ML20148L455

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Suppl 5 to Gap Appl for Constr Permit & Operating Lic for Subj Facil
ML20148L455
Person / Time
Site: Vogtle  Southern Nuclear icon.png
Issue date: 11/03/1978
From: Ehrensperger W
GEORGIA POWER CO.
To:
Shared Package
ML20148L213 List:
References
NUDOCS 7811200259
Download: ML20148L455 (45)
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Text

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BEFORE THE ,

0 UNITED STATES NUCLEAR REGULATORY COMMISSION NRC Docket Nos. 50-424, 50-425 In the Matter of GEORGIA POWER COMPANY SUPPLEMENT 5 TO APPLICATION FOR LICENSE UNDER THE ATOMIC ENERGY ACT OF 1954 AS AMENDED FOR l ALVIN W. V0GTLE NUCLEAR PLANT UNITS 1, 2 The Applicant, Georgia Power Company, hereby supplements its Application for a Construction Permit and Operating License, originally submitted on August 1,1972, by the addition of supplementary material attached hereto. r This supplement reflects the conclusions reached in our March 3,1978 meeting with the NRC Staff on the main steam and feedwater system design, and includes the 1.5 scaling factor used in the design of deeply-embedded i Category I structures.

O BY: W. E. Ehrensperge ,

Senior Vice President Sworn to and subscribed before me, this h day of November,1978 wv Notary /Pubild h[ V

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V My Commi:ttm Expires November 24,1979

. Georgia Pcwer Company Post Ottice Box 4545 230 Peacntree Street, N W.

At'anta. Georgia 30302 Telephone 404 522 6060 0 W. E. Ehrensperger Senior Vice Prestcent Power Supply November 17, 1978 Director of Nuclear Reactor Regulation ATTN: Roger S. Boyd, Director Division of Project Management U.S. Nuclear Regulatory Commission O Washington, D.C. 20555 NRC DOCKET NUMBERS 50-424 AND 50-425 CONSTRUCTION PERMIT NUMBERS CPPR-108 AND CPPR-109 ALVIN W. V0GTLE NUCLEAR PLANT-UNITS 1 AND 2 SUPPLEMENT 5 TO APPLICATION

Dear Mr. Boyd:

Georgia Power Company hereby files three (3) signed copies and sixty (60) conformed copies of Supplement 5 to its application for a Construction Permit'and Operating License for the Alvin W. Vogtle Nuclear Plant, Units 1 and' 2.

k This supplement reflects the conclusions reached in our March 3,1978 meeting with the NRC Staff on the main steam and main feedwater system design for the Vogtle Nuclear Plant. It also includes the 1.5 scaling factor that will be used to multiply the " envelope in-structure response spectra" generated for each deeply-embedded seismic Category I structure.

Yours truly, 408 W. E. Ehrensperg cc: R. A. Thomas s G. F. Trowbridge, Esq.

D. E. Dutton

' J. A. Bailey L. T. Gucwa B. L. Lex I. S. Mitchell, III H. A. Sindt x.

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DEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION e NRC Docket Nos. 50-424, 50-426 In the Matter of GEORGIA POWER COMPANY Q SUPPLEMENT 5 TO APPLICATION FOR LICENSE UNDER THE ATOMIC ENERGY ACT OF 1954

. AS AMENDED FOR ALVIN W. V0GTLE NUCLEAR PLANT O

V '

UNITS 1, 2 i

The Applicant, Georgia Power Company, hereby supplements its Application for a Construction Permit and Operating License, originally submitted on August 1,1972, by the addition of supplementary material attached hereto.

Thi, supplement reflects the conclusions reached in our March 3,1978 meeting with the NRC Staff on the main steam and feedwater system design, and includes the 1.5 scaling factor used in the design of deeply-embedded Category I structures.

BY: W. E. Ehrensperge ,

Senior Vice President M

Sworntoandsubscribedbeforeme,thisfy day of November,1978 bb-QzA Notary / Pub 1p p

My Commt:t!:n Exptros November 24,1979

lNSTRUCTION SHEET SUPPLEMENT NO. 5 m ALVIN W. VOGTLE NUCLEAR PLANT PRELIMINARY SAFETY ANALYSIS REPORT DO NOT REMOVE EXISTING WHITE PAGES l O l i

Replace Table of Contents pages 82 y and S2 vi with S5 v and S5 vi; and S2 xvii ;

and S2 xviii with S5 xvii throug'.1 S5 xviiia.

Insert Table of Contents pages S5 3-1 through S5 3-via ahead of page 3-1.

Replace Table of Contents pages S3 3-xv and S3 3-xvi with S5 3-xv through S5 3-xvia.

Insert Table of Contents pages S5 10-1 through S5 10-iv ahe i of page 10-1.

Insert Chapter 3 pages S5 3.1-5 and S5 3.1-6 ahead of page 3.1-5.

Insert Chapter 3 pages SS 3.2-1 and S5 3.2-2 ahead of page 3.2-1.

Insert Chapter 3 pages S5 3.2-3 through S5 3.2-4 ahead of page 3.2-3.

Insert Chapter 3 pages S5 3.5-31 and S5 3.5-32 ahead of page 3.5-31.

Insert Chapter pages SS 3.6-3 through S5 3.6-4a ahead of page 3.6-3.

Insert Chapter 3 pages S5 3.6-Sa through S5 3.6-7 ahead of page 3.6-Sa.

Replace Chapter 3 page S4 3.7-47 with S5 3.7-47.

Insert Chapter 10 pages S5 10.3-1 and SS 10.3-2 ahead of page 10.3-1.

Insert Chapter 10 pages SS 10.3-5 through S5 10.3-8 ahead of page 10.3-5.

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  • TABLE - OFiCONTENTS L (Continued) ' '

v() Section ' Title Page

2. 5. 2 ' VIBRATORY. GROUND MOTION:

2.5-21 '!

2. 5. 3 SURFACE / FAULTING' '2.5-47

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.G 2.5.4 STABILITY 1OF SUBSURFACE MATERIALS i 2.5-47' 2 '. 5 . 5 ' SLOPE STABILITY 2.5-50 2.5.6 . REFERENCES l2.5 ( 2.5.7 BIBLIOGRAPHY 2.5-53 2A GEOLOGY 2A-1 2B SEISMIC SURVEY 2B-1 2C SUBSURFACE AND FOUNDATIONS' 2C-1 .,

.2D. SELECTION OF' TEMPERATURE DIFFERENCE CATEGORIES TO DEFINE AVERAGE PASQUILL' .g STABILITY CATEGORIES BASED ON DATA s - 1

. COLLECTED AT THE FARLEY SITE- 2D-1

( ) 2E ON-SITE METEOROLOGICAL' DATA 2E-1' ,

3 DESIGN CRITERIA - STRUCTURES, COMPONENTS, . . .

EQUIPMENT, AND. SYSTEMS' 3.1-1 3.1 CONFORMANCE TO AEC GENERAL' DESIGN CRITERIA 3.1-1 r

3.1.1 CRITERION 1 - QUALITY STANDARDS AND RECORDS 3.1-1 l

3.1.2 CRITERION 2 - DESIGN BASES FOR PROTEC--

-TION AGAINST NATURAL PHENOMENA 3.1-3

() '3.1.3 3.1.4 CRITERION.3--' FIRE PROTECTION CRITERION 4 - ENVIRONMENTAL AND MISSILE 3.1 '

DESIGN BASES S5 3.1-6 lSSE 3.1.4 CRITERION.4 - ENVIRONMENTAL AND-MISSILE

(). DESIGN' BASES L3.1-6 3.1.5- - CRITERION 5 - SHARING OF STRUCTURES, - ,

SYSTEMS AND COMPONENTS 3.1-8

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S5 v i POST-CONSTRUCTION PERMIT . SUPPLEMENTARY" INFORMATION - NOVEMBER 17, 1978 y -..,,-m-g-aw,-, w m- v -- - , .g ..-w , . . " . , 4,e . . - - - , .w-- ,v.wy'y ay r- -wr

VNP TABLE OF CONTENTS (Continued)

Section Title Page O 3.1.6 CRITERION 10 - REACTOR DESIGN 3.1-9 3.1.7 CRITERION 11 - REACTOR INHERENT PROTECTION 3.1-11 3.1.8 CRITERION 12 - SUPPRESSION OF REACTOR POWER OSCILLATIONS 3.1-12 3.1.9 CRITERION 13 - INSTRUMENTATION AND CONTROL 3.1-13 3.1.10 CRITERION 14 - REACTOR COOLANT PRESSURE BOUNDARY 3.1-17 3.1.11 CRITERION 15 - REACTOR COOLANT SYSTEM DESIGN 3.1-19 3.1.12 CRITERION 16 - CONTAINMENT DESIGN 3.1-21 3.1.13 CRITERION 17 - ELECTRICAL POWER SYSTEMS 3.1-22 3.1.14 CRITERION 18 - INSPECTION AND TESTING OF ELECTRICAL POWER SYSTEMS 3.1-25 3.1.15 CRITERION 19 - CONTROL ROOM 3.1-26 3.1.16 CRITERION 20 - PROTECTION SYSTEM FUNCTIONS 3.1-28 3.1.17 CRITERION 21 - PROTECTION SYSTEM RELIABILITY AND TESTABILITY 3.1-30 3.1.18 CRITERION 22 - PROTECTION SYSTEM INDEPENDENCE 3.1-33 3.1.19 CRITERION 23 - PROTECTION SYSTEM FAILURE MODES 3.1-36 3.1.20 CRITERION 24 - SEPARATION OF PROTECTION AND CONTROL SYSTEMS 3.1-37 3.1.21 CRITERION 25 - PROTECTION SYSTEM REQUIRE- ll MENTS FOR REACTIVITY CONTROL MALFUNCTIONS 3.1-40 3.1.22 CRITERION 26 - REACTIVITY CONTROL SYSTEM REDUNDANCY AND CAPABILITY 3.1-42 O

S5 vi POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - NOVEMBER 17, 1978

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VNP TABLE OF CONTENTS (Continued)

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) Section Title Page j

9.4.7 DIESEL GENERATOR BUILDING 9.4-10 9.5 OTHER AUXILIARY SYSTEMS 9.5-1

[') 9.5.1 FIRE PROTECTION SYSTEMS 9.5-1

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9.5.2 COMMUNICATION SYSTEMS 9.5-4a 9.5.3 LIGHTING SYSTEMS 9.5-5

() 3.8-3 Liner Plate-Material Properties and Characteristics as Assumed in Design 3.8-39 3.8-4 Internal Structure 3.8-71 3.8-5 Internal Structure 3.8-72 3.8-6 Material Properties and Characteristics

!as Assumed in. Design. '3.8-80 )

3.8-7 Category I structure-Steel Factor. Load Combinktion. Working Stress Design, 3.8-93

. ai 3.8-8 Category I' Structure Concrete Factor 3 -Load Combination Ultimate Strength Design 3.8-94 I 3.8-9 Concrete Design Temperatures of Normal Shutdown for the Coolin'g' Towers 3.8-102 O ,

S5 3-xv POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - NOVEMBER 17, 1978-

VNP TABLE OF CONTENTS (Continued)

LIST OF TABLES Table Title Page 3.8-10 Ambient Temperature and Temperature Distributions in the Soil 3.8-102 3.8-11 Cool'ng i Tower Stress Analysis, Structure and Equipment Weight 3.8-105 3.8-12 Cooling Tower Stress Analysis, Earth Pressure at Rest 3.8-106 3.8-13 Cooling Tower Stress Analysis, Liquid Pressure 3.8-107 3.8-14 Cooling Tower Stress Analysis, Operating Thermal Load 3.8-108 3.8-15 Cooling Tower Stress Analysis, Live Loads 3.8-109 3.8-16 Cooling ' cower Stress Analysis, Accident Thermal Load 3.8-110 ll 3.8-17 Cooling Tover Stress Analysis Safe Shutdown Earthquake - Horizontal Load 3.8-111 3.8-18 Cooling Tower Stress Analysis Safe Shutdown Earthquake - Vertical Load 3.8-112 3.9-1 Design Loading Combination for ASME Code Class 2 and 3 Components and Supports Supplied by Westinghout., 3.9-11 3.9-2 Design Loading Combinations for ASME Section III Code Class 2 and 3 Components and Supports Outside the Westinghouse Scope of Supply 3.9-12 3.9-3 Stress Criteria for ASME Code Class 2 and Class 3 Piping (Code Case 1606) 3.9-13 3.9-4 Stress Criteria for ASME Code Class h 2 and Class 3 Vessels 3.9-14 3.9-5 Stress Criteria for ASME Code Class 2 and Class 3 Inactive Pumps 3.9-15 O

l SS 3-xvi l POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - NOVEMBER 17, 1978 l

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TABLE.OF CONTENTS (Continued)

LIST OF TABLES--

. Table Title Page'-

O 3.9-6 Stress Criteria-for ASME III Class-2 and 3 Active Pumps-

' 3.9-16.  :!

l 3.9-7' Stress Criteria for?ASME Code Class .

2 and Class 3 '. Valves (Active and' In-Active) 3.9-17. .l

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-POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - NOVEMBER 17,.1978 1 1

. VNP TABLE OF CONTENTS CHAPTER 10

.x_J Section Title Page 10 STEAM AND POWER CONVERSION SYSTEM 10.1-1 i,

) 10.1

SUMMARY

DESCRIPTION 10.1-1 .

10.2 TURBINE-GENERATOR 10.2-1 10.2.1 DFSIGN BASES 10.2-1

(,/ 10.

2.2 DESCRIPTION

OF TURBINE-GENERATOR EQUIPMENT 10.2-2 10.2.2.1 Turbine '10.2-2 10.2.2.2 Steam Extraction Connections 10.2-2 10.2.2.3 Generator 10.2-2a 10.2.2.4 A,utomatic Controls 10.2-3 10.2.2.5 Other Protective Systems 10.2-6

,- 10.2.2.6 Instrumentation 10.2 '

10.2.3 EVALUATION OF TURBINE-GENERATOR AND RELATED STEAM HANDLING EQUIPMENT 10.2-7 10.2.3.1 Summary Discussion of Anticipated Operating Concentrations of Radioactive Containments in the System 10.2-7 10.2.3.2 Access to the Turbine Area 10.2-8 10.3 MAIN STEAM SUPPLY SYSTEM S5 10.3-1 10.

3.1 DESCRIPTION

S5 10.3-1 O S5 10.3-1 S5

( I 10.3.2 PIPING COMPONENTS AND SYSTEMS tJ 10.3.3 VALVES S5 10.3-2 10.3.3.1 Safety Valves SS 10.3-2 rh

( ) 10.3 MAIN STEAM SUPPLY SYSTEM 10.3-1 10.

3.1 DESCRIPTION

10.3-1

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\~' S5 10-i POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - NOVEMBER 17, 1978

VNP l l

r TABLE OF CONTENTS (Continued)

Section Title Page O

10.3.2 PIPING COMPONENTS AND SYSTEMS 10.3-1 10.3.3 VALVES 10.3-2 10.3.3.1 Safety Valves 10.3-2 10.3.3.2 Atmospheric Power Relief Valve 10.3-3 10.3.3.3 Main Steam Line Isolation Valves S5 10.3-5 10.3.3.4 Valve Testing SS 10.3-7 S5 10.3.3.5 Pipe Testing S5 10.3-8 10.3.4 SYSTEM TESTING S5 10.3-8 10.3.3.3 Main Steam Line Isolation Valves 10.3-5 10.3.3.4 Valve Testing 10.3-7 10.3.4 SYSTEM TESTING 10.3-8 10.4 OTHER FEATURES OF STEAM AND POWER CONVERSION SYSTEM 10.4-1 10.4.1 MAIN CONDENSERS 10.4-1 10.4.1.1 Description 10.4-1 10.4.1.2 Radioactivity Considerations 10.4-2 10.4.1.3 Control Functions and Hydrogen Buildup 10.4-2 10.4.2 MAIN CONDENSERS EVACUATION SYSTEM ,

10.4-3 10.4.2.1 Description 10.4-3 10.4.2.2 Calculations of Radioactivity 10.4-3 10.4.3 TURBINE GLAND SEALING SYSTEM 10.4-3 10.4.4 TURBINE STEAM BYPASS SYSTEM 10.4-4 10.4.4.1 Design Bases 10.4-4 10.4.4.2 Description 10.4-4 S5 10-11 O

POST-CONSTRUCTTON PERMIT SUPPLEMENTARY INFORMATION - NOVEMBER 17, 1978

VNP-t N TST(Continu'e d)

' TABLE OF~ CONTE' x

Section- ' Title Page'  !

210.4.4.3') Evaluation 10.4-5

-10.4.5 (CIRCULATING WATER SYSTEM '10;4-5L

10. 4 '. 5. l ' Design Bases 10.4-5' 10.4.~5.2 Description 10.4 :r g '10.4.5.3 Plant Cooling-Wator' System. 10s4-6a-V' 10.4.5.4 : Makeup,' Dilution,;and ChemicaldTreatment 10.4-6a-

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10.4.5.5.' Emergency Cooling and Physical' Interaction- 10.4-8 10.4.6' CONDENSATE CLEANUP SYSTEM 10.4-8c 10.4.7 CONDENSATE AND FEEDWATER SYSTEM 10'4-8c 1014;7.1 Description- 10.4-8c. :f 10.4.7.2 -Steam Supply System 10.4-8d' 10.4.7.3 Auxiliary Feedwater System 10.4-8h

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10.4.8 SYSTEM GENERATOR BLOWDOWN PROCESSING SYSTEM '10. 4 10.4.8.1 Design Bases- 10.4-15 -j i

10.4.8.2 System Description 10.4-16  !

10.4.8.3-1 Safety Evaluation' 10.4-19 10.4.8.4 Tests and Inspections 10.4-20

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VNP TABLE OF CONTENTS (Continued)

LIST OF TABLES Table Title Page 10.1-1 Major Steam and Power Conversion Equipment Summary Description 10.1-3 10.1-2 Component Design Parameters (Per Unit Basis) 10.1-5 10.3-1 Main Steam Safety Valves (Per Steam Generator) 10.3-4 10.4-1 Steam Generator Blowdown Processing System O Codes and Classifications 10.4-10 10.4-2 Steam Generator Blowdown Processing System Major Component Parameters 10.4-11 10.4-3 Tower Blowdown Composition, Per Unit 10.4-8a 10.4-4 Single Failure Analysis Auxiliary Feedwater System 10.4-14a LIST OF FIGURES Figure Title 10.1-1 Heat Balance Diagram Turbine - VWO (Not Guaranteed) 10.1-2 Heat Balance Diagram Turbine Guarantee 10.1 ' P&I Diagram Main Steam System 10.1-4 P&I Diagram Steam Generator System (Sheets 1, 2 and 3) 10.1-5 P&I Diagram Extraction Steam System 10.1-6 P&I Diagram Condensate and Feedwater System 10.1-7 P&I Diagram Circulating Water System (Sheets 1 and 2) 10.1-8 P&I Diagram Turbine Plant Cooling Water System ,

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S5 10-iv POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - NOVEMBER 17, 1978 i I

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VNP Firefighting'. systems'are designed to' assure that..their' rupture-q.

b or. inadvertent operation will not significantly impair" systems l

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' important Lto . safety (see subsection 9.5) .

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The fire protection system' consists of a reliable, partially

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automatic system designed and installed in accord with the.

requirements'of the National Fire Protection Association, the O American Insurance Association, Nuclear Mutual Limited, and the /

l applicable ~ local-codes and' regulations. l 1

1 The fire protection system is provided with test hose valves-for periodic testing. All equipment is accessible for periodic inspection. The fire protection system is described in section 9.5.

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VNP 3.1.4 CRITERION 4 - ENVIRONMENTAL AND MISSILE DESIGN BASES Structures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents. These structures, systems, and components shull be appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit.

RESPONSE

Structures, systems, and components important to safety are designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant, main steam line break, and main SS feedwater break accidents. Criteria are presented in section 3.5, 3.6, and 3.9.2.7 while environmental conditions are presented in section 6.2.

These structures, systems, and components are appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit. I i

l Chapter 7 lists the motors, instrumentation, and associated l cables of protection and safety features systems located inside the containment. It gives the design requirements in terms of S5 3.1-6 O

POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - NOVEMBER 17, 1978

VNP 3.2 CLASSIFICATION OF STRUCTUR5S , COMPONENTS, AND SYSTEMS

,r') 3.2.1 SEISMIC CLASSIFICATION

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A.two-level system is used for the seismic classification of structures, components, and systems other than Westinghouse's scope of the Alvin W. Vogtle Nuclear Plant (VNP) :

l tT Category I Structures, Components, and Systems l Y-sl Category.II Structures, Components, and Systems l For Westinghouse design responsibility components refer to RESAR-3, sections 3.2, 3.9 and 5.2. In addition to the I'i standard systems described in RESAR-3 above, refer to tables

(_)

6.2-12, 9.3-2, 9.1-2, and 10.4-1 of the PSAR for the safety classification of the Westinghouse components in the containment spray system, boron recycle system, spent fuel pit cooling system and steam generator blowdown system.

The classification of pipes, valves, and fittings is shown on the piping and instrumentation diagrams in the appropriate PSAR sections.

3.2.1.1 Definitions

,s Seismic Category I structures, components, and systems are

(' ') defined in accordance with USAEC Safety Guide No. 29 as those necessary to assure:

The integrity of the reactor coolant pressure boundary.

The capability to shut down the reactor and maintain it in a safe shutdown condition.

The capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the guideline exposures of 10 CFR Part 100.

! )

(_/ Category I structures, components, and systems are designed to withstand the appropriate seismic loads and other applicable loads without loss of function. Category I structures are sufficiently isolated or protected from Category II structures to ensure that their integrity is maintained at all times.

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\_ Category II structures, components, and systems are those whose failure would not result in the release of significant (3

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S5 3.2-1 POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - NOVEMBER 17, 1978

VNP radioactivity and would not prevent reactor shutdown. All equipment not specifically listed as Category I is included as Category II. Specifically, all non-Category I systems, equipment and components that are installed in Category I structures are carefully examined to determine the degree, if any, of detrimental effect on safety related systems, should failure of the non-Category I equipment, systems or components occur. Where the detrimental effect can be shown to effect safe shutdown equipment then the non-Category I item is either upgraded to Category I or separated by distance or barricade.

The failure of Category II structures, components and systems may interrupt power generation.

3.2.1.2 Category I Structures Containment Structure Enclosure Building

  • Auxiliary Building Control Building Fuel Handling Building Nuclear Service Cooling Towers and Basins Diesel Generator Building S5l Auxiliary Feeduater Building Condensate Storage Tanks Refueling Water Storage Tank Reactor Makeup Water Storage Tank Pipe Tunnels and Electrical Cable Tunnels as follows:

A. Pipe Tunnels

1. Auxiliary Feedwater Piping
2. Nuclear Service Cooling Water Piping
3. Refueling Water Piping O
  • For structural integrity of the EB, refer to paragraph 3.8.4.4.1.

SS 3.2-2 POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - NOVEMBER 17, 1978

. VNP 3;2.1~.3.4 Component, Cooling;. Water System- .;

.,Cg ~~ Component-Cooling W5ter Heat Exchanger '(Tube ~ Side)

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9:%,.

(Shell ~ Side) ' ,

Component Cooling Water Pump- l Component Cooling Water Surge Tank O<

3.2.1.3.5' Engineering Safety'. Features .

a 1

Radiatiion Monitoring System l Containment-Isolation-System 3.2.1.3.6 Conventional Mechanical Section ]

Main Steam System Main Steam Piping (from Steam Generator up to and including five-way restraint) lSS Safety' Valves Isolation Valves

' Atmospheric Dump Valve Steam Blowdown and Sampling Piping (Steam Gener .

ator to and Including Isolation Valve) 3.2.1.3.7 Auxiliary and Main Feedwater. System. lS5 Auxiliary-Feedwater Pump -

Isolation Valves Piping'(from-Steam Generator up..to and including S5 five-way. restraint) 3.2.1.3'.8 Nuclear Service Cooling Water System Cooling Tower and Holdup Basin O

l O 4- .., .S5 3.2-3: . .

POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - NOVEMBER- 17,.1978

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VNP

'i Makeup-Pump Circulating Pump Piping, Valves, and Fittings' 3.2.1.3.9 Diesel Generation System Diesel Generator Diesel Fuel Storage Tank Diesel Fuel Day Tank l Diesel Generator Air Tank Diesel Fuel Transfer Pump Diesel Fuel Filter 3.2.1.3.10 Heating, Ventilating, and Air Conditioning A. Control Room Emergency' Supply Air Charcoal Filter Emergency Supply Air Absolute Filter Air Conditioning Unit B. Control Building Class I Switchgear Ventilation Supports Ventilation Fans O

O SS 3.2-4 g

POST-CONSTRUCTION. PERMIT SUPPLEMENTARY INFORMATION - NOVEMBER 17, 1978

VNP O

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.l POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - NOVEMBER 17, 1978

VNP 3.5.3.3 Tornado Missiles The characteristics of potential tornado missiles are given in paragraph 3.5.2.4.1.and tables 3.5-6 and 3.5-7.

Using the allowables and analytical techniques given in Appendix 3N, it is determined that a 2'-0" thick reinforced concrete wall could resist the impact effects of all the listed missiles without perforation, spalling or failure due to structural response.

3.5.3.4 Protection of Structures, Systems and Components Against External Missiles l The Category I structures listed below are analyzed for missile damage protection capabilities and for compliance with the protection criteria of this section. The bases for. selection of these structures for analysis are: (1) they house or service systems and components required for the safe shutdown of the reactor and to maintain it in a safe shutdown condition, (2) if damaged, they could cause uncontrolled release of radioactivity that could result in potential offsite exposure comparable to the guideline exposures of 10 CFR 100.

Containment excluding enclosure building Auxiliary building including penetration rooms and' MSIV/MFIV area S5 Control building including penetration rooms and MSIV/MFIV area Nuclear service cooling towers Condensate storage tank S5l Auxiliary feedwater pump building and piping tunnel Fuel Handling Building Diesel Generator Building and Fuel Facility To effect a safe shutdown during a tornado, all essential l

systems and components are protected with adequate barriers designed against external missiles which may be generated in a tornado as discussed in paragraph 3.5.2.4.1. The control, sensing, power supply and piping associated with safety train oriented systems in areas outside of the plant structures are located below plant grade.

O SS 3.5-32 POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - NOVEMBER 17, 1978

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may be considered (onL a reasonably conservative

(~'

i basis) in analyzing 1the accident seguence ifDit can be shown that unacceptable plant 1 conditions'will not' ensue or be, approached. -

C.- When analyzing.the damage inflicted following the

. postulated 1 pipe failure or; jet impingement, jr~N conservative assumptions are to be employed.

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D.

Pipe whipfis the. result ofLthe formation of a plastic

-hinge at the point of highest. moment in a piping run-that:has experienced.a circumferential pipe break.

It is assumed,that a whippin'g pipe has.the potential'

./~'t to move perpendicular to the' break orientation unless  ;

(_)" that' pipe has been restrained against motion'in that particular direction. A' whipping pipe ~is considered-to contain sufficient' energy to rupturet an impacted pipe of smaller nominal' pipe size and. lighter wall thickness.

E. In assessing accident effects, the loading cor.ditions of a. pipe or-branch run.inuterms of internal pressure and temperature, prior to postulated rupture, should be the normal and upset conditions associated with reactor operation.

F. Class 2 and 3 systems on the'Vogtle Plant in which

. , pipe breaks are postulated:to occur arefsupported such thatLthe following mechanistic break is not possible. The' mechanistic break assumed is a circumferential break wherein the pipe ends translate

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laterally with respect to.each other up to-a distance-of 1 diameter. To an< adjacent piece-of equipment, the-mechanistic break could potentially cause a full area blowdown load and-concurrently a full area thrust load. For~the Vogtle Plant the only relative motion of pipe-ends after a guillotine' rupture is due to thermal growth.

G. Limited breaks' (i'.e. , single-area breaks) are postulated for those portions-of the high-energy main steam'and main O. feedwater fluid piping systems between anchors adjacent-to containment isolation valves (including any rigid.

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connection to the containment penetration) . The. '

SS limited break. zone shall not exceed 60 feet, piping runs shall'be reasonably straight, and:the ASME. Boiler. .,

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and Pressure Vessel Code,Section III, safety Class 2, O forged anchor shall be~provided at the terminal-point ,

3 S5 3.6-3 I

POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - NOVEMBER 17,.1978

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VNP downstream of the isolation valves. Limited breaks are postulated for only those portions of the fluid piping system extending from the forged anchor, adjacent to the outside isolation valve, to the rigid pipe connection at the containment penetration.

These limited breaks are postulated provided the following design stress limits are met:

1. The maximum stress ranges, as calculated by the sum of equations (9) and (10) in S5 Paragraph MC-3652, ASME Code,Section III, considering normal and upset plant con-ditions, (i.e., sustained loads, occasional loads, and thermal expansion) and an OBE event, do not exceed 0.8 (1.2Sh+ A)
  • The maximum stress, as calculated by equation (9) in Paragraph NC-3652, under the loadings resulting from a postulated piping failure of fluid system piping beyond these portions of piping, does not exceed 1.8S h*
2. To prevent a breach of containment, the piping l run downstream of the valve outside the I containment and upstream of the penetration inside the containment is restrained such that i excessive pipe loads are not transmitted to the penetration, penetration piping and containment isolation valve following a postulated pipe ,

failure.

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3. For all high energy line penetrations, provisions l are made to permit full volumetric in-service i inspection of all longitudinal and circum-  !

ferential welds between the penetration and the l isolation valve.

4. When break locations are not postulate per 1.

above, longitudinal and circumferential welds on 4 the piping or branch runs are limited to a practical minimum. Transitions between different wall thicknesses are made with a gentle slope to diminish stress discontinuities. Piping restraints are attached to the piping in such a manner as not to cause excessive stress discontinuities.

O S5 3.6-4 POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - NOVEMBER 17, 1978

l VNP

-s H. The measures taken-for the protection of structures,

/) systems, and components important to. safety should I

\_) not preclude the conduct of inservice examinations of ASME Class 2 and 3 pressure-retaining components'as l required by the rules of ASME Boiler and Pressure.

Vessel Code - Section XI, " Inservice Inspection of

,_ - Nuclear Power Plant Components."

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\_) - I. In piping systems (moderate energy), the failure for the purpose.of assessing accident effects is assumed to be a " critical crack." .The critical crack postulated for these evaluations is defined to have

,3- an effect opening of'an area equivalent to the product.of one-half the pipe internal diameter and (N ') one-half the pipe wall thickness.

3.6.2.1 Design Basis Piping Break Criteria!for Postulated Breaks Inside tae Containment The criteria for postulated breaks inside the containment is defined for ASME Code Class 2 and 3 components as follows (this criteria also applies to Class 1 branch lines) :

3.6.2.1.1 Postulated Break Locations

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( , ) The locations for which breaks should be postulated ta occur for ASME Code Class 2'and 3 piping are based on the regions of.

piping runs or branch runs with the greatest. potential for failure under intensities associated with specified seismic events and normal operational plant conditions. On the basis of high stress intensities and related considerations which indicate a greater probability of failure relative to straight pipe, piping break locations should be postulated to occur at the terminal ends of the piping run or branch run. Similarly, breaks should be assumed to occur at any intermediate locations between terminal ends of piping runs or branch runs that exhibit stress intensities above conservatively derived limits

,.s based on the stress levels actually existing in the fluid

( ) system piping. The limits selected on this basis are N/ elastically calculated primary plus secondary stress intensities per Regulatory Guide 1.46.

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S5 3.6-4a POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - NOVEMBER 17, 1978

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- C. Longitiudinal'or circumferential'b'reaks.are postulated .,

. P vperpendicular to the#directionLof the maximun -

calculated-stress rather5than'having,both longitudinal ~.and circumferential-breaks at the same tlocation'.- ,

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. 3.6~.2~1.3-. Pipe Whip: Restraint ^ Measures  !

Measures for restraint against pipe whipping-as afresult of'the ,

design basis breaks postulated.to; occur;at the-locations' l specified in paragraph 3.6.~2sl.1 need not..be'provided-for' piping where anyj one of '.the.' following ; applies: . H O

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SS 3.6-Sa POST-CONSTRUCTION-PERMIT SUPPLEMENTARY.INFORMATION --NOVEMBER 17,.'1978

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' VNP' l A. The piping is physically separated'(or isolated) from l other piping or components by protective barriers or is restrained from whipping by plant design features such as concrete encasement; B. Following a single. break, the unrestrained pipe movement of either end of the ruptured pipe in any direction about a plastic hinge formed at the nearest pipe whip restraint cannot damage any structure, system, or component important to safety; C. The energy associated with the whipping pipe can be demonstrated to be insufficient to impair the. safety function of any structure, system, or component important to safety to an unacceptable level; D. Both of the following piping system conditions are met:

1. The design temperature is 200 F or less, and
2. -The design pressure is 275 psig or less.

3.6.2.2 Design Basis Piping Break Criteria for Postulated Breaks Outside the Containment' The design basis piping break criteria for breaks outside the containment 'is an follows:

3.6.2.2.1 High-Energy-Fluid Systems A. For piping systems that by plant arrangement and layout are isolated by remote location from structures, systems, and components important to safety, pipe breaks need not be postulated provided the requirements of paragraph 3.6.2.G are satisfied.

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B. For the main steam and feedwater piping systems that are enclosed in suitably designed concrete structures or compartments to protect structures, systems, and i components important to safety, pipe breaks should be postulated at the following locations in each piping or branch run within each protective structure:

1. A nonmechanistic single-area break of the l S5 largest line for area pressurization and environmental qualification of equipmen l

l SS 3.6-6 POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - NOVEMBER 17, 1978

2 ..

p VNP 74 2.: A minimum of one break of'the largest branch: lSS.

'( ,f .run within each' protective structure or

compartment at'a location that results in the

. maximum loading from the impact of the postulated -

ruptured pipe.and jet discharge force on wall, floor,.and roof of the structure or' compartment,

. ('T including internal pressurization, and taking 3 ,) intocaccount any piping. restraints provided to:

limit pipe motions.

C. ~ Break locations for the main; steam piping; located ino the main l steam tunnel shall be postulated only at both "T terminal points, two intermediate points, and in k'J proximity to the auxiliary.feedwater system if all

.the-following: conditions are met:  ;

1. Stress criteria in accordance with ANSI B31.1.0 assuming seismic Category 1 loads. SS
2. Tunnel, supports,'and restraints dynamically analyzed for SSE loads. >
3. Full volumetric inspection of all circum- '*

ferential and longitudinal. welds, either in.

the shop or in the field, of.all main steam piping in proximity to the diesel generator.

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building.

.A single pipe failure in'(seismic Category I or non- ,

seismic Category I) piping system'is postulated to occur as the initiating event (accident)'. Such an event is not concurrent with any other incident, 1 accident, or natural. phenomena. In ahigh energy ,l system (operating pressure.and temperature greater

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S5 3.6-7

-POST-CONSTRUCTION' PERMIT' SUPPLEMENTARY INFORMATION - NOVEMBER 17, 1978

VNP s- The variation of shear modulus with shear strain for the lower

( ) sand stratum is shown in figure 3.7-22C. This is based on the standard curve proposed by Seed and Idriss. The shear modulus corresponding to' low shear strain levels.(10-4 % or less) for this stratum was computed based on the meas" red shear wave velocity of 1800 ft/sec. To account for the variation in the

-( measured shear wave velocity with the depth, a range of. shear q ,) moduli with upper-bound values equal to 1.5 times the mean values and lower-bound values equal to the mean values divided by 1.5, will be used in the analysis.

S3 As discussed in paragraph 3.7.1.3.1, the damping val'ues for the compacted sand backfill, the clay marl bearing stratum and the

[ lower sand ;tratum shewn in figures 3.7-21, 3.7-22, and

\_ 3.7-22A wi11 be used the analysis.

In general, the soil perties are nonlinear in character. An iterative process is ;d to obtain equivalent linear properties which are strain dep 4ent. The methods generally used for such an analysis are inc12aed in the computer program FLUSH.

The generation of design time history motions is described in section 3.7.1.2. This ground motion is defined for the free field and applied at the finished grade level (El 220'-0") of the site.

The time history at the base of the idealized soil profile is 7 s, obtained through deconvolution analysis of'the design time t

'-'j history specified at finished grade level, using appropriate soil properties. The time history, thus obtained, is applied at the base of the soil-structure interaction system, with appropriate soil properties for soil-structure interaction analysis. The resulting time history response will be used to generate the in-structure response spectra at selected floor elevations. The analysis is performed with consideration S4 given to the variation of soil parameters as indicated above using appropriate cut-off frequencies such that the acceleration profile in the free field is realistic. .The " envelope in-structure response spectra" shall be' developed by enveloping the response spectra obtained by considering the variation of soil properties.

Response spectra corresponding to the free field' time history

. motions calculated at the elevations of Category I structural foundations are generated. Considering the variation of soil properties, " envelope response spectra" for each Category I foundation level is developed. A scaling factor of 1.5 has

/~ been established such that-when the " envelope response spectra"

( )} curves are multiplied by the scaling factor, the 60% design spectra curves.will be essentially enveloped. The " envelope S5 in-structure response spectra" curves generated for each deeply embedded seismic Category I structure will be multiplied by the scaling factor of 1.5 to obtah the " design in-structure f-'g response spectra" curves.

'LI SS 3.7-47 POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - NOVEMBER 17, 1978

VNP 10.3 MAIN STEAM SUPPLY SYSTEM 10.

3.1 DESCRIPTION

Steam at the outlet from the four steam generators, a total of 15,155,582 lb/hr at 985 psia, 542.8 F, and 0.25 percent moisture is routed in the main steam piping to the turbine at a total pressure drop of 20 psi at the turbine guaranteed steam O flow condition. As percent power demand increases, the steam generator outlet steam pressure decreases from 1107 psia, 556 F near zero percent power to 985 psia, 542.8 F at 100 percent power. Consequently, the turbine inlet pressure decreases as the load increases.due to this characteristic of the steam g

generators.of the NSSS.

The main steam piping headers conduct the total steam flow from the steam generators to the turbine; one line size is 38 inches, the other'is 44 inches as shown in figure 10.1-3 and 10.1-4. Each main steam line is sized to provide a required equal balanced steam pressure at the inlet to the turbine stop valves. The pipe sizes selected are based on different pipe lengths of the unsymmetrical steam piping layout with the present reactor building orientation to the turbine building.

The design pressure-temperature rating of the main steam piping is 1185 psig 600 F, which matches the design pressure-temperature rating of the steam generator secondary feedwater-steam side.

10.3.2 PIPING COMPONENTS AND SYSTEMS In evaluating the design of the main steam system-piping, the following components and/or systems are considered:

A. The attachment of the main steam piping to the steam generators takes into account the allowable nozzle loading movements and stresses as specified by the steam generator manufacturer for all four units operating or with units out of service.

k B. The routing, loading, and support of the main steam piping through the penetration section of the reactor containment building for various unit operating conditions are determincd.

C. The loading and support of the main steam line O isolation valves just outside the containment building are determined.

O S5 10.3-1 POST-CONSTM1CTION PERMIT SUPPLEMENTARY INFORMATION - NOVEMBER 17, 1978

VNP D. The' sequential blowing effect of the steam safety valves and atmospheric. power relief valves located in  ;

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the main steam piping between the containment building and the main steam line isolation valves and  ;

the resultant valve reaction thrusts on the piping is i determined.

E. The routing and support of interconnected steam l piping.to the reheaters, the turbine bypass system to l the condenser, and the main steam generator turbine driven feedwater pumps for normal, upset, and emergency conditions are determined.

'F. The interfacing of the main steam lines with the. main l turbina stop/ control valves is determined, taking _

l into account allowable loadings, movements, and stresses as specified by the turbine-generator .

manufacturer for normal conditions, on line stop i valve-testing conditions, and upset conditions.

G. For the foregoing systems and components relating to the main steam piping, the requirements are calculated as outlined by the specific applicable codes and classes as shown on the respective system diagrams.

Steam is conducted from each steam generator in 26-inch lines through the containment building, the lines being anchored at the containment wall. The lines are designed in accord with 55 subsection 3.6.2.G.2, up to and including the anchor forging at the end of the MSIV/MFIV area. The isolation valves and the spring-loaded safety and atmospheric power relief valves are located outside the containment building.

10.3.3 VALVES 10.3.3.1 Safety Valves Five spring-loaded safety valves in each 36-inch steam outlet line as shown on figure 10.1-4 are designed and selected in accord with the requirements of the ASME Boiler and Pressure' Vessel Code,Section III Nuclear Vessels. The lowest safety valve is set at 1175 psig, the design pressure of the steam generator (minus 10 psi piping loss). The highest safety valve setting is 105 percent of the design pressure at 1234 psig.

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1 9

S5 10.3-2 POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - NOVEMBER 17, 1978

VNP valve is suitable for remote adjustment of the relieving r's pressure. Control is automatic based on steam line pressure.

Local manual operators are provided in case of complete loss of

('") automatic control.

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/~'3/ 10.3.3.3 Main Steam Line Isolation Valves l

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Two automatically operated bi-directional main steam isolation i valves will be installed in the steam line from each steam S5 i generator outside of the reactor containment. The isolation valve system will prevent the uncontrolled blowdown of more f~]

(_,/

than one steam generator for. postulated steam line ruptures in accordance with subsection 3.6.2.G.2. This valve system ful- SS fills the requirements of General Design Criteria 57 of 10 CFR 50.

The following design bases apply:

A. Safety General Design Criteria 57 of Appendix A.to 10 CFR 50.

B. Design maximum valve closure time of 10 seconds. lSS

,_s C. Valves will be designed and manufactured to ASME l Section III, Class 2.

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D. Valves will be designed for seismic Category I.

If the break is within the containment, steam is discharged into the containment. The other steam generators would act to feed steam through the interconnecting piping to reverse the flow into the damaged line if reverse flow protection were not provided to prevent discharge of more than one steam generator.

Closure of the automatic bi-directional valves within 10 seconds l SS from receipt of signal prevents a reverse flow of steam.

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SS 10.3-5 POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - NOVEMBER 17, 1978

VNP If the break is downstream of the isolation valves, the S5l automatic isolation valves would close within 10 seconds from receipt of the initiating signal.

W For the condition of a steam generator tube rupture, the isolation valves serve to limit the total radioactive release to the environment by isolating the damaged steam generator.

A flow restrictor is installed within each steam generator outlet nozzle and is primarily used to limit the steam flow release in case of a main steam line rupture. Steam flow is measured using the pressure drop between the steam generators and pressure taps in the main steam lines downstream of the flow restrictors.

Three-inch lines connected upstream of the isolation valves (see figure 10.1-4) in the steam outlet line from three steam generators provide steam to the turbine-driven auxiliary feedwater pumps. This assures a source of steam to the turbine-driven auxiliary pump when the steam generators are isolated and are producing steam from reactor decay heat. The 3-inch steam piping to the auxiliary turbine feed pump is designed to the ASME Code Section III Class 2.

Main steam piping is cross-connected downstream of the main steam line isolation valves. The bypass valve around the main S5 steam isolation valves will be designed for Nuclear Safety Class 2 and seismic Category I. Branch piping from the cross-connections provides steam to the reheaters, gland steam sealing system, steam air ejectors, the steam generator feedwater pump turbines, and the turbine bypass steam to the condensers (see figure 10.1-3).

Main steam piping in the main steam tunnel downstream of the anchor forging at the end of the MSIV/MFIV area is designed in accord with the Power Piping Code ANSI-B31.1.0.

55 Anchor forgings shall be provided at both terminal points for main steam and feedwater piping located in the main steam tunnel.

In addition, pipe whip restraints shall be provided at the two intermediate high-stress points, as well as in the proximity of the auxiliary feedwater tunnel.

The specification for th'e main steam and feedwater line isolation valves will require each prospective vender to provide a detailed description of environmental qualification as well as the provisions for inservice tests. Environmental SS qualification will satisfy the environment which results from the breaks postulated in accordance with subsection 3.6.2.G.2. Inservice inspection tests will be performed periodically to demonstrate that these valves will function in accordance with design. The procedure will be available after a proposal is accepted.

O S5 la.3-6 POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - NOVEMBER 17, 1978

3

'VNPL 10.3.3.4  : Valve Testing

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^ -)" The steam safety valvesflocated?in the mainisteam piping-at'the 1

outlet.fromteach1 steam' generator are' individually-tested during initial startupfor;during1the shutdown operation by checking theLactual pop.and, closing pressures against:the required?

' design opening and closing pressures listed describingLthe N- safetylvalves.

.U Isolation 1 valve L leakage rates and acceptance criteria will be dependent on the; types-of. valve or valves purchased.. These valves will be-manufactured lin accordance with ASME SecHIII- ')

' Class 2..

I) The'inplant operational' testing will'be in.accordance withithe technical specification of;section -16.4.7. 1The inplantfleakage d

rate test'andLacceptance' criteria will;beidependent.upon1the type of valve selected.: After selection these tests will be included in the technical section of:the PSAR.

These documents will be specified in the. valve specificati'ons and more detailed information1will be available.when'a. proposal isfaccepted.

The opening and closing of the atmospheric power relief-valves j are likewise checked pri'or to-initial'startup or.during -j shutdown-

/~% 1 i/s The main steam line" isolation valves 1are located approximately twenty feet from the containment building-for.the following-reasons:-  !

1 n

A.

The main steam line safety and relief valves are to fj be placed directly on1the main steam line which: . .

requires that the; isolation' valve be located further l downstream. The safety and relief' valves alone i require approximately seventeen feet'of line. 'This-arrangement is considered'far more practical and-safer than placing the valves on'a header. j D

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1 S5 10.3-7 'I

, POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION -' NOVEMBER 17, 1978  ;

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I VNP l E: The safety valves are located outside the enclosure building so that a negative pressure'can be .

i maintained in the enclosure building. The enclosure  ;

building wall is approximately thirteen feet from 1 containment. Having the safety valves inside the enclosure building would require an additional ,

twenty-four enclosure penetrations per unit to discharge the safety and. relief valves to atmosphere.

Since umbrella type discharge stacks are to be used i as safety valve vents it would be impossible to maintain the negative pressure in the enclosure building.

C. The present arrangement of the safety and isolation valves provides a room to isolate these valves from other facilities. It also optimizes the support arrangements and expansion loops for the lowest practical stress loading. See figure 1.2-5.  ;

10.3.3.5 Pipe Testing Main steam piping located in the main steam tunnel in the S5 proximity of the diesel generator building will be volu-metrically inspected either in the shop or in the field. Tests and/or inspection will include circumferenti'al as well as j longitudinal welds.

10.3.4 SYSTEM TESTING l The various alarm and pressure trip points to isolate the steam (

generator feedwater pumps to prevent overpressurization are D checked by comparing design setpoints versus actual measured 1 trip settings. The main steam line is hydrostatically tested to confirm leaktightness. Visual inspection of pipe weld joints confirms the exterior condition of the weld. Pipeline i expansion and movement from the cold condition to the hot normal operating condition is checked by measuring movement i from field bench marks such as steel columns or pipe supports as specified on design isometric piping drawings indicating calculated movements along the x, y, and z axes.

e SS 10.3-8 e

POST-CONSTRUCTION PERMIT SUPPLEME'NTARY INFORMATION - NOVEMBER 17, 1978

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