ML20099H578

Rev 4 to Shoreham Defueled Sar

ML20099H578
Person / Time
Site:	Shoreham File:Long Island Lighting Company icon.png
Issue date:	07/31/1992
From:	LONG ISLAND LIGHTING CO.
To:
Shared Package
ML20099H569	List: ML20099H566 ML20099H578
References
NUDOCS 9208190100
Download: ML20099H578 (180)

Text

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4 EIEUiErI7

, 16 L SHOREHAM .

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DEryILID_JAFETY ANALYSIS REPORT fDSAR)

Insertion Instructions for Incorporating Revision 4 (NOTE: The electronic conversion process of the DSAR from the IBM 8100 System to the DEC System has resulted in different amounts of text on certain pages. These pages, included for completeness, are marked Revision 4, but do not have revision bars since no changes to the text have occurred.)

Replace the following pages of the DSAR with the attached pages. The revised pages are identified by revision number and, as appropriate, contain vertical line(s) in the right margin indicating the area of change.

REMOVE INSERT List of Effective Pages (3 pages) EP-1-1 and EP-1-2 Table of Contents (9 pages) Page i thru ix chapter 1

() 1-1 thru l-6 Chapter 2 1-1 thru l-7 2-1 and 2-2 2-1 and 2-2 2-2A 2-3 and 2-4 2-3 thru 2-5 Chanter 3 3-1 thru 3-6 3-1 thru 3-6 3-6A --

3-7 thru 3-16 3-7 thru 3-18 T3.2-1 (3-17 thru 3-23) T3.2-1 (1 thru 7 of 7)

Chanter 4 4-1 and 4-2 4-1 and 4-2 Chanter 6 6-1 6-1 6-1A 6-2 thru 6-4 6-2 thru 6-4 Chanter 7 7-2 thru 7-5 7-2 thru 7-5

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Page 1 of 2

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_. ,_j 9208190100 920721 PDR ADOCK 05000322 __

w. PDR

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BIRQYE ME  !

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Ehanter 8 8-1 thru 8-4 8-1 thru 8-4 8-4A Chapter 9 9-2 thru 9-4 9-2 thru 9-4 9-6 thru 9-8 9-6 thru 9-8 9-8A 9-9 9-9 9-9A --

9-10 thru 9-15 9-10 thru 9-15 9-15A 9-16 and 9-17 9-16 thru 9-19 Chanter 11 11-3 thru 11-21 11-3 thru 11-22 T11.1-1 (11-22) T11.1-1 (1 page)

T11.6.1-4 (11-23 and 11-24) T11.6.1-4 (1 and 2 of 2)

T11.6.3-1 (11-25) T11.6.3-1 (1 page)

T11.6.3-2 (11-26) T11.6.3-2 (1 page)

Chapter 12 10-3A 12-4 thru 12-6 12-4 thru 12-6 12-6A --

12-7 thru 12-10 12-7 thru 12-10 Chanter 13 13-1 13-1 13-1A --

13-2 thru 13-14 13-2 thru 13-28 T13.5.1-1 (1 thru 4 of 4) T13.5.1-1 (1 thru 3 of 3)

T13.5.1-3 (1 of 6) T13.5.1-3 (1 of 6)

T13.5.1-3 (4 of 6) T13.5.1-3 (4 of 6)

F13.1.1-1 F13.1-1 L F13.1.1-2 F13.5.1-1 Chanter 11 15-1 thru 15-5 15-1 thru 15-5 15-10 and 15-11 15-10 and 15-11 l- Chanter __12 17-1 thru 17-4 17-1 thru 17-26 F17.2.1-1 NUREG-0737 A111(39 pages) --

l THIS INSERTION INSTRUCTION IS TO.BE FILED IN THE FRON OF THE DSAR.

l p )- Page 2 of 2 lIg

SHOMEHAM DSAR LIST OF EFFECTIVE PAGES l es h NISION PAGE/ TABLE (T)/ FIGURE (F)

U 4 . . . . . . . . EP-1-1 and EP-1-2 4 . . . . . . . . I thru ix 4 . . . . . . . . 1-1 thru 1-7 3 . .. . . . . . T 1.2-1 (1 of 1) 4 . . . . . . . . 2-1-thru 2-5 4 . . . . . . . . 3-1 thru 3-18 4 . . . . . . . . T 3.2-1 (1 thru 7 of 7) 4 . . . . . . . . 4-1 and 4-2 3 . . .. . . . . 5-1 4 . . . . . . . . 6-1 thru 6-4 3 . . . . . . . . 7-1 4 . . . . . . . . 7-2 thru 7-5 3- . . . . . . . . 7-6 0 . . . . . . . . 7-7 3 . .. . . . . . 7-8 4 . - . . . . . . . 8-1 thru 8-4

( 0 . . . . . . . . 8-5 0 . . . . . . . . F 8.2.1-2 0 . . . . . . . . 9-1 4 . .. . . . . . 9-2 thru 9-4 3 . . . . . . . . 9-5 4 . . . . . . . . 9-6-thru 9-19 3 . . . . . . . . T 9.2.1-1 (1 & 2 of 2) 3 . . . . . . . . T 9.2.7-1 (1 of 1) 0 . . . . . . . . T 9.3.2-1 (2 pages) 0 . . .. . . . . T 9.5.9-1 (1 & 2 of 2) 0 . . . . . . . . F 9.1.2-1 3 . . . . . . . . 10-1 and 10-2 1 . . . . . . . .

11-1 3 . . . . . . . . 11-2 4 . . . . . . . . 11-3 thru=11-22 4 . . .. . . . . . T 11.1-1 (1 of 1)

T 11.6.1-4 (1 & 2 of 2) 4 . . . . . . . .

~4 . . . . . .. . . . .T 11.6.3-1 (1 of 1) 4 .... . . . . . T 11.6.3-2 (1 of 1) 0 . . . . . . . . F 11.6.3-1 0 .. . . . . . . F 11.6.3-2

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EP-1-1 Rev. 4

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SHOKEHAM DDAR LIST OF EFFECTIVE PAGES (Cont'd)  ;

./,,\ l V RELS1QH EMEl_ TABLE iT) / PLGMRE (F).

1 0 . . . . . . . . 12-1  !

3 . . . . . . . . 12-2 and 12-3 i 4 . . . . . . . . 12-4 thru 12-10 3 . . . . . . . . T 12.2 1 (1 of 1)

O . . . . . . . . T 12.2-2 (1 of 1) 0 . . . . . . . . T 12.3.4A (1 of 1) '

0 . . . . . . . . T 12.3.4B (2 pagos)

O . . . . . . . . T 12.3.4C (1 of 1)

O . . . . . . . . T 12.4-1 (1 of 1) 4 . . . . . . . . 13-1 thru 13-28 4 . . . . . . . . T 13.5.1-1 (1 thru 3 of 3) 0 . . . . . . . . T 13.5.1-2 (1 of 1) 4 . . . . . . . . T 13.5.1-3 (1 of 6) 0 . . . . . . . . T 13.5.1-3 (2 & 3 of 6) 4 . . . . . . . . T 13.5.1-3 (4 of 6) 0 . . . . . . . . T 13.5.1-3 (5 & 6 of 6) 4 . . . . . . . . F 13.1-1 0 . . . . . . . . 14-1 4 . . . . . . . . 15-1 thru 15-5 7

i'- 3 . . . . . . . . 15-6 0 . . . . . . . . 15-7 and 15-8 3 . . . . . . . . 15-9 4 . . . . . . . . 15-10 and 15-11 0 . . . . . . . . T 15.1.36-1 (1 of 1) 3 . . . . . . . . T 15.1.36-2 (1 of 1) 3 . . . . . . . . T 15.1.36A-1 (1 of 1) 0 . . . . . . . . F 15.1-1 0 . . . . . . . . F 15.l.36-1 0 . . . . . . . . F 15.1.36A-1 3 . . . . . . . . 16-1 4 . . . . . . . . 17-1 thru 17-26 O

EP-1-2 Rev. 4 July 1992

l SHOREHAM DSAR l

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(,) TABLE OF CONTENTS i l

Chapter /

Section Title Page i

1 UlTRODUCTION AND GENERAL DESCRIPTION OF PLbEl

1. 1 Introduction 1- 1

1. 2 General Plant Description 1- 5

1. 3 Comparison Tables 1- 7

1. 4 Identifit.ation of Agents and Contractors 1- 7

1. 5 Requirements for Further Technical Information 1- 7

1. 6 Material Incorporated by Reference 1- 7

1. 7 Symbols Used in Engineering Drawings 1- 7 2 SITE CHARA91 EEL 91.LCSi

3. 1 Geography and Demography .

2- 1

2. 2 Nearby Industrial, Transporation, and Military Facilities 2- 1

2. 3 Meteorology 2- 1

2. 4 Hydrologic Engineering 2- 2

,_, 2. 5 Geology & Seismology 2- 3

( 2A Boring Logs 2- 4

\~ 2B Scismicity Investigations 2- 4 2C A Reevaluation of the Intensity of the E.

Haddam, Conn. Earthquake of May 16, 1791 2- 4 2D Reevaluation of the Reported Earthquake at Port Jefferson, Long Island, New York 2- 4 2E Reevaluation of the Earthquake of October 26, 1845 2- 4 2F Reevaluation of the Earthquake of January 17, 1855 2- 4 2G Earthquakes Which Have Affected the Site with a Modified Mercalli Intensity of IV or Greater 2- 4 2H Report on Seismic Survey-Proposed Shoreham j Power Station LILCO 2- 4 2I Laboratory soils Testp 2- 5 2J banmary Report of Gootechnical Studies of -

Reactor Building-Foundation 2- 5

-2K Aircraft Crash Probability Study 2- 5 2L Report on Service Water System Soils 2- 5 2M Report on Densification of Service Water System Soile 2- 5 2N Hurricane Study 2- 5 7,

(,) i Rev. 4 July 1992

SHOREHAM DBAR tf }j TABLE OF CONTENTS Chapter /

Section Title Page 3 DESIGN OF STRUCTURES, COMPONENTS, EOUIPMENT, AND SYSTEMS

3. 1 Conformance to General Design Criteria for Nuclear Power Plants (10 CFR Part 50 App A) 3- 1

3. 2 Classification of Structures, Systems and Components 3-14

3. 3 Wind and Tornado Loading 3-15

3. 4 Water Level (Plood) Design 3-15

3. 5 Missile Protection 3-15

3. 6 Protection Against Lynamic Effects Associated with the Postulated Rupture of Piping 3-15

3. 7 Soismic Design 3-16

3. 8 Design 1of Seismic Category I Structures 3-16

3. 9 Mechanical Systems and Components 3-16 3.10 Seismic Qualif. of Seismic Category I Instrumentation.and Electrical Equipment 3-16 3.11 Environmental Design of Mechanical and Electrical Equipment 3-16 3.12 Separation Criterion for Safety Related

(~'s Mechanical and Electrical Equipment 3-17

(_,/ 3A Computer Programs for the Stress Analysis of Cat. I Structures, Dynamic and Static *

. Analysis, and Dynamic and Stress Analysis of Seismic Cat. I Piping Systems 3-18 3B NRC Regulatory Guides 3-18 3C Pipe Failure Outside Primary Containment 3-18 4 REACTOR

4. 1 Reactor Summary Description 4- 1

-4. 1. 1 Reactor. Vessel 4 4. 1. 2 Reactor Internal Components 4- 1

4. 1. 3 Reactivity Control System 4- 1

4. 1. 4 Analysis Techniques 4- 1

4. 4 Thermal and Hydraulic Design 4- 1

4. 5- Reactor Materials 4- 2

4. 6 Control Rod Drive Housing Supports 4- 2 5 REACTOR COOLANT SYSTEM

5. 1 Summary Description 5- 1 6 ENGINEERED SAFETY FEATURES

6. 1 General 6- 1

6. 2 Containment Systems 6- 1 t 11 Rev. 4 July 1992

SHOREHAM DSAR

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TABLE OF CONTENTS

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Chapter / '

Section Title Page

6. 2. 1 Containment Functional Design 6- 1

6. 2. 2 Containment Heat Removal System 6- 2

6. 2. 3 Containment Air Purification and cleanup System 6- 2

6. 2. 4 Containment Isolation System 6- 2

6. 2. 5 Combustible Gas Control in Containment 6- 2

6. 3 Emergency Core Cooling Systems 6- 2

6. 4 Habitability Systems G- 3

-6. 5 Main Steam Isolation Valve Leakage Control System 6- 3

6. 6 Overpressurization Protection 6- 3

6. 7 Main Steam Line Isolation Valves 6- 3 6, 8 Control Rod Drive Support System 6- 3

6. 9 Control Rod Velocity Limiters 6- 3 6.10 Main Steam Line Flow Restrictors 6- 3 6.11 Reactor Core Isolation Cooling System 6- 4 l 6.12- Standby Liquid Control System 6- 4 7 INSTRUMENTATI'ai AND CONTROLS '

7. 1 Introduction 7- 1

7. 1. 1 Identification and Classification 7- 3 l

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(_ 7. 1. 2 Identification of Safety Design Bases and Nonsafety Design Bases Criteria 7- 6 7.2 Reactor Protection System 7- 6

7. 3 Engineered Safety Feature System 7- 6

7. 4 Systems Required For Safe Shutdown 7- 7

7. 5 Safety Related Display Instrumentation 7- 7

7. 6 All Other Instrumentation Systems Required for Safety 7- 7

7. 6. 1 Description 7- 7

7. 7 Control Systems Not Required for Safety 7- 8 7A Plant Nuclear Safety Operational Analysis 7- 8  ;

7B ' Analog Transmitter / Trip Unit System for ESF i Sensor Trip Units 7- 8 8 ELECTRIC POWER

8. 1 Introduction e 8- 1

8. 1. 1 Utility Grid 8- 1 8 . - 1. 2 Interconnection to Other Grids 8- 1

8. 1. 3 Offsite Power System 8- 1

8. 1. 4 On Site AC Power System 8- 2

8. 1. 5 On Site DC-Power System 8- 2

8. 1. 6 -Identification of Safety-Related Systems 8- 2

8. 1. 7 Identification of' Safety Criteria 8- 2 l n

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SHOREHAM DSAR TADLE OF CONTENTS Chapter /

Section Title Page

8. 2 Offsite Power System 8- 3

8. 2. 1 Descr.iption 8- 3 C. 2. 2 Analysis 8- 3 l

8. 3 On Site Power System 8- 4

8. 3. 1 AC Power System 8- 4

8. 3. 2 DC Power System 8- 4 9 &UXILIARY SYSTEMS

9. 1 Fuel Storage and Handling 9- 1

9. 1. ' l - New Puol Storage 9- 1

9. 1. 2 Spent Fuel Storage 9- 1

9. 1. 3 Fuel Pool Cooling and Cleanup System 9- 2

9. 1. 4 Fuel Handling System 9- 2

9. 2 _ Water Systces 9- 3

9. 2. 1 Service Water System 9- 3

9. 2. 2 Reactor Building Closed Loop Cooling Water (RBCLCW) System 9- 4

9. 2. _ 3 Hakeup Water Domineralizer System 9- 4

9. 2. 4 Potable and Sanitary Water Systeps 9- 5

9. 2. 5 Ultimate Heat Sink 9- 5

9. 2. 6 Condensato Storage Facilities 9- 5

9. 2. 7 Turbine Building Closed Loop Cooling Water

(~)/

N- System 9- 5

9. 2. 8 Main Chilled Watel System 9- 5

9. 2. 9 Reactor Bldg Standby Vent Sys and Control Room AC Chilled Water System 9- 6

9. 3- Process Auxaliaries 9- 6

9. 3. 1 Compressed Air Sjstems 9- 6

9. .3. 2 Process Sampling System 9- 6

9. 3. 3 Equipment and Floor Drainage System 9- 7 l

9. 3. 4 Chemical, Volume control and Liquid Poison

' Systems _

9-~7

9. 3. 5 Failed Fuel Detection System 9- 7

9. 3. 6 Suppression Pool Pumpback System 9- 8 l

9. 4 Air Conditioning, Heating, Cooling, and Ventilation Systems 9- 8 l

9. 4. 1 Control Room Air Conditioning System 9- 8

9. 4. 2 Reactor Building Normhl Ventilation System 9- 8

9. 4. 3 Radwaste Building Ventilation 9- 9 I

9. 4. 4 Turbine Building Ventilation System and Station Exhaust System 9- 9 l .

9. 4. 5 Battery Room Heating and Ventilation 9- 9

9. 4. 6 Drywrll Air Cocling System 9- 9

-9. 4. 7 Screenwell Pump-House Heating and Ventilation 9- 9 iv Rev. 4 July 1992 O

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BHOREHaM DSAM  ;

n TABLE OF CONTENTH Chapter /

Section Title Page

9. 4. 8 Plant Heating 9-10

9. 4. 9 Primary Containment Purge System 9-10

9. 4.10 Diesel Generator Room Ventilation 9-10

9. 4.11 Relay Room, Emer. Switchgear Room &

Computer Room Air Cond. System 9-10 l

9. 5 Other Auxiliary Systems 9-10

9. 5. 1 Fire Protection System 9-10

9. 5, 2 Communications Systems 9-14

9. 5. 3 Lighting Systems 9-15 l

9. 5. 4 Diese] Generator Fuel Oil Storage and Transfer System 9-15

9. 5. 5 Diesel Generator Cooling Water System 9-16

9. 5. 6- Diesel Generator Starting System 9-16

9. 5. 7 Diesel Generator Lubrication System 9-16

9. 5. O Primary Containment Leakage Monitoring System 9-16

9. 5. 9 Storage of Gases Under Preasure 9-16 9A Fuel Criticality Analysis 9-18 9B . Evaluation of Spent Fuel Pool Makeup Requireuents 9-19 l

(> - 10 ETEAM AND POWER CONVFRSION SYSTEf4 Steam end Power Conversion System 10- 1

10. 1

10. 2 Turbine Generator 10- 1

10. 3 Main Steam Supply System 10- 1

10. 4- Other Features of Steam end Power -

Conversjon System 10- 1 10, 4. 1 Condenser 10- 1

10. 4. 2 Main Condenser Air Removal System 10- 1 -

10. 4. 3- Steam Scal. System 10- 1

10. 4. 4 Turbine Bypass System 10- 2

10. 4. 5 Circulating Water System 10- 2

10. 4. 6 Condencate Demineralizer System 10- 2 10, 4. 7 Condencate and Feedwater System 10- 2 11 BAQLQ]LCTIVE WASTE MANAGEMENT 111. 1 Radiation Source Termb 11- 1 Ell . 2 Radioactive Liquid Waste System 11- 2

11. 2. .1 Design. Objectives 11- 2

11. 2. 2 System Descriptions 11- 2

11. 2. 3- System Design 11- 3 11, 2. 4 Operating Procedures 11- 6 l

11. 2.-5 Performance Tests. 11- 6

11. 2. 6 Estimated Releases 11- 6

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_ SHORERAM DSAR

\-) TABLE OF CONTENTS o Cb.4pter/

6ectioP. Title Page

11. 2- 7 ' Release Points 11- 6

11. 2. 8 Dilution Factors 11- 6 11, V. 9 E%timated Doses 11- 7 l

11. 3 Gaseous WCata System 11* 7

11. 3. 1 Design Obj0ctives 11- 7

11. 3. 2 Cystem Descript$ons 11- 7

11. 3. 3 System nestgr. 11- 8 "

11. 3. 4 opercting Procedures 11- 8

11. 3. 5 Performance Tests 11- 8

11.  ?. 6 Estimated Releases 11- 8

11. 3. 7 Release Points 11- 8

11. 3. 8 Dispersion Factors 11- 8

11. 3. 9 Estimated Doses 11- 8

11. 3.10 Unmonitored kolease Points 11 ',

11. 4 Process & Effluent Radiation Monitoring System- l '. - 9

11. 5 Solid Waste System 11- 9

11. 1 Design Obj0ctives 11- 9

11. b. 2 Systom Input Source Terms 11- 9

(^3 11. 5. 3 Equipment Description 11- 9

\q$) 11. 5. 4 Expected Volumes 11-11

11. 5. 5 Packaging 11-11

11. 5. 6 Storage 11-11

11. 5. 7 Shipment 11-12

11. 6 offsite Aadiological Environmental Monitering Program 11-12

11. 6. 1 Objectivec of REMP 11-16

11. 6. 2 Potenti31 Pathways 11-18

11. 6. 3 Sampling Media, Locations, and Frcquency 11-19 11, 6. 4 (Not Used in the DSAR) 11-21 l

11. 6. 5 Data Analysis, Presentation, and Interpretation 11-21 p 11. 6. 6 Program Statistical Sensitivity 11-21 12 RADIATION PROTECTIOli

12. 1 Assuring that Occupational Radiation Exposures are ALARA 12- 1

' 12 . 2 Radiation Sources 12- 2

32. 2. 1 .Ccntained Sources 12- 2

12. 2. 2 Airborne Radioactive Material-Sources 12- 3 '

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12. 3 Radiation Protection Design Features 12- 4

12. 3. 1 Facility Design. Features 12- 4

12. 3. 2 Shielding 12- 5 l

12. 3. 3 Ventilation 12- 5 i

\- vi Rev. 4 July 1992

SHCREFAff DSAR

( TABLE OF CONTENTS

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Chapter /

Section Title Page

12. 3. 4 Radiation Monitoring Instrumentation 12- 5

12. 4 Dose Assessment 12- 7

12. 4. 1 Design Objectives 12- 7

12. 4. 2 Airborne Activity 12- 7 I

12. 4. 3 Occupational Dose Assessment 12- 8

12. 4. 4 Offsite Dose Assessment 12- 8

12. 5 Health Physics Program 12- 9 I 13 CONDUCT OF OPERATIONS

13. 1 Organizational Structure and Responsibilities 13- 1 l

13. 1. 1 Corportte Organization 13- 1

13. 1. 2 Operating Crgcnization 13- 3

13. 2 Training Program 13-18 13.-2. 1 Training to Support Maintenance in the Defueled Condition 13-18 l 13, 2. 2 . Training to Support Decommissioning Activities 13-19

13. 3 Emergency Planning 13-19

13. 4 Review and Audit 13-20

13. 4. 1 Site Review Committee 13-21

(~ 13. 4. 2 Independent Review Panel (IRP) 23-23 i 13. 5 Station Procedures 13-27

13. 5. 1 Administrative Control 13-27

13. 5, 2 Procedures 13-27

13. 6 Plant Records 13-28

~13. 7 Industrial Security 13-28 14 INITIAL TESTS AND OPERATIONS 15 ACCIDENT ANALYSIS

15. 1 General 15- 1

15. 1. 1 Generator Load Rejection 15- 2 l

15. 1. 2 Turbine Trip 15- 2

15. 1. 3 Turbine Trip with Failure of Generator Breakers to Open 15- 2

15. 1. 4- Main Steam Isolation Valve Closure 15- 2 l

15. 1. 5 Pressure Regulator Failure - Open 15- 2

15. 1. 6 Pressure Regulator Failure - Closed 15- 2 l

15. 1.'7 Feedwater Controller Failure-Maximum Demand 15- 2

15. 1. 8 Loss-of Feedwater Heating 15- 2 15, 1. 9 Shutdown Cooling (RHR) Malfunction-Decreasing Temperature 15- 2

15. 1.10= Inadvertent HPCI Pump Start 15- 4 July 1992

(.s) vii Rev. 4 L

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l SHORERAM DSAR

/"' TABLE OF CONTENTS Chapter /

section Title Page

35. 1.11 Continuous Control Rod Withdrawal During Power Range Operation 15- 4

15. 1.12 Continuous Control Rod Withdrawal During Reactor Startup 15- 4

15. 1.13 Lontrol Rod Removal Error During Refueling 15- 4

15. 1.14 Fuel Assembly Insertion Error During Refueling 15- 4

15. 1.15 off-Design Operational Transient Due to Inadvertent Loading of a Fuel Assembly into an Improper Location 15- 4

15. 1.16 Inadvertent Loading and Operation of Fuel Assembly in Improper Location 15- 4

15. 1.17 Inadvertent opening of a Safety Relief Valve 15- 4

15. 1.18 Loss of Feedwater Flow 15- 2 l

15. 1.19 Loss of AC Power 15- 3

15. 1.20 Recirculation Pump Trip 15- 3

15. 1.21 Loss of Condenser Vacuum 15- 3

15. 1.22 Recirculation Pump Seizure 15- 3

- 15. 1.23 Recirculation Flow Control Failure With Decreasing Flow 15- 3

15. 1.24 Recirculation Flow Control Failure With Increasing Flow 15- 4

15. 1.25 Abnormal Startup of Idle Recirculation Pump 15- 4

( 15- 3

15. 1.26 Core Coolant Temperature Increase 15, 1.27 AnticApated Transient Without Scre.m (ATWS) 15- 4 l

15. 1.28 Cask Drop Acc3 dant 15- 5 15, 1.29 Miscellaneous Small Releases Outside Primary Containment 15- 5 15, 1.30 Off-Design Operational Transient as a Consequence of Instrument Line Failure 15- 5

15. 1.31 Main Condenser Gas Treatment System Failure 15- 5

15. 1.32 Liquid Radwaste Tank Rupture 15- 5 I fi . 1.33 Control Rod Drop Accident 15- 4

15. 1.34 Pipe Breaks Inside the Primary Containment (Loss-of-Coolant Accident) 15- 5

15. 1.35 Pipe Breaks Outside the Primary Containment (Steam Line Break Accident) 15- 5

15. 1.36 Fuel Handling Accident 15- 7

15. 1.36A Worst Ca3e FLel Damag'e Event 15-10

15. 1.37 Feedwater System Piping Break 15- 4

15. 1.3C Failur< of Air Ejector Lines 15- 7 16 IXpJitECAL SPEQJFICATIONS

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viii Rev. 4 July 1992

SHOREHAM DSAR

(% TABLE OF CONTENTS d

Chapter /

Section Title Page 17 OUALITY ASSURANCE 17, 1 Quality Assurance During Design and Construction 17- 1

17. 2 Quality Assurance During the Decommissioning l Phase 17- 1

17. 2. 1 Organizations 17- 1

17. 2. 2 Shoreham Quality Assurance Program 17- 4

17. 2. 3 Design Control 17- 7

17. 2. 4 Procurement Document Control 17- 9

17. 2. 5 Instructions, Procedures, and Drawings 17-11

17. 2. 6. Document Control .

17-12 17, 2. 7 Control of Purchased Material, Equipment, and Services 17-13

17. 2, 8 Identification and Control of Materials, Parts, and Components 17-15

17. 2. 9 Control of Special Processes 17-16

17. 2.10 Inspection 17-17

17. 2.11 Test Control 17-18

17. 2.12 Control of Measuring and Test Equipment 17-19

17. ?. 13 Handling, Storage, and Shipping 17-21

_gs 17. 2.14 Inspection, Test, and Operating Status 17-21

! / 17. 2.15 Nonconforming Materials, Parts, or Components 17-22

17. 2.-16 Corrective Action 17-24

17. 2.17 Quality Assurance Records 17-24

17. 2.18 Audits 17-26

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~J 1x Rev 4 July 1992

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SHOREHAM DSAR

[v i CHAPTER 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT

1.1 INTRODUCTION

This Defueled Safety Analysis Report (DSAR) is an appendix to the Shoreham USAR and is submitted by Long Island Power Authority, hereafter known as LIPA, in support of the permanently defueled configuration of the Shoreham Nuclear Power Station as authorized by Facility Operating License NPF-82, i.e. the SNPS Possession only License or POL, as. transferred from the Long Island Lighting Company (LILCO).

Tho. description of the plant remains essentially unchanged from the description in Section 1.1 of the SNPS USAR. However, many of the sections which described systems needed to support power operation are significantly changed or excluded from the DSAR.

The DSAR format is the same as that used for the USAR (i.e. NRC Regulatory Guide 1.70, Rev. 1, 1972); however, commensurato with the level of activity of a defueled plant, the content is reduced.

}

The purpose of the DSAR is to provide a safety analysis for the storage and handling of Shoreham low burnup first cycle spent fuel. The DSAR confirms that fuel storage and handling systems, structures, components _and programs ensure that there ja no undue risk-to public health and safety _during normal and postulated accident conditions.

-The DSAR-assumes-that_the 560 fuel bundles comprising the

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Shoreham core are stored under water in the Shoreham spent fuel pool. The fuel bundles are held in seismic category I spent fuel racks within the stainless steel-lined spent fuel pool.- The spent fuel pool is located in the secondary containment of the Shoreham reactor building. The structures are designed to withstand seismic loads. ,

The Shoreham spent fuel is in a low burnup condition. .The Shoreham Nuclear Power Station operated during low power testing at power levels not exceeding 5% of rated power. The effective burnup of the fuel is approximately 2 full power days. This results in an estimated total core wide heat generation rate of approximately 550 watts-as of June 1939. The estimated fuel heat load will reduce-to approximately 250 watts by June 1991. Figure 15.1-1 depicts the fuel heat load versus time. Based on this low A

() 1-1 Rev. 4 July 1992 h

q SilCREllAM DSAR L) heat generation rate, systems for active cooling are not required, and only minimal capacity systems are required for pool water makeup to handle evaporation.

The Shoreham spent fuel contains limited quantities of radioactive materials that are available for release. As is -

I stated in DSAR Section 12.2, approximately 176,000 curies of radioactivity reside in the 560 fuel assemblies. Gaseous  !

activity in the fuel assemblies is primarily Krypton-85 (a noble gas with a 10.7 year half-life), and consists of approximately 1560 curies. The radioactive inventory estimation is based on a i two year decay from the last burnup period (completed June 7, l 1987), other sources of radioactivity outside the core are minor,.and include small amounts of contamination in the bottom of sumps, the suppression pool, inside the reactor pressure vessel, and in the radwaste systems.

Chapter 15 presents radiological analyses for those accider.ct, identified in the USAR which are applicable to the defue19d plant. In addition,.no other accident mechanisms were identified for the plant's defueled condition which are not bounded Oy Chapter 15. The events analyzed in Chapter 15 are:

~/ 1.

2.

Fuel Handling Accident (Fuel Bundle Drop)

Radwaste Tank Rupture The only design basis accident involving reactor fuel is a Fuel Handling-Accident, in which no heat generation takes place. As such, the activity available for release in this design basis accident-is primarily Krypton-85, and consists,of approximately 2.5. curies. In addition, a worst case radiological event is postulated in-which the entire gaseous activity of'the. core is released to the reactor building.. This event was postulated to conservatively bound any possible situation involving large-scale mechanical damage of the fuel.

The'results of the September 1989 spent-fuel radiological analysis described in DSAR Chapter 115 indicate that integrated doses are very small in comparison with 10CFR100 limits. For the worst case scenario in which all the gaseous activity is assumed to be released from the entire core, a spectrum of cases were analyzed as follows: operation of the standby ventilation system, operation of the normal ventilation system, and no

ventilation (modeled as puff release) . The results of the analyses indicate that the integrated whole body and skin doses, with. Reactor Building Normal Ventilation System operational, are less than approx.imately .03% of 10CFR100 limits. The results of 7-the radiological-ana)ysis for the worst case fuel damage scenario

- 1-2 Rev. 4 July 1992

SHOREHAM DSAR are depicted graphically in Figure 15.1.36A-1. In part.co -

, it  ;

was demonstrated that the reactor building standby ventilation j system operation does not provide an important filtering or ventilation safaty function and is therefore no longer required after fuel is +,7 red in the pool.

Based on this analysis, it has been found that the spent fuel pool provides a high degree of passive safety protection for Shoreham spent fuel. Active safety systems are not required to  !

mitigate postulated accidents; however, support systems are required to meet the intent of 10CFR50 Appendix A, General Design Criteria (see Chapter 3 for a listing) and Regulatory Guide 1.13.

. Supporting systems are required to provide for radiation monitoring, fuel pool makeup,-fuel pool cleanup, radwaste management, and normal bullaing' services. Thorofore a reclassification of safety systems is propoJed based on the importance to safety associated with each plant system with the plant defueled.-

The DSAR assumes that the Shoreham spent fuel from the initial core is to be stored for some interim period in the spent fuel pool contained within the SNPS reactor building.

The assumed configuration of principal plant systems is as

/)T

(_ follows:

1. All 560~ fuel bundles have been removed from the reactor and are being stored in seismic Category I spent fuel-racks in a the spent fuel storage pool. The total decay heat power of the-entire core has been determined to be approximately 550 watts as of June 1989 (reference DSAR Chapter 15).

2. As. described in-DSAR Chapter 9, the spent fuel-storage-pool 4 water level is maintained at its normal water level. Makeup

-will be furnished from the condensate transfer system or the domineralized~and makeup water system. The fuel pool cooling system;is not in-service due to the low heat-load in the pool. Water quality is maintained by the fuel pool cleanup system. The spent fuel pool transfer canal gates will remain installed. Fuel pool level'and temperature are alarmed in

-the Control Room.

3. The capability for fuel handling will be maintained as described in DSAR Chapter 9.

4. The Nuclear Boiler,-Reactor Protection, Emergency Core Cooling, and Primary containment systems are not required.

-This is discussed in DSAR Chapters 4, 5 and 6.

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SIIOREllAM DSAR

5. Two independent offsite AC power sources will be maintained i to supply reliable electric power. In addition, as discussed-in Chapter 8, blackstart combustion turbines exist nearby in l the Shoreham west site to supply emerger.cy power to the i plant. However, as discussed in DSAR Chapter 15, onsite Emergency Diesel Electric Power is not required to mitigate design basis accidents. AC Power is required by Technical Specifications to remain operable during fuel movement '

'(including one non-safety emergency diesel generator).

6. The normal. ventilation system (RBNVS) provides a controlled and monitored release capability but secondary containment integrity is no longer required as discussed in the DSAR l Chapter 15 Safety Analysis.

7. The steam and power conversion systems are not required to be operabic or functional.

8. Process and area radiation monitoring are maintained consistent with fuel storage and handling requirements, and are described in DSAR~ Chapters 11 and 12.

9. Radwaste Systems described in DSAR Chapter 11 are maintained to provide an appropriato level of radioactive liquid and

(-) solid waste management primarily due to operation of the spent-fuel pool.

10. Major systems that remain functional to provide non-safety related supporting services include: ,

a) Service Water (DSAR Chapter 9 and_10) b) Chilled Water Systems (DSAR Chapter 9) c) Cornpressed Air (DSAR Chapcer 10) d) HVAC Systems (DSAR Chapter 9)

The DSAR addresses the~following major pregrams:

1. Proposed revised Technical-Specifications (Appendices A and B) including the basis of the: specification is provided.

(DSAR Chapter 16) ,

2. . . Conduct of operations-and the-LIPA organizational structure is described in. Chapter 13. The ISEG functions are no longer-

. considered necessary for a defueled reactor.

3. The Quality Assuranca Program is maintained as described in

-DSAR Chapter 17.

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SHOREHAM DSAR O

4. The Fire Protection Program is maintained as described in .

DSAR Section 9.5.1 and the FHAR. With respect to overall nuclear safety, the primary focus of the Fire Protection Program is shifted from the protection of plant safe shutdown capability to the safety of stored irradiated fuel.

5. An offsite Radiological Environmental Monitoring Program (REMP) is maintained es described .. OSAR Section 11.6.

6. Changes to the LIPA Security Plan are being provided separately from the DSAR.

7. A LIPA Defueled Emergency Preparedness Plan is submitted separately to the NRC, adapted from the.LILCO plan previously reviewed and approved by the NRC for the defueled configuration of SNPS.

8. A-Shoreham Decommissioning Plan was submitted separately to

'the NRC and is incorporated by reference in this DSAR. See Section 1.2 for additional information.

1.2 GENERAL PLANT DESCRIPTION g

This section of the USAR is historically descriptive but the specifics of general and design criteria and modes of operation are generally no longer applicaole to the defueled plant. Design '

and operating information will be found in other sections of the DSAR e.g., Table 3.2-1.

Refer to the USAR for information on this subject. However, the systems,which will remain-operable for an extended time period in the defueled condition are listed in Table 1,2-1 of the DSAR.

All other systems will be either functional or-non-operable.

The following definitions apply:

1. Operable -System (s) maintained to meet Technical Specifications.-

2. Functional - Essential suppbrt system (s) not required p(J Technical-Specifications but necessary for minimal plant functions, habitabil'ity, and maintenance.

3. .-Nonoperable - Those systems not normally operated in the

,defueled-mode.- These systems will be in the deenergized .

l state. All systems Will be maintained consistent with the-

Decommiseloning Rule (no action will be taken which will affect the methods or options available for decommissioning

( 1-5 Rev. 4 July 1992 w

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i SHOREHAM DSAR f3 V

or increase the cost of decommissioning prior to approval of a decommissioning plan). These systems may be operated as necessary to support decommissioning activities as required.

The Shorehan Decommissioning Plan submitted on Decembar 29, 1990 as amended by letters SNRC-1832 (8/26/91), LSNRC-1855 (10/16/91),

LSNRC-1859 (11/27/91) and LSNRC-1874 (le/6/91) contains a detailed description of the plan for the decommissioning (i.e.

decontamination and dismantlement) of Shoreham's radioactive systems and structures. The Shoreham Decommissioning Plan as am'nded, is hereby incorporated by reference upon its approval by th; .C .

Where information on systems or structures appears in both the DSAR and the DP, the information in the DP must be ;onsidered governing. For example, the DP states 1, hat the following systems and structures are contaminated or activated and must be decommissioned:

Systems Control Rod Drive

• Process Sampling Il

• Core Spray

\l Residual Heat Removal

• Reactor Water Cleanup

• Liquid Radwaste

• Fuel Pool Cooling and Cleanup Condensate Demineralizer Reactor Recirculation _

8tructures

• Primary Containment Equipment / Floor Drains and Sumps Dryer and Separator Storage Pool

• Reactor Head Cavity Spent Fuel Storage Racks Spent Fuel Storage Pool '

• Radwaste Laydown Area Reactor Pressure Vessel and Internals The DSAR, however, also contains descripticas of these systems and structures which do not address how they will be affected by decommissioning.- The descriptions of the above systems and structures in the DSAR are, therefore, historical only and are superseded by the information in the DP.

1-6 Rev, 4 July 1992

-SHOREHAM DSAR c? b:

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-1.3 COMPARISON' TABLES The description contained under this heading in the latest revision of the Shoreham USAR remains unchanged. Refer to the USAR for information on this subject.

- 1. 4 -IDENTIFICATION OF AGENTS IED CONTRACTORS The. description contained under this heading in the latest revision of the Shoreham_USAR remains unchanged. Refer to the USAR for information n'this subject.

1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION The description contained under this heading in the latest reiision of the Shoreham USAR remains unchanged. Refer to the USAR for information on this subject. However, the status of systems wnich will remain operable for an extended time period in the defueled condition is described in Table 1.2-1 of the DSAR.

The systems described in this section are not required for the

-defueled-condition.

- (' 1 1.6 MATERIAL INCORPORATED BY REFERENCE

(~

The information contained under this heading in the latest .

revision of-the-Shoreham USAR remains unchanged. Refer to the USAR ft information on this subject.

.1. 7 -- SYMBOLS USED IN ENGINEERING DRAWINGS The'information contained under this heading in the latest revision of-the-Shoreham-USAR remains unchanged. Refer to the USAR for informationfon this subject.

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SHOREHAM DSAR

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% CHAPTER 2 SITE CHARACTERISTICS 2.1 GEOGRAPHY'AND DEMOGRAPHY The' description contained under this heading in the latest revision of Shoreham USAR remains unchanged. Refer to USAR for information'on this subject.

2.2 NEARBY INDUSTRIAL, TRANSPORTATION AND MILITARY FACILITIES The description-contained under this heading in the latest revision of Shoreham USAR remains unchanged. Refer to USAR for information on,this subject.

-2.3 METEOROLOGY The description contained under this heading in the latest revision of-Shoreham USAR remains unchanged except that the 33 ft. tower south of the plant will not be used. Additionally, the following information regarding the Operational Program applies to DSAR. Refer to USAR for other information on this subject.

,s

( 2.3.3.2 Operational Procram The operational-meteorological monitoring program uses instrumentation to determine wind-speed and -direction at 33- and 150-ft.-ambient air temperature at 33-ft and temperature differential (Temp 9-150-ft minus Temp 19 33-ft).. -These instruments are located on SNPS400 ft. peteorological tower

which Lis11ocated approximately 5100-ft WSW of. the reactor building;(Figure l2.1.1.1). The MET tower was positioned sufficiently close to SNPS to provide representative observations of_ released gaseous' effluents,.but far enough'away to minimize

. atmospheric disturbances caused by SNPS' structures.

' Wind-speed _and -direction at the_33-ft level, along with the temperature differential are transmitted to the Technical Support Center. .In addition to these parameters, wind-speed and

-direction at 150-ft., and temperature at 33-ft. are transmitted

.to the. Main Control Room and' entered into the RMS computer.

LAll' instrumentation-was either manufactured or supplied by Climatronics Corporation, Hauppauge, New York. The specifica-tions outlined in Regulatory Guide 1.23 were used'in the selection'of these instruments. Wind instrumentation includes F460 wind sets (three cup anemometers and direction vanes) at the f}

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2-1 Rev. 4 July 1992 l

SHOREHAM DSAR

(^h'

1. 33 and 150 ft. levels. Temperature sensors in shielded aspirators.are oriented in a northerly direction to limit the influence of solar insolation. A motor and fan draw a constant flow'of air at' ambient conditions over the sensor to ensure accurate measurements.

Observations from 33 ft. are used to model the dispersion of ground level release of activity, while data from 150 ft. are used'for elevated releases. The data obtained are used to project the dispersion of plant gaseous effluents based on Gaussian model and are included in requited periodic reports.

To ensure the operability of the system, semi-annual calibrations are performed by-a qualified vendor, and channel checks are performed by the operators on shift using qualitative assessment of the channel's behavior during operation. Operators do this by checking the chart recorders in the control room. This instrumentation includes:

1) Wind speed monitors at the 33-ft. and 150-ft. elevations;

2) Wind direction monitors at the 33-ft. and 150-ft. elevations;

3) Ambient temperature monitor at the 33-ft elevation; and

4) Differential air temperature monitor which uses the temperature data recorded at 33-ft. and 150-ft. elevations.

(3

' ( ,)

~

Meteorological sensors are replaced on a semi-annual basis with replacement sensors which have'been calibrated in the laboratory of a qualified vendor. Vendor personnel perform the sensor substitutions.under the direction of LIPA personnel. LIPA/LILCO technicians perform normal maintenance and inspection on instrumentation at the tower. Calibration and maintenance procedures have been developed for field testing and maintenance of.each meteorological channel at the Shoreham site.

Spare' sensors and auxiliary equipment are available for replacement of any malfunctioning components of the system. In the event that a Technica1' Specification meteorological tower instrument is~ damaged, causing one or more monitoring instrumentation channels to be inoperable for more than seven (7) days, refer to the Technical Specifications for the required action. .

2.4. HYDROLOGIC ENGINEERING The description contained under this heading in the latest revision of Shoreham USAR remains unchanged with the exception of-Subsections 2.4.8.1 and 2.4.11.5:

f} 2-2 Rev. 4 July 1992 y

E SHOREHAM DSAR b

2.4.8.1 Canals l The USAR requires that the Intake Canal bottom be monitored on a yearly basis, and dredging carried out when the results of the annual monitoring indicate cumulative sediment deposition has exceeded one (1) foot. This one (1) foot maximum sediment depth requirement is based upon anticipated sediment deposition of 3.2 feet during a low water Probably Maximum Hurricane (PMH) event. For the defueled condition, design for the PMH is not required since the decay heat load of the fuel is negligible.

Annual monitoring and dredging will not be required during the time that the plant is expected to be in the defueled condition.

This is based on the May 1990 Intake Canal soundings and the current rate of sediment deposition. However, the intake canal will continue to be used as a source of cooling water for normal plant needs (refer-to DSAR Section 9.2.1).

2.4.11.5- Plant Reauirements The USAR states that the required minimum safety related cooling water flow is 12,800 gpm supplied by two service water n'mps.

This minimum safety related flow is no longer requt ed ;r the defuel'ed condition since the RBSW system is considered non-safety related because it does not provide cooling water to any plant 3 equipment required to perform a safety function. One RBSW pump will be used to supply cooling water for normal plant needs (see DSAR Section 9.2.1).

2.5 GEOLOGY AND SEISMOLOGY The description contained under this heading in the latest revision of Shoreham USAR remains unchanged. Refer to USAR for information on this subject.

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() 2-3 Rev. 4 July 1992 l

SHOREHAM DSAR

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(

~'

)-

2A BORING LOGS The description contained under this heading-in.the latest revision of Shoreham USAR remains unchanged. Refer to USAR for information_on this subject.

-2B -SEISMICITY INVESTIGATIONS The description contained under this heading in the latest revision of Shoreham USAR_ remains unchanged. Refer to USAR for information on this subject.

2C A REEVALUATION OF THE INTENSITY OF THE EAST HADDAM, CONNECTICUT EARTHQUAKE OF MAY 16, 1971 The description contained under this heading in the latest revision of Shoreham USAR remains unchanged. Refer to USAR for information on this subject.

2D REEVALUATION OF THE REPORTED EARTHQUAKE AT PORT JEFFERSON, LONG' ISLAND, NEW YORK The description contained under this heading in the latest revision of Shoreham'USAR remains unchanged. Refer to USAR for

'T -information on this subject.

.[Q 2E REEVALUATION OF THE EARTHQUAKE OF OCTOBER 26, 1845 The description contained under this heading in_the latest revision of-Shoreham USAR remains unchanged. Refer to USAR for information on this subject.

2F REEVALUATION OF THE EARTHQUAKE OF JANUARY 17, 1855 The description contained under this heading _in the latest revision of Shoreham USAR remains unchanged. Refer to USAR for information on this subject.

2G' EARTHQUAKES WHICH HAVE AF1-ECTED THE SITE AREA WITH A MODIFIED MERCALLI' INTENSITY OF IV OR, GREATER The-descriptian contained under this heading in the latest

. revision of.Saoreham USAR remains unchanged. Refer to USAR for information on this subject.

2H REPORT ON SEISMIC SURVEY-PROPOSED SHOREHAM POWER STATION LONG ISLAND LIGHTING COMPANY The description contained under this heading in the' latest r- revision of Shoreham USAR remains unchanged. Refer to USAR for

\'

2-4 Rev. 4 July 1992

SHOREHAM DSAR

--O information on this subject.

21 LABORATORY SOILS TESTS The description contained under this heading in the latestRefer to USAR for revision of Shoreham USAR remains unchanged.

information on this subject.

2J

SUMMARY

REPORT OF GEOTECHNICAL STUDIES OF REACTOR BUILDING FOUNDATION The description contained under this heading Refer in thetolatest USAR for revision of Shoreham USAR remains unchanged.

information on this subject.

2K AIRCRAFT CRASH PROBAB7LITY STUDY The description contained under this heading Refer in thetolatest USAR for revision of Shoreham USAR remains unchanged.

information on this subject.

2L REPORT ON SERVICE WATER SYSTEM SOILS The description contained under this heading Refer in thetolatest USAR for O revision of Shoreham USAR remains unchanged.

information on this subject.

2M REPORT ON DENSIFICATION OF SERVICE WATER SYSTEM SOILS The description contained under this heading Refer in thetolatest USAR for revision of Shoreham USAR remains unchanged.

information on this subject.

2N HURRICANE STUDY The description contained under this heading Refer in thetolatest USAR for revision of Shoreham USAR remains unchanged.

information on this subject.

O 2-5 Rev. 4 July 1992

SHOREHAM DSAR i )-

- \J CHAPTER 3-DESIGN OF STRUCTURES. COMPONENTS.-EOUIMENT. AND SYSTEMS

-3.1 CONFORMANCE TO GENERAL DESIGN CRITERIA FOR NUCLEAR POWER PLANTS (10CFR Part 50, Appendix A)

The_ General Design Criteria (GDC), contained in the Shoreham USAR Section 3.1, were reviewed to establish those criteria that may be applicable to the otorage of SNPS low burnup_ cycle spent fuel -

in:the spent fuel pool. The following GDC are addressed:

I. Overall Reauirements GDC1 Quality Standards and Records GDC2 Design Bases for Protection Against Natural Phenomena GDC3- Fire Protection GDC4 Environmental and Dynamic Effects Design Bases

.II. frotection by Multiple Fission Product Barriers

[)

\_/

GDC13 GDC17 Instrnmentation and Control Electric Power Systems GDC18 Inspection'and Testing of Electric Power Systems GDC19 Control Room IV. Fluid Systems GDC44 . Cooling Water GDC45 -Inspection of Cooling Water System GDC46 Testing of Cooling Water System VI. Fuel and Radioactivity Control

-GDC60 Control of releases of radioactive material to the environment GDC61 Fuel: storage and handling and radioactivity control GDC62 Prevention of criticality in fuel storage and handling GDC63 Monitoring fuel and waste storage GDC64 Monitoring radioactivity releases

()

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3-1 Rev. 4 July 1992

SHOREHAM DS7.R

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'The following GDC were found not to be applicable to a defueled reactor:.

I Overall Recuirements GDCS Sharing of structures, systems, and components shoreham is a single unit, thus the above criterion does not apply.

II Protection av Multiple Fission Product Barriers GDC10 Reactor Design GDC11 Reactor Inherent Protection GDC12 Suppression of reactor power oscillations GDC14 Reactor' Coolant Pressure Boundary GDC15 -Reactor Coolant System GDC16 Containment Design The above criteria do not apply because the reactor and primary containment are not operable.

III Protection And Reactivity Control Systems

-GDC20 - 29 requirements apply only to an operating reactor

_ protection and reactivity control systems IV Fluid Systems GDC 30-43. address reactor and containment systems required for power operation only.

V Reactor Containment GDC 50- 57-address the primary containment design which is no longer _ required for a defueled reactor.

Aeolicable Criterion Conformance Ouality Standards and Records (Criterion 1)

Criterion Structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards comraensurate with the importance of the safety functions to be

-performed. 'Where generally recognized codes and-standards are used, they shall be identified and evaluated to determine their applicability,. adequacy, and sufficiency and shall be O,

supplemented or modified as necessary to escure a quality product 3-2 Rev. 4 July 1992

I i

l SliOREHAM DSAR l q

U in keeping with the-required safety function. A quality assurance program shall be established and implemented in order to provide adequate assurance that these structures, systems, and components will satisfactorily perform their safety functions.

Appropriate records of the design, fabrication, erection, and testing of structures, systems, and components important to safety shall be maintained by or under the control of the nuclear power unit licensee throughout the life of the unit.

Desian Conformance Structures, systems, and components are classified in Section 3.2. The LIPA QA program described in DSAR Chapter 17 assures that quality practices and documentation are maintained

. commensurate with the classification-that is identified in this Defueled Safety Analysis Report (DSAR).

Desian Basis for Protection Aaainst Natural Phenomena (Criterion 2)

Criterion Structures, systems, and components important to safety shall be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches

()\

k- . without loss-of capability to perform their safety functions.

~

The design-bases for these structures, systems, and components shall reflect:- (1) appropriate _ consideration of the most severe of the natural phenomena that have been historically reported for.

the site and: surrounding area, with sufficient margin for the limited accuracy, quantity, and period of time in which the historical data have been accumulated, (2) appropriate combinations of the effects of normal and accident = conditions with the effects of the natural phenomena, and (3) the importance of-the-safety functions to be performed.

Desian Conformance The spent fuel racks, fuel pool, and reactor building which are required to maintain the SNPS fuel in a safe condition are designed to withstand 1.atural phenomena as described in the USAR.

Because of'the low burnup condition of'the SNPS Cycle 1 spent fuel, the need-for support systems is limited (see Chapters 9, 15). Natural phenomena are described in Chapter 3 of the Shoreham USAR.

p)

( 3-3 Rev. 4 July 1992

SHOREHAM DSAR O

fire Protection (Criterion 3)

CI.iterion Structures,. systems, and components iuportant to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions.

Noncombustible and heat resistant materials shall he used wherever practical throughout the unit, particularly in locations such as tne containment and control room. Fire detection and fighting systems of appropriate capacity and capability shall be provided and designed to minimize the adverse effects of fires on structures, systems, and components important to safety. Fire fighting systems shall be designed to asstre that their rupture or inadvertent operation does not signifi-antly impair the safety capability of these structures, systems, and components.

Desian Conformance This-criterion is satisfied by the SNPS fire protection program which is described in Section 9.5.1 of this report and the USAR.

() Environmental and Missile Desian bases (Criterion 4)

Criterlom 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 shall be: appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that may result from equiprent failures and frou events and conditions-outside the nuclear power unit.

Desian Conformance Chapter 15 of this report.defin'es accidents that are applicable to spent fuel storage and fuel handling. The spent fuel is stored in the spent fuel storage pool. The pool structure, Reactor Building, and spent' fuel racks provide passive safety protection from missiles or other conditions that could cause l fuel mechanical damage. The structural design basis of the fuel l storage racks is discussed in Chapter 9 of the USAR. Additional

-information on the design of structures, systems, and components can be found in Chapter 3 of the Shoreham USAR.

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SHOREHAM DSAR-(

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i Instrumentation and Control (Criterion 13)  !

l Criterion Instrumentation shall be provided to monitor variables and-systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as-appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems.

Appropriate controls shall be provided_to maintain these variables and systems within prescribed operating ranges.

Desian Conformance Instrumentation is'provided to monitor spent fuel pool level and temperature as well as fuel pool cleanup. Instrumentation is provided for process ano effluent radiation monitoring, area and airborne radiation monitoring, and accident monitoring.

Radiation monitoring is maintained as described in DSAR Chapters 11 and 12.

' Electric Power Systems (Criterion 17)

Criterion

An onsite electric power system and an offsite electric power system shall be provided to permit functioning of structures, systems, and components important to safety. The safety function for1each system (assuming the cther system is not functioning) shall be to provide sufficient capacity and capability to assure that. (1) specified acceptable fuel design limits and design conditions of the reactor coolant pressure' boundary are not exceeded as a result'of anticipated operational occurrences and (2) the core is cooled and containment integrity and other vital functions are maintained in the event of postulated accidents.

I _The onsite electric power supplies, including the batteries, and the onsite electric distribution systel shall have sufficient independence, redundancy, and testabil_ty to perform their safety functions assuming a single failure.

Electric power from_the transmission network to the onsite electric distribution system shall be supplied by two physically independent circuits (not necessarily on separato rights of way) designed and located so as to minimize to the extent practical the_ likelihood of their simultaneous failure under operating and 3-5 Rev. 4 July 1992

SHOREHAM DSAR

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postulated accident.and environmental conditions. A switchyard common to both circuits is acceptable. Each of these circuits shall be_ designed _to be available in sufficient time following a loss of all onsite. alternating current power supplies and the other offsite-. electric power circuit, to assure that specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded. One of these

-circuits shall be designed to be available within a few seconds

'following a loss.of coolant accident to assure that core cooling, '

containment integrity, and other vital safety functions are maintained.

Provisions shall be included to minin'te the probability of losing electric power from any of the remaining supplies as a result of, or coincident with, the loss of power generated by the nuclear power unit, the loss of power from the transmission network,.or the loss of power from the onsite electric pnwer i supplies.

Desion'Conformance The criterion applies principally to the design of an operating reactor. As-demonstrated in DSAR Chapter 15,-active system are

[T N/

not-required to provide cooling or makeup functions in the event of postulated accidents including a seismic event. However, 4

operability of the electric power system will be required by Technical Specifications during fuel movement to provide for a controlled and monitored release capability in the event of a fue) drop accident. One offsite power transmission system will be maintained to provide power for support system operation. In addition, blackstart: combustion turbines exist nearby at Shoreham-West to provide reliable power in the unlikely event of a loss-of-offsite power occurs. One non-safety-Emergency Diesel Generator-Will be provided'during fuel handling operations. A further discussion-of electric power requirements can be found in Chapter 8.

Inspection and Testina of Electric Power Systems (Criterion 18) '

m criteriou Electric power systems important to safety shall-be designed to

-permit appropriate periodic inspection and testing of important >

areas and features, such as wiring, insulation, connections,_and suitchboards, to assess.the continuity of-the systems and the

-conditions of their components. The systams shall be designed with'a capability _to test periodically (1) the operability and l ((3):

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.SHOREHAM DSAR ry

.b functional performance of the components of the systems, such as onsite power sources, relays, switches, and buses, and (2) the operability of_the systems as a whole and, under conditie.is as close to design as practical, the full operation sequence that brings the systems into operation including operation of applicable portions of the protection system, and the transfer of power among the nuclear power unit, the offsite power system, and the onsite power system.

Desian Conformance Electric Power Systems will be tested and inspected in accordance with SNPS operating procedures and Technical Specifications. See Criteria 17 response.

Control Room (Criterion 191 Criterion A control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions,

-including loss of coolant accidents. Adequate radiation

-protection shall be provided to permit access and occupancy of j(_s)_ the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident.

Equipment at appropriate locations outside the control room shall be provided (1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to' maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures.

D_esian Conformance A control room is provided and equipped to_ operate the plant safely under normal and accident conditions.

Based on the results of radiological analyses provided in DSAR

-Chapter 15' control room shielding and ventilation functions are not required for the mitigation of postulated accidents.

Instrumentation available in the control room for accident monitoring and support system control are described in DSAR Chapter 7.

-/~T 3-7 Rev. 4 July 1992

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SHOREHAM DSAR f"N -  !

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Coolina Water (Criterion 44)

Criterion A system.-_to transfer heat from structures, systems, and

. components impsrtantLto safety, to an ultimate heat sink, shall be provided. The system safety function shall be to transfer the

. combined heat load of these structures, systems, and components under-normal operating and accident conditions.

Suitable redundancy in components and features, and suitable

. interconnections, leak detection, and isolation capabilities

.shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power operation (assuming onsite power is not available) the~ system's safety function can be accomplished, assuming a single failure.

Desian Conformance As demonstrated in Chapter _15 of this report, active cooling of the spent fuel pool is not required based on the low heat generation rate of-the low burnup spent fuel. Service varer and

.fs- other' support systems are expected to be normally available to

(' -

.provide plant building services; however, these systems do not fulfill a safety function, n Inspection of Coolina Water System

-(Criterion 45)

Criterion

_The cooling water system shall be designed to permit appropriate

-periodic inspection of-important components, such as heat exchangers and piping, to assure the integrity and capability of the system.

Desian Conformance The service water system which Fill be maintained functional is

. designed.to permit appropriate visual inspection in order to assurelthe integrity of system components. See Criterion 44 response.

l

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3-8 Rev. 4 July 1992 -

[JY

SHOREHAM DSAR Testina of Coolina Water System (Criterion 46)

Criterion The cooling water system shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leaktight integrity of its components, (2) the operability and parformance of the active components of the system, and (3) the operability of the system as a whole and, under' conditions as close to design as practical, the performance of full operational sequence that brings the system into operation for reactor shutdown and for loss of coolant accidents, including operation of applicable portions of the protection systems and the transfer between normal and emergency power sources.

Desian Conformance See Criterion 44 response.

Control of Releases of Radioactive Materials to the Environment q (Criterion 60)

Criterion The nuclear power unit design shall include means to control suitably the release of radioactive materials in gaseous and liquid effluents and to handle radioactive solid wastes produced during normal reactor operation, including anticipated operational' occurrences. Sufficient holdup capacity shall be provided for retention of gaseous and liquid effluents containing radioactive materials, particularly where unfavorable site environmental conditions can be expected to impose unusual operational limitations upon the release of such effluents to the environment.

Desian Conformance Because SNPS is not in normal operation, effluent releases are due primarily to maintenance of the spent fuel pool water quality. Means are provided to control and/or hold up the release of liquid and gaseous effluents as required. Fuel pool Ecleanup and appropriate radwaste' systems are provided and are

' described in Chapters 9 and 11. See also Criterion 61.

l

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s/  ;-9 Rev. 4 July 1992 I-1 i i i l

- . - . _- - . _ = -

SHOREHAM DSAR p-Fuel Storace and Handlina and Radioactivity Control (Criterion 61)

Criterion The fuel storage and handling, radioactive waste, and other systems _which may contain radioactivity shall be designed to assure adequate safety under normal and postulated accident conditions. These systems shall be designed, (1) with a capability to permit appropriate periodic inspection and testing et-components important to safety, (2) with suitable shielding for radiation protection, (3) with appropriate containment, confinement, and filtering systems, (4) with a residual heat removal capability having reliability and testability that reflects the importance to safety of decay heat and other residual heat removal, and (5) to prevent significant-reduction in fuel storage coolant inventory under accident conditions.

Desian Conformance Fuel Storace and Handlina I

The low burnup SNPS spent fuel is to be stored in the spent fuel storage pool located in the reactor building. The fuel racks and fuel pool structure are Seismic Category I. Systems required for safe fuel storage will be subject to appropriate inspection and testing requirements.

-Adequate shielding is provided by maintaining a minimum water depth over the active fuel. Dose rates at the refueling level without the effects of shielding were~ calculated to be approximately 1R/HR.

The'SNPS~ Secondary Containment is a Seismic Category I controlled leakage building surrounding'the fuel pool facility. The Reactor Bui'lding Normal Ventilation System (RBNVS) will be used to provide ventilation and a monitored release pathway, _ Because the gas activity-present in the fueJ and available for release is primarily noble gas (Kr-85), the-filtering role of the Reactor Building Standby Ventilation System (RBOVS) is not required.

Certain components of the RBSVS are needed to support operation of the RBNVS. These components will remain functional to provide

-these services. As discussed in Chapter 15, credible potential releases fromLaccidents are small in comparison to 10CFR100 limits, and neither the Reactor Building Standby Ventilation System nor secondary containment integrity-is reraired to reduce

, . offsite doses ~due to postulated accidents.

( 1 3-10 Rev. 4 July 1992 L m

. .m . ._ . . . . ~ _ _ _ _ _ . _ _ _ ,

b f>

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.4. SHOREHAM DSAR

&' j V  :

s. ...-

2 LRadiationLaonitoring is provided as described in Chapter 11 and  ;

12ltofdetectiradiological releases.

~

Because ofcthe extremely low residualiheat load (approximately

. 550 watts); nssociated withL the SNPS ' spent ~ fuel, active - fuel pool cooling is uoc required. Reliable-fuel pool makeup sources including condensate storage, demineralized water, and fire

= protection water,-.are. capable of-maintaining; pool water inventory

• toLcompensate for: evaporation. Chapter 9 conta' ins a complete "

discussion of makeup requirements.

- Th'e1.fuelipool isla Seismic category I structure. Systems that connect _to the pool 1(fue1Lpool cooling, fuel pool cleanup, etc.)

- have-been designed-to minimize the potential for draining of the pool inventory. _High and-low level alarms indicate pool water

= level changes in the: main control room.

-Radioactive Waste Systems ,

The~ radioactive-wasteLsystems-provide all equipment necessary to "

collect,. process, and prepare for disposal of all' radioactive

- liquids:and solid wasto produced as a. result of' spent fuel storage.: _The off-gas-system is not-needed._ Any Krypton 85 will

~

VD :be: retained within the-fuel cladding.- Should' pin-hole leaks -

\/ develop,_the'_ gases will be-handled by the-ventilation systems.

They will be discharged to atmosphere via the_ main plant vent.

The-radiological consequences 1of this type of release are

negligible. i

-- This accidentois bounded (by-the analysis of the Fuel Handling Accident ~(Section 15.1.36).

Liquid radwastestare: collected, classified, and treated as high conductivity,Llow? conductivity,_ chemical ors laundry wastes.

Processing includes filtration, i o n e x c h a n g e , .a n a l y s i s ,: and

dilution.; WetLsolid. wastes =are packaged.in steel containers or polyethylene-high' integrity containers. ; Dry solid.radwastes are compressedzand/or packed in steel drums or boxes.

JAccessibleLportions-of..the spent fuel pool area and-radwaste-L building ~have sufficient shieldJng'to maintain dose rates within

the;limitsJset forth in110CFR20.and 10CFR100. The radwaste bbilding isEdesigned to preclude accidental release of

' radioactive materials to the environs above those allowed by the applicable ~ regulations.:

The fuel: storage and-handling and radioactive' waste systems-are

' designed to assure' adequate safety under normal and postulated accident conditions._-'The design of these sy_ stems meets the i requirements of Criterion 61. H 3-11 Rev. 4 July 1992

.l g 4 I

n > m _ -

a

SHOREHAM DSAR

-U-Radwaste systens are designed to meet the limits for effluents set forth in 10CFR20-and 10CFR50.

Prevention of Criticality in Fuel _Storaae_ Hand 11na (Criterion 62)

Criterion Criticality in the fuel storage and handling system shall be prevented by physical systers or processes, preferably by use of geometrically safe configurations.

Desian Conforpance Appropriate plant fuel handling and storage facilities are provided to preclude accidental criticality for new and spent fuel. Criticality in spent fuel storage is prevented by the geometrically safe configuration of the storage rack. There is

-sufficient spacing between the assemblies to assure that the array,. when'ful]y loaded, is subtrantially subcritical. Fuel elements-areJ11mited by rack design to only top loading and designated fuel assembly positions.

Spent fuel is stored under water.in the spent fuel storage pool.

The racks in which spent fuel assemblies are placed are designed and arranged-to ensure subcriticality in the storage pool. Spent fuel is maintained at a suberitical multiplication factor k-eff of less than 0.95 for-both normal and abnormal storage conditions.

The fuel handling- system is designed to provide a safe, effective means of transporting and-handling fuel and to minimize the possibility of mishandling or misoperation.

The use:of: geometrically safe configurations for new and spent fuel-storage and the design of fuel handling systema precludes accidental criticality in accordance with. Criterion 62.

For further. discussion, see the following secti:in:

Section 9A Criticality Analysis 3-12 Rev. 4 July 1992-(J l i

i

SHOREHAM DSAR-

&,~)

. ,l

_Monitorina Fuel and Wasto Storage (Criterion _111 Criterion

' Appropriate systems shall be providedfin fuel storage and radioactive' waste syrtems and associated handling areas, (1) to detect conditions that may result in loss'of residual heat removal capability and. excessive radiation levels, and (2) to .

I initiate appropriate safety actions.

Design Conformance Appropriate. systems have been provided to neet the requirements of this criterion. A malfunction of the fuel pool cleanup system is alarmed in the main control room.- It is also alarmed in the

-radwaste control room on high pressure differential. Alarmed conditions include high/ low fuel pool level. The refueling level ventilation exhaust radiation monitoring system detects abnormal ,

. amounts of radioactivity. As demonstrated in Section 9A and

--Chapter 15 active--cooling of the spent _ fuel pool is not required because of the low heat generation rate.

Area radiation and sump levels are monitored _and alarmed to give

('/

(-

N. indication-of conditions that may result in excessive radiation levels in the fuel storage and radioactive waste system areas.

These systems satisfy the requirements of Criterion 63.

Monitorina Radioactivity Releases (Criterion 64)

Criterion

'?!eans shall be provided for monitoring the. reactor containment atmosphere, spaces containing components for recirculation of

~

loss of. coolant accident fluids, effluent discharge paths, and the plant: environs for radioactivity that may be released from

' normal; operations, including anticipated operational occurrences, andffrom postulat'ed accidents.

Desian conformance- t Means have been provided for monitoring radioactivity releases resulting from-normal and_ anticipated operational occurrences.-

The following station-release pathways are monitored:

1. Gaseous releases from-the station ventilation exhaust ,

2._ Liquid discharge to.the discuarge tunnel-m 3-13 Rev. 4 July 1992

. ( )=

l J

, i t >

~ - _ . _ . ._. .~ - - -

SHOREHAM DSAR

<(s- -

Radioactivity levels in the normal plant effluent discharge paths and in the environment are continually monitored during normal conditions by the various radiation monitoring systems and by the offsite radiological environmental monitoring programs.

The semiannual Effluent Release Report is submitted to the NRC.

This report includes specific information on the quantities of the principal radionuclides released to the environment.

Additional-discussion of radiation monitoring is contained in Chapters-11 and 12.

3.2 CLASSIFICATION OF STRUCTURES, SYSTEMS AND COMPONENTS Seismic Category I structures, systems, and components are those necessary to ensure:

1. The integrity of the reactor coolant pressure boordary

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

(,, 3. The capability to prevent or mitigate the consequences ofaccidents that could result in potential offsite exposures I comparable to the guideline exposures of 10CFR100, i

Criteria 1 and 2 do not apply to a defueled reactor with

-respect to the storage and handling of low burnup Shoreham cpent fuel. A set of postulated accidents has been identified and analyzed in Chapter 15 of this report that defines the potential for a radiological release. Based on this analysis it has been concluded that potentis' radiological releases are far below the exposure limits of 10CFR100. The analysis in Chapter 15 of this report assumes that the structural integrity of the filled fuel pool, fuel pool' liner, reactor building structure and fuel racks-together form a passive safety system that requires a seismic Category I designation. The Category I designation has been l

maintained for fuel handling equipment as well, i

L A reclassification of structures, systems, and components is provided in DSAR Table 3.2-1. Table 3.2-1 supplements the information provided in USAR Table 3.2.1-1. The quality group classification in USAR Table 3.2.1-1 reflects the original design basis. As analyzed in Chapter 15, active cooling of the spent fuel pool is not required and pool makeup requirements are minimal. Supporting systems are 7s 3-14 Rev. 4 July 1992

)

1 SHOREHAM DSAR r~N j N-required to maintain building habitability, provide radiation monitoring capability, and normal operating service functions.

Design Basis Earthquakes (DBE) and Operating Basis Earthquakes (OBE) are described in the Shoreham USAR Section 2.5.

Structures, systems, and components whose safety functior' require conformance to the quality assurance requirement > of 10CFR50, Appendix B, are summarized in Table 3.2-1 under the heading, LIPA Quality Assurance Category, with the notation I.

Modifications to QA Category II equipment and components at and above the 175' elevation in the reactor building shall be designed to withstand the DBE without failing in a manner that would result in an unacceptable impact to the spent fuel integrity or unacceptably damage the spent fuel whereby r4blic health and safety concern could be created.

A key of definitions is provided at the end of Table 3.2-1.

Chapter 17 discusses the graded level of Q.A. requirements I' for this equipment.

1

\w,)

3.3 WIND AND TORNADO LOADING The information contained in the USAR remains the same although the requirements to protect safe-shutdown equipment no longer exists.

3.4 WATER LEVEL (FLOOD) DESIGN The design of flood-protected structures remains the same although the requirements to protect safe-shutdown equipment no longer exist.

3.5 MISSILE PROTECTION

~

The deeign information contained in this section is unchanged.

However the spent fuel pool is the only area of the plant requiring mitsile protection.

3.6 PROTECTION AGAINST DYNAMIC EFFECTS ASSOCIATED WITH POSTULATED RUPTURE OF PIPING l

j In the defueled state high ener7y piping systems inside primary l

containment listed in USAR Table 3.6.1A-1 are no longer

/s pressurized and thus piping rupture need not be postulated.

~

3-13 Rev. 4 July 1992

SHORERAM DSAR O

3.7 SEISMIC DESIGN Seismic design methods remain the same; however, hydrodynamic load effects resulting from safety relief valve discharge and loss-of-coolant-accidents are no longer applicable for a defueled reactor.

3.8- DESIGN OF SEISMIC CATEGORY I STRUCTURES The design methods for seismic Category I structures such as the reactor building will remain as described in USAR Section 3.8 except that Safety Relief Valve (SRV) and LOCA hydrodynamic loads are no longer applicable to a defueled reactor.

3.9- MECHANICAL SYSTEMS AND COMPONENTS This section addresses methods and procedures used to qualify mechanical equipment. The information contained in this section is relevant only to reactor operating conditions and is, therefore, not applicable to the DSAR.

In the future,. mechanical equipment will be accorded the safety significance demonstrated by the classification in Table 3.2-1 of

()

bs ,/

the DSAR.

3.10 SEISMIC QUALIFICATION OF SEISMIC CATEGORY I INSTRUMENTATION AND ELECTRICAL EQUIPMENT Seismic Category I equipment is identified in Table 3.2-1 and is limited to structures and equipment' required to maintain the integrity of the fuel;in the spent fuel pool. As discussed in Section 3.2,.only the Reactor Building, fuel pool, fuel racks, and-fuel' handling equipment are required to be Seismic Category I. The instrumentation described in USAR Section 3.10 is no lon.ger required to be seismically qualified.

3.11 ENVIRONMENTAL DESIGN OF MECHANICAL AND ELECTRICAL EQUIPMENT Electrical Eculoment Environmen'tal Oualification Purnose The purpose of the Electrical Equipment Environmental Qualification Program for Shoreham is to provide assurance that electrical equipment important to safety as defined by 10CFR50.49 located in potentially harsh environments maintains functional operability when required to mitigate the consequences of a

) 3-16 Rev. 4 July 1992 o

d- L

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SHOREHAM DSAR n-.

N.A postulated accident or to bring the plant to a cold shutdown condition afterward. Since the fuel has been removed and stored in.the-fuel pool, LOCA or HELB cannot occur (see Chapter 15), and there is~no potential for creation of harsh environment.(i.e.,

the remaining design basis accidents discussed in Chapter 15 do not result in harsh environments). Based on these conditions, 10CFR 50.49 is not applicable, therefore the enviroa-2ntal

. qualification program is not required. Environmentally qualified electrical equipment will be designated Q.A. Category II.

3.12 SEPAF^ TION CRITERION FOR SAFETY RELATED MECHANICAL AND ELECTkICAL EQUIPMENT The systems described in this section are no longer required to fulfill a safety related function regarding the storage of spent fuel. Thus, there no longer exists a need to maintain separation criteria for these systems. Q.A. Category I equipment will be designated Q.A. Category II.

u( ) 3-17 Rev. 4 July 1992 I

L L - . _

SHOREHAM DSAR

'ej .

3A Computer Procrams for the Stress Analysis of Catecorv I Structures. Dynamic and Static Analysis, and Dynamic and Stress Analysis of Seismic Cateaorv I Pipina-Systems The description contained under this heading in the latest revision of_Shoreham USAR. remains unchanged. Refer to USAR for information on this subject.

3B NRC Reculatory Guides This section is described in the USAR. Specific topics are

. covered elsewhere in this DSAR.

3C Pipe Failure Outcide Primary Containment In the defueled state, piping systems outside primary containment which'were considered high energy systems are no longer pressurized. Pipe rupture need no longer be postulated.

/ T.

V

, ) 3-18 Rev. 4 July 1992

i r

SHOREHAM DSAR Y)

TABLE 3.2-1 EOUIPMENT ChASSIFICATION-EEp1T FUEL STORAGE LIPA QUALITY SYSTEM / ASSURANCE SEISMIC COMPONENT- CATEGORY CATEGORY COMMENTS I. Reactor System II N/A NR II NuclearfBoiler II N/A NR III Recirculation System II N/A NR IV Control Rod Drive Hydraulic System II N/A NR .j V Standby Liquid Control System II N/A NR  ;

VI Neutron Monitoring II N/A NR f VII Reactor Protection II N/A NR VIII Fixed Process. II N/A (1)

Airborne, and Effluent Radiation Monitors

( IX RHR. II N/A NR v__/

JC. . Core Spray: II N/A NR XI HPCI II N/A NR XII 'RCIC II N/A NR XIIIJFuel' Service Eauipment

1. Fuel preparation machine I. I i

-2.-General purpose grapple ' I I.

iXIV ' Reactor -Vessel Service

• Eauipment

1. System Line Plugs II N/A NR

2. Dryer & Separator sling and RPV-head strongback I I

3. Drywell head' lifting.

rig- I I 1 of 7 Rev. 4 July 1992-LI r

SHOREHAM DSAR TABLE 3.2-1 j/j^1. . (Continued)

EOUIPMENT CLASSIFICATION SPENT FUEL STORAGE LIPA QUALITY SYSTEM /- ASSURANCE SEISMIC COMPONENT CATEGORY CATEGORY COMMENTS

'XV In-vessel Service Eauipment 1.-Control rod grapple I I XVI Befuelina'Eauipment

1. Refueling platform I I (4).

2. . Refueling bellows, drywell- II N/A

3. Refueling bellows, cavity reactor II N/A

4. New Fuel Inspection

' Stand II N/A NR

' tp.,VII Storace Eauipment N,/

.1, New Fuel Storage Racks II N/A NR

2. Defective fuel.

storage container I I

3. Spent fuel pool, dryer /sep. pool,

-reactor-cavity liners I I 4.- Spent: fuel storage racks I I XVIII-Radwaste System II N/A XIX Reactor Water Cleanuo System II N/A NR XX - Fuel Pool Cleanuo Subsystem ,

1. Demineralizer vessel- II N/A

2. Filters II N/A 33 Pumps,_ purification

& transfer II N/A 4.. Piping II N/A -

5.- Valves II N/A 6.. Tanks, backwash

-storage and air

(~ accumulator II N/A

\_

2 of 7 Rev. 4 July 1992

k-.

-SHOREHAM DSAR r-(- TABLE 3.2-1

() (Continued)

EOUIPMENT CLASSIFICATION-SPENT FUEL STORAGE LIPA i QUALITY SYSTEM /. ASSURANCE SEISMIC COMPONENT CATEGORY CATEGORY COMMENTS XXI Fuel Pool Coolina Subsystem

1. Piping II N/A

2. Valves II N/A

. XXII Control Room Panels

1. Electrical modules II N/A

2. Cable II N/A XXIII. Local Panels i

1. Electrical modules II N/A

2. Cable II N/A

( iXIV Offaas System II N/A NR

'v_)

XXV-_ Service Water System II N/A XXVI Comoressed Air System II N/A XXVII Onsite Power Systems (USAR safety related)

a. Diesel Emergency Power Systems II N/A -(2)

b. AC Power Systems II N/A

c. Containment-Elec-

- trical Penetrations: II 'N/A- NR

d. Fire Stops II N/A

o. ~ DC Power Systems II N/A XXVIII Primary Containment Atmospher'e II N/A NR Control

-XXIX a) Reactor Buildinal Normal Ventilation II N/A b)' Reactor Buildina

-Standby Ventilation II N/A NR*

• "Certain. components such as fans and valves will .* in fy functional to support RBNVS operations.

.%/

3 of 7 Rev. 4 July 1992

SilOREllAM DSAR TABLE 3tl-1 h- (Continued)

EOUIPMENT CLASSIFICATIQH <

LPENT FUEL STORAGE LIPA QUALITY SYSTEM / ASSURANCE SEISMIO COMPONERI CATEGORY CATEGORY COMMENTS XXX Primary containment _2Mrgg II N/A NR XXXI Power Conversion II N/A NR  ;

XXXII Condeng31o Storace'and Transfer 11 N/A XXXIII Energ.gncy Sunnott Facilitie.g

1. TSC Bldg. II I

2. EOF II N/A HR(3)

3. OSC II N/A XXIV MSIV Leakane ContIgl II N/A NR O

L IXXV Miscellaneoua

1. FB Polar Crane I- I (4)

2. ECCS Loon Level II N/A NR XXXVI Eeactor_Buildinq glosed Loon Coolina II N/A NR XXXVII Eculoment and Flool Drains II N/A XXXVIII Miscellaneous Ventilation Eygtems 1, 125 Volt.DC Battery

' N/A room H & V_ II 2e Screenwell pumphouse

!!&V II N/A

3. Relay and emergency switchgear HM' II N/A

4. Control room ait con-ditioning,-incitiing

' filter trains II N/A

5. Diesel generator room II N/A S --

ventilation

L) 4 of 7 Rev. 4 July 1992

e SHORENAM DSAR  !

7 IAHLE_3.2-1 i (Continued)

EOUIPMENT CLASSIFICAT10lf SPENT FUEL STORAGE i l

LIPA '

QUALITY l SYSTEM / ASSURANCE SEISMIC

,Q.QMP_QllEllT CATEGORY CATEGORY COMMENTS XXXIX Area Radiation Monitorina ,

SXiltRE

1. All components II N/A

2. High Range Area II N/A NR XL Leak Detection System II N/A NR XLI Fire Protection System

1. Water spray deluge II N/A systems

2. Sprinklers, carbon dioxide systems II N/A

3. Portable and wheeled extinguishers II N/A

~}

XLII Sivil Structures

1. Reactor building I I

2. Office and service building II N/A

3. Screenwell II N/A

4. Control building II N/A (5)

5. Turbine. building II N/A (5)

6. Intake Canal II N/A

. bischarge tunnel II N/A

3. Discharge pipe and diffuser II N/A 9..Radwaste Building II N/A (5)

10. Auxiliary boiler and '

MG. set building II N/A L 11. Biological shielding II N/A (5)

12. Missile barriers II N/A

13. Waterproof doors- II N/A '

14. Site grading II N/A

( 15. Masonry walls (RB). II N/A (5) l 16. Masonry walls (non-RB) II N/A 7

5 of 7 Rev. 4 July 1992

1 SHOREHAM DSAR E E 3.2-1

.3's,{~~'

(Continued)

EOUIPMENT CLASSIFICATIOH SPENT J_,UEL STORAGE LIPA QUALITY SYSTEM / ASSURANCE SEISMIC COMPONENT CATEGORY CATEGORY COMMENTS XLIII Primary containment Structur.g II N/A 5 XLIV Safety-Parameter Displav system II N/A NR XLV Post Accident Sample System II N/A NR XLVI Containment Isola-tion Valve Posit, ion..

Indicator II N/A NR XCVII Accident Monitorina II N/A NR Instrumentation IHUREG 0578) l-l

(

v 6 of 7 Rev. 4 July 1992

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-SHOREHAM DSAR r"' TABLE 3.2-1

( ,S) (Continued)

ETE Ouality Assurance Cateaoryl I - Meets 10CFR50 Appendix B requirements (same as USAR).

.II - Meet requirements of industrial and engineering standards (commercial grade quality).

Seismiu cateaorv

.I - Equipment is designed in accordance with the seismic requirements for the DBE/OBE.

H/A. - Seismic requirements for DBE/0BE earthquake are not applicable to the equipment.

Comments:

NR - Not required (System secured from service or not required to support safe storage or handling of epent

() fuel).

Seismic events will not create a radiological (1) -

release due to passive protection provided by the spent fuel pool.

(2) - ~ Loss-of-offsite power will not create the potential for a radiological release as discussed in Chapter 15.

One emergency diesel generator will be maintained non-safety related operable, as required by Technical

~

Specifications:during fuel movement.

(3) -

Based on LIPA Defueled Emergency Preparedness plan, the EOF is not required.

(4) - Only structurally' safety related.

(5) .

Originally constructed as Seismic Category I; modifications will be analyzed for DBE to ensure

-integrity of Reactor Building.

l' l 7 of 7 Rev. 4 July 1992

SHOREHAM DSAR CHAPTER 4 REACTOR This Chapter includes reactor description, mechanical design, nuclear design, thermal and hydraulic design, reactor materials and control rod drive housing supports. In the plant's defueled condition, the fuel is not in the core and the reactor is '

depressurized. All sections of this Chapter are, therefore, not applicable to the DSAR. Fuel storage is addressed in DSAR Chapter 9. In particular, Section 9A addresses criticality and Section 9B addresses fuel pool make-up requirements.

4.1 REACTOR

SUMMARY

DESCRIPTION The HSS system is no longer needed for the de'tueled condition and i hence is depressurized.

4.1.1 -Beactor Vessel  ;

The reactor vessel design and description are covered in USAR Section 5.4.

4.1.2 Reactor Internal Comnonents

)

The reactor internal components are as described in the USAR.

The fuel rods and control rods are removed from the reactor.

4.1.3 Reactivity Control System This system is no longer needed as there is no fuel in the reactor vessel.

4.1.4 Analysis Techniaues The description contained under this heading in the latest revision of the USAR is no longer relevant in the plant's defueled condition.

- 4.4 THERMAL AND HYDRAULIC DESIGN The linear heat. generation rate (LHGR) limit of 13.4.kw/ft will not be' exceeded by the decaying fuel in the spent fuel pool.

Justification.for this limit can be found in Appendix A, of ,

General Electric Standard Application for Reactor Fuel (GESTAR II).

O 4-1 Rev. 4 July 1992

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SHOREHAM DSAR

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C .

4.5 REACTOR MATERIALS Neither the Control Rod System or Reactor Internal materials are of importance to the defueled plant conditions.

4.6 CONTROL ROD DRIVE HOUSING SUPPORTS There is no fuel ir the vessel in the defueled state and hence this system is not of concern.

O 4-2 Rev. 4 July 1992 l

SHOREHAM DSAR CllAPTER 6 ENGINEERED SAFETY FEATURES 6.1 GENERAL Because of the Defueled Plant Configuration, there is no longer a need for engineered safety features (ESP) systems at Shoreham.

This is substantiated by a review of the Design Basis Accidents and Postulated Transients. These are covered in Chapter 15.

This chapter discusses the effect of radiological accidents in the Secondary Containment. The Secondary containment is utilized for maintaining a controlled and monitored release point for the design basis accident, the Fuel Bundle Drop accident. In addition, a worst case release of the entire gaseous inventory of the fuel is postulated in Chapter 15 that bounds any possible ,

large scale mechanical-damage svent.

-6.2 CONTAINMEMT SYSTEMS 6.2.1 Containment Functional Desian 6.2.1.'1 Desian Basis 6.2.1.1.1 Safety Criteria The primary containment system is not req'Jired and will not be maintained functional as there will be no fuel within the primary

~

containment structure. The secondary containment will maintain a subatmospheric pressure for postulated radiological accidents to assure radiological monitoring of building releases. It is not needed to mitigate-the consequences of an accident.

6 . 2 .~ 1.1. 2 Desian Basis Accidents The major design basis accident identified which will affect the secondary containment is the_ Fuel Handling Accident (Fuel Bundle Drop). The results of this accident from a radiological standpoint are presented in Chapter 15. There are no pressure and-temperature effects of this accident and the RBNVS would continue to maintain a subatmospheric condition.

The other event which would have an effect;on the secondary containment is the loss of normal AC.

A loss of normal AC power _may_ result in loss of the subatmospheric conditions within the secondary containment and a-loss of spent fuel-pool water makeup capability. However, as

explained-in Chapter 15, should the loss of AC power occur as 6-1 Rev. 4 July 1992

4 SHOREHAM DSAR

.(^T-() part of any event which results in fuel damage, while the radioactive release to the atmosphere would not be monitored, the offsite dose consequences to the public would be insignificant.

With regard to loss of spent fuel pool water makeup capability, evaporative loss would be so slow that corrective action would be taken before loss of shielding is significant. There are no radiological consequences associated with the loss of normal AC power.

6.2.1.2 System Design The reactor building, which completely encloses the primary containment and acts as the secondary containment, is maintained at subatmospheric pressure by the RBNVS.

6.2.1.3 Desian Evaluation This entire subsection is not applicable as it deals with the  :

primary containment which is no longer maintained.

6.2.2 Containment Heat Removal System This subsection is not applicable as it deals with the primary containment whichois no longer maintained.

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L

. 6.2.3 Containment Air Purification and Cleanun Systems This subsection is not applicable as it deals with the filtration portion of the RBSVS which is no longer required.

6.2.4 gontainment Isolation System This subsection is no longer applicable as it deals with the primary containmentLisolation system. The primary containment is-no longer maintained.

6.2.5 Combustible Gas Control in Containment This subsection is no longer: applicable as.it is concerned with hydrogen combustion inside the primary containment.

6.3 EMERGENCY CORE COOLING SYSTEMS TheLemergency core cooling systems protect the core against hypothetical pipe breaks of various sizes. In the plant's present state, the fuel is not in the core and the reactor is depressurized.. Therefore, pipe breaks are not postulated and the emergency' core cooling-systems are not required and this section

'is not applicable to DSAR.

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6.3.2.2.3 ggrg Spray System The Core Spray (CS) System is described in the USAR. In the defueled status of the Shoreham Nuclear Power Station the CS System serves no function and is no longer maintained.

6.4 HABITABILITY SYSTEMS -

The systems, aside from the control room air conditioning portion, are no longer maintained because they are not needed since the fuel is stored in the spent fuel pool. The control room air-conditioning system is described in Section 9.4.1.

6.5 MAIN STEAM ISOLATION VALVE LEAKAGE CONTROL SYSTEM The main steam isolation valve-leakage control system (MSIV-LCS) is not required in the defueled state and is, therefore, not included in the DSAR.

6.6 OVERPRESSURIZATION PROTECTION gg The overpressurization protection system is not required in the defueled state and is,.therefore, not included in the DSAR (See

. ] Chapter 5. of DSAR).

6.7 MAIN STEAM LINE ISOLATION VALVES The main steam isolation valves (MSIVs) are not required in the defueled-state and are, therefore, not included in the DSAR (See Chapter 5. of DSAR).

6.8 CONTROL ROD DRIVE SUPPORT SYSTEM The control rod drive support system is not required in the defueled state and is, therefore, not included in the DSAR (See 1 Chapter 4 of DSAR).

a6 . 9 CONTROL ROD VELOCITY LIMITERS

.The-control. rod velocity limiters are not required in the

'defueled state and this Section is, therefore, not included in the DSAR (See Chapter _5 of DSAR).

6.10 MAIN STEAM LINE FLOW RESTRICTORS 1

The main steam line. flow restrictors are not required in the defueled state.and this Section is,-therefore not included in the DSAR,- (See. Chapter 5..of DSAR).

L 6-3 -Rev. 4 July 1992

SilOREllAM DSAR a-.

6.11 REACTOP CORE ISOLATION COOLING-SYSTEM The RCIC system is not required in the defueled state and is, therefore, not included in the DSAR (See Chapter 5. of DSAR).

6.12 STANDBY LIQUID CONTROL SYSTEM The standby liquid control system.is not required in the defueled state-and is, therefore, not included in the DSAR (See Chapter 4 of DSAR).

l O

O 6-4 Rev. 4 July 1992

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'\-) 7.1.1.1.6 Reactor-Manual control System This system is not needed to support the storage of the fuel in the fuel pool, therefore it is not included in the DSAR.

7.1.1.1.7 Reactor Vessel InstrumentatinD This system is not needed to support the storage of the fuel in the fuel pool, therefore it is not included in the DSAR.

7.1.1.1.8 Reactor Recirculation System This system is not needed to support the storage of the fuel in the fuel pool, tlerefore it is not included in the DSAR.

7.1.1.1.9 Feedwater Control System This system is not needed to support the storage of the fuel in the fuel pool, therefore it is not included in the DSAR.

7.1.1.1.10 Pressure Reculator and Turbine-Generator Controlg

-This system is not needed to support the storage of the fuel in the fuel pool, therefore it is not included in the DSAR.

/~T -

(_s/ 7.1.1.1.11 Egpote Shutdown System This system is not needed to support the storage of the fuel in the fuel pool, therefore it is not included in the DSAR.

7.1.1.1.12 Screenwell Pumnhouse Ventilation System The screenwell pumphouse ventilation system instrumentation and controls remain-functional and are designed to ventilate cach of the two rooms of the building using separate, 100 percent outside air ventilation systems.

-7.1.'1.1.13- Process' Computer System

=This. system is-not needed to support the storage of the fuel in the fuel pool, therefore it is not included in the DSAR.

7.1.1.1.14- Reactor Core Isolation Coolina System

/This system is not needed to support the storage of the fuel in

~the fuel-pool, therefore it is not included in the DS;.R.

~7.1.1.1.15 Standby Liauid control System This system is not needed to support the storage of the fuel in

()

7s the fuel pool, therefore it is not included in the DSAR.

July 1992 7-2 Rev. 4

l SHOREHAM DSAR

') l (V 7.1.1.1.16 Reactor Water Cleanun System This system is not needed to support the storage of the fuel in the fuel pool, therefore it is not included in the DSAR.

7.1.1.1.17 Leakaue Detection System This system is not needed to support the storage of the fuel in the fuel puol, therefore it is not included in the DSAR.

l 7.1.1.1.18 Reactor Shutdown Coolina Mode-RHR fvstem This system is not needed to support the storage of the fuel in '

the fuel pool, therefore it is not included in the DSAR.

7.1.1.1.19 Radwaste Syste,m Radwaste system instrumentation.and controls support manual processing and disposing of the radioactive procesa wastes.

7.1.1.1.20 Emeroency Diesel Generators

.This system is utilized to provide backup emergency power. Osia emergency diesel generator will be operable when fuel is being

) handled in the secondary containment.

7.1.1.1.21 Turbine Buildina Closed Loon Coolina Water System The turbine building closed loop. cooling water (TBCLCW) system instrumentation and controls remain functional to maintain the turbine building cooling water system at design temperature and nonitor system performance. The TBCLCW system also cools the equipment in the radwaste building and supports the station air Compressors.

7.1.1.1.22 Service Water System The service water system provides cooling for the plant components.- Instrumentation and controls for this system are provided to operate the system $n accordance with Section 9.2.

7.1.1~1.23 Recirculation Pumn Trio System This system is not needed to support the storage of the fuel in

.the fuel pool, therefore it is not included in the DSAR.

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I SHOREHAM DSAR v'O 7.1.1.1.24 Reactor Buildina Standbv Ventilation System The filtration portion of the system is not needed to support the storage of the fuel in the fuel pool. Certain fans ans air operated valves will remain functional to support RBNVS operation. See DSAR section 9.4 for additional information.

7.1.1.1.25 Reactor Buildina Closed Loon Coolina Water System This system is not needed to support the storage of the fuel in the fuel pool, therefore it is not included in the DSAR.

7.1.1.1.26 Primary Containment Atmosnheric control System '

This system is not needed to supporo the storage of the fuel in the fuel-pool, therefore it is not included in the DSAR.

7.1.1.1.27 Fuel Pool Coolina and Cleanun Systems Fuel pool cooling and cleanup systema instrumentation and controls remain unchanged except that the cooling portion is not required because evaporative cooling is sufficient to remove the small amount of decay heat.

7.1.1.1.28 CJ2ntrol Room Air Conditionino System The control room air conditioning (CRAC) system instrumentation and controls for one of the two redundant subsystems are functional to maintain the main control room at design temperature during normal and emergency conditions, monitor system performance, and permit manual as well as automatic initiation of an air supply fan.

7.1.1.1.29 Chiller Eau 1Dment Room Ventilation System This system remains operable to service the chiller equipment room located on the 63' elevation of the control building.

7.1'.1.1.30 -Diesel Generator Room Emeroency Ventilation Systems This system is needed to support the operation of the emergency diesel. generator during movement of fuel in the secondary containment.

7.1.1.1.31 Relav Room. Emeroency Switchaear Rooms. And Computer '

Room Air Conditionina Systgm The relay room, emergency switchgear rooms, and computer room air conditioning system instrumentation and controls for one of the two redundant subsystems are maintained functional to 7-4 Rev. 4 July 1992 I

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7s SHOREHAM DSAR t

automatically control the ventilation system to maintain these rooms at their design temperature and system performance.

7.1.1.1.32 Batterv Room Ventilation System The battery room ventilation system instrumentation and controls automatically control and monitor the ventilation system to maintain the battery room at its design temperature and monitor system performance. Each of the three battery rooms has its own ventilation system which will remove any generated hydrogen.

7.1.1.1.33 containment Sorav and Sunpression Pool Coolina This system is not needed to support the storage of the fuel in the fuel pool, therefore it is not included in the DSAR.

7.1.1.1.34 Rod Secuence Control System This system is not needed to support the storage of the fuel in the fuel pool, therefore it is not included in the DSAR.

7.1.1.1.3b Motor Control Center Room Ventilation System

("T The motor control center (MCC) room ventilation system

(_) instrumentation.and controls are maintained functional to provide automatic' control of the ventilation system to maintain the room at design temperature for habitability. Each of the two MCC rooms in the reactor building has its own ventilation system.

7.1.1.1.36 Motor Generator Room Ventilation Syst2E The motor generator (MG) room ventilation system instrumentation-and controls-remain functional to-maintain the room at design temperatures for habitability. Each of the four MG rooms in the reactor building has its own ventilation system.

7.1.1.1.37 Compressed Air System (SRV Accumulators)

This system is not needed to support the storage of the fuel in the fuel pool, therefore it is not included in the DSAR.

7.1.1.1.38 Main Steam Isolation Valve Leakace Control System This system is not needed to support the storage of the fuel in

- the fuel-pool, therefore it is not included-in the DSAR.

7.1.1.2: Classificati2D c -

Section 3'2 provides a reclassification of systems based on their importance'to safety.

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SHOREHAM DSAR CHAPTER 8 ELECTRIC POWER

8.1 INTRODUCTION

This chapter describes the details of the plant auxiliary power distribution system which is designed to provide adequate electrical power to all plant equipment. The defueled condition of the plant does not require the operation of any Class 1E power system. However, as stated in Section 8.3.1 item 2, a diesel generator and associated equipment shall remain operable while fuel handling is taking place.

8.1.1 Utility Grid The description contained under this heading in the latest revision of the Shoreham USAR remains unchanged in the defueled condition. For further information on this subject refer to the USAR.

8.1.2 Interconnection To Other Grids

/~N The description contained under this heading in the latest

(_) revision of the Shoreham USAR remains unchanged in the defueled condition. For further information on this subject refer to the USAR.

8.1 3 Offsite Power System While in the defueled condition the offsite power system provides

~

. power to all operating plant equipment. Power to the Shoreham Nuclear Power Station is provided from the LILCO system through 138KV or 69KV. circuits. The 138KV switchyard is arranged in a two bus configuration with circuit breakers and switches arranged to perm! isolation and/or repair of either bus section. Four 138KV circuits enter into the switchyard (two per bus) each containing a circuit breaker at the connection to its respective i bus. Two. separate rights-of-way are provided, each containing two of the 138KV~ circuits. The 69KV circuit from the Wildwood substation enters: the site sharing one of the aforementioned rights-of-way for a distance of one mile. This circuit, however, is mounted on separate towers and is separated from the 138KV circuits. The detailed description of the remaining offsite system remains as described in the USAR except as follows:

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SHOREHAM DSAR

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Three Brookhaven 80MW (each) Combustion Turbine units are located on LILCO SNPS property approximately 3600 feet fr... the 138KV switchyard. These units are connected into one of the 138KV Holbrook transmission lines and are available to provide an additional source of onsite power to the SNPS. (see figure 8.2.1-2)

The spare Reserve Station Service and Normal Station Service transformers will no longer be required.

8.1.4 Qaz_ Site AC Power System The station electrical power system includes electrical equipment j and connections required to provide power to and control the  !

operation of electrically driven station equipment in the defueled condition. A non-safety emergency diesel generator will provide backup AC power during fuel handling in the secondary '

containment.

8.1. 5; On Site-DC Power Systqm- ,

During the defueled condition, the 125V DC distribution systems do not have.a safety function.- However, a DC distribution system will be maintained operable during fuel handling. operations. It I O will remain functional at other times.

The 24V DC power source will no longer be required. This system '.

provides power to the Nuclear Source and Intermediate Rango

-Instrumentation which is no longer in service in the defueled condition.

8.1.6 Identification of Safety Related Systems The description contained under this heading in the latest revision of the Shoreham USAR will not be applicable in the defueled state.

Table 8.1.6-1 Identification of Safety Loads The basis for these tabulations,no longer exists. The electrical distribution system will remain in service to maintain power to cplant equipment on the site-in the defueled condition.

8 .1. 7 Identification of Safety Criteria The description contained under-this heading in the latest

. revision of the Shoreham USAR is not applicable in the defueled state..

-O- Rev. 4 July 1992 8-2

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SHOREHAM DSAR O

Table 8.1.7-1 Feculatorv Desian Criteria For Electric Power The basis for these tar'alations no longer exists. The electrical distribution system will remain in service to maintain power to plant equipment on the site in the defueled condition.

8.2 OFFSITE POWER SYSTEM e

8.2.1 Description The description contained under this heading in the latest revision of the Shoreham USAR remain unchanged except as follows:

Service buses 101, 102 and 103 are nct required to be maintained as safety related while in the defueled condition. They are reclassified as Category II.

8.2.1.1 One Line Diacrams and Physical Drawinas The information contained under this heading in the latest revision of the Shoreham USAR remains unchanged in the defueled condition.

8.2.1.2 Transmission Line The description contained under this heading in the latest revision of the USAR remains unchanged in the defueled condition except that.the safety related function of the busses (1R22-SWG-101, 102, and 103) no longer exists. They are reclassified as Q.A. Category II systems.

8.2.1.3 Station Switchyard The description contained under this heading in the latest revision of the USAR remains unchanged in.the defueled condition.

'For further information on this subject refer to the USAR.

' 8 . 2 .1. 4 - Transmission _Line Exits The description contained under'this heading in the latest revision of the USAR remains unchanged in the defueled condition except for the following:

The new Brookhaven Combustion Turbines are added to the existing transmission line configuration. (Figure 8.2.1-2) 8.2.2 Analysis The basis of the analysis no longer exists. The analysis as fC s_) described in the USAR is not required in the defueled condition.

8-3 Rev, 4 July 1992

SHOREHAM DSAR

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8.3 Onsite Power Systems ,

The plant power system is designed to provide an adequate source ,

of electrical power to all systems required to be operational in the defueled condition.

C.3.1 An Power Systems The general description of the plant electrical power (AC) systems is as providad in this section of the USAR. However, the safety related design criteria are no longer applicable. The following does apply:

1- Equipment, switchgear,-or buses built and designed to safety standards are not maintained as safety related but will be inspected in the defueled condition since they are required for the diesel to be classified as operable.

J 2- one diesel generator set shall remain operable during fuel handling in the secondary containment.

3- Required surveillances and tests will be performed in accordance with the Technical Specifications.

) 4- Adequate equipment protection and emergency measures are dvailable for the required plant electrical systems in the defueled condition.

Thc' equipment, switchgear, and buses have been reclassified to Q.A. Category II. Therefore, safety functions such as auto-stari , redundancy, etc. , are no longer required.

8.3.2 CC Power Systems 8.3.2.1 Descr30 tion The description contained under this heading in the latest ',

revision of the Shoreham USAR. remains unchanged in the defueled condition except as follows:

1- The 24V DC. system, providing power to source and intermediate range-nuclear instrumentation, is no longer used.

2- 'All class 1E/ safety related functions of the DC system are no longer classified as such.. i The batterirt are being maintained for those sfstems remaining ye - functional ca operable in the-defueled condition. Q.A. Category

! I equipment is now Q.A. Category II.

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8-4 Rev. 4 July 1992

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SHOREHAM DSAR 9.1.2.5 Radiolonical Considerationq The description contained under this heading in the latest revision of Shoreham USAR remains unchanged. Refer to USAR for information on this subject.

9.1.3 Fuel Pool Coolina and Cleanun Systen All of the equipment in this system will be retained for operation, but in a modified manner. Since the fuel pool cooling subsystem is designed to remove the decay heat produced by spent fuel assemblies, as described in the USAR, and only a negligible amount of heat is expected to be generated from the slightly irradiated spent fuel bundles stored there, the cooling mode is not required. Thus reactor building closed loop cooling water is not required.

Appendix 9B provides an evaluation of spent fuel pool makeup requirements.

However, the spent fuel pool cooling subsystem will be used in the makeup mode in order to provide normal makeup water to the fuel pool from the condensate storage tank using the condensate y transfer and storage system. Alternate makeup sources for the spent fuel pool are Domineralized and Makeup Water System, and Fire Protection Water System. The makeup p-de is described at the end.of USAR paragraph 9.1.3.2.1. ,

The fuel pool cleanup subsystem will be used as designed.

The fuel pool cannot be inadvertently drained because the pump suctions for the fuel pool' cooling and cleanup system are taken above elevation 168,--or about 7 feet below the normal water level. If'a break occurred in these lines, about 18 feet of

-water would remain above the fuel in'the pool. This is more than enough to provide-adequate shielding. Pump' returns to the pool are equipped with siphon breakers to prevent inadvertant pool drainage.

9.1.4 Egel-Handlina System e 9.1.4.1 Desian Basis See USAR. This section is identical to the USAR.

9.1.4.2 Eauioment Descriotion

Sco USAR. This-section is identical to the USAR.

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Sil0REllAM DSAR b

9.1.4.3 Description of Fuel Transfer  :

The fuel handling system provides a safe and effective means for transporting and handling fuel from the time it reaches the plant until '.t leaves the plant after post-irradiation cooling. The preceding subsection describes the equipment and methods used in fuel handling. The following paragraphs describe the integrated fuel transfer system, which enrures that the design bases of the fuel handling system and the requirements of Regulatory uuide 1.13 are satisfied.

9.1.4.3.1 Arrival of Fuel On Site No now fuel is expected to arrive on site.

Therefore this section of the USAR is not required.

9.1.4.3.2 Refuelina Procedure No refueling is planned. Therefore this section of the USAR is not required.

9.1.4.3.3 DeDarture of-Fuel from Site

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This section applies as written in the USAR.

In addition:

1. The' spent fuel will be removed from the site in certified fuel. shipping casks.

2. The casks will be leak tested prior to shipment.

The remainder of USAR Section 9.1.4 is applicable.

9.2 WATER-SYSTEMS 9.2.1 -Service Water System The Service Water (SW) System is as described in USAR Sections 9.2.1.1-thru 9.2.1.5 with the following changes because of the reduced heat removal requirements with the plant in the de-fueled state.

a) The RBSW system is considered non-safety related because it does'not provide-cooling water to any plant equipment-required.to perform a safety' function.

b) One RBSW pump will supply cooling water to one RBSVS/CRAC g-<- chiller. condenser, an emergency diesel generator, and to all Turbine. Building Service Water (TBSW) cooling loads. (See

(' j'em e below.) No service water is regaired for RHR, 9-3 Rev. 4 July 1992

SHOREHAM DSAR e

() RBCLCW, drywell cooling, and u.akeup water to the reactor vessel ultimate cooling connection (UCC). The testable check valve in the UCC will not require testing to verify forward flow. Emergency service water to the spent fuel pool is not required (per DSAR Chapter 15) because of the >

very low heat generation by the fuel.

c) Automatic start / initiation due to accident signals are not required.

d) The double isolation valves which split the RBSW from the TBSW subsystems will be opened to intertie the subsystems.

e) Normal operation will now consist of'only c1e RBSW pump in .

use because of the minimal heat load imposed by the TBCLCW system to support the station air compressors. It will supply. cooling water to one TBCLCW heat exchanger, and the circulating water pump bearing. Cooling water for the vacuum priming pump seal cooler is not required. The second RBSW pump will remain in standby.

f) The TBSW pumps are out of service since they are no longer required.

g) Table 9.2.1-1 has been revised.

3.2.1.5 Instrumentation Anolication This section remains unchanged except that only the instrumentation needed-for the Service Water System as described in 9.2.1 a) through g) is required.

9.2.2 Egact'or Buildina closed Loon Coolina Water (RBCLCW) System This system is not needed to support the storage of fuel in the spent fuel pool.

9.2.3 Makeun Water-Demineralizer System The description contained under this heading in the latest revision of the Shoreham USAR r,emains unchanged except as-follows:

1. SBLC, RBCLCW, seal water injection, and vacuum priming are no longer users of demineralized water in the defueled conditions L 2. .The'HPCI suction.line from the condensate storage tank is l not required to be maintained as safety related in the L defueled condition.

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9-4 Rev. 4 July 1992

SilOREHAM DSAR

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9.2.9 Reactor Buildina Standbv Ventilation And control Room Air Conditionino Chilled Water System Redundancy in this system is not needed since neither the RBSVS nor CRAC systems are safety related in the defueled condition.

The heat loada generated by the electrical equipment in the control room, relay room and the emergency switchgear room are greatly reduced, such that only one chiller is required to maintain the control room, relay room and switchgear room at design conditions. The operating chiller and associated pumps will be manually controlled from the control room. This system has been reclassified QA Category II. Aside from the above, the system design remains unchanged and further information can be found under the above heading in the Shoreham USAR.

9.3 PROCESS AUXILIARIES 9.3.1 Compressed Air Systems The description contained under this heading in the latest revision of the USAR remains unchanged in the defueled condition except for the following:

1. piping that has been installed as ASME III code class 2 is

,Q no longer considered safety related and is reclassified QA Ds,/ . Category II.

2.- Nitrogen will no longer be used for inerting the primary containment or for equipment within the primary containment.

3. Safety'related functions of the compressed air system no longer exist. No pneumatically operated valves are required for safe shutdown.

For further information on the compressed air system, refer to 2 the USAR.

9.3.2 Process-Samnlina System F/ ifocesc aampling System provides monitoring of certain ,

s\;4h9 oparations while fuel is in the spent fuel pool for l short or long term storage. The process monitoring is

' t; us< plished as necessary by means of measuring, analyzing and/or.

)

resseding for conductivity, pH, and silica concentration, as shown on DSAR Table 9.3.2-1.

9-6 Rev. 4 July 1992 i

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9.3.3 Equinment and Floor Drainace System With the Reactor defueled and the fuel assemblies stored in the Fuel Pool, large portions of the Equipment and Floor Drainage i system are not required.

Eystem Descrinti 2D This system is described in the USAR. Changes in status are addressed below.

Reactor Buildino The only source of radioactive waste to the Equipment and Floor Drainage System in the Reactor Building is the Fuel Pool and associated service equipment leakage. Sources in the USAR that are no longer applicable are the Drywell Equipment Drain System and the Reactor Recirculation Pumps Drainage System. The Drywell Equipment Drain Tank is no longer required. One or more floor drain sumps are no longer required, as applicable.

Turbine Buildina

() The Turbine Building Floor Drain and Equipment Drain Systems are no longer required, as applicable, except for the Decontamination Sump drains'and associated equipment. There is no steam and the turbine is no longer required, so that the only source of radioactive waste is the Chemical Laboratory.

Radwaste Buildina The Radwaste Building Equipment and Floor Drainage System is maintained operational. The Dirty Waste Sump and Pumps (IN52-TK 114 and 1N52-P-187A/B) and Regenerant Recovery Sump and Pumps (IN52-TK-115 and 1N52-P-181A/B) are no longer required.

9.3.4 Chemical Volume Control, and Liauld Poison Systgma The Standby Liquid Control Systpm is no longer required in the defueled condition. The RWCU System is also no longer required unless the Reactor is layed up wet.

9.3.5 Failed Fuel Detection System

.With the fuel in-the pool, the descriptio- in the USAR Section is no longer applicable.

O 9-7 Rev. 4 July 1992

i SHOREHAM DSAR l i

. k-(~)b  !

In the event of gross fuel rod failure in the fuel pool (see .

" Worst Case Fuel Damage Accident" in DSAR Chapter 15), the  ;

refueling floor process radiation monitors will detect this  !

. radioactivity if it becomes airborne.

9.3.6 Sucoression Pool Pumoback System

-This system not required to support storage of fuel in the fuel  :

pool.

9.4 AIR CONDITIONING, HEATING, COOLING, AND VENTILATION SYSTEMS 1 9.4.1 Control Room Air'Conditionina System The control Room AC system remains unchanged in design _and operating functions.- However, the system is reclassified to QA l

Category II, the filter portion of the system will no longer be required:and one of each of the redundant fans and ACUS will no longer be-required. The AC system will only function to-provide ,

an OSHA environment for.the operators during the fuel storage period. This requires the operation of only one RBSVS/CRAC chiller. Automatic initiation systems and interlocks for the habitebility portion of-the system will be non-operable and the

O AC system will be manually controlled from the control room. For further discussion on this system refer to the Shoreham USAR.

9.4.2- Reactor Buildina Normal-Ventilation System -

9.4.2.1 Desian Basis The RBNVS remains unchanged.in design and operating function ,

except.that the: system willionly:

1. Provide; ventilation by introducing _ filtered outside air into the reactor building at a rate of approximately 2.7 air changes per hour L2. Remove heat' generated by solar and external heat' transmission.~1ighting-an4 the fuel pool.-

'3.- induce slight negative pressure in the' reactor building to prevent.potentially contaminated air from escaping from the building without being monitored.

The-RBNVS mayibe operated-in a' recirculation mode in order to -

control Reactor Building, humidity. . This helps to protect equipment from: damage due to-corrosion. While operating in the recirculation mode, the operating functions, as discussed above, u are maintained except~that-supply air is limited to infiltration n- ' caused by the negative pressure inside the Reactor Building.

9-8 Rev. 4 July 1992

p _

_ , - . _ . .--.m .,r., . , , - ---,a- , _m m-,v_. .. .-.- .- _ . J -. ._

. - -. -=- _- . . -_. _ . . .. . _ - . . . . . . . - - , -

SHOREHAM DEAR Os For further discussion on this system-refer to the USAR.

9.4.3 Radwaste Buildina Ventilation The description contained under this heading in the latest Shoreham USAR remains unchanged, except that the charcoal exhaust filtration system is no longer required and one of the two redundant supply and exhaust fans, mechanical refrigeration units and circulating pumps are also no longer required. Refer to the USAR for information on this subject.

9.4.4- Turbine Buildino Ventilation System And Station Exhaust System A) Turbine Buildina Ventilation System This system is not required to support the storage of fuel in the spent fuel pool.

B) Statio" Exhaust System This system will expel the exhaust air from the radwaste building and the reactor building. However, only two fans will be

/~ required for this purpose, one fan operating and one fan on

_\ stendby. This will ensure that the Isokinetic nozzles located in i the. upper level'of the exhaust duct will see a sufficiently high velocity to be operational. For-further discussion regarding l- this-system refer to the Shoreham USAR.

9.4.5 Batterv Room Heatina And Ventilation The description contained under this heading in the latest revision of Shoreham USAR remains unchanged. Refer to USAR for information on this subject. This system is reclassified to Q.A.

Category II.

9. 4. 6' Drvwell Air Coolina System

.This system is not needed while-the fuel is stored in_the spent fuel pool.

• 9.4.7 .Screenwell Pump House Heatina And Ventilation The description contained under this heading in the latest revision of Shoreham_USAR remains unchanged. Refer to USAR for information on this subject. This sytem is reclassified to Q.A. ,

Category II.

9-9 Rev. 4 July 1992

SHOREHAM DSAR b

(_/ 9.4.8 Plant Heatina

The description contained under this heading in the latest revision of Shoreham USAR remains unchanged. Refer to USAR for information on this subject.

9.4.9 Primary Containment Purae System This system is not needed while the fuel is stored in the spent fuel pool.

9.4.10 Diesel _Generatt Room Ventilation The description containst under this heading in the latest revision of the Shoreha* JSAR is revised. This system is reclassified to Q.A. Cate;ory II and nonsoismic. The system is no longer safety related and the design bases for tornado missile protection and room temperature control are no longer applicable.

9.4.11 Relav Room. Emeroency Switchaear Room And Computer Room Air conditionina System The description. contained under this heading in the latest revision of Shoreham USAR remains unchanged with the exception that only one train of equipment will remain-functional. Refer

(-)

(_j to USAR for information on this-subject. This system is reclassified to QA Category II.

9.5 OTHER AUXILIARY SYSTEMS 9.5.1 Fire Protection System 9.5.1.1 Desian Basis The design basis section applies with the following addition:

The basic premise of the fire protection discussions in the USAR and FHAR is protection from fire for safety related areas including areas containing equipment or circuits that are (1) required for safe shutdown,=or (2) required to prevent or mitigate radiological releases pomparable to 10CFR 100 limits.

Since safe shutdown-is assured by non-operation of the plant, and all-of the nuclear fuel is in the fuel storage pool, the only remaining safety related area is the Reactor Building.

Structures, systems components and administrative controls in place to protect areas,-equipment or circuits previously identified as_ safety related will be maintained as required for

= property loss prevention purposes and_should be considered the same as those fire protection features described in_the USAR for protectir. of-non-safety related areas, p 9-10 Rev. 4 July 1992

=m. . __. . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _

SHOREHAM DSAR r

V} Three documents which were used in the design of the plant's fire protection features and continue to be part of the fire protection pr(Jram are:

1. Evaluation of the SNPS Fire Protection Program as compared to 10CFR50 Appendix R criteria submitted via SNRC 572 dated May 21, 1981.

2. Fire Hazards Analysis Report,

3. Cable separation Analysis Report:

SNRC 532 dated February 10, 1981 SNRC 811 dated April 13, 1983 1 However, the-term " safety related", as used in those documents and in USAR section 9.5.1, applies only to the Reactor Building.

Section 6 of the Fire Hazards Analysis Report (FHAR) contains technical requirements that formerly were fire protection technical specifications.

FHAR Chapter 6 reflects reductions in the technical requirements that are consistent with the text of this DSAR Section ).5.1.

Types of Fires The " types of fires" section applies with no changes.

Desian criteria The " design criteria" section applies with the following addition:

As discussed above,_this design will be maintained for property loss prevention purposes. However the " safety related" application of the listed documents, particularly NRC's Branch Technical Position APCSB 9.5-1 and Appendix A thereto, is limited '

to the Reactor Building.

Locations of Fires

'The " locations of fires" section applies with the following changes:

The rooms listed parenthetically as examples of safety related areas having a concentration of cables are reclassified to Q.A.

Category II. The rooms listed as examples of where oil fires could occur near safety related. equipment no longer fit that description because these areas are reclassified to QA Category f- ' II. Furthermore, the fire hazard associated with this equipment is significantly reduced while the equipment is not being used 9-11 Rev. 4 July 1992

, 1

t. i f.

SHOREHAM DSAR O because the ignition-sources associated with the operating i equipment have been eliminated.

Intensity of fires j This section applies without change, i

-Fire Characteristics This'section applies without change. ,

Buildina Arranaement and Structural Features The " building arrangement and structural features" section applies with the following changes: ,

In the response to NRC question 3, as shown in FHAR revision 3,

-SNPS has stated our intention to replace existing motorized fire  :

_ dampers with newly designed fire dampers. All of-+9e areas where these new dampers were,to be installed are-in the control *

. Building-and-are reclassified to Q.A. Category II. Therofore,-

this. proposed modification will not be implemented. The-CO2 ,

systems for those rooms lare in. electric lockout. When a fire is detected, the CO system controls would cause the dampers to O

2 close on an. electrical signal. -As a backup, the fusible link of each--of. the existing fire- dampers- is suf ficient to cause closure of.a. damper--in.the event of a fire,-thus assuring integrity of ithe fire barriers.

In contrast with this USAR section, an unprotected HVAC opening exists in the east wall of=each of the three-diesel generator rooms within 50 feet of an oil-filled (Reserve. Station Service) transformer. fhis deviation was reported to the NRC on Licensee-Event' Report 87-021. The preposed corrective action was to l- 1

' install a deluge _. water curtain-system below-the existing missile L shield wall.between the. transformer and the wall openings. Since the diesel generator rooms are reclassified to QA Category II,

-this modification will.not be implemented. The partial protection provided by the missile barrier is considered

-sufficient for non-safety relatpd areas.

F1 Seismic Desian This:section applies without change.

~

, Water Reauirementsl The' " water' requirementst'~ section applies with the following additional statement- l 9-12 Rev. 4 July.1992 ,

U- - , . _. ,, , _ 2_ , __ ___._...._____.._._._w._.z.._.____

SHOREHAM DSAR (D

L.) ~

Although some areas previously identified as safety related are reclassified to QA Categcry II, the water supply is not being

-reduced.

Codes and Standarda This section applies without changes. SNPS will continue to meet the requirements of the applicable NFPA codes for fire protection systems that remain functional.

9.5.1.2 System Description The " System Description" section applies with the following changes:

As discussed earlier, all fire protection features remain in place. Several rooms / areas listed in this section as safety related are reclassified to Q.A. Category II. Essential circuitry installed for safe shutdown of the plant is no longer needed for that purpose. No removal of such cable or change in its physical separation is contemplated. Similarly, the service water line inside the Reactor Building, where a spare connection exists for manual hookup to the fire protection water system, is gg reclassified to Q.A. Category II. Modifications that would

( ,) degrade its seismic design are not contemplated at this time.

9.5.1.3 Safety Evaluatiqn Electrical Insulation Fires This section applies without change.

Charegal Fires This section spp;4 1 without change.

Oil Fires The "cil fire- sec. ion of the safety evaluation applies with the following change: ,

As discussed earlier, th- fire hazards associated with non-operating equipment are significantly reduced because the primary ignition sources - electrical energy and hot surfaces -

are eliminated.

M erity. Intensity and Duration of Fires This section applies without changes.

9-33 Rev. 4 July 1992

SHOREHAM DSAR i .

U Time Estimates This section applies without changes.

Failure _Mgde and~ Effects Analysis This-section applies without changes.

-Accidental Initiation of Fire Protection System The " accidental initiation ci fire protection system" section applies with the following change:

Areas protected by CO2 systems are among those that are no longer considared safety 71ated.

Sinale Failure in Fire Protection Systems This section applies without change.

Pine Breaks'in Fire Protection Systems This section applies without changes.

n '

4, ) ~'t' lure of Fire Protection System Affectina Safety Related M yt.oment T::i.; sect on applies with the following change: {

't ' the areas listed, only the Feactor Building is still  !

considered safety related.

t 9.5.1.4 Tests and Inspections i

This section applies without changes.

9.5.1.5 Personnel Oualification and Trainina i

-This section applies without changes.

9.5.2 _ Communications. System j 9.5.2.1 p_eslan e Bases  :

This section of the USAR remains unchanged.

f3 N.)

! 9-14 Rev. 4 July 1992 l

I'

p- m SHOREHAM DSARL System-Description

~

-9.5.2.2

This section-of.the-USAR remains unchanged _except for the Lfo11owing

-

l '. - For7thelvery low frequency (VLF) portable radio systems, one

. low-powered VLF radio base station will be useo in conjunction with two mobile car units to provide offsite radio communications (instead of two VLF base stations and four mobile car units).

2.: iThe' Emergency-Operations _ Facility (EOF) is not required,

-since no emergency: requiring EOF activation can occur with '

the fuel in the Spent Fuel Pool.

9.5.2.3 Tests and Ins o_ections

~ This1section of the'USAR remains-unchanged.

9~.5.3 Ljahtina Systems iWhilein;thetdefueled condition this system will provide all the necessary.requiredLlighting to the plant and the. site. The

.' description of this system-inLthe USAR remains unchanged except

).

for;the-following:

l'. .f Section'9.5.3.2,.-item #2_- the standby AC lighting system-Lwill'receiveupower from plant service buses:which are powered

=from offsito.

-2.. LSame:section,.;iteu #5 - the fifth lighting subsystem will Ereceive; power.-from'DC1 battery-sources while_the-plant remains in theidefueled: condition.

3.: (The last paragraph of1the.same section, the independent-power;= sources for lighting, remains __ unchanged but the sourceLof?pcwer,will be from plant service' buses _and DC

battery sources if.needed.

19.5.4 Diesel Generator-Fuel Oil Storace and Transfer System An; emergency diesel generatorcis requ' ired to be operational'when

.  : irradiated fuel?: is1being handled-in the1 reactor building.

F  : Sections 925.4-9.5.7 Din-the USAR remain _ descriptive of the EDG auxiliaryfsystems:except that these systems and their components H  : Tare?classifiedLas QA1 Category II. Also, the. requirements of redundancy-to prevent malfunction or-failure of these systems and

~

L p Ltheir components, i.e.,: fuel _ storage-thnks,' fuel pumps, air start b .

tanks, etc.-and-7-Day: operability Post-LOCA are no longer

-applicab).e.

9-15 Rev. 4 July 1992

?

l SHOREHAM DSAR Oi

/ Statements in Sections 9.5.4-9.5.7 indicating that portions of these systems are designed to ASME Boiler Pressure Vessel Code, Section 1II,-Code Class 3, that they meet Seismic Category I requirements,.and that the concrete block-house and the two-foot thick concrete slab above the fuel storage tanks are seismic Category I and provide missile protection are also no longer applicable.- The USAR description of equipment design with respect to applicable codes is representative of the original design of these systems but these designs which were applicable to safety related equipment in an operating nuclear power plant will no longer be maintained as safety related equipment, based oon the DSAR Chapter 15 safety analysis.

9.5.5 Diesel Generator Coolina Water System 9.5.6 Diesel Generator Startina System 9.5.7 Djesel Generator Lubrication System 9.5.8 Primary Containment Leakace Monitorina System With the fuel in the Spent Fuel Pool, the Primary Containment Leakage Manitoring System is not required.

p 9.5.9 -Etorace of Gases Under Pressure d' The quantities'and type of gases stored in pressurized containers in'the defueled condition is reduced from that previously on hand. The design bases remain unchanged. Storage facilities are provided for.the following gases as thown in Table 9.5.9-1:

1. Carbon Dioxide-for fire protection.

2. H a l o n :1 3 0 1 f o r f i r e p r o t e c t i o n .

3. Air for instrument, control, breathing and service.

4. Nitregen for glycol and HW heating.

5. Propane for auxiliary boiler ignition.

The following gasr;s are no longer used or required to be stored in the defueled condition:

1. Hydrogen for main generator.

2. Hydragen and oxygen for gas analyzers.

3. Nitrogen for containment inerting.

4. Nitrogen:for druwell floor seals.

5. Nitrogen for sloctrohydraulic control.

6. Air for MSI\ accumulators (inboard and outboard).

7. Air f or long term accumulators.

t )

%.J .

9-16 Rev. 4 July 1992

SHOREHAM DSAR The statement in the USAR relative to maintenance and laboratory' gases remain unchanged.- The safety evaluation discussed in section 9.5.9.3 of the USAR is only. applicable for air for instruments, service breathing, and control and for carbon dioxide and-halon. Statements relative to the pressure relief valves and gas release hazards remain as discussed in the USAR.

Gas use for safe shutdown is no longer necessary in the defueled condition.

i O

9-17 Rev. 4 July 1992 J

- - ~ . -

Nr l ,

7 L

'SHOREHAM DSAR

% Appendix-

- 9A;. FUEL CRITICALITY ANALYSIS-The Shoreham-Spent Fuel _ Rack ~(SFR) is of a stainless steel and

-waterfneut'ron-flux trap _ design which uses no additional poison.

A; description of the storage racks is-provided11n 9.1.2. The

< criticality analysis of this rack design is described in detail 11n= Appendix 9A of the Shoreham USAR. .The reactivity results

- which are summarized in-USAR Table 9A-4 remain-_ valid for the conditions existingfat Shoreham after defueling. Furthermore, due"to the differences in U-235 enrichment between the SFR LdesignedLand the current'Shoreham fuel, a large' negative reactivity credit should be taken into account. This is Lexplained as follows:

The Snorehamf SFR design -is based on a. maximum U-235 enrichment of 3.1 wt. %. The resulting basic cell k is calculated to be~0.9129'without uncertainty and model

. adjustments (Table 9A-4, Appendix 9A, Shoreham USAR)._ The z

-Shoreham1 Cycle.1--fuel-loading uses three (3)- enrichments.

of the:560: fuel assemblies-in the core,E340. bundles-have the.

highestibundleJaverage.U-235 enrichment of 2.19 wt. %, 144' bundles of 1.76 wt._% and 76 remaining bundles uses natural uranium; If-the/six inch natural uranium segments at'the top and

• bottom of.the: fuel are excluded, the average' enrichment of a 2.19 wt'.- % bundle becomes 2.33 wt.-%. :Using this enrichment

-andslinearly. extrapolating the reactivity vs_._U-235

enrichment results given in: Figure-9A-5~of-Appendix 9A, Shoreham'USAR,--.theLreactivity difference between the SFR *

'designfenrichment:of 3.1-wt. % and the' current maxiumum loading-enrichmentiof 2.33 wt.-%-is'found to be about -6.0%

in 7 s ka '( ,k, -0. 060) . . This brings the basic cell kmunder

nominalistorage' conditions-for the< current-fuel down to LO.85, which11s well below theLregulatory acceptance-

. criterion.of;km 0. 9 5.- All"the corrective 1and uncertainty-

-adjustments listedLin Table-9A-4 of the Shoreham USAR remain applicable.-

Duringethe. period.from-July, 1985 to June, 1987, Shoreham went through three separate. stages of low power testing

~

l(less than- 5% lof- rated power) =,. which resulted- in a total -

~

core 1 exposure of.approximatelyL48 mwd /MT as determined;by a

series.of core-follow analyses. The net-effect of the core-exposure is aislight decrease in

reactivity"( -0.002 fin-

-ks)Lmainlyidue.to the offsetting contributions from the

~

formation of Sm-149 and the slight depletion of'the-burnable Gd poison'in the fuel bundles.. In light of-the large reactivityimargin described previously (km 0.85), no

~

additional credit will-be-claimed here.

9-18 Rev. 4 July 1992 4 _ __

SHOREHAM DSAR

/"M

(_,/ '9B EVALUATION OF SPENT' FUEL POOL MAKEUP REOUIREMENTS An analysis was performed to determine the rate of water loss from'the spent fuel pool.through evaporation under the worst case scenario described below. The following conservative assumptions are used in the analysis to maximize the calculated pool evaporation rate:

1) The spent fuel pool temperature is 110'F.

2) The ambient temperature above the spent fuel pool is conservatively assumed to have zero relative humidity.

3) The reactor building air flow exists due to normal ventilation system operation to maximize evaporation.

The result of the calculation shows that the maximum evaporation rate from the pool is approximately 0.6 gpm which translates to a pool level depletion rate of one foot per eleven. days. Based on.this very low maximum depletion rate, external cooling of the spent fuel pool is not required. Technical Specifications require that the water-level above the spent fuel be a least twenty-one feet. In addition, it should be noted that

-s pool water level is alarmed in the control room and

('- ) alarm response _ procedures exist to provide appropriate operator action.

I) 9-19 Rev. 4 July 1992

SHOREHAM DSAR

/~~l LJ This section is no. longer applicable since most of the waste streams would no longer exist.

11.2.2.2- Low Conductivity Waste Subsystem Waste Collector Subsystem This syster will receive all the influents as stated in the USAR except that no inputs will be received from the Condensate Demineralizer' System, Drywell-Equipment Drain System and the Phase Separator Tanks.

11.2.2.3 Hiah Conductivity Waste Subsystem Floor Drain Subsystem This system will not receive any influents from the Drywell Floor Drain System, the Turbine Building Floor Drain Sumps and the condensate Demineralizer System. The Waste Evaporator will not be utilized to process this waste. Floor drain influents will be processed through the Floor Drain Filters.

11.2.2.4 Recenerant Chemical Subsystem 5- -

In-this system the only_ equipment still required are the Chemical Waste. Sump, the Regenerant Liquid Evaporator Feed Tanks and their associated pumps. The regenerant evaporator is not required.

11.2.2.5 System Operational Analysis The analysis described under this heading in the latest version of the USAR is not applicable in-the defueled plant cor.dition.

11.2.3 System Desian 11.2.3.1 Eauipment Description

-This~Section remains _as presented in the USAR.

11.2.3.2 Applicable Codes and $tandards This Section remains as presented in the USAR.

'11.2.3.3 Radwaste Buildina This Section remains as presented in the USAR except that the Radwaste Building is now designated a Quality Assurance Category II, Non-Seismic Structure.

! ( Rev. 4 July 1992

! 11-3 l:

l

-~

SHOREHAM~DSAR

,n

('~') 11.2.3.4 Liauid Radwaste Eauipment Ouality Groun Classification

-This Section remains as presented in the USAR.

11.2.3.4.1 Conditions and Assumptions This accident (raised in USAR Section 11.2.3.4)' postulates the simultaneous failure of the liquid radwaste system tanks in or associated with the radwaste building. These tanks hold the radioactivity and potentially radioactive liquid waste from the floor drains, equipment drains, nonradioactive chemical wastes, and processed liquid effluents. The tanks (and their capacities) that are assumed to fail are:

1. Waste collector tanks: Two at 25,000 gal each (Contents are insignificant 1y radioactive).

2. -Floor drain tanks:= Two at 25,000 gal each (Contento are insignificant 1y radioactive).

3. Regenerant liquid and evaporator feed tanks: Two at 25,000 gal each (contents are insignificantly radioactive).

4. Recovery sample tanks: Two at 25,000 gal each (located

(~) outside the radwaste building contents are insignificantly

(_/ . radioactive)

5. Discharge waste sample tanks: Two at 25,000 gal each (located outside the radwaste building)

6. Spent resin-tank: One at 4,700 gal (Section 11.5)

The source concentrations in the above are described in DSAR Table 11.1-1.

11.2.3.4.2 Acpident Description The accident description can be considered as described in Section 11.2.3.4.2 of the USAR except the structure is now classed QA Category II. ,

11.2.3.4.3 Accident Analysis This section' remains as presented in the USAR except that:

1. A conservative airborne partition factor of 1.0E-03 is assumed for all isotopic activities listed in DSAR Table

-11.1-1, with the exception of Tritium (H-3), for which it is assumed that all the activity evolves.

A L 11-4 Rev. 4 July 1992

SHOREHAM DSAR 7s l(f

2. Ground release atmospheric dispersion factors are assumed,

-as given in USAR Table 15.1-3,'for the EAB.

J. The breathing rate of persons offsite is assumed to be 3.47E-04 cubic meters per second, consistent with Regulatory Guides-l'.3 and 1.25. For other age groups the breathing rate was obtained from the ratio of the maximum age group ra er,given in Regulatory Guide 1.109 (Reference J).

11.2.3.4.4 Results and Consecuences The doses resulting from the enalysis described above are as follows:

Dose, millirem Whole body Beta Maximum Sourca Gamma

• Skin Oroan**

Spent Resin' 1.8E-05 2.7E-06 1.3E-03 Tank-r^s Radwaste Filters 1.2E-07 1.7E-08 8.3E-06 A

Discharge Sample 3.1E-08 1.4E-08 ' 7E-06 Tanks Totals 123E-05 2.8E-06 1.3E-03

• External & internal pathways;-child is the limiting age group

• • Teen is the limiting age group, and lung is the limiting organ The consequences of the above postulated accident are clearly very low. These projected doses are far below those which justify Quality Group D non-seismic qualification of radwaste equipment-(i.e., 500 mrem whole body, or its equivalent to parts of the body)', in Reg. Guide 1.26, Rev. 1, and Reg. Guide 1.29, Rev. 1.

11.2.3.5 Instrumentation & Control The-description contained under this heading in the latest d-revision of Shoreham USAR remains unchanged. Refer to the USAR

() for information on this subject.

11-5 Rev. 4 July 1992 i l

1 _

4 SHOREHAM DSAR

%AI "11'.2.3.6 Shieldina Field Routed PiDe-4The description contained'under this heading.in the latest trevision of-Shoreham USAR remains unchanged. Refer to the-USAR ffor-information on this subject.

11.2.4 Operatina Procedureg Operating procedures including administrative control of liquid radwaste releases 1are as described in the USAR except the Radwaste Building is now classified as QA Category II.

11.2.5 -Performance Tesig-Performance tests'of equipment-are-as described in the USAR, "except for activity reduction factors'(DF), which are no longer F applicable. Only equipment that remains-in operation will be >

~ periodically: tested.

l 11. 2 . 6. Estimated Releaffgg

! Liquid effluent releases are expected to be minimal with the fuel gin thelspent fuel pool. - This-is basedon the fact that.during D' - - - -'

the:periodDfrom June 19881through May~1989, only-one release had

( anLisotopicLeoncentrationl greater than LLD.

W .The' quantity of the-annualorelease of contaminated liquids is

. conservatively estimated _by notingithat the discharge volume from-SNPS is1approximately 5,000,000 gallons per. year. Assuming the affluent concentration is consirtently equal to that found in the one1 sample abovefLLD.(7.83E-08 uCi/cc of;Co-60, from DSAR Table-L11.1-1)i the, estimated release'is:

1.SE-03 Ci]yr-of.-Co-60

11'.2.7.1 Release Points

~

The:descriptionfcontained-under this heading in the latest revision-ofMShoreham USAR remains unchanged. Refer to USAR for information.on this subject'. .-

11.~2.8 Dilution Factors-Under the plant's present condition, service water or circulating Lwater will-be used,fif1necessary, for dilution-so that the discharged l effluent concentration in the Long Island Sound will

.not exceed;thatiprescribed11n 10CFR20, Appendix-B, Table'II, Column 2.

u 11-6 Rev. 4 July 1992 I

m a ,p- e j T

SHOREHAM DSAR LJ Treated radioactive effluents are collected in the discharge

. sample tanks. The filled tank is sampled, and then discharged at a maximum ~ rate of 150 gpm for a period of approximately 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. If necessary, the treated effluent is diluted with about 8000.gpm of service. water prior to discharge into the sound.

Thus, if necessary a dilution factor of approximately 50 may be obtained during actual discharge.

No credit is taken for the external dilution factor, i.e. the mixing ratio in the Sound, for service water.

11.2.9 Estimated Doses offsite dosen due to liquid releases are expected to be minimal, as discussed in DSAR Section 11.2.6. An estimate of the yearly dose is_ conservatively obtained by assuming each batch liquid release contains the maximum batch activity concentration of 7.83E-08 uCi/cc of Co-60, and the release volume is approximately 5,000,000 gallons per year. Assuming no dilution, the resulting doses are as follows:

Whole Body 0.166 mrem (adult)

GI-LLI 1.43 mrem (adult)

[v Liver 0.074 mrem (adult)

As noted in Section 11.2.8, service water dilution remains available as necessary.

11.3 GASEOUS WASTE SYSTEM 11.3.1 Desian Obiectives With the fuel in the Spent Fuel Pool, the radioactive gaseous waste system is no longer required to meet either 10CFR20 or 10CFR50 Appendix I limits. '

11.3.2 System Descriptions With the fuel in the Spent Fuel Pool, and negligible amounts of radioactive halogens;in the fuei, the radioactive waste sources described no-longer apply, and the systems necessary to process them are not required.

Normal ventilation will be maintained in the Radwaste and Reactor l

Buildings with discharge through'the station ventilation exhaust duct.

,O -

~J 11-7 Rev. 4 July 1992 l

l

V '

l 1

i SHOREHAM DSAR j

.11.3.3 System Desian The process.offgas system, which is the system described in USAR Sections 11.3.3, 11.3.4 and 11.3.5, is not required with the fuel

-in the Spent Fuel' Pool.

11.3.I Operatino Procedures 11.3.5 Performance Tests 11.3.6 Estimated Releases In the plant's present state, no releases of radioactive gaseous effluents-are anticipated. This is evidenced by the fact that since the plant. achieved initial criticality in 1985, there have been-no recorded releases documented in the Semi-Annual Radiological Effluents Reports.

11.3.7 ' Release Points The description contained under this heading in the latest revision of the Shoreham USAR remains unchanged. Refer to the USAR for information on this subject, p.

I 11.3.8 Dispersion Factors The description contained under this heading in the latest revision of the Shoreham USAR remains unchanged. Refer to ths USAR forLinformation on-this subject.

11.3.9 Estimated Doses There will be no expected offsite doses because no releases of radioactive gaseous effluents are anticipated under the plant's present defueled-state.

11.3.10 Unmonitored Release Points The unmonitored gaseous release, paths as described in the USAR twould be-expected to occur during normal plant operation. In the defueled condition some pathways do exist on loss of ventilation systems. Doses in such an event would be insignificant due to the. low'radionuclide. inventory in the plant. (Ref: NED Safety Analysis No. 4170024)

'A D 11-8 Rev. 4 July 1992

SHOREHAM DSAR

(%.J9 11.4- PROCESS AND EFFLUENT RADIATION MONITORING SYSTEM The description contained under this heading in USAR only apply to those monitoring systems described in DSAR Section 12.3.4.

Refer to the USAR for further information. The changes to the USAR relating to the Radiation Monitoring System for the defueled condition are described in DSAR Section 12.3.4.

Sampling for halogens is not needed in the defueled condition.

11.5 SOLID WASTE SYSTEM 11.5.1 Desion Obiectives The description contained under this heading in the latest revision of the USAR remains unchanged as it is used to develop the basic, design criteria of the plant.

However, in the present plant configuration this system is no longer required except for the retractable fill pipes and the transfer carts in the ;ubicles (since no solidification of waste, per se, is needed). High Integrity containers (HICs) will continue to be used since some wastes will continue to be generated, and must be shipped. Also Dry Active Waste (DAW) will continue to be generated, and must be shipped. The volume of both will be significantly less than that given in the USAR.

~

It should beLnoted that waste will be generated from the Spent Resin Tank, Radwaste Filter and Floor Drain Filter, as described in Section 11.2, to be transferred directly into HICs or to a mobile solidification or dewatering vendor. The HICs are then

-transported by-the transfer carts out of their cubicles to be

-handled by the-overhead crane.

Tables 11.5.1-1D and 11.5.1.-2 thru 5 of the USAR are superseded by DSAR Table-11.1-1.

11.5.2 System-Input: source Terms The actual radwaste source terms in the plant's defueled condition are as follows:

The combined activity concentration in the spent resin tank, radwaste filters, and the floor drain filter is assumed to equal-the maximum in the most recent solid waste shipments during the period-November-December 1988. DSAR Table 11.1-1 lists the activity concentrations of radionuclides.

Figure 11.5.2-1 no longer applies.

11.5.3 Eauioment Description Ns/ 11-9 Rev. 4 July 1992 i

SHOREHAM DSAR j'} .

' ~ '

11.5.3.1 -General The only equipment remaining in use in this system is as follows:

4,700-Gallon Soent Resin Tank (SRT)

For_the defueled condition, this tank receives backwashed resin and filter media from the Radwaste Demineralizer and the Fuel pool cleanup Demineralizer and Filters. (This tank is also discussed in Section 11.2. It is included here since it is a direct feed to-the Solidification system.)

.The spent resin pump transfers the spent resin to HICs which are set on the Radwaste floor or in the pits in the floor. The HICs are then dewatered by portable air-operated diaphragm pumps.which draw suction from specially designed piping internals in the

-HICs. 'When convenient, HICs may be dewatered while in the fill cubicles.

Daler This equipment is furnished to compress miscellaneous dry active waste (DAW) into 55 gallon drums.

O Transfer Carts and Fill Pipes V.

-These carts position the HICs at various stations within the fill '

cubicle'during filling and_ dewatering operations. These are filled from the Radwaste Filters and Floor Drain Filters through fill pipes.-

A connection is provided to allow for solidification dewatering of resins 1by a mobile vendor.

No-other equipment in this Section of the USAR is required.

11.5.3.2.: Wet Wastes The first paragraph of this Section of the USAR no longer applies. The second paragraph remains applicable.

11.5.3.3 Dry Wastes This'Section of the USAR is applicable,_as some DAW will_ continue to.be generated.

11.5.3.4 Irradiated Reactor Components This Section of the USAR still applies.

O-G 11-10 Rev. 4 July 1992

^

SHOREHAM DSAR 11.5.3.5 Operatina Procedures This section of the USAR no longer applies except that:

1. SRT waste can be transferred into a high integrity container (HIC) where it can be dewatered by the in-house dewatering system to Federal and burial site limits. Ultimately, this waste will be shipped to burial siter.

2. The shipping container is located under the retractable fill pipe by first placing the conta.tner on the waste container transfer vehicle within its locating guides and then running the transfer vehicle to a preset position directly beneath the fill pipe. The' fill pipe is lowered over the container and the fill pipe splatter shield entirely covers the container opening. The remotely operated fill pipe is powered in the vertical dirrction s by pneumatic cylinders.

11.5.3.6 InstrumentatioD All instrumentation in this section is no longer needed except

, for the radiation monitors.

11.5.4 ' E x p e c t e d ' V o l u m e.s_

This Section of the USAR is superseded by the following:

'A conservative expected estimated volume of waste in HICs and carbon steel liners is 1,000 cubic feet per year buried volume.

See DSAR Table 11.1-1 for activities.

This statement and Table together supersede Table 11.1-1A of tne USAR.

DAW volume is conservatively estimated to be 1,000 cubic feet per year, buried volume. The DAW activity is negligible.

11.5.5 Packaaina .

The description contained under this heading in the latest revision of the.USAR remains unchanged. Refer to the USAR for information.on'this subject.

I 11.5.6 Storace The description contained under this heading in the latest revision of the-USAR remains unchanged. Refer to the USAR for information on this subject.

{ 11-11 Rev. 4 July 1992

.- . . ~ . . < - - . ~ - - -- - - - - . . .

3:;

SHOREHAM DSAR d,4 x

W 11.5.7 Shinment' The description-contained underEthis heading-in the latest revision ef the USAR remains unchanged. Refer to the USAR for J

inforration on this subject.

- 11. 6. OFFSITE RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM pISCUSSION Therobjectives of.SNPS' Offsite Radiological Environmental

_ Monitoring Program (REMP) are to identify and measure plant

_ generated radioactivity:in the-environment and to calculate the potential dose to the surrounding population. SNPS' REMP is designed-toicomply withithe: Plant's Offsite Dose Calculation Manual:-(ODCM) and NRC Regulatory Guide-4.15. REMP data is acquired-by sampling various_ media in the~ environment and then analyzing these samples for radioisotopes; Tables 11.6.3-1 and

-11.6.3-2 detail-the REMP sampling / analyzing program. Since REMP results1 vary for each sample-and location, several sampling locations-were-selected for each medium _using available "

. meteorological, land,7and water use data. .The range.of analyses

.e f- .

.performeduon a: sample depend on the type of sample taken.

Sampling-locations __are designated as either indicator or control.

Indicator locations provide. representative.neasurements of radiationgand radioactive materials for those exposure pathways

!and radionuclides (from SNPS) that' lead to the highest potential 1radistioniexposures. control locations are placed sufficiently

.far from1SNPS_so that-they=will;be=beyond the measurable

.influenco:of7SNPS orfanylother nuclear-facility.- This monitoring

programElmplementscSection JV.B.2 of Appendix-I to.10CFR Part 50, iby--verifying that_ measured concentrations of radioactive

'materialsland direct radiation.are representative of the actual contaminationnlevels and-doses to the public.

SNPS' REMP has~been subdivided.over three distinct time intervals:L Preoperational-REMP (prior to SNPS initially achieving criticality), operational REMP'(from initial.

! criticality until removal of the fuel-from the core), and JPost-DefuelEREMP '(after the core was transferred to.the spent l fuel--pool)_.

Preoperational REMP'was performed to identify and determine ibackgroundflevels of environmental activity around SNPS.

n <Preoperational,REMP also served;to verify that indeed the media beingLsampled and analyzed is sensitive to radiological fluctuations in SNPS' environs (indicator locations)-and to Lprovide effective monitoring of potential critical pathways.

((

11-12 Rev. 4 July 1992

SHOREHAM DSAR

- Preoperational and Operational REMP samples within the aquatic environment included surface water, algae, fish, invertebrates (clams, lobsters, etc.) and sediment. The atmospheric environment was campled for airborne particulates, iodine, and noble gases. -Milk, potable water, precipitation, game and food products were obtained from the terrestrial environment. Direct radiation was' measured using thermoluminescent dosimeters (TLDs).

The-range of analyces for each~ sample could include: gamma spectrometry, Sr-89 and Sr-90; I-131; H-3, gross beta, direct rediation and noble gases. Under Post-Defuel REMP, several of the above sample types, sampling locations and/or analyses are discontinued. The current-Post-Defuel REMP program is outlined in Tabl,s 11.6.3-1 & 11.6.3-2.

Preoperational REMP began in February 1977 and continued through 1984, although the official Preoperational REMP period; i.e. the time frame against which the data base from Operational REMP was compared,-occurred during 1983 and 1984. The Operational REMP began on February 15, 1985 when initial criticality was achieved.

Except for reactor operator _ training programs which required the reactor to operate at 'O.0% powar' (during January 1989), SNPS has not generated radioisotopes since the last 5.0% power test, completed on June 6, 1987. Comparisons between the Labove two phases of REMP were documented in each annual REMP

- report.

N) -

t-As_of August 9,_ 1989, SNPS' core was transferred to the spent fuel pool.as part_of the agreement between LILCO, state and local governments not to operate Shoreham. This transfer prevents criticality from being-reestablished. In addition, since SNPS' last 5.0% powcr test was completed during June 1987, per Ref. 9,

with the. exception _of.I-129 and Kr-85, all iodines and gaseous effluents ~have decayed away. Consequently, the surveillance requirements for SNPS' Post-Defuel REMP were reduced to below the

-operational level, dystification for Reducino REMP to Post-Defuel Surveillance i Levels Pursuant to Reg Guide 4.1, once the initial core of the licensee has reached.the point (in time),of maximum burnup, and the licensee has demonstrated (using results from environmental media or calculations) that the doses and concentrations associated with a particular pathway are-sufficiently small (comparable to preoperational levels), then the number of media campled in the pathway and-the frequency of sampling may be reduced to operational Tech Spec requirements. Since (as of August 9, 1989) the core has been in the spent fuel pool, the initial core has

" exceeded" the point of maximum burnup.

11-13 Rev. 4 July 1992

. - . . ~_ . . - . - - . -.

1

.:/

SHOREHAM DSAR O}

t 1It-should be.noted that the concept of " normal" Tech Spec requirements as referred to in Reg. Guide 4.1, refers to a fully operationalJstation with--normal surveillance-requirements. Reg.

Guide 4.1:does.not. account-for the. unique condition at SNPS. >

__ Consequently, the justification for reducing:REMP will be

-performed in'twi steps. Step one reduces Operational REMP to the level mandaten when-SNPS vas to become operational. Step two- '

reduces the_ surveillance program further, to_the revised '

requirements corresponding to the defueled condition.

Dose calculations to SNPS' environs (1983 - 1988) were performed b'y analyzing positive concentrations of radioactivity detected in collected samples._ Table 11.6.1-4 compares the radiological

' impact _from-each_ major pathway to the public-during SNPS' preoperational and operational REMPs. Specifically, the

" radiological impact during SNPS' 5.0% power testing program (1985

- 1987) c was . compared to preoperational lREMP.

In all cases,-the' calculated doses during both the operational and preoperational phases-of REMP were comparable. Therefore, no environmental; radioactivity was identified (during any of the

-5.0% power-tests)Las having-originated at SNPS. These results Lsatisfy:the criteria established in Reg. Guide-4.1 for reducing f '? post-defuel REMP to.the_ level originally-mandated by SNPS'

( license.. The sampling: points not required by the license are:

1) Game;. 4) Rain Water; and 2); -Aquatic. Plants;_ 5) Noble Gases.

, L3)  : Aquatic-Sediment;

? Justification for reducingL REMP to the- revised ' requirements

_(after;theicore was defueled);is given:,ased on the above i

Einformation;;i.e.,--the_ measured environmental 11mpact.due to 5.0%-

. power _ testing'wasicomparable to that of preoperational REMP, and

as'of1 August 19,E1989,1the core was removed from the reactor.

pressure: vessel. 'SNPS' -last.5.0% power-test was-completed-on June;6,-jl987,.and'per--Ref. 9, with the. exception of I-129 and Kr-85,-all-lodines.and gaseous effluents have since decayed away.

In addition,.radwaste system activities.are quite low (listed in

-DSAR_Sectionst11.li& 12.2). As.a result, the only remaining

=radioisotopesf(and their release pathways)-are:

Isotone (s) Source- Effluent Pathway

'1) =Kr Spent Fuel Gaseous

2) Solubles and Radwaste- Gaseous and Liquid

-Particulates y) _-- . _ .

11-14 Rev. 4 July 1992 l.

I L

e . .

SHOREHAM DSAR p

U SNPS' Post-Defuel REMP Surveillance Procram Outlitig (Steos 1 & 21

1) DIRECT RADIATION: Reduce from 36 to 18 locations Quarterly Surveillance Frequency

2) AQUATIC

a. Aquatic Plants and - Delete, not required Beach Sediments

b. Fish, Surface Water - Retain, may be impacted and Invertebrates from liquid release path to L.I. Sound Perform Semiannual surveillances as available

3) AIRBORNE

a. Iodine - Delete, insigo3ficant quantity

b. Particulates and - Retain, particulates and Gross Beta and solubles still exist.

c._ Noble Gas - Delete Noble Gas, not required.

(V")

Quarterly Surveillance Frequency ,

4) TERRESTRIAL

a. Precipitation, Soil, - Delete, not Tech Spec and Game required

b. Potable Wate'c - Delete, well water not impacted by discharges to L.I. Sound c._ Milk, Food products - Retain, long lived particulates Quarterly Surveillance for Milk, Annually for' Food

SUMMARY

/ CONCLUSION

1) Examination of the radiological impact to REMP locations which are to be' eliminated -- From 1983 (preoperational REMP).through 1988_(which encompasses SNPS' S.0% power

testing program)_---indicates no measured increase in environmental contamination; refer to Table 11.6.1-4.

2) As of August 9, 1989, the SNPS core was transferred to the spent fuel pool; thus, the initial core has reached maximum g

t w) burnup.

11-15 Rev. 4 July 1992

l e.

SHOREHAM DSAR

/^T

( 'l' Per Regulatory Guide 4.1, if the above two conditions are 3) met, then the operational phase of REMP may be reduced to

.the requirements that were written when SNPS was to be operated as designed.

4)1 .The post-operational REMP surveillance program may be reduced from Step 1 to the requirements as delineated in

.DSAR Table 11.6.3-1 (Step 2), developed after the.SNPS core was transferred to-the-spent fuel pool, because:

a) Criticality will not ce reestablished at SNPS. As of August 9, 1989, no additional fission / activation products will be generated; b) SNPS'.last 5.0% power test was completed on June 6, 1987, which means that with the exception of I-129 and Kr-85, all remaining gaseous effluents have decayed away; and c) the only possible release paths for the remainir.g soluble or particulate effluents is through eithcr the spent fuel pool cleanup or makeup water systems (independent systems with no direct release path to the general public), or the radwaste treatment systems

. ("N (liquid and gaseous pathways) through which effluents

\_ / are being or-could be processed.

11.6.1 Obiectives of REME 11.6.1.1 Preonerational REMP The objectives of the Preoperational REMP were:

1. To identify and determine baseline radiological characteristics in the environment around SNPS (these background levels were then compared with data collected during. actual plant operation);

2. To assure that the media being sampled and analyzed are

,- sensitive to fluctuations in the radiological l charreteristics of the environs at SNPS, and to assure that REMP will be responsive to' radioactive discharges from SNPS (i.e.,-to identify indicator locations and critical pathways);

3. To provide effective monitoring of critical pathways of radiological effluents to unrestricted areas; and I

4. To train personnel and. evaluate procedures, equipment and techniques which are utilized in the operational and Post-Defuel phase of REMP, including emergency response

~}

y j- capabilities.

11-16 Rev. 4 July 1992 f'.

h:

u SHOREHAM DSAR x :: ,

The years-1983_and-1984 served as tho. official preoperational period,Eas stipulated =in: Reference 8.- All data collected during othis period-were used'in developing a baseline for ultimate comparison with: operational 1 data. From the' levels and

' fluctuations of radioactivity analyzed in environmental samples

~it-was: concluded-that-sensitive indicators of radioactivity for

-the: environment around SNPS'had been selected. Sensitive indicators revealed minute quantities of radioactive fallout from

the October 1980 atmospheric nuclear weapons test by the People's  ;

Republic of China during 1980 and 1981, in addition to radioactivity remaining from two decades of' atmospheric testing.

Airborne particulate samples registered an increase in gross beta levels, along with identifying the gamma emitting isotopes Zr-95, Nb-95, Ru-103 and Co-141. Also in 1983-and 1984, REMP sampling identified low levels of iodine-131 in Port Jefferson Harbor area aquatic samples'.- This was attributed to local hospitals treating patients;for thyroid carcinoma.

Along.with these anomalies-in the environment, expected normal background radioactivity was' measured in REMP samples. Aquatic samples consisting-of surface water, fish, invertebrates, aquatic plants.andisediment.were chosen and reflected the normal r 1 background radiation found in this environment. The atmospheric L Lenvironment was sampled-for airborne particulate matter, iodine,

-and1 noble gases.1 All airborne'radioiodine analyses were below detectable: levels. In addition,. milk, potable water,-game, food products, beach sediments and rain water were sampled. The esults'obtained fr'om thefanalyses of these samples were typical of the: radioactivity _ values usually associated with-samples of Lthese types.: -All radiciodine-analyses of_ milk were below etectable levels. Direct radiation. levels were relatively low, and approximately'the same at all locations. No unusual radiological; characteristics were observed innthe environs of s SNPSLduring?1983~and 1984. A summary of the annual program results1for 1983 and 1984 is given in USAR Tables 11.6.1-1 and 2.

11.6.1.2' Operationil'REMP 1Theinbjectives_of Operational.REMP were:

1); ^ Identify-and-measure plant-related radioactivity in the

, environment for the calculation of potential doses to the public.

.2)_

Identify-excessive radionuclide concentrations of limited

{ duration, so_that appropriate action may be taken.

3). Determine-the long-term variation in radionuclide conc _entration, or

~

4) determine the; effects of plant-effluents on the environment.

11-17 Rev. 4 July 1992

SHOREHAM DSAR T'T s 5)_ Comply with regulatory requirements and-provide records to document _ compliance.

6) ' Comply with the REMP requirements as outlined previously.

Operational REMP used the Preoperational data base to identify plant-contributed radiation,-and to evaluate the possible effects of radioactive effluents on the environment. The Preoperational and Operational phases of REMP were designed to comply with Regulatory Guide 4.15 (5) and the associated Branch Technical Position (4).

Analyses of the environmental samples show results (8) consistent with those found during the preoperational years (1983 -

1984).

Sensitive indicators revealed minute quantities of radioacti._

fallout remaining from the October, _1980 atmospheric nuclear weapons test by the Peoples Republic of China. Radioactivity traces from'the previous two decades of international above ground atomic bomb _ testing were also recorded. Radioactivity increases from the accident at the Soviet Union's Chernobyl Nuclear Power Plant (during April, 1986) were also measured.

Along with these environmental anomalies, expected normal background radioactivity was measured in REMP samples between 1985 and 1988. USAR Table 11.6.1-3 summarizes results from REMP (g during 1985, and DSAR Table 11.6.1-4 presents a comparison of

. (_) preoperatonal and operational REMP data from 1983 through 1988.

11.6.1.3 Post-Operational REMP The objectives of Post-Defuel and Operational REMP are identical.

Differences in the execution of Post-Defuel REMP account for both the permanent defueling of SNPS, and experience gained during the preoperational and operational REMP phases.

11.6.2 Potential Pathways 11.6.2.1 Liauid Effluent Pathways The exposure pathways for liquid effluents are:

1. External exposure from radionuclides in water; and

2. Ingestion of fish and shellfish containing radionuclides.

The concentrations of radionuclides expected to be released to the service Water are listed in Section 11.2. Dilution of these concentrations in'Long Island Sound is discussed in Section 11'.2.8.

UtSAR Section 11.6.2.1 contains detailed discussions about the projected doses from various liquid pathways. With the updated i

} source terms as described in the DSAR (Sections 11.1 and 12.2),

\!

11-18 Rev. 4 July 1992

SHOREHAM DSAR js_) -future-doses 1from-liquid pathways are expected to be a small fraction of the doses presented in the USAR. See DSAR Section 11'.2.9ffor dose calculations.

11.6.2.2 Gaseous Effluent Pathways The exposure pathways for gaseous _ effluents are:

1) Submersion in a cloud of noble gas;

2) . Drinking milk from a milning animal pastured in an areas of long-lived particulates;

3) Eating leafy' vegetables on which particulates have deposited.

-The calculated air dose-(using REMP when SNPS was to operate as designed)'at'the north-northeast site boundary is 1.1 mrad /yr from gamma radiation and 1.2 mrad /yr from beta radiation. Doses from gaseous effluent pathways are summarized in USAR Table 11.6.2-3. Computational methods are discussed in Section 11.6.2.3.

A_ dairy survey is performed annually to determine the location of any milking-animal within-a 5-mile radius of SNPS. When a milking. cow or goat is found, annual doses are calculated using either current meteorological or activity release data, in accordance with-the methods specified in the Shoreham Offsite y~]s

[

% Dose, Calculation Manual.

11.6.2.3 Dose Computational Methods 11.6.2.3.1 Liauld Effluent Pathways

The discussion-contained in the latest vrrsion of the Shoreham USAR-(Section_ 11.6.2.3.1) continues-to apply.

11.6.2.3.2 Caseous Effluent Pathways The: discussion contained in the-latest version of-the Shoreham USAR (Section 11.6.2.3.2)- continues to apply.

~11.6.3 Samolina Media, Locations, and Frecuency Tyoical Post-operational REMP sampling locations and frequency are.given in Table 11.6.3-1. These locations are described in Table 11.6'.3-2 and are shown in Figures 11.6.3-1 and -2. By virtue of the liquid and gaseous effluents from the plant, REMP

-is divided up into four distinct categories: atmospheric, terrestrial, aquatic and direct radiation. Sampling media, locations, and frequencies-are discussed in the following sections.

(~h.

%J 11-19 Rev. 4 July 1992

SHOREHAM DSAR V 11.6.3.1 Samolir.a Media 11.6.3.1.1 Acuatic Environment The aquatic environment is examined by analyzing samples of: 1)

' Surface water; 2) Fish; and Invertebrates. Surface water samples are taken:in May and October using a Niskin Bottle. The samples are placed in new polyethylene bottles following three rinses with the sample medium prior to collection. When available samples of_ Winter Flounder,_Pseudopleuronectes americanus. Windowpane, Econhthalmus acuosunt Sea Robin, Prionotus spp, Little Skate, Raia erinacea. Blackfish, Tautoa onitia and Summer Flounder, Paralichthys dentatus are taken by trawl, sealed in plastic bags, frozen, and shipped to the analytical laboratory for-analysis.

When-available, invertebrate samples of American Lobster,-Homarus americanus. Squid,.Lolico praleii and Channeled Whelk, Busvcon canaliculata are_ collected by trawl._ Channeled whelk are also collected using pots. These-invertebrate samples are then sealed in plastic bags, frozen and shipped to the laboratory for analysis. . Blue Mussels Mvtilus edulis are collected by hand along jetties and soft-shell clama, Mya arenaria from Wading River-are shelled and sealed in plastic bags, frozen and shipped to the analytical laboratory.

3 -

11.6.3.1.2 Atmosoheric Environment

-The atmospheric environmentcis examined by-analyzing airborne

_particulates collected on Gelman Type A/E filters using low

-volume air samplers .(approximately 1 cfm). The samplers used

.are equipped with. vacuum recorders for sample volume correction and to indicate sample. validity and-maintenance problems when they occur. Should the sampler lose vacuum due to a leak the vacuum level reading will drop to zero. Since this may occur without a' corresponding loss of electric supply the exact-time of

.the maintenance problem will-be evident on the recorder chart.

Sample _ volumes are measured using-dry gas meters and corrected for differences between the actual pressure that the volume meter sees and the average atmospherig pressure. Sample volumes are corrected to standard pressure using average-weekly barometric pressure (measured at Environmental Engineering Department, Melville) andJair sampler vacuum readings. Time totalizers indicate the duration of tir; the sample is taken.

11.6.3.1.3 Terrestrial Environment The terrestrial environment is examined by analyzing samples of milk and food products. When available, milk samples are collected quarterly, except during the pasture season (May 11-20 Rev. 4 July 1992

SHOREHAM DSAR

. f).

l-through October) when the sampling is increased to monthly. Milk samples are prepared for shipment in accordance with the instruction-of;the laboratory performing the analy,is. Food products consisting of vegetables and fruit are collected from '

area farm stands and shipped fre=h to the laboratory.

11.6.3.1.4 Direct Radiation Direct radiation levels in the environs are measured with energy compensated calcium sulfate (CaSO4:Dy) TLDs, each containing four separate readout. areas. The TLDs are annealed prior to placeLent l in the field. One TLD is placed at each of the 18 locations, and exchanged on a quarterly bases; these locations correspond t; the 16' meteorological sectors in the general areas of the site boundary, plus'two control locations (actual locations arc listed

'in Table 11.6.3-1). .The units are then packaged and shipped to

-the-laboratory-for analysis.

11.6.3.2 Samolina Locations and Frecuency Typical REMP sampling locations and frequency are given in Table 11'.6.3-1. These locations are described in Table 11.6.5-2 and shown in Figures 11.6.3-1 and 11.6.3-2.

' [j\

\,

11.6.4' NOT USED IN THE DSAR (Data Incorporated Into Section 11.6.1) 11.6.5 Data Analysis. Presentation and Interoretation The discussion contained in the latest version of the Shoreham USAR (Section 11.6.5, 11.6.5.1, and 11.6.5.2) continues to apply.

~

11.6.6 Erogram Statistical Sensitivity The discussion contained in the latest version of the Shoreham USAR'(Section 11.6.6)-continues to apply.

REFERENCES In Section'11.6

'1) Regulatory Guide 4.1 " Programs for Monitoring Radioactivity in the Environs of. Nuclear Power Plants" J

M Not Used

3) Not Used

4) Radiclogical Branch Technical Position, Rev. 1, Nov. 1979

5) Reg. Guide 4.15, Rev. 1, February-1979, " Quality Assurance

, /

~

'For Radiological Monitoring Program (Normal Operation)

Effluent Streams and the Environment" 11 ^1 Rev. 4 July 1992

I l

j l,D[ SHOREHAM DSAR V

'6)- SNPS'Offsite Dose Calculation Manual 3/4.12 Radiological Environmental Monitoring 3/4.12.1 Monitoring Program Table 3.12.1-1 "REMP"

7) -Not Used

8) _ SNPS'-Operational REMP Annual Reports: January 1, to December 31, 1983, 1984, 1985, 1986, 1987, & 1988 issued by Nuclear Engineering and Environmental Engineering Departments of LILCO.

9) C-RPD-476, Rev. O, 10/21/88, "SNPS Core Thermal Power After Shutdown" O

11-22 Rev. 4 July 1992 i

i

SHOREHAM DSAR TABLE 11.1-1 Radwaste Sources Greater than LLD Soont Resin Tank. Radwaste Filter, & Floor Drain Filter The activity concentration is assumed to equal the maximum in the most recent HIC shipment (Nov-Dec 1988) and is'(From Reference 2):

Activity Isotope Concentration. uCi/cc  % of Activity i

• Cr-51 9.84E-04 58.46%

Mn-54 2.17E-05 1.29%

• Fe-55 4.19E-04 24.88%

• Co-57 7.92E-07 0.05%

Co-58 6.43E-06 0.38%

Co-60 1.09E-04 6.51%

• Fe 4.57E-05 2.71%

• Ni-63 6.41E-06 0.38%

• Sb-124 3.2SE-06 0.19%

• Zn-65 1.89E-05 1.12%

H-3 6.21E-06 0.37%

• C-14 3.94E-07 0.02%

- g ,/ - 0.01%

• Sr-90 1.69E-07

• Zr 1.52E-05 0.91%

• Nb-95 2.55E-05 1.51%

• Tc-99 4.79E-09 0.00%

• I-129 7.32E-10 0.00%

• Cs-137 1.34E-06 0.08%

• Co-144 2.95E-06 0.18%

• Pu-241 1.59E-05 0.95%

Discharce Waste Sample Tanks The activity concentration in these tanks is assumed to equal the maximum concentration measured in the past 12 months preceding May 1989 (from Ref. 3):

Activity Isotope Concentration, uCi/cc  % of Activity Co-60 7.83E-08 100.0%

Note: The remaining radwaste tanks (floor drain collector tanks, waste collector tanks, and recovery sample

-tanks) were all determined in Reference 4 to have isotopic concentrations less than LLD.

• Calculated based on generic scaling factor.

Rev. 4 July 1992

^

V .I_ _.

/m f !q,/

V \d SHOREHAM DSAR TABLE 11.6.1.-4 Comparison Of Operatiena1 - Preom er*ional._? IMP Data

-- Operational REMP ) (- Preoperational PIMP -)

(

1986 1985 1984 1983

_ Unit / Isotope 1988 1987 ,

SAMPLE TYPE 150 - 290 120 - 540 70 - 220._

pCi/1 (H-31 240 - 410 140 - 450 130 - 420 Potable dater _ 34.0 _{21,2 pCi/KofCs-137) 76.7 - 9270 35.1 - 6490 54 - 3230 992 - 4330 641 - 5340 Gmae 2.7 - 6.9 2.3 - s.?

2.8 --6.9 1.9 - 5.7 3.0 - 6.2 Direct (gamma) mrem Monthly 2.3 - 5.2 2.9 - 4.9 2.8 - 5.5 3.1 - 6.2 2.8 - 3 4 Ouarteriv 2.7 - 4.8 2.9 - 5.0 Radiation 4. 2 - 1. 5 - 54 5.0 - 44.0 4.0 - 32.0 5.0 - 360 6 - 47 Air: Gross Beta ( x 1. OE-3,) 1.3 - 1.4 0.11 - 0.27 LT O.8 LT O.07 pC1/m3 x 1.E-3 LT O.8 LT O.8 Particulate Sr-90 35 - 1230** _LT 10.0 LT 10.0 LT 30.0 Iodine-131 pCi/m3 x 1.E-3 LT_10.0 LT 10.0 6.8 - 27.

• 33. LT 20.0 LT 1.0 LT 1.0 LT 1.0 Aquatic pCi/Kg (Sr-90) 36 - 55

• 85.5

• 47.9

• 45. 69.7 - 140.

Plants pCi/Ko (Cs-l!7J_ LT 6.0 i

0.86 - 4.60 0.69 - 5.3 0. 9 - 7. '

0.76 - 6.00 0.61 - 5.70 0.98 - 13.0 pCi/1 (Sr-90) 12.9 - 14.1 7.0 - 8.9 + 4.4 9.6 - 14 pCi/1 (Cs-137) 6.00 - 14.8 5.90 - 11.5 Milk LT O.20 6A pCi/1 (I-1311 LT O.20 LT O.20 2.1 - 4.8 LT O.20 _ e LT 4.0 LT 3.0 NA

.~ s.O LT 4.0 LT 4.0 Food pC1/Kg (I-131) LT 5.0

• 24.7 LT 5.0

• 12.2 LT 5.0 Products 8 wet) (Cs-1371 LT 5.0 identified isotope.

• Ranges are not given since only one data point contained an

• • Evidence of Chernobyl accident.

Rev. 4 July 199; I of 2

-,-j i ._Y

2-l i!

j SHOREHAM DSAR TABLE 11.6.1.-4 , (Cont'd) t Comparison Of Operational - Precrerational REMS Data 1 (

Operational REMP ) (- Preoperational REMP -) j I- SAMPLE TYPE Unit / Isotope 1988 1987 1986 1985 1984 1983 Aquatic pCi/Kg (Sr-90) LT 1.0 LT 1.0

• 5.6 LT 1.0 LT O.9

• 86 l

Invertebrate tweti (Cs-1371 LT'5.0 34.8 - 36.2 NA NA NA NA l

j Beach pCi/Kg (Sr-90) LT.1.0 LT 1.0 LT 1.0 LT 1.0

• 3.3 LT 2.0 Sediment fdrvi (Cs-1371 LT 8.0 LT 8.0 LT 8.0 LT 8.0 LT 9.0 NA l.

Aquatic pCi/Kg (Sr-90) LT 2.0 LT 2.0 LT 2.0 LT 2.0

• 1.7 LT 3.0 l.

Sediment idrvi (Cs-1371 LT 10.0

• 21.7 LT 10.0

• 30.4 44.2 - 49.4 NA l

Serface Water. oci/l (H-31 + 190 170 - 430 180 - 290 180 - 220 50 - 270 90 - 28J Fish pCi/Kg (Sr-90) LT O.5 LT O.5 LT O.5 LT O.5 LT O.6 LT O.7 i oCi/Ko fCs-1371 7.11 - 17.5 11.0 - 25.8 10.2 - 13.8 7.70 - 17.4 8.4 - 21.4 8.8 - 19.1 Rain Water pCi/l (H-3) 130 - 490 130 - 410 120 - 190 140 - 320 80 - 970 90 - 270 4 pCi/1 iCs-1371 NA NA 1.40 - 12.4 NA NA NA Noble Cases gi/m3 (Kr-85) 28 - 44. 24 - 45 21 - 48 24 - 40 30 18 - 49 pCi/m3 (Xe-1331 LT 11.0 LT 11.0 LT 11.0 LT 11.0 LT 34.0 LT 40.0 J-l i

• Ranges are not given since only one data point contained an identified isotope.

i I

4 i

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2 of 2 Rev. 4 July 1%

i i

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1 SHOREHAM DSAR

(

r EDOREHAM DSAR TABLE 11.6.3-1 Jecond Steo Post-Ocerational Radiolocical Environmental Monitorino Procram (REMP1 Media Samolina' Locations Samplino Frecuency Analysis l

Direct 151,2A2,3S3,451,$S2, Quarterly Gamma Exposure Radiation (1) 6S2,7A2,8A3,951,10A1, 11A1,12A1,13S3,1452, 15S1,16S2,*5E2,*6El

' ' Fish and 3C1, 14C1, *13G2 Semi-annually Gamma-isotopic Inve- nurates (2) or when in season '

Fruis 8B1, 6B21, *12H1 At time of Annual Gamma-isotopic and Vegetables-(3) Harvest Airborne 6S2,2A2,351,781,*1101 Quarterly Gross-Ceta Particulates (4) and Gamma-isotopic Milk ($) lab 1,*10F1, or '8G2 Mthly during Gamme-isotopic Grating Season, Ortly, at all other times.

Surface Water 3C1 or 14C1, and *13G2 Semiannual Gamma-isotopic Grab Sample H-3

(*) Designates Control Locations (1). Eighteen monitoring rtations, DR1 through DR18, (16 indicator and 2 control) are used. One. indicator location is positioned in each meteorological sector near the site l>'undary. . One dosimeter or continuously measuring dose rate instrument is pisced at each location.

(2) At each Indicator location, one sample of each commercially and recreationally important species. One sample of same species in control location.

(3). Sample three different kinds of broad leafy vegetables grown nearest to two indicator locations -- having highout predicted average ground level D/Q (when milk samples not available). Also take one sample of same leafy vegetation grown nearest to Control Location.

(4) Three samples.(nest SNPS), one from each of the three Meteorological sectors having the largest annually averaged ground-level D/Q, are taken, one sample (near a community) also having the highest calculated annually sveragod ground-level D/Q-is taken. Establish one Control Location.

(5) Indicator samples from milking animale having highest potential dose.

_ Sample within 5 km distance (preferably), within 5 to 8 km where doses are calculated to exceed 1 mrem /yr:(second choice) or from 8 to 17 km. Control location is 15 to 30 km from SNPS and'in the least prevalent wind direction. +

Rev. 4 July 1992

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r L-.

SHOREHAM DSAR SHOREHAM DSAR TABLE 21.6.3-2 PEMP SAMPLING LOCATIONS DESIGNATION LOCATION IS1 Beach east of intake, 0.3 mile [H]

2A2 West end of Creek Road, 0.2 mile [NNE) 3C1 Fish and Invertebrates, Outfall Area, 2.9 miles ;aE) 351' Site Boundary, 0.1 mile [NE) 4S1 Site Boundary, 0.1 mile [ENE)

• SE2 -Calverton, 4.5 miles [E)

SS2' Site Boundary, 0.1 mile [E) 6B21 Condezella's Farm Stand, 1.8 miles [ESE)

• 6El LILCO ROW, 4.8 miles (ESE) 6S2 Site Boundary, 0.1 mile (ESE) 1A2 North Country Road, 0.7 n.ile [SE) 7B1 overhill Road, 1.4 miles [SE) 8A3 North Country Road, 0.6 mile [SSE) 881 Local Farm, 1.2 miles [SSE)

• 8G2 Dairy (Cow), 10.8 miles [SSE)

O 951 10A1 Service Road SNPS, 0.2 mile [S) horth Country Road, 0.3 mile [SSW)

• 10F1 Goat Farm, 9.2 miles [SSW) 11A1 Site Boundary, 0.3 mile [SW)

• 11G1 MacArthur Substation, 16.6 miles [SW) 12A1 Heteorological Tower, 0.9 mile-[WSW)

• 12H1 Background Farm, 26 miles [WSW)

!~ 13B1 Goat Farm, 1.9 miles [W)

• 1302 Fish and Invertebrates, Background, 13.2 miles [W) 13S3 Site Boundary,-0.2 mile [W) 14C1' Fish'and Invertebrates, Outfall Area, 2.1 miles [WNW)

-1452 St. Joseph's Villa, 0.4 miles [WNW) 15S1' Beach west of intake, 0.3 mile [NW) 16S2 Site Boundary, 0.3 mile [NNW)

• Designates control Locations

h' J Rev. 4 July 1992

l l

SHOREHAM DSAR l l

q b Egnetor Buildina There is no significant source of airborno activity assumed to i exist in the reactor building in the plant's present lefueled condition.

Turbine Buildina There is no source of airborne activity assumed to exist in the turbine building.

Radwaste Buildina There is no significant source of airborne activity assumed to exist in the radwaste building. ,

Further discussion regarding airborne activity is provided in sections 11.1 and 12.4.

REFERENCES cenoral Updated Safety Analysis Report (USAR) Shoreham Nuclear Power Station Revision 1, December 1987.

1.. ORIGEN2, Isotope Generation and Depletion Code, ORNL CCC-371, 7/80.

2. LILCO calculation C-RPD-476, rev. 0, 10/21/88.

3. LILCO calculation C-RPD-530, rev. O, 05/19/89.

4. LILCO calculation ~C-RPD-529, rev. O, 06/07/89.

5. QADMOD-G, Point Kernel Shielding Code, ORNL CCC-396, 12/79.

12.3 RADIATION PROTECTION DESIGN FEATURES

-12.3.1 Facility Desian Featuigg The description contained under'this-heading in the latest revision of.the Shoreham USAR remains unchanged as it is used to develop the basic design criteria of the plant. Refer to the

-USAR for information on this subject. However, the defueled condition, with low activity levels,= some design features are not necessarily utilized as-described in.the USAR. For example, liquid filters in the radwaste system do not usually require portable shielding or remote backwashing. Also, the radiation zone designations shown on USAR Figures 12.3.1-1 through -35 are not applicable for the plant's present condition.

(V~}

12-4 Rev. 4 July 1992

SHOREHAM DSAR

]'

y 12.3.2 Shielding The description contained under this heading in the latest revision of the Shoreham USAR remains unchanged as it is used to develop the bas.c design criteria of the plant. Refer to the USAR for information on this subject.

12.3.3 Ventilation The description contained under this heading in the latest revision of the Shoreham USAR remains unchanged. Refer to the i USAR for information on this subject. I 12.3.4 Radiation Monitorina Instrumentation In order to support the.ates ,~ .y *be fuel in the fuel pool, SNPS will need process aid aft. <t : Mation monitoring instrumentation, and are, Ane)nr sgh Jadiation monitoring ,

instrumentation.

-Process and Effluent _. Radiation.Monitorina System

.The process and effluent radiation monitoring system is designed in accordance with General Design Criterion 64. All normal paths for release of radioactive materials are monitored to ensure

(

\

compliance with the requirements of 10CFR20, 10CFR50, and Regulatory Guide 1.21.

Table 12.3.4A lists the monitors in service, a7d Table 12.3.4B provides data for each monitor.

Normally, nonradioactive systems that may become significantly ,

contaminated by. leaks from radioactive systems are monitored continually to ensure that no condition hazardous to the operating personnel or to the general public develops. For effluent streams that. discharge to the environs, sample points are' located downstream of the last point of possible-radioactive fluid addition to the effluent being monitored.

All monitors'in.the process and effluent radiation monitoring system detect gross activity levels and readout and alarm in the main control room. Alarms in the main control room are by annunciators and cathode ray tube (CRT) display.

There are three normal eff3uont release-points from the station that require radiation monitors: the station ventilation exhaust, the liquid radwaste effluent, and the reactor building salt water drain tank.

_N u

12-5 Rev. 4 July 1992

SHOREHAM DSAR ,

) Area Radiation and Airborne Radioactivity Monitorina s_/ Instrumentation This section contains a description of the area and airborne radiation monitoring systems. All channels have local readout by means of a log-ratemeter and local audible and visual alarms.

Each channel has high radiation and fail alarms which are annunciated locally and in the main control room. The area monitors are provided with an audio and visual alert and high radiation alarms. Monitors are placed in areas where personnel normally have access and where there is a possibly that radiation levels could become significant.

All airborno monitors are offline monitors and are designed in accordance with ANSI N 1331-1969, " Guide to Sampling Airborne Radioactive Materials in Nuclear Facilities." Sample lines are kept as short as possible to minimize plate out while allowing the monitor to be located in an accessible area.

Airborne radiation monitoring is provided where potentially radioactive sources exists. Each of these monitors is provided with an isokinetic nozzle which is sized to obtain a representative air sample at the normal flow in the ventilation duct.from which the sample is taken.

Table 12.3.4B lists the airborne monitors, and Table 12.3.4C O lists the area monitors.

Radiation Monitorino System comouters The RMS is equipped with redundant computers powered from U.P.S.

1Z97-INV-005/TSC Black Battery /69 kV primary feed. These units provide continual surveillance for all airborne, area, process, and effluent radiation monitore. Communication with the computer is through keyboard equipped CRT displays in the_ main control room, the health physics office, the process computer room, and the technical support center.

' Inservice Insnection. Calibration, and Maintenance The operability.of each channel-of the area and airborne RMS is checked periodically from the main control room or manually at the monitor. Both systems are checked periodically or as specified by the plant technical specifications.

Calibration'of-all! monitors is normally conducted at an interval of.18 months unless mandated sooner by Technical-Specification.

This calibration will allow indication in a low, mid, and high response range of each monitor.

12-6 Rev, 4 July 1992

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SHOREHAM DSAR

)

) 12.4 DOSE ASSESSMENT 12.4.1 Desian Obiectives The design of the shielding was originally based on conservative estimates of the occupancy time required in each area of tha plant, under operating conditions. An effort has been made to keep the dose to plant personnel as low as is reasonably achievable (ALARA) under all conditions, including the defueled i condition. Table 12.4-1 lists the six zone designations that were originally established, along with the maximum allowable dose rates and estimated occupancy times for each area. With the plant in its present condition, with spent fuel stored underwater in the pool,-there are no occupiable areas which are Zone III or

' higher.

12.4.2 Airborne Activity An area within the-Shoreham facility is described as an " airborne radioactivity area" if the sum of the concentrations of all airborne radionuclides divided by their respective Maximum-Permissible Concentrations (MPCs) (from 10CFR20, Appendix B,

' Table 1, Column 1) exceeds 0.05. At Shoreham, there are no

" airborne radioactivity areas" in the defueled condition. With the fuel in the spent fuel pool, and insignificant quantities of

' g- radioactive material elsewhere (see Sections 11.1 and 12.2), it s is.not expected that airborne radioactivity areas will exist in the future,_unless systems which_are currently anticipated to remain closed are opened to the atmosphere. In this instance, the radiation work _ permit _ procedure (see Section 12.5) will be applied to assure there is no release of contamination into the air.

With exposures reasonably expected to be much less than 2 MPCa-hrs per day and/or 10 MPCa-hrs per week, paragraph 103(a)

(3) of 10CFR20 indicates that exposure,-and the resulting internal doses, need not be included in the dose assessment to

-individuals. With no " airborne radioactivity areas" postulated, doses are thus t %en to be essentially Zara for the defueled condition.

It should be noted that the above conclusion will be confirmed in

~

actual practice by the whole body counting program (see Section 12.5). _ Procedures are in place for taking appropriate actions, including investigation, when any positive whole body count occurs in' excess.of 1% of the maximum permissible organ burden (MPOB), or 1% of the maximum permissible body burden (MPBB) .

12-7 Rev. 4 July 1992

l SHOREHAM DSAR O 12.4.3 Occuoational Dose Assessngnt  !

l Occupational dose at Shoreham is expected to be essentially zero for the defueled condition. This conclusion has three bases:

i

1) At present, the dose rates in occupiable areas are virtually l all less than 0.5 mrem /hr, as described in Section 12.3.  ;

There are no sources of radiation present which would cause the present dose rates to increase to any significant extent. r

2) In the defueled condition, occupancy in measurable dose rate areas is expected to be less than or equal to that in the recent past at Shoreham. However, physical decontamination activities (beginning in 1991) will result in occupancy levels substantially exceeding the levels between 1988 and 1990.

3) The recent collective station dose history at Shoreham is as follows (TLD data collected in response to the requirements of 10CFR20.407):

Ilme Period Dose. man-rem 1/1/86 - 6/30/86 0.562 7/1/86 - 12/31/86 3.123 1/1/87 - 6/30/87 0.341 O 7/1/87 - 12/31/87 1/1/88 - 6/30/88 0.065 0.050 7/1/88 12/33/88 0.000 1/1/89 - 6/30/89 0.020 7/1/89 - 12/31/89 0.075 Since February of 1987, when a change was made from R. S.

Landauer to Panasonic TLDs, doses have been insignificant, and due almost entirely to sma)1 statistical fluctuations rather than actual doses. -

Based on the above statements, it is anticipated that occupationa3 dose at Shoreham will be essentially zero until physical decontamination activitics commence. As such, the 1991 corporate ALARA goal was established at 2.5 man-rem. Doses will be measured as indicated in the' Health Physics Program, Section 12 . 5. -

L 12.4.4- Offsite Dose Assessment There are no sources (eg, N-16) in the defueled condition which, under normal (non-accident) conditions, could lead to offsite direct' doses, either by direct radiation or "skyshine", based on tho' source terms presented in Sections 11.1 and 12.2. As such, offsite-doses to the population are projected to be zero in the defueled condition. 'This conclusion will bc confirmed by the

'O- REMP, as described in Section 11.6.

12-8 Rev. 4 July 1992

i SHOREHAM DSAR 12.5 HEALTH PHYSICS PROGRAM The Shoreham Health Physics Program, tho intent of which is to provide for the protection of all permanent and temporary personnel and all visitore from radiation and radioactive materials in a manner consistent with Federal and State regulations during a*_' phases of operation, is described in Section 12.5 of the USAR. The program is applicable in its entirety to the defueled condition at Shoreham, with the following exceptions:

A) Handling of new fuel is no longer applicable to Shoreham.

(Reference USAR Section 12.5.1.2, Egrsonnel Experience and Oualifications. The basis of this change is that with the Settlement Agreement with New York State, no new fuel will be brought onsite.)

B) The laundry facility does not contain an automated respirator washer, unloading table for same, or-a respirator dryer. Cleaning of respirators is done by hand methods when necessary. Respirator fitting may at some time in the future be moved from the Annex Duilding to another onsite location. Protective clothing is to be cleaned either onsite or offsite, as conditions warrant.

('

(_,r) (Reference USAR Section 12.5.2.1, Location of Eauinment, Instrumentation and Facilities. The basis of this change is

-the fact that with no airborne areas currently identified,

.and none expected in the defueled conditior, requirements for respirator use are infrequent. Also, the need to clean protective clothing is significantly reduced.)

C) Deleted D) The numbers of detectors and monitoring instruments will not necessarily be maintained as indicated in USAR Section 12.5.2.2. Rather, the number maintained will be as required by the defueled plant's activities and number of personnel.

(Reference USAR Section 12.5.2.2, Tvoes of Detectors and Monitorina Instruments. Justification of this change is due to the near total decay of radiciodines at the site.

E) Radiation Work Permits arn required for work under any of the 'following conditions:.

1. Work in a posted radiation area.

2. Entry into a posted high radiation area.

() 3.

4.

Work in a posted contaminated area (see Item F below).

Entry into airborne radioactivity areas.

, 12-9 Rev. 4 July 1992

SHOREHAM DSAR

5. Breach of a radioactively contaminated system boundary.

k (Reference USAR Section 12.5.3.2, _ Radiation Work Permita.

The basis of this change is a change to station procedures.

F) Under the discussion of access control, add the definition of a contaminated area:

Contaminated Area

-Any area having removable beta / gamma-emitting radioactive material in excess of 1000 dpm/100 cq cm, or alphu-emitting radioactive material in excess of 20 dpm/100 sq cm.

(Reference USAR Section 12.5.3.3.1, Access Control.

The basis for this change is a modification to the station health physics procedures, as recommended by the Institute of Nuclear Power Operations, in their document INFO 85-001, rev.1.)

G) Under the discussion of access control, the " secondary access facility" no longer exists.

(Reference USAR Section 12.5.3.3.1, Access Control. The basis for this change is that as of September 1, 1989, the rN secondary access facility was taken out of service.)

\J The ALARA Review Committee (ARC) now administratively H) reports to the Resident Manager.

(Reference USAR Section 12.5.3.3.4, Post-Onerations Review.

The basis for this change is an organizational change. See Chapter 13 of the DSAR for further details.)

I) 1As stated in~DSAR'Section 12.1D, there is no longer a need to. provide dosimetry to personnel entering the RCA, unless they are required by an RWP.

(Reference USAR Section 12.5.3.5, Health Physics Trainina Erocram.)

It should be noted that some of the procedural requirements or commitments indicated under the USAR Health Physics Program will not apply in the defueled condition. For example, no areas requiring reevaluation for extra shielding are anticipated, due to the. low current-source terms (Reference.USAR Section 12.5.3.3). However, potential sources of radioactivity (during physical decontamination activities) warrant that the procedures

.and commitments remain in place.

12-10 Rev. 4 July 1992 l

__ _ ~- _ _ _ _ _ _

SHOREHAK DSAR CHAPTER 13

)

CONDUCT OF OPERATIONS 13.1 ORGANIZATIONAL STRUCTURE AND RESPONSI_BILITIES The description contained under this heading in the latest revision of the Shoreham USAR is superseded in its entirety by the following.

The Long Island Power Authority (LIPA), a non profit public entity, as the sole owner of the Shoreham Nuclear Power Station (SNPS) has assumed full responsibility for its maintenance and decommissioning. LIPA has contracted the New York Power Authority (NYPA) to manage the day-to-day maintenance and decommissioning of the SNPS. NYPA, also a non-profit public entity, is the sole owner and operator of the Indian Point 3 (IP3) and James A. Fitzpatrick (JAF) Nuclear Power Plants. LIPA and NYPA have entered into an agreement whereby several key corporate and site upper management personnel are coemployees of both LIPA and NYPA. This arrangement allows LIPA to establish and maintain technical cognizance through onsite residency.

Coomployed personnel shall have .nanagement responsibility for the safe conduct of operations at the Shareham site as a whole, as

well as_ individual management responsibilities in the areas of operations and maintenance, decommissioning, radiological controls, quality assurance, and licensing / regulatory compliance.

This coemployment status shall be maintained in order to provide assurance of a' safe and efficient maintenance and decommissioning process in conformance with Nuclear Regulatory Commission (NRC) requirements and the facility licensing commitments. Additional NYPA resources are~available to LIPA to support SNPS maintenance and decommissioning on a non-coemployed basis.

LIPA has also entered into separate agreements with the Long Island Lighting Company (LILCO) to secure the support of selected incumbent LILCO technical, administrative and management staff personnel, as well as offsite support services in areas such as training, emergency preparedness, environmental engineering, and other areas. LILCO is an investor-owned public utility and was responsible = for the original design, _ constrt: tion and licensing of SNPS.-

Figure 13.1-1 depicts the corporate and' plant organization of LIPA for the maintenance and decommissioning of the SNPS.

13.1.1 Corocrate Org_anization >

Figure 13.1-1 depicts the LIPA Corporate Organization for the management of the SNPS.

m 13-1 Rev. 4 July 1992

SHOREHAM'DSAR

. () 13.1.1.1 Coroorate Oraanizational Arranaement The LIPA Chairman has overall responsibility for the

~ administration-of LIPA, including all financial, legal, and public relations aspects. The Chairman is appointed by the Governor of the State of New York.

In meeting and supporting these responsibilities, the LIPA Chairman has a President of Shoreham Project reporting directly to him on matters relating to operations, engineering, decommissioning, quality assurance, and security. The President of Shoreham Project shall have an Executive Vice President of Shoreham Project (EVPSP) reporting directly to him.

The EVPSP shall.be a corporate project intermediary addressing administrative, budgetary, engineering, quality assurance, security,1and decommissioning activities at the SNPS. The EVPSP shall provide overall guidance and direction to the Shoreham Decommissioning Project,'shall_be the corporate executive responsible for the overall nuclear safety of the plant and shall

-have the authority.to take_such measures as may be needed to ensure acceptable performance of the staff jn operating, maintaining, and providing technical support to the plant to ensure nuclear safety. These operations are discharged by the Shoreham Plant Resident Manager and the managers of the operations and Maintenance, Decommissioning, Nuclear Operations O Support, Nuclear Quality Assurance,_ Finance and Administration, and Licensing / Regulatory compliance departments. Supplementary technical supportnis provided to these organizations under the direction of LILCO executive management by various offsite LILCO departments and divisions through appropriately defined LIPA Nuclear operations Corporate Policies.

The EVPSP shall_be a-coemployee of both LIPA and NYPA. -As a minimum,:the.EVPSP shall haveEa' Bachelor's degree'in science or-an engineering field associated with power production.-The EVPSP shall also have 10 years of experience associated with plant design and operation, at least 5 years of which shall be nuclear power plant experience.

The. Manager of: the -NQA Department has overall responsibility for nuclear quality 1 assurance (QA) hetivities directing'the

= activities of the Quality _ Control (QC) Manager and Quality Systems-(QS)_ Manager. The Manager,_NQA reports to a e EVPSP for ,

policy 1 matters, and-to the Resident Manager for personnel administration, budgetary control, and functional _ day to day .

Lassignments._ The Manager, NQA has direct access to.the EVPSP and' to the President of Shoreham Project 1for~ nuclear safety matters, as he deems necessary.

!OL 13-2 Rev. 4 July 1992 s

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SilOREHAM DSAR l 13.1.1.2 lechnical suonort I Technical support for the activities associe.ted with the maintenance and decommissioning of Shoreham is provided b3 several organizations. Onsite technical support is provided by the Nuclear Engineering Division of the Operations and Maintenance Department, and by the Decommissioning Engineering Division of the Decommissioning Department. Refer to sections-13.1.2.1.1 and 13.1.2.1.3, respectively, for a description of the technical support functions of these organizations.

Additional offsite technical support is also available as needed through LILCO and NYPA corporate engineering organizations, as well as through qualified outside contractors. ,

13.1.2 Operatina Oraanizatio.D The SNPS organization chart is shown in Figure 13.1-1. This chart depicts the titles and line of authority of the plant personnel in charge of the various plant departments. The station organization shall include all the technically trained personnel necessary to support all aspects of the maintenance and decommissioning of the plant.

13.1.2.1 Station Ornanization O The Resident Manager, reporting to the EVPSP, has complete responsibility for the safe, efficient, and dependable

,. ' maintenance and dacommissioning of the plant. The Resident Manager administers an organization of LIPA management employees skilled in the various disciplines required for nuclear plant maintenance and decommissioning. Management' employees in turn direct the actions and supervise the performance of station personnel at the plant, which are a composite of NYPA, LILCO and contractor employees under LIPA's umbrella organization. All plant personnel shown in Figure 13.1-1 shall meet or exceed the minimum qualifications of ANSI N18.1 1971 for comparable positions, as appropriate.for the permanently defueled status of SNPS.

l l 13.1.2.1.1 Onerations & Maintenance Decartment i

The Operations & Maintenance Department (O&M) provide onsite technical and administrative support for operations, maintenance, radiological controls, instrumentation-and controls, and engineering services.

The O&M department'is organized along the following division lines:

O 13-3 Rev. 4 July 1992 1'

, ._. _ _ _ . , . . _ __ _ .,-_ ___., _ . _ _ _ . _ , _ , _ . _ _ ~ , . , - -,

SHOREHAM DSAR Operations Division  ;

The operations Division (OD) shall be responsible for the operation and monitoring of station systems and equipment required for daily operations and maintenance, and for complying with the station license and regulations of governing agencies. I The OD will assist the Decommissioning Department, as necessary, I and shall review decommissioning activities related to l operations. The OD shall provide station administrative support l l

and assurance that the station is in compliance with the requirements of the License.

The OD shall be responsible for field engineering and providing i the criteria for post-modification / return to service testing of l station modifications. The OD shall prepara and issue procedures l and coordinate the implementation, testing and startup activities associated with station modificationa. The OD shall assure that the plant and systems modifications are properly installed, tested and demonstrated functional.

The OD will implement their portion of the station surveillance programs.

The OD will interface with the Decommissioning Department and other support organizations, as appropriate. This includes planning and scheduling support associated with plant O&M activities.

O- During off-shifts and in the absence of the Resident Manager or his designee, the OMD in the person of the Watch Engineer shall be the person-in-charge of station activities.

The OD will direct the Maintenance Division in fuel handling operations.

Maintenance Division The Maintenance Division _(MD) shall be responsible for maintaining the station's systems and equipment (i.e. -

mechanical, electrical, instrumentation, computer, etc.) and for implementation of approved modifications. This responsibility includes administering the local area computer network and l providing general computer support. The MD shall have a staff experienced in mechanical and electrical maintenance of large steam-electric generating stations. As necessary, the division's staff may be supplemented with competent maintenance personnel from outside contractors.

The MD shall be responsible for ense-ing that plant systems, applicable to the maintenance of SNLJ are maintained in accordance with the station license requirements and provide the Decommissioning Department with maintenance support throughout I

/~) the planning and execution of SNPS decommissioning.

(_/

13-4 Rev. 4 July 1992

S110REllAM DSAR In addition, the MD shall be responsible for ensuring the and testing of all instruments and O calibration, maintenance, control systems during daily maintenance and decommissioni activitics except for those instruments which are calibrated and maintained by the Radiological Controls Division. The MD will repair, test, and maintain all hardware, software, and firmware associated with security, and radiological monitoring systems and selected portions of the process radiation monitoring system.

The HD shall provide refueling bridge operators to work under the '

direction of the operations Division during fuel handling '

operations.

Radiological Controls Division The Radiological Controls Division (RCD) shall be responsible for establishing programs and procedures for protecting the public, station personnel, and the environment from the effects of radiation associated with normal maintenance of the plant and associated decommissioning activities. It shall provide mechanisms for ensuring radiation doses of station personnel and the public are maintained as low as reasonably achievable (ALARA) and assure proper handling, processing, and disposal of radioactive materials in compliance with applicable regulatory requirements.

6 The RCD shall be responsible for activities such as analysis and monitoring of potentially radioactive effluents to the environment associated with maintenance and decommissioning of the plant. The RCD shall work with NED in assessing radiation doses to the public,_and shall be responsible for station chemical and radiochemical activites.

The RCD shall be responsible for assisting the Emergency Director in evaluating an emergency condition. Continuing assessment actions will be taken for the purposes of 3dentification and characterization of the incident, prediction of offsite doses, if any, resulting from the incident, notification to offsite authorities, determination of appropriate corrective measures, and determination of escalation, reduction, or termination of the emergency.

The RCD shall be responsible for maintaining an effective waste reduction program and assurg regulatory compliance, in handling, packaging, storir. , and shipping of all radioactive waste generated during daily maintenance and' decommissioning activities at SNPS. The RCD shall maintain an adequate inventory of protective clothing and contamination control equipment, to support daily maintenance and decommissioning activities; and shall. maintain the plant as radiologically clean as possible through implementation of non-specialized decontamination processes.

13-5 Rev. 4 July 1992

SHOREHAM DSAR i g The RCD shall interface with all departments and with offnite i qf agencies and parties regarding procedures, techniques, and resources necessary to adequately maintain preparedness for any radiological emergency condition which could arise with the plant '

in the defueled condition or as a result of decommissioning ,

activities.

The RCD will prepare Radiation Work Permits, perform radiological '

surveillances, maintain personnel exposure records, calibrate and maintain all fixed and portable radiation detection instrumentation, and dispose of radioactive material properly.

The RCD shall be responsible for providing environmental support to the SNPS organizations, as necessary.

The RCD shall be responsible for developing conducting and evaluating the final site radiation survey.

Nuclear Engineering Division The Nuclear Engineering Division (NED) shall be responsible for providing design and engineering expertisc in station systems, structures, and equipment. This includes identifying problems and recommending corrective action or design changes, development of plant improvements, monitoring of plant modification implementation and performance, and issuance of approved O- engineering procedures and specifications for use, as required by 1 plant procedures.

The NED shall provide the design for station modifications.

-The NED shall perform, manage, direct, and provide design

-verification for engineering, design and safety analyses and perform / review engineering and safety evaluations.

The NED shall perform project engineering'for engineering studies and plant modifications, and prepare and monitor engineer.ng schedules and cost estimates for all engineering work related to the' maintenance of SNPS and shall provide engineering and technical support to the Decommissioning Department. The NED shall provide administrative and technical direction to outside engineering consultants.and LILCO's Office of Engineering that are performing activities related to chartered responsibilities.

[

The'NED shall' provide engineering support'for technical review of j spare and replacement parts, procurement and dedication of commercial grade items to nuclear application as required.

The NED will prepare and review licensing correspondence and L

submittals with regard to impact on design and safety analyses and licensing documents. The NED shall review and evaluate the-O technical adequacy of design,_ licensing and operating aspects of new or proposed regulatory requirements and industry experience.

13-6 Rev. 4 July 1992

SHOREHAM DSAR The NED shall also prepare and assure accuracy of the content of

/~ plant technical basis, design control and licensing documents (i.e., drawings, specifications, calculations, procurements, technical manuals, equiment and other controlled lists, and program descriptions) as required.

The NED shall be responsible for technical interface with the Decommissioning Department engineering personnel to assure that ,

decommissioning engineering plans, activities and station '

modifications are compatible with the existing Shoreham plant  :

design.

)

The NED shall be responsible for developing and maintaining the Radiological Environmental Monitoring, Process Control, Offsite Dose Calculation Manual, LIPA/LILCO Corporate Fire Protection Program (Nuclear) and ALARA programs.

13.1.2.1.2 Nuclear Operations Support Denartment The Nuclear Operations Support De artment (NOSD) provides LIPA with security, fire protection anu safety, emergency preparedness

-coordination, plant administration, records management, document control, and training support for the planning and execution of SNPS decommissioning.

The NOSD is organized along the following division lines:

Nuclear Security & Training (NST) Divir ion NST shall consist of the Nuclear Security, Emergency Preparedness, and Nuclear Training Sections.

Nuclear Security Section The Nuclear' Security (NS) section shall be responsible for establishing and implementing security plans, procedures, contingency plans, guard training. and qualification plans and programs necessary to comply with the rules and regulations of the governing regulatory agencies. NS ensures thnt utrict security is established and maintained to keep the station l buildings, equipment, materials, and personnel safe from injury, unauthorized used, or destruction. NS will-review appropriate plant modifications and decommissioning activities to ensure compliance with current and projected security requirements and commitme,ts.

NS will ensure security intrusion detection and access control systeme meet the requirements of the security plan. These systems protect the facility against sabotage or attack and provide and enforce a system of prevention of theft or loss of LIPA property through a protection of assets program. In O

addition, the NS will maintain security records concerning systems, equipment, and personnel.

13-7 Rev. 4 July 1992 I

\= - - - . _ _ . , - __. _ - _ - - _ - - -

SHOREHAM DSAR The N3 will paintain liaison with local, state, and federal law enforcement cgencies to assure coordination and support of the

(-) security plan.

Nuclear Training Section Nuclear Training (NT) shall be responsible for implementing and coordinating training required to establish qualifications for personnel assigned at SNPS. Individual Departments will establish position descriptions and qualification requirements within the Training and Qualification (T&Q) Program. Training to meet these requirements will be conducted, coordinated, and recorded by the Nuclear Training Coordinator (NTC). ,

Training will be established through contracts, the LILCO Training Center, or directly conducted through the NTC. Review and approval of qualifications will be the responsibility of the appropriate department. T&Q records will be developed, tracked within the LILCO T&Q Program and Controlled as QA records.

The'NTC shall be responsible for the development of training policy and procedures; and for the development, implementation, and evaluation of training programs for permanent, temporary, and contractor personnel. The NTC will ensure that training programs meet regulatory compliance, radiation dose minimization, worker safety, cost effectiveness, and plant staff qualification 3 requirements.

(O The NTC shall coordinate personnel training and prepare appropriate training materials for the following subjects and for other specialized training, as necessary: General Employee Training; Fitness for Duty; ALARA; security; fire protection and safety; emergency preparedness; decommissioning; quality assurance indoctrination; etc.

Although the applicable training will be directed through the NTC, the individual SNPS departments shall have the responsibility for ensuring that personnel under their direction are qualified to assume the responsibilities of their positions.

The NTC shall-make recommendations to the Resider.t Manager on training policies and procedures which impact on other departments or on maintenance and decommissioning activities of the plant.

Emergency Preparedness Section ,

The Emergency Preparedness Coordinator (EPC) shall be responsible for reporting on the status of the development and implementation of the emergency plan to_ protect the public and station personnel from the effects of-radiation exposure in the event of a 0, . radiological emergency during daily maintenance activities at the plant and during the planning and execution of SNPS decommissioning.

13-8 Rev. 4 July 1992

-pe----- myere-w----- -t-ae-rw --1 _m- -'n- ev*

SHOREHAM DSAR The EPC is the primary interface with LILCo's office of Corporate Services for matters pertaining to the development, maintenance, and revision of LIPA's Emergency Plan.

Fire Safety & Administration (FSA) Division The FSA shall consist of the Fire Protection & Site Safety, Plant Administration, and Records Management and Document Control Sections.

Fire Protection & Site Safety Section

.The Fire Protection & Site Safety Section (FPSS) shall be responsible for implementing the fire protection program, including drills, surveillance activities, nud maintenance of i fire equipment; and for providing effective health and safety programs for the employees, contractors, and visitors to SNPS, in  !

compliance with local, state, and federal laws remarding health, industrial safety, and fire protection. The FPho shall review and audit the installation and maintenance of fire protection and prevention equipment throughout decommissioning activities and shall-maintain fire protection and safety-records and files.

The FPSS shall manage an on-site medical /first-aid facility to provide competent emergency care and first-aid to minimize  !

medical complications from injury / illness. The FPSS shall 0

, develop medical unit policies and procedures. The FPSS shall analyze plant first-aid and medical equipment needs and establish a network'of available emergency equipment to optimize emergency response in conventional and radiation areas. The FPSS shall maintain records of all health and safety related documentation.

The FPSS shall maintain-information regarding quantities and types of hazardous materials stored and used on-site; and develop appropriate strategies and resources for protecting station

-personnel from unacceptable exposures to hazardous materials.

The FPSS-will coordinate hazardous material information with the l Radiological Enginecing Section of the Radiological Controls Division of the o&M Department, as appropriate, in accordance with applicable regulatory requirenents.

The FPSS shall be responsible fpr development and review of all Fire Protection, Hazardous Material, and Safety training material including the training of a fire brigade and Hazardous Material Response Team. . The FPSS shall maintain a liaison between local, state,;and federal vn:les to ensure efficient response in any L emet .yency. The FPc4 e ill be responsible for oporation and maintenance of intox.lyzer equipment as required to support the Fitness for Duty Program.

C:)

13-9 Rev. 4 July 1992

SHOREHAM DSAR fg plant Administrative Section (O&M) t~) The plant Administrative Section - O&M is supervised by the Plant i Administrative-Coordii. tor (PAC) and shall be responsible for ,

providing direction tt 'he office organization, including plant personnel records, plane filing system, office procedures, miscellaneous office equipment and supplies, and reproduction equipment for the Operations & Maintenance Department. The PAC administers the flow of correspondence, specifications and drawings into and out of the plant. The PAC maintains, updates, and distributes plant procedures.

The PAC shall be responsible for the supervision of the secretarial, clerical, and other administrative office personnel

- required for the operations & Maintenance Department.

The PAC will interface with'the Operations & Maintenance Department, as appropriate.

Records Management and Document Control Section Tho' Records Management and Document control Section (RMDC) shall be responsible for administration support and control for procedures, records management, and document control. RMDC shall establish, Implement, and maintain the SNPS Records Management

( and Document Control Programs consistent with applicable

\ requirements.

13.1.2.1.3 Decommissionina Department F

The Decommissioning Department (DD) provides LIPA with engineering, construction, and special process support for the planning and execution of SNPS decommissioning.

The Decommissioning Department is organized into the following divisions:

• Decommissioning Engineering Division Construction Division .

• Special' Processes Division.

Decommissioning Engineering Division The Decommissioning Engineering Division (DED) sbs11 be responsible for providing engineering support ft the implementation of the SNPS Decommissioning Plan. This include 6

,g- development'of engineering-packages and safety evaluations (j required'to accomplish the decommissioning of SNPS.

13-10 Rev. 4 July 1992

- - - - - - - _ - . - ~ - . . - - _ . - - - - - - _

SHOREHAM DSAR The Decommissioning Engineering Division shall be responsible for

((_/'] the day-to-day engineering support of decommissioning activities, direction and performance of the principal Architect / Engineer, and resolution of technical issues related to decommissioning.

The Decommissioning Engineering Division shall review all decommissioning activities performed by the Decommissioning Department and assure that the activities are in compliance with the requirements of project specifications and the License.

The Decommissioning Engineering Division shall coordinate and interface with other station departments and divisions as necessary.

The Decommissioning Engineering Division shall be responsible for cost and schedule control of all activities under its cognizance.

Construction Diviclon The Construction Division (CD) shall be responsible for the implementation and performance of dismantlement and construction support activities necessary for the decommissioning of the station's systems, equipment and structures in accordance with project specifications, station policies and procedures, applicable regulatory criteria, and the Decommissioning Plan.

The activities conducted by this division shall encompass those dismantlement techniques and construction support activities that O are considered standard industry techniques requiring little or no plant-specific development, demonstration or qualification prior to use.

The CD shall be the focal point for the acquisition and direction of decommissioning craft labor and shall be responsible for the performance of the decommissioning General Contractor and any other construction subcontractors not under the direction of the General Contractor. The CD shall also assist with plane maintenance as needed and requested.

The CD shall coordinate and interface with other station departments and divisions as necessary.

The CD shall be responsible for cost and schedule control of all activities under its cognizance.

The.CD staff may be supplemented with competent construction personnel from outside contractors as necessary.

Special processes Division The Special Processes (SP) Division shall be responsible for the specification, selection, implementation, and performance of specialized decontar;ination and dismantlement methods _and O processes to be used during plant-decommissioning activities, including' project management of the disposition of the spent fuel.

13-11 Rev. 4 July 1992

l Sii0REllAM DSAR l Special methods and processes are tnose which require some level

. N(~. of plant specific development, qualification, and/or demonstration prior to use at Shoreham. (They should not be confused with the term.special processes used in conjunction with quality assurance processes.) Examples of special processes include wire rope and underwater plasme, arc cutting to be used on the reactor pressure vessel, and ultra high pressure water or .

chemical decontamination. l l

The 1P Givision shall be responsible for the acquisition, direction and performance of the specialty contractors selected '

tta develop and implement the required decommissioning special pracosses.

The SP Division shall be responsibic for ensuring that plant systems, components, and structures on which special processes are performed are decommissioned in accordance with project spec'.f ications , station policies and procedures, applicable regulatory criteria and the Decommissioning Plan.

The SP Division shall be responsible for project management of Aho implementation of the option selected for disposition of the spent fuel. This includes responsibility to ensure that the option selected is implemented by the app opriate specialty 1 contractor (s) in a-safe and efficient manner in accordance with applicable regulatory criteria and the Decommissioning Plan, .

witid n schedule and budget. Further, the SP Division is

()'-

responsible to ensure all spent fuel disposition activities are properly integrated with other maintenance and decommissioning activities.

The-SP Division shall be responsible to coordinate and interface with other station departments and divisions as necessary.

The SP Division shall be responsibla for cost and schedule control of all activities under its cognizance.

13.1.2.1.4 Licensina/Reaulatory Comnliance Department The. Licensing / Regulatory Compliance Department (LRCD) provides LIPA'with guidance regarding regulatory, licensing, nuclear safety, and environmental compliance, licensing commitment identification and tracking, and nuclear regulatory and licensing information support for the planning and execution of SNPS

. decommissioning.

Tho LRCD monitors the status of all regulatory, licensing,

-safety, and environmental compliance activities, licensing commitment status, and generic and plant-specific information regarding developments in nuclear regulation and licensing.

The LRCD shall perform the following:

13-12 Rev. 4 July 1992

.- . _ . _ = _ . . - -- _- - - =.-- . ~ -.. -

SHOREHAM DSAR

• Provide the justification for LIPA's exemptions, exceptions,

(s s or proposed alternatives for compliance with NRC regulatory criteria.

• Prepare amendments to SNPS Defueled Technical Specifications  !

and licensing basis documents, and related justifications, hs required. l

• Advise LIPA personnel regarding the proper scope and content of safety evaluationn required under 10CFR50.59.

• Prepare No Significant Hazards consideration evaluations for submittal to the NRC for Defueled Technical Specifications changes and other license amendments.

• Review and concur with and/or prepare changes to SNPS Defueled Safety Analysis Report (DSAR) pursuant with LIPA policy, and-bring the DSAR up-to-date in accordance with the requirements of 10CFR50.71(e), as required.

• Maintain overall schedule for meeting regulatory requirements and commitments and, as appropriate, obtain schedulo extensions from the NRC for completion of decommissioning commitments.

• Develop licensing strategies for maintenance and decommissioning activities.

(

• Review and concur with criteria selected for plant and decommissioning modifications and activities.

Review surveillance documentation and maintain the Master Surveillance Schedule.

• Maintain-required li.enses and permits and coordinate renewals as necessary.

• Coordinate LIPA interface with nuclear industry organizations such as the Nuclear Utility Management and Re- fces Council (NUMARC) and other industry forums.

• Interface with agencies of the State of New York for matters pertaining to compliance with state regulations.

• Provide support in determinations for events potentially requiring NRC notification under 10CFR50.9 and 10CFR50.72 and prepare the-Licensee Event Reports required under 10CFR50.73.

l 13-13 Rev. 4 July 1992 W 9

SHOREHAM DSAR

• Perform the LRCD functions specified in NOC Policy 24, t'~h " Corporate Evaluation and Reporting Responsibilities (ms/ Pursuant to 10 CFR Part 21."

• Provide support for and coordinate activities of the Site Review Committee (SRC) and Independent Review Panel (IRP).

The LRCD shall be responsible for the receipt of incoming nuclear licensing corresponde'tce and regulatory documents (both plant- .

specific and generic) and for reviewing such information to determine if a corporate position or response is required. The LRCD shall distribute such information to appropriate site and corporate organizations, and shall establish a strategy and schedule for the development of input for any required corporate positions or absponses. The LRCD shall be responsible for assigning input development responsibilities or other required actions, and for the coordination of input development, corporate review and comment resolution. The LRCD shall review draft input for responsiveness, compliance with regulations and consistency with corporate policy, and shall assemble a final document package for signature by the Resident Manager or Executive Vice President-Shoreham Project,-as appropriate.

The LRCD shall be responsible for the overall management, staffing, coordination, strategy and conduct of Atomic Safety and Licensing Board (ASLB) litigation, as well as any other litigation pertaining to nuclear licensing or safety issues. The O' LRCD shall be the primary interface with LIPA's legal counsel, and shall work closely with other LIPA, NYPA and LILCO organizations to assign technical resources, select witnesses and develop the LIPA strategy for a given issue.

The LRCD shall interface with the licensing organizations of other utilities; and coordinate licensing positions with other utilities, particularly those who have decommissioned or are

' decommissioning a nuclear power plant. The LRCD shall be responsible for recommending appropriate licensing actions in concert with other utilities upon obtaining concurrences from appropriate LIPA personnel.

13.1.2.1.5 Nuclear Ouality Assurance Denartment The Nuclear Quality Assurance (NQA) Department provides LIPA with quality assurance and control related licensing commitments, a Quality Assurance (QA) Program, and other LIPA administrative

-policy support'for the planning and execution of SNPS decommissioning.

NQA is responsible for establishing end maintaining a quality assurance program, documented by wrf tton polic ies, procedures or instructions, that meets the requirements of Appendix B to 10 CFR

/ 50 and other NRC requiremelits for the SNPE. This' program sets

.\ . - forth the requirements for quality relat2d activities performed 13-14 Rev. 4 July 1992 s

l SHOREHAM DSAR '

l by the various departments at the plant and all applicable LIPA contractors.

Refer to Chapter 17, Quality Assurance, for a detailed explanation of the Nuclear Quality Assurance Department.

13.1.2.1.6 Finance & Administration Denartment Onsite support to all the technical departz*nts of the SNPS are

supplied by the Finance & Administration (FA) Department.

This includes providing administrative support to the plant departments in budgetary, accounting, procurement, and material control matters during operations and throughout decommissicaing of the plant.

The FA Department provides LIPA with administrative, budgetary, and procurement and material control support for the planning and execution of SNPS decommissioning.

FA is organized into the following divisions:

• Accounting Procurement / Contract Administration Project Controls

()

• Materials Management Accounting Division The Accounting Division (AD) is responsible for tracking the historical cost of maintenance and decommissioning.

The AD will receive, track, and facilitato payment of vendor invoices.

It will maintain and reconcile thorough, timely, and accurate accounting records as required by the Site Cooperation and Reimbursement Agreement and Management Services Agreement dated January 24, 1990. AD will also interface with NYPA, LIPA and LILCO accounting personnel and coordinato implementation of accounting for costs attributable to Shoreham in addition to support of various audits of Shoreham records.

The AD is responsible for management of the accounting-related modules contained in the Power Authority Reporting and Information' System (PARIS). This includes reconcilation of data

-in Shoreham PARIS and the interfaces between Shoreham PARIS and other systems.

O 13-15 Rev. 4 July 1992

. _-. m, _. _ -_ - - _

SHOREHAM DSAR

-Procurement / Contract Administration

)

The Procurement / contract Administration Division (PD) is responsible for procurement of the materials, equipment, and services required to maintain and decommission shoreham. It is also responsible for administering contracts governing such procurements.

For purposes of this description, Procurement refers to all of the actions chich are required to obtain goods (equipment and -

materials) and services (r.ofessional, consulting, technical, subscriptions, etc.) from outside sources. Contract Administration refers to the actions that are required to obtain labor-burdened contracts from outside sources, the management of the commercial elements of performance within each contract, and the closing of each contract with reconciliation of contract values among the User Group, Accounting, Project Controls, and the Contractor.

The PD will:

• ensure that schedule needs-and commitments are reflected in each contract / purchase order.

• ensure that work scope is clear]f incorporated in each contract / purchase order.

• document contractor performance.

• minimize claims and risk of loss.

• ensure contract records are preserved.

• negotiate scope changes to achieve the most favorable commercial conditions.

implement a back-charge program for recovering monies in.the event of vendor non-performance.

The PD will receive and process approved purchase requests. As required, it'will solicit and conduct the evaluation process of bids for purchases,-and award pyrchase orders and contracts to the preferable bidder. .It will also maintain-the files of record for each purchase order and contract issued, from purchase request through the life of the P.O. and Contract.

-The'PD will comply with al] established procurement regulations and LIPA's applicable Quality' Assurance Program requirements in its purchasing activities. The PD will maintain relations with vendors consistent with corporate ethical standards.

O 13-16 Rev. 4 July 1992

_ _ _ ~ . _ - - - - . , _ _ . - . _ _ __

S110REllAM DSAR Project Controls

() The Project controls Division (PC) is responsible for three functions:

• Planning and Scheduling Cost Estimating Budgeting and Cost Control The Planning and Scheduling function will develop and maintain strategic schedules for maintaining and decommissioning Shoreham.

Strategic schedules are milestone-level schedules required by internal managers to properly perform their functions. The PC 1 will communicato schedules to other groups and solicit their i input for updating these schedules.  !

-The Cost Estimating function will develop and maintain the total cost estimate for maintaining and decommissioning Shoreham. It will'also develop a system for maintaining, tracking and reporting the total cost estimate and other estimates required during maintenance and decommissioning. This group will also revise the estimate as required by the Management Services Agreement (MSA).

The Budgeting and Cost Control function will develop and maintain

[D

\- /

budgets as required by the MSA and by Shoreham managers to control the project. It will develop and maintain systems for tracking and reporting commitments and charges against these budgets, and for forecasting maintenance and decommissioning

. charges yet to be incurred.

Budgeting and cost control will assist la validating that services invoiced by vendors were performed and that costs incurred are representative of work performed. It will develop and analyze-financial data and report these analyses to the appropriate managers. This group will also ensure that the appropriate accounting codes are assigned to purchase requests, commitments and charges.

Materials Management .

The Materials Managem Jivision (MD) is administrative 1y responsible for reci ving, storing, controlling, and issuing material and equipment to be used in maintenance and decommissioning. It wiil control the on-site warehouse and interact with LILCo to reserve and obtain existing inventory to use in maintenance-and decommissioning.

The MD will communicate with tne Operations & Maintenance and

/~'

ds Decommissioning Departments tt plan materials needs and to schedule materials purchases to coincide with scheduled work L activities. It will also assist the Quality Assurance Department 13-17 Rev. 4 July 1992 l ..

I

-- ,~ n ,,-, , - _ . , . . , -, , _ . , - _ - , . . . , , - .

['

SHOREHAM DSAR in ensuring safety-related materials meet or exceed LIPA's

(')

L.)

requirements and are properly controlled.

The MD will maintain appropriate records of and documentation for LIPA-purchased and/or controlled materials. It will also review the adequacy of and compliance with LILCO's shelf life and inventory preventive maintenance programs.

13.1.2.2 Plant Personnel Respfnsibility and Authority The functions, responsibilities and authorities of key station personnel are delineated in the position descriptions contained in the LIPA Shorheam Nuclear Power Station Administrative Manual.

<he qualifications for the positions described therein meet the requirements of ANSI N18.1-1971, as appropriate for the permanently defueled status of SNPS, for comparable positions.

13.2 TRAINING PRQ_ GRAM The description under this heading in the Shoreham USAh is superseded by the following.

13.2.1 Trainina To Support Maintenange In The Defueled Condition

',m s

) In the defueled state, with the NRC operating license amended to remove operating authority, there is no requirement tr ,aintain accredited training programs since the plant is no P av ;

licensed to operate.

The LILCO Office of Training has non-nuclear training programs available to LIPA, developed via a " systematic approach to e training" method, which can be requested by the Shoreham plant management for training of operators, technicians, and mechanics, s

The Office of Training procedures outline the methods to be used s" to analyze training needs, and to establish or conduct required training. The Office of Training staff will be qualified in accordance with the LILCO " Training and Qualification Program".

Opera ors: Operators will be trained (or have been trained previcusly under LILCO .snership of the SNPS license) in the function and operation of those systems required to be operational during the defueled phase. The material used to conduct this training will be from the operator training program developed

• or nuclear operations.

This operatcr training program was originally developed by LILCO in order to license reactor operators and senior reactor-operators in accordance with 10CFR55 for low power and then full (k/)

power plant operation. Following isruance of the Shoreham Possession only License (POL), however, LILCO received an exemption from 10CFR55 allowing the licensed operator 13-18 Rev. 4 July 1992 l

SHOREHAM DSAR re-qualification programs to be reduced commensurate with the

(~N . more limited scope of activities authorized by the Shoreham POL.

\mj- When the POL was transferred to LIPA, in turn, LIPA received

-permission to eliminate altogether the need for operators to be licensed under 10CFR55, as well as permission to adapt the reduced-LILCO-licensed operator requalification program for use as a LIPA certified operator requalification program This approval also included permission for or? ators who c ~'CFRSS license qualifications under LILCO had not expired o initially 1 certified by LIPA for the remainder of thA qualification term without a'new examination.

Equipment Operator: Field operators will be trained (or have been trained) using portions of the Equipment Operator Training Program developed for nuclear operations. This training will include generic, non-nuclear theory, and the function and operation of thoes systems required to be operational during the defueled phase.

Control Technicians: Control technicians and computer technicians will be trained (or have been trained) in accordance with the Control Technician training program developed for power plant technicians.

Mechanics / Electricians: Mechanics / electricians attend formal training'as part of LILCO's maintenance training programs. These j- programs qualify mechanics / electricians as apprentices with journeyman qualAlications available in the area of welding, N2 rigging, machinery,-electrical, and general maintenance skills.

The Shoreham maintenance force will be trained and qualified in

'accordance with' existing LILCO maintenance training programs.

This program is not available for contract maintenance work forces; contractors would provide qualified mechanics and electricians.

Rad' Chem / Health Physics: The Radiochemistry and Health Physics technicians will be trained (or have been trained) using the training material developed for Health Physics and Rad Chem technicians for-nuclear operation. However, the training will be limited to fundamentals and task specific training as required to support Rad Chem, Health Physics, and Radwaste operations during the defueled condition.

13.2.2 Traininn To Suonort Decommissionina Activities The Training Program for decommissioning is described in Section 2.4 of the LIPA Shoreham Nuclear Power Station Decommissioning Plan.

13.3 EMERGENCY PLANNING

~The description under this heading in the Shoreham USAR is superseded by the following.

} .The-EmergeaNy Plan for the Shoreham Nuclear Power Station is submitted as a-corporate document titled "LIPA Defueled Emergency 13-19 Rev. 4 July 1992

SHOREHAM DSAR Q}

Preparedness Plan," which was adapted from the prior LILCO Defueled Emergency Preparedness Plan. Changes from the LILCO document were limited to those necessary to reflect LIPA as the Shoreham licensee.

13.4 REVIEW AND AUDIT The following information supersedes the information under this heading in the Shorebem USAR with respect to review and audit of activities conducteo ' aer the POL by LIPA.

A review and audit program, including in-plant and independent reviews, have been developed to: provide a system to ensure that plant design, operation, and decommissioning are consistent with company policies and rules, approved procedures, and license provisions;-review important proposed plant decommissioning changes, tests, experiments, and procedures; assure that unusual events are'promptly investigated and corrected in a manner that reduces the probability of recurrence of such events; and detect trends that may not be apparent to a day-to-day observer.

Review and audit during operating and decommissioning of the

(s Shoreham Nuclear Power Station (SNPS) is an integral part of the Long Island Power Authority (LIPA) Quality Assurance Program.

i J-Provisions are established for a comprehensive system of planned and periodic audits to verify-implementation of Quality Assurance Program requirements. These review and audit functions are fully described in Chapter 17, Quality Assurance, of this Defueled Safety-Analysis Report (DSAR). In addition, LIPA utilizes a formal committee method for review and audit cognizance, functioning at'two levels:

- 1. - At the station-operation level, the Site Review Committee (SRC)-

2. At the corporate level, the Independent Review Panel (IhP),

which-is independent of direct responsibility for plant maintenance and decommissi6ning.

The review and audit program has been established to assure that

-the operation and-decommissioning of the plant is in conformance with estri'ished procedures, license provisions, and quality assurance sequirements and to review and approve changes to station systems / equipment and procedures as described in the DSAR or tests and experiments, which do not constitute an unreviewed nuclear safety question, as defined in 10 CFR, Part 50.59. All unreviewed safety questions and changes to the Technical

( s). Specifications are reviewed by the IRP as described below.

13-20 Rev. 4 July 1992

SHOREHAM DSAR N

-l }-

L.J Reviews of nuclear safety related questions are made by the IRP as described-below.

-A continuing. review /is performed by the SRC to monitor plant

. maintenance, and plan future decommissioning activities, and to screen subjects that might be of interest to the IRP.

13.4.1 Site Review Committee The SRC.shall function to review plant operations and advise the Resident Manager on all matters related to nuclear safety, radiological and/or environmental protection, and decommissioning activities.

WRITTEN CHARTER A written charter has been prepared covering such areas as group responsibility, subjects requiring review, reporting requirements, and organization.

The charter of the SRC reflects the consideration that committee

.rst review responsibilities extend to all station activities and

' ')

t s

proposed-changes or modifications to station systems or equipment and are not limited to those designated safety related.

COMPOSITION The-SRC shall be composed of a chairman or alternate chairman and

-six or more members-or alternate members of the plant staff as designated by-the chairman.

ALTERNATES.

All alternate members shall be appointed in writing by the-SRC Chairman to serve on a temporary basis; however, no more than one alternate shall participate as a voting member in SRC activities at any one time.

MEETING FREQUENCY The SRC shall-meet at least once per calendar month and as convened by the-SRC Chairman or'his designated alternate.

QUORUM The quorum of the SRC necessary for the performance of the-SRC responsibility and authority provisions under the Defueled l,t. . Technical Specifications shall consist of the Chairman or his

'\/ designated alternate and four other members including alternates.

13-21 Rev. 4 July 1992

. -- = . -. _ . _ .

SHOREHAM DSAR O

s/m RESPONSIBILITIES The SRC shall be responsible for:

a. Review of (1) all proposed procedures and programs required by Defueled Technical Specification 6.7 and changes thereto, and (2) any other proposed procedures or changes thereto as determined by the Resident Manager to affect nuclear safety;

b. Review of all proposed tests and experiments that affect nuclear safety;

c. Review of all' proposed changes to the Possession only License and Defueled Technical Specifications;

d. Review of all proposed changes or modifications to unit systems or equipment that affect nuclear safety;

e. Investigation of all violations of the Defueled Technical Specifications, including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence, to the Chairman of the Independent Review Panel

/'"% (IRP) and the Executive Vice President-Shoreham Project; d f. Review of all reportable events;

g. Review of decommissioning activities and facility operations to detect potential nuclear safety hazards;

h. Performance of special reviews, investigations, or analyses and reports thereon as requested by the Resident Manager, any

. member of the SRC, or the Chairman of the IRP;

i. Review of the Security Plan and implementing procedures;

j. Review of the Defueled Emergency Preparedness Plan and implementing procedures;

k. Review of the Fire Protectibn Plan and implementing procedures;

1. -Review of the proposed changes to the Process Control Program (PCP);

m. Review of the proposed changes to the offsite Dose Calculation Manual (ODCM);

n. Review of the proposed major changes to radioactive waste systems; I

l 13-22 Rev. 4 July 1992

. . - .= .. -

SHOREHAM DSAR A

O

o. Review of Personnel Radiation Records annually to determine how exposures might be lowered consistent with ALARA principles. Document such considerations; and

p. Review of any accidental, unplanned, or uncontrolled onsite release of radioactive material, including the preparation of reports covering evaluation, recommendations, and disposition of the corrective action to prevent recurrence and the forwarding of these reports to the Executive Vice President-Shoreham Project and to the IRP.

q. Quality review of ALARA Review Committee (ARC) activities.

r. Review of proposed changes to the approved Decommissioning Plan.

The SRC shall:

a. . Recommend in writing to the Resident Manager approval or disapproval of items considered under items a through e and 1 through n above prior to their implementation.

ri b. Render determinations in writing with regard to whether or

_ f,' not each item considered under items a through e and 1 through n above constitutes an unreviewed safety question. .

c.- Provide written notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, to the Executive Vice President-Shoreham Project of disagreement between the SRC and the Resident Manager; however, the Resident-Manager shall.have. responsibility for resolution of such disagreements pursuant to Shoreham Defueled Technical Specifications.

d. Function to advise _the Resident Manager on all matters related to nuclear safety, radiological environmental operations, and decommissioning activities.

RECORDS The SRC shall maintain written minutes of each SRC meeting that, at a minimun,. document the results of all SRC activities performed under the " Responsibilities" Section of Defueled

-Technical Specification 6.5. Copies shall be provided to the Chairman of the IRP and the Executive Vice President-Shoreham Project.,

13.4.2 Independent Review Panel (IRP)

FUNCTION (vT The IRP shall function to pr: vide independent review and audit of designated activities in the areas of:

13-23 Rev. 4 July 1992

SHOREHAM DSAR

(_I' .

l a .- Nuclear engineering,

-b. Chemistry and radiochemistry,

c. Radiological safety, d .- Mechanical and electrical engineering, and

e. Quality assurance practices.

The~IRP shall report to and advise the Executive Vice President of.Shoreham Project.

WRITTEN CHARTER A written charter has been prepared covering such areas as group i responsibility, subjects requiring review, reporting l requirements, and organization.

The charter of the IRP reflects the consideration that IRP 1 activities are not limited.to items and functions that are designated as safety related. It is intended that IRP review and audit activities will.also cover non-safety related structures, systems, components, and plant computer software to ensure that

.the safety. significance'given to them in the DSAR, the Technical f- Specifications, and the Emergency Operating Procedures will be (S ,/ maintained during the operation of Shoreham.

COMPOSITION The IRP shall be composed of the permanent IRP Chairman and a minimum of four permanent IRP members. The chairman and all members of the IRP shall have qualifications that meet the requirements of Section 4.7 of ANSI /ANS 3;1-1978.

The membership shall include at least one individual from outside LIPA's or its contractors' organizations and at least one

. individual with substantial nuclear experience. The nuclear experience may be providad by the individual who is from outside LIPA's or'ite contractors' organizations.

MEETING FREQUEhCY .

The'IRP_shall-meet at least once per six months.

QUORUM LThe quorum of-the IRP necessary for the performance of the IRP review functions of the-Technical Specifications shall consist of theLChairman orfhis' designated alternate and at least three but not'less than-one-half of the IRP members present including-f"')

L>

alternates. No more than a minority of the quorum shall have line responsibil-ity for operation of the unit.

13-24 Rev. 4 July 1992

SHOREHAM DSAR

.O Y.l REVIEW The IRP shall review:

a. The safety evaluations-for (1) changes to equipment or systems and (2) tests-or experiments completed under the provisions of 10CFR50.59 to verify that such actions did not constitute an unreviewed safety question;

b. Proposed changes to procedures, equipment, or systems which involve an unreviewed safety question as defined in 10CFR50.59;

-c. Proposed tests or experiments which involve an unreviewed safety question as defined in 10CFR50.59;

d. Proposed changes to Technical Specifications of this Possession Only License;

e. Violations of codes, regulations, orders, Technical Specifications, license requirements, or-of internal

- procedures or instructions having nuclear safety l significance;

f. Significant deviations from normal and expected performance of station equipment that affect nuclear safety;

g. All REPORTABLE EVENTS;

h. All recognized indications of an unanticipated deficiency in some aspect of design or operation of structures, systems, or components that could affect nuclear safety; and-1.- Reports and meeting minutes of the SRC.

Audits of station activities shall be performed under the cognizance of the IRP. These audits and audit frequencies are as

-follows: ,

a. The conformance of station operation to provisions contained within the Technical Specifications and-applicable license conditions at least once per 12 months; b.. The. performance, training and qualifications of the entire staff at least;once per 12 mo' hs;

c. The results of actions taken to correct deficiencies

(~N occurring in unit equipment, structures, systems, or method l

(_) . of operation that affect nuclear safety, at least once per

year; L

13-25 Rev. 4 July 1992 i

SHOREHAM DSAR I/ \.

()

d. The performance of activities required by the Quality Assurance Program to-meet the criteria of Appendix B, 10 CFR Part 50, at least once per 24 months; e- The-fire protection programmatic controls including the implementing procedures, equipment and program implementation at-least once per 24 months utilizing either a qualified offsite licensee fire protection engineer (s) or an outside independent fire protection consultant;

f. Any other area of station operation considered appropriate by the~IRP, the President of Shoreham Project or the Executive Vice President of Shoreham Project;

g. The Radiological Environmental Monitoring Program and the results thereof at least once per 12 months;

h. The-Offsite Dose Calculation Manual and implementing procedures at least once per 24 months; and

i. The Process Control Program and implementing procedures for solidification of radioactive wastes at least once per 24 r months.

~

j. The performance of activities required by the Quality Assurance Program for effluent and environmental monitoring at least once per 12 months.

RECORDS Records of IRP activities shall be prepared, approved, and distributed as indicated below:

a. Minutes of each_IRP meeting shall be prepared, approved, and forwarded to the President of Shoreham Project and the Executive Vice President of Shoreham Project within 14 days following each meeting,

b. Reports of-reviews encompassed by Technical Specification 6.5.2.7 shall be prepared, approved, and forwarded to the President of Shoreham Project and the Executive-Vice President of Shoreham Project within 14 days following completion of the review.

c. Audit reports encompassed by Technical Specification 6.5.2.8 shall be forwarded-to the President of Shoreham Project,

. Executive Vice President of Shoreham Project and to the management positions responsible for the areas audited within f~ j 30-days after completion of the audit by the auditing

%# organization.

13-26 Rev. 4 July 1992

SHOREHAM DSAR

/%.,

( )

v 13.5 STATION PROCEDURES The description contained under this heading in the latest revision of the Shoreham USAR remains unchanged except for the following:

13.5.1 Administrative Control

1. Safety-related station procedures shall be processed through the Site Review Committea (SRC) and Nuclear Quality Assurance (NQA).

2. The Resident Manager shall approve Station Administrative Procedures, Security Plan Implementating Procedures, and Emergency Plan Implementing Procedures prior to implementation.

3. Other Station Operating Procedures shall be approved by the appropriate Division Manager or by the Operations and Maintenance Department Manager prior to implementation.

4. Revised Table 13.5.1-1 is attached.

,O t  !

'J 13.5.1.1 Normal Operations The NRB has been replaced by the IRP in accordance with 13.4.2 and a new Table 13.5.1-1 is supplied herein.

13.5.1.2 Routine Maintenance, Repairs, and Fuel Handlina LIPA QA personnel shall be responsible for the auditing of procurement documents to ensure that appropriate quality control requirements are fulfilled as defined in Section 17.2.

13.5.1.3 Modifications The LIPA Site Review Committee is responsible for review of.

proposed modifications to safety related systems or components.

13.5.2 Procedures Changes to subsections.of USAR section 13.5.2 as a result of the permanently defueled plant configuration are identified below.

Other information remains as described in the USAR.

V 13-27 Rev. 4 July 1992

SHOREHAM DSAR

-n 1

^ {Q .

13.5.2.1 Operatina-Procedures 1._ The General Operating Procedures now only describe integrated station operation. Startup and Shutdown are no longer pertinent.

+ 2. Operating Procedures are not necessarily performed by, or under the direction of, persons holding RO or SRO licenses.

13.5.2.1 Initial Test Procedures This section is no longer pertinent.

13.5.2.2 Shoreham Nuclear Power Station Emeroency Preparedness Plan The " Radiological Emergency Preparedness Plan" has been replaced with a "Defueled Emergency Preparedness Plan" as indicated in section 13.3 of the DSAR.

13.5.2.3 Temocrary Procedures

'A '

! Temporary procedures-for refueling are no longer required at

• Shoreham.

13.6 PLANT RECORDS The description: contained under this heading in the latest revision of Shoreham USAR remains unchanged-except that the

-Manager,J Operations and Maintenance Department or his designee shall!be responsible for the_ compilation of operating records and event records _as set forth in the Station Administration Procedures.

13.7 INDUSTRIAL SECURITY The Security Plan, Training and Qualification Plan, and the Safeguards Contingency Plan for,the Shoreham Nuclear Power

. Station have been submitted as separate documents. These documents are withheld from public disclosure pursuant to 10CFR2.79(d), " Rules of Practice." The Security Plan and the Safeguards Contingency' Plan ~are also withheld from public disclosure pursuant to 10CFR73.21, " Requirements for the Protection'of Safeguards Information."

A 13-28 Rev. 4 July 1992

SHOREHAM DSAR

- x_./ .

TABLE 13.5.1-1 PROCEDURES PROVIDED FOR SHOREHAM NUCLEAR POWER STATION A. Administrative Procedures shall be provided to cover the following types of administrative activites:

1. - -Authorities and Responsibilities for Safe Fuel Handling Operations

2. Equipment Control (e.g., locking and tagging)

3. Procedure Adherence and Temporary Change Method 4.- Procedure Review and Approval

5. Schedule for Surveillance Tests 6.- Shift and Relief Turnover - Recall of Pe*sonnel

7. Log Entries and Record Retention

8. Bypass of Safety Functions and Jumper Control

9. Operating Orders

10. Special Orders

11. Materials Control

12. Radiation Work' Permits

13. Access Control to Controlled Area

14. Personnel Training and Qualification O' B. Operatina e Procedur_es NJ

1. General Oparating Procedures have been provided to cover the foAlowing Integrated Plant Operating Activities:

(

a. Surveillance.

2. -

System Operating Procedures shall describe Startup, Normal Operating, and Shutdown for the designated system. Abnormal Operation, where1 required, shall be contained in'a section of the System operating Procedure. Procedures are available for operating the

~

systems listed in a through ac, below.

a. 138kV and 69kV Power System

b. Normal Station Service Transformer

-c. Reserve Station Sbrvice Transformer

d. Well Water-System

e. 4,160 V System

f. .480 V System

g. Station Lighting Panels

h. 120 V'ac Instrument Bus

.i. 120 V ac Reactor Protection System Bus

j. 120 V ac Uninterruptible Power Supply

k. 125 V de System

(

1 of 3 Rev. 4 July 1992 I

k l SHOREHAM DSAR e

p kms/ TABLE 13.5.1-1 (Cont'd)

B. Operatina Procedures (Cont'd.)

'l. -Reactor Building Normal Ventilation System (RBNVS)

m. Service Water n._ Radwaste (Liquid)

o. Radwaste (Solid)

p. Communications System

q. Condensate Transfer

r. Deluge and Sprinkler System

s. Demineralized Water Transfer

t. Equipment and Floor Drains

u. Fire Protection System

v. HVAC - Control Rooni W. HVAC -. Turbine Building

x. HVAC - Radwaste Building

y. Makeup Water Treatment

z. Station Air System aa. Smoke,-Temperature, and Flame Detection System ab. -Turbine Building Closed Loop Cooling System a c .- CRAC Chilled Water

3. _ Emergency Procedures have been provided for combatting

- the following potential emergency conditions:

\~# a. Acts of Nature

b. Abnormal Releases of Radioactivity

c. Fuel Handling Accident

d. Plant Fires

e. -Loss of Electrical Power

f. -Ioss of-Service Water

4. Abnormal Operation Procedures required to mitigate the consequences of the following abnormal conditions shall be contained in the appropriate System Operating Procedures (s):

a. None.

,i') ,

Qf 2 of 3 Rev. 4 July 1992

l i

SHOREHAM-DSAR ym ,/ TABLE 13.5.1-1 (Cont'd)

Note: Procedures not designated as emergency procedures shall be incorporated in the Abnormal Performance section of the appropriate system or general operating procedures.

C. Alarm Response Procedures (ARP)

Alarm Response Procedures shall be provided as required for-alarm windows in the main control room associated with the operation of safety related systems or equipment.

D. Fuel Handlina Procedures shall be provided to cover the following fuel handling activities:

1. Special Nuclear Materials Control and Accountability Procedures 2._ Spent Fuel Handling and Shipment

3. Handling and Storage of Sealed and Unsealed Sources E. Health Physics Procedures shall be provided to cover the following radiation protection activities:

1. Dose Rate Radiation Surveys t 2. Surface Radioactive Contamination Surveys

\ 3. Personnel Contamination Survey

4. Personnel Decontemination

5. Areas and Equipment Decontamination

6. Monitoring for and Collecting and Recording of Occupational l Radiation Exposure (ORE) data

.7 . Submission 1and Review of Suggestions by Plant Personnel for the Reduction of ORE 8.- Use of Protective Clothing and Respiratory Equipment F. Defueled Emercency Preparedness Implementina Procedures (DEPIPs) shall be provided to cover the following emergency plan activities:

1. Emergency Classification

2. Evacuation and Personnpl Accountability

3. -Operational Assessment and Damage Estimates

4. Support Systems and Activation

5. Surveys,-Analyses, Sampling, Assessment, and Actions

6. -Personnel and Equipment Decontamination

7. Notifications

8. Re-entry and Recovery

9. Emergency Organization, Drills, and Training L -

%)

! 3 of 3 Rev. 4 July 1992

TABLE 13.5.1-3 FORMAT FOR STATION PROCEDURES

\/ *SP Number Revision Eff. Date Sionature Date TPC No. Date Eff Date Exor Section Head

-Quality Control-Div. Mgr.

Resident Mgr.

Signature or N/A Tll.L3 1.0 PURPOSE A brief description of the purpose for which the procedure is intended should be clearly stated. If the procedure is used to satisfy, in any part, a Technical Specification surveillance requirement, indicate the

()%

(_ Technical Specification number here'.

-2.0 RESPONSIBILITY Indicate the_ person directly responsible for ensuring the proper implementation of the procedure.

3.0 DISCUSSION Provide a brief-description of the applicable component, system, or task in sufficient detail for a knowledgeable individual to perform the

. required function without direct supervision. Include a list of topics or a table of contents generally describing the extent or scope of the procedure, with-page location.

• For temporary procedures, SP NumbeE assignment is TP YYXYYY.YY.

l l

's 1 of 6 Rev. 4 July 1992 p:

l;

SHOREHAM DSAR

[^ k TABLE 13.5.1-3 (Cont'd)

L J-EVENT ORIENTED EMERGENCY PROCEDURE FORMAT Submitted: SP Number (Section Head)

Approved: Revision (Operations Manager)

Effective Date TITLE *

• Should be worded to- J ndicate the purpose of the procedure.

1. - SYMPTOMS: Symptoms should be included to aid in the identification of the emergency. This should include alarms, operating conditions, and probable magnitudes of parameter changes. If a condition is peculiar only to the emergency under consideration, it should be listed first.

j, AUTOMATIC ACTION: (Delete if not pertinent)

^~$ . IMMEDIATE ACTION: -These steps should specify immediate action for operation of controls or confirmation of automatic actions that are required to stop the degradation of conditions and to mitigate the' consequences of degraded conditions.

4. SUBSEOUENT ACTION: Steps should be included to return the reactor to a normal shutdown period under abnormal or emergency conditions.

5. FINAL CONDITIONS: These steps should-specify-the documentation, authorizations, and plant conditions that nust be completed-prior to resumption of Normal Operation, defined in 22XYYY.YY.

9

6. DISCUSSION: A brief explanation of the procedure.

This section should contain background information, causes, effects, and other information that may assist-in clarifying the procedure and analyzing symptoms.

Note: Attempt to get 1, 2 and'3 on cover page of procedure to allow rapid evaluation and action by the operator.

\

[v) 4 of 6 Rev. 4 July 1992

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, Figure 13.1-1 LIPA Corporate and Plant Organization- I

Defueled Safety Analysis Rep <xt' '

... Revision 4: ' July 1992.  ;

SHOREHAM DSAR d_ CHAPTER 15 N. ACCIDENT ANALYSIS 15.1 GENERAL Analytical Obiective Chapter 15 of SNPS USAR provides the results of analyses of the spectrum of transient and accident events which are postulated to occur with the plant-operating initially at up to maximum power.

The purpose of this analysis is to identify USAR transients and accidents that. apply to the storage and handling of the low burnup fuel. -

The analysis is based on the defueled condition of the plant, i.e., the fuel is removed from the core and is stored in the spent fuel pool. The total decay heat is approximately 550 watts, which is small enough that it could be removed by passive cooling and would not require the fuel pool cooling system.

Normal and emergency makeups are discussed in Chapter 9.

As the reactor will not be operated and the fuel is not in the reactor, most of the USAR Chapter 15 events cannot occur.

Approach to Safety Analysis The safety parameter evaluated for each transient of USAR Chapter 15 is the Minimum Critical Power Ratio (MCPR) which is a measure of fuel cladding integrity. Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) is the safety parameter for the_ reactor LOCA-related-accidents, and indicates whether the peak cladding '

temperature and the zirconium-water reaction is below the specified limits. As the decay power level is extremely low during spent' fuel storage, and will not increase, MCPR and MAPLHGR limits cannot be exceeded and are not applicable.

Those transients and accidents of USAR Chapter 15 which pose the potential'for a radiological release outside the primary containment are of primary concern.

Heat Generation-Analysis One result from the ORIGEN2 calculation is a graph of decay heat or thermal power (in watts), as a function of time. Results of 4 this analysis are presented in Figure 15.1-1 The calculated

-decay heat load as of' June 1989 is approximasaly 0.55 kw.

It must be recognized that there are some limitations in the ORIGEN2 model, and potential inaccuracies in the calculational processes of the code and its supporting data sets. For-O% instance, ORIGEN2 is a " point reactor" model, and cannot deal conveniently with the spatial variations in fuel enrichment and 15-1 Rev. 4 July 1992 h --

. _ _ _ _ _ _ - _ _ _ _ _ . _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ . _ .- _ _ _ . _ . _ .- - _ . _ _ _ . _ _ _ _ _ _ = - - - . -

SHOREHAM DSAR

.,m burnup. In addition, there are uncertainties associated with

(\ ') ~ averaging-of nuclear cross-section data within the thermal, resonance,_and fission neutron energy ranges. Nevertheless, it is not expected that large uncertainties should occur in heat load estimates, see the comparison of calculated to measures dose; rates in-DSAR Section 12.2. This gives evidence that the decay heat load calculations are reasonable, as the same analysis (ORIGEN2) was used to generate both sets of data.

Analytical Cateaorieg Each USAR Chapter 15 event is assigned to one of six analytical categories. The analytical categories and the events in each analytical category are discussed below.

1. Decrease in Core Coolant Temperature This analytical category of USAR Chapter 15 events includes the following events:

15.1.5 Pressure Reculator Failure - Open 15.1.7 Feedwater Controller Failure - Maximum Demand 15.1.8 Loss of Feedwater Heatina

- O() 15.1.9 Shutdown Coolino (RHR) Malfunction - Decreasino Temperature.

In the_ spent fuel storage condition, the pressure regulator, feedwater controller, feedwater heating system and'RHR-system are not_ operating and all four transients are, therefore, not applicable.

.2. Increase in Reactor Pressure Since the generator, turbine, main steam isolation valve, pressure regulator, feedwater system, condenser and RHR systems are not operating in support of nuclear fission, the following-transients are not applicable:

15.1.1 Generator Load Reiection 15.1.2. Turbine Ttin 15.1.3 Turbine Tric with Failure of Generator Breakers to Open 15.1.4 Main-Steam Isolatien Valve Closure 15.1.6 Pressure Regulator Failure - Closed p-w' 15.1.18 Loss of Feedwater Flow 15-2 Rev. 4 July 1992

N SHOREHAM DSAR gg 15.1.21 Loss of Condenser Vacuum A

}- 15.1.26 Core Coolant Temperature Increase The transient of this category applicable to rpent fuel storage is the following:

15.1.19 Loss of AC Power A loss of AC power condition can be postulated that will affect normal support systems. However, because of the very low heat generation rate (see Figure 15.1-1) and large thermal capccity of the pool active fuel pool cooling is not required. Loss of the spent fuel pool water makeup capability will result only in a very slow evaporative loss of the pool water. This evaporation rate is so slow that ample time exists to restore normal pool makeup sources so that pool level can be quickly restored. Thus, the passive protection provided by-the spent fuel pool and low fuel decay heat eliminate the need for active makeup requirements. (The rate of evaporation is discussed in Chapter 9.)

The loss of AC power will not in itself result in any release of radioactivity, as fuel movement is disallowed by Tech Specs when AC power is lost (and is virtually fr T . impossible in any event), and the decay heat of the core is (m,/ so low. Should the loss of AC power occur as-part of any other event which causes damage to the fuel in the pool, while the release in this case would not be monitored, the offsite dose consequences would be insignificant. Doses and

. dose rates are bounded by the " puff release" results given in Sections-15.1.36 and 15.1.36A.

3. Decrease in Reactor Coolant Flow Rate The recirculation pumps and recirculation flow controller are not operating in the defueled condition and therefore all the transients of this category are not applicable:

15.1.20 Recirculation Pumo Trin 15.1.22 Recirculation-Pum'o Seizure 15.1.23 Recirculation Flow Control Failure With Decreasina Flow

-4. Reactivity and Power Distribution Anomalies Events included in this category are those which cause rapid increase in power. Since the reactor is defueled, the

~s following events are not applicable:

\

15-3 Rev. 4 July 1992

SHOREHAM DSAR I 15.1.11- Continuous Control Rod Withdrawal Durina Power Rance Coeration 15.1.12 Cpntinuous Control Rod Withdrawal Durina Reactor Startuo L15.1.13 Control' Rod Removal Error Durina Refuelina 15.1.14 Fuel Assemb1v Insertion Error Durina Refuelina 15.1.15 Off-Desian Operational Transient Due to Inadvertent Loadina of a Fuel Assembly into an Improner Location 15.1.16- Inadvertent Loadina and ODeration of a Fuel Assembly in Improper Location 15.1.24 Becirculation Flow Control Failure with Increasina Flow 15.1.25- Abnormal Startun of Idle Recirculation Pump 15.1.33 Control Rod Drop Accident

5. Increase in Reactor Coolant Inventory b(,j Since the HPCI system is not required the following transient is not applicable:

15.1.10 Inadvertent HPCI Puno Start

6. Decrease in Reactor Coolant Inventory 6.A Events Not Acolicable to Spent Fuel Storace The' safety relief valve and the feedwater system are not operating in the defueled condition; therefore the following events are not applicable:

15.1.17 Inadvertent Openina of a Safety Relief Valve 15.1.37 Feedwater System-Pinina Break The following event is not a design basis event and is applicable _only to power operation:

15.1.27 Anticipated Transient Without Scram (ATWS)

The single failure-proof polar crane design eliminates the following event:

15-4 Rev. 4 July 1992-

E SHOREHAM DSAR I). 15.1.28 _qqsh_Rro.p Accident v

Instrument line, coolant line and steam line breaks present no consequences due to their lack of. interaction with the fuel and therefore the following events are not applicable:-

15.1.30 Off-Desian Onerational'Trangia"* Ap a Consecuence of Instrument Line Failure 15.1.34 Pioe Bree.ks Inside the Primary Containment (Loss-of-Coolant Accident) 1 15.1.35 Pine Breaks Outside the Primary Containment (Steam Line Break Accident)

-6.B Events Without Fuel Damace 15.1.29 Miscellaneous Small Releases outside Primary Containment 1

Releases that could result from piping failures outside the primary containment include the pipe breaks in the fuel pool cleanup system. The resulting offsite dose will be ,

negligible and are bounded by the Radwaste Tank Rupture accident.

p.

( ,) 15.1~.29.1 . Seismic Event Because the spent fuel pool structure and fuel racks and handling-equipment meet seismic category I requirements, a seismic event is not postulated to create a radiological release. However,-certain predecommissioning and decommissioning activities may involve the temporary use of QA/ Seismic Category II structures, systems and components which could fail during a seismic event, may damage fuel and may create a radiological release. No credible seismically induced' accident will exceed the bounding radiological release postulated in.Section 15.1.36A. Therefore, the radiological consequences of this very low probability event are bounded by those already analyzed and reported in Section 15.1.36A.

15.1.31- Main' Condenser Gas Treatment System Failure As the main condenser is not operating, there can be no offsite dose resulting from this event.

15.1.32 Licuid'Radweste Tank Ruoture Should accident occur radioactivity could be released to the environment but the effect would be negligible. The (e)

\/

accident analysis described in DSAR Section 11.2.3.4.2 and 11.2.3.4.3 proves this.

15-5 Rev. 4 July 1992

SHOREHAM DSAR 15.1.36.6.1.3 Radioloaical Effects

_n!

4 V Offsite Radiological exposures have been evaluated for the

-meteorological conditions, parameters, and assumptions given in Table 15.1.36-1. The results are given in Table 15.1.36-2.

Control Room Because the amount of radioactivity released is so small, the control room air intake monitors will not alarm and are not required. The control room HVAC system will continue to function in its normal operating mode. The resultant whole body and skin 30-day integrated doses are, at most, 9.59E-08 and 2.08E-04 mrem, respectively, well below the 10CFR50 GDC 19 limits.

Discussion It is seen in Table 15.1.36-2 that the (0-2 hour) EAB and (0-30 day) LPZ integrated doses are many orders of magnitude below 10CFR100 guidelines. Results are graphically shown in Figure 15.1.36-1. Furthermore, th' maximum (t=0) dose rates (whole body and skin) are very low and, with the exception ,

e of the RBNVS case, below Technical Specifications. This

(

) indicates that the HVAC system in use in the reactor building has no meaningful effect on radiological consequences to members of the public during a fuel handling accident with the present fuel source terms.

15.1.36A Worst Case Fuel Damace Event Scenario Several " worst case", extremely conservative scenarios were examined. Specifically, for the three reactor building HVAC cases analyzed in Section 15.1.36.5 (RBSVS operating, RBNVS operating, and puff release), instead of assuming the gap activity from 125 fuel rods is released (2.52 Ci Kr-85), it is assumed that all gaseous activity from the entire core in the spent fuel pool is relpased (1.56E+03 Ci Kr-85). This can only occur if all the fuel is postulated to be mechanically damaged and there is a complete release of gaseous isotopes. The assumption of a complete release of the gaseous inventory is also very conservative with respect to the Regulatory Guide 1.25 assumption of a 30% release fraction given the low burnup condition of Shoreham spent fuel. Doses and dose' rates are thus a factor of 617 higher than for the corresponding Regulatory Guide 1.25 cases.

G 15-10 Rev. 4 July 1992 l

SHOREHAM.DSAR

/ s All other: conditions and' parameters indicated in Table

(__) .15.1.36 apply to these_ cases. Results are given in' Table 15.1.36A-1.

Discussion Even with the highly conservative release quantity postulated above,_the calculated whole body and skin dose at the EAB and_LPZ are very small fractions (less than 0.031%)

of the 10CFR100 dose guidelines. Results are graphically shown in Figure 15.1.36A-1. Dose rates for the-postulated

. worst case scenaric are above current ODCM limits, but the duration of the high dose rates in the RBNVS and puff release cases is quite short (two hours or less).

t j

/ h

.'d 15-11 Rev. 4 July 1992

SHOREHAM DSAR

() CHAPTER 17 OUALITY ASSURANCE 17.1 QUALITY ASSURANCE DURING DESIGN AND CONSTRUCTION The description contained under this heading in the latest revision of the Shoreham USAR remains unchanged. Refer to USAR for information on this subject.

17.2 QUALITY ASSURANCE DURING THE DECOMMISSIONING PHASE The description of the Quality . ;Gurance Program during Shoreham Nuclear Power Station operational phase under this heading in the latest-revision of the Shoreham USAR is revised. However, many of the structures, systems and components designated as Quality Assurance Category I (safety related) in USAR Table 3.2.1-1 have been redesignated as Quality Assurance Category II in this DSAR.

The applicability of the USAR Section 37.2 Operational phase Quality Assurance Program as modified in this DSAR to the Quality Assurance (QA) Categories in DSAR Table 3.2-1 are as follows:

QA Category I -

The USAR Section 17.2 Quality Assurance

,_s Program as modified by DSAR Section 17.2

/ ) applies to the safety related

\J structures, systems and components which meets the intent of 10CFR50, Appendix B.

QA Category IIA -

Deleted (formerly safety related) ,

QA Category II - Appropriate measures are applied to these structures, systems, and components in (non safety accordance with QA corporate policy to related) assure that the safety significance given to them in the DSAR, Technical Specifications, and Emergency Operating Procedures are maintained.

The specific modifications of the USAR Section 17.2 applicable to the Shoreham decommissioning phase are as follows:

17.2.1 Oroanizations The Long Island Power Authority (LIPA) is responsible for the establishment and' execution of the Quality Assurance (QA) Program during the Shoreham decommissioning. LIPA has established the

-organization structure shown on Figure 13.1-1, LIPA Organization for Quality Assurance, to fulfill this responsibility. The

()S

( organization depicted in Figure 13.1-1 is subject to the QA 17-1 Rev. 4 July 1992 o _______

4 m ,.___,.--.--.m___-,e.------ 2 4 r-4.ii=.-si +,L- .+--+J-- --JAL1- W -'J,1. m- -m i

SHOREHAM DSAR 1 1

Program requirements set forth in the LIPA Quality Assurance Manual. Nuclear Quality Assurance (NQA) Department personnel' verify compliance by means of review, audit, surveillance,

-inspection, testing, or other appropriate methods.

The Executive Vice President-Shoreham Project reports directly to the LIPA President of Shoreham Project and is responsible for the overall direction, radiological and industrial safety, cost and schedule of the project. He is the corporate officer responsible for QA Program implementation and review, protection of occupational and public safety, and coordination with regulatory agencies.

He also has overall responsibility for the engineering, testing, licensing, modification, safety, reliability and maintenance, security and decommissioning of the Shoreham Nuclear Power Station and the implementation of the LIPA QA Program. He delegates the. administration of these functions to the Resident Manager who has delegated to the Department Managers (Operations and Maintenance; Decommissionir.g; Finance and Administration; Nuclear Operations Support and Licensing / Regulatory Compliance),

the responsibility to assure compliance with the QA Program requirements in their organizations.

. The Shoreham Plant Resident Manager reports directly to the j Executive Vice President - Shoreham Project, and has overall responsibility for day-to-day management of all station activities. Through his subordinates, he directs the technical, administrative and regulatory functions to accomplish all of the

-tasks and activitics comprising the LIPA project.

He also has the overall responsibility for the implementation of the LIPA QA-Program and maintenance of safety-related' structures, systems and components as defined in the DSAR Section 17.2.

The Nuclear Quality Assurance (NQA) Department Manager reports directly-to the Executive Vice President and has direct access to the LIPA President of-Shoreham Project as he deems necessary.

The NQA Department Manager is respcnsible to the Shoreham Plant Resident Manager for administration of the QA program. This

-organizational arrangement provides the necessary independence

-between personnel performing activities subject to.the controls of the QA program and those responsible for performing the checks, audits, and inspections. The NQA Department Manager is responsible for directing the activities of the Quality Control (QC) and Quality Systems (QS) Division Managers. His principal

! objective is to ensure that the Shoreham plant and all support l organizations establish and conform to adequate standards and l

procedures in accordance with the LIPA QA Manual. He has the

! authority to stop work when circumstances so warrant.

17-2 Pev. 4 July 1992

SHOREHAM DSAR n

! l-

'~ :The Manager, NQA Department, is responsible for :he development

-and imp 1ementation of the overall QA Program during the decommissioning of the Shoreham Nuclear Power Plant.

The QC Manager and QS Manager report to the Manager, NQA Department. This. organizational and functional relationship assures that the-LIPA QS personnel who audit or otherwise verify quality related activities are free from undue cost and schedull'ng_ influences and are independent of personnel who perform or are responsible for the activities.

The Manager, NQA Department, is responsible for maintaining a working interface and communication with other organizctions, regulatory agencies, consultants, contractors, inspection firms,

-and others as required to effectively execute the policies stipulated in the QA Program. He is responsible for assuring the establishment and continuous implementation of QA indoctrination and training programs for LIPA QA and other involved personnel.

The indoctrination and training will cover the quality related policies, procedures, and requirements applicable to the personnel involved. He is responsible for review and approval of applicable documents to assure the inclusion of appropriate quality requirements as indicated in Section 17.2.6. He is responcible for the performance of audits as described in Section T("N 17.2.18. In determining the applicability of the QA Program, the s

,) Hanager, NQA-Department shall consider the safety significance accorded to nonsafety related structures, systems, components, and plant computer software given to them in the Defueled Safety Analysis Report (DSAR), Technical Specifications, and Emergency Operating Procedures.

The Manager,ENQA Department, is responsible for defining the content and changes to the LIPA QA Manual subject to review and approval as' indicated in Section 17.2.6 and Appendix D of the

.LIPA-QA Manual.

The Manager, NQA Department is authorized to evaluate the manner in which all activities, both at the station and offsite, are conducted with respect to quality by means of reviews, audits, surveillance, and/or inspections. He shall perform this evaluation on a planned and periodic basis to verify that the QA Program is.being effectively implemented. He is responsible for periodically evaluating and reporting on the status and adequacy of-the QA Program to_the appropriate LIPA management.- He has the authority _and organizational' freedom to identify quality problems;-to initiate,. recommend, or providefsolutions through designated channels; and to verify implementation of solutions.

He has the authority to initiate stop work action or to control

~ further processing, delivery, or installation of nonconforming y- material through appropriate channels as described in the

( ,3E applicable QA procedure.

17-3 Rev. 4 July 1992

SHOREHAM DSAR

( ) The Manager, NQA Department is assisted in carrying out his

'~' responsibilities by the QC and QS Managers.

NQA is composed of engin! rs and technical and nontechnical personnel as needed. Additionally, the NQA staff shall be supplemented when necessary by consultants, contractors, or other organizations within LIPA. Line responsibility, coordination, and communication in such cases shall be through the QC and QS Managers.

The QC and QS Managers are jointly responsible for assuring full implementation of the LIPA QA Program, including additions and changes thereto. Each is responsible within his delegated scope of duties to establish and implement appropriate QA procedures and instructions; review applicable documents as indicated in Section 17.2.6; and perform audits, surveillances, and/or inspections as indicated in Sections 17.2.10 and 17.2.18. Each has, within his_ scope of responsibilities, the authority and organizational freedom to identify and report quality problems; to initiate, recommend, or provide solutions through designated channels; and to verify implementation of solutions. Each has the authority to initiate stop work action through appropriate channels or to control further processing, delivery, or installation of nonconforming material as described in the applicable QA procedures.

,, ~3

(_) 17.2.2 Shoreham Ouality Assurance Procram Responsibility for assuring that the Shoreham station will be decommissioned safely rests with LIPA. The LIPA Corporate Statement of Quality Assurance Policy imposes a QA Program designed to meet the requirements of Title 10 of the Code of Federal Regulations, Part 50, Appendix B, and identifies the QA Manual as the document that establishes the requirements for quality affecting activities during the decommissioning phase.

The QA Manual, which is distributed on a controlled basis to responsible managers and key supervisory and QA personnel, contains this corporate policy statement.

The QA Program is designed to assure that activities such as design, procurement, fabrication, shop inspection and testir.g, shipping, storage, construction', erection, cleaning, installation, fuel handling activities, equipment and system operation, maintenance, repair and modification of materials, structures, systems, components, services are accomplished in accordance with the criteria of 10 CFR 50, Appendix B. The QA Program is applied to the safety related structures, systems, and components as appropriate. Nonsafety related structures, systems, components, and services shall be accorded, as a minimum, the safety significance given to them in the DSAR, the Defueled Technical Specifications, and Emergency Ope ating (f- ) Procedures. This practice will assure # } it the safety 17-4 Rev. 4 July 1992

S!!OREllAM DSAR i

() significance accorded to nonsafety related structures, systems, and components is maintained during the decommissioning of Shoreham. Also, the Shoreham preventive and corrective maintenance program, the design change control program, procedures for procurement of equipment, and procedures for modification and removal of equipment from service saall ensure that LIPA cont) ues to accord to nonsafety related structures, systems, and components the safety significance given to them in the DSAR, Technical Specifications, and Emergency Operating l Procedures.

Thus, the responsible personnel implementing these programs and procedures shall, in exercising their 4udgment on the appropriate measures to be tpplied to nonsafety re4ated structures, systems, and components, do so in accordance with the corporate QA polic,.

Tha QA Program, described in the LIPA QA Manual, is supplemented by QA Procedures and Instructions, which provide the detailed ,

instructions and checklists necessary to implement or verify implementation of QA Program requirements. These procedures are delineated in Section 17.2.5. QA procedures are issued, reviewed and approved as shown in Table 17.2.6-1. The QA Manual, Frocedures, and Instructions shall be controlled in accordance with the requirements of Section 17.2.6.

O The QA Program requires that activities affecting quality shall be ac~omplished in accordance with documented policies, prot res, and instructions throughout the decommissioning of Shor n. These activities shall be accomplished under suitably conti led conditions. Controlled conditions include, as applicuble, appropriate equipment, suitable environmental conditions, and assu'...co that required prerequisites have been satisfied. Also conw.uered shall be the need for special controls, processes, and requirements for verification of quality by inspections, examinations, or tests.

The QA Procedures for decommissioning are derived from the program requirements established in the QA Manual. Organizations described in Section 17.2.1, performing activities that affect qualf11, shall prepare their procedures incorporating requirements of-the QA Manual and referenced codes, standards, and guides. These procedures shall also receive a QA review to assure that all program requirements have been addressed.

The corporate Statement of QA Policy, contained in the LIpA QA Manual, imposes the mandatory QA Program requirements-on all personnel and organizations performing activities affecting the quality of safety related-structures, systems, and components during the decommissioning of Shoreham. The Manager, NQA

. Department is responsible for periodically engaging an organization independent of the organization being reviewed to

-q assess LIPA quality related activities and to evaluate the scope, 17-5 Rev. 4 July 1992

SHOREHAM DSAR

(~)h

(_ implementation, and effectiveness of the QA Program. This periodic review assures that the program is adequate and complies with corporate QA policies, goals, objectives, and 10 CFR 50, Appendix B er'.teria. The requirement for independent nA Program evaluation is further imposed, as appropriate, on other organizations participating in the LIPA QA Program. The LIPA QA auditing program is described in Section 17.2.18.

The Manager, NQA Department is responsible for establishing and implementing the QA Program. Provisions have been established for the referral of quality related problems to the highest level of management necessary for resolution. The Manager, NQA Department, is responsibic for regularly asscosing the-status and adequacy of the QA Program. He shall, through written reports and or periodic meetings, inform the Executive Vice President -

Shoreham Project and Shoreham Plant Resident Manager, of the effectiveness of the QA Program and of significant quality trends.

These regular assessments shall be conducted in accordance with the requirements outlined in Section 17.2.18 and as detailed in '

Section 18 of-the_QA Manual. The requirement for regular QA Program evaluation shall be extended to other participating organizations for the pc.: ions of the prograr they are executing.

/~h The QA Program requires that procedures be established for the k-) indoctrination, training and, if appropriate, certification of personnel performing or verifying safety related activities.

These procedures shall document the scope, objective, and method of implementing the indoctrination and training program and contain provisions for documenting training sessions including content, date, attendance and results.

LIPA and/or supplier organizations shall provide for the initial qualification and refresher training of personnel to assure that they achieve and naintain proficiency to satisfactorily parform their safety-related functions. Training and qualification records shall be maintained.

Programs shall be established.for indoctrination, training and, if appropriate, certification of personnel performing or verifying safety-related activities. The NQA Department Manager shall provide for QA indoctrination and specific training of NQA personnel, and for assurance of the satisfactory QA Lindoctrination of other personnel engaged in safety-related activities.

Formal QA training shall be accomplished in accordance with written procedures. -These procedures shall describe the scope, objective and method of implementing the indoctrination and training program'and contain provisions for documenting training O

sessions. These training documents are to include content, date, 17-6 Rev. 4 July 1992

- .. . - . . ~ _ , - . - - , _ - ,.

l SHOREHAM DSAR

~x

_) attendance and results. Personnel proficiency shall be maintained as necessary by means of refresher courses, reexamination and/or recertification.

This QA Program is designed to comply wJth the requirements of 10 CFR 50 Appendix B, 10 CFR 50.55a, other applicable Federal regulations, applicable NRC Regulatory Guides, and ANSI and ANS Standards as committed to in the Shoreham Defueled Safety Analysis Report (DSAR). The requirements stated in the QA Manual are applicable to materials, structures, systems, components and services whose satisfactory performance is required for safe storage and handling of nuclear fuel.

Outsido contractors that perform safety related functions shall be required to comply with those portions of 10 CFR 50, Appendix B, and the LIPA Program that are applicable to the services provided. LIPA QA Procedures shall require that a review and evaluation report of a supplier's QA Program be available and accepted by LIPA NQA Departtixt prior to the issuance of purchase orders for safety items, or services to assure that the program meets the applicable elements of 10 CFR 50, Appendix B.

Compliance with QA Program requirements by both internal and external organizations shall be assured by a comprehensive system of audits and reviews performed by NQA under the direction of the 3

1

Manager, NQA Department. Significant changes to the DF' R that

\' may occur between general review cycles shall be transL.tted to organizations as defined in the applicable administrative procedures.

17.2.3 Desian Control The LIPA QA Program establishes measures to control design activities that affect the quality of safety related structurcs, systems, and components during the decommissioning phase. These measures are applicable to all organizations performing design, design review, or design audit activities including changes or modifications therete Section 3 of the LIPA QA Manual describes the QA Program requirementa established to provide this control.

The program requires that design and modification activities be accomplished in a planned, controlled, orderly manner in accordance with established procedures. Design control measures "

shall assure the translation of applicable design bases, regulatory requirements, codes, and standards (which includes the selection of suitable materials, parts, equipment, and processes) into specifications, drawings, and documented procedures and instructions. The program requires that the quality requirements be included in the design doc'2ments.

'Q 3 17-7 Rev. 4 July 1992

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - - _ _ _ _ _ .___ _ _. _____________-_____a

l SHOREHAM DSAR b

Deviations from or changes to specified quality requirements in design documor*9 shall be controlled. Suitable design control measures are 6:43. red for design analysis such as physics, stress, ther- voismic, hydraulic, radiation, and accident analysis; cce < _ollity of materials; accessibility for maintenance, repair and rework and acceptance criteria for inspections and testo. Design control procedures shall identify and control design interfaces both internal and external to LIPA.

Design verification, such as design reviews, alternative calculations, or qualification testing, shall be properly selected and accomplished. Responsibility for such verification is described later in this section. Where qualification testing of a prototype is used to verify adequacy of design, testing shall be performed under the most adverse design conditions. The program requires that design verification be performed by individuals or groups ot her than the original designer and the designer's immediate supervisor, but verification may be performed by individuals from the same organization.

Design changes shall be subject to design control measures commensurate with those applied to the original design. Design control measures shall provide for the suitable review and 7'~N selection of standard "off the shelf" commercial or previously

(_,) approved material, parts, equipment, and processes that are essential to safety related structures, systems, and components.

Design documents and revisions thereto shall be distributed to the responsible individuals in a timely and contr'lled manner to prevent inadvertent use of superseded documents. ontrol of design documents is further described in Section 17.2.6. Design documents and reviews, records, and changes thereto are collected, stored, and maintained in accordance with Section 17.2.17. Errors or deficiencies that may arise during the design process shall be addressed in accordance with Sections 17.2.15 and 17.2.16.

Organizations supplying equipment and/or services are responsible for imposing the applicable requirements of this section on their internal operations and on those vendors and contractors performing work within the scope of their activity as required by the procurement documents. These organizations are responsible for assuring by means of audit or surveillance that design control as defined in their respective programs is oeing effectively implemented. LIPA is responsible for assuring program adequacy and implementation by external suppliers through planned and perioaic audits, e

17-8 Rev. 4 July 1992 i l

SHOREHAM DSAR O The design change control programs also include provisions to ensure that nonsafety related structures, systems, components, and plant computer software shall continue to be accorded the safety significance given to them in the DSAR, Defueled Technical Specifications, and Emergency Operating Procedures.

The OMD and DED Managers are responsible for determining, initially, whether proposed modifications or repairs involve unroviewed safety questions or changes in technical specifications as described in 10 CFR 50.59. This determination shall be reviewed by the Site ReviPW Committee (SRC) and forwarded to the Resident Manager for approval. Procedures shall provide documentation and control of such determinations.

Technical evaluation, including design verification, shall be the responsibility of the appropriate organization. The LIPA HQA Department is responsible for verifying overall program establishment and implementation through planned and periodic audits.

17.2.4 Procurement Document Control The LIPA QA Program provides for the control of procurement documents for safety related material, equipment, and services f~T whether purchased by LIPA or suppliers, during decommissioning.

s_/ Section 4 of the LIPA QA Manual describes the QA Program requirements established to assure procurement document control.

The program requires that procedures establish measures to assure control of the preparation, review, approval, and concurrence of procurement documents. Document control procedures as described in Section 6 and as delineated in Table 17.2.6-1 shall be applied to procurement documents including changes and revisions. . The procurement' documents shall be reviewed by qualified personnel, as defined within this section, assuring the adequacy of the quality requirements. The review shall be utilized to assure that the quality requirements, including preparation, review, and approval, have been properly defined, that the procured items are inspectable and controllable, and that the acceptance criteria are adequately specified. ,

The program requires that procurement documents such as purchase specifications contain or reference the design bases technical requirements, which include codes, industry standards, and regulatory requirements; material and component identification requirements; drawings and/or specifications; test and inspection requirements;.and special process instructions. In addition, procurement documents are required to identify the following:

17-9 Rev. 4 July 1992

_ .- ._. ._ - -_._- _- m _ _ _ _ _ _ _ __

SHOREHAM DSAR

(

1. Requirements of 10 CFR 50, Appendix B, with which the supplier QA Program must comply.

2. The document requirements for drawings; specifications; procedures; personnel and procedure qualifications; material, chemical, and physical test results; and inspection and test records that must be prepared, maintained, submitted, or made available for review and/or approval.

3. The requirements for the retention, control maintenance, and/or_ delivery of records.

4. LIPA's right of access to supplier's facilities and records for source inspection and audits.

Procurement documents for spare or replacement parts shall be subject to program requirements equivalent to those used for the original equipment or to those specified by a properly reviewed and approved revision.

Tho'LIPA Finance and Administration Department is responsible for the commercial aspects-associated with procuring items or services, which includes the processing of purchase orders.

0

, LIPA's organizations are responsible for assuring that the procurement documents contain technical and quality requirements as indicated above. Authorized release, assuring acceptability of'both technical and quality content, is required prior to releasing a purchase order.

LIPA's Shoreham organizations shall prepare those procurement documents pertaining to their scopes of responsibilities and shall present those documents to the Finance and Administration Department for processing. The NQA Department is responsible for reviewing the procurement documents for quality requirements, and for the review of and concurrence with selected suppliers' QA Programs.

Consultants, architect-engineers, testing companies, etc.

(collectively " contractors"), assigned responsibility by LIPA for procurement activities associated with safety related material, equipment, or services _shall impose the control requirements indicated above.- LIPA's contractors shall establish the p requirements in procedures,-instructions, drawings, etc.- These requirements shall be imposed cn1 the contractors' internal l operations and on any vendors or contractors performing work l 'within the scope of their activities as required by the

, procurement documents.- The contractors shall assure the adequacy of program implementation through audit or surveillance. LIPA

/~3 shall verify _ program adequacy and implementation by suppljers

. (,/ through planned and periodic audits consistent with the complexity,11mportance, and-quality of items or services.

17-10 Rev. 4 July 1992

SHOREHAM DSAR l ) Personnel exercising their judgment with regard to procurement of

\J- nonsafety related structures, systems, components, and plant computer software shall assure that the safety significance accorded to them in the DSAR, Technical Specifications, and the Emergency Operating Procedures is maintained throughout the decommissioning of Shoreham.

17.2.5 Instructions. Procedyres, and Drawinog The LIPA QA Program establishes provisions ioc activities affecting the quality of safety related structures, systems, and components during decommissioning to be accomplished and controlled in accordance with instructions, procedures, and drawings. Section 5 of the LIPA QA Manual describes the QA Program requirements for the control of instructions, procedures, and drawings. Organizational procedures delineate the sequence of actions to be accomplished in the preparation, review, approval, and control of instructions, procedures, ar;d drawings.

Suppliers, vendors, and contractors have the responsibility for establishing instructions, procedures, drawings, and other documents to control the quality related activities of their own operations and those of their subsuppliers, as required by the procurement documents. A description of the associated procurement document control requirements is in Section 17.2.4.

p

(~j LIPA organizations are responsible for establishing instructions, procedures, and drawings or for utilizing established procedures, instructions, and other documents to control the quality relat :

activities they perform. The required station procedures are described in Section 13.5 of the DSAR. All responsible organizations establish provisions such that the development and implementation of instructions, procedures, and drawings, including changes thereto, are clearly identified and controlled.

The LIPA NQA Department is responsible for performing review, surveillance, and audit functions to verify that the instructions, procedures, drawings, and other documents used for safety related structures, systems, and components are controlled to meet the requirements of 10 CFR 50, Appendix B.

Activities affecting the quality of safety related structures, systems, and components are defined in specifications, instructions, procedures, drawings, and other documents. These documents include qualitative and quantitative acceptance criteria for the activity being conducted. These criteria are used to control quality affecting activities; and to define special process controls, codes, standards, and regulatory requirements.

O 17-11 Rev. 4 July 1992 l

SHOREHAM DSAR f'~))

\

The LIPA NQA Department reviews all safety related test, calibration, special process, maintenance, modification and repair procedures, drawings and specifications, and changes thereto, with respect to quality requirements as indicated in Section 6 and delineated in Table 17.2.6-1.

17.2.6 Document Control The LIPA QA Program provides for the control of documents, including changes thereto, which affect the quality of safety related structures, systems, and components during decommissioning. The applicable documents include, but are not limited to, the QA Manual; QA Procedures and Instructions; the Defueled Safety Analysis Report; design drawings; component specifications; procurement documents; supplier technical manuals, procedures and instr u itions. Section 6 of the LIPA QA Manual describes the QA Progre: requiremonts established to assure document control.

The program requires that a document contro2 eystem be established in accordance with approved procedures and instructions for review, approval, and issuance of the documents, including changes thereto, to assure that they are adequate and incorporate the quality requirements prior to release. LIPA

(~N organizations that issue, review, and approve documents shall T ,)

s establish provisions for the identification of individuals or groups responsible for performing review, approval, issuance, or revision activities.

The program requiren that changes to documents be reviewed and approved by the organization responsible for conducting the original review and approval or, as deemed necessary by LIPA, such changes will be reviewed and approved by another qualified and responsible organization. In the event that another qualified organization is charged with the responsibility for revision, that organization shall have access to pertinent background information for adequate understanding of the requirements and intent of the original document. Procedures and instructions provide measures to assure the prompt distribution of approved changes and revisions, including control of obsolete or superseded documents to prevent their inadvertent use. The program requires that the documents be available at the location where the activity will be performed prior to the start of work.

Change or revision identification will be established and verified through the utilization of document distribution lists.

Updating and distribution to personnel of such lists will be consistent with the nature of the document.

Suppliers of safety related items and services are responsible gs for imposing the above document control requirements on their

() internal operations and on those vendors and contractors 17-12 Rev. 4 July 1992 I l

i "u - - - ' - -u "- '

.. _ . . . . . .._.misuiud

SHOREHAM DSAR performing work within the scope of their activities as required by the procurement documents. Suppliers shall assure program adequacy and implementation through planned and periodic audits.

LIPA is responsible for assuring program adequacy and implementation by external suppliers through planned and periodic audits.

LIPA organizations that issue, review, and approve documents, including changes thereto, are responsible for establishing and implementing a document control system in accordance with the requirements indicated above. The LIPA NQA Department is responsible for assuring overall program adequacy and implementation through planned and periodic audits.

17.2.7 Control of Purchased Material, Eauinment, and Services The LIPA QA Program establishes measures to assure that safety related material, equipment, and services procured during decommissioning either directly or through contractors, conform to the procurement document requirements. Section 7 of the LIPA QA Manual describes the QA Program requirements established to provide this control.

The program establishes provisions for source evaluation and selection. Source evaluation and selection may be based upon historical quality performance data, source surveys or audits, or

\- source qualification programs. This evaluation and selection process-will determine the supplier's capability to supply the item or service in compliance with the design, manufacturing, and quality requirements as stipulated in the procurement documents.

Measures are' established to provide for both a technical and quality evaluation of those suppliers providing safety related components or services._ LIPA's Shoreham organizations shall perform the technical evaluation, and-the HQA Department shall perform the quality evaluation. These functions may also be accomplished through the utilization of-qualiffed independent organizations. Personnel performing the evaluations, such as

. auditors, shall be qualified. Source evaluation and selection

, information shall be documented and filed.

The program provides for source, inspection, surveillance, and audit of suppliers to assure conformance to procurement document requirements. The inspections, surveillance, and audits shall be conducted in accordance with documented procedures. Source inspection procedures provide for instructions to be established for specifying the characteristics to be witnessed, inspected or verified, and accepted; for indicating responsibility; and for determining-documentation requirements.

Source audits or surveillance shall be conducted, as necessary, O to assure compliance with quality requirements. Source inspection or. audit may not be necessary when the quality of the 17-13 Rev. 4 July 1992

, , . . - - - - - - . - - . ~ . . - ,- . -

SHOREHAM DSAR f3 item can be verified by review of test reports, inspection upon receipt, or other means.

The program requires that receiving inspection be accomplished in accordance with documented procedures and instructions. The receiving inspection procedures and instructions establish measures to assure that the item is properly identified and corresponds to the receiving documentation, that the item and acceptance records are determined to be acceptable in accordance with the inspection instructions prior to use, that the receiving ~

i documentation is available at the plant prior to use, and that j the inspection status is identified as indicated in Section j 17.2.14. The QA Program specifies that procurement documents <

require suppliers to furnish documentation identifying any )

procurement requirements that have not been met together with a description of those nonconformances marked " accept as is" or

" repair". Responsible NQA and technical personnel shall perform a review and approval of the supplier's recommended disposition.

Nonconforming items shall be identified and controlled as indicated in Sec' ion 17.2.15. Inspections shall be conducted based upon the nature of the item being procured.

When required by code, regulation, or contract requirements, documentary evidence that items conform to procurement

(~ requirements shall be available and readily retrievable at the

(_N.) plant. -This documentary evidence shall specifically identify the item-and codes and/or specifications met by the item. When not precluded by other requirements, such documentation may take the form of written certification of conformance identifying the-requirements met by the items. LIPA QA Procedures require that suppliers' certificates _of_conformance be periodically evaluated by audits or tests to-assure that they are valid.

Suppliers of safety related material,. equipment, and services are responsible for imposing the control requirements indicated above on their internal operations and on any vendors or contractors performing work within the scope of their activities as required by the procurement documents, suppliers shall assure through audit or surveillance the adequacy of program implementation.

The'LIPA Finance and Administration Department is responsible for commercial aspects associated with procuring items or services.

The-LIPA organizations that requisition items and/or services and the NQA Department are responsible for assuring that the procurement documents contain the information as required above.

Procedures have been established to control spare and replacement part procurement documents,.through technical and NQA review, to ensure that the controls for safety related items are equal to or better than the original equipment. The QA Program requires that a-technical evaluation and NQA review be performed to determine fs which requirements are to be applied to the procurement of spare -

ard replacement parts when the original equipment requirements

('-

L are not known.

17-14 Rev. 4 July 1992

l SHOREHAM DSAR

() Procurement document control is described in Section 17.2.4.

LIPA shall assure program adequacy and implementation of suppliers through planned and periodic audits consistent with the complexity, importance, and quality of the item or service. The LIPA NQA Department is responsible for evaluating suppliers.

This evaluation shall include utilization of qualified independent organization surveys. Source inspection, as necessary, shall be conducted by LIPA or qualified independent organization. The LILCO Project Organization is responsible for receipt of items at the station.

The NQA Department, which is responsible for conducting receiving inspections of items with respect to quality requirements, assures overall program establishment and implementation through planned and periodic audits and surveillances.

17.2.8 Identification and Control of Materials. Parts. and Components The LIPA QA Program requires the establishment of an identification and control system to prevent the use of defective, unapproved, or incorrect safety related material, parto, and components. Section 8 of the LIPA QA Manual describes the QA Program requirements established for this purpose.

including O The program requires that the identification system, unique part or mark numbers developed construction phases, be maintained and expanded as necessary during decommissioning. A system for identification and control of materials, parts, and components (including partially fabricated subassemblies) shall be based on documented procedures and/or instructions.- Identification is referenced in specifications, drawings, purchace orders, or other appropriate documents providing traceability to associated documentation such as manufacturing and inspection documents, deviation reports, heat numbers, and mill test reports. The identification may be placed either on the item or on records directly and readily traceab1n to the item. Physical identification shall be used to the maximum extent possible and shall be applied in such a manner as not to affect the function of the item. Verification of identification shall be accomplished at appropriate stages throughout fabrication, assembly, shipping and receiving, and prior to installation.

During decommissioning, suppliers of safety related material, parts, and components are responsible for establishing a system

.of identification and control that addresses the requirements outlined above. Suppliers are responsible for imposing the requirements on their internal operations and on those vendors and contractors performing work within the scope of their

~

activities-as-stipulated in the procurement documents. Suppliers

.shall assure, through audit or surveillance, the adequacy of 17-15 Rev. 4 July 1992

SHOREHAM DSAR

(~ program implementation. LIPA shall assure program adequacy and I

( implementation through planned and periodic audits of the suppliers.

I The Material Management Division (MMD) is responsible for maintaining and expanding the identification and control system for safety related material, parts, and components. If a design change is necessary, the MMD is responsible for supplying identification requirements to the associated organizations and for assuring the continued implementation of the established identification and control system. The NQA Department is responsible for assuring overall program establishment and implementation through planned and periodic audits, surveillance, and inspections at the station.

17.2.9 Control of Special Processes The LIPA QA Program imposes on organizations performing special processes the requirement to develop a system of special process controls. Special processes include, but are not limited to, special inspection or test processes, welding, heat treating, nondestructive examination (NDE) , decontamination and radiological and chemical analyses. Section 9 of the LIPA QA Manual describes the QA Program requirements established for control of special processes.

() The program requires that organizations performing special processes on safety related equipment at the nuclear power station or at an offsite facility do so using approved procedures, instructions, or the equivalent, and that equipment and personnel be qualified in accordance with applicable codes, standards, specifications, or special requirements. Special process. procedures, in addition to providing for the qualification of equipment and personnel, shall provide for the documentation of' accomplished activities. Where special processes are not covered by existing codes or standards, or where certain item quality' requirements exceed the requirements of established codes or-standards, the necessary qualification of personnel, equipment, or procedures shall be required. Special process procedures and qualification records shall be filed, maintained, and available for verification.

Suppliers of equipment and services whose scope of activity

-includes utilization and control of special processes are

. responsible for imposing these requirements on their internal operations and on those suppliers, vendors, or contractors performing work within the scope of their activity as required by the procurement documents. Special process controls shall be submitted-to the suppliers for approval as specified in the procurement documents. . Suppliers shall verify through audit or surveillance the' adequacy of program implementation.

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17-16 Rev. 4 July 1992 u

SHOREHAM DSAR LIPA shall verify overall program adequacy and impicmentation by f'))

x_ internal organizations and suppliers through planned and periodic audits.

17.2.10 Jnsocction The LIPA QA Program provides for inspection of activities that affect the quality of safety related structures, systems, and components during decommissioning. Section 10 of the LIPA QA Manual describes the QA Program requirements established for inspection.

It provides for an inspection program to be implemented in accordance with applicable procedures, instructions, and checklists. Inspections shall be performed by individuals other than those who performed or directly supervised the activity being inspected. Inspection procedures, instructions, or checklists contain identification of responsibility for performance of the inspection, method of inspection, characteristics to be inspected, acceptance / rejection criteria, verification, evaluation, and documentation of the results of the 11:spection. The program requires that inspection procedures or instructions be made available for use, with supporting documents such as drawings and specifications, prior to the performance of inspection operations. Information concerning inspections shall ex be obtained from design specifications, drawings, and/or other controlled documents including codes, standards, and regulatory

('~') requirements. The inspections are conducted by inspectors who have been qualified and certified in accordance with codes, standards, and/or LIPA training programs. The inspection program requires that inspector qualifications be kept current. The respective managers shall be responsible for certifying their inspection personnel.

When notification or hold points are established in procurement or other documents, the inspection program requires that: ,

1. Work does not progress beyond the hold point until released by the designated authority.

2. The notification and acknowledgement has been satisfied prior to continuation of work.

Inspection of rework, repair, replacement, or modification activities shall be conducted in accordance with inspection requirements or by means of an approved alternative. Such alternatives shall be evaluated on both a technical and quality basis. When direct inspection is not possible, provisions are established for indtreet control by monitoring of processing methods, equipment and personnel.

/9 V

17-17 Rev. 4 July 1992

l SHOREHAM DSAR Suppliers of safety related material are responsible for imposing the above requirements on their internal operations and on those vendors or contractors performing work within the scope of these activities as required by the procurement documents. Suppliers shall assure, through audit or surveillance the adequacy of program implementation.

LIPA, through planned and periodic audits, surveillance, and participation in celected inspections shall verify conformance of inspection programs delegated to external organizations. When inspections or other safety related activities are conducted by LIPA or an outside contractor at the station, the NQA Department is responsible for verifying that the inspection program complies with the requirements outlined above. The LIPA NQA Department is responsible for reviewing maintenance and modification procedures to assure that requirements such as the need for inspection, identification of personnel, and documentation of results have been addressed.

17.2.11 Test Control The LIPA QA Program establishes provisions to assure that testing required to demonstrate satisfactory inservice performance of safety related structures, systems, and components is conducted

(~h in accordanca with an approved, documented test program. Section

\s / 11 of the-LIPA QA Manual describes the QA Program requirements established for test control during the decommissioning phase.

It is required that the test program be identified, documented, and accomplished in'accordance with procedures that are written, approved, and controlled. The basis for determining when proof, preoperational, and operational tests are required to demonstrate satisfactory inservice performance are addressed-in Section 17.2.14 and in the LIPA QA Manual. The QA Program has established that modifications, repairs, and replacements shall be tested in accordance with the original design and testing requirements or acceptable alternatives. Technical and NQA reviews provide assurance that the testing does accomplish this end. The test procedures contain or reference the requirements

and acceptance limits from the applicable design or procurement documents. The procedures establish provisions to assure that prerequisites for a given test have been met. Prerequisites include

Test equipment is adequate and in satisfactory operating condition; test _ instrumentation has been properly calibrated; personnel are trained, qualified, and certified if

-necessary for the various test functions; preparation, condition, and completeness of the. item to be used have been satisfactorily accomplished; suitable environ ~*ntal conditions are available; provisions for data acquisit$3 . ave been established; if necessary, mandatory inspection hold points for witness by the O designated authority are included; appropriate acceptance and/or 17-18 Rev. 4 July 1992 L -_

S!!OREHAM DSAR rejection criteria are established; and methods for documencina data and results are established. The program requires that test results be documented in sufficient detail to prevent misinterpretation, that they be evaluated to the established criteria, and that the acceptance status be identified by a qualified, responsible individual or group. Test records shall be appropriately filed upon completion of the test and evaluation.

Suppliers of safety related material and services are responsible for imposing the above requirements on their internal operations and on those vendors and contractors performing work within the scope of their activities as stipulated in the procurement documents. Suppliers shall assure, through audit or surveillance, the adequacy'of program implementation. LIPA shall verify program adequacy and implementation by external suppliers through planned and periodic audits. '

Responsibility for the station testing programs has been assigned to the Operations and Maintenance Departi it during Shoreham's decommissioning. The LIPA HQA Department is responsible for verifying overall program establishment and implementation through planned and periodic audits and surveillances.

e 17.2.12 Control of Me q_urina and Test Eauinment The LIPA QA Program imposes requirements for control of measuring and test equipment on organizations whose activities affect the quality of safety related structures, systems, and components.

The program requires calibration control for the measuring and test instruments, tools, gauges, fixtures, reference and transfer standards, and nondestructive test equipment. Section 12 of the LIPA QA Manual defines the QA Program requirements established forfcontrol of measuring and test equipment.

The program requires that calibration procedures describe the technique, frequency, and maintenance for measuring and test equipment.' The QA Program requires procedures to establish methods for identification of. measuring and test equipment and associated calibration data including provisions to assure that

! documented control system to in@icate the date of the next calibration. The frequency of calibration is established for measuring and test equipment on an individual basis or generic grouping thereof. It is based-upon the type of equipment, required accuracy, stability characteristics, purpose,-degree of usage, experience, manufacturers' recommendations, and recognized industry standards. The reference and transfer standards are traceable to nationally recognized standards and, for any 7xceptions, provisions are established to document the basis for y calibration. The calibration program requires that, in the event j (' an instrument is found to be out of calibration, an investigation i

shall be conducted and documented to determine the validity of l . previous measurements. It is required that calibration records l

l 17-19 Rev 4 July 1992

SHOREHAM DSAR

,m k_) be established and maintained to provide objective evidence that measuring and test equipment is being controlled, calibrated, and maintained in accordance with approved procedures.

Provisions assure that calibrating standards have an accuracy, range, and stability adequate to verify that the equipment being colibrated is within specified tolerance and can meet all other specified requirements.

The reference standard used as the working (shop) standard shall have a tolerance not greater than one-fourth the specified tolerance of the measuring and test equipment being calibrated, except when equipment acceptable for nuclear power plants applications is not commercially available. In those cases, instruments of equal or greater accuracy shall be used. The reference standards used to calibrate the working (shop) standards shall have an accuracy greater than that of the working (shop) standard. When reference standards used to calibrate the working (sPop) standard have an accuracy equal to that of the working (shop) standard, the basis for the use of standards having the same accuracy shall be documented by responsible management.

Procedures shall be written to control and monitor the use of eS measuring and test equipment and reference standards to assure l i that the above requirements are maintained within the limitations N/ noted. These procedures also assure that permanently installed operating instrumentation is calibrated against measuring and test equipment having a tolerance not greater than the specified tolerance of the installed instrumentation.

During decommissioning, suppliers of equipment and services whose scope of activity includes the utilization of measuring and test equipment on safety related structures, systems, and components are responsible for imposing the above control requirements on their internal operations and on those vendors and contractors performing work within the scope of their activities as required by the procurement documents. Suppliers shall assure, through audit and surveillance, the adequacy of program implementation.

LIPA shall verify program adequacy and implementation through planned and periodic audits of suppliers.

LIPA station organizations such as Radiochemistry, Health Physics and Maintenance are responsible for maintaining control over the M&TE they utilize and for complying with the applicable requirements of this section.

The LIPA NQA Department is responsible for verifying program establishment und implementation through planned and periodic audits and surveillance.

(~)

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17-20 Rev. 4 July 1992 i

SHOREHAM DSAR 17.2.13 Handlina. Storace, and Shippina The LIPA QA Program imposes control requirements on organizations whose scope of activity includes the handling, storage, and shipment of safety related structures, systems, and components during Shoreham's decommissioning. Section 13 of the LIPA QA Manual describes the QA Program requirements established for handling, storage, and shipment. Certain requirements are applied as necessary to non safety-related materials, equipment and services.

The program requires that organizations performing han 111ng, storage, and shipping activities (including cleaning, packaging, and preservation) do so using written procedures or instructions.

These procedures shall be developed in accordance with applicable design and specification requirements and shall provide for control of the aforementioned activities to preclude damage, lost, er deterioration of safety related materials, components, and equipsant. Special environmental conditions (such as special coverings, inert gas atmosphere, allowable moisture content, and temperature level) shall be detailed, and their existence shall be verified and documented. Provisions for necessary cleaning operations, as required by the nature of the material or equipment, shall be included and their verification documented.

O Special handling requirements shall be provided and controlled to ensure safe and adequate handling, including associated verification and documentation. The procedures or instructions provide for inspection operations to verify conformance to established critoria, use of qualified personnel, and associated documentation. In addition, the procedures and instructions shall provide for the controlled release of safety related material, components, or equipment from storage for shipment or -

installation and for the verification and documentation thereof.

The progcam requiremento are applicable to the stages of fabrication, manufacturing, and installation associated with decommissioning. Suppliers are responsible for imposing the requirements, as specified in the procurement documents, on their internal' operations and on those vendors and contractors performing work within the scope of their activities. Suppliers alrs assure the adequacy of program implementation.

The LIPA NQA Department shall verify overall program adequacy and implementation by internal crqanizations and by suppliers through planned and periodic audits.

17.2.14 Inspection. Test, and Operatina Status The.LIPA QA Program provides measures for indicating the inspection, test, and operatinrj status of safety related O structures, systems, and co.?ponents. Section 14 of the LIPA QA 17-21 Rev. 4 July 1992 1

SHOREHAM DSAR

() Manual describes the QA Program requirements r identification and control of inspection, test, and operating status, i

The QA Program requires that organizations responsible for fabrication, storage, installation, testing, and operation of safety related components and systems identify and control the inspection, test, and operating status of these items. The status is identified and controlled through the utilization of l status indicators (such as tags, markings, logs, shop travelers, l stamps, inspection, or test records).

In addition, the QA Program requires the establishment of measures to control the use of the status indicators, inclu' ling responsibility and authority for their application and removal and the unique identification of the individual involved.

Associated procedures establish provisions to assure the performance of required tests and inspections including requirements that the identification of the status be known at any given time. The bypassing of required inspections, tests, and_other critical operations is controlled through station administrative procedures. These administrative procedures shall be reviewed by the NQA Department. Procedures establish measures to indicate the operating status to prevent inadvertent operation of safety related systems, equipment, and components. They establish provisions so that the identification of operating status is known at any given time.

(U~%

The programs assure that functions performed out of sequence are adequately documented and do not compromise system integrity.

Procedures provide for the positive identification and control of nonconforming items in accordance with Section 17.2.15, to prevent their inadvertent use.

Tne program requirements are applicable to stages of fabrication, installation,. testing, and operation associated with the decommissioning. Suppliers are responsible for imposing the requirements, as specified in the procurement documents, on their internal operations and on those vendors and contractors

performing work within the scope of their activities. Suppliers also assure through audit or surveillance the adequacy of program implementation. ,

The LIPA NQA Department shall verify overall program adequacy and implementation by internal organizations as well as by suppliers through planned and periodic audits.

17.2.15 Fonconformina Materials. Parts, or Comnonents The LIPA QA Program imposes requirements for control of nonconforming safety related material, parts, and components.

~ These requirements are applicable to organizations whose (3j- activities affect the quality of such safety related items during 17-22 Rev. 4 July 1992

l SHOREHAM DSAR

(,)

'/

the decommissioning phase. Section 15 of the LIPA QA Manual

- describes the QA Program requirements established to assure control on nonconforming items to prevent their inadvertent use or installation.

The QA Program requires that a control system be established to address nonconformances in accordance with documented, approved procedures. The procedures establish measures to assure that nonconforming items and services are properly identified, documented, reviewed, segregated if practical, dispositioned, and reported to affected organizations.

In addition, the procedures establish provisions for designation of responsibility and authority for approval of the dispositioning of nonconforming items. The program requires that nonconforming items be documented and that such documentation include a clear identification of the nonconformance, a description of the nonconformance, the appropriate disposition including the approval signature, and the applicable inspection and test requirements. Nonconforming items shall be clearly identified as such and placed in a controlled segregated area, when practical, until proper disposition has been effected.

Nonconforming items may be dispositioned by accepting "as is,"

scrapping, repairing, or reworking. The acceptability of r^T repaired or reworked nonconforming items is verified by

(_) reinspection. The reinspection of the item shall be in accordance with the original inspection requirements or by acceptable alternatives. The program requires that the appropriate repair, rework, and inspection procedures be documented. Nonconformance reports verifying the " accept as is" or " repair" disposition shall be made part of the required inspection records.

Suppliers of safety related materials, parts, and components are responsible for imposing the above requirements on their internal operations and on those vendors and contractors performing work within the scope of their activities as required by the procurement documents. They also assure, through audit or surveillance, program adequacy and implementation. LIPA is responsible for conducting audits to verify program adequacy and implementation by suppliers. The LIPA NQA Department is responsible for assessing the adequacy and implementation of suppliers' nonconformance control systems. This assessment is in addition to technical reviews of applicable nonconformance reports by other LIPA organizations. Safety related nonconformance reports shall be analyzed periodically to determine the existence of quality trends. Trends, if any, shall be reported to the appropriated LIPA management.

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17-23 Rev. 4 July 1992

SHOREHAM DSAR O When a LIPA organization discovers a nonconformance related to a LIPA activity, it is that organization's responsibility to generate and control a nonconformance report in accordance with  !

the requirements stated herein. In general, the organization responsible for the nonconforming condition is responsible to provide an acceptable disposition. The reporting organization and the NQA Department are required to review and accept the disposition before it may be implemented.

17.2.16 Corrective Action The LIPA QA Program provides measures to assure that conditions adverse to quality are promptly identified, reported, and corrected. Section 16 of the LIPA QA Manual describes the QA Program requirements for corrective action and control thereof.

The program provides for a corrective action system implemented through the use of approved written procedures. The procedures provide for identification and documentation of deficiencies, including nonconformance reports, and determination of the need for corrective action. The procedures provide for reporting significant conditions adverse to quality, assessment of their probable root causes, and that the preventivo and corrective g- actions taken be documented and reported to appropriate levels of

- (j management for review and assessment. Follow-up action shall be taken to assure proper implementation and timely closcout of corrective action.

Suppliers are responsible for establishing and implementing a corrective action program commensurate with the function they perform. The supplier systems provide measures that comply with the requirements outlined above and are imposed on internal operations as well as on vendors and contractors performing work within the scope of their activities as required by the procurement documents. Suppliers also assure, through audit or surveillance, the adequacy of implementation. LIPA shall verify overall-program adequacy and implementation through planned and periodic audits.

The LIPA NQA Department shall'bp informed of corrective action determinations associated with safety related structures, systems and components. In addition, the NQA Department is responsible for verifying proper implementation of internal corrective action associated with safety related structures, systems, and components.

17.2.17 Ouality Assurance Records The LIPA QA Program imposes requirements on organizations _

l (~% performing safety related functions for QA records, which furnish l (_) documentary evidence of the quality of items and of activities l

17-24 Rev. 4 July 1992

affecting quality during the decommissioning phase. Section 17 of the LIPA QA Manual describes the QA Program requirements established for QA recordc.

The program requires that records documenting evidence of the quality of items and activities include results of reviews, inspections, tests, audits, and material analyses; monitoring of work performance; qualification of personnel, training procedures, and equipment; maintenance and modification activities; abnormal occurrences; and other documentation such as drawings, Lpecifications, procurement documents, calibration procedures and reports, nonconformance reports, and corrective action reports. Requirements for identification, transmittal, retention, and maintenance of quality related records subsequent to completion of work or prior to release of material or equipment for installation are to be indicated in procurement documents, specifications, procedures, or instructions and are to be consistent with applicable codes and standards. The program requires that inspection and test records specify a description of the type of observation, identification of the inspector or data recorder, evidence of completion or verification of manufacturing, inspection or test operation, the data and results of the inspection or test, information related to nonconformances,-and acceptability of the item inspected or tested.

The permanent plant filing system, developed during the design and construction phases and maintained during the operational phase, is known as the Shoreham Records Management System and is under the direction of the Nuclear Operations Support Department.

This system assures that QA records are readily identifiable and retrievable. The QA Program requires that the record storage facilities be constructed or located, and secured to prevent damage or loss of records due to fire, flooding, or environmental

. conditions.such as temperature or humidity or, alternatively, to maintain duplicate records stored in a separate remote location.

Suppliers performing safety related activities are responsible for imposing requirements for the generation, collection, stora90, and maintenance of QA records on their internal operations and on those vendors and contractors performing work within the scope of their activities as specified in the procurement documents. Suppliers also assure, through audit or purveillance, the adequacy of. program implementation.

The NQA-Department shall verify overall program adequacy and implementation by LIPA internal organizations and suppliers through planned and periodic audits.

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l 17-25 Rev. 4 July 1992 l

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SHOREHAM DSAR O

# 17.2.18 Audits The LIPA QA Program establishes provisions for a comprehensive system of planned and periodic audits to verify implementation of program requirements. Section 18 of the QA Manual describes the QA Program requirements for audits.

The program requires that a comprehensive system of audits be established for both internal and external functions that affect safety related structures, systems, and components to verify compliance with QA Program requirements as well as with approved QA procedures, the Shoreham Defueled Technical Specifications, administrative controls, and regulatory requirements. Audits shall include evaluations of quality related practices, effectiveness of implementation, conformance to policy, work areas,. activities and processes, and reviews of documents and records.

Audits shall be conducted to predetermined schedules. These schedules shall be reviewed, published annually, and ?pdated as required. Audit frequency shall be based on the status, safety and importance of the audited activity and results of prior audits. Audits shall be scheduled to ensure that implementation of QA Program requirements and related supporting procedures

(\

(_)

receive a comprehensive audit at least every two (2) years.

Those applicable elements of the QA Program in which quality related activities are more intensive and impacting upon daily operation shall be audited at 1 cast annually. Audits of nonroutine operations such as major modifications shall be y scheduled as necessary.

Audits shall be conducted in accordance with written, approved procedures, plans, and checklists by qualified personnel-not directly responsible for the area being audited. Audits shall provide for objective evaluation of the status and adequacy of the area audited.

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17-26 Rev. 4 July 1992 l