURES 19 12.3.1 Facility Design Features 12.3.2 Shielding 12.3.3 Ventilation 12.3.4 Area Radiation and Airborne Radioactivity Monitors 12.3.5 References 12.4 DOSE ASSESSMENT 19 12.5 HEALTH PHYSICS PROGRAM 19 O

xxvi

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O Chapter /

Section Title Volume 13 CONDUCT OF OPERATIONS 19 i

4 i

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Section Title Volume 14 INITIAL TEST PROGRAM 14.1 TEST PROGRAM 20 14.1.1 Administrative Procedures (Testing) 14.1.2 Administrative Procedures (Modifications) 14.1.3 Test Objectives and Procedutes 14.1.4 Fuel Loading and Initial Operation 14.1.5 Administrative Procedures (System Operation) 14.2 SPECIFIC INFORMATION TO BE INCLUDED IN SAFETY ANALYSIS REPORTS 20 14.2.1 Summary of Test Program and Objectives 14.2.2 Organization and Staffing 14.2.3 Test Procedures 14.2.4 Conduct of Test Program 14.2.5 Review, Evaluation and Approval of Test Results 14.2.6 Test Records 14.2.7 Conformance of Test Programs with Regulatory Guidec 14.2.8 Utilization of Reactor Operating and Testing Experiences in the Develop-ment of Test Program 14.2.9 Trial Use of Plant Operating and Emergency Procedures 14.2.10 Initial Fuel Loading and Initial Criticality 14.2.11 Test Program Schedule 14.2.12 Individual Test Descriptions O

xxviii

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Section Title Volume 15 ACCIDENT ANALYSES 15.0 GENERAL 21 15.0.1 Analytical Objective 15.0.2 Analytical Categories 15.0.3 Event Evaluation 15.0.4 Nuclear Safety Operational Analysis (NSOA) Relationship 15.0.5 References 15.1 DECREASE IN REACTOR COOLANT TEMPERATURE 21 15.1.1 Loss of Feedwater Heating 15.1.2 Feedwater Controller Failure -

Maximum Demand 15.1.3 Pressure Regulator Failure - Open 15.1.4 Inadvertent Safety / Relief Valve

/g Opening

\x/ /

15.1.5 Spectrum of Steam System Piping Failures Inside and Outside of Containment in a PWR 15.1.6 Inadvertent RHR Shutdown Cooling Operation 15.1.7 References 15.2 INCREASE IN REACTOR PRESSURE 21 15.2.1 Pressure Regulator Failure - Closed 15.2.2 Generator Load Rejection 15.2.3 Turb.ine Trip 15.2.4 MSLIV Closures 15.2.5 Loss of Condenser Vacuum 15.2.6 Loss of Offsite AC Power 15.2.7 Loss of Feedwater Flow 15.2.8 Feedwater Line Break 15.2.9 Failure of RHR Shutdown Cooling O

G XXiX

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Section Title Volume 15.3 DECREASE IN REAC1gR COOLANT SYSTEM FLOW RATE 21 15.3.1 Recirculation Pump Trip 15.3.2 Recirculation Flow Control Failure -

Decreasing Flow 15.3.3 Recirculation Pump Seizure 15.3.4 Recirculation Pump Shaft Break 15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 21 15.4.1 Rod Withdrawal Error - Low Power 15.4.2 Rod Withdrawal Error at Power 15.4.3 Control Rod Maloperation (System Malfunction or Operator Error) 15.4.4 Abnormal Startup of Idle Recirculation Pump 15.4.5 Recirculation Flow Control with Increasing Flow 15.4.6 Chemical and Volume Control System Malfunctions 15.4.7 Misplaced Bundles Accident 15.4.8 Spectrum of Rod Ejection Assemblies 15.4.9 Control Rod Drop Accident (CRDA) 15.5 INCREASE IN REACTOR COOLANT INVENTORY 21 15.5.1 Inadvertent HPCS Startup 15.5.2 Chemical Volume Control System Malfunction (or Operator Error) 15.5.3 BWR Transients Which Increase Reactor Coolant Inventory O

XXX

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< Chapter /

}i Section Title Volume j 15.6 -DECREASE IN REAPTOR COOLANT INVENTORY 21 i

i 15.6.1 Inadvertent Safety / Relief Valve l Opening

15.6.2 Failure of Small Lines Carrying Primary Coolant Outside Containment i 15.6.3 Steam Generator Tube Failure 15.6.4 Steam System Piping Break Outside Containment 15.6.5 Loss-of-Coolant Accidents (Resulting from Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary) - Inside Containment

!' 15.6.6 Feedwater Line Break - Outside Containment 15.7 RADIOACTIVE RELEASE FROM SUBSYSTEMS AND COMPONENTS 21 j 15.7.1 Radioactive Waste System Leak or Failure

} 15.7.2 Liquid Radioactive System Failure 15.7.3 Postulated Radioactive Releases Due to Liquid Radwaste Tank Failure 15.7.4 Fuel-Handling Accident 15.7.5 Spent Fuel Cask Drop Accidents I APPENDIX 15A PLANT NUCLEAR SAFETY OPERATIONAL

! ANALYSIS 21

! APPENDIX 15B BWR/6 GENERIC ROD WITHDRAWAL ERROR ANALYSIS 21 l

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TADLE OF CONTE!h'S (ccntinued)

Chapter / > \

Section Titla Vo7 fume 16 STANDARD TECilNICnL SPECIFICATIONS FOR GENERAL i ELECTRIC BOILING WATER REACTORS 16.1 DEFINITIONS 22' i 16.1.1 Action ' s .

\

16.1.2 Averaga Planar Exposure -

16.1.3 Average P} anar Linear IIcat Generation Rate .

16.1.4 Channel cdlibration ,

16.1.5 Channel Chech ,

16.1.6 Channel Functicqa1 Test 16.1.7 Core Alteration ., s s ' "

16.1.8 Critical Power Ratio .

~

16.1.9 Dose Equivalent I-131 ,

16.1.10 E-Average Dirintegration Energy s 16.1.11 Emergency Core Cooling System (ECC9) , ,

Response Time 16.1.12 Frequency Notation s 16.1.13 Identified Leakage 16.1.14 Isolation System Response Time ,

16.1.15 Limiting Control Rod Pattern 16.1.16 Linear IIcat Generation Rate s s _

16.1.17 Logic System Functional Test 16.1.18 Maximum Total Peaking Factor s 16.1.19 Minimum Critical Power Ratio

16.1.20 Operable - Operability , s 16.1.21 Operational condition (Conditiont 16.1.22 Physics Test s 16.1.23 Pressure Boundary Leakage, ,

16.1.24 Primary containmt.it Integrity ,

s 16.1.25 Rated Thermal Power 16.1.26 Reactor Protection System Responce. '

Time 16.1.27 Recirculation Pump Trip System ' '

Response Time xxxii -

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'i

^T Chapter /

1 Section Title Volume

, g 16.1.28 Reportable occurrence t

16.1.29 Rod Density

. 16.1.30 Secondary Containment Integrity 16.1.31 Shutdown Margin 16.1.32 Staggered Test Basis 16.1.33 Thermal Power 16.1.34 Total Peaking Factor

,  % , 16.1.35 Unidentified Leakage i6.2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM i SETTINGS 22 1- 16.2.1 Safety Limits 16.2.2 Limiting Safety System Settings 16.B2 SAFETY LIMITS 22 16.B2.1 Bases 16.B2.2 Limiting Safety System Settings 16.3/4 LIMITING CONDITIONS FOR OPERATION AND

' SURVEILLANCE REQUIREMENTS 22 16.3/4.0 Applicability 16.3/4.1 Surveillance Requirements 16.3/4.2 Power Distribution Limits 16.3/4.3 Instrumentation 16.3/4.4 Reactor Coolant System 16.3/4.5 Emergency Core Cooling Systema 16.3/4.6 Containment Systems 16.3/4.7 Plant Systems 16.3/4.8 Electrical Power Systems 16.3/4.9 Refueling Operations 16.3/4.10 Special Test Exceptions O

XXXiii

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Chapter /

Section Title Volume 16.B3/4.0 Applicability 16.B3/4.1 Reactivity Control Systems 16.B3/4.2 Power Distribution Limits 16.B3/4.3 Instrumentation 16.B3/4.4 Reactor Coolant System 16.B3/4.5 Emergency Core Cooling System 16.B3/4.6 Containment Systems 16.D3/4.7 Plant Systems 16.B3/4.8 Electrical Power Systems 16.B3/4.9 Refueling Operations 16.B3/4.10 Special Test Exceptions 16.5 DESIGN FEATURES 22 16.5.1 Site 16.5.2 Containment 16.5.3 Reactor Cote 16.5.4 Reactor Coolant System 16.5.5 Meteorological Tower Location 16.5.6 fuel Storage 16.5.7 Component Cyclic or Transient Limit 9

XXXiV

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Section Title Volume i

17 QUALITY ASSURANCE 17,1 QUALITY ASSURANCR DURING DESIGN AND CONSTRUCTION 22 i

! 17.2 OUALITY ASSURANCE DURING Ti!E OPERATING PIIASE 22 i

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