Text

Rev 3 to Shoreham Defueled Sar
	ML20082M508
Person / Time
Site:	Shoreham File:Long Island Lighting Company icon.png
Issue date:	08/26/1991
From:	; LONG ISLAND LIGHTING CO.
To:	;
Shared Package
ML20082M500	List: ML20082M508; SNRC-1828, Forwards Rev 3 to Shoreham Defueled SAR & Rev 5 to Fire Hazards Analysis Rept;
References
	NUDOCS 9109050232
	Download: ML20082M508 (246)
	v • d • e

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DSAR - REVISION 3 LIST OF EFFECTIVE PAGES REVISION PAGES , 3 . . . . . . .. 1-1 THRU 1-5 /gf J

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0 . . . . . . .. 1-6 - I 3 . . . . . . .. Table 1,2-1 0 . . . . . . .. 2-1 3 . . . . . . .. 2-2 and 2-2A 0 . . . . . . .. 2-3 and 2-4 0 . . . . . . .. 3-1 and 3-2 3 . . . . . . .. 3-3 0 . . . . . . .. 3-4 and 3-5 3 . . . . . . .. 3-6 and 6A 0 . . . . . . .. 3-7 and 3-8 3 . . . . . . .. 3-9 0 . . . . . . .. 3-10 thru 3-12 3 . . . . . . .. 3-13 thru 3-15 0 . . . . . . .. 3-16 3 . . . . . . .. 3-17 thru 3-23 3 . . . . . . .. 4-1 and 4-2 3 . . . . . . .. 5-1 3 . . . . . . .. 6-1 thru 6-3 0 . . . . . . .. 6-4 3 . . . . . . .. 7-1 THRU 7-6 0 . . . . . . .. 7-7 3 . . . . . . .. 7-8 3 . . . . . . .. B-1 THRU 8-4A 0 . . . . . . .. 8-5 0 . . . . . . .. F8.2.1-2 0 . . . . . . .. 9-1 and 9-2 3 . . . . . . .. 9-3 thru 9-6 1 . . . . . . .. 9-7 3 . . . . . . .. 9-8 thru 9-9A 1 . . . . . . .. 9-10 3 . . . . . . .. 9-11 and 9-12 1 . . . . . . .. 9-13 and 9-14 3 . . . . . . .. 9-15 and 9-15A 1 . . . . . . .. 9-16 and 9-17 3 . . . . . . .. T 9.2.1-1 (1 & 2 of 2) 3 . . . . . . .. T 9.2.7-1 0 . . . . . . .. T 9.3.2-1 (2 pages) 0 . . . . . . .. T 9.5.9-1 (1 & 2 OF 2) 0 . . . . . . .. F 9.1.2-1 r V

3 . . . . . . .. 10-1 AND 10-2 1 . . . . . . .. 11-1 3 . . . . . . .. 11-2 thru 11-5 0 . . . . . . .. 11-6 p,~) y 3 . . . . . . .. 11-7 thru 11-10 N' O . . . . . . .. 11-11 3 . . . . . . .. 11-12 thru 11-15 0 . . . . . .. 11-16 thru 11-19 1 . . . . . . .. 11-20 3 . . . . . . .. 11-21 0 . . . . . . .. 11-22r" a

3 . . . . . . .. 12 -2 3 va " H" O . . . . . . .. 11-24 3 . . . . . . .. 11-25 0 . . . . . . .. 11-26 0 . . . . . . .. T 11.6.3-1 0 . . . . . . .. F 11.6.3-2 0 . . . . . . .. 12-1 3 . . . . . . .. 12-2 thru 12-3A 0 . . . . . . .. 12-4 and 12-5 3 . . . . . . .. 12-6 thru 12-10 3 . . . . . . .. T 12.2-1 0 . . . . . . .. T 12.2-2 0 . . . . . . .. T 12.3.4-4A 0 . . . . . . .. T 12.3.4-4B (2 pages) 0 . . . . . . .. T 12.3.4C 0 . . . . . . .. T 12.4-1 p} 3 . . . . . . .. 13-1 thru 13-4 N_ / 0 . . . . . . .. 13-5 3 . . . . . . .. 13-6 thru 13-8 0 . . . . . . .. 13-9 3 . . . . . . .. 13-10 0 . . . . . . .. 13-11 thru 13-14 0 . . . . . . .. T 13.5.1-1 (1 thru 4 of 4) 0 . . . . . . .. 7 13.5.1-2 0 . . . . . . .. T 13.5.1-3 (1 thru 6 OF 6) 3 . . . . . . .. F 13.1.1-1 3 . . . . . . .. F 13.1.1-2 0 . . . . . . .. F 13.5.1-1 0 . . . . . . .. 14-1 0 . . . . . . .. 15-1 and 15-2 3 . . . . . . .. 15-3 0 . . . . . . .. 15-4 3 . . . . . . .. 15-5 and 15-6 0 . . . . . . .. 15-7 and 15-8 3 . . . . . . .. 15-9 0 . . . . . . .. 15-10 3 . . . . . . .. 15-11 (O

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r , 0 . . . . . . .. T 15.1.36-1 3 . .. . . . .. T 15.1.36-2 3 . . . . . . .. T 15.1.36A-1 0 . . . . . . .. F :.5.1-1 () 0 . . . . . . .. F 15.1.36-1 0 . . . . . . .. F 15.1.36A-1 3 . . . . . . .. 16-1 2 3 . . . . . . .. 17-1 thru 17-3 0 .. . . . . .. 17-4 3 . . . . . . .. F 17.2.1-1 l l l lC h

l l l t ATTNCHMENT 3 TO SNNC-1664 The S1orelam Nuclear Power Station Defuelec Safety Analysis Report

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SHOREHAM DSAR j - TABLE OF CONTENTS Chapter / Title Page Section 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT 1, 1 Introduction 1- 1 General Plant Description 1- 4

1. 2 Comparison Tables 1- 5

1. 3 Identification of Agents and Contractors 1- 5

1. 4

1. 5 Requirements for Further Technical 1- 5 Information

1. 6 Material Incorporated by Reference 1- 5 2 SITE CHARACTERISTICS Geography and Demography 2- 1

2. 1

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

2. 3 Hydrologic Engineering 2- 2

2. 4 Geology & Seismology 2- 2A

2. 5 Boring Logs 2- 3 2A Seismicity Investigations 2- 3 2B A Reevaluation of the Intensity of the E. 2- 3

{ 2C *

  \                           Haddam, Conn. Earthquake of May 16, 1791 2D                    Reevaluation of the Reported Earthquake at      2- 3 Port Jefferson, New York 2E                    Reevaluation of the Earthquake of October       2- 3 26, 1845 2F                   Reevaluation of the Earthquake of January       2- 3 17, 1855 2G                   Earthquakes Whice Have Affected the Site        2- 3 with Modified Mercalli Intensity    = IV Report on Seismic Survey-Proposed Shoreham      2- 3 2H Power Station LILCO                             2- 4 2I                    Laboratory Soils Test 2J                    Summary Report of Geotechnical Studies of       2- 4 Reactor Building Foundation Aircraft Crash Probability Study                2- 4 2K Report on Service Water System Soils            2- 4 2L                                                                   2- 4 2M                   Report on Densification of Service Water System Soils                                    2- 4 2N                   Hurricane Study 3                    DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS

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

O Rev. 3 July 1991 L

SHOREHAM DSAR TABLE OF CONTENTS

 -(/                             (Continued)

Chapter / Section Title Page

3. 2 Classification of Structures, Systems and 3-1.

Components

3. 3 Wind and Tornado Loading 3-14

3. 4 Wa'7r Level (Flood) Design 3-14

3. 5 Missile Protection 3-14

3. 6 Protection Against Dynamic Effects 3-14 Associated with the Fostulated Rupture of Piping

3. 7 Seismic Design 3-14

3. 8 Design of Seismic Category I Structures 3-14

3. 9 Mechanical Systems and Components 3-14 3.10 Seismic Qualif. of Seismic Category I 3-15 Instrumentation and Electrical Equipment 3.11 Environmental Design of Mechanical and 3-15 Electrical Equipment 3.12 Separation Criterion for Safety Related 3-15 Mechanical and Electrical Equipment 3A Computer Programs for the Stress Analysis 3-16 of Cat I Structures and Piping Systems I( ) 3B NRC Regulatory Guides 3-16 3C Pipe Failure Outside Primary Containment 3-16 4 REACTOR

4. 1 Reactor Summary Description 4- 1 Reactor Vessel 4- 1

4. 1. I Reactor Internal Components 4- 1

4. 1. 2 4- 2

4. 1. 3 Reactivity Control System Analysis Techniques 4- 2

4. 1. 4

4. 4 Thermal and Hydraulic Design 4- 2 Reactor Materials 4- 2

4. 5 Control Rod Drive Housing Supports 4- 2

4. 6 5 REACTOR COOLANT SYSTEM Summary Description 5- 1

5. 1 6 FNGINEERED SAFETY FEATURES General 6- 1

6. 1 6- 1

6. 2 Containment Systems Containment Functional Design 6- 1

6. 2. 1 6- 2 6.2.2 Containment Heat Removal System

6. 2. 3 Containment Air Purification and Cleanup 6- 2 System Containment Isolation System 6- 2

6. 2. 4 6- 2

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

   \       6. 3      Emergency Core Cooling Systems                 6-  3

6. 4 Habitability Systems Rev. 3 July 1991

N SHORERAM DSAR [G TABLE OF CONTENTS (Continued)

  \ ,)

Chapter / Section Title Page

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

6. 6 Overpressurization Protection 6- 3

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

6. 8 Control Rod Velocity Limiter 6- 3

6. 9 Main Steam Line Flow Restrictor 6- 3 6.10 Reactor Core Isolation Cooling System 6- 3 6.11 6.12 Standby Liquid Control System 6- 4 l 7 INSTRUMENTATION AND CONTROLS Introduction 7- 1

7. 1

7. 1. 1 Identification and Classification of 7- 1 Nonsafety Related Systems Identification of Safety Design Bases and 7- 6 ,

7. 1. 2 Nonsafety Design Bases Criteria l I

Reactor Protection System 7- 6

7. 2 Engineered Safety Feature System 7- 6

7. 3 Systems Required For Safe Shutdown 7- 7 k-'-

7. 4 Safety Related Display Instrumentation 7- 7

7. 5

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

7. 6. 1

7. 7 Control Systems Not Required for Safety 7- B 7A Plant Nuclear Safety Operational Analysis 7- 8 Analog Transmitter / Trip System for ESF 7- 8 7B Sensor Trip Units 8 ELECTRIC POWER 1

Introduction 8- 1 i

8. 1 Utility Grid 8- 1

8. 1. 1 Interconnection to Other Grids 8- 1

8. 1. 2 8- 1

8. 1. 3 Offsite Power System On Site AC Power System 8- 2

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

8. 1. 5 Identification of Safety Related System 8- 2

8. 1. 6 8- 2A

8. 1. 7 Identification of Safety Criteria Offsite Power System 8- 3

8. 2 Description 8- 3

8. 2. 1 8- 4

8. 2. 2 Analysis On Site Power System 8- 4

8. 3 AC Power System 6- 4

8. 3. 1 8- 4

8. 3. 2 DC Power System IO

   'Ns /     9           AUXILIARY SYSTEMS Fuel Storage and Handling                           9- 1

9. 1 Rev. 3 July 1991

SHORERAM DSAR 'I TABLE OF CONTENTS

\                                         (Continued)

Chapter / Section Title Page New Fuel Storage 9- 1

9. 1. 1 Spent Fuel Storage 9- 1

9. 1. 2

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

9. 1. 4 Water Systems 9- 3

9. 2 Service Water System 9- 3

9. 2. 1

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

9. 2. 3 Potable and Sanitary Water Systems 9- 5

9. 2. 4 Ultimate Heat Sink 9- 5

9. 2. 5 9- 5

9. 2. 6 Condensate Storage Facilities

9. 2. 7 Turbine Building Closed Loop Cooling Water 9- 5 System Main Chilled Water System 9- 5

9. 2. 8

9. 2. 9 Reactor Didg Standby Vent Sys and Control 3- 6 Room AC Chilled Water System Process Auxiliaries 9- 6

9. 3

[- Compressed Air Systems 9- 6 g 9. 3. I 9- 6

9. 3. 2 Process Sampling System Equipment and Floor Drainage System 9- 6

9. 3. 3

9. 3. 4 Chemical, Volume Control and Liquid Poison 9- 7 Systems Failed Fuel Detection System 9- 7

9. 3. 5 9- 7

9. 3. 6 Suppression Pool Pumpback System

9. 4 Air Conditioning, Heating, Cooling, and 9- 7 Ventilation Systems Control Room Air Conditioning System 9- 8

9. 4.1

9. 4. 2 -Reactor Building Normal Ventilation System 9- 8 Radwaste Building Ventilation 9- B

9. 4. 3

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

9. 4. 5 Battery Room Heating and Ventilation 9- 9 Drywell Air cooling System 9- 9

9. 4. 6

9. 4. 7 Screenwell Pump House Heating and 9- 9 Ventilation Plant Heating 9- 9 9.4.8 9- 9

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

9. 4.10 Diesel Generator Room Ventilation 9- 9

9. 4.11 Relay Room, Emer. Switchgear Room &

Computer Room Air Cond. System 9-10

9. 5 Other Auxiliary Systems Fire Protection System 9-10

9. 5. I 9-14 Communication Systems 0- 9.

9.

3. 2

5. 3 Lighting Systems 9-14 Rev. 3 July 1991

SHOREHAM DSAR () - TABLE OF CONTENTS (Continued) . Chapter / Section Title Page

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

9. 5. 5 Diesel Generator Cooling Water System, 9-15

9. 5. 6 Diesel Generator Starting System 9-15

9. 5. 7 Diesel Generator Lubrication System 9-15 ;

9. 5. 8 Primary Containment Leakage Monitoring 9-15 =

System

9. 5. 9 Storage of Gas Under Pressure 9-15 r 9A Fuel Criticality Analysis 9-16 {

9B Evaluation of Spent Fuel Pool Makeup 9-17 ! Requirements 70 STEAM AND POWER CONVERSION SYSTEM , Steam and Power Conversion System 10- 1

10. I l

Turbine Generator 10- 1 ;

10. 2 Main Steam Supply System 10- 1

10. 3 Other Features of Steam and Power 10- 1 ;

i 10. 4 ; Conversion System I 10- 1 l [( Condenser l

10. 4. 1 10- 1

10. 4. 2 Main Condenser Air Removal System 10, 4. Steam Seal System 10- 1 l 3 !

Turbine Bypass System 10- 2

10. 4. 4 10- 2 i

10. 4. 5 Circulating Water System 10- 2

10. 4. 6 Condensate Demineralizer System t

10. 4, 7 Condensate and Feedwater System 10- 2 l

l 11 RADIOACTIVE WASTE MANAGEMENT - Radiation Source Terms 11- 1 l 11. 1 11- 2

11. 2 Radioactive Liquid Waste System 11, 2. I Design Objectives 11- 2 ,

System Descriptions 11- 2 l

11. 2. 2 11- 3 i

11. 2. 3 System Design 11- 5 Operating Procedures !

11. 2. 4 11- 6

11. 2. 5 Performance Tests 11- 6 Estimated Releases ,

ll. 2. 6 11- 6

11. 2. 7 Felease Points 11- 6 l

11. 2. 8 Dilution Factors 11- 6

11. 2. 9 Estimated Doses 11- 7 !*

11. 3 Gaseous Waste System l

Design objectives 11- 7

11. 3. I 11- 7 System Descriptions

11. 3. 2 11- 7 t

11. 3. 3 System Design 11- 7 11, 3. 4 Operating Procedures 11- 7 l 11. 3. 5 Performance Tests 11- 7

[] N 11. 3. 6 Estimated Releases 11- 8

11. 3. 7 Release Points r

Rev. 3 July 1991 ! l

   .                                                                            T

SHOREHAM DSAR TABLE OF CONTENTS (Continued) Chapter / Section Title l'ag e

11. 3. 8 Dispersion Factors 11- 8' Estimated Doses 11- 8

11. 3. 9 Unmonitored Release Points 11-'8

11. 3.10 .

11. 4 _ Process & Effluent Radiation Monitoring 11- 8 System Solid Waste System 11- 8

11. 5 II, 5. I Design Objectives 11- 8 System Input: Source Terms 11- 9

11. 5. 2 Equipment Description 31- 9

11. 5. 3

11. 5, 4 . Expected Volumeu 11-11

11. 5. 5 Packaging 11-11 11, 5. 6 -Storage' 11-11

11. 5. 7 Shipment 11-11 Offsite Radiological Environmental 11-11

11. 6 Monitoring Plan 11, 6. 1 Objectives of REMP 11-16

11. 6. 2 Potential Pathways 11-18 11, 6. 3 Sampling Media, Locations, and Frequency 11-19 11, 6. 4 Not Used'in the DSAR 11-20 '

Data Analysis, Presentation, and 11-21

11. 6.-5 Interpretation 11, 6. 6 Program Statistical Sensitivity 11-21 12 RADIATION PROTECTION

12. 1 Assuring that Occupational Radiation 12- 1 Exposures are ALARA Radiation Sources 12- 2

12. 2 12- 2

12. 2. 1 Contained Sources Airborne Radioactive !!aterial Sources 12- 3

12. 2. 2 12- 4 12._3 Radiation Protection Design Features

     '12,     3. 1      Facility-Design Features                       12- 4 Shielding                                      12- 4

12. 3. 2 12--5

12. 3. 3- Ventilation

12. 3. 4 Radiation Monitoring Instrumentation 12- 5

12. 4 Dose Assessment 12 Design Objectives 12 _6-

     -12. 4. 1                                                          7

12. 4. 2 Airborne Activity Occupational Dose Assessment 12- 7

12. 4. 3 12- 8 12, 4. 4 Offsite Dose Assessment 12 12. 5 Health Physics-Program 13 CONDUCT'OF OPERATIONS Organizational Structure of Applicant 13- 1 3 13. 1 13- 1

13. 1. 1 Corporate Organization 13- 1A

13. 1. 2 Nuclear Operations Support Organization Rev. 3 July 1991

SHORERAM DSAR () TABLE OF CONTENTS (Continued) Chapter / Page

          .Section                             Title

13. 1. 3 Nuclear Engineering Organization 13- 2

13. 1. 4 Operating Organization 13- 2

13. 1. 5 Qualification Requirements for Station 13- 4 Personnel

13. 2 Training Program 13- 4

13. 2. 1 Program Description 13- 4 13, 3 Emergency Planning 13- 5 13, 4 Review and Audit 13- 5

13. 4. 1 Review and Audit - Construction 13- 5

13. 4. 2 Review and Audit - Test and Operation 13- 6

13. 4 3 Shoreham Independent Safety Engineering 13-11 Group

33. 5 Station Procedures 13-11

13. 5. 1 Administrative Control 13-11 >

13. 5. 2 Procedures 13-12

13. 6 Plant Records 13-14

13. 7 Industrial Security 13-14 g .

14 INITIAL TESTS AND OPEIATIONS , ACCIDENT ANALYSIS 15 '

15. 1 General 15- 1 General Load Reduction 15- 2

15. 1. 1

15. 1. 2 Turbine Trip 15- 2

15. 1. 3 Turbine Trip and Failure of Generator 15- 2 Breakers to Open Main Steam Isolation Valve Closure 15- 3 ,

15. 1. 4

15. 1. 5 Pressure Regulatory Failure - Open 15- 2 '

l 15. 1. 6 Pressure Regulatory Failure - Closed 15- 3 Feedwater-Controller Failure-Maximum Demand 15- 2 ; l 15. 1. 7 ' Loss of Feedwater Heating 15- 2

15. 1. 8

Shutdown Cooling (RHR) 15- 2

15. 1. 9 Malfunction-Decreasing Temperature l 15. 1.10 Inadvertent HPCI Pump Start 15- 4 l 15. 1.11 Continuous Control Rod Withdrawal During 15- 4 r

I Power Range Operatio.r.

15. 1.12 continuous Control Rod Withdrawal During 15- 4 ,

Reactor Startup

15. 1.13 Control Rod Removal Error During Refueling 15- 4
15. 1.14 Fuel-Assembly Insertion Error During 15- 4 l Refueling: 15. 1.15 Off-Design Oper Transient Due to Inadvertent 15- 4 ,

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

15. 1.17 Inadvertent opening of a Safety Relief Valve 15- 4 l Rev. 3 July 1991 i

SHOREHAM DSAR TABLE OF CONTENTS (' ) (Continued) Chapter / Title Page Section 15, 1.18 Loss of Feedwater Flow 15- 3 Loss of AC Power 15- 3

15. 1.19 15- 3

15. 1.20 Recirculation Pump Trip Loss of Condenser Vacuum 15- 3

15. 1.21 15- 3

15. 1.22 Recirculation Pump Geizure

15. 1.23 Recirculation Flow Control Failure - 15- 3 Decreasing Flow Recirculation Flow Control Failure With 15- 4

15. 1.24 Increasing Flow

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

15. 1.26 Core Coolant Temperature Increase 15- 3

15. 1.27 Anticipated Transient Without Scram (ATWS) 15- 5 Cask Drop Accident 15- 5

15. 1.28

15. 1.29 Miscellaneous Small Release Outside Primary 15- 5 Containment

15. 1.30 off-Design Operational Transient as a 15- 5 Consequence of Instrument Line Failure

  /"'       15. 1.31    Main Condenser Gas Treatment System Failure        15- 5
                                                                          '15-~ 5 ,
           -15. 1.32    Liquid Radwaste Tank Rupture                       15- 4

15. 1.33 Control Rod Drop Accident 15, 1.34 Pipe Breaks Inside the Primary Containment 15- 5 (Loss-of-Coolant Accident) 15- 5

15. 1.35 Pipe Breaks Outside the Primary Containment (Steam Line Break Accident) 15- 7

15. 1.36 Fuel Handling Accident Worst Case Fuel Damage Event 15-10 l

15. 1.36A 15- 4

15. 1.37 Feedwater System Piping Break Failure of Air Ejector Lines 15- 7

15. 1.38 e

16 TECHNICAL SPECIFICATIONS 17 QUALITY ASSURANCE 17- 1

17. 1 Quality Assurance During Design and Construction 17- 1

17. 2 Quality Assurance During the Operational Phase 17- 2

17. 2. 1 Organization 17- 3

17. 2. 2 Quality Assurance Program 17- 3

17. 2. 3 Design Control 17- 3

17. 2. 4 Procurement Document Control

17. 2. 5 Instructions, Procedures, and Drawings 17- 3 ;

17- 3 17, 2. 6 Document Control 17- 3

    ,s       17, 2. 7     Control of Purchased Material, Parts, and fy,         17. 2. 8 Services Identification and Control of Special             17- 3 Processes Rev. 3     July 1991

i 1 SHOREHAM DSAR TABLE OF CONTENTS (/' (Continued) Chapter / Page Section Title Control of Special Processes 17- 3

17. 2. 9 17- 3

17. 2.10 Inspection 17- 3

17. 2.11 Test Control

17. 2.12 Control of Measuring and Test Equipment 17- 4

17. 2.13 Handling, Storage, and Shipping 17- 4 17, 2.14 Inspection, Test, and operating Status 17- 4

17. 2.15 Nonconforming Materials, Parts, or 17- 4 Components Corrective Action 17- 4

17. 2.16 17- 4

17. 2.17 Quality Assurance Records 17, 2.18 Audits 17- 4 O

(V f Rev. 3 July 1991 .

SHOREHAM DSAR [\ CHAPTER 1 INTROLUCTION 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 Lighting Company, hereafter-known as LILCO, in support of its application to amend Facility Operating License NPF-82 as described in SNRC-1664. The 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, commensurate with the level of activity of a defueled plant, the content is reduced. l The purpose of the DSAR is to provide a safety analysis for the storage and handling of Shoreham low burnup first cycle spent i fuel. The DSAR confirms that-fuel storage and handling systems,

     /~' structures, components and programs ensure that there is no undue

, k,) risk to public health and safety during normal and postulated accident conditions. The DSAR assumes that the 560 fuel bundles comprising the 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 the structures-are secondary containment of the -l Shoreham reactor building. 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 This burnup of the fuel is approximately 2 full power days. results in an estimated total core wide heat generation rate of l approximately 550 watts as of June 1989. The estimated fuelFigureheat load will reduce to approximately 250 watts by June 1991. 15.1-1 depicts the fuel heat load versus time. Based on this low 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. L i O l-1 Rev. 3 July 1991

SHOREilAM DSAR The Shoreham spent fuel contains limited quantities of radioactive materials that are available for release. As is 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 two year decay from the last burnup period (completed June 7, 1967). 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 accidents identified in the USAR which are applicable to the defueled plant . In addition, no other accident mechanisms were identified for the plant's defueled condition which are not bounded by Chapter 15. The events analyzed in Chapter 15 are:

1. Fuel Handling Accident (Fuel Bundle Drop)

2. Radwaste Tank Rupture The only design basis accident involving reactor fuel is a Fuel Handline Accident , in which no heat generation takes place. As ye~s such, the activity available for release in this design basis i 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 This event was postulated to released to the reactor building.

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 15 indicate that integrated doses are very small ir comparison with 10CFR100 limits. For the worst case scenario ir shich 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 approximately .03% of 10CFR100 limits. The results of the radiological analysis for the worst case fuel damage scenario are depicted graphically in Figure 15.1.36A-1. In particular, it l was demonstrated that the reactor building standby ventilation system operation does not provide an important filtering or ventilation safety function and is therefore no longer required after fuel is stored in the pool. Based on this analysis, it has been found that the spent fuel f[\~/) pool provides a high degree Active of passive safety protection for safety systems are not required to Shoreham spent fuel. t 1-2 Rev. 3 July 1991

SHORERAM DSAR h 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 building services. Therefore a reclassification of safety systems is proposed based on the importance to safety associated with each plant system with the plant defueled. The DSAR assumes that the Shoreham spent fuel f rom 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 follows:

1. All 560 fuel bundles have been removed from the reactor and are being stored in seismic Category I spent fuel racks in 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 water level is maintained at its normal water level. Makeup p(f, ' will be furnished from the condensate transfer system or the

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

5. Two independent offsite AC power sources will be maintained In addition, as discussed to supply reliable electric power.

in Chapter 8, blackstart combustion turbines exist nearby in the Shoreham west site to supply emergency power to the 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 h~')

(s and monitored release capability but secondary containment integrity is no longer required as discussed in the DSAR Chapter 15 Safety Analysis. Rev. 3 July 1991 1-3

s SHORERAM DSAR

7. The steam and power conversion systems are not required to be operable 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 appropriate level of radioactive liquid and r;olid waste management primarily due to operation of the spent fuel pool.

10. Major systems that remain functional to provide non-safety related supporting services includes a) Service Water (DSAR Chapter 9 and 10) b) Chilled Water Systems (DSAR Chapter 9) c) Compressed Air (DSAR Chapter 10) d) HVAC Systems (DSAR Chapter 9)

The DSAR addresses the following major programs:

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

(n (DSAR Chapter 16) Nx-

2. Conduct of operations and the LILCO organizational structure is described in Chapter 13. The ISEG functions are no longer considered necessary for a defueled reactor.

3. The Quality Assurance Program is maintained as described in DSAR Chapter 17. l

4. The Fire Protection Program is maintained as described in DSAR Section 9.5.1 and the FHAR.

5. An offsite Radiological Environmental Monitoring Program (ITJT) is maintained as described in DSAR Section 11.6.

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

7. A Defueled Emergency Plan was submitted separately for NRC l review and approval via SNRC-1651.

1.2 GENEFN PLANT DESCRIPTION This section of the USAR is historically descriptive but the specifics of general and design criteria and modes of operation are generally no longer applicable to the defueled plant. Design (~s) and operating information will be found in other sections of the (d

  '       DS AR e .g . , Table 3. 2-1.

1-4 Rev. 3 July 1991

SHOREHAM DSAR 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 support system (s) not required per Technical Specifications but necessary for minimal plant functions, habitability, and maintenance. l

3. Nonoperable - Those systems not normally operated in the defueled mode. These systems will be in the deenergized state. All systems will be maintained consistent with the Decommissioning Rule (no action will be taken which will affect the methods or options available for decommissioning or increase the cost of decommissioning prior to approval of l a decommissioning plan.)

l 1.3 COMPARISON TABLES l i fe~s 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 AND CONTRA. TORS 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. 1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION 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. However, the status of l systems which will remain operable for an extended time period in the defueled condition is described in Table 1.2-1 of the DSAR. l l The systems described in this section are not required for the defueled condition. l 1.6 MATERIAL INCORPORATED BY REFERENCE The i, formation contained under this heading in the latest revision of the Shoreham USAR remains unchanged. Refer to the USAR for information on this subject. fl V l-5 Rev. 3 July 1991

i SHOREHAM DSAR- > l I [ 1.7 . SYMBOLS USED IN ENGINEERING DRAWINGS ! The information contained under this heading in the latest f revision of the Shoreham USAR remains unchanged. Refer to the .! USAR for information on this subject. l I, i h i

I

                                                                   +.                      l I

D ' L o l 1-6

SHORERAM DSAR TABLE 1.2-1 f) OPERABLE PLANT SYSTEMS IN THE DEFUELED CONFIGURATION{FOR AN EXTENDED PERIOD OF TIME (RB) Cranes, Hoists and Elevators Reactor Building Superstructure Process Radiation Monitoring

Area Radiation Monitoring
Servicing Aids (Fuel)

Refueling Radwaste* Fire Protection (Mechanical)* Meteorological Monitoring Station Transformer (NSS) Non-Segregated Buses Metal Clad Switchgear Load Centers and Unit Substations Fire Detect & Station Security

l (Electrical /I&C) 13B/69kv Switchyard Pot. Transf.

138kv Switchyard Relay Panels O Reactor Building Reactor Building Ventilation

Reactor Building Standby Ventilation (shared l

portion only)* Seismic Monitoring Emergency Diesel Generator l 125V D.C. Distribution l I l l OPERABLE: Sy stem (s) maintained to meet Technical Specifications.

                                                                             & PCP.                  l

Needed to fulfill requirements of FRAR, ODCM, O

Rev. 3 July 1991

SHOREHAM DSAR-CHAPTER 2-h SITE CHARACTERISTICS [ 2.1 GEOGRAPHY AND DEMOGRAPHY t The description con,tained under this heading in the latest ! revision'of Shoreham USAR' remains unchanged. Refer to USAR for l information on this subject. j i 2.2 NEARBY INDUSTRIAL, TRANSPORTATION AND MILITARY FACILITIES l 1 The description contained under this heading in the latest ! revision of Shoreham USAR remains unchanged. Refer to USAR for j 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 r ft. tower south of the plant will not be used. Additionally, the j following information regarding the Operational Program applies i to DSAR. Refer to USAR.for other information on this subject. j r

 $        2.3.3.2     Operational Procram                                        _

The operational meteorological monitoring program uses ! j instrumentation to determine wind-speed and -direction at 33- and l l 150-ft. ambient air temperature at-33-ft and temperature [ differential (Temp @ 150-ft minus Temp 9 33-ft). These i instruments are located on SNPS' 400 ft. meteorological tower l which is located approximately 5100-ft WSW of the reactor j building (Figure 2.1.1.1). The MET tower was positioned ! sufficiently close to SNPS to provide representative observations i of released gaseous effluents, but far enough away to minimize } atmospheric disturbances caused by SNPS' structures. l t l Wind-speed and -direction at the 33-ft level, along with the ! temperature differential are transmitted to the Technical Support l Center. In addition to these parameters, wind-speed and ;

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

to the Main Control Room and entered into the RMS computer. t A11' 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 33 and 150 ft. levels. Temperature sensors in shielded aspirators are oriented in a northerly direction to limit the l (() influence of solar insolation. A motor and fan draw a constant 2-1

SHORERAM DSAR

  )    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 required periodic reports. To ensure the operability of the system, semi-annual calibrations l are mrformed by a qualified vendor, and channel checks are peric.med by the operators on shif t 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 ten.erature monitor at the 33-ft elevation; and

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

I 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 I'\ substitutions under the direction of LILCO personnel. 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 l replacement of any malfunctioning components of the system. In i the event that a Technical 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 EYDROLOGIC 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: 3 2.4.8.1 Canals 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 h exceeded one (1) foot. This one (1) foot maximum sediment depth , requirement is based upon anticipated sediment deposition of 3.2 2-2 Rev. 3 July 1991 l i

SHOREHAM DSAR 9,/ 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 monitor:r. 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 Requirements The USAR states that the required minimum safety related cooling water flow is 12,800 gpm supplied by two service water pumps. This minimum safety related flow is no longer required for the defueled condition since 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. 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 l() information on this subject. b 2-2A Rev. 3 July 1991

s SHOREHAM DSAR 2A BORING LOGS The description contained under this heading Refer ir. thetolatest USAR for revision of Shoreham USAR remains unchanged. information on this subject. 2B SEISMICITY INVESTIGATIONS

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

2C A REEVALUATION OF 'r"E INTENSITY OF THE EAST HADDAM, CONNECTICUT EARTHQUAKE OF MAY 16, 1971 The description contained under this heading Refer in thetolatest USAR for revision of Shoreham USAR remains unchanged. information on this subject. 2D REEVALUATION OF THE REPORTED EARTHQUAKE AT PORT JEFFERSON, LONG ISLAND, NEW YORK The description contained under this heading Refer in thetolatest USAR for revision of Shoreham USAR remains unchanged. information on this subject, ip_D OCTOBER 26, 1845 (/ 2E REEVALUATION OF THE EARTHQUAKE OF The description contained under this heading Refer in thetolatest USAR for revision of Shoreham USAR remains unchanged. information on this subject. 2F REEVALUATION OF THE EARTHQUAKE OF JANUARY 17, 1855 The description contained under this heading Refer in thetolatest USAR for revision of Shoreham USAR remains unchanged. inforn.ation on this subject. 2G EARTHQUAKES WHICH HAVE AFFECTED THE SITE AREA WITH A MODIFIED MERCALLI INTENSITY OF IV OR GREATER The description contained under this heading Refer in thetolatest USAR for revision of Shoreham USAR remains unchanged. information on this subject. 2H REPORT ON SEISMIC SURVEY-PROPOSED SHOREHAM POWER STATION LONG ISLAND LIGHTING COMPANY The description contained under this heading Refer in thetolatest USAR for revi ion of Shoreham USAR remains unchanged. Px information on this subject. 2-3

i SHOFEMAM DSAR j i ( 2I IABORATORY SOILS TEbTS The description contained under this heading in thc. latest revision of Shoreham USAR remains unchanged. Refer to USAR for ! information on this subject. 2J 811MMARY REPORT OF GEOTECHNICAL STUDIES OF REACTOR BUILDING FOUNDATION t The description contained under this heading in the latest ( revision of Shorehem USAk remains unchanged. Refer to USAR for inforanation on this subject. l 2K AIRCRATT CRASH PROBABILI'lY STUDY The description contair.ed under this heading in the latest l revision of Shoreham USAR remains unchanged. Refer to USAR for i j fnformation on this subject. ) i 2L REPORT ON SERVICE WATER SYSTEM SOILS i c The description contained under this heading in the latest ; revision of Shoreham USAR remains unchanged. Refer to USAR for ! information en this subject. l l' t 2M REPORT ON DENSITICATION OF SERVICE WATER SYSTEM SOILS , The description contained under this heading in the latest ! revision of Shoreham USAR remains unchanged. Refer to USAR for l information on this subject. HURRICANE STUDY 2N , The description contained under this heading in the latest h revision of Shoreham USAR remains unchanged. Refer to USAR for ! information on this subject. . i i I i i

  .(O i

2-4

l i ' SHORERAM DSAR

l CHAPTER 3 i

DESIGN OF STRUCTURES, COMPONENTS, EQUIMENT,_AND SYSTEMS l 3.1 CONFORMANCE TO GENERAL DESIGN CRITERIA FOR NUCLEAR POWER i PLANTS (10CTR 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 t.o the storege of SNPS low burnup cycle spent fuel in the spent fuel pc - The following GDC are addressed:

1. Overall Requirements GDC1 Quklity Standards and Records GDC2 Design Bases for Protection Against Natural Phenomena GDC3 Fire Protection GDC4 Environmental and Dynamic Effects Design Bases II. Protection by Multiple Fission Product Barriers GDC13 Instrumentation and Control I GDC17 Electric Power Systems GDC18 Insp0ction and Testing of Electric Power Systems GDC19 Control Room IV. Fluid Syste,ms GDC4( Cooling Water GDC45 Inspection of Cooling Water System GDC46 Testing.of Cooling Water System VI. Tuel-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 h'ndling GDC63 honitoring fuel and waste storage GDC64 Monitoring radioactivity releases

      -The following GDC were found not to be applicable to a defueled teactor:

I overall Requirements GDC5 Sharing of struatures, systems, and components j< Shoreham is a single unit, thus the above criterion does not apply. 3-1

S!!ORERAM DSAR ' I i (~ Protection By Multiple Fission Product Barriere II GDC10 Reactor Design GDC11 Reactor Inherent Protection CDC12 Suppression of reactor power oscillations , GDC14 Reactor Coolant Pressure Boundary 2 GDC15 Reactor Coolant System ; GDC16 Containment Design The above criteria do not apply because the reactor and primary ; containment are not operable. l' III Pro.tection And Reactivity control systems GDC20 - 29 requirements apply only to an operating reactor protection and reactivity control systems l IV Fluid Systems - GDC 30-43 address reactor and containment systems required ; for povpr operation only. ! V Fenetor Containment t I GDC 50- 57 address the primary containment design which is no (, (")T longer required for a defueled reactor. , l ! Applicable Criterion Conformance Quality Standards and Records (Criterion 1) j Criterion Structures, systems, and components important to safety shall bc designed, fabricated, erected, and tested to quality standards ; commensurate 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 . supplemented or modified as necessary to assure a quality product ! in keeping with the required safety function. A quality t 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. - Design Conformance Structures, systems, and components are classified in Section 3.2. The LILCO OA program described in DSAR Chapter 17 assures 3-2 l L - . _ . - _ - - _ - _ _ _ _ - _ _

SHOREllAM DSAR that quality practices and documentation are maintained commensurate with the classification that is identified in this Defueled Safety Analysis Report (DSAR). l Design Basis for Protection Against 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 without loss of capability to perform their safety functions. The design bases for these structures, systems, and components shall reflects (1) appropriate consideration of the most severe of the natural phenomena that have been historically reported for the site and surrounding area, with r.ufficient margin for the limited accuracy, quantity, and peri;d of time in which the ' historical data have been accumulatt 1, (2) appropriate combinations of the ef fects of normal and accident conditions with the effects of the natural phenomena, and (3) the , importance of the safety functions to be performed. Design conformance (tC') y' The spent fuel racks, fuel pool, and tecctor building which are ' required to maintain the SNPS fuel in a safe condition are designed to withstand natural 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. Fire Protection (Criterion 3) Criterion Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and ef fect of fires and explosions. Noncombustible and heat resistant materials shall be used wherever practical throughout the unit, particulcrly in locations such as the 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 assure that their rupture or inadvertent operation does not significantly impair the safety capability of these structures, systems, and components. ((~

3-3 Rev. 3 July 1991

_ _ _ . . .. -..m % SHORERAM DSAR h x- Desfon conformance This criterion is satisfied by the SNPS fire protection program which is described in Section 9.5.1 of this report and tae USAR. Environmental and Missile Design bases (Criterion 4) criterion , Structures, systems, and components important to safety shall be designed to accommodate the ef fects 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 proteeted against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit. Design Conformance Chapter 15 of this report defines accidents that are applicable to spent fuel storage and fuel handling. TheThepool spent fuel is stored in the spent fuel storage pool. structure, ( Reactor Building, and spent fuel racks provide passive safety ( protection from missiles or other conditions that could cause The structural design basis of the fuel fuel mechanical damage. 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. Instrumentation and Control (Criterion 13) 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 Jts associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges. Design Conformance Instrumentation is provided to monitor spent fuel pool level and temperature as well as fuel pool cleanup. Instrumentation is provided for process and effluent radiation monitoring, area and airborne radiation monitoring, and accident monitoring. Radiation monitoring is maintained as described in DSAR Chapters 11 and 12. 3-4 L _

SHOREHAM DSAR

 !       Electric power Systems (Criterion 17)                                  4 Criterion An onsite electric power system and an offsite electric power system shall be provided to permit functioningThe  of structures, safety function systems, and components important to safety.
         .for each system (assuming the other system is not functioning) shall be to provide-sufficient capacity and. capability to assure that    (1) specified acceptable fuel design li'mits 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.

The ensite electric power supplies, including the batteries, and the ansite electric distribution system shall have sufficient indspandence, redundancy, and testability 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 separate rights of way) designed and located so as to minimize to the extent practical the likelihood of their simultaneous failure under operating and A switchyard postulated accident and environmental conditions. 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 ofOne theof these reactor coolant pressure boundary are not exceeded. 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. l Provisions shall be included to minimize the probability of losing electric power from any of the remaining supplies as a l ' 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 power supplies. Design-Conformance The criterion applies principally to the design of an operating reactor. As demonstrated in DSAR Chapter 15, active system are not required to provide cooling or makeup functions However, in the event of postulated accidents including a seismic event. operability of the electric power system will be required by O. Technical Specifications during fuel movement to provide for a controlled and monitored release capability in the event of a 3-5 l L

SHOREllAM DSAR fiel drop accident. One offsite power transmission system will

    ) be maintained to provide power for support system operation. In l

addition, blackstart combustion turbines exist nearby at shoreham-West to provide reliable power in the unlikely event of a loss-of-of fsite 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 B. l Inspection and Testing of Electric Power Systems

      ~(Criterion 18)

Criterion 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 switchboards, to assess the continuity of the systems and the conditions of their components. The systems shall be designed with a capability to test periodically (1) the operability and 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 conditions as close to design as practical, the full operation sequence that brings the systems into operation including operation of gO 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. Design 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 19) 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 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 l 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 conditicn during hot shutdown, and 3-6 Rev. 3 July 1991 L

SHOREHAM DSAR (2) with a potential capability for subsequent cold shutdown of the reactor through the use of su!. table procedures. 3-6A Rev. 3 July 1991 b - . _-

S!!ORERAM DSAR Design 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. Cooling Water (Criterion 44) Criterion A system to transfer heat from structures, systems, and components important to 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 (rS shall be provided to assure that for onsite electric power system ('j 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. Design 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 water and other support systems are expected to be normally availabic to provide plant building services; however, these systems do not fulfill a safety function. Inspection of Cooling Water System Teriterion 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. Deuten Conformance (_) The service water system which will be maintained functional is designed to permit appropriate visual inspection in order to 3-7 L

SHOREHAM DSAR assure the integrity of system components. See criterion 44

    )   response.                                                               .

Testing of Cooling ~W~ater System

~~

TCriterion 46)_ critertog The cooling water system shall be designed to permit appropriate periodic pressure and functional testing to assure (1) (2) the the structural and leaktight integrity of its components, operability and performance 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. Design Conformance See Criterion 44 response. l Control of Releases of Radioactive Materials to the Environment

~'

() TDriterion 60) griterion The nuclear power unit dnsign shall include means to control suitably the release of radioactive materials in gaseous and liquid effluents and to handle radioective solid wastes produced during normal reactor operation, Sufficientincluding holdup anticipated capacity shall be operational occurrences. 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. l Desiun 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 upFuel the pool release of liquid and gaseous effluents as required. cleanup and appropriate radwaste systems are provided and are described in Chapters 9 and 11. See also Criterion 61. O 3-6 L .- . - - - - - _ - _

SHORERAM DSAR Fual Storage and Handling and Radioactivity Control (Cdterion 61) t criterion-The fuel storage and handling, radioactive waste, and other systmas which may contain radioactivity shall be designed to asmae adequate safety under normal and postulated accident - comutions. These systems shall be designed, (1) with a capellity to permit appropriate periodic inspection and testing of components important to safety, (2) with suitable shielding for radiation protection, (3) with appropriate containment, confinement, and filtering systems, (4) with a residual heat rammal 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. De sign - con f ormance Fuel Storage and Handling 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 O safe fuel storage will be subject to appropriate inspection and testing requirements. Adepate shielding is provided by maintaining a minimum water depd1 over the active fuel. Dose rates at the refueling level widmut the ef fects 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 f acility. The Reactor , Building Normal Ventilation System (RBNVS) will be used-to provide ventilation and a monitored release pathway. Because the l gas activity present in the fuel and available for release is prhurily noble gas (Kr-85) , the filtering role of the Reactor Building Standby Ventilation System (RBSVS) is not required. Cutain components of the RSSVS 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 from accidents are small in comparison to 10CFR100-1hdts, and neither the Reactor Building Standby Ventilation Systen nor secondary containment integrity is required to reduce offsite doses due to postulated accidents. Radiation monitoring is provided as described in Chapter 11 and 12 to detect radiological releases. Because of the extremely low residual heat load (approximately 550 watts) associated with the SNPS spent fuel, active fuel pool 3-9 Rev. 3 July 1991 e

SHOREHAM DSAR r Y Reliable fuel pool makeup sources 9 cooling is not required. including condensate storage, demineralized water, and fire protection water, are capable of maintaining pool water inventory to compensate for evaporation. Chapter 9 contrins a complete discussion of makeup requirements. The fuel pool is a Seismic Category I structure. Systems that connect to the pool (fuel pool 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 waste systems provide all equipment necessary to collect, process, and prepare for disposal of all radioactive liquids and solid waste produced as a result Any of spent fuel85 will Krypton storage. The off-gas system is rot needed. be retained within the fuel cladding. Should pin-hole leaks devalop, the gases will be handled by the ventilation systems. They will be discharged to atmosphere via the main plant vent. The radiological consequences of this type of release are negligible. This accident is bounded by the analysis of the fuel Handling Accident (Section 15.1.36). Liquid radwastes are collected, classified, and treated as high (r^} (_j conductivity, lou conductivity, chemical or laundry wastes. processing includes filtration, ion exchange, analysis, and dilution. Wet solid wastes are packaged in steel containers or polyethylene high integrity containers. Dry solid radwastes are compressed and/or packed in steel drums or boxes. Accessible portions of the spent fuel pool area and radwaste building have sufficient shielding to maintain dose rates within the limits set forth in 10CFR20 and 10CFR100. The radwaste

building is designed to preclude accidental release of l 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 I accident conditions. The design of these systems meets the requirements of Criterion 61. Radwaste systems are designed to meet the limits for effluents set forth in 10CFR20 and 10CFR50. l Prevention of Criticality in Fuel Storace Handlina l ((riterion62)_ Criterion (s\ b 3-10 l b

SHOREHAM DSAR 3 Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of' geometrically safe configurations. Design conformance Appropriate plant fuel handling and storagefor facilities are new and spent provided to preclude accidental criticality' prevented by the fuel. Criticality in' spent fuel storage is geometrically safe configuration of the storage rack. There is sufficient spacing between the assemblies to assure that the array, when fully loaded, is substantially suberitical. ruel elements are limited by rack design to only top loading and designated fuel assembly positions. Spent fuel is stored under water fuel in the spent assemblies fuel storage are placed are designed pool. The racks-in which spent , and arranged to ensure suberiticality in the storage pool. Spent fuel is maintained at a suberitica3 multiplication factor k-eff of less than 0,95 for both normal and abnormal storage , conditions. The fuel handling system is decigned 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 systems precludes ' accidental criticality in accordance with criterion 624 For further discussion, see the following section: Section 9A Criticality Analysis Monitorine ruel and Weste Storage (Criterion 63) ; Criterion Appropriate systems shall be provided in fuel storage and (1) to radioactive waste systems and associated handling areas, ; I detect conditions that may result in loss of residual heat removal capability and excessive radiation levels, and (2) to ; initiate appropriate safety actions. i Design Conformance Appropriate systems have been provided to meet the requirements . of this criterion. A malfunction of the fuel pool cleanup system is alarmed in the main control room. It is also alarmed Alarmed in the

                '     radwaste control room on high pressure differential.

conditions include high/ low fuel pool level. The refueling level O - ventilation exhaust radiation monitoring system detects abnormal amounts of radioactivity. As demonstrated in Section 9A and ; 3-11 L '! E . _ _ _ _ :

SHOREHAh DSAR 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 indication of conditions that may iesult in excessive radiation levels in the fuel storage and radioactive vaste system areas. These systems satisfy the requirements of Criterion 63. Monitoring Radioactivity Releases (Criterion 64) Criterion Means 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, and from postulated accidents. Design Conformance Means have been provided for monitoring radioactivity releases resulting from normal and anticipated operational occurrences. The following station release pathways are monitored () 1. 2. Gaseous releases from the station ventilation exhaust Liquid discharge to the discharge tunnel 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. l The semiannual Effluent Release Report is submitted to the NBC. 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 CLASSITICATION OF STRUCTURES, SYSTEMS AND COMPONENTS Seismic Category I structures, systems, and components are those i necessary to ensure:

1. The integrity of the reactor coolant pressure boundary

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

3. The capability to prevent or mitigate the consequences of

    -             accidents that could result in potential offsite exposures comparable to the guideline exposures of 10CTR100.

3-12

~_ _ - - - .- __- - . _ - _ - __ - - __ _ __ _ _ . _ _ _ _ _ _ _ . . t i SHOREHAM DSAR ! criteria 1 and 2 do not apply to a defueled reactor with i respect to the storage and handling of low burnup Shoreham spent fuel. A set of postulated accidents has been , identified and analyzed in Chapter 15 of this report that l defines the potential for a radiological release. Based on this analysis it has been concluded that potential radiological releases are far below the exposure limits of l 10CTR100. The analysis in Chapter 15 of this report assumes j 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 ! maintained for fuel handling equipment as well. 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 AS, active cooling of the spent fuel pool is not required and pool makeup requirements are minimal. Supporting systems are 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. St ructu re s , systems, and components whose safety functions require conformance to the quality assurance requirements of 10CTR50, Appendix B, are summari:ed in Table 3.2-1 under the heading, LILCO Quality Assurance Category, with the notation I. Modifications to CA 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 impact the spent fuel pool integrity, damage the spent fuel or damage or interfere with the operation of OA Category I structures or components. A key of definitions is provided at the end of Table 3.2-1. Chapter 17 discusses the graded level of 0.A. requirements l for this equipment. l O 3-13 Rev. 3 July 1991

i SilOREllAM DSAR , A V 3.3 WIND AND TOPNADO LOADING The inf ormation contained in the USAR remains the same although , the requirements to protect safe-shutdown equipnent no longer l 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 design information contained in this section is unchanged. However the spent fuel pool is the only area of the plant requiring missile protection. l 3.6 PROTECTION AGAINST DYNAMIC EFFECTS ASSOCIATED WITil POSTULATED RUPTURE OF PIPING In the defueled state high energy piping systems inside primary containment listed in USAR Table 3.6.1 A-1 are no longer pressurized and thus piping rupture need not be postulated. , k) 3.7 SEISMIC DESIGN Seismic design methods remain the sames 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 3.0 ' except that Safety Relief Valve (SRV) and LOCA hydrodynamic loads are no longer applicable to a defueled reactor, a 3.9 MECRANICAL 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. i In the future, mechanical equipment will be accorded the safety significance demonstrated by the classification in Table 3.2-1 of the DGAR. L 3-14 Rev. 3 July 1991 L

SHOREHAM DSAR l 3.10 SEISMIC QUALIr1 CATION OF SEISMIC CATEGORY I 1NSTRUMENTATION 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 Duilding, 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 longer required to be seismically qualified. l 3.11 ENVIRONMENTAL DESIGN Or MECHANICAL AND ELECTRICAL EQUIPMENT Electrical Equipment Environmental Qualification Purpose The purpose of the Electrical Equipment Environmental Qualification Program for Shoreham is to provide assurance that electrical equipment important to safety as defined by 10CTR50.49 located in potentially harsh environments maintains functional operability when required to mitigate the consequences of a postulated accident or to bring the plant to a cold shutdown condition afterward. Since the fuel has been removed and stored g in the fuel poni, LOCA or HELD cannot occur (see Chapter 15), and g' 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, 20CTR 50.49 is not applicable, therefore the environmental qualification program is not required. Environmente11y qualified electrical equipment will be designated O.A. Category II. l 3.12 SEPARATION CRITERION FOR SAFETY RELATED MECHANICAL AND ELECTRICAL EQUIPMENT The systems descrioed 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 O.A. Category II. l l 3-15 Rev. 3 July 1991 'L _

r S110RERAM DSAR 3A Computer Programs for the Stress Analysis of Category I ! and Dynamic and LH uctures, Dynamy~c and' Static Analysis

I StresisTAnalysis oM Seismic Category I Pj, ping Systems }

The description contained under this heading in the latest ; revision of Shoreham USAR remains unchanged. Refer to USAR for ; information on this subject. r 3 B' NRC Regulatory Guides l This section is described in the USAR. Specific topics are ! covered elsewhern in this DSAR. 3C Pipe railure outside Primary containment In the defueled state, piping systems outside primary contt:inment i which were considered high energy systems are no longer pressurized. Pipe rupture need no longer be postulated. ! i , e

        'O                                                                                                                                                                                             !

I p a . 3-16 t-- . . _ _ _ . _ - - _ _ _ _ . _ _ _ _ . . . _ _ _ _ . - . _ _ _ _ . . _ . . . . _ _ _ . _ - _ _ _ _ . . . . - _ _ _ _ _ . - . _ . . _ . . . - _ . . _ . - .

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

i SHOREMAM DSAR TABLE 3.2-1 EQUIPMENT CLASSIFICATION [ SPENT FUEL STORAGE I 5 LILCO j QUALITY [ SYSTEM / ASSURANCE SEISMIC i COMPONENT CATEGORY CATEGORY COMMENTS ( I I. Reactor System II N/A NA ! II Nuclear Boiler II N/A NR I III Recirculation System II N/A NR f IV -Control Rod Drive Hydraulic System II N/A NR [ V Standby Liquid Control System II N/A- NR i VI Neutron Monitoring II N/A NR ! VII Reactor Protection II N/A NR ! VIII Fixed Process, II N/A (1)

                               ' Airborne, and Effluent Radiation                                                                          i t

Monitors g I IX RHR II N/A Nh X Core Spray II N/A NR , XI . HPCI II N/A NR f

XII RCIC II N/A NR [ t XIII Fuel Service Equipment

1. Fuel preparation !

machine I I i

2. General purpose r

grapple I I XIV Reactor Vessel Service l Equipment

1. System Line Plugs II N/A NR

2. Dryer & Separator sling and KPV head strongback I I ,

l 3. Drywell head lifting ! rig I I l

I l 3-17 Rev. 3 July 1991 I T E. I

SHOREHAM DSAR i (/~S 5 ,) TABLE 3.2-1 ; 3 Continued) i EQUIPMENT CLASSIFICATION SPENT TUEL STORAGE , LILCO QUALITY SYSTEM / ASSURANCE SEISMIC COMPONENT CATEGORY CATEGORY COMMENTS ; i IV In-vessel Service Equipment ; i

1. Control rod grcpple I I j XVI Refueling Equipment l

1. Refueling platform I I

2. Refueling bellows, dryvell '4 I N/A j

3. Refueling bellows, cavity reactor II N/A -

4. New Fuel Inspection Stand II N/A NR XVII Storage Equipment f

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

2. Defective fuel storage container I I ,

3. Spent fuel pool, -

dryer /sep. pool, i reactor cavity liners I I l ;

4. Spent fuel storage ;

racks I I ; XVIII Radwaste System II N/A 3 XIX Reactor Water Cleanup System II N/A NR IX Tuel Pool Cleanup Subsystem

1. Demineralizer vessel II N/A

2. Filters II N/A

3. Pumps, purification

                    & transfer                                                                             II         N/A

4. Piping II N/A

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

        \           storage and air accumulator                                                                            II         N/A 3-18              Rev. 3 July 1991-L..

SHORERAM DSAR TABLE 3.2-1 (Continued) EQUIPMENT CLASSIFICATION SPENT FUEL STORAGE LILCO QUALITY SYSTEM / ASSURANCE SEISMIC COMPONENT CATEGORY CATEGORY COMMENTS XXI Fuel Pool Cooling 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

1. Electrical modules II N/A

2. Cable II N/A

( XXIV Offous System II N/A NR XXV Service Water System II N/A XXVI Compressed 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 Il N/A

c. Containment Elec-trical Penetrations II N/A NR

d. Fire Stops II N/A

e. DC Power Systems II N/A XXVIII Primary Containment Atmosphere II N/A NR Control XXIX- a) Reactor Building Normal Ventilation II N/A b) Reactor Building Standby Ventilation 11 N/A N R*

Certain components such as fans and valves will remain

           )                     functional to support RENVS operations.

3-19 Rev. 3 July 1991 u-

S!!ORERAM DSAR

     )                                       TABLE 3.2-1                                                                          ,

(Continued) EQUIPMCNT CLASSITICATION SPENT FUEL STORAGE LILCO QUALITY SYSTEM / ASSURANCE SEISHIC COMPONENT CATEGORY CATEGORY COHMENTS XXI Primary Containment Purge II N/A NR l XXXI Power Conversion II N/A NR l XXXII Condensate Storage and i Transfer II N/A I XXXIII Emergency Support Facilities

1. TSC Bldg. II I

2. EOT II N/A NR(3)

3. OSC II N/A

() XXIV XXXV MSIV Lenkage Control Miscellaneous II N/A NR l

1. RB Polar Crane I I (4)

2. ECCS Loop Level II N/A NR XXXVI Reactor Buildine N/A NR Closed Loop Cooling II l XXXVII Equipment and Floor Drains u II N/A l l XXXVIII Miscellaneous Ventilation l

Eystems

1. 125 Volt DC Battery room H & V II N/A l

2. Screenwell pumphouse II N/A l ,

l B&V

3. Relay and emergency l l

! switchgear H&V II N/A

4. Control room air con-ditioning, including filter trains II N/A

5. Diesel generator room II N/A I('N

  \                ventilation                                                                                                    ,

3-20 Rev. 3 July 1991 i

SHOREllAM DSAR TABLE 3.2-1 (Continued) , EQUIPMENT CLASSITICATION SPENT FUEL STORAGL: LILCO QUALITY SYSTEM / ASSUPANCE SEISMIC COMPONENT CATEGORY CATEGORY COMMENTS XXXIX Area Radiation Monitoring System ,

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 de1% e II N/A systems

2. Sprinklers, carbon dioxide systems II N/A (O V

3. Portable and wheeled extinguishers II N/A XLII Civil 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 ,

7. Discharge tunnel II N/A

8. Discharge pipe and diffuser II N/A

9. Radwaste Building II N/A (5) l

10. Auxiliary boiler and MG set building II N/A

11. Biological shielding II N/A

12. Missile barriers II N/A

13. Waterproof doors II N/A

14. Site grading II N/A

15. Masonry walls II N/A l bi '

3-21 Rev. 3 July 1991

SHORERAM DSAR h) - TABLE 3.2-1 (Continued) l EQUIPMENT CLASSIPICATION SPLN1 FUEL STORAGE l LILCO QUALITY SYSTEM / ASSURANCE SEISMIC COMPONENT CATEGORY CATEGORY COMMENTS ILIII Primary Containment Structure II N/A (5) l KLIV safety Parameter Display System II N/A NR l ILV Post Accident Sample System II N/A NR XLVI Containment Isola-tion Valve Position Indicator II N/A NR l XCVII Accident Monitoring II N/A NR l ' Instrumentation (NORIG 0578) f[ , O 3-22 Rev. 3 July 1991

SHORERAM DSAR

         )                                                                                                                     TABLE 3.2-1 (Continued)-

REY LILCO Quality Assurance Categoryt I - Meets 10CFR50 Appendix B requirements (same as USAR). II - Meet requirements of industrial and engineering standards (commercial grade quality). Seismic Category ) I - Equipment is designed in accordance with the seismic requirements for the DBE/OBE. N/A - Seismic requirements for DBE/0BE earthquake are not applicable to the equipment. Comments: NR - Not requirad (System secured from service or O' not required to support safe storage or handling of spent fuel). (1) - Seismic events will not create a radiological 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 LILCO Defueled Emergency Plan, the EOF is not required.

                                                                 -(4)                --              --only structurally safety related.                                                      l-(5)               -                  Originally constructed as Seismic-Category It modifications will be analyzed

- for DBE to ensure integrity of Reactor Building.

 'O 3-23                  Rev. 3 July 1991

SHORERAM 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 SUKHA.RY DESCRIPTION The NSS system is no longer needed for the defveled condition and hence is depressurized. 4.1.1 Reactor Vessel The reactor vessel design and description are covered in USAR Section 5.4.

 .       4.1.2     Reactor Internal Components The reactor internal components are as described in the USAR.                                The fuel rods and control rods are removed from the reactor.

t I e i [ i 4-1 Rev. 3 July 1991 f i

SHORERAM DSAR (3 If l 4.1.3 Reactivity Control System This system is no longer needed as there is no fuel in the f reactor vessel. : 4.1.4 Analysis Techniques The description contained under this heading in the latest ! revision of the USAR is no longer relevant in the plant's j defueled condition. I 4.4 THEAMAL AND HYDRAULIC DESIGN , The linear hett 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 Gene {gJElectricStandardApplicationforReactorTuel (GESTAR l i II). . 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 in the vessel in the defueled state and hence this system is not of concern. l 1 4-2 Rev. 3 July 1991

SHOREHAM DSAR {}

             \                                                                                 CHAPTER 5 REACTOR COOLANT SYSTEM 5.1

SUMMARY

DESCRIPTION The reactor coolant system includes those systems and components that contain or transport fluids coming from or going to the . reactor core. In the plant's defueled condition, the fuel is not ! in the core and the reactor is depressurized. Therefore, the reactor coolant system is not required. l I I i6 1: I t O ! v [ l i l l l l t l i p i O 5-1 Rev. 3 July 1991

    - i s tm-g+a +c w as -- =    gr- n w  y ~ _ .e w-,-,-wes       ,, , w-y-3.,-+e --,m,.wwe
                                                                                      ,      r a---*v-   *F-y--i ~'e- - *ev-i w.-C-wtwny-psy-www-T=--m'+ie-p-&*y*--+,

SHOREHAM DSAR CHAPTER 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 Transiento. 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 event. 6.2 CONTAINMENT SYSTEMS [ t 6.2.1 Containment Functional Design

A 6.2.1.1 Design Basis i I

6.2.1.1.1 Safety Criteria . t The primary containment system is not required and will not be i maintained functional as there will be no fuel within the primary [ containment structure. The secondary containment will maintain a t_ ' j subatmospheric pressure for postulated radiological accidents to ; j assure radiological monitoring of building releases. It is not needed to mitigate the consequences of sn accident. l 6.2.1.1.2 Design Basis Accidents The major design basis accident identified which will affect the secondary containment is the Fuel Handling Accident (Fuel Bundle l Drop). The results of this accident from a radiological ; l t standpoint are presented in Chapter 15. There are no pressure l and temperature effects of this accident and the RBNVS would [- l continue to maintain a subatmospheric condition. The other event which would have an effect on the secondary l containment is the loss of normal AC. l O ! 6-1 Rev. 3 July 1991 f i !

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

SHOREHAM DSAR O A loss of normal AC power may result in loss of the-substmospheric 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 part of any event which.results in fuel damage, while the radioactive releaee to the atmosphere would not be monitored, the offsite dose c.>ns_equences 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 shiel(ing is significant. There are no radiological consequences ns ociated wft! the loss of normal AC power. O . O I I 6-1A Rev. 3 July 1991

                                           ,.m   ,,,--y-w           yq   ,v,----a,-,      ,- n-pm,,-~wm     -m-      -mp--w      w y  ' y ,

SHORERAM DSAR 6.2.1.2 System Design The reactor building, which completely sneloses the primary containment and acts as the secondary containment, is maintained at subatmospheric pressure by the RBNVS. 6.2.1.3 Design 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 which is no longer maintained. . r 6.2.3 Containment Air Purification and Cleanup Systems ; This subsection is not applicable as it deals with the filtration i i portion of the RBSVS which is no longer required. ! 6.2.4 Containment Isolation System

This subsection is no longer applicable as it deals with the primary containment isolation system. The primary containment is , no longer maintained. j i 6.2.5 Combustible Gas Control in Containment l This subsection is no longe'r applicable as it is concerned with , hydrogen combustion inside the prin ary containment. l i l 6.3 EMERGENCY CORE COOLING SYSTEMS l The emergency 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. r 6.3.2.2.3 Core Spray System l 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. : s/ l 6-2 July 1991 - Rev. 3 l

SHOREHAM DSAR J

                            '6. 4 EABITABILITY SYSTEMS                                                      !

The systems, aside from the- control room air conditioning i' 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 f I

                            .The main ' steam isolation valve-leakage control system (MSIV-LCS)              I is not required in the defueled state and is, therefore, not included in the DSAR.                                                          '

c 6.6 OVERPRESSURIZATION PROTECTION l' 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 i 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                   1 defueled state and is, therefore, not included in the DSAR (See Chapter 4 of DSAR) .                                                          l 6.9 CONTROL ROD VELOCITY LIMITERS                                            ,

The control rod velocity limiters are not required in the j defueled state and this -Section is, therefore, not included in < the DSAR (See Chapter- 5 of DSAR) . f i 6.10 MAIN STEAM LINE FLOW RESTRICTORS t L The main steam line flow- restrictors are not required in the L defueled state and this Section is, therefore not included in the ; l DSAR, (See Chapter 5. of DSAR) . [ I I

6.11 REACTOR CORE ISOLATION COOLING SYSTEM The ECIC system is not -required in the defueled state and is, therefore, not included in the DSAR (See Chapter 5. of DSAR) . { i i t 6-3 Rev. 3 July 1991 l

SHOREHAM DSAR IV 6.12 STANDBY LIQUID CONTROL SYFTEM 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). i l l l C) 6-4

SHOREHAM DSAR CEAPTER 7 INSTRUMENTATION AND_ CONTROLS

7.1 INTRODUCTION

This chapter presents the details of the control and instrmnentation systems in the plant except radiation monitoring systems, and electrical power systems which are described in Chapters B ,11, and 12. 7.1.1 Identification and Classification All of the instrumentation and control systems listed below which were classified in the USAR as Q. A. Category I have been reclassified as Q. A. Category II and no longer have a safety function.. USAR Figures 7.1.1-1 and 7.1.1-2 are no longer applicable due to the 3efueled status of the plant. Chapter 3 provides the reclassification of systems and components. 1.1.1.1 Identification of Individual Systems This section identifies the individual systems which are retained operable or functional to meet the requirements of the defueled plant. 7.1.1.1.1 Reactor Protection System

This system is not needed to support the storage of the fuel in the fuel pool. It is not included in the DSAR. 7.1.1.1. 2 Nuclear Steam Supply Shutoff 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. 7.1.1.1.3 Emergency Core Cooling System

        'This system is not needed to support the storage of the fuel in the fuel pool. It is not included in the DSAR.

7.1.1.1. 4 Neutron Monitoring System l This system is not needed to support the storage of the fuel in the fuel pool. It is not included in the DSAR. 7.1.1.1. 5 Refueling Interlocks This system is not needed to support the storage of the fuel in the fuel pool. It is not included in the DSAR. O 7-1 Rev. 3 July 1991

i SHOREHAM DSAR i 1 l ((/~~Ty 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 Instrumentation 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, therefore 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 Regulator and Turbine-Generator Controls This system is not needed to support the storage of the fuel in the fuel pool, therefore it is not included in the DSAR. 7.1.1.1.11 Remote Shutdown System (/C } 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 Pumphouse Ventilation System The screenwell pumphouse ventilation system instrumentation and controls remain functional and are designed to ventilate each 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 Cooling System This system is not needed to support thb storage of the fuel in the fuel pool, therefore it is not included in the DSAR.

      ., s 7-2                Rev. 3 July 1991 l

l SHORERAM DSAR 7.1.1.1.15 Standby Liquid 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. 7.1.1.1.16 Reactor Water Cleanup System This system is not needed to rupport the storage of the fuel in the fuel pool, therefore it is not included in the DSAR. . 7.1.1.1.17 Leakage Detection 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. 7.1.1.1.18 Reactor Shutdown Cooling Mode-RHR 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. 7.1.1.1.19 Radwaste System Radwaste system instrumentation and controls support manual processing and disposing of the radioactive process wastes, k) 7.1.1.1.20 Emergency Diesel Generators This system is utilized to provide backup emergency power. One emergency diesel generator will be operable when fuel is being handled in the secondary containment. 7.1.1.1.21 Turbine Building Closed Loop Cooling Water Systela 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 monitor system performance. The TBCLCW system also cools the equipment in the radwaste building and supports the station air l compre ssor s. 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 in accordance with Section 9.2. 7.1.1.1.23 Recirculation Pump Trip 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. l 7-3 Rev. 3 July 1991

j SilOREHAM DSAR k) 7.1.1.1.24 Reactor Building Standby Ventilation System The filtration portion of the system is not needed to support the storage of the fuel in the fuel pool. Certain fans and air operated valves will remain functional to support RBNVS operation. See DSAR section 9.4 for additional information. 7.1.1.1.25 Peactor Building closed _ Loop cooling 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. l 7.1.1.1.26 Primary Containment Atmospheric 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. l 7.1.1.1.27 Fuel Pool Coolin,g_and Cleanup Systems Fuel pool cooling and cleanup systems instrumentation and controls remain unchanged except that the cooling portion is not required because evaporative cooling is suf ficient to remove the small amount of decay heat. 7.1.1.1.28 Co_ntrol Room Air Condition 1no 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 f an. 7.1.1.1.29 Chiller Equipment 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 Emergency 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 Relay Room, Emergency Switchgear Rooms, And Computer Room Air Conditioning System 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 automatically control the ventilation system to maintain these p( g) rooms at their design temperature and system performance. 7-4 Rev. 3 July 1991 l

SHOREHAM DSAR l gr^N

  '--  7.1.1.1.32   Battery Room Ventilation System

( The battery room ventilation system instrumentation and controls ! automatically control and monitor the ventilation system to l 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. l 7.1.1.1.33 Containment Spray and Suppression Pool Cooling This system is not needed to support the storage of the fuel in i the fuel pool, therefore it is not included in the DSAR. 1 l 7.1.1,1.34 Rod Sequence 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. l 7.1.1.1.35 Motor Control Center Room Ventilation System 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. (/'~T Q Motor Generator Room Ventilation System

i 7.1.1.1.36 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 ventilaf.lon 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. l 7.1.1.1.38 Main Steam Isolation Valve Leakage Contrcl 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. l 7.1.1.2 Classification Section 3.2 provides a reclassification of systems based on their importance to safety. l ps C 7-5 Rev. 3 July 1991

1 SHOREHAM DSAR f

     ) 7.1.2    Identificacion of Safety Design Bases and Nonsafety Design Bases Criteria The following sections remain as stated in the USAR except that the systems are now reclassified as OA Category II per Chapter 3 and the safety design bases no longer apply:

7.1.2.1.12 Screenwell Pumphouse Ventilation System l 7.1.2.1.19 Radwaste System 7.1.2.1.20 Emergency Diesel Generators 7.1.2.1.21 TBCLCW System 7.1.2.1.22 Service Water System 7.1.2.1.27 Fuel Pool Cooling and Cleanup System The cooling portion of this system is not required for the defueled plant. 7.1.2.1.28 Control Room Air Conditioning System 7.1.2.1.29 Chiller Equipment Room Ventilation System i f-w 7.1.2.1.30 Diesel Generator Room Emergency Ventilation System l 7.1.2.1.31 Relay Room, Emergency Switchaear Room, and Computer Room Air Conditioning System 7.1.2.1.32 , Battery Room Ventilation System 7.1.2.1.35 Motor Control Center Room Ventilation System 7.1.2.1.36 Motor Generator Room Ventilation System All other systems listed in subsections of the USAR under 7.1.2 are not needed. 7.2 REACTOR PROTECTION SYSTEM This section is not needed to support the storage of the fuel in the fuel pool. 7.3 ENGINEERED S AFETY FEATURE SYSTEM This section is not needed to support the storage of the fuel in the fuel pool.

   'tO V

7-6 Rev. 3 July 1991 {

SHORERAM DSAR h 7.4 SYSTEMS REQUIRED FOR SAFE SHUTDOh'N This section is not needed to s'tpport the storage of the fuel in, l the fuel pool. 7.5 SAFETY RELATED DISPLAY INSTRUMENTATION l This section is not needed to support the storage of the fuel in the fuel pool. l 7.6 ALL OTHER INSTRUMENTATION SYSTEMS REQUIRED FOR SAFETY The following sections remain as stated in the USAR with the exception that safety related systems have been reclassified to nonsafety related systems: 7.6.1 Description only the cleanup portion of the fuel pool cooling and cleanup l system is required. 7.6.1.2 Fuel Pool Cooling and Cleanup System Instrumentation and Controls l (("T The cooling portion of this system is no longer required because (_/ the transferred fuel in the pool has minimal decay heat and therefore active cooling and circulation of water in the spent fuel pool is not required. Fuel Pool Level Instrumentation: The spent fuel pool level is indicated andThe alarmed purpose on of high and this low levels in the main control room. instrumentation is to ensure that the water level in the spent fuel pool is maintained at sufficient height to provide shielding for normal building occupancy. If the low level alarm annunciates, the control room operator will notify the fuel handling personnel to evacuate. To ensure that the refueling floor personnel know what the radiation levels are on the refueling floor, three area radiation monitors are provided and are set to alarm at 5 mr/hr. Fuel Pool Temperature Instrumentation The spent fuel pool temperature is indicated and alarmed on high temperature in the main control room. The purpose of this instrument is to ensure that the maximum bulk pool temperature does not exceed 125'F design temperature. Based on the low fuel heat load it is not expected that the pool could reach this

 ;m       temperature.

s. 7-7

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

s d 4 i SHORERAM DSAR .D's/. 7.6.2.2.1 General Functional Requirements

          . Conformance of Fuel Pool Cooling and Cleanup System i

4 Instrumentation and Controls - . All other USAR Part 7.6 sections not listed above are not needed ,l in the defueled condition. f 7.7 - CONTROL SYSTEMS NOT REQUIRED FOR SAFEIT i The following sections remain as stated in the USAR: 7.7.1.6 Liquid Radwaste Control System Instrumentation and Controls (since limited power generation has created 3 radioactive waste). f '7.7.1.7 Turbine Building Closed Loop Cooling Water System Instrumentation and Controls I [ 7.7.1.11 Refueling Interlocks Instrumentation and Controls

            .7.7.2.6.1         General Functional Requirements Conformance for Liquid j'~

Radwaste System Instrumentation and Controls i

7. 7. 2. 7 .1 General-Functional Requirements Conformance for Turbine Building Closed Loop Cooling Water System i ke"N)- Instrumentation and-Controls

{' I 7.7.2.11.1 General Functional Requirements Conformance for Refueling Interlocks Instrumentation and Centrols

             - All other USAR Part 7.7 sections not listed above are not needed.

7A Plant Nuclear Safety Operational Analysis This section is not needed to support-the storage of the fuel in the. fuel pool, i-7B Analog Transmitter / Trip Unit System for Engineered Safeguard Sensor Trip Units This section is not needed to support the storage of the fuel in the fuel pool. O 7-8 Rev. 3 July 1991

1 SHOREHAM DSAR

 ,a 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

 /%   The description contained under this heading in the latest N)      revision of the Shoreham USAR remains unchanged in the defueled condition. For further information on this subject refer to the U S AR.

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 f rom the LILCO system through 13BKV or 69KV circuits. The 138KV switchyard is arranged in a two bus configuration with circuit breakers and switches arranged Four to permit isolation and/or repair of either bus section. 138KV circuits enter into the switchyard (two per bus) each containing a circuit breaker at the connection to its respective 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 theThis aforementioned circuit, however, rights-of-way for a distance of one mile. 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: Three Brookhaven 80MW (each) Combustion Turbine units are located on LILCO SNPS property approximately 3600 feet from the 138KV g- switchyard. These units are connected into one of the 138KV k-]J 8-1 Rev. 3 July 1991

4 SHORERAM DSAR h Holbrook tranrmission-lines and -are available to provide an ! additional source of onsite power to the SNPS. (see figure j 8.2.1-2) t The spare Reserve Station Service and Normal Station Service i transformers will no longer be required. S.I.4 On Site AC Power System j The station electrical power system includes ' electrical equipment and connections required to provide power to and control the ! operation of electrically driven station equipment in the i defueled condition. A non-safety emergency diesel generator will I provide backup AC power during fuel handling in the secondary l containment. B.1.5 On Site DC Power System e During the defueled condition, the 125V DC distribution systems do not have a safety function. However, a DC distribution system f will be maintained operable during fuel handling operations. It ; will remain functional at other times. [ t

 .gs    The 24V DC power source will no longer be required. This system prowides power to the Nuclear Source and Intermediate Range

((_) Instrumentation which is no longer in service in the defueled i condition. l S.I.6 Identification of Safety Related Systems t The description contained under this heading in the latest revision of the Shoreham USAR will not be applicable in the : defueled state. Tabl.e : 8.1.6-1 Identification of Safety Loads t The basis for these tabulations no longer exists. The electrical ! distribution system will remain in service to maintain power to l plant equipment on the site in the defueled condition. l 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. l I 8-2 Rev. 3 July 1991 l v

                          *-                              .n .__.       _
             ~~'

n.

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

i i SHOREHAM DSAR j Table 8.1.7-1 Regulatory Design Criteria For Electric Power The basis for these tabulations no longer exists. The electrical l distribution system will remain in service-to maintain power to- l plant equipment on the site in the defueled condition. j 8.2 OFFSITE POWER SYSTEM j 8.2.1 Description The description contained under this heading in the latest l revision of the Shoreham USAR remain unchanged except as follows: l Service-buses 101, 102 and 103 are not required to be maintained l ! as safety related while in the defueled condition. They are : reclassified as Category II. i i s 8.2.1.1 One Line Diagrams and Physical Drawings The information contained under this heading in the latest I revision of the Shoreham USAR remains unchanged in the def ueled { condition. l ! ! 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 'l l except that the safety related function of the busses !

(1R22-SWG-101, 102, and 103) no longer exists. They-are l l reclassified as O.A. Category II systems. l ;

i 8.2.1.3 Station Switchyard ; i lThe description contained under this heading in the latest ! I revision of the USAR remains unchanged in the defueled condition. { For further information on this_ subject refer to the USAR. ; I ' 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 j except for the following: I f l I O l l l 8-3 Rev. 3 July 1991 1 I t ;

                                              - +--              w        ,.- . -     r,,

SHOREllAM DSAR O k# 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 described in the USAR is not requi .3 in the defueled condition. 8.3 Onsite Power Systems The plant power system is designed to provide an adequate source of electrical power to all systems tequired to be operational in the defueled condition. B.3.1 AC Power Systems The general description of the pl. ant electrical power (AC) systems is as provided in this section of the USAR. However, the safety related design criterie 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 (~D for the diesel to be classified as operable. l\-] 2- One diesel generator set shall remain operable during fuel l handling in the secondhry containment. 3- Required surveillances and tests will be performed in accordance with the Technical Specifications. 4- Adequate equipment protection and emergency measures are available for the regoired plant electrical systems in the defueled condition. The equipment, switchgear, and buses have been reclassified to Q.A. Category II. Therefore, safety functions such as auto-start, redundancy, etc., are no longer required. 8.3.2 DC Power Systems 8.3.2.1 Description The description contained under this heading in the latest revis2on of the Shoreham USAR remains unchanged in the defuelec condition except as follows: 1- The 24V DC system, providing power to source and intermediate range nuclear instrumentation, is no longer used. (-}

  %/       2-  A14 class 1E/ safety related functions of the DC system are no longer classified as such.

8-4 Rev. 3 July 1991

SHOREHAM DSAR (!O

 '-   The batteries are being maintained for those systems0.A.

functional or operable in the defueled condition. remaining Category l I equipment is now Q.A. Category II. I O fV : i

f\h 8-4A Rev. 3 July 1991

. _ . - . _ .-. -~ , -. .. -. _ .. . .. . . . ~ .. .. .- . I i t SHOREHAM DSAR i i 5m t For further information on this subject, refer to the USAR. Tables - 8.3.1.1, thru table 8.3.1.7A - related Emergency Diesel ; Generator System loads, demands, sequencing and margin test results are no longer applicable in the defueled condition. l l Tables - 8.3.2.1 and 8.3.2.2 - related to plant design basis loads are no longer applicable in che defueled condition. . I

                                                                                                                                                        .l i

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    -                                                  SHOREHAM DSAR                                                                                         ?
                                                                                                                                                             ?

CHAPTER 9 f AUXILIARY SYSTEMS i 9.1 FUEL STORAGE AND HANDLING

      ,9.1.1                     New Fuel Stot&qe.                                               .

l Since no new fuel will-be received, this section of the USAR is not required in the defueled condition. , i

        -9.1.2                    Spent Fuel Storage                                                                                                         :
                                                                                                                                                             ?

9.1.2.1 Design Bases , i

1. Spent fuel storage space is designed to accommodate 2,176 fuel assemblies (390 percent of the full core fuel load);

however, currently only 1,420 storage cavities have been : installed and are available to receive spent fuel assemblies ; (See Figure - 9.1.2-1 revised for DSAR) . The remainder of this Section, paragraphs 9.1.2.1.2 through , 9.1.2.1.10,- is identical to the USAR. j j} f 9.1.2.2 Facilities Description i

Spent fuel storage racks provide a place in the spent fuel The pool for storing spent fuel received from the reactor vessel. : location of .the spent fuel pool within the reactor buf1 ding is i shown on Figure 3.8.1-4. The general arrangement of the storage I space, illustrated on DSAR Figure 9.1.2-1, will permit the i storage.of 2,176 fuel assemblies (the current installed capacity in the spent fuel pool 11s for 1,420 fuel assemblies) plus 144 l control rods. The-remainder of this Section is identical From to that in the USAR, this point to the up to'the first paragraph of page 9.1-8. ; end of page 10, the text is deleted. 9.1.2.3 Safety Evaluation l j This section remains identical to that in the dSAR except that in l the DSAR Appendix 9A provides the criticality analysis. . 9.1.2.4- Tests and Inspection , i The_ description contained under this heading in the latest revision of Shoreham USAR remains unchanged. Refer to the USAR l p j(,) for information on this-subject. { 9-1 1

SHOREHAM DSAR I f k(N- -9.1.2.5 Radiologiccl Considerations i The description contained-under this heading in the latest .

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

t 9.1.3 Fuel Pool Cooling and Cleanup System All of the equipment in this system will be retained for Since the fuel pool cooling l operation, but in a modified manner. l subsystem is designed to remove the decay heat produced by spent fuel assemblies, as described in the USAR, and only a negligible i amount of heat is expected to be generated from the slightly irradiated spent fuel bundles stored there, the cooling mode is i not required Thus reactor building closed loop cooling water is ! l not remd" . 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 i fuel pool from the condensate. storage tank using the condensate transfer and storage system. Alternate makeup sources for the i'

  +-      spent fuel pool are Demineralized and Makeup Water System, Fire k ,')   Protection Water System, and the Service Water System. The makeup mode 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 i above elevation 168, or about 7 feet below the normal water level. If a break occurred in these lines, about 18 feet of 4 f water would remain above the fuel in the pool.- This is more than enough to provide adequate shielding. Pump returns to the pool i are equipped with siphon breakers to prevent inadvertant pool l drainage. l 9.1.4 Tuel Handling System ! 9.1.4.1 Design Basis : i See USAR. This section is identical to the USAR.

Equipment Description 9.1.4.2 : See USAR. This section is identical to the USAR. i 9-2 l [

. . ~ . - . .. - -.- - - . - -... - - - . - . - ~. . . . - . - - . - . - .. - - h r r SHOREHAM DSAR i r 9.1.4.3 - Description of Fuel Transf er ( The fuel handling system provides- a safe and effcetive means for . transporting and handling fuel from the time it reaches the plant ! until it 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 ensures that the design bases of the ! fuel handling system and the requirements of Regulatory Guide l l 1.13 are satisfied. , l

                                                                                                                                                                        -t 9.1.4.3.1                     Arrival of Fuel On Site No new fuel is expected to arrive on site.                                          Therefore this                                              j section of the-USAR is not required.                                                                                                            l 9.1. 4. 3 . 2                 Refueling Procedure t

No refueling is planned. Therefore this section of the USAR is ; not required. l 9.1. 4 . 3 . 3 Departure of Fuel from Site This section applies as written in the USAR. . () In addition: l l

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

I;

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 i I 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 i reduced heat removal requirements with the plant in the de-fueled l state. i a) The RBSW system is considered non-safety related because it -i does not provide cooling-water to any plant equipment required to perform a safety function. ! i b) One RBSW pump.will supply cooling water to one RBSVS/CRAC chiller condenser, an emergency diesel-generator, and to all j Turbine Building Service Water (TBSW) cooling loads. (See l () item e below.) No service water is required for RHR, RBCLCW, drywell cooling, and makeup water to the reactor vessel l t 9-3 Rev. 3 July 1991

                                                    ,       - - - , . .-         _, -,         , ,_. .--- .r,  - - . _ _ . . y,_m.,. ., ,-            . . _ . , . _ _ ,

SHORERAM DSAR ; Y 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 one 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. I f) The TBSW pumps are out of service since they are no longer ! required. f._s g) Table 9.2.1-1 has been revised. N- 9.2.1.5 Instrumentation Application 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 Reactor Building Closed Loop Cooling Water (RBCLCW) System This system is not needed to support the storage of fuel in the ' spent fuel pool. r 9.2.3 Makeup water Demineralirer System The description contained under this heading in the latest revision of the Shoreham USAR remains unchanged except as follows:

1. SBLC, RBCLCW, seal water injection, and vacuum priming are no longer users of demineralized water in the defueled condition.

2. The HPCI suction line from the condensate stcrage tank is not required to be maintained as safety related in the defueled

( condition. r~T l

9-4 Rev. 3 July 1991 l I

SHORERAM DSAR O' For further information relative to this system refer to the USAR. , 9.2.4 Potable and Sanitary Water Systems 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 l USAR. , Ultimate Heat Sink i 9.2.5 With the fuel in the Spent Fuel Pool, the ultimate heat sink i (Long Island Sound) no longer has any safety significance, since the decay heat of the fuel is insignificant. However, the ultimate heat sink will continue to be used as a source of ' cooling water for normal plant needs (refer to DSAR Section 9.2.1). , 9.2.6 Condensate Storage Facilities While in the defueled condit. on the condensate storage f acilities provide makeup water for the fuel storage pool. The description of this system in the USAR remains unchanged except as follows:

1. Condensate, feedwater, reactor systems, HPCI and RCIC will no longer be users.

2. HPCI test discharge and CRD pump return lines to the CST are ;

not required to be active.

3. The first three paragraphs of USAR 9.2.6.3 are no longer applicable. [

4. The last paragraph of USAR 9.2.6.4 and 9.2.6.5 are no longer applicable.

9.2.7 Turbine Building Closed Loop Cooling Water System The description contained under this heading in the latest revision of the Shoreham USAR remains unchanged except that DSAR Table 9.2.7-1 lists the coolers still receiving cooling water flow while the plant is in the defueled state. For further information on this subject refer to the USAR. 9.2.8 Main Chilled Water System This system will not be maintained as an operable system since it is not neede? with the plant in the defueled condition. 9-5 Rev. 3 July 1991

i SHORERAM DSAR ! 9.2.9 Reactor Building Standby Ventilation And Control Room Air Conditioning Chilled Water System Redundancy in this system is not needed sir.ce neither the RBSVS nor CRAC systems are safety related in the defueled condition. The heat loads generated by the electrical equipment in tha 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: rg 1. Piping that has been installed as ASHE III code class 2 is no K/ longer considered safety related and is reclassified QA Category II. l ,

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 the USAR. 9.3.2 Process Sampling System The Process Sampling System provides monitoring of certain process operations while fuel is in the-spent fuel pool for either short or long term storage. The process monitoring is accomplished as necessary by means-of measuring, analyzing and/or recording for conductivity, pH, and silica concentration, as shown on DSAR Table 9.3.2-1. 9.3.3 Equipment and Floor Drainage System With the Reactor defueled and the fuel assemblies stored in the Fuel Pool, large portions of the Equipment and Floor Drainage O System are not required. 9-6 Rev. 3 July 1991

SHOREKAM DSAR

) ) ' System Description                                                    -

This system is described in the USAR. Changes in status are addressed below. Raactor Building 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 Building 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 n,o steam and the turbine is no longer-required, so that the only source of radioactive waste is the Chemical Laboratory. Radwaste Building The Radwaste Building Equipment and Floor Drainage System is {) maintained operational. The Dirty Waste Sump and Pumps (1N52-TK 114.and 1N52-P-187A/B) and Regenerant-Recovery Sump and Pumps (1N52-TK-115 and 1N52-P-181A/B) are no longer required. 9.3.4 Chemical, Volume Control, and Liquid Poison Systems The Standby Liquid Control System 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 description in the USAR Section is no longer applicable. In the eventLof 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 Suppression Pool Pumpback System This system not required to support storage of fuel in the fuel pool. 9.4 AIR CONDITIONING, HEATING, COOLING, AND VENTILATION SYSTEMS 9-7 Rev. 1 Aug. 1990

i

                                                                                                                                 ,i SIDRERAM DSAR 9.4.1       Control Room Air Conditioning-System The Control Room AC system remains unchanged in design and i

operating functions. However, the system is reclassified to QA ; Category II, the filter portion of the system will no longer be l required and one of each of the redundant f ans and ACUS will no ! longer be required. The AC system will only function to provide i an OSRA environment for the operators during the fuel storage j period. This requires the operation of only one RBSVS/CRAC l chiller. Automatic initiation systems and interlocks for the habitability portion of the system will be non-operable and-the l ! 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 Building Normal Ventilation System { 9.4.2.1 Design Basis The RBNVS remains unchanged in design and operating function except that the system will only: t

1. Provide ventilation by introducing filtered outside a!r into I j

the reactor building at a rate of approximately 2.7 air i changes per hour ; O- 2. Remove heat generated by solar and external heat tr ansmi s sion, lighting and 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. i The RBNVS may be operated in a recirculation mode in order to f; control. Reactor Building humidity. This helps to protect i equipment from damage due to corrosion. While operating in the recirculation mode, the operating functions, as discussed above, f are maintained except that supply air is limited to infiltration caused by the negative-pressure inside the Reactor Building. t For further discussion on this system refer to the USAR. l l 9.4.3 Radwaste Building Ventilation The description contained under this heading in the latest [ Shoreham USAR remains unchanged, except that the charcoal exhaust f 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. [ l l 9-8 Rev. 3 July 1991 [ i

l SilOREllAM DSAR O 9.4.4 Turbine Building Ventilation System And Station txhaust System A) Turbine. Building Ventilation system This system is not required to support the storage of fuel in the spent fuel pool. O i l l O i 9-8A Rev. 3 July 1991 _ . .. _ _ _ .._ __._... _ _.. _ .._.--. . _ . _ __ _. _ .-._.,__--._.- _ _ -. _ , _._ . ~ _ _ .-._- . _ --

I SHOREllAM DSAR B) Station Exhaust System : This system will expel the exhaust air from the radwaste building and the reactor building. However, only two fans will be i required for this purpose, one f an operating and one f an on i standby. This will ensure that the Isokinetic nozzles located in the upper level of the exhaust duct will see a suf ficiently high , velocity to be operational. For further discussion regarding , this system refer to the Shoreham USAR. 9.4.5 Battery Room Heating And_yentilation 3 t The description contained under this heading in the latest revision of Shoreham USAR remains unchanged. Refer to USAR for infonmation on this subject. This system is reclassified to Q.A. l Category II. l l 9.4.6 Drywell Air Cooling System j t Thas system is not needed while the fuel is stored in the spent l i fuel pool. 9.4.7 Screenwell Pump House Heating Ard Ventilation :

 ;()    The description contained under this heading in the latest revision of Shoreham USAR remains unchanged. Refer to USAR for f

l information on this subject. This sytem is reclassified to Q A. Category II. l 9.4.8 Plant Heating 7 The description contained under this heading in the latest revision of Shoreham USAR remains unchanged. Refer to USAR for ! i l information on this subject. 9.4.9 Primary Containment Purge System This system is not needed while the fuel is stored in toe spent i fuel pool, f 1 9.4.10 Diesel Generator Room Ventilation t The description contained under this heading in the latest revision of the Shoreham USAR is revised. This system is , reclassified to Q. A. Category II and nonseismic. The system is no longer safety related and the design bases for tornado missile [ protection and room temperature control are no longer applicable, j 9.4.11 Relay Room, Emergency Switchgea.r Room And l Computer Room Air Conditioning System The description contained under this heading in the latest revision of Shoreham USAR remains unchanged with the 9xception that only one train of equipment will remain functional. Refer i

9-9 - -Rev.-3-July 1991 ,

SHOREHAM DSAR O to USAR for information on this subject. This system is reclassified to QA Category II. lO l lO l

                                            . . _ .._~. _~ _ .--..._. -_ _ .-.9,_ .- _ ._ . _ _-.._,.__ _ _

- - - . - - -__ - _ _ - - - - - ~ _ - _ . _ _ - - _ - _ - - - - SHOREHAM DSAR i. I 9.5 OTHER AUXILIARY SYSTEMS f Tire Protection System

                                                                                                                                                    ,j 9.5.1                                                                                                                         i 9.5.1.1          Design Basis l

The design basis section applies with the following addition l The basic premise of the fire protection discussions in the USAR ; and THAR is protection from fire for safety related areas (1) ; including areas containing equipment or circuits that are i required for, safe shutdown, or (2) required to prevent or '

                       ' mitigate radiological releases comparable to 10CTR 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. 7 Structures, systems components and administrative controls in place to protect areas, equipment or circuits previously l* 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 ' protection of non-safety related areas. i Three documents which were used in the design of the plant's fire i protection features and continue to be part of the fire j protection program are: Evaluation of the SNPS Tire Protection Program as compared to 1. 10CTR50, Appendix R criteria submitted via SNRC 572 dated May [ 21, 1981,

2. Fire Hazards Analysis Report. \

h

3. Cable Separation Analysis Report: !

SHRC 532 dated February 10, 1981 SNRC Bil dated April 13, 1983 lt However, the term " safety related", as used in those documents and in USAR section 9.5.1, applies only to the Reactor Building. l i Section 6 of the Fire Hazards Analysis Report (THAR) contains technical requirements that formerly were fire protection j technical specifications, , THAR Chapter 6 reflects reductions in the technical requirements i that are consistent with the text of this DSAR Section 9.5.1. 1 r Types of Fires The " types of fires" section applies with no changes. Design Criteria 9-10 Rev. 1 Aug. 1990 t i

SHOREHAM DSAR r The ' design criteria" section applies with the following addition: As discussed _above, this design will be maintained for property loss prevention r>.rposes. However, the " safety related" application of the listed documents, particularly HRC'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 0. A. Category II. The rooms listed as examples of where oil fires l could occur near safety related equipment no longer fit that description because these areas are reclassified to QA Category II. Furthermore, the fire hazard associated with this equipment l 1s significantly reduced while the equipment is not being used because the ignition sources associated with the operating equipment have been eliminated. ( Intensity of fires This section applies without change. Fire Characteristics This sec+. ion applies without change. Building Arrangement and Structural reatures The ' building arrangement and structural features" section applies with the following changes: In the response to NRC question 3, as shown in FRAR revision 3, SNPS has stated our intention to replace existing motorized fire dampers with newly designed fire dampers. All of the areas where these new dampers were to be installed are in the Control Building and are reclassified to Q.A. Category II. Therefore, The CO l this proposed modification will not be implemented.Whenaftheis systems for those rooms are in electric lockout. detected, the CO, system controls would cause the dampers to close on an electrical signal. As a backup, the fusibic link of each of the existing fire dampers is sufficient to cause closure of a damper in the event of a fire, thus assuring integrity of the fire barriers. Rev. 3 9-11 July 1991

SHOREHAM DSAR 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. This deviation was reported to the NRC on Licensee Event Report 87-021. The proposed corrective action was to install a deluge water curtain system below the existing missile shield wall between the transformer and the wall openings. Since the diesel generator rooms are reclassifisd to QA Category II, l this modification will not be implemented. The partial protection provided by the missile barrier is considered sufficient for non-safety related areas. Seismie Design This section applies without change. Water Recuirements The " water requirements" section applies with the f ollowing additional statement: Although some areas previously identified as safety related are reclassified to QA Category II, the water supply is not being l reduced. Codes and Standards This section applies without changes. SNPS will continue to meet the requirements of the applicable NTPA 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 reclassified to Q.A. Category II. Modifications that would l degrade its seismic design are not contemplated at this time. 9.5.1.3 Safety Evaluation Electrical Insulation Fires ' This section applies without change. I l 9-12 Rev. 3 July 1991 l

SHORERAM DSAR 4 Charcoal Tires This section applies without change. . Oil Fires The

oil fires" section of the safety evaluation applies with the following change As discussed earlier, the fire hazards associated with non-operating equipment are significantly reduced because the primary ignition sources - electrical energy and hot surfaces -

                 .are eliminated.

Severity, Intensity and Duration of Tires This section applies without changes. Time Estimates This section applies without changes. Failure Mode and Effects Analysis This section applies without changes. Accidental Initiation of Fire Protection System The

accidental initiation of fire protection system" section applies with the following change:

Areas protected by CO 3 systems are among those thtt are no longer considered safety related. Sinole railure in rire Protection Systems This section applies without change. Pipe Breaks in Tire Protection Systems This section applies without changes. Tailure of Fire Protection System Affecting Safety Related Eculpment This section applies with the following change: Of the areas listed, only the Reactor Building is still considered safety related. (O _/ Aug. 1990 9-13 Rev. 1

g SHORE!1AM DSAR

      -~            9.5.1.4         Tests and Inspections
   \-~#            This section applies without changes.                                                          ,

9.5.1.5 Personnel Qualification and Training This section applies without changes. 9.5.2 Communications System 9.5.2.1 Design Bases

                   *Thic section of the USAR remains unchanged.

9.5.2.2 System Description This section of the USAR remains unchanged except for the following:

1. For the very low frequency (VLT) portable radio systems, one low-powered VLF radio base station will be used in conjunction with two mobile car units to provide offsite radio communications (instead of two VLF base stations and four mobile car units).

2. The Emergency operations racility (Eor) is not required, f

since no emergency requiring EOF activation can occur with 6 the fuel in the Spent ruel Pool. 9.5.2.3 Tests and Inspections This section of the USAR remains unchanged. 9.5.3 Lighting Syst2ms While in the defueled condition this system will provide all the necessary required lighting to the plant and the site. The description of this system in the USAR remains unchanged except for the following:

1. Section 9.5.3.2, item (2 - the standby AC lighting system will receive power from plant service busas which are powered from offsite.

2. Same section, item 45 - the fifth lighting subsystem will receive power from DC battery cources while the plant remains in the defueled condition.

3. The last paragraph of the same rection, the independent power sources for lighting, remains unchanged but the source of power will be from plant service buses and DC battery sources if needed.

9-14 Rev. 1 Aug. 1990 i

t I r SHOREHAM DSAR i 9.5.4 Diesel Generator Fuel Oil Storage ano Transfer System ( An emergency diesel generator is required to be operational when ! irradiated fuel is being handled in the reactor building. ? Sections 9.5.4-9.5.7 in the USAR remain descriptive of the EDG auxiliary systems except that these systems and their components i are classified as QA Category II. Also, the requirements of redundancy to prevent malfunction or failure of these systems and their components, i.e., fuel storage tanks, fuel pumps, air start t tanks, etc. and 7-Day operability Post-LOCA are no longer j applicable. ! Statements in Sections 9.5.4-9.5.7 indicating that portions of these systems are designed to ASMI Boiler Pressure Veseel Code, l Section III, Code Class 3, that they meet Seismic Category I requirements, and 't the concrete block house and the two-foot l thick concrete sla' . cove the fuel storage tanks are Scismic j Cate90ry I and provide missile protection are also no longer applicable. The USAR description of equipment desisn with ; j respect to applicable codes is representative of the original _ design of these systems but these designs whieu were applicable t j to safety related equipment in an operating nucinar power plant will no longer be maintained as safety related equipment, baced r [ on the DSAR Chapter 15 safety an* qais. , Diesel Generator Cooling Water System ; 9.5.5 _. 9.5.6 Diesel Generator Starting System ,

9. 5.7 Diesel Generator Lubrication Systam 4

9.5.8 Primary containment Leakage Monitoring System s With the fuel in the Spent ruel Pool, the Primary Containment ] Leakage Monitoring System is not required, i t 9.5.9 Storage of Gases Under Pressure The quantities and type of gases stored in pressurized containers ! in the defueled condition is reduced from that previously on . The design bases remain unchanged. Storage facilities are l hand. provided for the following gases as shown in Table 9.5.9-1: ,

1. Carbon Dioxide for fire protection. !

2. Halon 1301 for fire protection.

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

4. Nitrogen for glycol and HW heating. l

5. Propane for auxiliary boiler ignition.

l I 9-15 Rev. 3 July 1991

SHOREHAM DSAR (- The following gases are no longer used or required to be stored in the defueled condition:

1. Hydrogen for main generator.

2. Rydrogen and oxygen for gas analyzers.

3. Nitrogen for containment inerting.

4. Nitrogen for drywell floor seals.

5. Nitrogen for electrohydraulic control.

6. Air for MSIV accumulators (inboard and outboard).

7. Air for long term accumulators.

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 nLe for-safe shutdown is no longer necessary in the defueled

!                            condition.
       'O 9

b i e i I i. [ l I L !

9-15A -- ---S.ev 4 July 19 91 -

SHOREHAM DSAR () Appendix 9A TUEL CRITICALITY ANALisis The Shoreham Spent Fuel Rack (STR) is of a stainless steel and water neutron flux trap design which uses no additional poison. A description of the storage racks is provided in 9.1.2. The criticality analysis of this rack design is described in detail The reactivity results in Appendix 9A of the Shoreham USAR, which are summarized in USAR Table 9A-4 remain valid for the rurthermore, conditions existing at Shoreham after defueling,

              .due to the differences in U-235 enrichment between the SrR designed and the current Shoreham fuel, a large negative reactivity credit should be taken into account. This is explained as follows:

The Shoreham SrR design is based on a maximum U-235 enrichment of 3.1 wt. t. The resulting basic cell k is calculated to be 0.9129 without uncertainty and model adjustments (Table 9A-4, Appendix 9A, Shoreham USAR). The Shoreham Cycle 1 fuel loading uses three (3) enrichments. Cf the 560 fuel assemblies in the core, 340 bundles have the highest bundle average U-235 enrichment of 2.19 wt. 4, 144 bundles of 1.76 wt. t and 76 remaining bundles uses natural uranium. i If the six inch natural uranium segments at the top and bottom of the fuel are excluded, the average enrichment of a l i 2,19 wt. t bundle becomes 2.33 wt. 1. Using this enrichment and linearly extrapolating the reactivity vs. U-235 I enrichment results given in rigure 9A-5 of Appendix 9A, Shoreham USAR, the reactivity difference between the SrR design enrichment of 3.1 wt. t and the current maxiumum loading enrichment of 2.33 wt. t is found to be about -6.0% in k ( k -0.060). This brings the basic cell k under nominal storage conditions for the current fuel down to f 0.85, which is well below the regulatory acceptance criterion of k 0.95. All the corrective and uncertainty adjustments listed in Table 9A"4 of the Shoreham USAR remain applicabic. During the period from July, 1985 to June, 1987, Shoreham went through three separate stages of low power testing (less than 5% of rated power), which resulted in a total core exposure of approximately 48 mwd /MT as determined by a series of core-follow analyses. The net effect of the core exposure is a slight k) mainly due to the : decrease in reactivity ( -0.002 in i offsntting contributions from the formation of Sm-149 and the I slight depletion of the burnable Gd poison in the fuel bundles. : i In light of the large reactivity margin described previously (k '

0. 6 5) , no additional credit will be claimed here.

l I ' 9-16 Rev. 1 Aug. 1990 l l

t l SHORERAM DSAR f^ 9B IVALUATIO11 OT SPENT FUEL POOL MAKEUP REQUIREMEllTS An analysis was performed to determine the rate of water loss ' fram the spent fuel pool through evaporation under the worst case r,cenario described below. The following conservative assumptions ; are used in the analysis to maximize the calculated pool , evaporation rates !

1) The spent fuel pool temperature is 110*T. (

2) The ambient temperature above the spent fuel pool is :

        -     conservatively assumed to have zero relative humidity.                                                                                    l

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 i this very low maximum depletion rate, external cooling of the r spent fuel pool is not required. Technical Specifications l' require that the water level above the spent fuel be a least twenty-one feet. In addition, it should be noted that pool water level is alarmed in the control room and alarm response l procedures exist to provide appropriate operator action. i P r 1 l i l + l ! t t i 9-17 Rev. I Aug. 1990

                                                                                        -m-g   --     .,,-

t - -i.--- .- m-,

     --                                         .__--.9,                 _

s k i SE2.AM DSAR { TABLE 9.2.1-1 i ' SERVICE WATER S7FTEM COMPONENT DESIGN DATA l i- Nominal j Capacity Number of Components l1 Each- Utilized in I i I Component Quantity (gpm) pefueled Condition l 1 4 8,600 1 Reactor Building Service Water Purps Turbine Building Service Water Pumps 3 8,000 Reactor Building Subsystem Components: Reactor Building Service Water Strainers 4 250 1 1 700 1 l Diesel Generator Jacket Coolers 3 Residual Heat Removal Heat Exchangers 2 8,000 ---

     '      Reactor Building Closed Loop Cooling         2            6,370               ---

Water Heat Exchangers . 4 525 1 Reactor Building Standby Ventilation System Chiller Condensers Main Chilled Water System Chiller 3 1,500 --- Condensers 1 400 Drywell Cooling Booster Heat Exchangers 2 1,450 --- 1 of 2 Rev. 3 July,1991

I ( i l SMG DSAR i TABLE 9.2.1-1 ? i

          -j                                                                                           SERVICE WATER SYSTEM COMPONENT DESIGN DATA (Cont'd.1 l

1 Nominal ! Capacity Number of Compenents { Each- Utilized in Ouantity (qpel Defueled Condition i Component f ? Turbine Building Subsystem Components:

i. .

2 420 - i Turbine Building Service Water Strainers 4 6 Note 1 Cire Water Pump' Bearing Cooling i 1 185 -- l ! Fish Retention Pool 2 14,200 1 ^ Turbine Building Closed Loop Cooling Water lieat Exchangers 3 100 - Vacuum Priming Pumps Seal Water Coolers 1 l t i Note 1: One cire water pump bearing cooler will be needed if a circ water pump is used tc, provide dilution of chlorine. t, 1 i i i i 1 1 I b 1 Rev. 3 July 1991 2 of 2 r ,+-,,.,,ow--,,w,,.,,,- w. -, nn w n, . m .--w n--.,, ,v e - e m . --n,,, om.-,m.-,m,,wn,m.,,,, mm -n, e . - ,.,---,.--.,.,-e ,--...,,.,,-e..mnw-n.-w,.,-,-,..m,+---.,nn,,,-,nw

TABLE 9.2.7-1 LIST OF COOLERS BEING SERVICED BY TURBINE BUILDING CLOSED LOOP COOLING WATER SYSTEM Component coolers Quantity I Air compressors 3 l Sample temperature bath coolers 2 IAI Reactor feed pump turbine oil coolers 4 IAI Main turbine lube oil coolers 2 2 IAI Offgas vent coolers Waste evaporator overhead condenser I IAI AI Regenerant evaporator overhead condenser 1 Hechanical seal cooler heat exchangers on various 10 radwaste process pumps Office & Services Building chillers 2 l [\ (A) These systems are not active but flow to them is used to dissipate excess pump capacity (i.e. system balancing). O fd Rev. 3 July 1991

3 I I StoraWI DSMt { i TAB 1I 9.3.2-1 PrecrSS E*JTLDG SYSTm. l

                                                                                                                       %Te of Descrir*.ia. of_ samle                                                                                      Samle                            Pttriew E M SY5' TIN I

hogararant Liquid and OC, C441, Sanple regenerant i Evaporator Teod Tanks

Grab 11guld evaporator -l hecirmilation Line f W tanks for ;

pw. ass data i Discharae Waste Sanple OC, Grab Sanple dischartre i Tanks 7ecirculation Line vaste sanple tank for i process data j Waste Cc11ector Tanks CC, Grab Sarple vasta l Recirculatior. Line collector tank for l process data T1oor Drain Collector CC, Grab Sanple floor drain Tanks Ibeirculation Line collector tank for 3 process data Fe:cnery Sample 'Janks CC, G:.ab Sample floor drain ! Eecirculation Line collector tank for process data ladweste Dc:r.ineralizer Grab Deriteralizer ; efficiency l, Outlet Grab T11ter efficiency Radwaste filter Effluent Tinal Discharoe Sa pling Grab Process data, prior to discharge Point T1oor Drain Tilter Effluent Grab Process data Laundry Drain Tanks Grab Process data Discharoe ynn? DmIhuALI2fm syrrm Irdividual Det.ineralize.r Cr, Grab, Darineralizer IX efficiency ard makeup Effluents water quJLlity Dilute Ct.ustic for CC Caustic concentration EL9eneration O

w...,c,w,+y.m e---,we,m-,4,msene=,-+ge-myse,+w,am-"m,y-g,-,,my-t----ewtw &et=* e T 4PTv v vt - :rNt--7 vWt*,7 -We r v F 1 r a-ee y y

SHOPDPM DSAR TABLE 9.3.2-1 (cont'd) O PacCESS FATLING SYSToi , Type of . . Description of senple frole Punote . Dilute Acid for DC Acid carecentratico Rooereraticn Wasta Beoeraration Cril

pH of nonradioactive l Neutralizing Tank ard awfeneration wastes, Acid and Caustic Wasta prio to disc.happe ,

Sump Beturn fin root etTRUP WWD4 Dardneralizer Inlet . OC, Grab 3 dication of fuel pool water quality Derdneralizer outlet CC, Grab Domineralizer officiency AtTILIAW BoitrR sysTD4 Auxiliary Loiler Blewdcun Grab Water gaality, boiler performance Auxiliary Boller reed Grab Water quality data Peps Discharce Auxiliary Leiler Condensate Grab Cordensate quality Au):111ary Loiler Steam Grab Toiler perfontence data . Not Water for lieating Grab Water quality data COOLING hTTIR WEDS Turbine Ba11 ding Closed CC Cooling water quality Icop Cooling Water Heat Exchanger Discharge Note: CC - Continucras Conductivity tbnitoring Grab - Grab Sample IX. - Interrittant Silica >bnitoring Cpil - Continuous pil Monitoring i i O L._.-_,.._.__._.___..__._.._._._,_._.,___ f

1 Stl096%. Tutz 9.5.9-1 i SIDRNZ Or CE aster TPEREPE

(bntainer Oper. Max Est. Ter* Met Ehecyr Belemee(2)6 Voltre if 7!tiptured (ft 11x10 Design Press. Press. (1) No.

Looetion l (PSIG) (PSIG Ccmtain._ (ft3) each One Tank All Tanks i Gas (PSIG) T Carbon Dioxide (3) I 320 230.8 230.8 Yard Fire rivb-ct. ion 363(4) 300 341 1 i i

Halon 1301 1 Firn TiuGKLlon 2 0.74 .002 .004 Mesctor Blog.

j Beactor Shutrbwn 2650 600s70*F 1250 1 i j Fire FivhCdon TSC conputers (3) 4 .25(5) .50(5) TSC Bldg. 600 360 600 2

Above Floor 2 1.8 .14(5) .28(5)

Belcw Floor 1000 600 1000 ) I Air J j Iwn-.i. & 3 415 11.8 35.4 turbim Bldg. 125 145 ! Service axeivers 150 i ! Cbntrol Rnn i Breathable 10 1.73 1.17 11.67 Turbi m Bley. 10000 2400 4000 1 Air W A 2216 3693 5 0.30 0.16 0.80 B 5000 11 0.205 0.005 0.051 h M y cont. SW Acetrolator 145 95 115 i 1 i . i i i l of 2 i

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

O SHONE. gJt taste 9.5.9-1 (Cont'd) STUPXE OF GA3 UNDER PPESSURE Est. Tanit PhK F- m yy > 1mmen(2)6 I Cont. h Oper. Max I Design Press. Press. (1) 7:o. Voltsee if Rtytured (ft itzt10 (ISIG) (PSIG Contain. {f t3) each One Tanit All Tanks locatim c.is (PSIG) Nibum. 1.04 2.06 Turtrine Bldg. 10000 2600 4000 2 1.73 Ilot Water Ih ting 4000 2 1.73 1.04 2.05 1hrbine Bl&y. ; Glycol IIeating 10000 2600 P:tipane 220 2 85 1.25 2.51 Tard Aux. Bir Ignition 480 204 [ i i l i Notes (1) Safety valve set point. (2) Reversible Mlahmtic expensim fItse MKintan container press:tre. ' (3) Stored as liquified gas. ' (4) Mixirvann working pressure. - (5) nix energy release calculated m tesis of 5/3 wereting press

ining equal to munx press. This does not significantly alter the l max. energy release mavber. *

(6) Variable quantity and type of welding gas tanks not listed. Type and Wty is variable based m mintenance requh_a.s. l I i L f i l I 2 of 2 [ t I

             ._            _. _              m. _ __ _ _ . ~ _                     . _ . . . . _ . - - - _ _                                                               _ _ ._ _ .._ . __ _ _ . _ . _ _ .        _

r e 2 DRIDBE TR AVEL & , r y _ 478.00* REF.INsipt LiwtR(si'.6*) ' O --

it SS 56 De 56 CONTA0L , f 400 ARE A i b ( 4 8' ! 52 DS 94 96 to

                                                                                               .T R OL L E Y T R A 'f L                    FUEL AllCMILY                                                                                                !

ETEE.PWTIONE l Y ( T Y P.)

                                                                                                                                                                                                                         /                              l i

f . g- 32 e6 96 96 to

O b ,

f.a l 31 96 96 96 94 1 # l O 1c 7

l 96 96 96 96 [

                                                               ,                                                                                                                                                                                        s
                   *                                                           " '                                                                              "                                 TEMPORARY

[ CONTROL ROD AMEA(99) i

                                                                  ,                                                                                                                         AND CASK 1TORAGE                                            !

AhD LOADING AREA ! N 68 56 56 l w .. y NOTERe 2176 FUEL STOR AGE Pollfl0NS ! 144 CONTROL RCD STORAGE PollTIONS ,

                                                            ~
                                                                             ,                               FIGURE 9.1.2-1                                                                                                                             !

HIGH DENSITv RACK PLACEMENT l IN SPENT FUEL POOL ! SHOREHAM NUCLEAR POWER STATION O otruuto sarEry mAtvsis ReeORT > i

 . . _ _ _ _ - . _ _ . _ _ _ _ - _ . _ . _ _ _ _ _ .                                                               _ _ _ _ _ _ . - _ - . . , - _ . _ . . _ . . ~ . . _ .                                     _ . _ . _          .. ..         - . ~ - -I

SHOREHAM DSAR CHAPTER 10 STEM! AND POWER CONVERSION SYSTEM i 10.1 STEAM AND POWER CONVERSION SYSTEM The purpose for which the steam and power conversion system was built.no longer exists. The components of this system as described in the USAR will not be required in the defueled condition. 10.2 TURBINE GENERATOR The purpose for which the turbine-generator system was built no longer exists. The components of this system as described in the USAR will not be required in the defueled condition. There is no longer a concern for turbine generated missiles. 10.3 MAIN' STEAM SUPPLY SYSTEM The purpose for which the main steam supply system was built no longer exists. The components of this system as described in the (; USAR will not be required in the defueled condition. In the defueled condition the main steam system will not serve any safety-related function and therefore will be reclassified as Q.A. Category II, l 10.4 OTHER FEATURES OF STEAM & POWER CONVERSION SYSTEM 10.4.1 Condenser The purpose for which the cendenser was built no longer exists. The components of this system as described in the USAR will not be used in the defueled condition. 10.4.2 Main Condenser Air Removal System The purpose for which the main condenser _ air removal system we? built no longer exists. The components of this system as described in the USAR are not required-in the defueled condition. 10.4.3 Steam Seal System-The purpose for which the steam seal system was built no longer exists. The components of this system as described in the USAR are not required in the defueled condition, i O - 1 10-1 Rev. 3 July 1991 r

 - . -       w    w,eew   ,-,-,,...r          y _ .-. r.,..     .    -.--..,--w.,  ...m,.-.,,_..,.,...,-m   ..,...,,----ww.e.e....,w.,    -
                                                                                                                                              , -  o,r--w-r - ..rsm.-- e,-, ,v-,,y.   --w,,-er,-,.m...,-p

          -- - ._ _ -_ _ _ - - ____ _ -                            .  - -    _ - _           _ _ _ _ -          - _ . - .          .       _ _ - - - - =

l SHOREHAM DSAR h 10.4.4 Turbine Bypass System ' The purpose for which the turbine bypass system was built no : longer exists. The components of this system as described in the USAR are not required in the defueled condition. ! The portion of the bypass system upstream of the bypass valves l was built to ASME III Cede Class 2 criteria. As the function of t l the bypass system no longer exists in the defueled condition, the bypass system is reclassified O.A. Category II. l f i 10.4.5 Circulating Water System I The purpose for which the circulating water system was built no i longer exists. The components of this system as described in the USAR will not be required in the defueled condition. The only i exception is that a circulating water pump and the circulating water discharge system may be used te provide dilution capacity [ l for elimination of liquid radwaste and SPDES limits on chlorine and suspended solids to the Long Island Sound. , 10.4.6 Condensate Domineralizer_ System l Since there is no fuel in the Reactor and no Reactor steam I produced, there is no need for the Condensate Demineralizer l 4() System. However, the Acid and Caustic Storage Tanks (IN52-TK-035 l and -TK-036) will remain operable to provide regeneration chemicals for the continued operation of the Demineralizer and t Makeup Water System (P21). The Chemical Waste Sump (IN52-TK-113) will remain operable as a pathway for further treatment of non-radioactive regenerant waste from the Demineralizer and Hakeup Water System. 10.4.7 Condensate and reedwater System I The purpose for which the condensate and feedwater system was built no longer exists. The components of this system as ! described in the USAR will not be required in the defueled condition. , Piping built and designed to ASME III Code Class 1 is considered Inservice 0.A. Category II while in the defueled condition. l inspection according to ASME XI need not be performed while in l the defueled condition. i

I I) 10-2 Rev. 3 July 1991

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

SHORIHAM DSAR CHAPTER 11

RADIOACTIVE WASTE MANAGEMENT J1.1 RADIATION SOURCE TERMS The description contained under this heading in the latest revision of Shoreham USAR remains unchanged as it is used to develop the basic design criteria of the plant. However, the actual source terms in the plant's present defueled condition are as follows:

a. Liquid Radioactivity Sources As of August 1989, since all SNPS' fuel had been placed in the spent fuel pool, there were no liquid sources with nuclide concentrations greater than the Lower Limit of Detection (LLD),

outside of radwaste streams. It must be recognized that in the future some concentrations greater than LLD will be seen (e.g., as sludge at the bottom of sumps is processed to Radwaste). l However, these should be minor and temporary occurrences. Sources related to the decontamination and decommissioning should also be minor, as the degree of overall plant contamination is low. These liquid sources would be dealt with in accordance with the Liquid Radwaste, ALARA, and Health Physics programs as discussed in DSAR Sections 11.2, 12.1, and 12.5, respectively. i( ) Isotopic concentrations above the LLD levels in the Radwaste System as of 6/30/89 are indicated in Table 11.1-1, from References 2, 3 and 4.

b. Gesecus Sources There are no detectable gaseous sources at SNPS, either present or anticipated. This statement is supported by the fact that the Semi-Annual Radiological Effluent Release Report for the first and second quarter 1989 (Reference 1) indicates there were no detectable releases during the six-month period, either from the offgas system or the various building exhaust systems.

c. Activated Materials Sources l

It is expected that materials which were located in the reactor vessel during low power testing (eg, control rods, TIPS, IRMs, and the like) have been activated to some extent. With the l ~ exception of some portions of the liquid radwaste system (10 mrem /hr maximum) , dose rates outside of plant systems are very low, less than 0.5 mrem /br. Thece low dose rates are indicative of a low deposition of sources within plant systems. O 11-1 Rev. 1 Aug. 1990

SHOREHAM DSAR h There may be a minor amount of activation source material deposited within plant systems. However, the level of this activity, and indeed of the activation products within the reactor vessel itself, are not considered significant compared to the spent fuel sources described in Section 12.2. REFE RENCES General Updated Safety Analysis Report (USAR) Shoreham Nuclear Power Station Revision 1, December 1987.

1. ' Semiannual Radioactive Ef fluent Release Report - First and Second Quarter 1989", transmitted by letter SNRC-1619, B/29/89.

2. "SNPS HIC Package Data for November / December 1988", 6/5/89, Memorandum L. Hall to T. Gillett.

3. " Transmittal of Data for Dose Projection", 5/16/89.

Memorandum, P. Lynch to M. Beer.

4. Gamma Spectrometer Scan of Floor Drain Collector Tanks, Waste Collector Tanks, and Recovery Sample Tanks, 6/15/09, Memorandum M. Ma to T. Gillett.

 '(7 s) 11.2    RADIOACTIVE L2 QUID WASTE SYSTEM 11.2.1    Design objectives The Radioactive Liquid Waste System is described in the USAR.

With the Reactor defueled and the fuel assemblies stored in the Tuel Pool, the sources, quantity and activity of the radioactive waste are greatly diminished. Certain portions of the Radioactive Liquid Westo System are not required. 11.2.2 System Descriptions The estimated influent to the radwaste system is reduced fron 25,000 gpd to approximately 2,000 gpd. The regenerant chemical subsystem is no longer required, except for the Chemical Waste Sump, the Regenerant Liquid and Evaporator l Feed Tanks and Pumps. The Waste Evaporator portion of the Floor Drain Subsystem is not required. The Phase Separator System serving the RWCU System is not l (~]

  %J required.

l 11-2 Rev. 3 July 1991

SHOREHAM DSAR 11.2.2.1 Summary This section is no longer applicable since most of the waste , I streams would no longer exist. 11.2.2.2 2,ov Conductivity Waste fubnystem waste Collector Subsystem This system 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. l 11.2.2.3 High 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 Regenerant Chemical Subsystem ,7~

-       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 condition. 11.2.3 System Design 11.2.3.1 Equipment Description This section remains as presented in the USAR. 11.2.3.2 , Applicable Codes and Standards This Section remains as presented in the USAR. 11.2.3.3 Radwaste Building This Section remains as presented in the USAR except that the Radwaste Building is now designated a Quality Assurance Category II, Non-Seismic Structure. 11.2.3.4 Liquid Radwaste Equipment Quality Group Classification This Section remains as presented in the USAR. 11-3 - -

                                                           -Rev r-3  July 1991

SHORERAM DSAR . fr 11.2,3.4.1 Conditions and Assumptions This accident (raised in USAR Section 11.2.3.4) pobtulates the ! simultaneous failure of th6 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 ef fluents. The tanks (and their capacities) that are assumed to f ail are:

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

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

3. Regenerant liquid and evaporator feed tanks: Two at 25,000 gal each (contents are insignificant 1y radioactive) . .

4. Recovery sample tanks: Two at 25,000 gal each (located i outside the radwaste building contents are insignificant 1y radioactive)

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

V f. Spent resin tanks one at 4,700 gal (Section 11.5) j The source concentrations in the above are described in DSAR Table 11.1-1. + 11.2.3.4.2 Accident Description The accident description can be considered as described in : Section 11.2.3.4.2 of the USAR except the structure is now j 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 f actor 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 l assumed that all the activity evolves. l

2. Ground release atmospheric dispersion factors are assumed, as ;

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

3. The breathing rate of persons ci'-A.te is r.ssumed to be e

3.47E-04 cubic meters per second, consistent with Regulatory Guides 1.3 and 1.25. For other age groups the breathing rate ' was obtained from the ratio of the maximum age group rates given in Regulatory Guide 1.109 (Reference J). i 11-4 Rev. 3 July 1991

i SHOREKllM DSAR 11.2.3.4.4 Results and Cor.sequet ces f i The doses resulting from the ana;/ sis described above are as , follows: { i Dose, millirem __

Whole body Beta Maxist.um bource Gamma

skin organ **

Spent Resin 1.8E-05 2.7E-06 1.3E-03

Tank Radwaste Filters 1.2E-07 1.7E-08 8.3E-06 ,

Discharge setnple 3.1E-08 1.4E-08 7.72-06 l Tanks . Totals 1.8E-05 2.8E-06 1.3E-03 _i i

External & internal pathwayst child is the limiting- ;

age group l

                                              ** Teen is the limiting age group, and lung is the                            ,

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

                    .11.2.3.5               Instrumentation & Control                                                       ;

i The description contained under this heading in the latest revision of Shoreham USAR remains unchanged. Refer to the USAR l for information_on_this subject.- l 11.2.3.6 Shielding Field Routed Pipe The-description contained under this heading in the latest }

                    -revision of.Shoreham USAR remains unchanged.                                Refer to the USAR          [

for information on this subject. 5 11.2.4 Operating Procedures , Operating procedures including _ administrative control of liquid ! radwaste releases are as described in the USAR except the Radwaste Building is-now. classified as CA Category II. t 11-5 Rev. 3 July 1991 - i i

_ . - - - . . - .- . _ .. _ - . - - - _ . - - ,-. - _ ,_ D

l SilOREHAM DSAR

 ,/ q                                                                                                                                                                                     '

ks_) 11.2.5 Performance Tests Performance tents of equipment are as described in the USAR, , enccpt for activity reduction factors (DF), which are no longer applicable. Only eqvipment that remains in operation will be periodically tested. 11.2.6 Estimated Releases '

fuel Liquid effluent releases are expected to be This is based on the fact that duringminimal with the , in the spent fuel pool. the period from June 1988 through May 1989, only one release had an isotopic concentration greater than LLD. ; The quantity of tb n,mual release of contaminated liquids is ; conservatively estimated by noting that the discharge volume from SNPS is approximately 5,000,000 gallons per year. Assuming the effluent concentratiot is consistently equal to that found in the one sample above LLD g7.83E-08 uCi/cc of Co-60, from DSAR Table 11.1-1), the astimated release is: 1.5E-03 C1/yr of Co-60 11.2.7 Felease Points ' I) The description contained under this heading in the latest revision of Shoreham USAR remains unchanged. Refer to USAR for information on this subject. I 11.2.8 Dilution Factors Under the plant's present condition, service dater or circulating water will be used, if necessary, for dilution so that the discharged effluent concentration in the Long Island Sound will not exceed that prescribed in 10CFR20, Appendix B, Table II, Column 2. Treated radioactive effluents are collected in the discharge sampic tanks. The filled tank is sampled, and then discharged at a maximum rate of 150 gpm for a period of approximately 2.5 , hours. If necessary, the treated effluent in 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. Estimated Doses t 11.2.9 ['\d) offsite doses due to liquid releases areAnexpected to be minimal, estimate of the yearly as discussed in DSAR Section 11.2.6. 11-6 l l

d SHOREHAM DSAR fn) dose is conservatively obtained by assuming each batch liquid

\,

release contains the maximum batch activity concentration of i 7.83E-08 uCi/cc of Co-60, and the release volume is approximately 5,000,000 gallons per year. Astuming no dilution, the resulting doser are as follows: 1 Whole Body 0.166 mrem (adult) GI-tLi 1.43 mrem (adult) Live. 0.074 mrem (cdult) 4 As noted in Section 11.2.0, service water dilution remains available as necessary. 11.3 GASEOUS WASTE SYSTEM J 11.3.1 Design Obiectives , With the fuel in the Spent Fuel Pool, the radioactive gaseous waste system is no longer required to meet either ICCFR20 or 10CFR50 Appendix I limits. l 11.3.2 System Descriptions S With the fuel in the Spent Fuel Pool, and negligible amount. a1 1 I radioactive halogens in the fuel, the radioactive waste sou - a g( ) described no longer apply, and the systems neceerary to proce + r them are not Jequired. Nc7 mal ventilation will be maintained in the Radwaste and Reactoi Buildings with discharge through the station ventilation exhaust l duct. 11.3.3 System Design The process offgas system, which is the system described in USAR Sections 11.3s3, 11.3.4 and 11.3.5, is not required with the fuel in the Spent Fuel Pool. 11.3.4 Operating 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 i since the plant achieved initial criticality in 1985, there have been no recorded releases documented in the Semi-Annual Radiological Effluents Reports. a 11-7 Rev. 3 July 1991

SHORERAM DSAR f )

 \xs/ 11.3.7    Release Points The description contained under this heading in the latest                '

sevision of the Shoreham USAR remains unchanged. Refer to the USAR for information on this subject. 11.3.8 Dispersion Factors 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. 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 would be expected to occur during normal plant operation. In the defueled condition some pLthways do exist on loss of ventilation systems. Doses in such an event would be insignificant due to - the low radionuclide inventory in the plant. (Reft NED Safety . (,) (/~% Analysis No. 4170024) i 11.4 PROCESS AND EFFLUENT RADIATION MONI't, RING 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 S'OLID WASTE SYSTEM 11.5.1 Design objectives The description contained under this heading in the latest revision of the USAR remains unchanged as it is used to develop t 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 cubicles (since no solidification of waste, per se, is needed). High Integrity containers (HICs) will continue to be used since some wastes willActive Also Dry continue Wasteto be (DAW) will I7 s generated, and must be shipped. k-} continue to be generated, and must be shipped. The volume of both will be significantly less than that given in the USAR. 11-8 Rev. 3 July 1991 -

   . _- .    -.-              . - -.- -- -              ..- .- -,~ -. -                        . .-__ - - -           . . - - . .

SHORERAM DSAR ( It'should be noted 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 11.5.3-1B 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 i 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 Equipment Description () 11.5.3.1 General The only equipment remaining in use in this system is as follows: 4,700 Gallon Spent Resin Tank (SRT) For the defueled condition, this tank receives backwashed resin l and filter media from the Radwaste Demineralizer and the Fuel Pool Cleanup Demineralizer and Filters. (This tank is also l 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 The HICs-are set.on the Radwaste floor or in the pits in the floor. are then- dowatered 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.

                                                                                                                                      .i O

11-9 Rev. 3 July 1991 yr-., , ,.% ,-y ,y. . .,,

SHOREHAM DSAR i) Baler This equipment is furnished to compress miscellaneous dry active waste (DAW) into 55 gallon drums. Transfer Carts and Fill Pipes 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 by 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 USAK no longer applies. The second paragraph remains applicable. 11.5.3.3 Dry wastes gs This Section of the USAR is applicable, as some DAW will continue l Ilv ) to be generated. ! 11.5.3.4 Irradiated Reactor Components l This Section of the USAR still applies. 11.5.3.5 Operating Procedures i 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 bitrial sites.

2. The shipping container is located under the retractable fill pipe by first placing the container 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 direction by pneumatic cylinders.

lO

  %J 11-10             Rev. 3   July 1991

SHORERAM DSAR , s 11.5.3.6 Instrumentation All instrumentation in this section is no longer needed except , , for the radiation monitors. 11.5.4 Expected Volumes , i This Section of the USAR is superseded by the following: A conservative exp'ected 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. ; r This statement and Table together supersede Table 11.1-1A of the , USAR. j DAW volume is conservatively estimated to be 1,000 cubic feet per ' year, buried volume. The DAW activity is negligible. t 11.5.5 Packaging The description contained under this heading in the latest revision of the USAR remains unchanged. Refer to the USAR for [ information on this subject. l I 11.5.6 Storage The description contained under this heading in the latest ; revision of the USAR remains unchanged. Refer to the USAR for information on this subject. l 11.5.7 shipment ; The description contained under this heading in the latest , f revision of the USAR remains unchanged. Refer to-the USAR for ! information on this subject. i. 11.6 OFFSITE RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM DISCUSSION i i The objectives of SNPS' Offsite Radiological Environmental ! Monitoring Program [REMP) are to identify and measure plant ' generated radioactivity in the environment andSNPS' to calculate REMP is the i potential dose to the surrounding population. designed to comply with the 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 results vary for each sample and location, several sampling !'

[) (_/ locations were selected for each medium using available 11-11 7 f

SHOREHAM DSAR meteorological, land, and water use data. The range of analyses performed on a sample depend on the type of sample taken. Sampling locations are designated as either indicator or control. Indicator locations provide representative measurements of radiation and radioactive materials for those exposure pathways and radionuclides (from SNPS) that lead to the highest potential radiation exposures. Control locations are placed sufficiently far from SNPS so that they will be beyond the measurable influence of SNPS or any other nuclear facility. This monitoring program implements Section IV.B.2 of Appendix I to 10CFR Part 50, by verifying that measured concentrations of radioactive materials and direct radiation are representative of the actual contamination levels and doses to the public. SNPS' REMP has been subdivided over three distinct time inte rvals : Preoperational REMP (prior to SNPS' initially achieving criticality), Operational REMP (from initial criticality until removal of the fuel from the core), and Post-Defuel REMP (after the core was transferred to the spent fuel pool). Preoperational REMP was performed to identify and determine background levels of environmental activity around SNPS, Preoperational REMP also served to verify that indeed the media being sampled and analyzed is sensitive to radiological ' fluctuations in SNPS' environs (indicator locations) and to provide effective monitoring of potential critical pathways. Preoperational and Operational REMP samples within the aquatic ' environment included surface water, algae, fish, invertebrates (clams, lobsters, etc.) and sediment. The atmospheric environment was sampled 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 analyses for each sample could includes gamma l spectrometry, Sr-89 and Sr-90; I-131; H-3, gross beta, direct radiation and noble gases. Under Post-Defuel REMP, several of the above sample types, sampling locations and/or analyses are l discontinued. The current Post-Defuel REMP program is outlined in Tables 11.6.3-1 & 11.6.3-2. Preoperational REMP began in February 1977 and continued through 1984, although the of ficial Preoperational REMP period; i.e. the l 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% power' (during January 1989), SNPS has not generated radioisotopes since the last 5.0% power test, (% completed on June 6, 1987. Comparisons between the 11-12 Rev. 3 July 1991 j 1 - . l

s SHOREHAM DSAR above two phasec of REMP were documented in each annual REMP report. 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% power test was completed during June 1987, per Ref. 9, with the exception of I-129 and Kr-85, all iodines and gaseous l effluents have decayed away. Consequently, the surveillance requirements for SNFS' Post-Defuel REMP were reduced to below the operational level.

Justification for Reducing REMP to Post-Defuel Surveillance 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 sampled in the pathway and the frequency of sampling Since may be reduced to operational Tech Spec requirements. (as of August 9, 1989) the core has been in the spent fuel pool, the initial core has

() " exceeded" the point of maximum burnup. It should be noted that the concept of " normal" Tech Spec requirements as referred to in Reg. Guide 4.1, refers to a fully operational station 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 l l performed in two steps. Step one reduces Operational REMP to the i level mandated when SNPS was 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 by 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) was compared to precperational REMP.

1 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) as having originated at SNPS. These results satisfy the criteria established in-Reg. Guide 4.1 for reducing post-defuel REMP to the level originally mandated by SNPS' O license. The sampling points not required by the license are: 11-13 Rev. 3 July 1991 i

                                   .    -     ~      .        . - _           - .   .

SHOREHAM DSAR

1) Game; 4) Rain Water; and

2) Aquatic Plants; 5) Noble Gases. ;

3) Aquatic Sediment; Justification for reducing REMP to the revised requirements (after the core was defueled) is given based on the above information; i.e. , the measured environmental impact due to 5.0%

power testing was comparable to that of preoperational REMP, and as of August 9, 1989, the core was removed from the reactor pressure vessel. SNPS' last 5.0% power test was completed on June 6, 1987, and per Ref. 9, with the exception of I-129 and Kr-85,alliodinesandgaseouseffluentshavesincedecayedaway.l In additien, radwaste system activities are quite low (listed in DSAR Sections 11.1 & 12.2). As a result, the only remaining . radioisotopes (and their release pathways) ares l Isotope ( s) Source Effluent Pathway j

1) Kr-85 Spent Fuel Gaseous

2) Solubles and Radwaste Gaseous and Liquid ,

Particulates SNPS' Post-Defuel REMP Surveillance Program Outline - (Steps 1 & 2) Leduce from 36 to 18 locations ! II ) 1) DIRECT RADIATION: Quarterly Surveillance Frequency

2) AQUATIC

n. Aquatic Plants and - Delete, not required i Beach Sediments

b. Fish, Surface Water - Retain, may be impacted and Invertebrates from liquid release path to L.I. Sound t

Perform Semiannual surveillances as available

3) AIRBORNE

a. Iodine - Delete, insignificant quantity

b. Particulates and - Retain, particulates and Gross Beta and solubles still exist,

c. Noble Gas - Delete Noble Gas, not required.

Quarterly Surveillance Frequency O 11-14 Rev. 3 July 1991

s i r SHOREHAM DSAR ,

 -,"-                                                                                                l 4)~    TERRESTRIAL                                                                    i
  \,

a. Precipitation, Soil, - Delete, not Tech Spec I and Game required :

b. Potable Water - Delete, well water not impacted by discharges to L.I. Sound

c. Milk, Food products - Retain, long lived ,

particulates , L Quarterly Surveillance for Milk, i Annually for Food ' j

SUMMARY

/ CONCLUSION                                                                       l l

1) Examination _of the radiological impact to REMP locations which are to be eliminated -- From 1983 (preoperational REMP) l i

through 1988 (which encompasses SNPS' 5.0% 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 pools thus, the initial core has reached maximum j burnup.

  ,~      3) _ Per Regulatory Guide 4.1, if the above two conditions are                             ,

( met, then the operational phase of REMP may be reduced to the : l requirements that were written when SNPS was to be operated l l as designed. I ! 4) The post-operational REMP surveillance program may be reduced . from Step 1 to the requirements as delineated in DSAR Table l 11.6.3-1 (Step 2), developed after the SNPS core was transferred to the spent fuel pool, because l l

a) Criticality will not be reestablished at SNPS. As of August 9, 1989, no additional fission / activation t l 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 , l Kr-85, all remaining gaseous effluents have decayed i away; and c) the on3y possible release paths for the remaining soluble or particulate effluents is through either 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

                      -(liquid and gaseous pathways) through which effluents are being or could be processed.
      )

i 11-15 Rev. 3 July 1991 ; r

i i SHORERAM DSAR

  T S(N J 11.6.1 Objectives of REMP 11.6.1.1 Preoperational REMp_                                                                          ,

The objectives of the Preoperational REMP were:

1. To identify and determine bassline 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 characteristics of the environs at SNPS, and to assure that REMP will be responsive to radioactivo 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 areast and

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 capabilities.

.n)

(, The years 1983 and 1984 served as theAll official preoperational data collected during period, as stipulated in Reference 8. this period were used in developing a baseline for ultimate comparison with operational 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 revected 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 iso'eopes 2r-95, 3b-95, Ru-103 and Ce-141. Also in 1983 and 1984, REMP sampling identified low levels ;f 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 and sediment were chosen and reflected the The normal atmospheric background radiation found in this environment. environment was sampled for airborne particulate matter, fodine, and noble gases. All airborne radiciodine analyses were below ( detectable levels. In addition, milk, potable water, game, The food

   \            products, beech sediments and rain water were sampled.

11-16

SHOREHAM DSAR results obtained from the analyses of these samples were typical

           ) of the radioactivity values usually associated with samples of these types. All radioiodine analyses of milk were below                                               '

detectable levels. Direct radiation levels were No relatively low, unusual and approximately the same at all locations. l radiological characteristics were observed in the environs of i SNPS during 1983 and 1984. A summary of the annual program results for 1983 and 1984 is given in USAR Tables 11.6.1-1 and 2. 11.6.1.2 operational REMP The objectives of Operational REMP weres i

1) Identify and measure plant-related radioactivity in the .

environment for the calculation of potential dor.es to the l public.

2) Identify excessive radionuclide concentrations of limited ;

duration, so that appropriate action may be taken. l

3) Determine the long-term variation in radionuclide .

concentration, or j

4) determine the effects of plant effluents on the environment. F

5) Comply with regulatory requirements and provide records to ;

document compliance. j j t

6) Comply with the REMP requirements as outlined previously. l i

Operational REMP used the Preoperational data base to identify l l plant-contributed rcdiation, and to evaluate the possible effects l of radioactive effluents on the environment. The Preoperational ; and Operational phases of REMP were designed to comply with l 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 radioactive 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 during 1985, and DSAR Table 11.6.1-4 presents a comparison of preoperatonal and operational REMP data f rom 1983 through 1988. 11-17 l

4 SBORERAM DSAR (fr) (, 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 Liquid 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. USAR Section 11.6.2.1 Tontains detailed discussions about the projected doses from various liquid pathways. With the updated source terms as described in the DSAR (Sections 11.1 and 12.2), future doses from liquid pathways are expected to be a small fraction of the doses presented in the USAR. See DSAR Section (b]' A 11.2.9 for 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 milking 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 beca 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 Dose Calculation Manual. Dose Computational Methods (_) 11.6.2.3 11-18

SHOREHAM DSAR . ib) 'N 11.6.2.3.1 Liquid Effluent Pathways The discussion contained in the latest version of the Shoreham

USAR (Section 11.6.2.3.1) continues to apply.

11.6.2.3.2 Gaseous Ef fluent Pathways The discussion contained in the latest version of the Shoroham USAR (Section 21.6.2.3.2) continues to apply. 11.6.3 Sampling Media, Locations, and Frequency typical Post-Operational REMP mampling locatione 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 i sections. 11.6.3.1 Sampling Media 11.6.3.1.1 Aquatic Fnvironment

/~ \  The aquatic environment is examined by analyzing samples of:   1)     -

k- 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, Scophthalmus aguosus, Sea Robin, Prionotus spp, Little Skate, Raja erinacea, Blackfish, Tautog onitis and Summer Flounder, Paralichthys dentatus are taken by trawl, sealed in plastic bags, frozen, and shipped to the analytical laboratory fcr analysis. When available, invertebrate samples of American Lobster, Homarus americanus, Squid, Loligo pealeii and Channeled Whelk, Busycon canaliculata are collected by trawl. Channeled whelk are also collected using pots. These invertebrate samples are then sealed in plastic bags, irozen and shipped to the laboratory for analysis. Blue Mussels Mytilus edulis are collected by hand along jetties and soft-shell clams, Mya arenaria from Wading River are shelled and sealed in plastic bags, frozen and shipped to the analytical laboratory. in Q 11-19

SHORIEAM DSAR 11.6.3.1.2 Atmospheric Environment

The atmospheric environment is examined by analyzing airborne particulates collected on Gelman Type A/E filters using low volume air samplers (approximately I c f m) . The samplers used l 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 atmospheric pressure. Sample volumes are corrected to standard pressure using average weekly barometric pressure (measured at Environmental Engineering Department, > Melville) and air sampler' vacuum readings. Time totalizers indicate the duration of time 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 Milk r through October) when the sampling is increased to monthly. l samples are prepared for shipment in accordance with the

    '     instruction of the laboratory performing the analysis. Food products consisting of vegetablec and fruit are collected from area farm stands and shipped fresh 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 by LILCO prior to placement in the field. One TLD is placed at each of the 18 locations, and exchanged on a quarterly bases; these locations correspond to the 16 meteorological sectors in the general areas of the site boundary, plus two control locations (actual locations are listed in Table 11.6.3-1). The units are then packaged and shipped to the laboratory for analysis. ! 11.6.3.2 Sampling Locations and Frecuency l 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. 11.6.4 NOT USED IN THE DSAR (Data Incorporated Into Section l 11.6.1)

      -s G/

11-20 Rev. 1 Aug. 1990

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

SHOREEAM DSAR l t

  )  11.6.5    Data Analysis, PreFentation and Interpret & tion                                    ,

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 Program 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 i in the Environs of Nuclear Power Plants" l

2) Not Used i

3) Not Used j

4) Radiological 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"

6) SNPS Offsite Dose Calculation Manual l

'- 3/4.12 Radiological Environmental Monitoring 3/4.12.1 Monitoring Program Table 3.12.1-1 "REMP" !

7) Not Used l

SNPS' Operational REMP Annual Reports: January 1, to December

8) '

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" ;

i 11-21 Rev. 3 July 1991

SHORERAM DSAR TABLE 11.1-1 k'-') Radwaste Sources Greater than LLD Spent Resin Tank, Radwaste Filter, & Floor Drain Filter The activity concentration is assumed to equal the maximum in the most recent FIC shipment (Nov-Dec 1988) and is (From Reference 2): Activity Concentration, uCi/ec t of Activity . Isotope 9.84E-04 58.46%

       *Cr-51                                                   1.29%

Mn-54 2.17E-05 4.19E-04 24.88%

       *Te-55                                                   0.05%
       *Co-57                    7.92E-07 6.43E-06                       0.38%

Co-58 6.51% Co-60 1.09E-04 4.57E-05 2.71%

       *Fe-59                                                   0.38%
       *N1-63                     6.41E-06 3.25E-06                      0.19%
       *Sb-124                                                   1.124
       *2n-65                     1.89E-05 6.21E-06                      0.37%

H-3 0.02%

        *C-14                     3.94E-07 1.69E-07                       0.014 Vs     *Sr-90 1.52E-05                       0.914

() *2r-95

        *Nb-95                    2.55E-05                       1.51%

4.79E-09 0.004

        *7c-99                                                   0.00%
        *I-129                    7.32E-10 1.34E-06                       0.086
        *Cs-137                                                  0.18%
        *Ce-144                   2.95E-06 1.59E-05                       0.95%
        *Pu-241 Discharae 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 Concentration, uCi/cc t of Activity Isotope 7.83E-08 100.0% Co-60 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.

s./ 11-22

h DSM

                                                                                                                                                                                           ' DUKE 11.6.1.-4 Copparison Of Operational - Precoerational REMP Data i

( Operational T:EMP ) (- hwai.ional R!f@ -) i i 1987 1986 1985 1984 1983 Unit /I w h ve 1988 _ l SAMPIE TYPE i pC1/1 240 - 410 140 - 450 130 - 420 150 - 290 120 - 540' 70 - 220 Potable Water (11-3) _ _ 76.7 - 9270 35.1 - 6490 54 - 3230 992 - 4330 641 - 5340 34.0 - 6310 Game pCi/Kg(Cs-137) mtun Monthly 2.3 - 5.2 2.8 - 6.9 1.9 - 5.7 3.0 - 6.2 2.7 - 6.9. 2.3 - 5.7 Direct (gama) 2.9 - 4.9 2.8 - 5.5 3.1 - 6.2' 2.8 - 5.4 Radiation Quarterly 2.7 - 4.8 2.9 - 5.0

                                                                                                                                                                                                                             .5.0 - 360                  6 - 47                            4.2 - 61.                          5 - 54 Air: Gross Beta                                                     [x1.0E-3]                                   5.0 - 44.0                                  4.0 - 32.0 LT 0.8                                      LT 0.8.                      0.11 - 0.27                   LT 0.8                                  LT 0.07                    1.3 - 1.4 Particulate Sr-90 K i/m' x 1.E-3                                                                                                                                                                                                                                                                  LT 30.0 LT 10.0                                     LT 10.0                            35 - 1230**             LT 10.0                              'LT 10.0                                             ;

Iodine-131 pC1/m2 x 1.E-3 i LT 1.0 LT 1.0 6.8 - 27.

33.. LT 20.0

                           . 7quatic                                                          pCi/Kg (Sr-90)                                         LT 1.0 t

B5.5
47.9
45. 69.7 - 140. 36 - 55 Plants pCi/Kg (Ce-137) LT 6.0 0.61 - 5.70 0.98 - 13.0 0.86 - 4.60 0.69 - 5.3 0.9 -- 7.7 0.76 - 6.00 pCi/1 (Sr-90) 12.9 - 14.1 6.00'- 14.8 5.90 - 11.5 7.0 - 8.9
4.4 9.6 - 14 V.11k pCi/1 (Cs-137)

LT 0.20 LT 0.20 tot pC1/1 (I-131) LT 0.20 LT 0.20 2.1 - 4.8 LT 4.0 LT 4.0 LT 4.0 LT 4.0 LT 3.0 19L Food pCi/Kg (I-131) LT 5.0 LT 5.0

12.2 ' LT 5.0 LT 5.0
24.7 -

Products (wet) (Ca-137) L

Ranges are not given since only one data point contained an identified iwM.

I

                                  ** Evidence of C=:udifl accident.

11-23 Rev. 3 July 1991 t i g- w r w- -%-e'-,.-er . - , , - +we, , - - , , s,,-w-ye , - , eg-4, e- ,4- e ,.,.y .= we- a w ws - .- e e *t-+--e ,aw.-.- , - ,- e - w z wem w vn-- +,w,.,s-,..e_.%+,e .-,--.e .-e.---, ..-+,r-----v-e.4,w=,. -,.eww-=-,----*-w , . - ~ . y ,s

M DSAR TABIE 11.6.1. -4 (Obnt'd) Otmmrifeon Of Operational - Fi+1stional REMP Data { Operational PEMP - ) (- Preoperational REPP -) 1987 1986 1985 1984 1993 SNEE TITE Unit / Isotope 1988 LT 1.0

5.6 LT 1.0 LT 0.9
86 Aquatic pCi/I'g (Sr-90) LT 1.0 LT 5.0 34.8 - 36.2 m m m m Invertrbrate (wet) (Ce-137)

LT 1.0 LT 1.0 LT 1.0 LT 1.0

3.3 LT 2.0 Beach pCi/Kg (Sr 90)

(dry) (Cs-137) _ LT 8.0 LT 8.0 LT 8.0 LT 8.0 Itf 9.0 m Gedinumt LT 2.0 LT 2.0 LT 2.0

1.7 LT 3.0 A;uatic pCi/Kg (Sr-90) LT 2.0 (dry) LT 10.0
21.7 LT 10.0
30.4 44.2 - 49.4 m Sedinent (Cs-137) pCi/l (H-3)
190 170 - 430 180 - 280 180 - 220 50 - 270 90 - 280 Surface Water '

LT 0.5 LT 0.5 LT 0.5 LT 0.5 LT 0.6 LT 0.7 Fish pC1/Kg (Sr-90) pC1/Kg (Cs-137) 7.11 - 17.5 11.0 - 25.8 10.2 - 13.8 7.70 - 17.4 8.4 - 21.4 8.8 - 19.1 pC1/1 (II-3) 130 - 490 130 - 410 120 - 190 140 - 320 80 - 970 90 - 270 Rain Water pCi/l (Cs-137) m m 1.40 - 12.4 m m m 28 - 44 24 - 45 21 - 48 24 - 40 30 18 - 49 Noble Ga9es pC1/m3 (Kr-85) . LT 40.0 gri/mS (Xe-133) LT 11.0 LT 11.0 LT 11.0 LT 11.0 LT 34.0

Ranges are not given since only cme data point contdned an idsmtified isottpe.

11-24

SHORDIAM DSAR SHORDIAM DSAR TABLE 11.6.3-1 em Second Step Post-operational Radiological Fnviremental &nitoring Program (Ra m Media Sampling Locations Sanpling Frequency Analysis Direct IS1,2A2,3S1,4S1,5S2, Quarterly Gama Dcposure Radiaticn (1) 6S2,7A2,BA3,9S1,10A1, 11A1,12A1,13S3,14S2, 15S1,16S2,*5E2,*6El Fish and 3C1,14C1, *13G2 Semi-annually Gama-isotopic Invertebrates (2) or when in season Fruits, BB1, 6B21, *12H1 At time of Annual Gama-isotopic and Vegetables (3) Harvest Airborne 6S2,2A2,3S1,7B1,*11G1 Quarterly Gross-Beta Particulates (4) and Ga:me-isotopic Milk (5) 13B1,*10F1, or *8G2 Mthly during Garma-isotopic Grazing Season, Ortly. at all other times. Surface Water 3C1 or 14C1, and *13G2 Semiannual Gam a-isotopic Grab Sanple H-3 (*) Designates Control Iocations (1) Eighteen ronitoring stations, DR1 through DR18, (16 indicator and 2 control) are used. One indicator location is positioned in each meteorological sector near the site boundary. One dosineter or continuously measuring dose rate instrumnt is placed at each location. (2) At each Indicator location, one sarple of each corrercially and recreationally important species. One sample of sane species ir. control location. (3) Sa.ple three different kinds of broad leafy vegetables grown nearest to two indicator locations - having highest predicted average ground level D/Q (when milk samples not available) . Also take one sample of same leafy vegetation grown nearest to Control Iocation. (4) Three sanples (near SNPS), one from each of the three Meteorological sectors having the largest annually averaged ground-level D/0, are taken. One sample (near a comunity) also having the highest calculated annually averaged ground-level D/Q is taken. Establish one Control Iocation. (5) Indicator samples frun milking animals having highest potential dose. Sanple within 5 km distance (preferably), within 5 to 8 km where doses are e calculated to exceed 1 mrem /yr (second choice) or fran 8 to 17 km. Control I location is 15 to 30 km fran SNPS and in the least prevalent wind direction. P 11-25 Fev. 3 July 1991

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

s SHOREHAM DGAR SWRDRM DSAR TABIE 11.6.3-2 REMP SAMPLDG IDCATIWS DESIGATION IDCATIN

          '181           Beach east of intake, 0.3 mile [N]

2A2 West and of Creek Road, 0.2 mile [NNE) 3C1 Fish and Invertabrates, Outfall Area *, 2.9 milac [NE) . 3S1 Sita Boundary, 0.1 adle [NE) 4$1 Site Boundary, 0.1 mile [DE)

         *SE2            Calverton, 4.5 rA1mm [E]

5S2 Site Boundary, 0.1 mile [E] 6B21 Condezella's Farm Stand, 1.8 niles [ESE) f 86El LIICO PCW, 4.8 miles [ESE) , 652 Site Boundary, 0.1 mile [ESE) l 7A2 North Country Road, 0.7 mile [SE) 7B1 Cverhill Ibad,1.4 miles [SE) . BA3 North Country Ibad, 0.6 mile [SSE) ; local Farm,1.2 miles [SSE) Q BB1 Dairy (Ccw),10.6 miles [SSE) f

          *BG2 951           Service Road SNPS, 0.2 mile [S)                                   ;

10A1 North Country Road, 0.3 mile [SSW) ,

          *10F1           Goat Fam, 9.2 miles [SSR) 11A1          Site Boundary, 0.3 mi.le [SW)                                      ,
          *11G1           MacArthur Substation,16.6 miles [SW) 12A1         Metacrological 'Jbwer, 0.9 rile [WSW)
           *12H1          Backgrtund Farm, 26 miles [WSW)                                   ,

13B1 Gbat Farm,1.9 miles [W) .

           *13G2          Fish and Invertebrates, Background,13.2 miles [W) 13S3         Site Boundary, 0.2 rile [W)                                        #

14C1 Fish ard Invertebrates, Outfall Area, 2.1 miles [NNR) 1452 St. Joseph's Villa, 0.4 miles (W%') 15S1 Beach west of intake, 0.3 mile [NW) . i 16S2 Site Bourdaz'f, 0.3 mile [N%')

Desigrates control locations ;

t i 5 11-26 , f

S

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4 l l l y' Q l

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                                                                                                                               - u.. -2 OFF StTE SAMPLff00 LOCATICIES RAOJOLOGICAL ENVIRONMEMTAL MOMITORMOG PROGRAM
                                                                                                       -                    ' SHOREMAM NUCLEAR MMNER STATIOII sceu1=ius                                                        , DEFUELED SAFETY ANALYSIS REPORT
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l SHOREHAM DSAR

  ^'                                                               CHAPTER 12 RADIATION PROTECTION 12.1                           ASSURING THAT OPERATIONAL RADIATIO.N EXPOSURES ARE AS LOW AS REASONABLY ACHIEVABLE The Shoreham ALARA. Program, the intent of which is to maintain
                                               ^

operational radiation exposuresis (ORES) to levels described as low as in the USAR, is Section reasonably achievable (ALARA), 12.1. The program is applicable in its entirety to Shoreham in a defueled condition, with the following exceptions: A) Within the Nuclear Engineering Support Organization, the radiation protection function is found within the Nuclear Analysis function, as described in Chapter 13 of the DSAR. (Reference USAR Section 12.1.1.3.2, Review The basis of change is a of this Modification of Operations. reorganization of the Nuclear Engineering Support Organization.) B) Several of the items listed under Section 12.1.2, Design

                               , Considerations, are no longer applicable or have been Jl N-T revised.               Specifically, charcoal has been removed from the Radwaste vent filters (considerations 10 & 12),

the Radiation Monitorine System is now discussed in DSAR Section 12.3 (consideration 11), demin water hose stations are located on the radwaste building floor (consideration 7), shielded radwaste shipping casks and remote handling techniques are not generally used (considerations 15 and 16), and the discussion of gaseous radwaste sources in consideration 15 is no longer relevant, as the offgas systen is shut down. The (Reference USAR Section 12.1.2, Design Considerations. justification for lack of charcoal is the tact that as per DSAR Sections 11.1 and 12.2, Shoreham does not possess a meaningful quantity of radioiodines. The low levels of radioactivity in the solid radwaste do not justify using shielded casks and remote handling.) O V 12-1

c SHORERAM DSAR ]i I ( ' C) -All visitors within the Protected Area are escorted by ! qualified _ personnel. Those visitors requiring access to the [ radiologically controlled area (RCA) are given an ' appropriately abbreviated indoctrination in protection j against radiation, prior to their entry into the RCA. ) ,. J (Reference USAR Section 12.1.3.1.1, Use of Individual ! Personnel Monitoring Devices. The basis of this change is a ! more stringent application of security requircments for ! visitors to Shoreham). : i D) With the generally very low dose rates associated with the ! plant's defueled condition, there is no. longer a requirement [ to have all personnel (permanent and temporary) equipped with ! cpproved dosimetry devices upon their entry to the : radiologically controlled area (see Section 12.5.2.1, Access j Control. Rather, only individuals working on a Radiation ; Work Permit (RWP) are required to use approved dosimetry f devices. That the requirements of 10CFR20,202 are met by this approach will be assured by the ongoing station radiation surveillance program (as described in USAR Section 12.5.3.1), as well as the posting of thermoluminescent i dosimeters (TLDs) in' general access areas of the RCA. ! l

  .                        (Reference USAR Section 12.1.3.1.1, Use of Individual                                                                   i Personnel Monitoring Devices.                   This change is justified by                                              l I

(

the low dose rates seen presently at Shoreham, and by the !

very low historical man-rem data in Section 12.5 of the . DS AR. ) j t It should be noted that the Shoreham station's original physical ! design for radiation protection (e.g., shield walls, i penetrations, sample stations, etc.) remains generally unchanged from that described in the USAR, Sections 12.1 and 12.3. This is despite the fact that the actual source strengths and unshielded dose rates do not necessitate the degree of protection afforded. , The physical design is based upon the presumption of plant operations, with the associated source terms and unshielded dose rates as described in the USAR. Although they will not generally , be needed, operational considerations describad in the USAR (e.g, *

                    -the precautions for.high dose rate jobs -- in excess of 100 mrem /hr) will be maintained.                                                                                                 l 12.2    RADIATION SOURCES                                                                                                    j[

12.2.1 Contained Sources ! l Fuel Sources-

                    -The Shoreham reactor core has undergone three periods of low                                                                  [

power (0-5%) testing over the past four years. The low power ! tests are summarized below: l 12-2 Rev. 3 July 1991 k

t r SHOREHAM DSAR k Specific Burnup Power

Test Period Duration MWD /MT Range.%

7/7-10/7/85 93 days 27.8 0.0 - 3.3 8/5-8/30/86 26 days 13.8 0.0 - 4.0 5/26-6/6/87 12 deys 6.7 0.0 - 3.5 Total Tli T ; The detailed profiles of the above three low power test periods ; have been input to the ORIGEN2 (Reference 1) b..erup code, along with the physical characteristics of the react fuel and bundle l structural elements. Results of this analysis (heference 2) are i given in Table 12.2-1. The activities correspond to two and four years decay (June 1989 and 1991, respectively) after the last burnup period, and reflect total core inventories for those ; isotopes with greater than 10 curies. That the source strengths given in Table 12.2.-l are reasonable is evidenced by neasurements taken during defueling activities in ; 1989. Dose rate measurements vere taken at one foot from a number of spent fuel bundles, and the maximum values from each ! bundle were tabulated. Dose rates at one foot its a function of , bundle burnup) were calculated from the 1989 source terms l ' presented in Table 12.2-1, using the point kernel code OADMOD (Reference 5), and the resu3.ts compared to the measured dose rates. Results are given in Table 12.2.-2. That the calculated and measured bundle maximum dose rates agree within about 10% on s(e~s)average gives assurance that the calculated source terms in Table i l 12.2-1 are reasonably accurate. As can be seen from the **11e 12.2-1 only long-lived isotopes remain from d.: original tinides and fission / activation products created, along h their equilibrium daughters. By far the most radiologically ignificant, from a gamma dose rate standpoint, are the Cs-137/Ba-137m pair; about 80% of the whole ; body dose rate from a spent fuel bundle is dae to the Ba-137m photon (Reference 3). For dose assessment of accidental gescous releases (e.g., a postulated fuel handling accident), only Kr-85 is meaningful (Reference 4). 12.2.2 Airborne Radioactive Material Sources ? The statements ! low apply when systems are closed up. When potentially contaminated systems are opened, the RWP controls, as stipulated in DSAR Section 12.5, will minimize airborne sources. l

Zero power activities were intermittently performed in January 1989 for Reactor Operator training purposes. This operation is assumed to have virtually no impact on spent fuel inventories.

i,Q V 12-3 Rev. 3 July 1991 l D

                            ..  .=

SiiOREHAM DSAR (Or' Reactor Bu?' sing , There is no significant source of airborne activity assumed to exist in the reactor building in the plant's present defueled condition. t P b i s : i Ch LJ 12-3A Rev. 3 -7uly 1991 i

I i SHOREHAM DSAR h$ Turbine Buildinq } i There is no source of airborne activity assumed to exist in the ; turbine building. Radcaste Building . I j There is no signifzcant nource of airborne activity assumed to

            , exist in the radwaste building.                                                                        !

Purther discussion regar.'.ing airborne activity is provided in l sections 11.1 and 12.4. [ t REFERENCES General I, L Updated Safety Analysis Report (USAR) Shoreham Nuclear Power ! Station Revision 1, December 1987. i b ORIGEN2, Isotope Generation and Depletion Code, ORNL CCC-371, j 1.

                   .7/80, t
        ~

2. LILCO calculation C-RPD-476, rev. O, 10/21/88. ;

LILCO calculation C-RPD-530, rev. O, 05/19/89, f

3. !

4. LILCO calculation C-RPD-529, rev. O, 06/07/89. ;

CADMOD-G, Point Kernel shielding Code, ORNL CCC-396, 12/79. 5. r 12.3 RADIATION PROTECTION DESIGN FEATURES ; 12.3.1 Facility Design Features The description contained under this heading in the. latest I

revision of the Shoreham USAR remains unchanged as it is used to ;

develop the basic design' criteria of the Refer to the plant. the USAR for information on this subject. However, defueled ! j condition, with low activity levels, some design For features example, are not - necessarily utilized as described in the USAR. ' liquid filters in the radwaste system do notAlso, usually require the radiation i portable shielding or remote backwashing. ; rone designations shown on USAR Figures 12.3.1-1 through -35 are ; not applicable for the plant's present condition. : i 12.3.2 Shieldinq ! 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. l l 12-4 . t f I

SHOREHAM DSAR 12.3.3 Ventilation The description contained cader this heading in the latest revision of the Shoreham USAR remains unchanged. Refer to the USAR for information on this subject. 12.3.4 Radiation Monitoring Instrumentation In order to support the. storage of the fuel.in the fuel pool, SNPS will need process and effluent radiation' monitoring instrumentation, and area and airborne radiation monitoring instrumentation. Process and Effluent Radiation Monitoring 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, and Table 12.3.4B provides data for each monitor. O Normally, nonradioactive systems that may become significantly Aq,) contaminated by leaks from radioactive systems are monitored continually to ensure that no condition hazardous to theFor operating personnel or to the general public develops. effluent streams that discharge to the environs, sample points are located downstream of the lar point of possible radioactive fluid addition to the effluent be.ng 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 effluent relsase points from the station that require radiation monitors: the station ventilation exhaust, the liquid radwaste effluent, and the reactor building salt water drain tank. Area Radiation and Airborne Radioactivity Monitoring Instrumentation This section contains a description of the area All channels and have airborne local readout by radiation monitoring systems. 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. Theand area r ,O monitors are provided with an audio and visual alert high l radiation alarms. Monitors are placed in areas where personnel l 12-5

SHOREHAM DSAR

       \ ,)     normally have access and where there is a possibly that radiation Itvels could become significant.                                                                .

All airborne monitors are offline monitors and ars designed in accordance with ANSI N 13.1-1969, ' 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 acceesible area. Airborne radiation monitoring is provided where potentially radioactive sources exists. Each of these monitors is provided with an isokinstic nozzle which is sized to ottain a representative air sample at the normal flow in the ventilation duct from which the sample is teken. Table 12.3.4B lists the airborne monitors, and Table 12.3.4C lists the area monitors. Radiation Monitoring System Computers The RMS is eculpped with redundant computers powered from U.P.S. 1297-INV-005/TSC Black Battery /69 kV primary feed. These units provide continual surveillance forCommunication all airborne, area, process, with the computer and effluent radiation monitors. is through keyboar6 equipped CRT displays in the main control room, the health physics office, the process computer room, and ((o} the technical support center, inservice Inspection, Calibration, r .a Maintenance The operability of each channel of the area and af.rborne 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 specificat.cns. Calibrction of all monitors is norr. ally conducts 1 at an interval of 18 months unless mandated sooner by Technical Specification. l This calibration will allow indication in a low, mid, and high response range of each monitor. 12.4 DOSE ASSESSKENT l f 12.4.1 Design objectives The design of the shielding was originally based on conservative estimates of the occupancy time required in each area of the 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 condition. Table 12.4-1 lists the six zene designations that vere originally established, along with the maximum allowable Os dose rates and estimated occupancy times for each area. With the 12-6 Rev. 3 July 1991 1 - .

f 1 SHOREHAM DSAR > ( plant in its present condition, with spent fuel stored underwater ! j in the pool, there are no occupiable areas whi7h are zone III or higher. ; i { I t i k

I L k I I- i e i I t I I l t i i L I t i I r i. l l, l & I i 12-64 Rev. 3 July 1991 ! l . . . . .. , - - . , - . , .,

                                                                       - - , ~     . . . - -
                                                                                                       - . - . - - - , . . ~ . - . . . - - - . . . . - . - . - - . . ~ . . . . - - . - ,            -

~ - - __ - __ - __ _ _ _ . SHORERAM DSAL 12.4.2 Airborne Activity An area within the Shoreham f acility is described as an *airbort.e radioactivity area' if the sum of the concentrations of all airborne radionuclides divided by their respective Maximum Permissible Concentrations (MPCs) (from 10CTR20, Appendix D, Table 1, Column 1) exceeds 0.25. At Shoreham, there are noWith

           " airborne radioactivity areas" in the defueled condition.

the fuel in the spent fuel pool, and insignificant quantitiesitof radioactive material elsewhere (see Sections 11.1 and 12.2), is not expected that airborne radioactivity areas will exist in the future, unless systems which are currently In anticipated to this instance, remain closed are opened to the atmosphere. 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-brs per day and/or 10 MPCa-brs per week, paragraph 103(a) (3) of 10CTR20 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 taken to be essentially zero 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 actier.s, including investigation, when any positive whole body count occurs in excess of it of the maximum permissible organ burden (MPOB), or it of the maximum permissible body burden (MPBD). 12.4.3 Occupational Dose Assessment occupational dose at Shoreham is expected to be essentially zero for the defueled condition. This conclusion has three bases:

1) At present, the dose rates in occupiable areas are virtually 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.

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 1cvels between 1988 and 1990.

(~ 3) The recent collective station dose history at Shoreham is as (TLD data collected in response to the n quirements ds /T follows of 10CTR20.407): 1 12-7 Rev. 3 July 1991

SHORIllAM DSAR Time Period Dose, man-rem 1/1/86 - 6/30/06 0.562 7/1/86 - 12/31/86 3.123 1/1/87 - 6/30/87 0.341 7/1/87 - 12/31/87 0.065 1/1/88 - 6/30/88 0.050 7/1/88 - 12/31/88 0.000 1/1/89 - 6/30/89 0.020 7/1/89 - 12/31/89 0.075 l Since February of 1987, when a change was made from H. S. Landauer to Panasonic TLDs, doses have been insignificant, and due almost entirely to small statistical fluctuations rather than actual doses. Based on the above statements, it is anticipated that occupational dose at Shoreham will be essentially zero until As such, the 1991 physical decontamination activities commence. corporate ALARA goal was established at 2.5 man-rem. Doses will be measured as indicated in the Health Physics Program, Section 12.5. t 12.4.4 offsite Dose Assessment

                       ) There are no sources (eg, N-16) in the defueled condition which, f(                  under normal (non-accident) conditions, could lead to offsite                                                  *'

direct doses, either by direct radiation or "skyshine", based on the 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 be confirmed by the RERP, as described in Section 11.6. 12.5 HEALTH PHYSICS PROGRAM The Shoreham Health Physics Program, the intent of which is to provide for the protection of all permanent and temporary personnel and all visitors from radiation and radioactive i materials in a manner consistent with Federal and State regulations during all 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 j A) Handling of new fuel is no longer applicable to Shoreham. l (Reference USAR Section 12.5.1.2, Personnel Experience and Qualifications. The basis of this change nois

that with the Settlement Agreement with New York State, + new fuel will be brought onsite.) D . h 12-8 .Rev. 3 July 1991 l _. _- - - - --- _ . -. =

SHOREHAM DSAR 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 ka 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. (Reference USAR Section 12.5.2.1, Location of Equipment, Instrumentation and Facilities. The basis of this change is the fact that with no airborne areas currently identified, and none expected in the defueled condition, 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, Types of Detectors and Monitoring Instruments. Justification of this change is due to the near total decay of radiciodines at the site.

E) Radiation Work Permits are 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. Work in a posted contaminated area (see Item F below).

4. Entry into airborne radioactivity areas.

5. Breach of a radioactively contaminated system boundary.

l (Reference USAR Section 12.5.3.2, Radiation Work Permits. The basis of this change is a change to station procedures. l F) Under the discussion of access control, add the definition of l a contaminated area: I l 12-9 Rev. 3 July 1991

S110REllAM DSAR () Contaminated Area Any area having removable beta / gamma-emitting' radioactive material in excess of 1000 dpm/100 sq cm, or alpha-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 INPO 85-001, eiv.1.) M Under the discussion of access control, the " secondary access facility" no loncer exists. (Reference USAR Section 12.5.3.3.1, Access Control. The basis for this change is that as of September 1, 1989, the secondary access facility was taken out of service.) H) The Corporate ALARA Review Committee (CARC) now I administrative 1y reports to the Vice President, Office of Corporate Services and Vice President, Office of Nuclear. (Reference USAR Section 12.5.3.3.4, Post-operations Review. '() The basis for this change is an organizational change. See

N> Chapter 13 of the DSAR for further details.) I) As 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 Training Program. For justification, see DSAR Section 12.1.3.1.1.) 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 pources of radioactivity (during physical decontamination activities) varrant that the procedures and commitments remain in place. 'O 12-10 Rev. 3 July 1991

SHORERAM DSAR k TABLE 12.2-1 - Fuel Source Terms i ISOTOPE CURIES HAL1-LIFE 1989 1991 Inventory Inventory . H-3 1.77E+02 1.58E+02 1.23E+01 years Mn-54 3.36E+01

3.13E+02 days ,

Fe-55 8.06E+02 4.82E+02 2.70E+00 years ' co-60 5.64E+02 4.34E+02 5.27E+00 years Ni-63 4.2BE401 4.22E+01 1.00E+02 years Kr-85 1.56E+03 1.37E+03 1.07E401 years ; 1.54E+01

5.0$E+01 days Sr-89 Sr-90 1.37E+04 1.31E+04 2.86E+01 years :

Y-90 1.37E+04 1.31E+04 6.41E+01 hours 6.81E+01

5.85E+01 days Y-91 ,

t Zr-95 1.48E+02

6.40E+01 days ,

3.49E+02

3.51E+01 days !

Nb-95 Ru-106 5.98E+03 1.51E+03 3.68E402 days Rh-106 5.98E+03 1.51E+03 2.99E+01 seconds Sn-119m 3.30E+02 5.86E401 2.93E+02 days , ( Sb-125 1.45E+03 8.79E+02 2.77E+00 years . Te-125m 3.53E402 2.15E+02 5.80E401 days 1.49E+01

9.35E+00 hours Te-127 Te-127m 1.52E+01
1.09E+02 days Cs-134 1.33E+02 6.79E+01 2.06E+00 years Cs-137 1.48E+04 1.41E+04 3.02E+01 years Ba-137m 1.40E+04 1.34E+04 2.55E+00 minutes Ce-144 3.55E404 5.98E+03 2.84E+02 days Pr-144 3.55E+04 5.98E+03 1.73E+01 minutes .

Pr-144m 4.26E+01 8.55E+01 7.20E+00 minutes i Pm-147 2.95E+04 1.74E+04 2.62E+00 years Sm-151 3.60E+02 3.55E+02 9.00E+01 years Eu-154 1.18E+01 1.01E+01 8.80E+00 years ' Eu-155 4.47E+01 3.3BE+01 4.96E+00 years U-234 1.02E+02 2.04E+01 2.45E+05 years !' Th-234 3.38E+01 3.38E+01 2.41E+01 days Pa-234m 3.3BE+01 3.3BE+01 1.17E+00 minutes U-238 3.38E+01 3.38E+01 4.47E+09 years Pu-239 2.77E+02 2.77E+02 2.41E+04 years ! Pu-241 5.58E+01 5.07E+01 1.44E+01 years (3 Total 1.76E+05 9.07E+04 ,. i~\) Note: Only isotopes with activity greater than 10 curies are , listed. ,

Decayed Away Rev. 3 July 1991 ,

_v . -- mw w r -

                                                                                                                                                                                                                              .pm ,

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I I

l S!!OREMAM DSAF r

    - b)                                                                                                   TABLE 12.2 '                                                                       -

I l Comparison of Measured (maximum) vs. Calculated Spent Fuel Bur 31e Dose Rates

j Cell Burnup Dose I ntes, rem /hr Measured /

Date Number awd/st** Measur_f.$ Calculated Calculated l i 07-46 0.0045 0.4 0.4 1.00 l 7/14/89 0.4 1.75= ! 7/19 01-32 0.0047 0.7 34 0.0385 5.1 3.5 1.46 l 7/21 3.9 1.41 ; 7/24 25-48 0.0421 5.5 45-10 0.0126 1.6 1.2 1.33 7/26 0.9 0.8 1.13 17/26 47-14 0.0082 15-44 0.0398 3.2 3.7 0.86 . 7/26 6.0 1.00- l 7/28 43-30 0.0547 6.0 21-44 0.0516 5.3 5.6 0.95 l 7/29 5.6 0.68

                                      .7/31                                    09-34        0.0514                             3.8 39-16        0.0607                             6.0                                   6.6                        0.91                                  )

8/01 6.0 6.8 0.88 j

                                      .3/01                                    39-32        0.0624 average                           1.11                                   ;

in water at 12 inches from fuel bundle {

                                           ** gwd/st a gigawatt-days /short ton i

i I i

                                                                                                                                                                                                                                     ~f r

I t l i t

1. .

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

 ._ __   .._______-__._______._.___m.__.._._.____m.___..                                    .-

SHOREHAM DSAR ! O TABLE 12.3.4A : PROCESS AND EFFLUENT MONITORS i

PM 13 Liquid RW Discharge . PM 21 Low Range Station Vent Monitor PM 29 Gas Peactor Building Vent .

               .            PM-30'.Part                  . Reactor Building Vent-                  '

PM 41 Part Station Vent PM 42 Gas Station Vent PM 55 Gas Radwaste Vent " PM 56 Part Radwaste-Vent . PM 79 Saltwater Drain Tank

4 O : a T

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            \*)                                                                                                .J
                                                                                                                 )                                                                                                  N(\

Q

SIDFDW1 DSAR intz 12.3.4n - IMTA R1R UTUT7T #D T1003l5 RADIATKN Mh6 Ptwf tw Fayr (2) Actim P&mul geratim Pimitorirg Bunctim Incatim Type of ?twitor Scruitivity Taken en Alarm Statim vmtilntim Ddast Ptmitor ffm1 relex:e for ikwmtrena of tie last Offline pts 10 tCIfcc. Investi ete ==I all hw level pt9ms point of activity Particulate p (Kr-85) correct cmae of hish activity etthesits to the min plant vmt aircrean release rate enreed-irt iman-m-teclnical specifica-- tien done rate limit. IJguld Fahrste Ptnitor activity relea9e Effhwsit pipe prior Offilne Ligrid 10 tC1/cc twh-,cic clemrre of rate during plarrd to dL% into the (Ce-137) Ifqzid wete dischnrme Effhrnt 13 quid wete distforge cirudatirg mter systna ulve enre dira 100R20 periods 1buits to - a c a ictal Steful. Fer:cter niilding Salt Pinitor nctivity relce of Ikwmtress of the Offline Ligrid 104 tC1/cc Imestigte and correct water Prain Tark service witer drained durire collecticn tzA (Ce-137) cmune of hish activity rninterace relenne rate exceeding 100R20 limits to

                                                                                                                                                                                  -amhicted arms Emetor niildirm                                               Ptnitor all kw level          Effheit dtet prior     Offline pe        10 tC1/cc                                  IrmatiJutte mui crrrect Ventilatim                                                     pesarms efflumes in the      to diminrne into the   Fartio11 ate      puses (Kr45) casse of hish activity renctor b1dg vmtilatim       station vent                                                                         releene rate duct Faheste Thrilding                                             finitor all kw level          Efflumt duct prior      Offline pm        104 tCI/cc                                 Irwestipte erl correct vent flatim                                                    pramm effluents in the        to dLh into the        Particulate      games (Kr 85) caisse of hish activity r:whente b1dg. vmtilatim      statim vent                                                                          release rate.

dict.

e em (- s us m yri\ u ) TAME 12.3.48 TATA R11 UTTIT?.T #O Nx, RADIATIOt MhTITES Mmfter R1rme (2) Actim Trpe of Mnitor Sen=itivity Tahm m Alam Mvtitorirg Puticn Isratim Ibet-ket h e Dvr:tren of tb Iast Offlisv sa: 10 tCf/cc N/A I- Statim vmttfatim Rnitor fimi releam for Particulate & M) DJrust (im Rw) all met effitets to point of xtivity the min pimt vmt air-sttwri derrirg m accidet (1) Orw' init inless STecified M** (2) py meric rnre is a en of 4 decades abe semitivity. i t i 4 1 1 1

1 . L - O m'O i, ; i: 1 NEE 12.3.4C ARFA MNITUtS t il ' i Area finitor AlertjHIph Setpnints Ramy* (d) (sur-rii/hr) Imcatimt i 1D21-FM101 Flocr Drain Sisap Tardt 5/100 0.1-10nD Remetty 514. el 8-0 '. '

           -010   Be1 rool CIcanup Ptsge                5/ICU          0.1-1000     Reactor B14. el 112-9                              l 0.1-Inno     Reactor Eldet. el 130-9                            i
           -012   Mel Ibol Fqdgwnt Area                 3/100
           -013   Centmimtal Fgdp. Sternss-             5/100          0.1-10nD     Remeter B1% el 130-9
           -014   Rw 1 Storage Pbol                     5/10           0.1-1000     Mescter B1dg. el 175-9
           -015   Reactor mal Irvictlation Stornre      5/100          0.1-10nD     Rentter Bldg. el IM
           -022   Owwistry I.Lokmy Mezzmine             1/3            0.01-100     Bester Bay el 31-0
           -024   Rahev:te B1dg. Dramtadmtim Area       5/100          0.1-1000     Rahseite B1dg. el 15 4
           -026   Storage vaults for Omtainers          10/1n0         1.0-10'      Rahmete B1dg. el 15-6
           -027   Sanple Roam                           20/100-        0.1-1000     Rabsiste Eldg. el 37-6 Rabsuute B1dg. el 37-7                             i
           -028    F1mr Drain Filter Am                 5/100          0.1-10re
           -029   Rahstste Filter and redneralizer Area 5/100          0.1-1        Rahswete 51dg. el 37-6                             f 1
           -032   Faheste BIdg. Dedneralizer Arnt       20/100         0.1-1        Rahauste Bldg. el 15-6      .
           -033    Raheste Bldg. Bolst Aret             3/100          0.1-1         Rmhaete B1dg. el 304
           -035                                         5/100          0.1-1         Maetter Bldg. el LLO Fed Iwnt Drain Tank Area
           -0% TIP Prive Rotu                           5/100           1.0-1        Renctor B1dg. el 75-7
           -037 TIP Drive Area                          5/100          0.1-1         Resctor 31dg. el 78-7
            -038 New 1%el Storap.e Aren                 2/10           0.1-1         Reutter Bldg. el 17'P9                            y f%2 Fahawte Bldg. Ptsap Area               3/100          0.1-1         Radismete B1dg. el 15 4                           i t

f i t i I i

SHOREHAM DSAR to + SHOREHAM DSAR 7pble 12.4-1 RADIATION EONES (Original plant design basis)

                                                  'Maxianm        Estimated Allowable      occupancy Dose Rate       Time tone              tone Description            (mrem /hr)   (hr/wk1 pesignation    _

Unrestricted Area - less than Unlimited I continuous Access lD . 2 II Unrestricted Area - less than 50 Periodic Access 2 III Restricted Area - less than 20 () Controlled Trequent Access 5 IV Radiation Area ~ less than 5 Controlled Infrequent 20 Access less than 1 V Radiation Area - Controlled Infrequent 100 Access VI High Radiation Area - greater than - Not Normally Accessible 100 0

i SHORERAM DSAR , CHAPTER 13 (( ) I CONDUCT OF OPERATIONS i i 13.1 ORGANIZATIONAL STRUCTURE OF APPLICANT The description contained under this heading in the latest revision of the Shoreham USAR changed to be as described below. 13.1.1 Corporate organization ; A) The office of Nuclear organization is shown on DSAR pigure ! 13.1.1-2. Executive responsibility for the management of the i Shoreham Station is exercised through the Vice President, 5 Office of Corporate Services and Vice President, Office of Nuclear. The Vice President, Office of Corporate Services and Vice President, Office of Nuclear has corporate ! responsibility for overall plant nuclear safety and authorit" ! to take such measures as may be needed to ensure acceptable ! performance of the staff in maintaining and providing l technical support to the plant to ensure nuclear safety. He l reports to the President, Chief operating Officer, who These is [ responsible to the Chairman, Chief Executive Officer. operations are discharged by the Plant Manager, Shoreham i Nuclear Power Station, and the Managers of the Nuclear l l IO Operations Support and Nuclear Engineering organizations. Supplementary technical support is provided to these i l f organizations under the direction of the Vice President, ' Office of Corporate Services, and Vice President, Office of Nuclear by other LILCO organizations through appropriately ; defined Nuclear Operations Corporate Policies. l [ sports ; B) The Manager, Nuclear Quality Assurance (NOA' > l functionally to the Vice President,0ffice c. .arporata ' Services and Vice President Office of Nuclear and maintains direct access to the President of the Company as he deems { necessary. l C) The Nuclear Engineering organization has been modified to l [ reflect a reduced level of activity in the defueled s condition. It no longer includes an Engineering Assurance l l, function. ! D) The Safety Engineering and Reliability crganization within i the NQAD ic eliminated. This includes the ISEG and Reliability Sections. E) The Director, Office of Training and the Manager, Nuclear Emergency Preparedness Division report to the Vice President, 13-1 Rev. 3 July 1991 i

 .__.____._m_--.                                                                                                         , , . , , , , _ , , . . _ . . . _ _ . _ . - . .    . . _ . . - _ . - . . _     _  .. ._.     -                            -

SHOREHAM DSAR Office of Corporate Servicos and Vice President, Office of Nuclear. , F) DSAR Figure 13.1.1-1 shows revised direction of executive responsibility. 13.1.2 Nuclear Operations Support Organization The Nuclear operations Support organization consists of the following organizational units: t O . l l l l l l 10 13-1A Rev. 3 July 1991

                     ,n.. , ,, , , ,

r SHOREHAM DSAR Licensing, Nucicar Contracts & Material Controls, Security, and l , j Nuclear Financial Services. These units have as many staff specialists as required to support Shoreham. l t Nuclear Operations Support personnel provide expertise for ! supplementary support functions such as licensing and regulatory f activities, including assessing evolving regulations, managing l all nuclear litigation and evaluating regulatory documents for j impact on plant design. Other responsibilities cover cost t control, estimating, budget and cost administration for the . Of fice of Nuclear, nuclear records management and administration : of site clerical administrative personnel. The units are also i responsible for nuclear contract development and administration, administration of site warehouses, spare parts and inventory control. The responsibilities of Security are described in the Security Plan. 13.1.3 Nuclear Engineering Organization j The Nuclear Engineering organization consists of the following > units: Nuclear Systems, Nuclear Analysis and Nuclear Projects and Administration. Responsibilities include, (1) Systems, Mechanical & I&C Engineering, (2) Procurement Support, (3) General administrative support for procedures, training and document control, (4) Nuclear Analysis Support for the following . technical functions: Radiological Engineering & Health Physics, '! Radiological Monitoring Program (REMP), and Engineering and [ Nuclear Puels Engineering. l l The Nuclear Engineering organization also coordinates work [ performed by off-site support Engineering which includes [ Corporate Engineering and outside contractors. Outside [ contracturs include the original plant architect - engineer and [ i the NSSS Vendor. 13.1.4 operating organization The Shoreham Nuclear Pows: Station organization, as shown in ! Figure 13.1.1-2, consists of 4 units: [ i Operations, Maintenance, Radiological Controls & Operations Staff l 13.1.4.1 operations Operations is responsible for complying with the rules and ! regulations of the governing regulatory agencies and the l l monitoring of the station performance. It is composed of ; Operations, System Engineering, Modification Engineering and Work Planning & Scheduling. [ t 13-2 Rev. 3 July 1991 f i _ , , , , . . _ , - . . - - _ _ _ . _ . .. . - - . , . . . - _ , ~ , , . _ - . , . , _ , . _ . . , _ ,

___~._ .- ____ _- - - . - - - - . - - - . - - - - - ..- S!!OREHAM DSAR Operation activities of this unit primarily consist of the ( routine operation of the station systems and equipment. l l The Systems Engineering unit is part of operations and its i responsibilities include rapidly providing specialized, , operationally oriented technical expertise in station systems and ' equipment. Systems Engineering also performs support functions for other operating organizations as appropriate. The Modification Engineering unit is responsible for the coordination and implementation of station modification activities including post-modification testing and system return to service. The Work Planning and Scheduling unit performs planning and scheduling associated with plant activities. 13.1.4.2 operations Staf f The operations Staff unit consists of Administrative Support, Compliance, and Fire Prottetion and Safety. e The 7dministrative Support unit provides station administrative , support for procedures and document control.

,()       The Compliance unit implements the station surveillance programs and reviews surveillance activities to ensure compliance with the station's Technical Specifications.

Fire Protection and Safety is responsible for implementing the Plant Fire Protection Program and for coordinating The the activities of the Fire Brigade and the Site Safety Committee. Supervisor holds the position as Fire Protection Program Manager responsible for maintaining compliance with applicable Federal, State, and local government regulations regarding station fire protection and personnel safety. 13.1.4.3 Maintenance Maintenance is responsibic for maintaining the Station's mechanical, electrical, instrumentation, and computer systems. The Instrunent and control unit is responsible for the l calibration, maintenance, and testing of instruments and control l and computer systems in the nuclear power station. The Maintenance unit has a staff experienced in mechanical and electrical maintenance of large steam-electric generating stations. Additionally, it can be supplemented with additional competent maintenance personnel from other LILCO power stations or organizations, or outside contractors, as may be required. 13-3 Rev. 3 July 1991

SHORERAM DSAR 13.1.4.4 Radiological Controls ( } Radiological Controls is made up of Health Physics, Radiochemistry, and Radwaste units. Radiological Controle is responsible for the protection of the public, station personnel, and the environment from the effects of exposure to radiation. It assures that the radiation doses of station personnel and the public are maintained as low as is reasonably achievable (ALARA). Radiological Controls is also responsible for detection and control environmental releases f rom the station, and station chemical and radiochemical activities. In addition, Radiological Controls is responsible for the proper processing, packaging, storage, and shipment for burial of radioactive waste. 13.1.5 pualificat ion Requirements for Station Personnel This section is revised from that in the USAR. All responsible station personnel, both supervisory and non-supervisory meet the requirements of ANSI 18.1-1971. 13.2 TRAINING PROGRAM 13.2.1 Program Description The purpose of the accreditation program is to assist INPO member utilities in maintaining training programs that produce well-qualified, competent personnel to operate the nation's nuclear power plants (INPo 08-001). In the defueled state, with the NRC operating license amended to remove operating authority, there is no requirement to maintain accredited training programs since the plant is no longer licensed to operate. The Office of Training has non-nuclear training programs available, developed via a " systematic approach to training" method, which can be requested by the Shoreham plant management for training of operators, technicians, and meet. nics. The Office of Training procedures outline the methods to be used to analyze training needs, and to establish or conduct required training. The Office of Training staff will be qualified in accordance with the " Training and Qualification Program". O 13-4 Rev. 3 July 1991

SHOREHAM DSAR , Operators: Operators will be trained 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 licensed operator training program developed for . l nuclear operations. , Equipment Operators rield operators will be 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 those 4 systems required to be operational during the defueled phase. control Technicians: Control technicians and-computer technicians will be trained in accordance with the Control Technician training program developed for power plant , technicians. i Hechanics/ Electricians: LILCO mechanics / electricians attend ! formal training as part of LILCO's maintenance training programs. i These programs qualify mechanics / electricians as apprentices with i journeyman qualifications available in the area of welding, ; rigging, machinery, electrical, and general maintenance skills. i The Shoreham maintenance force will be trained and qualified in ! accordance with existing LILCO maintenance training programs. t This program is not available for contract maintenance work O- forces; contractors would provide qualified mechanics and i electricians. l Rad Chem / Health Physics: The Radiochemistry and Health Physics technicians will be trained using the training material developed , for Health Physics and Rad Chem technicians for nuclear ; operation. However, the training will be limited to fundamentals I and task specific training as required to support Rad Chem, Health Physics, and Radwaste operations during the defueled condition. 13.3 EMERGENCY PLANNING The emergency plan for the Shoreham Nuclear Power Station is submitted as a separate document entitled, "Defueled Emergency Preparedness Plan". 13.4 REVIEW AND AUDIT The description contained under this heading in the latest revision of the Shoreham USAR remains unchanged except as described below. 13.4.1 Review and Audit - Construction () No Change. 13-5

SliORElWi DSAR ( 13.4.2 Review and Audit - Test and Operation A) In 13.4.2.1, change the ROC membership, alternates and quorum requirements as follows: Membershipt A chairman, alternate chairman and four members of the Plant Staff as designated by the Chairman. Alternate members will also be designated by the Chairman from among the Plant Staff. Quorums The Chairman or his designated alternate and two members or alternates. Only one alternate shall participate in a vote in ROC activities at any one time. B) USAR paragraph 13.4.2.2, Nuclear Review Board (NRB) , is revised as follows: The function of the NRB is to provide the management of Long Island Lighting Company, through the Vice President, Office of Corporate Services and Vice President, Office of Nuclear, a inechanism for independently ascertaining that activities related to the nuclear station are performed safely and officiently in accordance with company policies and regulatory requirements. its initial The NRB is established and functionalt (D membership comprised LILCO and consultant personnel. Collectively, the membership has been selected to have the experience and capability to function effectively in the areas of responsibility as designated in license documents. The objectives are to ensure that a representative Office decision of is reached on each issue and that Vice President, Corporate Services and The ViceNRB President, membershipOfficeis of Nuclear selected so is appropriately advised. that a majority of members are not directly responsibile for plant activities. All members, whether LILCO employees or consultants, are afforded equal voting statusOffice alongof with a defined route to advise the Vice President, Office of Nuclear, of Corporate Services and Vice President, the assessment of dissenting voters.

1. Written Charter A written charter has been prepared covering such areas as group responsibility, subjects requiring review, reporting requirements, and organization.

t l 13-6 Rev. 3 July 1991

SHORERAM DSAR The charter of the NRB reflects the consideration that (.O' NRB activities are not limited to items and that are designated as safety related. It is intended functions that NRB review and audit activities will also cover nonsafety related structures, systems, components, and plant computer software to ensure that the safety significance given to them in the DSAR, the Technical Specifications, and the Emergency Operating Procedures will be maintained during the operation of Shoreham. l 2. Membership The NkB will consist of theAs NRB Chairman a group, theyand will at least four permanent members. collectively havo the competence reqvsyed to review problems in the following aroast nuetear anyAreering, chemistry and radiochemistry, radiologic >i safety, mechanical and ciectrical engineering,.anc QA practices. The Chairman will be appointed by the Vice President, office of Corporate Services end Vice President, office of Nuclear. The Chairman of the NRB is responsible Membership for appointing individuals to NRB membership. appointments are to be such that the collective membership includes the experience and capability noted in the foregoing subsection. Membership appointments are f\ subject to concurrence by the Vice President, Office of Corporate Services and Vice President, Office of Nucicar. In the event a regular member is not able to participate in NRB activities, designated alternates are Any authorized nominated to act in the place of the regular member. alternates shall be appointed in writing by the Chairman of the NRB to serve on a temporary basis. The NRB may obtain recommendations from scientific or technical personnel employed by LILCO or other consultant organizations whenever the NRB Chairman considers it necessary to obtain further scientific or technical assistance in carrying out its responsibility. Such individuals shall function as staf f to the NRB, performing tasks and submitting reports as assigned by the action of the NRB. Minimum qualifications of HRB members are as follows:

a. The Chairman will be a college graduate or equivalent and will have at least 10 years of experience in the power generation field.

b. Other members of the NRB and their designated alternates will be graduate engineers or equivalent (O and will have at least 3 years experience in the 13-7 Rev. 3 July 1991

Sil0REllAM DSAR app opriately related scientific, technical, (() engineering, or power generation field. Hemuers, their designated alternates, may possess competence or in more than one specialty area.

c. If sufficient competence in the specialty areas as described in this subsection is not available within LILCO, the review and audit functions will be performed or supplemented by outside consultants or organizations.

The minimum quorum cf the NRB necessary for the performance of review and audit functions shall consist of the Chairman (or his designated alternate) and at least three members, including alternates. Less than a majority of the quorum shall have line responsibility Afor the operation of the Shoreham Nuclear Power Station. quorum shall be considered filled if conference telephone communications are established with the requisite number of members or alternates at remote locations. No more than two alternates shall participate as voting membern in NRB activities at any meeting.

3. Meeting rrequency The NRS shall meet at least once per six months.

() Any member may request a special NRB meeting to consider a matter believed to involve a safety or rediological environmental problem.

4. Records I

a. Minutes shall be recorded for all meetings of the NRB. The minutes shall identify all documentary material reviewed and the finding:, recommendations, and actions taken by the NRB. Meetings shall be ,

numbered in sequence, and minutes of meetings shall be distributed to the president; the Vice President, Office of Corporate Services and Vice President, 3 Of fice of Nuclear; and NRB members within 14 days following each meeting. ,

b. Reports of audits rubmitted to or conducted under the '

cognizance of the NRB, including recommendations of the NRB, shall be made in writing to the Vice t President, Office of Corporate Services and Vice President, Office of Nuclear; and to the management positions responsible for the areas audited within 30 l days after completion of the audit. 13-8 Rev. 3 July 1991 h

                                 $1!OREHAM DSAR

5. Review Responsibilities The NRB shall reviews

a. The safety evaluations for (1) changes to equipment or systems and (2) tests or experiments completed under the provision of 10CTR Section 50.59, to verity that such actions did not constitute an unreviewed safety question. ,

b. Proposed changes to procedures', equipment, or systems that involve an unreviewed safety question as defined in 10 CrR, Section 50.59.

c. Proposed tests or experiments that involve an unreviewed safety question as defined in 10 CTR, Section 50.59.

d. Proposed changes to the shoreham Technical Specifications or the Shoreham Station operating License.

e. Violations of applicable codes, regulations, orders, Technical. Specifications, license requirements, or

 <                internal procedures or instructions having nuclear 8\               safety 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.

1. Reports and meeting minutes of the Shoreham ROC.

6. Audit Responsibilities Audits of Shoreham Station activities required by Technical Specifications shall be performed under cognizance of the NRB. These audits shall encompasst I

I i

a. The conformance of station operation to provisions contained within the Shoreham Technical Specifications and applicable license conditions at least once per 12 months.

 >O 13-9

l i SHOREKAM DSAR l r P b. The performance, training, and qualifications of the entire station staff at least once per 12 months. i (

c. The results of actions taken to correct deficiences !

I occurring in station equipment, structures, systems, or methods of' operation tant affect nuclear safety at least once per year,

d. The performance of activities required by the QA Program to meet the criteria of 10CTR50, Appendix B, 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 independent fire protection i consultant, j

f. Any other area of station operation considered l appropriate by the NRB or the Vice President, Office i of Corporate Services and Vice President, Office of Nuclear. ,

f g. The radiological environmental monitoring program and ! the results thereof at least once per 12 months. l (t *[

h. The Offcite Dose Calculation Manual and implementing procedures at least once per 24 months. l

i. The Process Control Program and implementing procedures for solidification of radioactive wastes at least once per 24 months.

i

j. The performance of activities required by the Quality l Assurance Program for effluent and environmental ;

monitoring at-least once per 12 months. 1 l

7. Authority .

i The NRB is organizationally responsible to the Vice ! President, Office of Corporate Services and Vice l President, Office of Nuclear. ! 1

8. Procedures j Written administrative procedures for the operation of the NRB will- be prepared and maintained.

Those items submitted to the NRB as described in O ! Paragraphs 5(b) through 5(d) above, reviewed by, and- . iv accepted by the NRB will be resolved as follows: l t 13-10 Rev. 3 July 1091 I .= _ _ ._ _ _- _ _ _ . . . .____._____.__i

SHOREHAM DSAR

a. If the NRB is of the opinion that a proposed change, test or experiment does not require approval by the NRC under the terms of the license provisions, it so reports in writing to the Plant Manager, together with a statement of the reasons for its decision.

The Plant Manager may then proceed with the change, test, or experiment. b'. If the hRB is of the opinion that approval of the NRC is required, the Shoreham Nuclear Power Station staff, assisted by other LILCO nucitar organizations or by concultants, shall prepare a request for such approval, including an appropriate safety analysis in support of the request in accordance with approved procedures. If, in the course of any additional reviews of facility operations, the NRB determines that a variation from the Technical Specifications or an unreviewed safety question exists, the NRD shall immedie.tely notify the Plant Manager, who shall take the necessary steps to ensure nuclear safety. 33.4.3 Shoreham Independent Safety Engineering Group I is The Shoreham Independent Safety Engineering The ISEG was Group (ISEG) required to be eliminated for the DSAR Phase. established by NUREG-0737, TMI ActionThe PlanISEG Requirements by each was an independent applicant for an operating license. organization dedicated to improving plant safety through examinations, reviews and audits of plant operations, modification, maintenance and operating characteristics, and NRC and other industry sources of plant design and operating experience and information that may indicate areas for improving plant safety. During the DSAR Phase, Shoreham will not be operated and the fuel will remain in the spent fuel pool until removed from the plant. In the Shoreham defueled configuration, only maintenance and minor modifications will be performed. The principal function of the ISEG, to improve plant safety during operations, is no longer The remaining activities are adquately covered under applicabic.the LILCO Quality Assurance Program for Shoreham which will remain unchanged. 13.5 STATION PROCEDURIS 13.5.1 Administrative Control

The description contained under this heading in the latest that: lj% revision of the Shoreham USAR remains unchanged except 13-11

SHOREHAM DSAR l O 1 . Safety-related station procedures shall be processed through the Review of Operations Committee (ROC) and Nuclear Quality i l Assurance (NQA). j [

2. The Plant Manager shall approve Station Administrative l Procedures, Security Plan Implementating Procedures, and i Emergency Plan Implementing Procedures prior to [

implementation. Other Station operating Procedures shall be approved by the f 3. appropriate Division Manager or by the Plant Manager prior to i implementation. See DSAR Piqure 13.5.1-1. Refer to the latest revision of the USAR for other information on this subject. ! i 13.5.1.1 Normal o operations The description contained under this heading in the latest revision of Shoreham USAR remains unchanged with the exception ! that the NRB has been revised in accordance with 13.4.2C and a new Table 13.5.1-1 is supplied herein. l 13.5.1.2 Routine Maintena_nce2_ Repairs, and ruel Handling The description contained under this heading in the latest revision of Shoreham USAR remains unchanged. Refer to the USAR ! for information on this subject. 1 13.5.1.3 Modifications The description contained under this heading in the latest revision of Shoreham USAR remains unchanged. Refer to the USAR l for information on this subject. . Procedures 13.5.2 13.fo2 ; operating Procedures The de.cription contained under this heading in the latest revisfor af the Shoreham USAR remains unchanged except that the General operating Procedures now only describe integrated station operation. Startup and Shutdown are no longer pertinent. > Operating Procedures are not necessarily performed by, or under i the direction of, persons holding RO or SRO licenses. I 13.5.2.2 Alarm Response Procedures The description contained under this heading Refer in thetolatest the USAR revision of Shoreham USAR remains unchanged. for information on this subject. 13-12

SHOREHAM DSAR V((7 13.5.2.3 Initial Test Procedures This section is no longer pertinent. 13.5.2.4 Maintenance Procedures The description contained under this heading Refer in thetolatest the USAR revision of Shoreham USAR remains unchanged. for information on this subject. 13.5.2.5 Instrument and control Systems Procedures The description contained under this heading Refer in thetolatest the USAR revision of Shoreham USAR remains unchanged. for information on this subject. 13.5.2.6 Surveillance Procedures The description contained under thin heat ng in thetolatest Refer the USAR revision of Shoreham USAR remains unchanged. for infornation on this subject. 13.5.2.7 Shoreham Nuclear Power Station Emeroency Preparedness [Aa,3

, ,,n
   -                 The description contained under this heading in                                thetolatest Refer        the USAR revision of Shoreham USAR remains unchanged.

for information on this subject. 13.5.2.8 Health Physics Procedures

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

13.5.2.9 Chemistry Procedures The description contained under this heading Refer in thetolatest the USAR revision of Shoreham USAR remains unchanged. for information on this subject. 13.5.2.10 Reactor Encineerino Procedures Procedures that describe the methods of nucicar performance and evaluation are originated, reviewed and approved in accordance with revised DSAR Figure 13.5.1-1. 13.5.2.11 Plant Security Procedures The description contained under this heading Refer in thetolatest the USAR [~ - revision of Shoreham USAR remains unchanged. ( for information on this subject. l 13-13

I a i 1-i SHOREHAM DSAR- j 13.5.2.12 Radioactive Waste Management Procedures f The description contained under this heading Refer in thetolatest the USAR revision of Shoreham USAR remains unchanged. for information on this subject. Temporary Procedures l 13.5.2.13 i

                  , The description contained under this heading Refer                           in thetolatest the USAR                     !

revision of Shoreham USAR remains unchanged. for information on this subject. l

                                                                                                                                        -l 13.5.2.14                             Temporary Changes To Approved Station Procedures a

The description contained under this heading Refer in thetolatest the USAR I revision of Shoreham USAR remains unchanged. f for information on this subject. lj 13.6 PLANT RECORDS i The description contained under this heading in the latest , revision of Shoreham UFAR remains unchanged. . Refer to the USAR l for information on this subject. r () 13.7 INDUSTRIAL SECURITY The Security Plan, Training and Qualification Plan, and the Safeguards Contingency Plan for the Shoreham Nuclear These Power ! Station have been submitted as separate documents. l documents are withheld from public disclosure pursuant to 10CTR2.0 9 (d) , " Rules of Practice." The Security Plan and the l i Safeguards - Contingency Plan are also withheld f rota public- ; disclosure pursuant to-10CPR73.21, " Requirements for the j Protection of Safeguards Information." ! [ i 13-14

. _ . - . . . _        _ _ - , _ _ . . _ - . _ _ . _ . _ ,                        u _,_ _         _ . . _ _          __ c , . ; ._ -

                                                                                    .._ _. _. . ~ _ _ _ _ _ _ _ .

SHOREHAM DSAR TABLE 13.5.1-1 PROCEDURES PROVIDED FOR SHOREHAM WU,qlEAR POWER-STATION A. Administrative Procedures shall be provided to cover the Tcilowing types of administrative activites:

l. Authorities and Responsibilitie.s for Safe Operation and Shutdown

2. Equipment Control (e.g., locking and tagging) 3, Procedure Adherence and Temporary Change Method C _ Procedure Review and Approval 5 Schedule for Surveillance Tests

6. Shift and Relief Turnover - Recall of Personnel

7. Log Entries and Record Retention

3. Bypass of Safety Functions and Jumper Control

9. Operating Orders

10. Special Orders

11. Materials Control

12. Radiation Work Permits

13. Access Control to Centro 11ed Area

14. Personnel Training and Qualification B. Operatino Procedures t

1. General Operating Procedures have been provided to_ cover the following Integrated Plant Operating Activities

a. Surveillance.

2. System Operating Procedures shall describe Startup, Normcl Operating, and Shutdown for the designated system. Abnormal Operation, where required, shall be

                              -contained in a section of the System Operating Proccdnt. Procedures are available for operating the systems listed in a through ad. below.
                              -a. 138kV and:69kV Power System

b. Normal Station Service Transformer

c. Reserve Static < Service Transformcr

d. Well Water Syvoem

e. 4,160 V System

f. 480 V System

g. Station Lighting Panels

h. 120 V ac Instrument Bus

1. 120 V ac Reactor Protection System Bus

j. 120 V ac Uninterruptible Power Supply

k. 125 V de System

1. Fuel Pool Cooling

       /- ,

In . Raactor Building Normal Ventilation System (RBNVS)

n. Service Water
o. Radwastef(Liquid) 1 of 4

P k SHOREHAM DSAR A\ T.gqT 13.5.1-1 (Cont'd) s B. Operating Procedures (Cont'd.)

p. Radwaste (Solid)

q. Communications System

r. Condensate Transfer

s. Deluge and Sprinkler System

t. Demineralized Water Transfer

u. Equipment and Floor Drains t

v. Fire Protection System ;

w. HVAC - Control Room

x. HVAC - Turbine Building

y. HVAC - Radwaste Building

z. Makeup Water Treatment na. Station Air S) stem ab. Smoke, Temperaturt, and Flame Detection System Turbine Bui.'. ding Closed Loop Cooling System ac.

ad. CRAC Chilled Water ,

3. Emergency Procedures have been provideo for combatting the following potential emergency conditions: ;

a. Acts of Nature d

b. Abnormal Releases of Radioactivity

c. Fire in Control Room

d. Fuel Handling Accident

e. Plant Fires

f. Loss of Electrical Power

g. Loss of Instrument Air

h. Loss of Service Water

i. Loss of Turbine Building Closed Loop Cooling Water

j. Secondary Containment Control -

k. Radioactive Release Control M armal Operation Procedures required to mitigate the

4. :

coa. quences of the following abnormal conditions shall be contained in the appropriate System Operating Procedures (s):

a. None.

Notes Procedures not designated as emergency procedures shall be incorporated in the Abnormal Performance section of the appropriate system or general l' operating procedures. O V 2 of 4

SHOREHAM DSAR- I TABLE 13.5.1-1 (Cont'd) ( 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.        Maint'enance Procedures                                         +

{ t Maintenance Procedures shall be provided to cover the j following maintenance activities.

1. Control of Welding Processes, Materials, and Welder Qualifications

2. Preventive and Corrective Maintenance of Safety Related Equipment [

i E. Instrument and Control Procedures shall be provided to cover - tIie following instrumentation and control activities: l l i

1. Measuring and Test Equipment j

2. Protective Relaying

3. Instrument Records i

4. Surveillance Testing j

5. Preventive Maintenance of Process Instrumentation

     #()                                                                                                                                                                                     t F.         Fuel Handlino Procedures shall be provided to cover the                                                                                                          j Tollowing fuel handling activities:                                                                                                                              i

1. Special Nuclear Materials Control and Accountability Procedures

2. Spent Fuel Handling and Shipment l

3. Handling and 'torage-of Sealed and Unsealed Sources !

G. Health Physic- P-ocennres-shall be provided to cover the following radCIIEU5~Erotection activities ! o 1. Dose Rate Radiation Surveys '

2. Surface Radioactive Contamination Surveys Personnel Contamination Survey i

3. '

4. T3rsonnel Decontamination j

5. Areas and Equipment Decontamination

6. Monitoring for and Collecting and Recording of !

Occupational Radiation Exposure (ORE) data l

7. Submission and Review of Suggestions ~by Plant Personnel l for the Reduction of ORE

8. Use of Protective Clothing and Respiratory Equipment l 1

I 3 of 4 i

 .   ---.-,-.m      , , .        m  -

c- ,,--- -, . _ _- ,--,,-m..-..<e~ .,,,,..,,,,,,_._,,w. -,.-.,,..-,--,.-,....-,y-,,- ,- ,r ,wy ,~,-me-. ,, - y-

_. . - .- ~ _. . .. - .- . . - . =.-.. - - - - . ... ~ . - _ . . . - -._ ... I j SHOREHAM DSAR TABLE'13.5.1-1-(Cont'd) , i i E. Defueled Emergency Preparedness Implementing Procedures- ; (DEPIPs) shall be provided to cover the following-emergency ! plan activities:

1. Emergency Classification ;

2. Evacuation and Personnel 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 i B. Re-entry and Recovery

9. Emergency Organization, Drills, and Training (O

b F i h I f . 'O , 4 of 4-i

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

SHOREHAM DSAR i!gp) TABLE 13.5.1-2 OK / RESPONSIBILITY FOR ORIGINATION OF STATION PROCEDURES Procedure Type Responsible for Origination Administrative Appropriate Section Head / Unit Manager Operating Appropriate Section Head Alarm Response Appropriate Section Head Maintenance Appropriate Section Head Instrument and Control Appropriate Section Head Health Physics Appropriate Section Head Radiochemistry Appropriate Section Head Reactor Engineering Appropriate Section Head Solid Radica 've Waste Appropriate Section Head

 ,/'N;  Handling and L..ipping C/                                          Appropriate Section Head Gaseous and Liquid Radioactive Waste Effluent Control Fuel Handling                        Appropriate Section Head Surveillance                         Appropriate Section Head Security                             Appropriate Section Head j

1 of 1

1 1 SHOREHAM DSAR k' TABLE 13.5.1-3 FORMAT FOR STATION PROCEDURES

                                                      *SP Number Revision                                           Eff. Date Signature               Date
                                                               'PC NO                             Date Eff         Date Expr Soction Head Quality control Div. Mgr.                                                                                                     __

Plant Mgr. . ___ S2gnature or N/A TITLE

  .r~S.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 indicate the part, a Technical Specification surveillance requirement, 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 Number assignment is TP XX.XXX.XX.

1 of 6 I

SHORERAM DSAR

,p (y    )

TABLE 13.5.1-3 (Cont'd) 4.0 PRECAUTIONS General precautions should be listed in this section before the description of the actual procedure. Precautions should be established, as applicable, to alert the individual performing the task to those situations in which mersures should be taken early or when care should be ekercised to prr;+ cific i equipment and/or personnel. Precautionary notes applicabP A

                                                                                                                                                 ;     .e steps in the procedure should be included prior to that c main body of the procedure and should be clearly identifit',

5.0 PREREQUISITES It is necessary to identify those independent actions or procedures which shall be completed and plant conditions which shall exist prior to performing the procedure. Prerequisites applicable only to specific section of a procedure should be so identified. 6.0 LIMITATIONS AND ACTIONS Limitations on the parameters being controlled and appropriate (

   ,'N      corrective measures to return the parameter to normal should be
  \I        specified when applicable.

7.0 MATERIALS OR TEST EQUIPMENT Special tools, instrumentation, measuring devices, materials, etc. required to accomplish the work should be identified in this section. 8.0 PROCEDURE Etep-by-step instructions in the degree of detail necessary forThese performing the required function or task should be provided. shall be numbered sequentially. Note 1: Operatina Procedures (Table 13.5.1-1, Sections B.2 and B.4) shall, as appropriate, be divided into two categorins: Normal Performance shall include step-by-step instructions to complete the required operation. Subcategories may include startup, routine operation at power, rotation of eculpment, and shutdown. Abnormal Performance shall include instruction to recugnize the existence of and to correct out-of-normal conditions that occur during the normal performance. Included ecy be a rs statement of the out-of-normal condition, including limits of (,) parameters and/or alarm annunciator action. 2 of 6 J

SHOREllAM DSAR

/~T                                        TABLE 13.5.1-3 (Cont'd)

[V Note 2: Maintenance and/or Calibration Procedures If technical manual instructions are written in sufficient detail to permit a safe and logical accomplishment of the required task, applicable sections of the technical manual may be referenced. Note 3: Surveillance Procedures The step-by-step instructions, with appropriate signoff or checkoff provisions for each step, shall be provided to ensilre the proper performance of the surveillance activity. 9.0 ACCEPTANCE CR_ITERIA Specific acceptance criteria against which the test results shall be Acceptance judged for approval / disapproval must be stated clearly. criteria may contain qualitative data (i.e., a given event does or does not occur) and/or quantitative data (such as set points, calibration curves, tolerances, etc.) as appropriate for the type of device being tested. A 0.0 FINAL CONDITIONS

 ')
 '         Provide a listing of those tasks required to return the applicable component or system to operational status and to compile the proper documentation of the procedure. Where applicable, verification of completion will be provided by a signature.

11.0 REFERENCES

This section contains applicable references including appropriate sections of the USAR, Technical Specifications, QA Manual, flowother diagrams or other drawings, manufacturer's equipment manuals, station procedures, and system descriptions. 12.0 APPEtjDICES Applicable appendices (in the form of checklists, data sheets, diagrams, etc.) should be included when necessary to support the proper implementation of the procedure, k'T

 \J 3 of 6

SHOREHAM DSAR TABLE 13.5.1-3 (Cont'd) t(O V EVENT ORIENTED EMEDGENCY PROCEDURE FORMAT SP Number Submitted ' (Section Head) Revision Approved: (Operations Manager) Effective Date ,, TITLE *

                                                                                        *Should be worded to indicate the purpose of the procedure.

1. SYMPTOMS:

Symptoms should be included to aidThis in the should identification of the emergency. include alarms, operating conditions, and If probable magnitudes of parameter changes. a condition is peculiar only to the emergency under consideration, it should be listed

                    ~h                                                                                                                                               first.
 <[V                                                                                                                                                                 (Delete if not pertinent)

2. AUTOMATIC ACTION:

Il'REDIATE ACTION:

                                                                                                                                                                   'These steps should specify immediate action

3. 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.

SUBSEQUENT ACTION: Steps should be included to return the reactor

4. to a normal shutdown period under abnormal or emergency conditions.

FINAL CONDITIONS: These steps should specify the documentation,

5. authorizations, and plant conditions that must be completed prior to resumption of Normal Operation, defined in 22.XXX.XX.

DISCUSSION: A brief explanation of the procedure. 6. 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 O)(                                                                               rapdi evaluation and action by the operator, j

4 of 6

                                                                                                                       .t s

SHOREHAM DSAR

  <-                                                                TABLE 13.5.1-3 (Cont'd) f k--                    -

SYMPTOM ORIENTED OPERATING EMERGENCY PROCEDURE FORMAT

                                                                          ** TITLE Submitted:                                                              1.

(Section Head) Approved _ '2. . 3. TPC No Effect Expir. Date of Date TPC of TPC 00 Should be worded to indicate the purpose of the procedure. 1.0 FURPOSE (A brief description of the purpose for which the procedure is intended should be clearly stated). 2.0 ENTRY CONDITIONS O) l s_ (This section should specify the plant conditions and/or plant procedures which identify the need for performing this procedure). 3.0 OPERATOR ACTIONS (These steps should specify operation of controls or confirmation of automatic actions that are required to fulfill the purpose of this procedure). 2 w 5 of 6

SHORERAM DSAR TABLE 13.5.1-3 (Cont'd) a ALARM RESPONSE PROCEDURE (ARP) ' FORMAT Submitted: ARP (Section Head) (Windown Number). Approved .

      .            (Operations Mgr.)                              (Panel Number)

(Panel Sub-Section)

 -Effective Dates Revisions-ALARM TITLE Instr. No.                                 Set Point:    Trip Reset OSSIBLE CAUSE                             IMMEDIATE ACTION
    ,ist those conditions which               List the immediate actions           -

might-have initic+=d the alarm; in order for each possible list the1most pcobable cause-first. cause.

  -SUBSEQUENT ACTION List the procedures by title and-number that would give follow-up action.

Note 1: ALARM RESPONSE PROCEDURES - will--include specific instructions to mitigste the consequences of the condition indicated by the

             -alarmed annunciator. Alarm Response Procedures should be filed int numerical sequence in Appendix I to Volume II of the Ctation Operation Manual.

REFERENCES

   -The procedure with which the-ARP is associated should be identified.

The reference drawing (s) that details the input and/or control signal to the annunciator and/or its initiating device (s) should be identified. O 6 of 6

PRESIDENT j AND

     !,                                                                                                            C.O.O.

I r t f 5 1 VICE PRESIDENT l i OFFICE OF 4 , CORPORATE SERVICES I & VICE PRESIDENT , l OFFICE OF NUCLEAR j i t i 1 i l l l MANAGER ouAUTY MANAGER ASSURANCE-VIC R I ENT NUCl_ EAR EMERGENCY PLANT NUCLEAR NUCLEAR ORGANIZATION c

               & DIRECTOR '                                                                              IV1ANAGER        EN    tEERING                                     (SEE RGURE                           i PREPAREDNESS                                                                                      SUPPORT                        17.2.1-1)

OF TRAINING ~ c 1 FIGURE 13.1.1-1 i DIRECTION OF EXECUTIVE RESPONSIBILITY i SHOREH AM NUCLEAR POWER STATION # I DEFUELED SAFETY ANALYSIS REPORT 3 MANAGER l MANAGER REVISION 3 JULY 1991 l MANAGER TRAINING e OPERATION & PRODUCTIONi SIMMON TRAINING , SERVICES t-i l l..._.._,.___....__..._._,__,-..______.,..-,..___,__,,..___.__,__

O . U- , 6 ___. , I VICE PRESIDENT ; OFFICE OF CORPORATE SERVICES

                               & VICE PRESIDENT                                                  j.

I OFFICE .OF NUCLEAR I

i. l 71 l l .

i QUALITY MANAGER MANAGER ASSURANCE NUCLEAR PLANT N CHR ORGANINION - j OPERATIONS MANAGER ENGINEERING c (SEE FIG RE r SUPPORT t F l i LICENSING - OPERATIONS ' SYSTEMS l. l- , CONTRACTS i NUCLEAR

             & MATERIAL   MAINTENANCE      :

CONTROL I ANALYSIS ,. i ( -- l NUCLEAR RAD I AL N FINANCIAL PROJEC & i ADMINISTRATION SERVICES FIGUnE 13.1.1-2 orFice oF uuctEAn oRoAnizArion I. r

suonEHAu wuctEAn powen STArion i i OPERATIONS DEFUELED SAFETY ANALYSIS nEPoRT ,I SECU RITY STAFF REVISION 3 JULY 1991 .

~~

4

                                                      .~          .

t ORIGINATOR DEVELOPS PROCEDURE [ 6' $ ELECTED REVIEWER COW.ENTS l h * ! B > r RESOLVE COMMENTS r NON 5AFETY ! 5AFETY ' APPROVAL ROUTE RELATED RELATED SELECTION , ROC MEMBERS DIV. AND/OR

                   "                                                              PLANT MANAGER                  !

REVIEW AND COMENT 3 ti APPROVAL D i RESOLVE COMENTS ! g- - W ROC APPROVAL i . t DIV. AND/OR ' PL ANT MANAGER . AP P ROVAL !$$UE [ i l ISSUE

                                                                                                   .             i t

NQA DEPARTMENT . i 4 FIGURE 13.5.1-1 PROCEDURE FLOW DI AGRAM SHOREHAM NUCLEAR POWER STATION fq'v DEFUELED SAFETY ANALYSIS REPORT .

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

, a f

s SHOREHAM DSAR . CHAPTER 14 l i INITIAL TESTS AND OPERATIONS r i This chapter is not needed due to the defueird condition of the

l

                                 - plant.

l 4 I e d i

6 P t t

i. 5 I r

                                                                                                                                                                                                               .         t I

5

                                                                                                                                                                                                                     -h i

, f I L Ai

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                                                                                                                                                                                                                     .1 4

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                                                                                                                                                                                                                        +
                                                                                                                                                                                                                         ?-

r> I f 24-1 r i v I I

4 l SHOREHAM DSAR ^l/

 - '                                      CEAPTER 15                                  l i

ACCIDENT ANALYSIS i 15.1 GENERAL , i Analytical Objective > t Chapter.15 of SNPS USAR provides the results of analyses of ' the ' l spectrum of transient and accident events which are postulated to j 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 f rom the core and is stored in the r 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. i i 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.               ;

i ( ! 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 i 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 l 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 i One result from the ORIGEN2 calculation is a graph of decay heat ' or thermal power (in watts) , as a function of time. Results of this analycis are presented in Figure 15.1-1. The calculated j decay heat load as of June 1989 is approximately 0.55 kw. i 15-1

SHOREHAM DSAR (( )'It must be recognized that there are some limitations For in theORIGE processes of the code and its supporting data sets. instance, ORIGEN conveniently with the spatial variations in fuel enrichment and burnup. In addition, there are uncertainties associated with averaging of nuclear cross-section data within the thermal,Nevertheless, it resonance, and fission neutron energy ranges.

        'is not expected that large uncertainties should occur in heat load estimates. See the comparison         of calediated to measures This  gives evidence that the dose rates in DSAR Section 12.2.

decay heat load calculations are reasonable, as the same analysis (ORIGEN2) was used to generate both sets of data. Analytical Categories 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 C'colant_._ Temperature This analytical category of USAR Chapter 15 events includes the following events:

y'~h 15.1.5 Pressure Regulator Failure - Open D 15.1.7 Feedwater Controller Failure - Maximum Demand l 15.1.8 Loss of Feedwater Heating 15.1.9 Shutdown Cooling (RHR) Malfunction - Decreasing i 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 RER I

systems are not operating in support of nuclear fission, the following transients are not applicable: 15.1.1 Generator Lead Rejection l Turbine Trip 15.1.2 i 15.1.3 Turbine Trip with Failure of Generator Breakers to s 15-2

SHOREHAM DSAR 15.1.4 Main Steam Isolation Valve closure ((v') Pressure Regulator Failure - Closed 15.1.6 15.1.18 Loss of Feedwater Flow , 15.1.21 Loss of Condenser vacuum 15.1.26 Core Coolant Temperature Increase The transient of this category applicable to spent 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 capacity 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 , Ip(_)s 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 impossible in any event), and the decay heat of the core is 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 Pump Trip 15.1.22 Recirculation Pump Seizure 15.1.23 Recirculation Flow Control Failure With Decreasing h t Flow l V 15-3 Rev. 3 July 1991

k SHORERAM DSAR k( 4. Reactivity and Power Distribution Anomalien

Events included in this category are those which cause rapid increase in power. Since the reactor is defueled, the following events are not applicable . 15.1.11 Continuous Control Rod Withdrawal Durinq _ Power Range operation 15.1.12 Continuous control Rod Withdrawal During

Reactor Startup 15.1.13 Control Rod Removal Error During Refueling 15.1.14 Fuel Assembly __ Insertion Error During Refueling 15.1.15 Off-Design Operational Transient Due to inadvertent Loading of a Fuel Assembly into an Improper Location 15.1.16 Inadvertent Loading and Operation of a Fuel Assem5Iy in Improper Location 15.1.24 Recirculation Flow Control Failure with Increasing Flow

     )          15.1.25    Abnormal Startup of Idle Recirculation Pump 15.1.33     Control Rod Drop Accident

5. Increare in Reactor Coolant Inventory

               -Since the HPCI system is not required the following transient-is not applicable:

15.1.10 Inadvertent HPCI Pump Start

6. Decrease in Reactor Coolant Inventory 6.A Events Not Applicable to Spent Fuel Storage 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 Inadver. tent opening of a Safety Relief Valve 15.1.37     Feedwater System Pipina Break The following event is not a design basis event and is rs             applicable only to power operation:

d 15-4

l S!!OREHAM DSAR 15.1.27 Anticipated Transient Without Scram ( ATWS) (f(j The single failure-proof polar crane design eliminates the following event: 15.1.28 Cask Drop Accident 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 applicables : 15.1.30 off-Design operationa3. Iransient as a Consequence of Instrument Line Failure 15.1.34 Pipe Breaks Inside the Primary Containment (Loss-of-Coolant Accident) 15.1.35 Pipe Breaks Outside the Primary Containment lSteam Line Break Accident! 6.B Events Without Fuel Damage , 15.1.29 Miscellaneous Small Releases outside Primary , Containment Releases that could result from piping failures outside the ([~) x- 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. 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. 15.1.31 Main Condenser Gas Treatment System Fpilure As the main condenser is not operating, there can be no offsite dose resulting from this event. 15.1.32 Liquid Radwaste Tank Rupture Should accident occur radioactivity could be released to the environment but the effect would be negligible. The accident analysis described in DSAR Section 11.2.3.4.2 and 11.2.3.4.3 proves this. 15-5 Rev. 3 July 1991 _ _ _ _ ___m_ --_

F SHOREHAM DSAR (O ! t i 5 (This Page Intentionally Left Blank) ; i t , r 3 h

 'O 15-6          Rev. 3 July 1991           r
                                                                            ,--m-,

t SHOREHAM DSAR 15.1.38 Failure of Air E$ector Lines As the main condenser is not operating, this accident is no longer a design basis event. 6.C Events with Fuel Damage . 15.,1.36 Fuel Handling Accident 15.1.36.1 Identification of causes The fuel handling accident is assumed to occur as a consequence of a failure of the fuel assembly lifting mechanism, resulting in the dropping of a raised fuel assembly onto the top of the spent fuel racks. , 15.1.36.2 Startine conditions and Assumptions Accidents that result in the release of radioactive materials directly to the secondary containment can occur when fuel is being handled. In this case, radioactive material released as a result of fuel damage is available for transport directly to the secondary containment. Table 15.1.36-1 presents the parameters used in this analysis, h) 15.1.36.3 Accident Description The most severe fuel handling accident from a radiological viewpoint is the dropping of a fuel assembly onto other fuel assemblies. The sequence of events is as follows: Approximate Event Elapsed Time Fuel assembly is being handled 0+ 1. by refueling equipment. The assembly drops. Some of the fuel rods in both 1 min. . 2. the dropped assembly and another assembly are damaged, resulting in the release of gaseous fission products to the fuel pool and eventually to the secondary containment atmosphere.

3. The reactor building refueling 1 min.

floor ventilation exhaust radiation monitoring system may alarm to alert plant personnel. Operator actions begin 5 min. 4. 15-7

SHORERAM DSAR I(p x_ 15.1.36.4 Identification of operator Actions

1. The operator will initiate the evacuation' of the secondary containment and securing of Secondary Containment doors, if necessary.

2. The fuel handling foreman will instruct personnel to go immediately to the radiation protection personnel decontamination area, if necessary..

3. The fuel handling foreman will make the operator aware of the accident.

4. The operator will initiate action to determine the extent of potential radiation doses by measuring the radiation levels in the vicinity of or close to the secondary containment.

5. An HP technician will post the appropriate radiological control signs at the entrance to the sece'idary containment.

6. Before entry to the secondary containment is made, a careful study of conditions, radiation levels, etc.,

will be performed. ( 15.1.36.5 HVAC Scenarios Considered As set forth in Section 15.1.36.6, the quantity of gaseous fission products in the fuel's gap which is released will Calculations not be large (2.52 Ci of Kr-85 only). indicate that the reactor building refueling floor exhaust radiation monitoring system would not alarm and consequently the PBSVS will not be actuated (i.e., the As a result, analyses were RBNVS continues to operate). performed assuming either RBSVS or RBNVS system operation. Secondary containment discharge rates are 167 and 6500 percent / day for the RBSVS and RENVS cases, respectively. As a comparison case, a " puff" release over a short period of time (2 hours, as suggested by Regulatory Guide 1.25), has been analyzed. Although thir is not a design basis case, it is useful to compare it with the two HVAC cases, Results for all three cases (RBSVS, RBNVS, and puff release) are given in the following sections. 15.1.36.6 Analysis of Effects and Consequences 15.1.36.6.1 Evaluation Methods The analytical methods and associated assumptions used to (~N evaluate the consequences of this accident are consistent

  \m-)

with Regulatory Guide 1.25. The assumptions and parameters ; are given in Table 15.1.36-1. 15-8

SHORERAM DSAR h\m- 15.1.36.6.1.1 Methods, Assumptions, and Conditions The assumptions used in the analysis of this accident are listed below:

1. The fuel assembly is dropped from the maximum height allowed by the fuel handling equipment.

2. The entire amount of potential energy, referenced to the top of the spent fuel racks, is available for application to the fuel assemblies involved in the accident. This assumption neglects the dissipation of some of the mechanical energy of the falling fuel assembly in the water above the racks and requires the complete detachment of the assembly from the fuel hoisting equipment. This is possible if the fuel assembly handle, the fuel grapple, or the grapple cable breaks.

3. None of the energy associated with the dropped fuel assembly is absorbed by the fuel materici (uranium dioxide).

15,1.36.6.1.2 Results and Consequences I ((~m) 15.1.36.6.1.2.1 Fuel Damage _ The analysis of USAR Section 15.1.36.5.1.2.1 applies to this accident. In that section of the USAR, it was assumed that 125 fuel rods would fail as a resultThe of dropping same the assumption fuel assembly into the reactor vessel. is applied here. 15.1.36.6.1.2.2 Fission Product Release From Fuel Fission determinedproduct from releases the 1989 for the fuel inventory in handling accident are l Table 12.2-1. Specifically, it is seen that only Kr-85 is of any The only significance with respect to gaseous releases. other gaseous isotope in this table is H-3, which wouldUsing the add, at most, 0.1% to the skin dose from Kr-85. above number of failed rods, and the assumptions given in Regulatory Guide 1.25, the quantity of Kr-85 released, is as follows: Release = 1.56E+03Ci x 125 damaged rods 62 rods / bundle x 560 bundles in core x 1.5 radial peaking factor x 30% in gap = 2.52 Ci (( 15-9 Rev. 3 July 1991

SHOREHAM DSAR- '\ 15.1.36.6.1.3 -Radiolocical Effects 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 2.0BE-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, the maximum (t=0) dose rates (whole body and skin) are very low and, with the exception of the RENVS 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 This in the spent fuel pool is released (1.56E+03 Ci Kr-85). can only occur if all the fuel is postulated to be mechanically damaged and there is a complete release of gasecus 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. 15-10

                                                  -         e-*--g wee --

SHORERAM DSAR

;.              All other conditions'and_ parameters indicated in Teble 15.1.36-1 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 0.0314) of the - 10CFR100 dose guidelines. Results are graphically shown in Figure 15.1.36A-1. Dose rates for the postulated worst case scenario are above current ODCM limits, but the duration of l

                -the-high dose rates in the RBNVS and puff release cases is quite short (two hours or less) .

10 ' lO 15-11 Rev. 3 July 1991

s SHORERAM DSAR r' TABLE 15.1.36-1 ( FUEL HANDLING ACCIDENT - PARAMETERS

FOk POSTULATED ACCIDENT ANALYSES Conservative (NRC) Assumptions I. Data and' assumptions used to estimate radioactive source from postulated accidents See this Chapter A. Power level 1.5 B. Peaking factor C. Fue2 damaged 125 rods D. Release of activity from fuel 30% Kr-85 E. Iodine fractions (1) Organic N/A (2) Elemental N/A (3) Particulate N/A II. Data and assumptions used to estimate activity released A ( A. Secondary contain- See Section ment discharge rate (t/ day) 15.1.36.5 B. Adsorption and filtra-tion efficiencies N/A (1) Elemental iodine C. Recirculation system parameters (1) Flow rate N/A (2) Mixing efficiency N/A III. Dispersion data A. EAB and LPZ distances 311/3,220 (meters) 3 B. X/Os (sec/m ) EAB (0-2 hr) 1.36E-03 LPZ (0-8 hr) 2.50E-05 (6-24 hr) 1.75E-05 (1-4 days) 7.80E-06 (4-30 days) 2.45E-06 IV. Dose data A. Method of dose calculation Regulatory Guide 1.25 B. Dose conversion assumptions Regulatory Guide 1.25 s C. Doses and Dose Rates Table 15.1.36-2

SHORERAM DSAR

   -                                                                                           TABLE 15.1.36-2
 'N /                                                                                       FUEL HANDLING ACCIDENT RADIOLOGICAL CONSEQUENCES Whole Body Dose, rem                                                                           Skin Dose, rem EVAC                                                                         10CFR100                                                                           10CFR100 Scenario                              EAB                 LPZ                 Limit                             EAB                           LPZ                Limit

RBSVS 1.14E-07 1.22E-08 2.50E+01 9.90E-06 1.06E-06 3.00E+02 Operates Maximum (t = 0) Dose Rates, mram/hr Whole Body Gamme Skin ODCM ODCM l EAB LPZ Limit EAB LPZ Limit 6.10E-05 1.12E-06 5.70E-02 5.30E-03 9.74E-05 3.42E-01 Whole Body Dose, rem Skin Dose, rem RBNVS 10CFR100 10CFR100 Operates EAB LPZ Limit EAB LPZ Limit
1.74E-06 3.22E-08 2.50E+01 1.52E-04 2.80E-06 3.00E+02 I Maximum (t = 0) Dose Rates, mrem /hr Whole Body Gamma Skin CDCM ODCM l EAB LPZ Limit EAB LPZ LLmit 4.79E-03 8.81E-05 5.70E-02 4.17E-01 7.66E-03 3.42E-01 Whole Body Dose, rem Skin Dose, rem Puff 10CFR100 10CFR100 Release EAB LPZ Limit EAB LPZ Limit
1.75E-06 3.22E-08 2.50E+01 1.52E-04 2.80E-06 3.00E+02 Maximum (t = 0) Dose Rates, mrem /hr Whole Body Gamma Skin ODCM ODCM l EAB LPZ Limit EAB LPZ Limit 8 *sE-04 1.61E-05 5.70E-02 7.61E-02 1.40E-03 3.42E-01
The skin dose limit is set equal to the thyroid limit.

it 1 Rev. 3 July 1991

l i S110REllAM DSAR , TABLE 15.1.36A-1

                                        " WORST CASE' TUEL DAMAGE ACCIDENT                                                                                                                                 l RADIOLOGICAL CONSEQUENCES I

Whole Body Dose, rem Skin Dose, rem 10CFR100 f HVAC 10CFR100 t Scenario EAD LPZ Limit EAD LPZ Limit f j RBSYS 7.03E-05 7.50E-06 2.50E+01 6.11E-03 6.52E-04 3.00E+02 l Operates Maximum ( t = 0 Dose Rates, mram/hr i Whole Body Gamma Sksn , ODCM ODCM l , EAB LPZ Limit EAB LPZ Limit j 3.76E-02 6.92E-04 5.70E-02 3.27E+00 6.01E-2 3.42E-91 l Whole Body Dose, rem Skin Dose, rem RBtNS Operates 10Cth100 10Crh100 i EAB LPZ _ Limit _ EAB LPZ Limit ! 1.08E-03 3.99E-05 2.50E+01 9.35E-02 1.73E-03 3.00E+02 Maximum (t = 0) Dose Rates, mrem /hr l Whole Body Gamma Skin l ODCM ODCM l EAB LPZ Limit EAB LPZ Limit 7 2.96E-00 5.44E-02 5.70E-02 2.57E+02 4.73E+00 3.42E-01 [ Whole Body Dose, rem Skin Dose, rem i Puff Release 10CFRT6U 10Cra100 1 EA3 LPZ Limit EAD LPZ Limit [ 1.00E-03 1.99E-05 2.50E+01 9.39E-02 1.73E-03 3.00E+02 Maximum (t = 0) Dose Rates, mrem /hr Whole Body Gamma Skin ; ODCM ODCM l EAB LPZ Limit _EAB LPZ Limit i 5.40E-01 9.93E-03 5.70E-02 4.70E+01 8.63E-01 3.42E-01

Skin dose limit set equal to thyroid limit I

r Rev. 3 July 1991 l

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S110REllAM DSAR CilAPTER 16 (O TECHNICAL SPECII'ICATIONS The SNPS Technical Specifications are found in Appendix A of the SNPS license, f (J l j tv l i l 16-1 Rev. 3 July 1991 a&F- t we.m-

SHOREHAM DSAR ! () CHAPTER 17 QUALITY ASSURANCE 17.1 QUALITY ASSURANCE DURING DESIGN AND CONSTRUCTION l t The description contained under this heading in the latest revision of the Shoreham USAR remains unchanged. Refer to USAR l for information on this subject. 17.2 QUALITY ASSURANCE DURING THE OPERATIONAL PHASE i The description of the Quality Assurance Program during Shoreham . Nuclear Power Station operational phase under this heading in the ! latest revision of the Shoreham USAR is essentially unchanged. [' However, many of the structures, systems and components designated as Quality Assurance Category I (safety related) in t USAR Table 3.2.1-1 have been redesignated as Quality Assurance . Category II in this DSAR. The applicability of the USAR Section l , 17.2 Operational phase Quality Assurance Program as modified in l this DSAR to the Quality Assurance (QA) Categories in DSAR Table ; 3.2-1 are as follows: l ' i QA Category I - The USAR Section 17.2 Quality Assurance Program as modified by DSAR Section 17.2 . (' applies to the safety related structures, ! systems and components which meets the i intent of 10CFR50, Appendix B. i QA Category IIA - Deleted (formerly aafety l related) , QA Category II - Appropriate measures are applied to these (non safety structures, systems, and components in related) eccordance with QA corporate policy to assure that the safety significance given to them in the DSAR, Technical Specifications, l i and Emergency Operating Procedures are i maintained. The specific modifications of the USAR Section 17.2 applicable to the Shoreham DSAR phase are as follows: ; ( t i 17-1 Rev. 3 July 1991 _ _ _ . _ . _ . _ _ _ . ~ .- __. .__ _ _ . _ _ .._ _

                                                                            -I SHOREMAM DSAR                             {.

17.2.1 Organization A) Deleted j t

3) Deleted c) Deleted D) Deleted .

El Deleted !; F) Deleted-G) The Vice President, Office of Corporate Services and Vice ! President, Office of Nuclear, reports to the President and- 3 provides-SNPS with training, emergency preparedness and nondestructive examination services. T. lese services are ! subject to-the policies and requirumente of the LILCO QA t, Program. H)- The LILCO organization structure for the Shoreham DSAR Phase l is shown in rigure 17.2.1-1. } t I) Deleted ; i l ( J) Deleted r t I

                                                                           -i i

i (O 17-2 Rev. 3 July 1991 i i i

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SHOREHAM DSAR 3 (} 17.2.2 A) Quality Assurance Program The ISEG organization is eliminated from the NOAD. B)- The structures, systems, and components designated as QA Category 1 (safety related) in USAR Table 3.2.1-1 which no longer have a safety function are redesignated as QA Category II. These structures, syrtams, and components will no longer require a 10 CPR 50, Appendix B QA program. These structures, systems, and components will be afforded the requirements associated with non-safety systems. l 17.2.3 Design control No change. 17.2.4 Procurement-Document control No change. 17.2.5 Instructions, Procedures, and Drawings No change. 17.2.6 Document control No change. 17.2.7 Control of Purchased Material, Equipment, and Services No change. 17.2.8 Identification and Cor. trol of Material, Parts and Components No change. 17.2.9 Control of Special Processes No change. 17.2.10 Inspection No change. 17.2.11 Test Control No change. P 17-3 Rev. 3 July 1991

I i S110REllAM DSAR l (( control of Measuring and Test Equipmen_t I 17.2.12 No change. 17.2.13 Inspection, Test, and operating status , No change. , Test, and Operating Status ! 17.2.14 Inspection i i No change. ; i

                       ' 17.2.15                               Nonconforming Materials, Parts, or components                                                                                                                                                  j i

No change. t 17.2.16 co m etive Action t

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DIRECT ACCESS AS REQUIRED DAY TO DAY FUNCTION AL REPORTlHO F IGU R E 17.2.1 1 LILCO ORG ANi:ATION FOR QUALITY ASSUR ANCE SHOREH AM NUCL E AR FOWER STATION DEFU E L E D SAF E T Y AN ALYSIS REPOR T R EVIS'ON 3 J ULY 19 91 9

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i i f SHOREHAM DSAR i NUREG.0737 TMI ACTION PLAN REQUIREMENTS ; _ TABLE OF CONTENTS liig Title f Introduction i i I.A.l.1 shift Technica' Advisor I.A.l.2 Shift Supervisor Administrative Duties I . A . i . *3' Shift Manning I.A.2.1 Immediate Upgrade of Reactor Operator and Senior Reactor Operator Training-and Qualification I.A.2.3 Administration of Training Programs I.A.3.1 Revise Scope and Criteria for Licensing Examinations I.B.I.2 Evaluation of Organization and Management I.C.1 Guidance for the Evaluation and Development of O Procedures for Transients and Accidents I.C.2 Shift and Relief Turnover Procedures I.C.3 Shift Supervisor Responsibilities I.C.4 Control Room Access I.C.5 Procedures fo't Feedback of Operating Experience to Plant Staff 1.C.6- Procedures for Verification of Correct Performance of Operating Activities I.C.7 NSSS Vendor Review of Procedures I.C.8 Pilot Monitoring or Selected Emergency Procedures for Near Term Operating License Applicants I.D.1 Control Room Design Reviews I.D 2 Plant Safety Parameter Display Console I.G.1 Training Requirements During Low Power Testing ( II.B.1 Reactor Coolant System Vents i

SHOREHAM DSAR ( _ TABLE Or CONTENTS (CONT'D)

Item Title , II.B.2 P4 ant Shielding to Provide Access to Vital Areas and Protect Safety Equipment for Post Accident Operation II.D.3 Post Accident Sampling II.B.4 Training for Mitigating Core Damage

         !!.B.7    Analysis of Hydrogen Control II.D.1 Performance Testing of BWR and PRR Relief and Safety Valves II.D.3 Relief and Safety Valves Position Indication II.E.4.1 Containment Dedicated Penetrations II.E.4.2 Containment Isolation Dependability II.F.1 Additional Accident Monitoring Instrumentation II.F.2 Identification of and Recovery from Conditions Leading to Inadequate Core Cooling II.K.l.5   Safety Related valve Position II.K.l.10 Operability Status II.K.l.22  Auxiliary Heat Removal System Procedure II.K.l.23  RV Level, Procedures II.M.3.3 FailuretoofClose Valve                   Power Operated Relief Valve or Safety II.K.3.13 Separation of HPCI and RCIC System Initiation Levels                Analysis and Implementation II.K.3.15  Modify Break Detection Logic to Prevent Spurious Isolation of HPCI and RCIC Systems II.K.3.16 Reduction of Challenges and Failures of Relief valves                reasibility Study and System Modification II.K.3.17  Report on Outage of ECC Systems                   Licensee Report and Proposed Technical Specification Changes rs    II.K.3.lB f,y)             Modification of Automatic Depressurization System Logic - reasibility for Increased Diversity for Eome Event Sequences 11
                                                                                          =

SHOREHAM DSAR ( TABLE OF CONTENTS fQ0NT'D)

                                           .I_t e m     Title II.K.3.21   Restart of Core Spray and LPCI Systems on Low Level II.K.3.22   Automatic Switchover of Reactor    Core Isolation Verify Procedures and Cooling System Suction Modify Design II.K.3.24   Confirm Adequacy of Space Cooling for High Pressure coolant Injection and Reactor Core Isolation Cooling Systems II.K.3.25   Effect of Loss of AC Power on Pump-Seals II.K.3.27   Provide Common Reference Level for Vessel Level Instrumentation II.K.3.28  Study and Verify Qualification of Accumulators on ADS Valves II.K.3.30  Revised Small Break LOCA Methods to Show Compliance

() II.K.3.31 with 10CFR50.46, Appendix K Plant Specific Calculations to Show Compliance with 10C'/R50.46 II.K.3.44 Evaluation or Anticipated Transients with Single railure to verify No ruel railure II.K.3.45 Evaluation of Depressurization with other Than rull ADS II.K 3.46 Response to List of concerns from ACRS Consultant (Mr. C. Michelson) III.A.l.1 Upgrade Emergency Preparedness III.A.l.2 Upgrade Licensee Emergency Response racilities III.A.2 Improving Licensee Emergency Preparedness - Long Term III.D.l.1 Primary Coolant Sources Outside the Containment Structure III.D.3.3 In plant Radiation Monitoring III.D 3.4 Control Room Habitability lii

4 ! S!!OREHAM DSAR j f NUREG-0737 TMI ACTION PLAN REQUIREMENTS ; I. OPERATIONAL SATETY : i I.A.1.1 Shift Technical Advisor i NRC Position Each licensee shall provide an on-shift technical advisor to the ! shift supervisor. The shift technical advisor (STA) may serve i more than one unit at a multi-unit site if qualified to perform ' the advisor function for the various units. ! The STA shall have a bachelor's degree or equivalent in a ; scientific or engineering discipline and have received specific ; training in the response and analysis of the plant for transients [ and accidents. The STA shall also receive training in plant i design and layout, including the capabilities of instrumentation j and controls in the control room. The licensee shall assign i normal duties to the STAS that pertain to the engineering aspects of assuring safe operations of the plant, including the review ! and evaluation of operating experience. [ h g- LILCO Position i

Under the conditions of the LILCO and New York State Shoreham settlement the plant is shutdown and defueled. STA staffing is therefore not required. [ I.A.1.2 Shift Supervisor Administrative Duties t i NBC Position _ ; i Review the administrative duties of the shift supervisor and delecate functions that detract from or arc subordinate to the ' management responsibility for assuring safe operation of the plant to other personnel not on duty in the control room.

1. The highest level of corporate management of each licensee !

shall issue and periodically reissue a management directive that emphasizes the primary management responsibility of the ; shift supervisor for safet operation of the plant under all i l conditions on his shift and that clearly establishes-his i command duties. 2 Plant procedures shall be reviewed to assure that the duties, responsibilities, and authority of the shift supervisor and ( l control room operators are properly defined to effect the ! establishment of a definite line of command and clear I delineation of the command decision authority of the shift - supervisor in the control room relative to other plant l managment personnel. Particular emphasis shall be placed on i the following: }

SHORERAM DSAR {'

a. The responsibility and authority of the shift supervisor shall be to maintain the broadest perspective of operational conditions affecting the safety of the plant as a matter of highest priority at all timos when on duty in the control room. The principle shall be reinforced that the shift supervisor should not become totally involved in any single operations in times of emergency when multiple operations are required in the control room. -

b. The shift supervisor, until properly relieved, shall remain in the control room at all times during accident situations to direct the activities of control room operators. Persons authorized to relieve the shift supervisor shall be specified.

c. If the shift supervisor is temporarily absent from the control room during routine operations, a lead control room operator shall be designated to assume the control room command function. These temporary duties, responsibilities, and authority shall be clearly specified.

3. Training programs for shift supervisors shall emphasize and reinforce the responsibility for safe operation and the O management function that the shift supervisor is to provide for assuring safety.

4. The administrative duties of the shift supervisor shall be reviewed by the senior officer of each utility responsible for plant operations. Administrative functions that detract from or are subordinate to the management responsibility for assuring the safe operation of the plant shall be delegated to other operations personnel not on duty in the control room. ,

LILCO Position In lieu of a shift Supervisor SNPS uses a Watch Engineer. The administrative duties are minimized by the plant being shutdown and defueled with some systems being preserved / protected. Additionally, the administrative duties have little effect on safe operation of the plant. See the USAR for additional information. A Corporate Management Directive that emphasizes the primary t management responsibility of the Watch Engineer for the safe operation of the plant under all conditions is issued by the Assistant Vice President, Nuclear Operations annually.

i SHOREHAM DSAR

                  .#7 I.A.1.3                 Shift Manning NRC Position                                                                                                    ,

Assure that the necassary number and availability of personnel to ; man the operations shifts have been designated by the licensee. Administrative procedures should be written to govern the movement of key individuals about the plant to assure that i qualified individu&ls are readily available in the event of an abnormal or emergency situation. This should consider the

recommendations on overtime in NUREC-0578. Provisions should be made for an aide to the shift supervisor to assure that, over the long term, the shift supervisor is free of routine administrative : duties. At any time a licensed nuclear unit is being operated in Modes 1-4 for a PWR (Power Operation, Startup, Hot Standby, or Hot Shutdown, respectively) or in Modes 1-3 for a BWR (Power ? Operation, Startup, or Hot Shutdown, respectively), the minimum shift crew shall include two licensed senior reactor operators (SRO), one of whom shall be designated as the shift supervisor, : two licensed reactor operators (RO), and two unlicensed auxiliary l operators (AO). For a multi-unit station, depending upon the station configuration, shift staffing may be edjusted to allow : credit for licensed senior reactor operators and licensed reactor l

                    'l(~N       operators to serve as relief operators on more than one unit; i    however, these individuals must be properly licensed on each such unit. At all other times, for a unit loaded with fuel, the minimum shif t crew shall Anelude one shif t supervisor who whall be a licensed senior reactor operator (SRO), one licensed reactor operator (RO), and one unlicensed auxiliary operator (AO),                                                     t Adjunct requiremants to the shift staffing criteria stated above                                               j are as follows:

a. A shift supervisor with a senior reactor operator's licenso, who is also a member of the station supervisory t staff, shall be onsite at all times when at least one unit is loaded with fuel. l

b. A licensed senior reactor operator (SRO) shall, at all ;

times, be in the control room from which a reactor is 4 being operated. The shift cupervisor may from time-to-l t

                                                       . time act as relief operator for the licensed senior reactor operator assigned to the control room.

c. For any station with more than one reactor containing fuel, the number of licensed senior reactor operators onsite shall, at all times, be at least one more than the number of control rooms from which the reactort are being operated.

i

d. In addition to the licensed senior reactor operators specified in a., b., and c. above, for each reactor

j I SHOREllAM DSAR containing fuel, a licensed reactor operator (RO) shall be in the control room at all times.

e. In addition to the operators specified in a., b., c., and

d. above, for each control room from which a reactor is being operated, an additional licensed reactor operator l (RO) shall be onsite at all times and available to serve as relief operator for that control room. As noted above, this individual may serve as relief operator for each unit being operated from that control room, provided he holds a current license for each unit.

f. Auxiliary (non-licensed) operators shall be properly '

qualified to support the unit to which assigned.

g. In addition to the staffing requirements stated above, shift crew assignments during periods of core alterations shall include a licensed senior reactor operator (SRO) to directly supervise the core alterations. This licensed 1 senior reactor operator may have fuel handling cuties but ';

shall not have other concurrent operational duties. Licensees of operating plants and applicants for operating licensees shall include in their administrative procedures

 -r-     (required by licence conditions) provisions governing required shift staffing and movement of key individuals about the plant.

d' l These provisions are required to assure that qualified plant personnel to man the operational shifts are readily available in the event of an abnormal or emergency situation. These administrative procedures shall also set forth a policy, . the objective of which is to operate the plant with the required ! staff and develop vorking schedules such that use of overtime is avoided, to the extent practicable, for the plant staff who perform safety-related functions (e.g., senior reactor operators, , reactor operators, health physicists, auxiliary operators, I&C ! technicians and key maintenenance personnel). ! IE Circular No. 80-02, " Nuclear Power Plant Staff Work Hours,' , dated February 1, 1980 discusses the concern of overtime work for members of the plant staff who perform safety-related functions. j The staff recognizes that there are diverse opinions on the amcunt of overtime that would be considered permissible and that there is a lack of hard data on the effects of overtime beyond the generally recognized normal 8-hour working day, the effects of shift rotation, and other factors. NRC has initiated studies I in this area. Until a firmer basis is developed on working hours, the administrative procedures shall include as an interim measure the following guidance, which generally follows that of IE Circular No. 80-02, in the event that overtime must be used (excluding extended periode of shutdown for refueling, major maintenance or major

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

9 i SHOREHAM DSAR ! k ' plant modifications), the following overtime restrictions should ~ j be followed:- An individual shall not be permitted to work more than I (1) 12 hours straight (not including shift turnover time). (2) There should be a break of at least 12 hours (which can i include shift turnover time) between all work periods. . (3) An individual shall not work more than 72 hours in any l 7-day period. (4) An individual shall not work more than 14 consecutive days without having 2 consecutive days of f. ; However, recognizing that circumstances may arise requiring deviation from the above restrictions, such deviation may be authorized by the plant manager or his deputy, or higher levels of management in accordance with published procedures and with appropriate documentation of the cause. l If a reactor operator or senior reactor operator has been workir.g i more than 12 hours during periods of extended shutdown (e.g. , at ,' duties away from the control board), such individual shall not be assigned shift duty in the control room without at least a l 12-hour break preceding such an assignment. We encourage the development of a staffing policy that would ; permit the licensed reactor operators and senior reactor i operators to be periodically assigned to other duties away from , the control board during their normal tours of duty. l If a reactor operator is required to work in excess of 8 > continuous hours, he shall be periodically relieved of primary duties at the control board, such that periods of duty at the board do not exceed about 4 hours at a time. l l , The guidelines on overtime do not apply to the shift technical ' i advisor provided he or she is provided sleeping accomodations and a 10-minute availability is assured. ; operating license applicants shall complete these administrative procedures before fuel loading. Development and implementation of the administrative procedures at operating plants will be i reviewed by the Office of Inspection and Enforcament beginning 90 ' days after July 31, 1980. LILCO Position < i l The Shoreham Station Procedures implement the following , h 1. The minimum shift compicment consists of three operators and i a sufficient nunber of extra people in order to meet the Emergency Plan and Fire Protection Plan requirements. f i i

         - . . -          , _ . _ , , ,            x.,_ ... . , _ _ . . . . _ _ , . _ . . _ _ _ _ . - _ . . _ _ . . . . _ _ _ _ , . . . _       . _ . ,               ,

4 i

                            .                                              SHOREHAM DSAR b(k /Q    2.      The shift schedule conforms to the guidelines provided in the                                                 ,

Shoreham Station Procedure entitled Station Operations - Overtime Selection as it applies to the scheduling and ute of overtime.

3. The movement in the plant by members of the shift complement l are such that they may be easily an'd rapidly informed and/or ;

contacted and dispatched by the operators in the event an emergency situation arises. i The above items ensure that qualified plant personnel are available to man operational shifts. The Shoreham Station Procedure entitled Station Operations - , overtime Selection, implements the requirements of the Technical Specifications (Chapter 16) . The Operators are trained and qualified as outlined in the , Shoreham Station Procedure entitled Operations Section Training and Qualification Program. Since SNPS contains only one unit and since no other units are operated by LILCO, the zequirement that .! Auxiliary (nonlicensed) operators be properly qualified to support the unit to which assigned is not a problem at Shoreham. The Shoreham Nuclear Power Station is in complete compliance with \ 4

         ,f-s)   the portions of this Task Action Item that apply to a shutdown                                                        ,

j > and defueled plant. I.A.2.1 Immediate Uparade of Reactor Operator and Senior keactor Operator Trainino and Qualification NRC Positi,on Effective May 1, 1980, an applicant for a senior reactor operator r ( S'RO) license shall have four years of responsibic power plant experience of which at least two years shall be nuclear power , plar.t experience. Six months of the nuclear power plant experience shall be at the plant on which the applicant is ; licensing. A maximum of two years power plant experience may be i fulfilled by academic or related technical training, on a : one-for-one time basis. l Effective December 1, 1980, an applicant for a senior reactor

operator (SRO) license shall have held an operator's license for l one year. .

Effective August 1, 1980, an applicant for a senior reactor operator (SRO) license shall have three months of shift training as an extra man on shift. An applicant for a reactor (RO) license shall have three months training on shift as an extra person in the control room. (

Effective August 1, 3980, training programs shall be modified to i provides I

i SHORE!IAM DSAR , (_) 1. Training in heat transfer, fluid flow, and [' thermodynamics,

2. Training in the use of installed plant systems to control or mitigate an accident in which the core is severely '

damaged,

3. Increased emphasis on reactor and plant transients. !

Effective May 1, 1980, certifications that operator license applicants have learned to operate the controls shall be signed by the highest of corporate management for plant operation. LILCO Position . There will not be any applicants for SRO or RO licenses. The SNPS operator training program will provide training for ; subjects applicable to a shutdown and defueled plant and also for fuel handling operations. I.A.2.3 Administration of Trainino Procrams NRC Position kD

  \' #

Applicant 6 for operator licenses will be required to grant

permission to the NRC to inform their facility management

regarding the results of examinations. ; contents of the licensed operator requalification program shall be modified to include instruction in heat transfer, fluid flow, i thermodynamics, and mitigation of accidents involving a degraded , I Core. The criteria for requiring a licensed individual to participate i in accelerated requalification shall be modified to be consistent with the new passing grade for issuance of a license. I Requalification programs shall be modified to require specific reactivity control manipulations. Normal control manipulattens, such as plant or reactor startups, must be performed. Control L manipulations during abnormal or emergency operations shall be , walked. Simulator examinations will be included as part of the licensing ' examinations. t LILCO Position i It is LILCO's position that permanent members of the training j i ,( staff who teach systems, integrated responses, or transients be ; l \ qualified or certified to teach in the appropriate subject area.

       -r-       - ~ --- ,        -n.--.- - - .      , ,    ,,   . - . , . , _ .                                  _ , , _   _ _ _ _ _    _ _ _

SHOREHAM DSAR k( LILCO does not intend to require either quest lecturers who r u experts in particular subjects (reactor theory, instrumentati;n, v thermodynamics, health physics, chemistry, etc.) to successfully complete an NRC SRC examinations or system experts, such as an , instrument and control supervisor teaching the contro3 rod drive ! oystem to successfully complete an NRC SRO examination. The degree of training provided will be commensurate with the tasks required to,be performed. I.A.3.1 Revise scope and criteria for Licensino Examinations i NRC Position Applicants for operator licenses will be required to grant i permission to the NEC to inform their facility management i regarding the results of examinations. l contents of the licenses operator requalification program shall be modified to inc.lude instruction in heat transfer, fluid flow, l thermodynamics, and mitigation of accidents involving a degraded l Core. The criteria for requiring a licensed individual to participate ; in accelerated requalification shall be niodified to be consistent ! () with the new passing grade for issuance of a license. Requalification programs shall be modified to recuire specific Normal control manipulations, ; reactivity control manipulations. ; such as plant or reactor startups, must be performed. Control i manipulations during abnormal or emergency operations shall be walked. Simulator examinations will be included as part of the licensing i examinations. LILCO Position Licenses ! LILCO vill not have applicants for operator licenses. will be limited to fuel handling operations. i I.B.I.2 Evaluation of Orcanization and Manaaement i NRC Position Each applicant for an operating license shall establish an onsite independent safety engineering group (ISEG) to perform independent reviews of plant operations. I The principal function of the ISEG is to examine plant operating characteristics, NRC issuances, Licensing Information Service advisories, and other appropriate scurces of plant design and O operating experience information that may indicate areas for

improvinc plant safety. The ISEG is to perform independent i E

i [ SHOREHAM DSAR k' review and audits of plant activities including maintenance, modifications, operational problems, and operational analysis, j and aid in the establishment of programmatic requirenents for l plant activities. Where useful improvements can be achieved, it l 1s expected that this group will develop and present detailed ! recommendations to corporate management for such things as j revised procedures or equipment modifications. :

                   ' Another function of the ISEG is to maintain surveillance of plant                            j operations and maintenance activities to provide independent                                  ;

verifications that these activities are performed correctly and i that human errors are reduced as far as practicable. ISEG will i then be in a position to advise utility management on the overall ! quality and safety operations. ISEG need not perform detailed audits of plant operations and shall not be responsible for i sign-off functions such that it becomes involved in the operating ! organization. The new ISEG shall not replace the plant operations review committee (PORC) and the utility's independent review and audit ! group as specified by current staff guidelines (Standard Review l Plan, Regulatory Guide 1.33, Standard Technical Specifications). { Rather, it is an additional independent group of a miniraum of ; five dedicated, full-time engineers, located onsite, but l reporting offsite to a corporate official who holds a high-level, i technically oriented position that is not in the management chain l for power production. The ISEG will increase the available >

technical expertise located onsite and will provide continuing, { systematic, and independent assessment of plant activities. ; Integrating the Shift Technical Advisors into the ISEG in some way would be desirable in that it could enhance the group's ! contact with and knowledge of day-to-day plant operations and ! provide additional expertise. However, the shift technical l adviser on shift is necessarily a menter of the operating staff ! and cannot be independent of it. It is expected that the ISEG may interface with the quality assurance (QA) organization, but preferably should not be an integral part of the QA organization. The functions of the ISEG require daily contact with the operating personnel and continued access to plant facilities and records. The ISEG review functions can, therefore, best be carried out by a group physically located onsite. However, for utilities with multiple sites, it may be possible to perform portions of the independent, safety assessment function in a centralized location for all the utility's plants. In such cases, an onsite group still is required, but it may be slightly smaller than would be the case if it were performing the entire f independent safety assessment function. Such cases will be reviewed on a case-by-case basis. AT this time, the requirements for establishinc an ISEG is being applied only to applicants for operating licenses in accordance

SHOREHAM DSAR i kI) with Action Plan item 1.B.1.2. The staff intends to review this

   '--                 activity in about a year to determine its effectiveness and to                          i see whether changes are required. Applicability to operating plants will be considered in implementing long-term improvements                        i in organization and management for operating plants (Action Plan item I.B.1.1).
                      'LILCO Position                                                                         >

LIbCO does not have an ISEG. The functions of an ISEG do not apply to shutdown and defueled reactor which will not be operated. ; I.C.1 Guidance for the Evaluation and Development of Procedures ; for Transients and Accidents ; NRC Position In letters of September 13 and 27, October 10 and 30, and November 9, 1979, the Office of Nuclear Reactor Regulation required licensees of operating plants, applicants for operating ; licenses n' 'dcensees of plants under construction to perform

l analyses of taansients and accidents, prepare emergen-" procedure quidelines, upgrade emergency procedurcs, including ; :edures for operating with natural circulation conditions, anu to conduct ,

fr~N operating retraining (see also Item I.A.2.1). Emergency i ( ,) procedures are required to be consistent with the actions ; necessary to cope with the transients and accidents analyzed. Analyses of transients and accidents were to be completed in , early 1980 and implementation of procedures and retraining were t to be completed 3 months after emergency procedure guidelines were established; however, some difficulty in completing these requirements has been experienced. Clarification of the scope of the task and appropriate schedule revisions are being developed. In the course of review of these matters on Babcock and Wilcox : (b&W)-designed plants, the staff will follow up on the bulletin and orders matters relating to analysis methods and results, as listed in NUREG-0660, Appendix C (see Table C.1, items 3, 4, 16, , 18, 24, 25, 26, 27; Table C.2, items 4, 12, 17, 18, 19, 20; and Table C.3, items 6, 35, 37, 38, 39, 4*, 47, 55, 57). Based on staff reviews to date, there appear to be some recurring r deficiencies in the guidelines being developed. Specifically, the staff has found a lack of justification for the approach used , (i.e., symptom , event , or function-oriented) in developing diagnostic guidance for the operator and in procedural ' development. It has also been found that although the guidelines take implicit credit for operation of many systems or components, ; they do not address the avilaability of these systems under expected plant conditions nor do they address corrective or e alternative actions that should b perforned to mitigate the event should these systems or components fail.

  !{ }

i

SHOREHAM DSAR I ) The analysis conducted to date for guideline and procedure development contain insufficient information to assess the extent to which multiple failures are considered. NUREG-0578' concluded that the single-failure criterion was not considered appropriate for guideline development and called for the considreation of multiple failures and operator errors. Therefore, the analyses that support guideline and procedure development should consider the occurrences of multiple and consequential failures. In general, the sequence of events for the transients and accidents and inadequate core cooling analyzed should postulate multiple f ailures such that, if the f ailures were unmitigated, conditions of inadequate core cooling would result. Examples of multiple failure events include:

1. Multiple tube rupture in more than one steam generator and tube rupture in more than one steam generators ;

i

2. Failure of main and auxiliary feedvaters i

3. Failure of high-pressure recctor coolant makeup system; \

4. An anticipated transient without scram (ATWS) event following a loss of offsite power, stuck-open relief !

valve or safety relief valve, or loss of main feedwater; I () 5. Operator errors of omission or commission. , The analyses should be carried out far enouch into the event to ! assure that all relevant thermal / hydraulic /neutronic phenomena r are identified (e.g., upper head voiding due to rapid cooldown, i i steam generator stratification). Failures and operator errors during the long term cooldown period should also be addressed. The analyses should support development of guidelines that define j a logical transition from the emergency procedures into the ; inadequate core cooling procedure including the use of I instrumentation to identify inadequate core cooling conditions, i Fationale for this transition thould be discussed. Additional i information that should be submitted includes:

1. A detailed description of the methodology used to develop I the guidelines; [

Associated control function diagrams, sequence-of-event i 2. diagrams, or others, if used; .

3. The bases for multiple and consequential failure ,

considerations; ;

 ,         4. Supporting analysis, including a description of any computer codes usea;

f SHOREHAM DSAR i () 5. A description of the applicability of f l owners' group or vendor submittals may be referenced as aporopriate to support this reanalysis. review, a brief description of the submittals and justification required. procedures described in task action plan Item I.C.8 (NUREG-0660). For PWRs, this will involve review of the loss-of-coolant, steam generator-tube rupture, loss of main feedwater, and inadequate core cooling procedures. The adequacy of each PWR vendor's guidelines will be identified to each NTOL during the emergency procedure review. Since the Followine approval of analysis and guidelines and the pilot monitoring of emergency procedures, the staff will advise their plants. Consideration will be given to human factors review, as described in task action plan Item I.C.8, required or applicants for operating licenses will be as described in the long-term plan. Depending submitted to support development of emergency procedures for each reactor type or vendor, this transition may take place at different times. For example, if the GE quidelines are shown to plants and applicants will then have the option of implementing the long-term plan in a manner consistent with their operating schedule, provided they meet the final date required for implementation. This may require a plant that was reviewed for an operating license under Item I C.9 to revise its emergency any generic results to plant-specific applications. ; If owners' group or vendor submittals have already been forwarded to the staf f for , l of their adequacy to support guideline development is all that is ! l Pending staff approval of the revised analyais and guidelines, the staff will continue the pilot monitoring of emergency , . : ! ! ! analysis and guidelines i submitted by the General Electric Company (GE) owners' group that i comply with the requirements stated above have been reviewed and i approved for trail implementation of six plants with applications l for operatinc licenses pending, the interim program for BWRs will i consist of trial implementation on these six plants. l t ' all licensees of the adequacy of the guidelines for application to ! engineering and system operational characteristics, such as i information transfer under stress, compatibility _with operator i training and control room design, the time required for component i and system response, clarity of procedural actions, and centrol room personnel interactions. When this determination has been made by the staff, a long-term plan for emergency procedure ; will be made available. At that time, the reviews currently being conducted f on NTOLs under Item I.C.8 will be discontinued, and the review ! ! on the information , : ! l be effective on the six plants chosen for pilot monitoring, the j long-term plan for BWRs may be complete in early 1981. Operating : : ! ; ; i

              <          procedures again prior to the f2ual implementation date for Item I.C.B. The extent to which the long-term program will include                                                                    .,

review and approval of plant-specific procedures for operating ! plants has net been established. Our objective, however, is to !

SilOREllAM DSAR ; minimize the amount of plant-specific procedure review and ; approval required. The staff believes this objective can be acceptably accomplished by concentrating the staff review and approval on generic guidelines. A key element in meeting this ; objective is use of staff-approve ( generic guidelines and j guideline revisions by licensees to develop procedures. For this approach to be ef f ective, it is imperative that, once the staf f has issued approval of a guideline, subsequent revisions of the guideline should not be implemented by licensees until reviewed and approved by the str.ff. Any changes in plant-specific , procedures basud on unapproved guidelines could constitute an ! unreviewed cafety issue under 10CTR50.59. Deviations from this approach on a plant-specific basis would be acceptable provided the basis is submitted by the licensee for staff review and ' approval. In this caso, deviations from generic guidelines { should not be implemented until staf f approval as formally i ' received in writing. Interim implementation of analysis and procedures for small-break loss-cf-coolant accident and inadquate core cooling should remain on the schedule contained in l NURIG-0578, Eecommendation 2.1.9. LILCO Position l The description contained under this heading in the latest revision of Shoreham USAR remains unchanced with the following i exception: LILCO will not continue to participate in the DWR owners' Group program to develop Emergency Procedure Guidelines l for GE Boiling Water Ecactors. Refer to USAR for information on j this subject. 4 1.C.2 Ehift and Felief Turnover Procedures NRC Porition . The licensees shall review and revise as necessary the plant 5 procedure for shift and relief turnover to assure the followina: l

1. A checklist shall be provided for the oncoming and offacing control room operators and the oncoming shift ;

supervisor to complete and sign. The following items, as a minimum, shall be included in the checklist: l

a. Assurance that critical plant parameters are within allowable limits (parameters and allowable limits I shall be listed on the checklist).

l

b. Assurance of the availabilty and proper alignment of i l all systems essential to the prevention and mitigation of operational transients cnd accidents by a check of the control console (wlat to check and criteria for acceptable states shall te -luded on

(~ ( the checklist). l l

f FliGREHAM DSAR f N"i c. Identification of systems and components that are in , a degraded c.cna if operation permitted by the Technical Specifications. For such systems and ; components, the length of time in the degraded mode shall be compared with the Technical Specifications action statement (this shall be recoreded as a ; r separate entry on the checklist).

2. Checklists or logs shall be provided for completion by the ongoing and oncoming auxiliary operators and technicians.

such checklists and logs shall include any equipment under maintenance or test that by themselves could upgrade a system , critical to the prevention and mitigation of operational transients and accidents or initiate an operational , transients (what to check and criteria for acceptable status shall be included on the checklist), and

3. A system shall be established to evaluate yhe effectiveness of the shift and relief turnover procedure (for example, periodic independent verification of system alignments).

LILCO Position 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.C 3 Shift Supervisor Responsibilities NRC Position In the letters of September 13 and 27, October 10 and 30, and November 9, 1979, NBC required licensees and applicants to review and revise as necessary plant procedures and directives to assure that the duties, responsibilities, and authority were properly : defined to establish a definite line of command and clear ' delineation of the command decision authority of the supervisor in the control room relative to otner plant management personnel. These letters also emphasized the primary management responsibility of the shift supervisor for safe operation of the plant. Training programs for the shift supervisor were required to emphasize and reinforce the responsibility for sfae operation and management functionof the shift supervisor to assure safe operation of the plant. LILCO Position The description contained under this heading in the latest Refer to USAR for revision of Shoreham USAR remains unchanged.

!()    information on this subject.

SHOREFAM DSAR ip L\_/ I.C 4 Control Room Access NBC Position The licensee shall make provisions for limiting access to the control room to those individuals responsible for the direct operation of the nuclear power plant (e.g., operations supervisor

               'shif t supervisor, and control room operators), to techntcal advisors who' may be requested or required to support the operation, and to predesignated NRC personnel. Provisicns shall include the following:

1. Develop and implement an administrative procedure that establishes the authority and respcnsibility of the person in charge of the cintrol room to limit access.

2. Develop and implement precedures that establish a clear line of authority and responsibility in the control room in the event of an emergency. The line of succession for the person in charge of the contro1 room shall be established and lim'.ted to persons possessing a current senior reactor operator's license. The plan shall clearly define the lines of communication and authority s

for plant management personnel not in direct command (f

    ,.7
          '                     operations, including those who report te b?.ations
     ^~                         outside of the control room.

LILCO Position A Shoreham Station Procedure on Control Room Conduct establishes s se authority and responsibility cf the person in charge of the t mtrol room to limit access to the control room. The same procedure estsblishes the line of authority and responsibility in the control room in the event of an emergency. I.C.5 Procedures for Feedback of Operatino Experience to Plant Staff NRC Position In accordance with Task Action Plan I.C.5, Procedures for Feedback of Operating Experience to plant Staff (NUREG-0660), each applicant for an operating license shall prepare procedures to assure that operating information pertinent to plant safety originating both within and outside the utility organization is continually supplied to operators and other personnelThese and is incorporated into training and retraining programs. procedures shall: (_j (1) Clearly identify organizational responsibilities for review of operating experience, the feedback of pertinent inforcation to operators and other personnel, and the incorporation of such information into training and j retraining programs; i

I SHOREHAM DSAR p Identify the administrative and technical review steps b(_).. "",- ~(5)

                  ~          necessary in translating recommendations by the operating
                            ' experience assessment group into plant actions (e.g.,

changes to procedures, operating orders); (3) Identify the recipients of various categories of information from operating experience (i.e., supervisory personnel, shift technical advisors, operators, maintenance personnel, health physics technicians) or otherwise provide means through which such information can be readily related to the job function of the recipients: (4) Provide means to assure that effected personnel become aware of and understand information of sufficient importance that should not wait for emphasis through routine training and retraining programs; (5) Assure that plant personnel do not routinely receive extraneous and unimportant information on operating experience in such volume that it would obscure priority information or otherwise detract from overall job performance and proficiency; (6) Provide suitable checks to assure that conflicting or s contradictory information is not conveyed to operators and other personnel until resolution is reached; and, (V ) (7) Provide periodic internal audit to assure that the feedback program functions effectively at all levels. Each utility shall carry out an operating exp3rience assessment function that will involve utility personnel having collective competence in all areas important to plant safety. In connection with this assessment function, it is important that procedures exist to assure that important information on operating experience originating both within and outside the organization is continually provided to operators and other personnel, and that it is incorporated into plant operating procedures and training and retraining programs. Those involved in the assessment of operating experience will

sfew information from a variety of sources. These include operating information fron the licensee's own plant (s) ,

publications such as IE Bulletins, Circulars, and Notices, and pertinent NRC or industrial assessments of operating experience. In some cases, information may be of sufficient importance that it must be dealt with promptly (through instructions, changes to operatinc and emergency procedures, issuance of special changes to operating and emergency procedues, issuance of special precautions, etc.) and must be handled in such a manner to assure fr's that operations management personnel would be directly involved Eq,) in the process. In many other cases, however, important information will become available which would be brought to the attention of operators and other personnel for their genercl

I SHOREHAM DSAR O information to assure continued safe plant operations. Since the total volume of information handled by the assessment group may - be large, it is important that assurance be provided that ; high-priority matters are dealt with promptly and that discrimination is used in the feedback of other information so that personnel are not deluged with unimportant and extraneous information to the detriment of their overall proficiency. It is , important, also, that technical review be conducted to preclude , premature dissemination of conflicting or contradictory . information. ! LILCO Position The description contained under this heading in the latest revision of the USAR remains unchanged with the following exceptions:

1. The membership and quorum requirements of the ROC are given '

in the Technical Specifications.

2. The Nuclear Review Board and ISEG have been eliminated.

Refer to Section 13.4.2 and 13.4.3 for justification. <

3. The Training Division Manager has been changed to Office of Training.

4. The responsibilities assicned to the Plant Manager, Division i Manager and Section Heads have been reassigned to plant management personnel (Section Heads and above).

5. The Training Administrative Supervisor administers the circulation of required reading lists.

I.C.6 Procedures for Verification of Correct Performance of , Operating Activities NRC Position It is required from NUREG-0660 that licensees' procedures be i reviewed and revised, as necessary, to assure that an effective system of verifying the correct performance of operating activities is provided as a means of reducing human errors and improving the quality of normal operationr. This will reduce the frequency of occurrence of situations Such a that could verification system result may in or include , contribute to accidents. automatic system status monitoring, human verification or operations, and maintenance activities independent of the people or performing the activity (see NUREG-0585, Recommendation 5) ' both. () Implementation of automatic status monitoring if required will reduce the extent of human verification of operations and maintenance activities but will not verification in all instances. Theeliminate procedures theadopted need for by such the

SHOREHAM DSAR licensees may consist of two phases - one before and one after installation of automatic status monitoring equipment, if - requirco, in accordance with item I.D.3 of NUREG-0660. l An acceptable program for verification of operating activities is described below. , The American Nuclear Society has prepared a draft revision to ANS l Standard N 18.7-1972 ( ANS) . Administrative Controls and Quality Assurance for the Operational Phase or Nuclear Power Plants. A ' second proposed revision to Regulatory Guide 1.33, Quality Assurance Program Requirements (Operation), which is to be issued for public comment in the near future, will endorse the latest , draft revision to ANS 3.2 subject to the following supplemental provisions: ; (1) Applicability of the guidance of Section 5.2.6 should be ' extended to cover surveillance testing in addition to . maintenance. ; (2) In lieu of any designated senior reactor operator (SRO), the ' authority to release systems and equipment for maintenance ' or surveillance testing or return-to-service may be delegated to an on-shift "90, provided provisions are made , tu ensure that the shif'. pervisor is kept fully informed of system status. (( ) (3) Except in cases of significant radiation exposure, a recond qualified person should verify correct implementation or o.' equipment. ; equipment control measures such as tagging (4) Equipment control procedures should include assurance that control-room operators are informed of changes in equipment t status and the effects of such changes, (5) For the return-to-service of equipment important to safety, a second qualified operctor should verify proper systems alignment unless functional testing can be performed without compromising plant safety, and can prove that all equipment, valves, and switches involved in the activity are correctly aligned. NOTE: A licensed operator possessing knowledge of the systems involved and the relationship-of the systems to plant safety would be a qualified person. The staff is investigating the level of qualification necessary for other operators to perform these functions. For plants chat have or will have automatic system status monitoring as discussed in Task Action Plan Item I.D.3,

 /r -

NUREG-0660, the extent of human verification of operations and I! maintenance activities will be reduced. However, the need for such verification will not be eliminated in all instances.

SHOREHAM DSAR (N/ f LILCO Position The description contained under this heading in the latest revision of the USAR remains unchanged with the following exceptions:

1. The Watch Engineer;

a. gives permission to release plant systems or equipment for maintenance, surveillance tests or return to rervice.

b. must be informed of changed in equipment status and the effect of such changes.

c. provides final acceptance for return to service.

2. Considerations for shutdown margin and decay heat removal are not applicable.

I.C.7 USSS Vender Review of Procedures . NRC Position Applicants for near-term operating licenses will be required to obtain NSSS vendor review of low-power and power-ascension test

,. , 3 and emergency procedures (see Regulatory Guide 1.33, Appendix A, 1'-)   Section 6) as a further verification of the adequacy of the procedure =                 After trial use of this requirement on a few pending operating license applications, the staff will decide whether its further use or expansion to include procedure review by the A-E is desirable.                This decision will be made in light of the long-term program described in Item I.C.9.                                   See also Table Col, Item 4a and Table C.3, Item 50 of NDREG-0660.

LILCO Position Emergency procedures were prepared using the Emergency Procedures Guidelines developed by the Emergency Procedures Subgroup of the BWR Owners' Group and the General Electric Company. Low-power and power-ascension test and emergency procedures will not be used at Shoreham. I.C.8 Pilot Monitorino of Selected Emeroency Procedures For Near-Term Operatino License Applicants NRC Position Correct emergency procedures as necessary based on the NRC audit of selected plant emergency operating procedures (e.g., small pg break loss-of-coolant accident, loss of feedwater, restart of N) engineered safety features following a loss of ac power and steam-line break).

SHOREHAM DSAR ([% LILCO Position The generic guidelines prepared by the BWR Owners' Group and approved by the NRC for trial implementation at the Shoreham Nuclear Power Station have been. incorporated into the Shoreham Emergency Operating Procedures (refer to NUREG-0737 Item I.C.1). The completed procedures have received an extensive in-house review, and were subjected to simulator verification. The verified procedures were submitted for NRC review. Any comments or corrections found necessary as a result of the NRC audit were evaluated and implemented, as appropriate. This item was closed in Section 3.1.1 of Inspection Report 83-09. I.D.1 Control Room Desion Reviews NRC Position In accordancv with Task Action Plan I.D.1, Control Room Design Reviews (NUREG1-06 6 0 ) , all licensees and applicants for operating licenses will be required to conduct a detailed control-room design review to identify and correct design deficiencies. This detailed control-room design review is expected to take about a year. Therefore, the Office of Nuclear Reactor Regulation (NRR) requires that those applicants for operating licenses who are unable to complete this review prior to issuance of a license make preliminary assessments of their control rooms to identify significant human factors and instrur..entation problems and establish a schedule approved by NRC for correcting deficiencies. These applicants will be required to complete the more detailed control room reviews on the same schedule as licensees with operating plants. NRR is presently developing human engineering guidelines to assist each licensee and applicant in performing deteiled control-room review. A draft of the guidelines has been published for public comment as NUREG/CR-1580, " Human Engineering Guide to Control Room Evaluation." The due date for comments on this draft document was September 29, 1980. NRR will issue the final version of the guidelines as NUREG-0700, by February 1981, after receiving, reviewing, and incorporating substantive public comments from operating reactor licensees, applicants for operating licenses, human factors engineering experts, and other interested parties. NRR will issue evaluation criteria, by July 1981, which will be used to judge the acceptability of the detailed reviews performed and the design rodifications implemented. Applicants for operating licenses who will be unable to complete the detailed control-room design review prior to issuance of a license are required to perform a preliminary control-room design assessment to identify significant human factors problems. Applicants will find it of value to refer to

      .(^}     the draft document NUREG/CR-1580, " Human Engineerino Guide to x_/

Control Ecom Evaluation," in performing the preliminary assessment. NRk will evaluate the applicants' preliminary assessmens includinc the performance by NRR of onsite review /

SHOREHAM DSAR

 ,m
 '-          audit. The NRR onsite review / audit will be on a schedule consistent with licensing needs and will emphasize the following aspects fo the control room:

1. The adequacy of information presented to the operator to reflect plant status for normal operation, anticipated operational occurrences, and accident conditions;

2. The. groupings of displays and the layout of panels;

3. Improvements in the safety monitoring and human factors enhancement of controls and control displays;

4. The communications from the control room to points outside the control room, such as the onsite technical support center, remote shutdown panel, offsite telephone lines, and to other areas within the plant for normal and emergency operation.

5. The use of direct rather than derived signals for the presentation of process and safety information to the operator;

6. The operability of the plant from the control room with

 .,r~s                   multiple failures of nonsafety-grade and nonseismic Q                        systems;

7. The adequacy of operating procedures and operator training with respect to limitations of the instrumentetion displays in the control room;

8. The categorization of alarms, with unique definition of safety alarms.

9. The physical location of the shift supervisor's office either adjacent to or within the control-room complex.

Prior to the onsite review / audit, NRR will require a copy of the applicant's preliminary assessment and additional information which will be used in formulating the details of the onsite review / audit. LILCO Position LILCO has performed the required preliminary design assessment of the Shoreham control room and remote shutdown panel. The intent of the review is to identify significant human factors and instrument problems and to establish a schedule for correcting any deficiencies.

         )      LILCO retained the General Physics Corporation, as a human

{s/. factors consultant, to perform the preliminary assessment. General Physics has been involved in similar audits LaSalle, at the followinc near-term operating license BWR's: Susquehanna, and Zimmer. l r

                                                                                                                                         ...J

i SHOREHAM DSAR l I The criteria and checklists used at Shoreham considered the draft NUREG/CR-1580 ' Human Engineering Guide to Control Room- [ Evaluation" and the BWR Owners' Group Control Room Human Factor j Review guidelines. l t The preliminary assessment report detailing the resulting findings and a schedule for correcting deficiencies was submitted to the NRC March 12, 1981, SNRC-536. j i NRC performed a five day onsite review / audit of the Shoreham control room beginning March 30, 1981. A fit.al report of their l j findings (CRDR/A report) was issued May 10, 1981. : A Detailed Control Room Design Review (DCRDR) has been performed. [ The items identified by this review have been placed on hold due l~ to the LILCO and New York State Shoreham settlement. 7.D.2 riant Safety Parameter Display Console l

                                                                                                             ,s NRC Position In accordance with Task Action Plan I.D.2, Plant Safety Parameter Display-Console (NUREG-0660), each applicant and licensee shall install a safety parameter display system (SPDS) that will                                         i display to operating personnel a minimum set of parameters which                                   l define the safety status ofthe plant. This can be attained e

9 through continuous indication of direct and derived variables as [ necessary to assess plant safety status. These requirements are defined in NUREG-0696 which was published { in February 1981. :

                                                                                                             ^
                                                                                                             ?

LILCO Position Since the Shoreham Nuclear Power Station will not be operating , and because its reactor is defueled, the SPDS is not needed. [ I.G.1 Trainina Recuirements Durina Low Power Testinc t NRC Position . Define and commit to a special low power testing program approved- ! i by the NRC to be conducted at power levels no greater than five percent to obtain additional technical information and ; supplemental training. , LILCO Position 3 A low power test program will not be conducted. This itec is not ; applicable to LILCO. L

__ ,m_ . . - _ _ . , _ . . - . , .-

SHOREHAM DSAR ((p/~N 's- II.B.1 Reactor Coolant System Vents This system is specific to the reactor coolant system and is not neaded to support the storage of the fuel in the fuel pool. II.B.2 Plant Shielding to Provide Access to Vital Areas and Protect Safety Equipment for Post-Accident Operation NRC Position The NRC position on the above is as given in the Shoreham USAR. LILCO Position In a fashion similar to the Shoreham USAR, LILCO has determined that the only areas after an accident where access may be needed are the Radwaste and Main Control Rooms and the Technical Support Center (TSC) full-time, and the Reactor Building refueling deck part-time. The basis of this is that, as seen in Chapters 11 and 15 of the DSAR, the only design basis events which remain credible are the Fuel Handling Accident (FRA) as described in Regulatory Guide 1.25, and the Liquid Radwaste Tank Rupture Accident. As discussed in Chapters 15 and 11, respectively, neither of these involve the release of large goantities of radioactivity, and thus there is no need The for the Post other Accident areas suggested

   '( 3

() Sampling and Analysis Facility (PASF). as vital post accident in NUREG-0737 do not apply for Shoreham or are not needed, for the reasons given in Section II.B.2 of the USAR. Furthermore, systems such as the hydrogen recombiner are as clearly no longer required in the defueled condition, documented in Chapter 6 of the DSAR. Source Term and Results The radioactive source terms (quantities and source distributions) for the design basis accidents are described in Section 11.2 for the Liquid Radwaste Tank Rupture Accident, and in Chapter 15.1.36 for the FHA (and Worst Case Fuel Damage Accident). Because the cubicles where a Liquid Radwaste Tank Rupture Accident could occur are exhausted to a process air header, which is then processed through a filter train before release through on the station vent, the postulated accident would have no affect general Radwaste Building habitability. Specifically, the Radwaste Control Room would be unaffected, since it is well For 1solated from the cubicles where the accident could occur. details, see Section 9.4.3 of the USAR. For the FHA and Worst Case Fuel Damage Accidents, the

                 /~T         habitability criteria for full occupancy of the Main ControlThe     Room
                 \-)         and Technical Support Center are clearly not challenged.

30-day ir.tegrated doses are:

i SHORIHAM DSAR g-

 \_/            Control Room Dose, mrem Whole Body                            S'k in Puel Handling Accident                                                                        9.59E-05                          2.08E-01 Worst Case Fuel Damage                                                                        5.92E-02                          1.28E+02 Accident Technical support Center Dose, mrem Whole Body'                         Skin Fuel Handling Accident                                                                       5.02E-05                         1.04E-01 Worst Case Fuel Damage                                                                        3.10E-02                        6.42E+01 For any fuel damage accident with the spent fuel in the pool, clearly the highest dose rate area in the plant will be the refuelina floor. With the Reactor Building Normal Ventilation System (RDNVS) operating throughout the event, airborne concentrations and associated dose rates are quickly dissipated in the Reactor Building.                                                   The time-dependent dose rates are es follows:

() Fuel Handling Accident Time After Accident, hrs Whole Body Dose Rate, mrem /hr Skin 0 7.53E-02 7.llE+00 2 3.16E-04 2.98E-02 8 2.32E-ll 2.20E-09 Worst Case Fuel Damage Accident Dose Rate, mrem /hr Time After Accident, hrs Whole Body Skin 0 4.65E+01 4.39E+03 2 1.95E-01 1.84E+01 8 1.43E-08 1.36E-06 Integrated 30-day doses, even assuming full occuparcy for the Reactor Building refueling deck area, are as follows: Dose, mrem Whole Body Skin Fuel Handling Accident 2.75E-02 2.60E+00 Worst Case Fuel Damage 1.70E+01 1.60E+03 (_/ The above are clearly of no concern for post-secident plant habitability. l

SHOREHAM DSAR 1 (D N_/ Harsh Environment Due to the lack of safety-related equipment in the Radwaste Building, there are no harsh environment concerns there. Neither the Fuel Handling nor the Worst Case Fuel Damage Accident involve the release of any meaningful quantity of heat energy or chemically ha:ardous material. Furthermore, the gamma and beta

        . doses given above are insignificant                    insofar as environmental qualification is concerned. As such, the credible accidents considered do not result in harsh environment in the Reactor Building or elsewhere.

II.B.3 Post-Accident Sampling This system samples the reactor and containment and is not needed to support the safe storage of the fuel in the fuel pool. II.B.4 Trainina for Miticating Core Damaae This trainina is not needed to support the storage of the fuel in the fuel pool because it applies to an operating reactor. II.B.7 Analysis of Hydrocen Control This system is not needed to support the storage of the fuel in j the fuel pool because the primary containment is not required for a defueled reactor. II.D.1 Performance Testino of BWR and PWR Relief __and Safety Valves This system is not needed to support the storage of the fuel in the fuel pool because the reactor is defueled and unpressuri:ed. II.D.3 Relief and Safety Valves Position Indication This system is not needed to support the storage of the fuel in the fuel pool. (See II.D.1) II.E.4.1 Containment Dedicated Penetrations This system is not needed to support the storage of the fuel in the fuel pool because the primary containment is not required. II.E.4.2 Containment Isolation Dependability This system is not needed to support the storage of the fuel in the fuel pool. (See item above) II.F.1 Additional Accident Monitorina Instrumentation x_/ Eefer to Chapters 12 and 15 regarding instrumentation requirements for a defueled reactor. i

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

s t I SHOREHAM DSAR ; v

      )             Identification of and Recovery from Conditions Leading to                                             l II.F.2 Inadequate Core Cooling                                                      .

l t This item is specific to an operating reactor and is not needed ; to support the safe storage of the fuel in the fuel pool. . II.K.1.5 Safety Related Valve Position These valves are not needed to support the storage of the fuel in j the fuel-pool. j l II.K.l.10 Operating Status j This section applies without change. . (plant Staff to confirm) , II.K.1.22 Auxiliary Heat Removal System Procedure , This procedure refers to an operating reactor and is not needed to support the storage of the fuel in the fuel pool. l t II.K.1.23 RV Level, Procedures f The reactor vessel level procedures are not required. ; i II.K.3.3 Failure of Power Operated Relief Valve or Safety Valve j to Close l This system is not needed because the reactor is not pressurized. II.K.3.13 Separation of HPCI and RCIC System Initiation _ Levels - l Analysis and Implementation The HPCI and RCIC systems are not needed to support the storage \ of the fuel in the fuel pool. ; II.K 3.15 Modify Break Detection Logic to Prevent Spurious Isolation of HPCI and RCIC Systems i This system is not needed to support the storage of the fuel in , the fuel pool (See item above). ! l II.K.3.16 Reduction of Challenges and Failures of Relief Valves ;

                                   - Feasibility Study and System Modification This system is not needed to support the storage of the fuel in                                               ;

the fuel pool. , [NSSS will no longer challenge SRV's.) >

  \                                                                                                                       1 6
                      ..___ _ _ -             _ m - .-     -
                                                                                 - - , .                             4 -

SHOREHAM DSAR fr\ Report on Outage of ECC Systems -Licensee Report and II.K.3.17 Proposed Technical Specification Changes This system is not needed to support the storage of the fuel in the fuel pool. Refer to Chapters 9 and 15 for system requirements.

     ,II.K.3.18' Modification of Automatic Depressurization System Locic - Feasibility for Increased Diversity for Some E~ent- Sequences This system is not needed to support the storage of the fuel in the fuel pool because the reactor is not pressurized.

II.K.3.21 Restart of Core Spray and LPCI Systems on Low Level This system is not needed to support the storage of the fuel in the fuel pool. [CS and LPCI not required for defueled reactor.) II.K.3.22 Automatic Switchover of Reactor Core Isolation Cooling System Suction - Verify Procedures and Modify Design This system is not needed to support the storage of the fuel in the fuci pool. II.K.3.24 Confirm Adequacy of Space Coolina for High-Pressure Coolant Injection and Reactor Core Isolation Cooling b9 stems This system is not needed to support the storage of the fuel in the fuel pool. II.K.3.25 Effect of Loss of AC Power on Pump Seals This system is not needed to support the storage of the fuel in the fuel pool. [This issue deals with Reactor Recirculation Pumps which are no longer needed.) II.K.3.27 Provide Common Reference Level for Vessel level Instrumentation This system is not required for a defueled reactor. ADS II.K.3.28 Study and Verify Oualification of_ Accumulators or Valves This system is not needed because the reactor is unpressurized.

SHOREHAM DSAR ('~) II.K.3.30 Revised Small-Break LOCA Methods to Show Compliance with 10CFRSO.46, Appendix ,K : LOCA's are not possible for a defueled reactor. II.K.3.31 Plant-Specific Calculations to Show Compliance with 10CFR50.46 This item is not needed to support the storage of the fuel in the fuel pool (See item above). , II.K.3.44 Evaluation or Anticipated Transients with Sincie Failure to Verify No Fuel Failure only the loss of Normal AC Power is applicable to the defueled plant configuration. See Table I. Without any inventory makeup to the pool the evaporation rate would be approximately .6 gpm. , Chapter 15 c_monstrates that there are no radiological ; consequences associated with this event. II.K.3.45 Evaluation of Depressurization with Otaer Than Full i ADS . This system is specific to operating reactor conditions and is not needed to support the storage of the fuel in the fuel pool. a [,

 'A       II.K.3.46    Response to List of Concerns from ACRS Consultant (Mr.     :

C. Michelson) This system is not needed to support the storage of the fuel in ! the fuel pool. [These questions address concerns with protecting the fuel in the Vessel Core.) , III. EMERGENCY PREPAREDNESS (Refer to LILCO Defueled Emergency Plan) III.A.l.1 Uporade Emercency Preparedness NRC Position ; The overall state of emergency prepardness for nuclear power plant accidents will be upgraded, including the integration of emergency preparedness onsite and offsite, accordingApproval to the of NRC/ FEMA Memorandum of Understanding (item III.B). the overall state of preparedness will be required prior to issuance of an operating license. The review and upgrading for operating reactors is under way, h ') Six NRC teams were formed in September 1979 to iraplement the

           " Action Plan for Promptly Improving Emergency Preparedness" (SECY 79-450). That Action Plan identifies the elements required for i

i _ m.

SHOREHAM DSAR C t

 '    promptly improving licensee emergency preparedness and for ensuring the capability of offsite agencies to take appropriate emergency actions. In the short term, the teams are making an integrated assessment of licensee, local, and State capabilities and interfaces based on: (a) a review of existing plans and a meeting in the site area to communicate upgraded criteriaThis                                 and to identify -               '$censees the areas requiring improvements.

includes an opportunity for expression of concerns by the public through an open mehting. An objective of the teams is to help improve working relationships and communications concerning emergency plan development among all parties. The criteria being used by the NRC teams reflect a number of the recommendations made as a result of the TMI-2 accident by the President's Commission and the NRC Special Inquiry Group; and (b) a review of upgraded licensee, local, and State plans submitted by the licensee after the site visit is summarized in a safety evaluation report. This includes an identification of areas requirina improvement, a schedule for implementation of the improvements, and a specification of any required interim measures. The review of upgraded plans encompasses the points in SECY-19-450 and reflects any input from the Federal Regional Advisory Committees (RAC). Items on local or State plans requiring improvement to meet the upgraded criteria of NUREG-0654 but which are adequate to meet the essential planning elements of (' "KRC Guide and Checklist," NUREG-75/111, and Supplement 1 (_)g thereto, are not being required for issuance of licenses for low-power testing. The above actions are in progress and will be completed in FY 1980. In the longer term, beginning in FY 1981, an integrated assessment of the implementation of the plans will be performed. This assessment will take into account comments and revices by the RAC as a result of State plan concurrence efforts, including critiques of energency exercises. The results of the Office of Inspection and Enforcement (IE) special team efforts to evaluate Licensee health physics procrams during 1980-81 will be factored into the review. This longer term review of emergency preparedness will consist of three parts: (a) a review of implementing procedures, including inplant and offsite personnel and equipment. The review of these procedures will be done by the team. Subsequently, periodic reviews and inspections will be performed by IE; (b) observing and critiquing exercises involving licensee, local, and State capabilities; and (c) observing and critiquing exercises involving licensee, local, State and Federal capabilities. For new operating license applicants, this must be completed before full-power licensing and within about five years for operating reactorc. ERF has sent letters to operating reactors, operating license applicants, and holders of construction permits requestina

   ;                 information recarding time estimates for evacuation of areas
   ^                 arcund plants to determine the difficulty of implementinc protective measures for the public.

f l

SHORERAM DSAR /^N LILCO Position Refer to the emergency plan for the Shoreham site which is being submitted for review and approval as a separate document entitled, Defueled Emergency Preparedness Plan, via letter SNRC-1651. The information contained in this document supersedes in its entirety the information originally submitted as part of the FSAR. III.A.1.2 Upgrade License Emeroency Response Facilities NPC Position Each operating nuclear power plant shall maintain an onsite Technical Support Center (TSC) separate from and in close proximity to the control room that has the capability to display and transmit plant status to those individuals who are knowledgeable of and responsible for engineering and management support of reactor operations in the event of an accident. The center shall be habitable to the same degree as the control room for postulated accident conditions. The licensee shall revise his emergency plans as necessary to incorporate the role and locations of the TSC. Records that pertain to the asbuilt conditions and layout of structures, systems, and components shall be readily available to personnel in the TSC. ('J_]

'm An Operational Support Center (OSC) shall be established separate from the control room and other emergency response facilities as a place where operations support personnel can assemble and report in an emercency situation to receive instructions from the operating staff. Communications shall be provided between the OSC , TSC , EOF, and control room.

An Emergency Operations Facility (EOF) (Near-Site) will be operated by the licensee for continued evaluation and coordination of all licensee activities related to an emergency having or potentially having environmental consequences. The EOF shall be located within 20 miles of the TSC to permit periodic face-to-face communication between management personnel in the TSC and the 20F. The EOF structure shall be well engineered for the design life of the plant. If the EOF is located within 20 miles of the TSC it shall have an isolatable ventilation system with HEPA filters and a backup EOF shall be located within from 10 to 20 miles of the TSC. If the EOF is located between 10 and 20 miles of the TSC, no isolatable ventilation system or backup EOF is required. The facility will have sufficient space to accomodate representatives from Federal, State and local governments as appropriate. In addition, the major State and local response agencies may provide for data analysis jointly with the operator at this location. The EOF will provide f-s (- information needed by Federal, State, and local authorities for implementation of offsite emercency plans in addition to a centralized meetinc location for key representatives from the acencies. Recovery operations shall be managed from this facility. Press facilities also may be available at the EOF.

SHOREHAM DSAR ( O( h LILCO Position The ' Permanent Emergency Response Facilities Design Criteria and Description" applicable to this DSAR cats be found in the Defueled Emergency Preparedness Plan which is being submitted separately via SNRC-1651. The facilities changes principally involve: , 14 The elimination of the EOF.

2. Moving the ENC to the Corporate Information Department in :

Hicksville.

3. The following information regarding TSC habitability:

As presently designed, the SNPS TSC meets the habitability criteria of GDC 19, for all credible design basis accidents (DBAs) envisioned with the fuel in the pool. The DBA on which this conclusion is based is the fuel handling accident (FRA), as described in detail in DSAR Section 15.1.36. Because the l accident releases are so low, it is (conservatively) assumed that ' the TSC's HVAC system is not isolated (i.e., 7000 cfm of unfiltered intake and exhaust continues throughout the accident). > A conservative, ground-level X/0 to the TSC intake is assumed, 7.86E-04 seconds per cubic meter. (~N (

   )   Whole body gamma and beta doses are due to Kr-85.             The gamma 1         doses are computed based on a finite cloud model in the TSC, plus a semi-infinite cloud surrounding the building, which has 18 inches of concrete shielding all around. The beta doses are based on the semi-infinite cloud model suggested by the NRC,core        Reg. )

Guide 1.3. The only radiciodine determined to be in SNPS' is I-129, with an inventory of approximately 4 mil 11 curies. Thyroid doses are computed using a conversion factor for I-129 derived in a fashion consistent with Reg. Guide 1.109 rev. 1, and a breathing rate of 3.47E-04 cubic meters per second (1.25 cubic meters per hour). The resulting doses, and the associated GDC 19 Criteria, are as follows: Dose, rem Whole Body l Gamma Beta Thyroid Results 5.02E-08 1.04E-04 1.21E-07 25 300 300 GDC 19 Criteria III.A.2 -Improvino Licensee Emergency Preparedness --Lono-Term 7-NPC Position ' f Each nuclear facility shall upgrade its emergency plans to provide reasonable assurance that adequate protective measures i

l l l SHOREEAM DSAR l , (} v can and will be taken in the event of a radiological emergency. Specific criteria to meet this requirement

                                       " Criteria         are delineated for Preparation and                     in NUREG-0654 (FERA-REP-1) .

Evaluation of Radiological Emergency Response Plans and Preparation in Support of Nuclear Power Plants'. In accordance with Task Action Plan item III.A.I.1, ' Upgrade Emergency Preparedness," each nuclear power facility was required to immediately upgrade its emergency plans with criteria provided October 10, 1979, as revised by NUREG-0G54 (FEHA-REP-1, issued for interim use and comment, January 1980) . New plans were submitted by January 1, 1980, using the October 10, 1979 criteria. Reviews were started on the upgraded plans using NUREG-06 5 4. Concomitant to these actions, amendments, were developed to 10CFR part 50 and Appendix E to 10CFR These Part 50, newto provide the long-term implementation requirements. rules were issued in the Federal Register on August 19, 1980, with an effective date of November 3, 1980. The revised at nuclear rules delineate requirements for emergency preparednest reactor facilities. NUREG-0 6 5 4 (FEMA-REP-1), ' Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants," provides detailed items to be 7s ) included in the upgraded emergency plans and, along with the revised rules, provides the meteorological criteria, means for

'w/

providing for a prompt notification to the population, and the need for emergency response facilities (see Item III.A.1.2). Implementation of the new rules levied the requirement for the licensee to provide procedures implementing the upgraded emergency plans to the NRC for review. Publication of Revision 1 to NUREG-0654 (FEMA-REP-1), which incorporates the many public comments received is expected in October 1980. This is the document that will be used by NRC and FEMA in their evaluation of emergency pla- submitted in accordance with the new NRC rules. NUREG-0654, Revision 1; NUREG-0696, "Punctional Criteria for Emergency Response Facilities;" and the amendments to 10CFR Part 50 and Appendix E to 10CFR Part 50 regarding emergency preparedness, provide more detailed criteria for emergency plans, design, and functional criteria for emergency response facilities and establish firm dates for submission of upgraded emergency These plans for installation of prompt notification systems. revised criteria and rules for the upgrading of emergency preparedness at nuclear power facilities. LILCO Position

/s

(,) Refer to the emergency plan for the Shoreham Site which is being submitted as a separate document entitled, "Defueled Emergency Preparedness Plan", via letter SNRC-1651. The information contained in this document supersedes in its entirety the information originally submitted as part of the USAF.

SHOREHAM DSAR III.D.l.1 Primary Coolant Sources outside the Containment Structure NRC Position Applicants shall implement a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as-

       ' low-as-practical levels. This program shall' include the following:

Immediate leak reduction (a) Implement all practical leak reduction measures for all systems that could carry radioactive fluid outside of containment. (b) Measure actual leakage rates with system in operation and report them to the NRC. Continuing Leak Reduction -- Establish and implement a program of preventive maintenance to reduce leakage to as-low-as-practical levels. This program shall include periodic integrated leak tests at intervals not to exceed each refueling cycle. ' Applicants shall provide a summary description, together with l initial leak-test results, of their program to reduce leakage from systems outside containment that would or could contain primary coolant or other highly radioactive fluids or gases during or following a serious transient or accident. Systems that should be leak tested are as follows (any other plans system which has similar functions or postaccident characteristics even though not specified herein, should be l included): l Residual heat removal (RHR) Containment spray recirculation High-pressure injection recirculation Containment and prtmary coolant sampling Reactor core isolation cooling Makeup and letdown (PWR's only) l Waste gas (includes headers and cover gas system outside of containment in addition to decay or storage system) Include a list of systems containing radioactive materials which are excluded from program and provide justification for exclusion.

S!!ORERAM DSAR l C'i V Testing of gaseous systems should include helium leak detection or equivalent testing methods. Should consider program to reduce leakage potential release paths due to design and operator deficiencies as discussed in our letter to all operating nuclear power plants regarding North Anna and related incidents, dated October 17, 1979. LILCO Position The purpose of this program is to minimize leakage of Since primary in the coolant sources outside of the primary containment. FIPS condition there is no ' primary coolant" per se, the leakage prevention program becomes irrelevant and annecessary. III.D.3.3 In-plant Radiation Monitoring NRC Position Each licensee shall provide equipment and associated training and procedures for accurately determining the airborne iodine concentration in areas within the facility where plant personnel may be present during an accident.

Effective monitoring or increasing iodine levels in the buildings f~

  \- /    under accident conditions must include the use of portable instruments using sample media                       that will collect iodine silver zeolite) for the following selectively over xenon (e.g.,

reasons:

a. The physical size of the auxiliary and/or fuel handling buildino precludes locating stationery monitoring instrunentation at all areas where airborne iodine concentration data might be required.

b. Unanticipated isolated " hot spots" may occur in locations where no stationary monitoring instrumentation is located,

c. Unexpectedly high background radiation levels near stationary monitoring instrumentation after an accident may interfere with filter radiation readings.

d. The time required to retrieve samples after an accident may result in high personnel exposures if these filters are located in high-dose-rate areas.

After January 1, 1981, each applicant and licensee shall have the capability to remove the sampling cartridge to aNormally, low-background, counting low-contamination area for further analysis. rooms in auriliary buildings will not have sufficientlyInlow the low s_ backgrounds for such analvses following an accident. background area, the sample should first be purged of any entrapped noble gases using nitrogen gas or clean air free of I l

f [ r SHOREHAM DSAR

noble gases.- The licensee shall have the_ capability to measure ! accurately the iodine concentrations present on these samples under accident conditions. There should be sufficient samplers 1 to sample all vital areas. ' L LILCO Position i Revise to state "The source term calculations for spent fuel ; (discussed in DSAR Section 12.2.1) indicate only a very small ; amount of iodine is present in the fuel, about 4.0 mil 11 curies of I-129 (total core). There is no measurable quantity of radiciodine elsewhere in the Reactor, Radwaste, or Turbine Buildings. As such, inplant measurement of radiciodine during and after an accident is unnecessary, and no provisions are made l to perform such analyses." 1

                               ..                                                                   t
      . III.D.3.4         Control Room Habitability NRC Position                                                                            ;

r In accordance control with the Task room habitability, Action Plcn licensees shallitem III.D that assure 3.4 and control l g room operators will be adequately protected against the effects 4 of accidental release of toxic and radioactive gases and that the ! nuclear power plant can be safely operated or shut" down ControlunderRoom," l design basis accident conditions (Criterion 19, j of Appendix A, " General Design Criteria for Nuclear Power , Plants", to 10 CFR Part 50). ! All licensees must make submittal to the NRC regardless of whether or not they meet the criteria of the referenced Standard

Review Plans (SRP) sections. The new clarification specifies that licensees that meet the criteria of the SRP's should provide the basis for their conclusion that SRP 6.4 requirements are met. .

j Licensees may establish this basis by referencing past submittals ' to the NRC and/or providing new or additional information to supplement past submittals. j [ All licensees with control rooms that meet the criteria of the - ; following sections of the Standard Review Plan 2.2.1.2.2.2 Identification of Potential Hazards in Site Vicinity, 2.2.3 : i Evaluation of Potential Accidents, and 6.4 Habitability Systems, shall report their findings regarding the specific SRP Sections as explained below. The following documnets should be used for guidance: ! l7

1. Regulatory Guide 1.78, " Assumptions for evaluating the Habitability of Regulatory Power Plant Control RoomJune (s During a Postulated Hazardous Chemical Release",

1974.

2. Regulatory Guide 1.95, " Protection of Nuclear Power Plant Control Room Operators Against an Accident Cholorine Release"; and,

SHOREHAM DSAR r"N

   'b i

3. K. G. Murphy and E. M. Campe, " Nuclear Power Plant Control Room Ventilation System Design for Meeting General Design Criterion 19, *13th AEC Air Cleaning Conference, August 1974.

4. NUREG- 0570, " Toxic Vapor concentrations in the Control Room Following a Postualted Accidental Maledce.

In providing the ba' sis for the habitabii.ay finding, Licensees shouldlicensees however : may reference their past submittals. ensure that submittals reflect All the curren that the information requested in Attachment.1 is provided. ' licensees with control rooms that do not meet the criteria of the above-listed references, Standard Paview Plans, Regulatory Guides, and other references shall perform the necessary evaluations and identify appropriate modifications. Each licensee submittal shall include the results of the analyses of control room concentrations from postulated accidental release of toxic gases and control room operator radiation exposures from airborne radioactive ,aterial The. and direct radiation resulting from toxic gas accident analysis should . design-basis accidents. () be performed for all potential hazardous chemical releasesoccu r

   \/     boundary. Regulatory Guide 1.78 lists the chemicals ,ost commonly encountered in the evaluation of control room habitability but is not all inclusive.

The design-basis-accident (DBA) radiation source term leakage should be and (LOCA) containment for the loss-of-coolant accidentleakage (ESP) contribution outside , engineered safety featureas described in Appendix A and Bof Standard Review containment Plan Chapter 15.6.5. In addition, boiling-water reactor (BWR) facility evaluations should add any leakage from the main steam (i.e., valve-stem leakage, valve seat isolation valves (MSIV) leakage, main steam isolation valve leakage control systemto the co release) LOCA. This should not be construed as altering the staff (Rev. 2) recommendations in Section D of RegulatoryOther Guide 1.96 DBAS should be regarding MSIV leakage-control systems. reviewed to determine whether they might constitute a more-severe control-room hazard than the LOCA. . In addition to the accident-analysis results, wjich should either . identify the possible need for control -room modifications or provide assurance that habitability systems will operate under all postulated conditions to permit the control-room operators to remain in the control room to take appropriate actions required s'- s by General Design Criterion 19, the licensee should su the adequacy of the habitability systems.information that should be prov evaluation. L_

i i SHOREHAM DSAR I r-) s ATTACHMENT 1 INFORMATION REQUIRED FOR CONTROL-ROOM HABITABILITY LVALUATION Control-room mode of operation, i.e., pressurization and (1) filter recirculation for radiological accident isolation or chlorine release

          ~(2)    Control-room characteristics                    .

(a) air volume control room (b) control-room emergency zone (control room, critical files, kitchen, washroom, computer room, etc.) (c) control-room ventilation system schematic with normal and emergency air-flow rates (d) infiltration leakage rate (e) high efficiency particulate air (HEPA) filter and charcoal adsorber efficiencies (f) closest distance between containment and air intake (g) layout of control room, air intakes, containment building, and chlorine, or other chemical storage facility with dimensions (h) control-room shielding including radiation streaming from penetrations, doors, ducts, stairways, etc. (1) automatic isolation capability-damper closing time, damper leakage and area (j) chlorine detectors or toxic gas (local or remote) (k) self-contained breathing apparatus availability (number) (1) bottled air supply (hours supply) (m) emergency food and potable water supply (how many days and how many people) (n) control-room personnel capacity (normal and emergency) (o) potassium iodide drug supply (3) Onsite storage of chlorine and other hazardous chemicals () (a) (b) total amount and size of container closest distance from control-room air intake l

1 l SHOREHAM DSAR rs i

(4) Cffsite manufacturing, storage, or transportation facilities, of hazardous chemicals (a) identify facilities within a 5-mile radius (b) distance from control room (c) quantity of hazardous chemicals in one container (d) frequency of hazardous chemical transportation traffic (truck, rail, and barge) (5) Technical specifications (refer to standard technical specifications) (a) chlorine detection system (b) control-room emerg ency filtration system including the capability to maintain the control-room pressurization at 1/8 in, water gauge, verification of isolation by test signals and damper cicsure times, and filter testing requirements.

    /~3     LILCO Position O        Habitability Systems Final Decision Design Bases The original plant design bases of the control room's habitability systems, as described in USAR Section III.D 3.4, still apply in general. However, due to the small quantity of radioactivity released during the design basir accident (the fuel handling accident) , the control room's remote intakes and standby charcoal filtration system are no longer required to meet General Design Criteria 19.      Doses, assuming the control room's HVAC system continues to function as during normal operation, are indicated in Chapter 15 of the DSAR.

System Desien The design of the control room HVAC system is as described in DSAR Section 9.4.1. As stated above, the remote intakes and the Inlet duct standby filtration system are no longer required.As such, instrumentation is no longer required as well. discussion of these items in USAR Section III.D.3.4 no longer applies. During the design basis accident, the control room EVAC system will continue to function as during normal plant j operations.

            Desian Evaluation This section is as described in USAE Section III.D.3.4, except                                       l that the remote intakes and standby filtration systems are no

I I l SHOREHAM DSAR lO longer required. Also, as per DSAR Section 9.2.9, only one chilled water system is required with the spent fuel in the pool. Tests and Inspections Tests and inspections of the control room HVAC system

                                                                                                        ' hat     are as the standby described in USAR Section III.D.3.4, except
   ' filtration systems are no longer required.                                                       ,

Standby Charcoal Filtration Trains This equipment is no longer required. O O

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ML20082M508

Text

1.1 INTRODUCTION

SUMMARY

7.1 INTRODUCTION

8.1 INTRODUCTION

SUMMARY

11.0 REFERENCES

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