ML20005F251

Shoreham Nuclear Power Station Defueled Sar.

ML20005F251
Person / Time
Site:	Shoreham File:Long Island Lighting Company icon.png
Issue date:	01/05/1990
From:	LONG ISLAND LIGHTING CO.
To:
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ML20005F164	List: ML20005F163 ML20005F246 ML20005F248 ML20005F251
References
NUDOCS 9001160092
Download: ML20005F251 (230)

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ATTACHMENT 3 TO SNRC-1664 L

The Shoreham Nuclear Power Station  !

Defueled Safety Analysis Report O  ;

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1 10-P Obo 22 PDC

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CHOREHAM DSAR TABLE CF CONTENTS

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Chapter /  ;

E Section Title Page l 1 INTRODUCTION AND GENERAL DESCRIPTION OF 1' PLANT i 1. 1 Introduction 1- 1

1. 2 General Plant Description 1- 4

1. 3 Comparison Tables 1- 5

1. o4 Identification of Agents and contractors 1- 5 l 1. 5 Requirements for Further Technical 1- 5 I

Information

1. 6 Material Incorporated by Reference 1- 5

1. 7 Symbols Used in Engineering Drawings 1- 6 2

e 2 SITS' CHARACTERISTICS  :

2. 1 Geography and Demography 2- 1

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

2. 3 Meteorology 2- 1

2. 4 Hydrologic Engineering 2- 2 i

2. 5 Geology & Seismology 2- 2 2A Boring Logs 2  :*

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2B ' Seismicity Investigations 2- 3 l 4 2C A Reevaluation of the Intensity of the E. 2- 3 '

Haddam, Conn. Earthquake of-May 16, 1791 l 2D Reevaluation of the Reported Earthquake at 2- 3 l- 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 Which Have Affected the Site 2- 3 r With Modified Mercalli Intensity >= IV 1HE Report on Seismic Survey-Proposed Shoreham 2- 3 Power Station LILCO 2I Laboratory Soils Test 2- 4 2J Summary Report of Geotechnical Studies of 2- 4 Reactor Building Foundation 2K Aircraft Crash Probability Study 2- 4 2L Report on Service Water System Soils 2- 4 2M Report on Densification of Service Water 2- 4 System Soils 2N Hurricane Study 2- 4 3

3- DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS

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

() 3. 2 Classification of Structures, Systems and Components Wind and Tornado Loading 3-12 3-14

3. 3

3. 4 Water Level (Flood) Design 3-14

3. 5 Missile Protection 3-14

3. 6 Protectitn Again:t Dynamic EtfOct3 3 A020citt d with th3 Postulct d Ruptura of Piping

3. 7 Seismic Design 3-14

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3. s. Design of seismic Category I Structures 3-14  :

E '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 Equipr.ent L 3A Computer Programs for the Stress Analysis 3-16 of Cat I Structures and Piping Systems 3B NRC Regulatory Guides 3-16 j 3C Pipe Failure outside Primary Containment 3-16 4

l: 4 REACTOR ,

4. 1 Reactor Summary Description 4- 1 '

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4. 1. 1 Reactor Vessel 4- 1

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

4. 1. 3 Reactivity Control System 4- 2

4. 1. 4 Analysis Techniques 4- 2

4. 4 Thermal and Hydraulic Design 4- 2

4. 5 Reactor Materials 4- 2

4. 6 Control Rod Drive Housing Supports 4- 2 5 '

5 REACTOR COOIANT SYSTEM

5. 5. 7 Residual Heat Removal System 5- 1 6

6 ENGINEERED SAFETY FEATURES L 3 -

6. 1 General 6- 1 l' 6. 2 Containment Systems 6- 1

6. 2. 1- Containment Functional Design 6- 1

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

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

6. 2. 4 Containment Isolation System 6- 2

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

6. 3 Emergency Core Cooling systems 6- 2

6. 4 Habitability Systems 6- 3

6. 5 Main Steam Isolation Valve Leakage control 6- 3 System i

6. 6 Overpressurization Protection 6- 3

6. 7 Main Steam Line Isolation Valves 6- 3

6. 8 Control Rod Drive Support System 6- 3 )

6. 9 Control Rod Velocity Limiter 6- 3 6.10- Main Steam Line Flow Restrictor 6- 3 6.11 Reactor Core Isolation Cooling System 6- 3 6.12 Standby Liquid Control System 6- 4 7

7. INSTRUMENTATION AND CONTROLS

7. 1 Introduction 7- 1

7. 1. 1 Identification and Classification of 7- 1 Nonsafety Related Systems

7. 1. 2 Identification of Safety Design Bases and 7- 6 O- Nonsafety Design Bases Criteria 7- 6

7. 2 Reactor Protection System

7. 3 Engineered. Safety Feature System 7- 6 l

7. 4 Systems Required For Safe Shutdown 7- 7 l

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7.f5 Saf;ty Relcted Dicylcy In;trumentGtien 7- 7

'7. 6- All oth;r In3trumentaticn System 3 R: quired 7- 7 fcr Ocf0ty.

7. 6. 1 Description 7- 7 O'

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

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78 Analog Transmitter / Trip System for ESF- 7- 9 Sensor Trip Units S-S ELECTRIC POWER S. 1 Introduction 8- 1 S. 1. 1 -Utility Grid 8- 1 S. l. 2 Interconnection to Other Grids 8- 1 S. 1. 3 Offsite Power System 8- 1 S. 1. 4' On Site AC Power System 8- 2

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

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

8. 1. 7 Identification of Safety Criteria 8- 2 S. 2 offsite Power System 8- 3

8. 2. 1 Description 8- 3

8. 2. 2 Analysis 8- 4

8. 3 On Site Power system 8- 4

8. 3. 1 AC Power System 8- 4

8. 3. 2 DC Power System 8- 4 9

9 AUXILIARY SYSTEMS

9. 1 Fuel Storage and Handling 9- 1 l

9. 1, 1 New Fuel Storage 9- 1

9. 1. 2 Spent Fuel Storage 9- 1

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

9. 1. 4 Fuel Handling System 9- 2 dh

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9. 2 Water Systems 9- 3

9. 2. 1 Service Water System 9- 3

9. 2. 2 Reactor Building Closed Loop Cooling Water 9- 4 (RBCLCW) System 9.-2. 3 Makeup Water Domineralizer System 9- 4

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

9. 2. 5 Ultimate Heat Sink 9- 5

9. 2. 6 Condensate Storage Facilities 9- 5

9. 2. 7 Turbine Building Closed Loop Cooling Water 9- 6 System

9. 2. 8 Main Chilled Water System 9- 6

9. 2. 9 Reactor Bldg Standby Vent Sys and Control Rm 9- 6 A/C Chilled Wtr System l 9. 3 Process Auxiliaries 9- 6

9. 3. 1 Compressed Air Systems 9- 6

9. 3. 2 Process Sampling System 9- 7

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

9. 3. 4 Chemical, Volume Control and Liquid Poison 9- 7 Systems

9. 3. 5 Failed Fuel Detection System 9- 8

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

9. 4 Air Conditioning, Heating, Cooling, and 9- 8 Ventilation Systems 9.4.1 Control Room Air Conditioning system 9- 8

,/"'g 9. 4. 2 Reactor Building Normal Ventilation System 9- 8 U 9. 4. 3 Radwaste Building Ventilation 9- 9

9. 4. 4 Turbine Bldg Ventilation System and Station 9- 9 Exhaust System 9.4.5 Battery Room Heating and Ventilation 9- 9

9. 4.:6- Drywall Air C Oling Syst03 99

9. 4. 7 Ocreenwall Pump Hsu30 H Sting Cnd 9- 9 Ventilation m .

9. 4. 8 Plant Heating 9- 9 s '

9. 4. 9' Primary Containment Purge System 9-10 L 9. 4.10 Diesel Generator Room Ventilation 9-10 L 9. 4.11 Relay Room, Emerg. Switchgear Room & 9-10 L

Computer Room' Air Cond. Systes *

9. 5 Other Auxiliary Systems 9-10

9. 5. 1 Fire Protection System 9-10 g

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9. 5. 2 Communication Systems 9-14 ,

L 9. 5. 3 Lightfag Systems 9-15 L 9. 5. 4 Diesel Generator Puol Oil Storage and 9-15 Transfer System ,

o 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 9A Pue4 Criticality Analysis 9-17 9B Evaluation of Spent Fuel Pool Makeup 9-18 Requirements 10 10 STEAM AND POWER CONVERSION SYSTEM

10. 1 Steam and Power Conversion System 10- 1

10. 2 Turbine Generator 10- 1 l 10. 3 Main Steam Supply System 10- 1 i 10. 4 Other Features of Steam and Power Conversion 10- 1

,Q System U 10. 4. 1 Condenser 10- 1

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

• 10.H4. 3 Steam Seal System 10- 1 .

10. 4. 4 Turbine Bypass System 10- 2 l 10. 4. 5 Circulating Water System 10- 2

10. 4. 6 Condensate Domineralizer System 10- 2

10. 4. 7 Condensate and Feedwater System 10- 2 11 11 RADIOACTIVE WASTE MANAGEMENT -

11. 1 Radiation Source Terms 11- 1

11. 2 Radioactive Liquid Waste System 11- 2

11. 2. 1 Design Objectives 11- 2

11. 2. 2 System Descriptions 11- 2 L

11. 2. 3 System Design 11- 3

11. 2. 4 operating Procedures 11- 5

11. 2. 5 Performance Tests 11- 6

11. 2. 6 Estimated Releans 11- 6

11. 2. 7 Release Points 11- 6

! 11. 2. 8 Dilution Factors 11- 6 l 11. 2. 9 Estimated Doses 11- 6

11. 3 Gaseous Waste System 11- 7 l 11. 3. 1 Design Objectives 11- 7 L 11. 3. 2 System Descriptions 11- 7 l

11. 3. 3 System Design 11- 7

-Operating Procedures

11. 3. 4 11- 7

! 11. 3. 5 Performance Tests 11- 7 l 11. 3. 6 Estimated Releases 11- 7

11. 3. 7 Release Points 11- 8

11. 3. 8 Dispersion Factors 11- 8

11. 3. 9 Estimated Doses 11- 8 1

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11. 3o10 11- 8 ilo 4 Proceso & EffluOnt Rodiotien Manitcring 11- 8

. System

11. 5 Solid Waste System 11- 8

. 11. 5. 1 Design Objectives 11- 8 11, 5. 2 System Input: Source Terms 11- 9

11. 5. 3 Equipment Description 11- 9

11. 5. 4 Expected Volumes 11 11.-5. 5 Packaging 11-11

11. 5.-6 Storage 11-11'

11. 5. 7- Shipment 11-11

11. 6 Offsite Radiological Environmental 11-11 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 11-21

11. 6. 5 Data Analysis, Presentation, and 11-21 Interpretation-

11. 6. 6 Program Statistical Sensitivity ,

11-21 12 ,

12 RADIATION PROTECTION

12. 1 Assuring that occupational Radiation 12- 1 Exposures are AIARA

12. 2 Radiation Sources 12- 2

12. 2. 1 Contained Sources 12- 2.

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

12. 3 Radiation Protection Design Features 12 4'

12. 3. 1 Facility Design Features 12- 4

12. 3. 2 shielding 12- 4

12. 3. 3 Ventilation 12- 5

12. 3. 4 Radiation Monitoring Instrumentation 12-'5

12. 4 Dose Assessment 12- 6

12. <4. 1 Design Objectives 12- 6

. 12. 4. 2 Airborne Activity 12- 7

12. 4. 3 Occupational Dose Assessment 12- 7

12. 4. 4 Offsite Dose Assessment 12- 8

12. 5 Health Physics Program 12- 8 13 13 CONDUCT OF OPERATIONS

13. 1 Organizational Structure of Applicant 13- 1

13. 1. 1 Corporate Organization 13- 1

13. 1. 2 Engineering and Administrative Support 13- 1 Organization

13. 1. 3 Operating Organization 13- 2

13. 1. 4 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 13, 5 Station Procedures 13-11

13. 5. 1 Administrative Control 13-11

13. 5. 2 Procedures 13-12 ,

-13. 6 Plant Racords 13-14

13. 7 Industrial Security 13-14

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14 14 INITIAL TESTS AND OPERATIONS 15' 15 ACCIDENT ANALYSIS '

O. 15. 1

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

15. 1. 3 Turbine Trip 15- 2

15. 1. 3 Turbine Trip and Failure of Generator 15- 2

. Breakers to Open

15. 1. 4 Main Steam Isolation Valve closure- 15- 3

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

15. l'. 6 Pressure Regulatory Failure - Closed 15- 3

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

15. 1. 8 Loss of Feedwater Heating 15- 2

15. 1. 9 ' Shutdown Cooling (RHR) 15- 2 Malfunction-Decreasing Temperature

15. 1.10 Inadvertent HPCI Pump Start 15- 4

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

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 Refueling

15. 1.15 Off-Design Oper Transient Due to Inadvertent 15- 4 Loading of a Fuel Assembly

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

15. 1.17 Inadvertent Opening of a Safety Relief Valva 15- 4

/~ 15. 1.18 Loss of Feedwater Flow 15- 3

15. 1.19 Loss of AC Power 15- 3

15. 1.20 Recirculation Pump Trip 15- 3

15. 1.21 Loss of Condenser Vacuum 15- 3

15. 1.22' Recirculation Pump Seizure 15- 3

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

15. 1.24 Recirculation Flow Control Failure With 15- 4 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 Transiant Without Scram (ATWS) 15- 5

15. 1.28 Cask Drop Accident 15- 5

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 l 15. 1.32 Liquid Radwaste Tank Rupture 15- 5

! 15. 1.33 Control Rod Drop Accident 15- 4 L 15. 1.34 Pipe Breaks Inside the Primary Containment 15- 5 (Loss-of-Coolant Accident)

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

15. 1.36 Fuel Handling Accident 15- 7 Worst Case Fuel Damage Event 15-10

() 15. 1.36A

15. 1.37 15, 1.38 Feedwater System Piping Break Failure of Air Ejector Lines 15- 4 15- 7 16 16 16 TECHNICAL SPECIFICATIONS 1

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17- QUALITY ASSURANCE

17. 1 Quality A00uranc3 During Design Cnd 17- 1 Construction L

17. 2 Quality Assurance During the operational 17- 1 Phase

17. 2. 1 Organization 17- 2 (

17. 2. 2 . Quality Assurance' Program 17- 3

17. 2. 3 -Design control 17- 3

17. 2. 4 Procurement Docurent Control. 17- 3 1

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

17. 2. 6 Document Control 17- 3

17. 2. 7 Control of Purchased Material, Parts, and 17- 3 services

17. 2. 8 Identification and Control of Special 17- 3 Processes

17. 2. 9 . Control of Special Processes 17- 3

17. 2.10 Inspection 17- 3

17. 2.11 Test Control 17- 3

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

17. 2.16 Corrective Action 17- 4

17. 2.17 Quality Assurance Records 17- 4

17. 2.18 Audits 17- 4 1 O

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SHOREHAM DSAR  ;

l CHAPTER 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT f

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.

This report is intended to provide sufficient information to enable the NRC Staff to issue the license amendment as requested

, in SNRC-1650.

l The purpose of-the DSAR is to provide a safety analysis for the E

storage and handling of Shoreham low burnup first cycle spent fuel. The DSAR confirms that fuel storage and handling systems, ,

structures, components and programs ensure that there is no undue 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 secondary containment, the Shoreham reactor building. The structures are designed to withstand seismic loads.

The Shoreham spent fuel is in a low burnup condition. The Shoreham Nuclear Power Station operated during low power testing at power-levels not exceeding 5% of rated power. The effective burnup of the fuel is approximately 2 full power days. This results in an estimated total core wide heat generation rate of I approximately 550 watts as of June 1989. The estimated fuel heat load will reduce to approximately 250 watts by June'1991. Figure

\ I-1 (taken from DSAR Section 15.1) depicts the fuel heat load g versus time. Based on this low heat generation rate, systems for ,

K active cooling are not required, and only minimal capacity systems are required for pool water makeup to handle evaporation. ,

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L e-BROREHAM D8AR The Shoreham spent fuel contains limited quantities of l radioactive materials that are available for release.: As is i stated in DSAR Section 32.2, approximately 176,000 curies of l radioactivity reside in the 560 fuel assemblies. Gaseous activity in the fuel assemblies is primarily Krypton-85 (a noble i 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, j 1987). Other sources of radioactivity outside the core are minor, and include small amounts of contamination in the bottom of sumps, the suppression pool, inside the reactor pressure vessel, and in the radwaste systems.

Chapter 15 presents radiological analyses for those 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 l'

Chapter 15. The events analyzed in Chapter 15 are:

1. Fuel Handling Accident (Puel Bundle Drop)

2. - Radwaste Tank Rupture The only design basis accident involving reactor fuel is a Fuel

-( ) Bandling Accident, in which no heat generation takes place. As such, the activity available for release in this design basis accident is primarily Krypton-85, and consists of approximately l 2.5 curies. In addition, a worst case radiological event is postulated in which the entire gaseous activity of the core is H released to the reactor building. This event was postulated to conservatively bound any possible situation involving large-scale J

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-in comparison with 10CFR100 limits. For the worst case scenario in which all the gaseous activity is assumed to be released from the entire core, a spectrum of cases were analyzed as follows: operation of the standby ventilation system, operation of the normal ventilation system, and no o

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 1.1-2. In particular, it 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 pool provides a high degree of passive safety protection for L

Shoreham spent fuel. Active safety systems are not required to 1-2 L _ ..

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n SHOREHAM DSAR 1 mitigate postulated accidentar however, support systems are-required to meet the intent of 10CFR50 Appendix A, General Design 1

Criteria ((see Chapter 3 for a-listing) and Regulatory Guide 1.13.

o Supporting systems are required to provide for radiation- J L monitoring, fuel pool' makeup, fuel pool cleanup, radwaste

-management, and normal building services. Therefore a reclassification of safety systems is proposed based on the l importance to safety associated with each plant system with the

> plant defueled.

The DSAR' assumes that the Shoreham spent fuel from the initial core is to be stored for some-interim period in the spent fuel-poo1~ contained within the SNPS reactor building.

I The assumed configuration of principal plant systems is.as follows:

L 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 l

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

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water level is maintained at its normal water level. Makeup will be' furnished from the condensate transfer system or the

! domineralized and makeup water system. The. fuel pool cooling l~ 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 l

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 L

Cooling, and Primary Containment systems are not required and are-in a protected state. This is discussed in DSAR Chapters 4, 5 and 6.

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

(: in Chapter 8, blackstart combustion turbines exist nearby in l

the Shoreham west site to supply 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 Technic 1 Specifications to remain operable during fuel movement.

( 6. Secondary containment integrity will be maintained utilizing the Normal Ventilation System to provide a controlled and monitored release capability as discussed in Chapter 15 Safety Analysis.

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" I SHOREHAM DSAR

7. The steam and power conversion systems are not required to be operable or' functional and are protected from significant degradation as described in.DSAR Chapter 10.

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

10. Major systems that remain functional to provide non-safety related supporting services include a) Service Water-(DSAR Chapter 9 and 10) b) Chilled Water Systems (DSAR Chapter 9) c) 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 l B) including the basis of the specification is provided.  ;

! (DSAR Chapter 16) l

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 DEAR Chapter 17. A new Quality Assurance Category IIA is defined in DSAR Chapter 3 for systems, structures, and L components that no longer fulfill a safety function in l support of a defueled reactor.

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

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

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

i L 7. A Defueled Emergency Plan is being submitted separately for NRC review and approval via SNRC-1651.

l 1.2 GENERAL PLANT DESCRIPTION The descriptions and design criteria contained under this heading in the latest revision of the Shoreham USAR remain unchanged.

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id 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.

LAll other systems will-be_either functional or protected from degradation.

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 preservation concerns.

3. Protected - Those systems not to be operated in the defueled mode. These systems will be left in a deenergized safe state and layed-up in accordance with System Lay-up Implementation Package (SLIPS), which specify maintenance and custodial services necessary to protect them pending disposition of LILCO's operating. license.

1.3 COMPARISON TABLES The description contained under this heading in the latest revision of the-Shoreham USAR remains unchanged.

Refer to the USAR for'information on this subject.

' 2.4 IDENTIFICATION OF AGENTS AND CONTRACTORS 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.

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

The systems described in this section are not required for the defueled condition.

1.6 MATERIAL INCORPORATED BY REFERENCE The information contained under this heading in the latest f .

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

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? i.r7 .SHOREHAM DSAR  !

l; -: mJ I- l'.7 . SYMBOLS USED IN ENGINEERING DRAWINGS Tho'information= contained'under this heading in the latest -

revision of the-Shoreham USAR remains unchanged.

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Refer'to the j

-USAR-for information on this subject.

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SHORERAM DSAR l

TABLE 1.2-1

. l STATUS-OF PLANT SYSTEMS IN THE DEFUELED CONFIGURATION FOR AN E__XTENDED PERIOD OF TIME l

OPERABLE (RB) Cranes, Hoists and Elevators Reactor Building Superstructure "

i Process Radiation Monitoring Area Radiation Monitoring Servicing Aids (Fuel)-

Refueling l

Radwaste Fire Protection (Mechanical)

Meteorological Monitoring

! Station Transformer (NSS) '

L Non-Segregated Buses

-Metal Clad Switchgear Load Centers and Unit substations I Fire Detect-& Station Security

' /'%

i (Electrical /IEC) 138/69kv Switchyard Pot. Transf.

138kv. Switchyard Relay Panels Reactor Building p Reactor Building Ventilation Reactor Building Standby Ventilation-(shared portion only)

Seismic Monitoring 1

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OPERABLE: System (s) maintained to meet Technical Specifications.

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i.- ~

l O Q Ox '

DSAR. FIGURE 1.1 - 1 SNPS Spent Fuel Decay Heat Load 1200 1080 ,

t l

l l

o a4o se f 720 o

_3 "o

8 1 4so o

o

$ seo 240

= _

June 1989 - -

o  :  :

1 1.5- 2 3 5 7 10 15 Time from Last Burnup years 8

+ _ __ - _ _ _ - _

- - - _ _ - _ _ , - - - -_____-__.__a

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

e e

SHOREHAM DSAR CHAPTER 2 SITE CHARACTERISTICS 2.1- GEOGRAPHY AND DEMOGRAPHY ,

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

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

2.3 METEOROLOGY

'The description contained under this heading in the latest revision of Shoreham USAR remains unchanged except that the 33 ft. tower south of the plant will not be used. Additionally, the following information regarding the Operational Program applies <

Refer to USAR for other information on this subject.

to DSAR.

2.3.3.2 Operational Program .

The operational meteorological monitoring program uses instrumentation to determine wind-speed and -direction at 33- and 150-ft. ambient air temperature at 33-ft and temperature i differential (Temp 9 150-ft minus Temp 9 33-ft). These instruments-are located on SNPS' 400 ft. meteorological tower which is located approximately 5100-ft WSW of the reactor building (Figure 2.1.1.1). The MET tower was positioned ,

sufficiently close to SNPS to provide representative observations of' released gaseous effluents, but far enough away to minimize atmospheric disturbances caused by SNPS' structures.

!. Wind-speed and -direction at the 33-ft level, along with the

' temperature differential are transmitted to the Technical Support Center. ~In addition to these parameters, wind-speed and

-direction at 150-ft., and temperature at 33-ft. are transmitted to the Main Control Room and entered into the RMS computer. .

All 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 influence of solar insolation. A motor and fan draw a constant 2-1

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

i b -

bf SHOREHAM DSAR flow of air at ambient conditions over the sensor to ensure

. accurate-measurements.

1 1: 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 model

l. the dispersion of plant gaseous effluents and are used as input  :

L' to required periodic reports.

To ensure the operability of the system, quarterly calibrations are performed by.a qualified vendor, and channel checks are -,

-performed by the operators on. shift using qualitative assessment

- of the channel's behavior during operation. Operators.do this by i checking the chart recorders in the control room. This L instrumentation includes:

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

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

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

4) Differential air temperature monitor which uses the L

temperature data recorded at 33-ft. and 150-ft. elevations. 3 1

L (') - Meteorological sensors are replaced on a quarterly basis with L 's /- Adentical equipment which have been calibrated in the laboratory of a qualified vendor. Vendor personnel perform the actual sensor substitutions under the direction of LILCO technicians.

LILCO technicians perform the normal monthly maintenance

. procedures on instrumentation at the base of the tower.

Calibration and maintenance procedures have been developed for L field testing and maintenance of each meteorological channel at l

the Shoreham site.

o L Spare sensors and auxiliary equipment are available for rapid  ;

L replacement of any malfunctioning components of the system. In L the event that a meteorological tower is damaged, with one or more monitoring instrumentation channels 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 I revision of Shoreham USAR remains unchanged. Refer to USAR for i information on this subject.

2.5 GEOLOGY AND SEISMOLOGY f The description contained under this heading in the latest

() revision of Shoreham USAR remains unchanged.

'information on this subject.

Refer to USAR for l

l 2-2

,J SHOREHAM DSAR

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

-2B SEISMICITY. INVESTIGATIONS l

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

information on this subject.  !

2C A REEVALUATION OF THE INTENSITY OF THE EAST HADDAM, CONNECTICUT EARTHQUAKE OF MAY- 16, 1971  :

The description contained under this heading in the latest

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

2D REEVALUATION OF THE REPORTED. EARTHQUAKE AT PORT JEFFERSON, LONG ISLAND, NEW YORK q(~N The description contained under this heading in the latest

\_/' revision of Shoreham USAR remains unchanged. Refer to USAR for information on this subject.

2E REEVALUATION OF THE EARTHQUAKE OF OCTOBER 26,.1845

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

2F' REEVALUATION OF THE EARTHQUAKE OF JANUARY 17, 1855 L The description contained under this heading in the latest I revision of Shoreham USAR remains unchanged. Refer to USAR for L

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

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

information on this subject.

Refer to USAR for 2-3 l

.. e

%% P t

[I SHOREHAM DSAR 22 LABORATORY-SOILS TESTS The description contained under this heading in the latest <

Refer to USAR for revision of Shoreham USAR remains unchanged.

information on this subject.

2J

SUMMARY

REPORT OF GEOTECHNICAL STUDIES OF REACTOR BUILDING l FOUNDATION The description contained under this heading in the latest revision of Shoreham USAR remains unchanged. Refer to USAR for information on this subject.

- 2K AIRCRAFT CRASH PROBABILITY STUDY

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

. information on this subject.

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

\ .

information on this subject.

2M REPORT ON DENSIFICATION OF SERVICE WATER SYSTEM SOILS The description contained under this heading in the latest revision of Shoreham USAR remains unchanged. Refer to USAR for information on this subject.

2N IIURRICANE STUDY The~ description contained under this heading in the latest L

revision of Shoreham USAR remains unchanged. Refer to USAR for l

information on this subject.

L i

O l

2-4

x i

SHOREHAM DSAR CHAPTER 3 DESIGN OF STRUCTURES, COMPONENTS, EQUIMENT, AND SYSTEMS 3.1 CONFORMANCE TO GENERAL DESIGN CRITERIA FOR NUCLEAR POWER PLANTS (10CFR Part 50, Appendix A)

The General Design Criteria (GDC), contained in the Shoreham USAR Section 3.1, were reviewed to establish those criteria that may be applicable to the storage of SNPS low burnup cycle spent fuel in the spent fuel pool. The following GDC are addressed:

I. Overall Requirements E GDC1 Quality Standards and Records l GDC2 Design Bases for Protection Against Natural Phenomena j GDC3 Fire Protection GDC4 Environmental and Dynamic Effects Design Bases II. Protection by Multiple Fission Product Barriers GDC13 Instrumentation and Control H 1

GDC17 Electric Power Systems GDC18 Inspection and Testing of Electric Power Systems GDC19 Control Room IV. Fluid Systems _

GDC44 Cooling Water GDC45 Inspection of Cooling Water System GDC46 ' Testing of Cooling Water ~ System l

VI. Fuel and Radioactivity Control I GDC60 Control of releases of radioactive material to the environment J GDC61 Fuel storage and handling and radioactivity control 1 GDC62 Prevention of criticality in fuel storage and  !

handling i GDC63 Monitoring fuel and waste storage GDC64 Monitoring radioactivity releases The following GDC were found not to be applicable to a defueled reactor:

I Overall Requirements GDC5 Sharing of structures, systems, and components shoreham is a single unit, thus the above criterion does not apply.

3-1

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

I I l

SHOREHAM DSAR II Protection By Multiple Fission Product Barriers GDC10 Reactor Design. ,

GDC11 Reactor Inherent Protection GDCl2 Suppression of reactor power oscillations GDC14 Reactor Coolant Pressure Boundary GDC15 Reactor Coolant System ,

GDC16 Containment Design

-The above criteria do not apply because the reactor and primary containment are not operable.

III Protection And Reactivity Control Systems GDC20 - 29 requirements apply only to an operating reactor protection and reactivity control systems IV Fluid Systems L

GDC 30-43 address reactor.and containment systems required for power operation only.

V Reactor Containment

(}

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

. Applicable criterion Conformance Quality Standards and Records (Criterion 1)

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

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

O Design Conformance Structures, systems, and components are classified in Section 3.2. The LILCO QA program described in DSAR Chapter 17 assures 3-2

l y .4 SHOREHAM DSAR l l

l that quality practioes and documentation are maintained l ' commensurate with the classification that is identified in this. l p Defueled Safety Analysis Report (DEAR). A new Q.A. Category IIA l

is provided for USAR safety-related structures, systems, and E camponents that no longer fulfill a safety function for.a defueled reactor.

-Desion Basis for Protection Against Natural Phenomena (Criterion-2)

-Criterion l'

Structures, systems, and components important to safety shall be designed to withstand the effects of natural phenomena such as L 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 l of the natural phenomena that have been historically reported for the site and surrounding area,'with sufficient margin for the I

L limited accuracy, quantity, and period of time in which the l historical. data have been accumulated, .(2) appropriate combinations of the effects of normal and accident conditions (3) the with the effects of the natural phenomena, and importance of,the safety functions to be performed.

Desian Conformance

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

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

L Tire Protection (Criterion 3) i Criterion Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions.

Noncombustible and heat resistant materials shall be used wherever practical throughout the unit, particularly in locations such as the containment and control room. Fire detection and i fighting systems of appropriate capacity and capability shall be provided and designed to minimize the adverse effects of fires on O 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

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

0 ,

SHOREHAM DSAR Design Conformance ,

This~ criterion is satisfied by the SNPS fire protection program which is described in section 9.5.1 of this report and the USAR.

L- ' Environment'al and Missile Design bases (Criterion 4) n Criterion' l- Structures, systems, and components important to safety aball be

. designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss of coolant accidents. ,These structures, systems, and components shall be appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids,'that may result from equipment failures and from events and conditions outside the nuclear power unit.

L Design Conformance l

Chaptar lL5.of this report defines accidents that are applicable to spent fuel storage and fuel handling.. The spent fuel is.

l O.L stored in the spent fuel storage pool. The pool structure, E . Reactor Building, and spent fuel racks provide passive safety l protection from missiles or other conditions that could cause fuel mechanical damage. The structural design basis of the fuel 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 its 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 a

i SHOREHAM DSAR Electric-Power Systems (Criterion 17) criterion l

I' An onsite electric power system and an offsite electric power j system shall be provided to permit functioning of structures, l systems, and components important to safety. The safety function j 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 limits and design conditions of.the reactor coolant pressure boundary are not exceeded as a result of anticipated operational occurrences and ,

(2) the core is cooled and containment integrity and other vital l functions are maintained.in.the-event of postulated accidents. l, The onsite electric power supplies, including the batteries, and the onsite electric distribution system shall have sufficient independence, 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)

-( f 2

designed and located so as to' minimize to the extent practical

.the likelihood of their simultaneous failure under operating and postulated accident and environmental conditions. A switchyard-common to both circuits is acceptable. Each of these circuits shall be designed to be available.in sufficient time following a lose:of.all onsite alternating current power supplies and the

-other offsite electric power circuit, to assure that specified acceptable fuel design limits and design conditions of the reactor' coolant pressure boundary are not exceeded. One of these circuits shall be designed to be available within a few seconds following a loss of coolant accident to assure that core cooling, containment integrity, and other vital safety functions are maintained.

Provisions shall be included to minimize the probability of losing electric power from any of the remaining supplies as a result of, or coincident with, the loss of power generated by the nuclear power unit, the loss of power from the transmission L 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 in the event O of postulated accidents including a seismic event. However, operability of the electric power system will be required by Technical Specifications during fuel movement to provide for a controlled and monitored release capability in the event of a 3-5

~ -

, . =i SHOREHAM DSAR fuel drop eccident.- Two offsite power transmission system will ,

be maintained to provide power for support system operation. In  !

addition, blackstart combustion turbines exist nearby at Shoreham-West to provide reliable power in the unlikely event of a loss-of-offsite power occurs. Emergency Diesel Generators are not required.- A further discussion of electric power requirements can be found in Chapter 15.

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 o 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

-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 L 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 be provided (1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of '

the reactor through the use of suitable procedures.

3-6

p _ J 't l

SHOREHAM DSAR

- Design Conformance h[*

j A~ contre!. Acom is provided and equipped to operate the plant l.

safely under normal and accident conditions.

. s Based on the results of radiological analyses provided in DSAR i

~

Chapter 15 control room shielding and. ventilation functions are not required for the mitigation of postulated accidents. 1 Instrumentation available in the control room for accident.

monitoring and support system control are described in DSAR i Chapter 7.

Cooling Water (Criterion 44)

Criterion -

L 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 L

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

1 L f~s suitable redundancy-in components and features, and suitable

• L -( j interconnections, leak detection, and isolation capabilities shall be provided to assure'that for onsite electric power system t operation (assuming offsite power is not available) and for L

offsite electric power operation (assuming onsite power is not .'

L L

available) the system's safety function can be accomplished, '

assuming a single' failure.

L

- Design Conformance I-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 available to -

provide plant building services; however, these systems do not fulfill a safety function.

L Inspection of Cooling Water System L (Criterion 45)

Criterion .

The cooling water system shall be designed to permit appropriate periodic inspection of important components, such as heat exchangers and piping, to assure the integrity and capability of the system.

Design Conformance L The service water system which will be maintained functional is designed to permit appropriate visual inspection in order to 3-7

I I

l

p. SHOREHAM DSAR k

assure the integrity of system components. See Criterion 44 response.

Testing of Cooling Water System _

Criterion 46?

Criteripa The cooling water system shall be designed to permit appropriate  !

periodic pressure and functional testing to assure (1) the >

st.r'setural and leaktight integrity of its components, (2) the opetebility and performance of the active components of the hystem, and (3) the operability of the system as a whole and, under conditions as close to design as practical, the performance 1 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 eraergency power ,

sources.

Design Conformance See Criterion 44 response.

Cont;rol of Releases of Radioactive _ Materials to the Environment

~

TCrii;erion 60)

Criterio_n The nuclear power unit design shall include means to control suitably the release of radioactive materialg in gaseous and liquid effluents and to handle radioactive solid wastes produced during normal reactor operation, including anticipated operational occurrences. Sufficient holdup capacity shall be provided for retention of gaseous and liquid affluents 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.

Design Conformance Because SNPS is not in normal operation, effluent releases are due primarily to maintenance of the spent fuel pool water quality. Means are provided to control and/or hold up the release of liquid and gaseous effluents as required. Fuel pool clehnup and appropriate radwaste systems are provided and are described in Chapters 9 and 11. See also Criterion 61.

O 3-8

v  ;

SHOREHAM DSAR f Fuel Storage and Handling and Radioactivity Control TCriterion 61)

Criterion The fuel storage and handling, radioactive waste, and other j systems which may contain radioactivity shall be designed to assure adequate safety under normal and postulated accident conditions. These systemr. shall be designed, -(1) with a capability to permit apprcpriate periodic inspection and testing  ;

of components important to sufety, (2) with suitable shielding i for radiation protection, (3) with appropriate containment, confinement, and filtering systems, (4) with a residual heat removal capability having reliability and testability that.

reflects the importance to safety of decay heat and other residual heat removal, and (5) to prevent significant reduction in fuel storage coolant inventory under accident conditions.

Dgsign Conformance .

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

Adequate shielding is provided by maintaining a minimum water depth over the active fuel. Dose rates at the refueling level without the effects of shielding were calculated to be approximately 1R/HR. i The SNPS Secondary Containment is a Seismic Category I controlled leakage building surrounding the fuel pool facility. The Reactor Building Normal Ventilation System (RBNVS) will be used to provide ventilation and secondary containment negative pressure.

Because the gas activity present in the fuel and available for release is primarily noble gas (Kr-85) , the filtering role of the -

Reactor Building Standby Ventilation System (RBSVS) is not ,

required. Certain components of the RBSVS are needed to support operation of the RBNVS. These components will remain functional to provide these services. As discussed in Chapter 15, credible potential releases from accidents are small in comparison to 10CFR100 limits, and the Reactor Building Standby Ventilation System is not required to reduce offsite doses due to postulated accidents.

Radiation monitoring is provided as described in Chapter 11 and O 12 to detect radiological releases. l Because of the extremely low residual heat load (approximately ,

550 watts) associated with the SNPS spent fuel, active fuel pool l 3-9

1 l

l i

{} SHOREHAM DSAR cooling is not required. Reliable fuel pool makeup sources j including condensate storage, demineralized water, and fire protection water, are capable of maintaining pool water inventory  ;

to compensate for evaporation. Chapter 9 contains a complete discussion of makeup requirements. j i'

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 of spent fuel storage. The off-gas system is not needed. Any Krypton 85 will )

be retained within the fuel cladding. Should pin-hole leaks develop, the gases will be handled by the ventilation systems.

They will be discharged to atmosphere via the main plant vent. i

. The radiological consequences of this type of release are '

, neglioible. 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 conductivity, low conductivity, chemical or laundry wastes.

Processing includes filtration, ion exchange, analysis, and 7 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 radioactive materials to the environs above those allowed by the applicable regulations.

The fuel storage and handling and radioactive waste systems are designed to assure adequate safety under normal and postulated -

accident conditions. The design of these systems meets the requirements of Criterion 61.

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

_revention of CrM.icality in Fuel Storage Handling P

TCHterion 62)

Criterion 3-10

- - - - .e.--. . _-_.,.n., _ . _

4 SHOREHAM DSAR 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 storage facilities are provided to preclude accidental criticality for new and spent fuel. Criticality in spent fuel storage is prevented by the geometrically safe configuration of the storage rack. There is sufficient spacing between the assemblies to assure that the array, when fully loaded, is substantially suberitf,;al. Fuel elements are limited by rack design to only top loading and designated fuel assembly positions.

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

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

r-' The fuel handling system is designed to provide a safe, effective i 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 62.

For further discussion, see the following section:

Section 9A Criticality Analysis l

Monitoring _ Fuel and Waste Storage (Criterion 63) l criterion Appropriate systems shall be provided in fuel storage and radioactive waste systems and associated handling areas, (1) to detect conditions that may result in loss of residual heat removal capability and excessive radiation levels, and (2) to initiate appropriate safety actions.

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 in the radwaste control room on high pressure differential. Alarmed conditions include high/ low fuel pool level. The refueling level ventilation exhaust radiation monitoring system detects abnormal amounts of radioactivity. As demonstrated in Section 9A and 3-11

c i

,e m SHOREHAM DSAR

\J ,

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

Area radiation and sump levels are monitored and alarmed to give I indication of conditions that may result in excessive radiation levels in the fuel storage and radioactive waste system areas.

These systems satisfy the requirements of Criterion 63. ,

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.

Desigg_Conformance Means have been provided for monitoring radioactivity releases resulting from normal and anticipated operational occurrences.

The following station release pathways are monitored:

-( } ,

1. Gaseous releases from the station ventilation exhaust

2. Liquid discharge to the 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.

The semiannual Effluent Release Report is submitted to the NRC.

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

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

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

1. The integrity of the reactor coolant pressure boundary

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

3. The capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the guideline exposures of 10CFR100, 3-12

--rr gw-+

i i

SHOREHAM DSAR  !

[u.)i b 4

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

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  !

defines the potential for a radiological release. Based on thin analysis it has been concluded that potential i radiological releases are far below the exposure limits of i 10CFR100. The analysis in Chapter 15 of this report assumes

• that the structural integrity of the filled fuel pool, fuel  ;

pool liner, reactor building structure and fuel racks together form a passive safety system that requires a seismic i Category I designation. The Category I designation has been ,

maintained for fuel handling equipment as well.

A reclassification of structures, systems, and components is e provided in DSAR Table 3.2-1. Table 3.2-1 supplements the information provided in USAR Table 3.2.1-1. The quality group classification in USAR Table 3.2.1-1 reflects the original design basis. As analyzed in Chapter 15, active  !

cooling of the spent fuel pool is not required and pool makeup requirements are minimal. Supporting systems are E required to maintain building habitability, provide radiation monitoring capability, and normal operating service  ;

i O: functions.

l- Design Basis Earthquakes (DBE) and Operating Basis

( Earthquakes (OBE) are described in the Shoreham USAR Section 2.5.  ;

Structures, systems, and components whose safety functions require conformance to the quality assurance requirements of 10CFR50, Appendix B, are summarized in Table 3.2-1 under the heading, LILCO Quality Assurance Category, with the notation l

l I.

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

A new designation, Q.A. Category IIA, is utilized for systems, structures and, components originally Q.A. Category I that are no longer required to meet 10CFR50 Appendix B in the defueled condition. Chapter 17 discusses the graded level of 0.A. requirements for this equipment.

l l

A Q.A. Category IIA designation thus indicates that the item l was re-classified with respect to USAR Table 3.2.1-1.

A IIA and II classification provided together for a given system on Table 3.2-1 implies that the portion of the system that was originally Q.A. Category I in the USAR is now Q.A.

Category IIA and the Q.A. Category II portion remains ,

s unchanged.

1 3-13 1

I SHOREHAM DSAR 3.3 WIND AND TORNADO LOADING l

! The information contained in the USAR remains the same although  ;

L the requirements to protect safe-shutdown equipment no longer

! exists.

! 3.4 WATER LEVEL (FLOOD) DESIGN l

The design of flood-protected structures remains the same ,

although the requirements to protect safe-shutdown equipment no l longer exist. j 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. That protection is adequately l

provided by the reactor building wall and roof structures and i also by the spent fuel pool structure.

l L

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

In the defueled state high energy piping systems inside primary containment listed in USAR Table 3.6.lA-1 are no longer pressurized and thus piping rupture need not be-postulated. That protection is adequately provided by the reactor building wall and roof structures and also by the spent fuel pool structure. l 1

I 3.7 SEISMIC DESIGN l

i seismic design methods remain the same; however, hydrodynamic load effects resulting from safety relief valve discharge and l

loss-of-coolant-accidents are no longer applicable for a defueled i reactor. ,

3.8 DESIGN OF SEISMIC CATEGORY I STRUCTURES The design methods for seismic Category I structures such as the J c

L reactor building will remain as described in USAR Section 3.8 l l except that Safety Relief Valve (SRV) and LOCA hydrodynamic loads  !

l are no longer applicable to a defueled reactor. l l

l 3.9 MECHANICAL SYSTEMS AND COMPONENTS l 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, l therefore, not applicable to the DSAR. l In.the future, mechanical equipment will be accorded the safety significance demonstrated by the classification in Table 3.2-1 of the DSAR.

3-14

v c' SHOREHAM DSAR

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3.30 SEISMIC QUALIFICATION OF SEISMIC CATEGORY I INSTRUMENTATION AND ELECTRICAL EQUIPMENT Seismic category I equipment is identified in Table 3.2-1 and is limited to structures and equipment required to maintain the integrity of the fuel in the spent fuel pool. As discussed in Section 3.2, only the Reactor Building, fuel pool, fuel racks, and fuel handling equipment are required to be seismic. category I. The instrumentation described in USAR Section 3.10 is no longer required to be seismically qualified. This equipment is given a Q.A. Category IIA designation (See Chapter 17 and Table 3.2-1) so that deviations from original seismic requirement can be tracked.

3.11 ENVIRONMENTAL DESIGN OF MECHANICAL AND ELECTRICAL EQUIPMENT i

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 10CFR50.49

' 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 L condition afterward. Since the fuel has been removed and stored I in the fuel pool, LOCA or HELB cannot occur (see Chapter 15), and there is no potential for creation of harsh environment (i.e.,

the remaining design basis accidents discussed in Chapter 15 do not result in harsh environments) . Based on these conditions, 10CFR 50.49 is not applicable, therefore the environmental qualification program is not required. Environmentally qualified electrical equipment will be designated O.A. Category IIA so that deviations from the EQ program can be tracked.

l 3.12 SEPARATION CRITERION FOR SAFETY RELATED MECHANICAL AND ELECTRICAL EQUIPMENT The systems described in this section are no longer required to fulfill a safety related function regarding the storage of spent fuel. Thus, there no longer exists a need to maintain separation criteria for these systems. Q.A. Category I equipment will be designated Q.A. Category IIA whereby deviations can be tracked in accordance with the LILCO Q.A. Program.

O 3-15

o.  !

SHORERAM DSAR 3A Computer Programs for the Stress Ana19 sis of Category I '

Structu;res, Dynam;,c and Static Analys),s. and Dynamic and 5 tress Analysis o:f Seismic Category I P;, pine Systems I

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

information on this subject.

F 35 NRC Regulatory Guides This section is described in the USAR. Specific topics are j

covered elsewhere in this DSAR. i 3C Pipe Failure Outside Primary Containment

. i In the defueled state, piping systems outside primary containment which were considered high energy systems are no longer '

pressurized. Pipe rupture need no longer be postulated.

'I C)  :

O 3-16

( SHOREHAM DSAR s

TABLE 3.2-1 j EQUIPMENT CLASSIFICATION SPENT FUEL STORAGE LILCO ,

QUALITY SYSTEM / ASSURANCE SEISMIC COMPONENT CATEGORY CATEGORY COMMENTS I. Reactor System IIA & II W/A NR II Nuclear Boiler IIA & II N/A NR III Recirculation System IIA & II N/A NR IV Control Rod Drive Hydraulic System IIA & II N/A NR V Standby Liquid Control System IIA & II N/A NR VI Neutron Monitoring IIA & II N/A NR VII Reactor Protection IIA N/A NR VIII Fixed Process, IIA & II N/A (1)

Aarborne, and E:! fluent Radiation i Monitors 4 IX RHR IIA & II N/A NR .

X Core Spray IIA & II N/A NR XI HPCI IIA & II N/A NR XII RCIC IIA & II N/A NR XIII F_uel Service Equipment

1. Fuel" preparation machine I I

2. General purpose grapple I I XIV Reactor Vessel Service Equipment

1. System Line Plugs IIA N/A NR

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

3. Drywell head lifting rig I I 3-17

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

SHOREHAM DSAR l

(

TABLE 3.2-1 (Continued)

EQUIPMENT CLASSIFICATION

• SPENT F_UEL STORAGE LILCO QUALITY SYSTEM / ASSURANCE SEISMIC ,

COMPONENT CATEGORY CATEGORY COMMENTS XV In-vessel Service Equipment

1. Control rod grabple I I XVI Refueling Equipment

1. Refueling platform I I

2. Refueling bellows, drywell II N/A i 3. Refueling bellows,

! cavity reactor II N/A

( 4. New Fuel Inspection Stand II N/A NR ,

1 XVII Storsoe Equipment I

1. New Fuel Storage Racks IIA N/A NR

2. Defective fuel ,

storage container. I I '

3. Spent fuel pool, dryer /sep. pool, reactor cavity I I l

4. Spent fuel storage I I racks XVIII Radwaste System IIA & II N/A XIX Reactor Water Cleanup System IIA & II N/A XX Fuel Pool Cleanup Subsystem

1. Demineralizer vessel II N/A

2. Filters II N/A

3. Pumps, purification

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

5. Valves II N/A

6. Tanks, backwash storage and air accumulator II N/A 3-18

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

4

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SHOREHAM DSAR

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TABLE 3.2-1  ;

(Continued) l l

EQUIPMENT CLASSIFICATION

~~ SPENT FUEL STORAGE LILCO QUALITY SYSTEM / ASSURANCE SEISMIC COMPONENT CATEGORY CATEGORY COMMENTS XXI Fuel Pool Cooling Subsystem

1. Piping IIA N/A

2. Valves . IIA N/A ,

XXII _ Control Room Panels i

1. Electrical modules IIA N/A

2. Cable IIA N/A XXIII Local Panels f' l.

2.

Electrical modules Cable IIA IIA N/A N/A XXIV Offgas System IIA N/A NR

{ XXV Service Water System IIA & II N/A ,

XXVI Compressed Air System IIA & II N/A L XXVII Onsite Power Systems (USAR safety related)

a. Diesel Emergency Power Systems IIA N/A NR(2) l b. AC Power Systems IIA N/A

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

d. Fire Stops IIA N/A

e. DC Power Systems IIA N/A XXVIII Primary Containment Atmosphere IIA N/A NR Control l

l XXIX a) Reactor Build jin Normal Ventilation II N/A l b) Reactor Building Standby Ventilation IIA N/A NR*

l I

l

• Certain components such as fans and valves will remain functional to support RBNVS operations.

l 3-19

4 i

SHOREHAM DSAR TABLE 3.2-1 (Continued) j EQUIPMENT CLASSIFICATION SPENT FUEL STORAGE j LILCO QUALITY l SYSTEM / ASSURANCE SEISMIC ,

' COMPONENT CATEGORY CATEGORY COMMENTS XXX Primary Containment Purgg IIA & II N/A NR XXXI Power Conversion IIA & II N/A NR XXXII Condensate Storage and Transfer I, IIA & II N/A ,

XXXIII Emergency Support Facilities

1. TSC Bldg. II I l 2. EOF II N/A NR(3)

3. OSC II N/A (f

XXIV MSIV Leakage Control IIA & II N/A NR XXXV Miscellaneous

1. RB Polar Crane I I

2. ECCS Loop Level IIA N/A NR XXXVI Reactor Building Closed Loop Coo:.ing IIA & II N/A NR I

XXXVII gE uipment and Floor Drains IIA & II N/A XXXVIII Miscellaneous Ventilation l

Systems i

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

2. Screenwell pumphouse H&V IIA N/A l- 3. Relay and emergency L switchgear H&V IIA N/A

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

5. Diesel generator room IIA N/A NR ventilation 3-20 l

I l

SHOREHAM DSAR TABLE 3.2-1  :

(Continued)

EQUIPMENT CLASSIFICATION SPENT FUEL STORAGE LILCO QUALITY SYSTEM / ASSURANCE SEISMIC '

COMPONENT CATEGORY CATEGORY COMMENTS XXXIX Area Radiation Monitoring System ,

1. All components II N/A

2. High Range Area IIA N/A NR XL Leak Detectio_n System IIA N/A NR

+

XLI Fire Protection System

1. Water spray deluge II N/A systems

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

3. Portable and wheeled I extinguishers II N/A l

l XLII Civil Structures l

1. Reactor building I I l

2. Office and service building II N/A

3. Screenwell IIA' N/A

4. Control building IIA N/A

5. Turbine building II N/A

6. Intake Canal II N/A

7. Discharge tunnel II N/A

8. Discharge pipe and diffuser II N/A

9. Radwaste Building I I

10. Auxiliary boiler and N/A

! MG set building II l 11. Biological shielding IIA N/A

12. Missile barriers IIA N/A

13. Waterproof doors IIA N/A

14. Site grading II N/A

15. Masonry walls IIA N/A l 3-21

( SHOREHAM DSAR I

T_ABLE 3,2-1 t (Continued)

EQUIPMENT CLASSIFICATION

$ PENT FUEL STORAGE LILCO i QUALITY

• SYSTEM / ASSURANCE SEISMIC COMPONENT CATEGORY CATEGORY COMMENTS XLIII Primary Containment Structure IIA N/A NR l XLIV Safety Parameter JDisplay System IIA & II N/A NR XLV Post Accident Sample System II N/A NR XLVI Containment Isola-t on Yalve Position n lestor IIA & II N/A NR

,O XCVII-Accident Monitoring IIA N/A NR Instrumentation (NUREG 0578)

I l'

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l SHOREHAM DSAR

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TABLE 3.2-1 l (Continued)  ;

.EEI LILCO Quality Assurance Category:

1 - Meets 10CFR50 Appendix B requirements (same  :

as USAR). [

IIA - Systems which were safety related are now hon-safety related. Deviations from safety i related requirements will be documented (see Section 17.2.?).

II - Meet requirements of purchase specification (same as USAR).

Seismic Category

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

N/A - Seismic requirements for DBE/OBE earthquake -

l are not applicable to the equipment.

Comments:

NR - Not required (System secured from service or 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.

(3) - Based on LILCO Defueled Emergency Plan, the EOF is not required.

3-23 l

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SHOREHAM DSAR

(~N y/

CHAPTER 4 REACTOR This Chapter includes reactor description, mechanical design, nuclear design, thermal and hydraulic design, reactor materials  ;

and control rod drive housing supports. In the plant's defueled )

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

4.1 REACTOR

SUMMARY

DESCRIPTION The NSS system is no longer needed for the defueled condition and hence is depressurized.

4.1.1 Reactor Vessel The reactor vessel design and description are covered in Section

- 5.4.

4.1.2 Reactor Internal Components Fuel Rod i

J-A fuel rod consists of uranium dioxide (U0 3 ) pellets and a l zircaloy-2 cladding tube. A fuel rod is made by stacking pellets

( into a zircaloy-2 cladding tube that has been evacuated and backfilled with helium. The tube is sealed by welding zircaloy end plugs in each end of the tube.

The BWR fuel rod is designed as a pressure vessel. The ASME Boiler and Pressure Vessel Code,Section III, is used as a guide in the mechanical design and stress analysis of the fuel rod.

The rod is designed to withstand the applied loads, both external ,

and internal. The fuel pellet is sized to provide sufficient volume within the fuel tube to accommodate differential expansion between the fuel and cladding. Overall fuel rod design is conservative in its accommodation of the mechanisms affecting fuel in a BWR environment.

i Fuel Bundle Each fuel bundle contains 62 fuel rods and two water rods, which

("' are spaced and supported in a square (8 x 8) array by a lower and upper tie plate. The fuel bundle has two important design features:

4-1

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SHOREHAM DSAR  ;

I

1. The bundle design places minimum external forces on a fuel rods each fuel rod is free to expand in the axial direction.

2. The unique structural design permits the removal and replacement, if required, of individual fuel rods.

The fuel assemblies that make up the core are designed to meet l all criteria for core performance and to provide ease of handling. Selected fuel rods in each assembly differ from the j others in uranium enrichment. This arrangement produces more I uniform power production across the fuel assembly, thereby significantly reducing the amount of heat transfer surface required to satisfy the design thermal limitations.

1 4.1.3 Reactivity Control System This system is no longer; needed as there is no fuel in the i reactor vessel. l 4.1.4 Analysis Techniques The description contained under this heading in the latest revision of the USAR remains unchanged. Refer to the USAR for information on this subject.

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

Justification for this limit can be found in Appendix A, of i GeneygJElectricStandardApplicationforReactorFuel(GESTAR l II).

l l 4.5 REACTOR MATERIALS l

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.

4-2 l

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

I t

I

() SHOREHAM DSAR CHAPTER 5 REACTOR COOLANT SYSTEM 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 and all sections, including Appendices, of USAR Chapter 5 are not, applicable to the DSAR. The possible exception is that if the reactor is layed up wet, the RWCU System will be utilized.

5.5.7 Residual Heat Removal System ,

The Residual Heat Removal (RHR) System is described in the USAR.

In the defueled status of the Shoreham Nuclear Power Station the '

RHR System serves no function. This system is protected.

5-1

- (' ' SHOREHAM DSAR l \ s l

CHAPTER 6 l

ENGINEERED SAFETY FEATURES i 6.1 GENERAL ,

Because of the Defueled Plant Configuration, there is no longer a need for engineered safety features (ESP) systems at Shoreham.

This is substantiated by a review of the Design Basis Accidents ,

and Postulated Transients. These are covered in Chapter 15. ,

This chapter discusses the effect of radiological accidents in {

the Secondary Containment. The Secondary Containment is utilized for maintaining a controlled and monitored release point for the design basis accident, the Fuel Bundle Drop accident. In  :

addition, a worst case release of the entire gaseous inventory of the fuel is postulated in Chapter 15 that bounds any possible large scale mechanical-damage event.

6.2 CONTAINMENT SYSTEMS 6.2.1 Containment Functional Design 6.2.1.1 Design Basis 6.2.1.1.1 Safety Criteria The primary containment system is not required and will not be maintained functional as there will be no fuel within the primary containment structure. The secondary containment will maintain a subatmospheric pressure for postulated radiological accidents to assure radiological monitoring of building releases. It is not '

needed to mitigate the consequences of an accident.

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

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

A loss of normal AC transient may result in loss of subatmospheric conditions within the secondary containment.

O However, as explained in Chapter 15 the time period before any 6-1

l I

i SHOREHAM DSAR i

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^~/ ]

significant loss of fuel pool water level is long, and l appropriate corrective action is postulated to be taken before l fuel damage occurs. There are no radiological consequences associated with this event.

6.2.1.2 System Desion ]

The reactor building, which completely encloses the primary containment and acts as the secondary containment, is maintained -

at subatmospheric pressure by the RBNVS.

6.2.1.3 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.

6.2.3 Contairient Air Purification and Cleanup Systems This subsection is not applicable as it deals with the filtration

. ') portion of the RBSVS which is no longer required.

6.2.4 containmont Isolatio_n System This subsection is no longer applicable as it deals with the

primary containment isolation system. The primary containment is no longer maintained.

6.2.5 Combustible Gas control in Containment L

l This subsection is no longer applicable as it is concerned with  !

! hydrogen combustion inside the primary containment.

6.3 EMERGENCY CORE COOLING $YSTEMS 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 l is not' applicable to DSAR.

l 6.3.2.2.3 Core Spray System

~~

(Q./ 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. The CS System is recured in a protected state.

6-2

SHOREHAM DSAR

('

C 6.4 EABITABILITY SYSTEMS ,

The systems, aside from the control room air conditioning ,

portion, are secured in a protected state because they are not '

needed since the fuel is stored in the spent fuel pool. The control room air conditioning system is described in Section 9.4.1.

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

6.6 OVERPRESSURIZATION PROTECTION

! The overpressucization prctection 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 L

The main steam isolation valves (MSIVs) are not required in the f')

defueled state and are, therefore, not included in the DSAR (See Chapter 5. of DSAR) . .

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

6.9- CONTROL ROD VELOCITY LIMITERS  ;

The control rod velocity limiters are not required in the defueled state and this Section is, therefore, not included in the DSAR (See Chapter 5 of DSAR) .

6.10 MAIN STEAM LINE FLOW RESTRICTORS The main steam line flow restrictors are not required in the defueled state and this Section is, therefore not included in the l DSAR, (See Chapter 5. of DSAR).

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

O 6-3 e , - -

.. . s . ..

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SHOREHAM DSAR j y

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

t 4

9 8 4

I

(

6 4

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6-4

i SHOREHAM DSAR CHAPTER 7 ,

INSTRUMENTATION AND CONTROLS

7.1 INTRODUCTION

This chapter presents the details of the control and I instrumentation systems in the plant except radiation monitoring i systems, and electrical power systems which are described in i Chapters 8, 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 IIA. USAR Figure 7.1.1-1 is no l longer applicable due to the defueled status of the plant. r 7.1.1.1 Identification of Individus1 Systems i >

This section identifies the individual systems which are retained >

and presents a general description of their instrumentation and

(

..() control functions.

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.

L 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 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 Interlockg 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

SHOREHAM DSAR 7.1.1.1.6 Reactor Manual Control System l

Reactor vessel instrumentation monitors and transmits information concerning key reactor vessel operating variables. Portions of this system will only be used if a wet layup of the reactor ,

vessel is utilized.  ;

1 7.1.1.1.7 Reactor Vessel Instrumentation This system will be used only if a wet layup of the reactor vessel is utilized.

7.1.1.1.8 Reactor Recirculation System 1 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 j This system is not needed to support the storage of the fuel in the fuel pool, therefore it is not included in the DSAR. t 7.1.1.1.10 Pressure Requiator 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 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 i 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.

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

7-2

SHOREHAM DSAR l

'( )

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 Clasnup System The reactor water cleanup (RWCU) system instrumentation and controls provide manual initiation of system equipment to i maintain high water purity in the reactor water. This system will be used only if the reactor system is placed in a wet layup >

condition.

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 l

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 i processing and disposing of the radioactive process wastes.

l 7.1.1.1.20 Emergency Diesel Generator _s This system is not needed to support the storage of the fuel in the fuel pool.

7.1.1.1.21 Turbine Building Closed Loop Cooling Water System

! The turbine building closed loop cooling water (TBCLCW) system .

Instrumentation and controls remain functional to maintain the turbine building cooling water system at design temperature and monitor system performance. The TBCLCW system also cools the equipment in the radwaste building and supports the station pressurized air compressors.

7.1.1.1.22 Service Water System

! Under normal conditions the service water system provides cooling for the plant components.

7.1.1.1.23 Recirculation Pump Trip __ System 1

This system is not needed to support the storage of the fuel in j the fuel pool.

l 7-3

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i SHOREHAM DSAR 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 Reactor Building Closed Loop Cooling Water System This system is not needed to support the storage of the fuel in the fuel pool.

7.1.1.1.26 Primary Containment Atmospheric Control System  :

This system is not needed to support the storage of the fuel in i the fuel pool.

7.1.1.1.27 Fuel Pool Cooling and Cleanup Systems Fuel pool cooling and clea6up systems instrumentation and i controls remain unchanged except that the cooling portion has been secured because evaporative cooling is sufficient to remove the small amount of decay heat.

7.1.1.1.28 Control Room Air Conditioning System l l

The control room air conditioning (CRAC) system instrumentation and controls 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 the air supply fans.

7.1.1.1.29 Chiller Equipment Room Ventilation System

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

7.1.1.1.30 Diesel Generator Room Emergency Ventilation Systems f This system is not needed to support the storage of the fuel in the fuel pool and is protected.

.7.1.1.1.31 Relay Room, Emergency Switchgear Rooms, And Computer RoomAi(ConditioningSystem The relay room, emergency switchgear rooms, and computer room air conditioning system instrumentation and controls are maintained functional to automatically control the ventilation system to

'(~3 maintain these rooms at their design temperature and system

-(/ performance.

7-4 1

l

() SHOREEAM DSAR 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 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 4 L

the fuel pool.

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.

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. .

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

7.1.1.1.37 Compressed Air System (SRV Accumulators) _

This system is not_needed to support the storage of the fuel in the fuel pool.

o 7.1.1.1.38 Main Steam Isolation valve Leakage Control System This system is not needed to support the storage of the fuel in the fuci pool.

7.1.1.2 Classification Section 3.2 provides a reclassification of systems based on their importance to safety. Q.A. Category IIA applies to those systems that no longer fulfill a safety function for a defueled reactor.

7-5

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f

SHOREHAM DSAR

- 7.1.2 ' Identification of Safety Design,3pses and Nonsafety Design Bases Criteria The following sections remain as stated in the USAR with the exception-that Q.A. Category I systems or portions of systems have been reclassified as Q.A. Category IIA:

4 7.1.2,1.7 Reactor Vessel Instrumentation TIf the reactor is laped'Ep wet) 7.1.2.1.12 '5greenwell Pumphouse Ventilation Syctem 7.1.2.1.16 vi.ctor Water Cleanup System TIf the, reactor is layed up wet) 7.1.2.1.19 Radwaste System 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 7.1.2.1.31 Relay Room, Emergency Switchgear 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 SAFETY FEATURE SYSTEM This section is not needed to support the storage of the fuel in the fuel pool.

7-6

w ,

SHOREHAM DSAR 7.4 SYSTEMS REQUIRED FOR SAFE SHUTDOWN This section is not needed to support the storage of the fuel in

'the-fuel pool.

7.5 SAFETY'RELATED DISPLAY INSTRUMENTATION This'section is not needed to support the storage of the. fuel in the fuel pool..

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 system is required.

7.6.1.2 Fuel Pool Cooling and Cleanup System Instrumentation and-Controls 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 Instrumentations The-spent fuel. pool level is indicated and alarmed on high and low levels in~the main control room. The purpose of this 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 temperature.

i 7-7

7 s

SHOREHAM DSAR

' f.

7.6.2.2.1' ' General Functional Requirement,s Conformance of ruel Pool Cooling and Cleanup System Instrumentation and Controls All other USAR Part 7.6 sections not listed above are not needed in the defueled-condition.

7.7 CONTROL SYSTEMS NOT REQUIRED FOR SAFETY The following sections remain as stated in the USAR:

7.7.1.6 Liquid Radwaste_ Control System Instrumentation and Controls

'7.7.1.7 Turbine Building closed Loop Cooling Water System Instrumentation and Controls 7.7.1.8 Reactor Water Cleanup System Instrumentation and Controls Required functional to support wet layup of reactor only.

7.7.1.9 Reactor Vessel Instrumentation As It Pertains To Water 1 9' Level Instrumentation Only _

Required. functional for wet layup of reactor systems.

7.7.1.11 Refueling Interlocks Instrumentation and Controls 7.7.2.6.1 General Functional Requirements Conformance for Liquid Radwaste System Instrumentation and Controls 7.7.2.7.1 General Functional Rectlirements Conformance for Turbine Buildino Closed Loop Cooling Water System Instrumentation and Controls 7.7.2.B.1 General Functional Requirements Conformance for Reactor Water Cleanup System Instrumentation and Controls Required functional to support wet layup of reactor only.

7.7.2.9.1 General Functional Requirements Conformance for Reactor Vessel Instrumentation as it pertains to_ water level instrumentation only Required functional to support wet layup of reactor only.

l 7-8

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1

./Y .SHOREHAM DSAR .

I Q ):

L7.7.2.ll.1 General Functional Requirements Conformance for Refueling Interlocks Instrumentation and Controls 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.

7B~ Analog Transmitter / Trip Unit System for Engineered Safeguard

~

Sensor Trip Units .

1 This section 1s not needed to support the storage of the fuel in the fuel pool.

L 1.

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SHOREHAM DSAR CHAPTER 8-e j

f, ELECTRIC POWER

8.1 INTRODUCTION

l! '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 lE power system. Therefore, as stated in Section 8.3.1 item 2, diesel generator and safety related equipment will not be required while L the plant is in the defueled mode.

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

USAR.

8.1.3 Offsite Power System While~in-the defueled condition the offsite power system provides power,to all operating plant equipment. Power to the Shoreham Nuclear Power Station is provided from the LILCO system through 138KV or 69KV circuits. The 138KV switchyard is arranged in a

.two bus configuration with circuit breakers and switches arranged to permit isolation and/or repair of either bus section. Four

, 138KV circuits enter into the switchyard (two per bus) each containing a circuit breaker'at the connection to its respective L

bus. Two separate rights-of-way are provided, each containing two of the 138KV circuits. The 69KV circuit from the tildwood l

substation enters the site sharing one of the aforementioned rights-of-way for a distance of one mile. This circu'., however, is mounted on separate towers and is separated from the 138KV l

circuits. The detailed description of the remaining offsite j

system remains as described in the USAR except as follows:

Three Brookhaven 80MW (each) Combustion Turbine units are located

(~N cn LILCO SNPS property approximately 3600 feet from the 138KV L A switchyard. These units are connected into one of the 138KV l

I 8-1

i. -- - .--

() SHOREHAM DSAR  ;

Holbrook transmission lines and are available to provide an additional source of onsite power to the SNPS. (see figure 8.2.1-2)

-The spare Reserve Station Service and Normal Station Service 3 transformers will no longer be required.

8.1.4 OnSite AC Power System i

^

The station' electrical power system includes electrical equipment and connections required to provide power to and control the -i operation'of electrically driven station equipment in the defueled condition. However, the Emergency Diesel Generators are not required to operate to maintain safe conditions.

8.1.5 -On Site DC Power System >

The DC system consists of two independent systems containing-total of 6 separate'and independent 125V DC battery sources.

L L These-are designed to provide a minimum of two hours of DC

l. emergency power. However, since the defueled status of the system-has reduced the emergency power load on these batteries, this will extend the duration of power supply from each source.

(G~T.

The onsite DC power system as described in the USAR remains unchanged in the defueled condition except for the following:

a) The'24V DC power source will no longer be required, This system'provides power to the. Nuclear Source and Intermediate n Range Instrumentation which is no longer in service in the defueled condition.

b) T'e n safety related battery source and equipment is not required with the plant in the defueled state.

8.1.6 Identification of Safety Related Systems  ;

The description contained under this heading in the latest revision of the Shorcham USAR will not be applicable in the defueled state.-

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

8.1.7 Identification of Safety Criteria fg V The description contained under this heading in the latest revision of the Shoreham USAR is not applicable in the defueled state.

8-2

SHOREHAM DSAR l{ }

~

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

8.2 OFFSITE. POWER SYSTEM 8.2.1' Description The description contained under this heading in the latest revision of the Shoreham USAR remain unchanged =except as follows:

Service buses- 101, 102 and 103 described in'DSAR Section 8.2.1.2 are not required to*be maintained as safety related while in the defueled condition. They are reclassified as Category IIA. 1 8.2.1.1 One Line Diagrams and Physical Drawings The information. contained under this heading in the latest revision of the Shoreham USAR remains unchanged in the defueled L condition except as follows:

) 1 - Figure 8.2.1-1 Main one line diagram The diesel generators shown are not operational in the defueled b condition.

2 - Figure-8.2.1-2'One line diagram of 138kV and 69kV systems in Shoreham Area

'Added to the system are the New Brookhaven Combustion Turbines.

8.2.1.2 Transmission Line The description contained under this heading in the latest revision of the USAR remains unchanged in the defueled condition except.that the safety related function of the busses l'

l (1R22-SWG-101, 102, and 103) no longer exists. They are L reclassified as Q.A. Category IIA systems.

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

For further information on this subject refer to the USAR.

8.2.1.4 Transmission Line Exits The description contained under this heading in the latest revision of the USAR remains unchanged in the defueled condition except for the following:

8-3

L l 4 SHOREHAM DSAR.

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 7

. described in'the USAR is not required in the defueled condition.

8.3 Onsite Power Systems u

The' plant power system is designed to provide an adequate source of electrical power to.all systems required to be operational in the defueled condition.

8.3.1 AC-Power Systems The Plant electrical power (AC) layout configuration as designed /

described in the USAR is not changed in the defueled condition, except as follows:

1- Equipment, switchgear, or buses built and designed to safety standards are not maintained as safety related or service inspected in the defueled condition.

2- Diesel generator sets with safety related equipments are no longer needed for the plant in the defueled condition. (see USAR 8.3.1.1.5 onsite standby power supply).

3- Adequate equipment protection and emergency measures are

.available for the required plant. electrical systems in the defueled condition.

Appropriate layup measures have been taken to protect the longevity of the various electrical equipments. The equipment, switchgear, and bus have been reclassified to Q.A. Category IIA.

B.3.2 DC Power Systems 8.3.2.1 Description The description contained under this heading in the latest revision of the Shoreham USAR remains unchanged in the defueled condition except as follows:

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

2- All class IE/ safety related functions of the DC system no longer exist.

Appropriate layup measures have been taken to protect the longevity rf the various DC electrical equipment. Q.A. Category I equipment is now Q.A. Category IIA.

8-4

~

SHOREHAM DSAR li4^0 Thefnew Brookhaven Combustion Turbines =are addei~to the existing transmission.line configuration. (Figure 8.2.1-2)

8. 2'. 2 Analysis'

- The ' basis of the analysis ru)' longer exists. The analysis as described ine the USAR is not' required in the defueled condition.

N 8.3 'onsite Power Systems ,

The plant power system is designed to provide an adequate source of electrical power to all systems required to be operational in the'defueled condition.

8.3.1 AC Power Systems The Plant electrical power (AC) layout configuration as designed /

described in the USAR is not changed in'the defueled condition, except as follows:

1- Equipment, switchgear, or buses built and designed to safety standards are not maintained as safety related or service inspected in the defueled condition.

2- Diesel generator sets with safety related equipments are no longer needed for the plant in the defueled condition. (see USAR.8.3.1.1.5 onsite standby power supply).

3- Adequate equipment protection and emergency' measures are

-available for the required plant electrical systems in the defueled condition.

Appropriate layup measures have been taken to protect the longevity of the various electrical equipments. The equipment, switchgear, and bus ~have been reclassified to 0.A. Category IIA.

B.3.2 DC Power Systems

'B.3.2.1 Description The description contained under this heading in the latest revision of the Shoreham USAR remains unchanged in'the defueled condition except as follows:

1- The 24V DC system providing power to source and intermediate X range nuclear inskrumentationj is no longer used. x 2- All class IE/ safety related functions of the DC system no longer exist F Appropriate layup measures have been taken to protect the '

longevity of the various DC electrical equipment. Q.A. Category I equipment is now Q.A. Category IIA.

8-4

't.

> l SHOREHAM DSAR 1'd

. P' . ( ..

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.spplicable in-the defueled condition.-

Tables --8.3.2.1 and 8.3.2.2 - related to plant design basis loads are no longer applicable in the defueled condition.

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L/ : SHOREHAM DSAR s

CHAPTER 9 AUXILIARY SYSTEMS j

'9.1 FUEL STORAGE AND HANDLING 9.1.1 New' Fuel Storage Since no new fuel will be received, this section of the USAR is not required in the defueled condition.

9.1.2 Spent Fuel Storage '

9.1. 2. l' 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.

9.1.2.2 Facilities Description

' Spent fuel. storage racks provideia place in the spent fuel pool for storing spent fuel received from the reactor vessel. The location of the spent; fuel pool within the reactor building is shown on Figure 3.8.1-4. The general arrangement of the storage soace, illustrated on DSAR Figure 9.1.2-1, will permit the storage of 2,176 fuel assemblies (the current installed capacity in the spent fuel pool'is for 1,420 fuel assemblies) plus 144 control rods.

The remainder of this Section is identical to that in the USAR, up to the first paragraph of page 9.1-8. From this point to the end of page 10, the text is deleted.

9.1.2.3 Safety Evaluation This section remains identical to that in the USAR except that in the DSAR Appendix 9A provides the criticality analysis.

9.1.2.4 Tests and-Inspection The description contained under this heading in the latest

/~'. revision of Shoreham USAR ren:ains unchanged. Refer to the USAR for information on this subject.

9-1

G f(j SHOREHAM DSAR 9.1'.2.5' Radiological Considerations The description contained under this heading in the latest .

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

~9.1.3 Fuel Pool Cooling and Cleanup System All of the equipment in this system will be retained for operation, but in a_ modified manner. Since the fuel pool cooling subsystem is designed to remove the decay heat produced by spent fuel assemblies, as described-in the USAR, and only a negligible amount of heat is expected to be generated from the slightly irradiated spent fuel bundles stored there, the cooling mode is '

not required. Thus reactor building closed loop cooling water is not required.

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

However, the spent fuel pool cooling subsystem will be used in the makeup mode in order to provide normal makeup water to the

\ fuel pool from the condensate storage tank using the condensate

~(d

' transfer and storage system. Alternate makeup sources for the l spent fuel pool are Demineralized and Makeup Water System, Fire I

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.

1 The fuel-pool cannot be inadvertently drained because the pump suctions for the fuel pool cooling and cleanup system are taken above elevation 168, or about 7 feet below the normal water level. If a break occurred in these lines, about 18 feet of water would remain above the fuel in the pool. This is more than enough to provide adequate shielding. Pump returns to the pool are equipped with siphon breakers to prevent inadvertant pool l

drainage.

9.1.4 Fuel Handling System 1

9.1.4.1 Design Basis See USAR. This section is identical to the USAR.

9.1.4.2 Equipment Description See USAR. This section is identical to the USAR.

O_

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SHOREHAM DSAR 9.1.4.3 Description of Fuel Transfer The fuel handling system provides a safe and effective means for transporting-and handling _ fuel from the time it reaches theThe. plant until it leaves the plant after post-irradiation cooling.

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.13.are satisfied.

9.1.4.3.1 Arrival of Fuel On Site No new fuel is expected to arrive on site. Therefore this section of the USAR'is not applicable.

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

9.1.4.3.3 Departure of Fuel from Site

() This section applies as written in the USAR.

In addition:

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

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

The remainder of USAR Section 9.1.4 is applicable.

9.2 WATER SYSTEMS 9.2.1 Service-Water System

-The Service Water (SW) System is as described in USAR Sections 9.2.1;l thru 9.2.1.4 with the following changes because of the reduced heat removal requirements with the plant in the de-fueled state.

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

b) One RBSW pump will supply cooling water to one RBSVS/CRAC chiller condenser. No service water is required for RHR, O- -

diesel engine cooling, RBCLCW, drywell cooling, and makeup water to the reactor vessel ultimate cooling connection (UCC). The testable check valve in the UCC will not require 9-3

~'

SHOREHAM DSAR fi l(d(

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) All pumps and motor operated valves will' actuate upon manual initiation signals only. Automatic start /initiatior,due to accident signals will be defeated..

-d)- The double isolation valves which split the RBSE from the TBSW subsystems may be opened to intertie the subsystems _as l required. l e) Normal operation will now consist of only one TBSW pump in use because of the minimal heat load imposed by the TBCLCW i system to support the station air compressors. It will l supply cooling water to.one TBCLCW heat exchanger, the l circulating water pump bearing and the fish retention pool. l I

Cooling water for-the vacuum priming pump seal cooler is not required. The second TBSW pump would remain in automatic standby while-the third pump would be off and rotato in the operational mode with the other two pumps.

f)- The only tests and inspections to be performed on the TBSW system in the-defueled_ condition are those that are deemed to 1

be required for proper operation and maintenance.

g) 'Tabld 9.2.1-1 has been revised.

Section 9.2.1.5 remains unchanged.

9.2.2 Reactor Building Closed Loop Cooling Water (RBCLCW) System l None of the equipment that uses RBCLCW (refer to USAR Section 9.2.2.2) is required to operate with the fuel in the spent fuel L

pool. The one possible exception is the RWCU pump which may be needed if the reactor is layed up wet. Therefore, this system'is not required for operation in the defueled condition. -

9.2.3 Makeup Water Demineralizer System 1

1 The description contained under this heading in the latest l

)- revision of the Shoreham USAR remains unchanged except as l

' l follows:

l l

l l L 1 l) u 9-4 L _-

I SHOREHAM DSAR

" L/ .

1. SBLC, RBCLCW, seal water injection, and vacuum priming are no longer considered principal users of domineralized water in '

the defueled condition.

l~

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

condition, y

o For further information relative to this system refer to the h USAR.

9.2.4 Potable and Sanitary Water Systems The description contained under this heading in the latest

~

! revision of the Shoteham USAR. remains unchanged in the defueled L -condition. For further information on this subject refer to the E USAR.

9.2.5 Ultimate Heat Sink With the fuel in the Spent Fuel Pool, the ultimate heat sink (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 l

' cooling water for normal plant needs (refer to DSAR Section 9.2.1).

9.2.6 Condensate Storage Facilities L

While in the defueled condition the condensate storage facilities l

provide makeup water for the fuel storage pool. The description I of this system in the USAR remains unchanged except as follows:

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

2. HPCI test discharge and CRD pump return lines to the CST are l not required to be active.

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

1

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

!O l

l 9-5

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.i' SHOREHAM DSAR

- 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 in the defueled condition. The only exception is that many of the coolers listed l

. in DSAR Table 9.2.7-1 will normally be valved out of service  ;

- while~the plant remains in the defueled state.

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

'9.2.8 Main chilled Water System.

l This system will not be maintained as an operable system since it 11s not needed with the plant in the defueled condition.

9.2.9- Reactor. Building Standby Ventilation System And Control Room Air Conditioning Chilled Water System Redundancy in this system is not needed since the RBSVS system is not-required to operate in the defueled condition. The heat loads generated by.the electrical equipment in=the control room, relay room and the emergency switchgear room are greatly reduced, such that only one chiller is required to maintain the control p)i s room, relay room and switchgear room at-design conditions. The operating chiller and associated pumps will be manually controlled from the control room. 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 I

The description contained under this heading in the latest revision of the USAR remains unchanged in the defueled condition except for the followings

( 1. Piping that has been installed as ASME III code class 2 is no j longer considered safety related and is reclassified QA t l

Category IIA.

2. Nitrogen will no longer be used for inerting the primary

) containment or for equipment within the primary containment.

I 3. Safety related functions of the compressed air system no

1. longer exist. No pneumatically operated valves are required for safe shutdown.

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

9-6

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e .

SHOREHAM DSAR

9. 3. 2 - Process Sampling System 4 - The Process Sampling System provides' monitoring of certain L ' process- operations while fuel is in the spent fuel pool for either short or long term' storage. The process monitoring is L 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 System are not required.

System Description

L This system is described in the USAR. Changes in status are addressed below.

Reactor Building

(~3 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 Equipme'nt Drain Systems are no longer required, as applicable, except for the Decontamination Sump drains and associated equipment. There is no steam and the turbine is no longer required, so that the only source of radioactive waste-is the Chemical Laboratory.

Radwaste Building The Radwaste Building Equipment and Floor Drainage System is -

maintained operational. The Dirty Waste Sump and Pumps (IN52-TK 114 and 1N52-P-187A/B) and Regenerant Recovery Sump and Pumps (IN52-TK-115 and 1N52-P-181A/B) are no longer required.

9.3.4 Chemical, Volume Control, and Liquid Poison Systems The Standby Liquid Control System is no longer required in the

()

t L defueled condition. The RWCU System is also no longer required unless the Reactor is layed up wet.

9-7

eM ,

eX-SHOREHAM DSAR.

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

" Worst Case Fuel Damage Accident" in DSAR Chapter 15), the refueling floor processiradiation 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 9.4.1 Control Room Air Conditioning System The Control Room AC system remains unchanged in design and operating functions. However, the system is reclassified to QA Category IIA and the filter portion of the system wil no longer

)-sJ be required. The AC-system will only function to provide an OSHA 3,,) environment for the operators during the fuel storage period.

This requires the operation of only one RBSVS/CRAC chiller. All e

automatic initiation systems and interlocks for the habitability E portion of the system will be secured and the AC system will ba 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:

L 1.- Provide ventilation by introducing filtered outside air in the reactor building at a rate of approximately 2.7 air changes-per hour u

2. Remove heat generated by solar and external heat transmission, lighting and the fuel pool.

l

3. Support monitor for radioactive release through the exhaust air system.

4. Induces negative pressure in reactor building for secondary containment integrity.

z()

For further discussion on this system refer to the USAR.

l l

9-8

3 .

SHOREHAM DSAR 9.4.3 Radwaste Building-Ventilation The description. contained under this heading in the latest Shoreham-USAR remains unchanged, except that the charcoal exhaust filtration system is no longer required. Refer to the USAR for, information on this subject.

9 . 4 . 41 Turbine Building ventilation System And Station

-Exhaust System A) - Turbine Building Ventilation System This system is not required to support the storage of fuel in the. .

spent fuel pool. .,

B) Station Exhaust System This system will accelerate the exhaust air from the radwaste building and the reactor building. 'However, only one fan will be needed for.this purpose, allowing one fan.to be secured and still maintain a fan on standby. This will ensure-that the: Isokinetic nozzles located in the upper level.of the exhaust duct will see a

.- sufficiently high velocity to be operational. For further E .

discussion regarding this system refer to the Shoreham USAR.

9.4.5 . Battery _ Room Heating And Ventilation The description contained under this heading in the latest

. revision of Shoreham USAR remains unchanged. Refer to USAR for information on this subject.. This system is reclassified to Q.A.

Category IIA.

9.4.6 'Drywell Air Cooling System This system is not needed while the fuel is stored in the. spent fuel pool.

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

Category IIA.

9.4.8 Plant Heating The description contained under this heading in the latest revision of Shoreham USAR remains unchanged. Refer to USAR for

() information on this subject.

9-9

SHOREHAM DSAR

-9.4.9 Primary Containment Purge System This system is net needed while the fuel is stored in the spent fuel pool.

9.4.10 Diesel Generator Room Ventilation This system is not needed while the fuel is stored in the spent fuel pool. This system is reclassified to Q.A. Category IIA.

l 9.4.11 Relay Room, Emergency Switchgear Room And L Computer Room Air Conditioning System

-The description contained under this heading in the latest revision of Shoreham USAR remains unchanged. Refer to USAR for l information-on this subject. This system is reclassified to QA Category = IIA.

L l 9.5 OTHER AUXILIARY SYSTEMS 9.5.1 Fire Protection System Design Basis l

o ~9 .5.1.1 The design basis section applies with the following additions-I The basic premise of the fire protection discussions in the USAR and FHAR is protection from fire for safety related areas l including areas containing equipment or circuits that are (1)

L required for safe shutdown, or (2) required to prevent or mitigate radiological releases. comparable to 10CFR 100 limits.

Since safe shutdown is assured by non-operation of the plant, and all of the nuclear fuel is in the fuel storage pool, the only

! remaining safety related area is the Reactor Building. J l

Structures, systems components and administrative controls in ]

l place to protect areas, equipment or circuits previously i identified as safety related will be maintained as required for i l

property loss prevention purposes and should be considered the 1 t same as those fire protection features described in the USAR for j protection of non-safety related areas. i

)

.Three documents which were used in the design of the plant's fire protection features and continue to be part of the fire j protection program are. ,

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

L

2. ' Fire Hazards Analysis Report.

1 \

l i L 9-10 )

l

o. . . .- - .._ _ .._ _ _ _ _ __ _ _ ______._____ __.___________________ _ _ _ _ _ _ _

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

m ,

g bi SHOREHAM DSAR

3. Cable Separation Analysis Report:

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

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

I FHAR Chapter 6 reflects reductions in the technical requirements l that are consistent with the text of this DSAR Section 9.5.1. R l

Types of Fires _

l The " types of fires" section applies with no changes. l l

Design Criteria

.The " design criteria" section applies with the following additions l

("% As discussed above, this design will be maintained for property

(_/

loss prevention purposes. However, the " safety related" application of the listed documents, particularly NRC's Branch- J Technical Position APCSB 9.5-1 and Appendix A thereto, is limited l

-to the-Reactor Building. 4 Locations of-Fires The." locations of fires" section applies with the following changes:

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

y Category IIA. The rooms listed as examples of where oil fires j

could occur near safety related equipment no longer fit that description because these areas are reclassified to QA Category IIA. Furthermore, the fire hazard associated with.this equipment is significantly reduced while the equipment is not being used j because: the ignition sources associated with the operating equipment have been eliminated.

Intensity of fire _s This section applies without change.

Fire Characteristics This section applies without change.

I 9-11

1 f

SHOREHAM DSAR TBuilding Arrangement and Structural Features Thel" building arrangement.and structural features" section

' applies with the following changes:

In the response to NRC question 3, as shown in FEAR revision 3, SNPS has stated our intention to replace existing motorized fire dampers:with newly designedLfire 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 IIA. Therefore, The C0 this proposed modification will not systems forLthose rooms are in electric lockout.be implemented.-When a fife is detected, the CO, system controls would cause the dampers to .

c close on an electrical signal. As a backup, the fusible link of each of the.existiny fire dampers is sufficient to cause closure '

of a damper-in the' event of a fire, thus assuring integrity of the fire barriers.

L In contrast with this USAR section, an unprotected HVAC opening exists in the' east wall of each of the three diesel generator u ' rooms within 50 feet of an' oil-filled (Reserve-Station Service).

l ' transformer. This deviation'was reported to the NRC on Licensee 87-021. The proposed corrective action was to

rst Event Report o -i,,,/ ' 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 reclassified to QA Category IIA, l

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

~

sufficient.for non-safety related areas.

Seismic Design This section applies without change.

Water Requirements The " water requirements" section applies with the following additional statement:

Although some areas previously identified as safety related are reclassified-to QA Category IIA, the water supply is not being reduced.

Codes and Standards L

This section applies without changes. SNPS will continue to meet

! the requirements of the applicable NFPA codes for fire protection I systems that remain functional.

'O .

9-12 li ll

4 SHOREHAM DSAR 9.5.1'.2 System Description L

' The " System Description" section applies with the following changest.

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 IIA. Essential circuitry installed for safeNo shutdown of the plant is no longer removal of such cable or change in -

- needed'for that purpose.

its physical separation is contemplated. Similarly, the service water line inside the Reactor Building, where a spara connection exists for manual hookup to the fire protection water system, is reclassified to Q.A. Category IIA. Modifications that would degrade its seismic' design are not contemplated at this time.

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

Charcoal Fires 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.

j Severity, Intensity and Duration of Fires This section applies without changes.

Time Estimates

- This-section applies with.sut changes.

Failure Mode and Effects Analysis This section applies without changes.

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

9-13

I? ,_

l f

l 1

SHOREHAM DSAR Areas protected by CO systems are among those.that are no longer 0 considered safety rel$ted.

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

Pipe Breaks in Fire Protection Systems This section applies without changes. j Failure of Fire Protection System Affecting' Safety j Related Equipment _s I

This section applies with the following changer l Of the ereas listed, only the Reactor Building is still considered safety related.

9. 5.1. 4 ' Tests and Inspections This section applies without changes. q Personnel Qualification and Training

( ). 9.5.1.5 This section applies without changes.

9.5.2 Communications System 9.5.2.1- Design Bases This section of the USAR remains unchanged.

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

1. For the very. low frequency (VLF) 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 Facility (EOF) is not required, since no emergency requiring EOF activation can occur with the fuel in the Spent Fuel Pool.

Tests and Inspections

(} 9.5.2.3 This section of the USAR remains unchanged.

9-14

)

SHOREHAM DSAR 1

9.5.3 Lighting Systems I

While in the defueled condition this system will provide all the I necessary required lighting to the plant and the site. The description of this system in the USAR remains unchanged except I for the following l

1. Section 9.5.3.2, item #2 - the standby AC lighting system i will receive power from plant service buses which are powered

!. from offsite.

2. Same section, item 45 - the fifth lighting subsystem will receive power from DC battery sources while the plant remains L

in the defueled condition.

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

! 9.5.4 Diesel Generator Fuel Oil Storage and Transfer System

- Since emergency power is no longer required with the fuel in the O Spent Fuel Pool, the Emergency Diesel Generators are not required, and sections 9.5.4 - 9.5.7 of the USAR no longer apply.

l 9.5.5 Diesel Generator Cooling Water System 9.5.6 Diesel Generator Starting System l

9.5.7 Diesel Generator Lubrication System 9.5.8 Primary Containment Leakage Monitoring System With the fuel in the Spent Fuel Pool, the Primary Containment Leakage Monitoring System is not required.

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 hand. The design bases remain unchanged. Storage facilities are 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. I

4. Nitrogen for glycol and HW heating. i

() 5. Propane for auxiliary boiler ignition.

9-15

r <

E.N SHOREHAM DSAR L'V The following gases are no longer used or required to be stored .

i in the defueled conditions

1. Hydrogen for main generator.

~2. Hydrogen and oxygen for gas analyzers.

3 . -- Nitrogen for containment inerting.

4. Nitrogen for drywell floor seals..

5. Nitrogen for electrohydraulic control.

g'-

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

7. ' Air for'long term accumulators.

8. Air for standby diesel generators.

uThe-statement in the USAR relative to maintenance and laboratory

. gases. remain unchanged. The safety. evaluation discussed in section 9.5.9.3 of the USAR is only applicable for air for instruments, service breathing; and control and for carbon  ;

' dioxide and halon. Statements relative to the pressure relief valves and gas release hazards remain as discussed in the USAR.

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

i l

L l

II l^

O 9-16

-+3 s.

SHOREHAM DSAR' V(~)

Appendix 9A FUEL CRITICALITY ANALYSIS The Shoreham Spent Fuel Rack (SFR) 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 in Appendix 9A of the Shoreham USAR. The reactivity results which are summarized in USAR Table 9A-4 remain valid for the >

conditions existing at Shoreham after defueling. Furthermore, due to the differences in U-235 enrichment between the SFR i designed and the current Shoreham fuel, a large negative ,

reactivity credit should be taken into account. This is explained as follows:

The Shoreham SFR design is based on a maximum U-235 '

enrichment of 3.1 wt. %. The resulting basic cell k is calculated to be 0.9129 without uncertainty and model adjustments (Table 9A-4, Appendix 9A, Shoreham USAR). The Shoreham Cycle 1 fuel loading uses three (3) enrichments. Of the 560 fuel assemblies in the core, 340 bundles have the ,

highest bundle average U-235 enrichment of 2.19 wt. %, 144 l bundles of 1.76 wt. %-and 76 remaining bundles uses natural

'.( )- . uranium.

l l

,If the six inch natural uranium segments at the top and bottom of the fuel are excluded, the average enrichment of a 2.19 wt. % bundle becomes 2.33 wt. %. Using this enrichment and linearly extrapolating the reactivity vs. U-235 enrichment results given in Figure 9A-5 of Appendix 9A, Shoreham USAR, the reactivity difference between the SFR design enrichment of 3.1 wt. % and the current maxiumum i loading enrichment of-2.33 wt. % 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 i 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 applicable.

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 decrease in reactivity ( -0.002 in k) mainly due to the l-offsetting contributions from the formation of Sm-149 and the slight depletion of the burnable Gd poison in the fuel bundles.

In light of the large reactivity margin described previously (k O. 0.85), no additional credit will be claimed here.

9-17 l

, . . '7 >

'\,;s.

j SHOREHAM DSAR 93 'Jy,ALUATION OF SPENT FUEL POOL MAKEUP REQUIREMENTS An analysis was performed to determine the rate of water loss .

from the spent fuel pool through evaporation under the worst case

.are scenario The following conservative assumptions l

used described below.to maximize in the analysis the calculated pool .

evaporation rates

1) ,The spent fuel pool temperature is 110*F. t L 2) The ambient temperature above the spent fuel pool is  :

conservatively assumed to have zero relative humidity.

3) The reactor building air flow exists due to normal i

ventilation system operation to maximize evaporation.

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

' level is alarmed in the control room and alarm response procefures exist to provide appropriate operator action.

l l

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L 9-18 L

L l t' _ _ . . _ _ - _ _ _ . .. . -_-

O o-coS- O^- '

' TABLE 9.2.1-1 SERVICE WATER SYSTEN COMPONENT DESIGN DATA i

! Nominal Capacity Neunber of C-- - rar.ts l Each Utilized in l Component Quantity (ops) Defueled_ Condition I Racctor Building Service Water Pumps 4 8,600 1 Turbine Building Service Water Pumps 3 8,000 . 1 i

! ROcctor Building Subsystem Components I

Reactor Building Service Water Strainers 4 250 1 l

Diesel Generator Jacket Coolers 3 700 --

Residual Heat Removal Heat Exchangers 2 8,000 --

Reactor Building Closed Loop Cooling 2 6,370 --

Water Heat Exchangers

' Reactor Building Standby Ventilation 4 525 1 System Chiller Condensers 2

i Main Chilled Water System Chiller 3 1,500 --

Condensers 1 400 Drywell Cooling Booster Heat Exchangers 2 1,460 --

1 of 2 ,

i i i

O ,m & _ o *

' TABLE 9.2.1-1 i SERVICE WATER SYSTEM COMPONENT DESIGN DATA (Cont'd.)

4 Nominal Capacity Number of Components  !

Each Utilized in i component Quantity (qpe) Defueled Condition q l Turbine Building Subsystem Components:

i Turbine Building Service Water Strainers 2 420 1 s .

Circ Water Pump Bearing Cooling 4 6

[

l l Fish Retention Pool 1 185 1" l

Turbine Building Closed Loop Cooling 2 14,200 1 Water Heat Exchangers Vccoum Priming Pumps Seal Water Coolers 3 100 -- t l

.)

l l

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i 2 of 2 ,

4 4 j

r. )

[' TABLE 9.2.7-1 J L. 1 LIST OF COOLERS CAPABLE OF BEING SERVICED BY TURBINE BUILDING CLOSED LOOP COOLING WKTEE~3YEYEM j i

Component Coolers Quantity ,

Hydrogen coolers 4 -i Excitor alternator cooler 1 Generator leads cooler 1 Generator stator winding coolers 2 ,

EHC coolers .

2 Air'cempressors 3 ,

Condensate pump motor thrust bearing coolers 2 Condensate booster pump lube oil coolers 4 Sample temperature bath coolers 2  :

() Condensate air removal pump oil coolers 2 Condensate Air removal pump sealing water heat 2 exchangers Reactor feed pump turbine oil coolers 4 Main turbine lube oil coolers 2 Offgas vent coolers 2 Waste evaporator overhead condenser 1 Regenerant evaporator overhead condenser 1 Waste evaporator distillate cooler 1 Regenerant evaporator distil3 ate cooler 1 Waste evaporator bottoms cooler 1 Regenerant evaporator bottoms cooler 1 Mechanical seal cooler heat exchangers on various > 10 radwaste process pumps

1

~ .  ;

SHORDDM DSAR h-V) t TABIZ 9.3.2-1 PROCESS SAMPLING SYS'11!M  ;

Type of ,

Saniple i Descriptien of Sarnple h RADMIAS'IE SYSTIM Regenerant Liquid and CC, CpH, Sanple regenerant ,

Evaporator Food Tanks Grab liquid evaporator Recirculation Line feed tanks for process data Discharoe Waste Sanple CC, Grab Sanple discharge Tanks Tecirculation Line, waste sanple tank for process data Waste. Collector Tanks CC, Grab Sanple waste Recirculation Line collector tank for

l. process data

! Floor Drain Collector CC, Grab Sanple floor drain l Tanks Becirculation Line collector tank for process data l Pectnery Sanple Tanks CC, Grab Sanple floor drain Recirculation Lirn collector tank for '

process data Radwaste Domineralizer Grab Demineralizer outlet efficiency Radwaste Filter Effluent Grab Filter efficiency Final Discharue Sanpling Grab Process data, prior Point to discharge Floor Drain Filter Effluent Grab Process data Laundry Drain Tanks Grab Process data Discharoe MA1 TUP DDIINERALIZER SYSTEM .

Individual Domineralizer CC, Grab, Demineralizer Effluents IX efficiency and makeup water quality D CC Caustic concentration "ilute caustic

• " for O

~ - - - . - - . ,-.

.} ?

'.fm SHORCHPM DEAR

-( ") . TABLE 9.3.2-1 (Cont'd) i 1

PROCESS SAMPLING SYS'IEM i

Type of ,

D3scription of Sanole Sa!wls M l Dilute Acid for CC Acid conoantration Booeneration i Waste Regeneraticm CpH pH of nonradioactive Neutralizing Tank and regeneration wastes, .

Acid and Caustic Waste prio to discharge W Return PUEL POOL CLEAWP SYSTM Domineralizar Inlet CC, Grab Indication of fuel pool water quality Domineralizer Outlet CC, Grab Domineralizer efficiency l AUXILIARY BOILER SYSTD4 i l

, Auxiliary Boiler Blowdown Grab Water quality, boiler  !

' performance l l

Auxiliary Boiler Feed Grab Water quality data Pmps Discharge Auxiliary Boiler Condensate Grab Condensate quality l l

Auxiliary Boiler Steam Grab Boiler performance data ,

1 i

l Hot Water for Heating Grab Water quality data i 1 COOLING WATER SYSTD4S )

l Turbine Building closed cc Cooling water quality l l Icop Cooling Water Heat i

! Exchanger Discharge Note: OC - Continuous Conductivity bbnitoring l Grab - Grab Sanple ,

l IX - Intentittent Silica Monitoring CpH - Continuous pH Monitoring O -

L l

l

O _A o

1RBEE 9.5.9-1 e y g a m (6)

Container Oper. Max Est. Tark Max Ehergy Release (2)*

Design Press. Press. (1) No. Voltane if Ruptured (ft itst10 .j Gas (PSIG) (PSIG) (PSIG Contain._ (ft3) each One Tanit All Tanks Iecetion Carbon Dicacide(3)

Fire Fivixction 363I *I 300 341 1 320 230.8 230.8 Yard Halon 1301 l

-1 Fire Fxvixct. ion Reactor Bldg.

Reactor Shutdbun 2650 600970'F 1250 2 0.74 .002 .004 Fire Fruixction TSC Computers (3)

Above Floor 600 360- 600 2 4 .25(5) .50(5) T9C Bldg.

Below Floor 1000 600 1000 2 1.8 .14(5) .28(5)

Instrinnent &

1 Service Receivers 150 125 145 3 415 11.8 35.4 Tiarbine Bldg.

Control Room i Breathable Air System A 10000 2400 4000 10 1.73 1.17 11.67 1brbine Bldg.

B 5000 2216 3693 5 0.30 0.16 0.80 i

l SRV Accumlator 145 95 115 11 0.205 0.005 0.051 Primary Cont.

d i l

l

! Id2 l

SIEIWIWWE r

'DeEE 9.5.9-1 (Oant'd)

S10f812 0F GAS (MER PIESSENE Container Oper. Max Est. Tank feet Ehergy D=Imame(2)6 Design Press. Press. (1) 10 0 . Dbime if Ruptureti (ft 1kst10 Gas (PSIG) (PSIG) (PSIG Contain. (ft3) endt One Tank All Tanks Iecetjan

. Nitrogen Hot Water Heating 10000 2600 4000 2 1.73 1.04 2.08 Stuttane Bldg.

Glym1 Heating 10000 2600 4000 2 1.73 1.04 2.08 Turbine Bldg.

Propane

Aant. Bir Ignition 490 204 220 2 85 1.25 2.51 Yard l

Notes (1) Safety valve set point.

(2) Reversible arliahatic expensim fra nuorimum container pressure.

(3) Stored as liquified gas.

(4) Maximaan working peessure.

(5) Max energy release calculated on basis of 5/3 operating press being equal to nex press. This does not significantly alter the max. energy relcese rumber.

(6) Variable quantity and type of welding gas tanks not listed. ^

Type and quantity is variable hamari on maintenance requia e.

2 of . 2

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

E l

BRID6E TR AVEL y 4

374.00* A EF.Ih8IDE LlWER(BI'- 6") __

i. . . . . ;grap, a <..>

n . . . _rj l l

JlMEY TRAVEL FgAgMg '

/  ;

( T Y P.)

V l

- = ... = = =-  !

8 Ia  :

s n u u = =  !

, 'I O 8

. 's Ts a

u u s

u o . ,,,,,,,,, .

CONTROL ROD AREA (99)

AND CASK STORASE q .. ..- -

AND LCADING AAEA r -

1 NOTES 8 2176 FUEL STOR AGE POSITIONS 144 CONTROL A00 STORA9E POSITIONS ,

, FIGURE 9.1.2-1 HIGH DENSITY RACK PLACEMENT IN SPENT FUEL POOL O SHOREHAM NUCLEAR POWER STATION DEFUELED SAFETY ANALYSIS REPORT

---d - --'*-ww' - --

• --~wM'e'

i SHORERAM DSAR CHAPTER 10 STEAM AND POWER CONVERSION SYSTEM 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 N 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 IIA.

10.4 OTHER FEATURES OF STEAM & POWER CONVERSION SYSTEM 10.4.1 Condenser The purpose for which the condenser was built no longer exists.

The components of this system as described in the USAR will not I be used in the defueled condition.

10.4.2 Main Condenser Air Removal System l The purpose for which the main condenser air removal system was built no longer exists. The components of this system as described in the USAR are not required in the defueled condition.

l 10.4.3 Steam Seal System 1

The purpose for which the steam seal system was built no longer exists. The components of this system as described in the USAR l are not required in the defueled condition.

I

' - I l 10-1 l

I i -

SHOREHAM DSAR 10.4.4 Turbine Bypass System  ;

The purpose for which the turbine bypass system was built no i 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 was built to ASME III cc2 criteria. As the. function of the i bypass system no longer exists in the defueled condition, the bypass system is reclassified Q.A. Category IIA. ,

10.4.5 Circulating Water System The purpose for which the circulating water system was built no longer exists. The components of this system as described in the t

USAR will not be required in the defueled condition. The only exception is that the circulating water discharge system will be

'used to provide dilution capacity for elimination of liquid radwaste and SPDES limits on chlorine and suspended solids to the Long Island Sound. .

10.4.6 Condensate _Demineralizer System t.(

. Since there is no fuel in the Reactor and no Reactor steam produced, there is no need for the Condensate Demineralizer  ;

System. .This system will be protected. However, the Acid and l Caustic Storage Tanks (lN52-TK-035 and -TK-036) will remain operable to provide regeneration chemicals for the continued operation of the Demineralizer and Makeup Water System (P31).

l l

I The Chemical Waste Sump (lN52-TK-ll3) will remain operable as a pathway for further treatment of non-radioactive regenerant waste from the Demineralizer and Makeup Water System.

10.4.7 condensate and Feedwater 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 cci is considered O.A.

Category IIA while in the defueled condition. Inservice inspection according to ASME XI need not be performed while in the defueled condition.

l 10-2

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SHOREHAM DSAR  ;

-(

CHAPTER 11 RADIOACTIVE WASTE MANAGEMENT  ;

11.1 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 i 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 and the suppression pool are processed to Radwaste). However, these should be minor and temporary occurrences. Sources related to the decontamination L

(, 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. ,

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.

l b. Gaseous 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 It is expected that materials which were located in the reactor

! vessel during low power testing (eg, control rods, TIPS, IRMs, I and the like) have been activated to some extent. With the exception of some portions of the liquid radwaste system (10 mrem /hr maximum), dose rates outside of plant systems are very g

low, less than 0.5 mrem /hr. These low dose rates are indicative of a low deposition of sources within plant systems.

! 11-1

v  :

() SHOREHAN DSAR l

There may be a minor amount of activation source material L deposited within plant systems. However, the level of this f 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.

REFERENCES General Updated Safety Analysis Report . (USAR) Shoreham Nuclear Power I Station Revision 1, December 1987.

1. " Semiannual Radioactive Effluent Release Report - First and l Second Quarter 1989", transmitted by letter SNRC-1619, -

8/29/89.

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.

T4 . . Gamma Spectrometer Scan of Floor Drain Collector Tanks, Waste Collector Tanks, and Recovery Sample Tanks, 6/15/89, t -Memorandum M.'Ma to T. Gillett.

I 11.2 RADIOACTIVE LIQUID WASTE SYSTEM i 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 Fuel Pool, the sources, quantity and activity of the radiosctive waste are greatly diminished. Certain portions of the i Radioactive Liquid Waste System are not required.

11.2.2 System Descriptions The estimated influent to the radwaste system is reduced from 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 Feed Tanks and Pumps.

The Waste Evaporator portion of the Floor Drain Subsystem is not required.

O The Phase Separator System serving the RWCU System is not required unless RWCU is required if reactor is layed up wet.

(. 11-2

() SHORERAM DSAR 11.2.2.1 Summary l

This section is no longer applicable since most of the waste i streams would no longer exist.

I 11.2.2.2 Low conductivity waste subsystem 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 Domineralizer System, Drywell Equipment Drain System and the Phase Separator Tanks (unless the reactor is layed up wet).

11.2.2.3 High Conductivity Waste Subsystem l

l 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 Domineralizer System. The Waste Evaporator will not be utilized to process this waste. Floor drain influents will be i

(} processed through the Floor Drain Filters.

11.2.2.4 Regenerant Chemical Subsystem

! In this system the only equipment still required are the ghemical j yaste gump, 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 l

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.

11.2.3.4 Liquid Radwaste Equipment Quality Group Classification This Section remains as presented in the USAR.

11-3

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i  !

i SHOREHAM DSAR i

11.2.3.4.1 Conditions and Assumptigng This accident (raised in USAR Section 11.2.3.4) postulates the simultaneous failure of the liquid radwaste system tanks in or ,

associated with the radwaste building. These tanks hold the i radioactivity and potentially radioactive liquid waste from the floor drains, equipment drains, nonradioactive chemical wastes, l and processed liquid affluents. The tanks (and their capacities)  :

that are assumed to fail are:

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

2. Floor drain tanks: Two at 25,000 gal each (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 outside the radwaste building contents are insignificant 1y radioactive) ,

() 5. Discharge waste sample tanks:

(located outside the radwaste building)

Two at 25,000 gal each j

6. Spent resin tanks one at 4,700 gal (Section 11.5)

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

11.2.3.4.2 Accident Description The accident description can be considered as described in Section 11.2.3.4.2 of the USAR.

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

1. A conservative airborne partition factor of 1.0E-03 is assumed for all isotopic activities listed in DSAR Table 11.1-1, with the exception of Tritium (H-3), for which it is assumed that all the activity evolves.

2. Ground release atmospheric dispersion factors are assumed, as given in USAR Table 15.1-3, for the EAB.

3. The breathing rate of persons offsite is assumed to be O 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).

11-4

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. i 1

SHOREHAM DSAR

)

11.2.3.4.4 Results and Consequences The doses resulting from the analysis described above are as follows:  ;

Dose, millirem Whole body Beta Maximum. ,

Source Gamma

• Skin Organ ** .

Spent Rosin 1.8E-05 2.7E-06 1.3E-03 ,

Tank ,

Radwaste Filte,rs 1.2E-07 1.7E-08 8.3E-06 ,

l Discharge Sample 3.1E-08 1.4E-08 7.7E-06 Tanks Totals 1.8E-05 2.8E-06 1.3E-03 1

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

• • Teen is the limiting age group, and lung is the l

limiting organ

(

The consequences of the above postulated accident are clearly very low. These projected doses are far below those which justify Quality Group D, non-seismic  ;

qualification of radwaste equipment (i.e., 500 mrem whole body, or its equivalent to parts of the body), in Reg.

Guide 1.26, Rev. 1, and Reg. Guide 1.29, Rev. 1.

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

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.

11.2.4 Operating Procedures

. Operating procedures including administrative control of liquid  ;

- () radwaste releases are as described in the USAR.

11-5

(} SHOREHAM DSAR 11.2.5 Performance Tests Performance tests of equipment are as described in the USAR, except for activity reduction factors (DF), which are no longer applicable. Only equipment that remains in operation will be periodically tested.

11.2.6 Estimated Releases Liquid effluent releases are expected to be minimal with the fuel in the spent fuel pool. This is based on the fact that during the period from June 1988 through May 1989, only one release had an isotopic concentration greater than LLD.

The quantity of the' annual release of contaminated liquids is conservatively estimated by noting that the discharge volume from I SNPS is approximately 5,000,000 gallons per year. Assuming the effluent concentration is consistently equal to that found in the i l one sample above LLD (7.83E-08 uCi/cc of Co-60, from DSAR Table l i

11.1-1), the estimated release ist l 1.5E-03 Ci/yr of Co-60 i

() 11.2.7 Release Points The description contained under this heading in the latest revision'of Shoreham USAR remains unchanged. Refer to USAR for l

information on this subject. l 11.2.8 Dilution Factors Under the plant's present condition, service water or circulating t

water will be used, if necessary, for dilution so that the discharged effluent concentration in the Long Island Sound will not exceed that preLcribed in 10CFR20, Appendix B, Table II, j Column 2.

Treated radioactive effluents are collected in the discharge sample tanks. The filled tank is sampled, and then discharged at a maximum rate of 150 gpm for a period of approximately 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. If necessary, the treated effluent is diluted with about 8000 gpm of service water prior to discharge into the sound.

Thus, if necessary a dilution factor of approximately 50 may be t 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 l 11.2.9 Offsite doses due to liquid releases are expected to be minimal, as discussed in DSAR Section 11.2.6. An estimate of the yearly l

l 11-6 l

l SHORERAM DSAR j

[]

\/

l dose is conservatively obtained by assuming each batch liquid release contains the maximum batch activity concentration of 7.83E-08 uCi/cc of-Co-60, and the release volume is approximately 5,000,000 gallons per year. Assuming no dilution, the resulting doses are as follows:

Whole Body 0.166 mrom (adult)

GI-LLI 1.43 mrom (adult)

Liver 0.074 mrom (adult) i As noted in Section 11.2.8, service water dilution remains available as necessary.

11.3 GASEOUS WASTE SYSTEM 1

11.3.1 Design Obiectives -

l l 1 I With the fuel in the Spent Fuel Pool, the radioactive gaseous i waste system is no longer required to )

i

1. meet either 10CFR20 or 10CFR50 Appendix I limits or l 1

l 2. ensure plant operability or availability.

I 11.3.2 System Descriptions With the fuel in the Spent Fuel Pool, and negligible amounts of radioactive halogens in the fuel, the radioactive waste sources described no longer apply, and the systems necessary to process them are not required.

Normal ventilation will be maintained in the Radwaste and Reactor Buildings with discharge through the station ventilation exhaust duct.

11.3.3 System Design The process offgas system, which is the system described in USAR Sections 11.3.3, 11.3.4 and 11.3.5, is not required with the fuel in the Spent Fuel Pool.

l 11.3.4 Operating Procedures 11.3.5 Performance Tests 11.3.6 Estimated Releases j

In the plant's present state, no releases of radioactive gaseous effluents are anticipated. This is evidenced by the fact that j

since the plant achieved initial criticality in 1985, there have been no recorded releases documented in the Semi-Annual Radiological Effluents Reports.

11-7

a SHOREHAM DSAR 11.3'.7 Release Points ,

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

i

'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 Dosgg There will be no~ expected offsite doses because no releases of j radioactive gaseous effluents are anticipated under the plant's  !

present defueled state.

I L

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 pathways do exist on loss of secondary

()- containment.

11.4 PROCESS AND EFFLUENT RADIATION MONITORING SYSTEM

(

The description contained under this heading in USAR only apply to those monitoring systems described in DSAR Section 12.3.4. .

Refer to the USAR for further information. The changes to the USAR relating to the Radiation Monitoring System for the defueled l condition are described in DSAR Section 12.3.4.

Sampling for halogens is not needed in the defueled condition.

-11.5 SOLID WASTE SYSTEM 11'.5.1 Design Obiectives

The description contained under this heading in the latest l

revision of the USAR remains unchanged as it is used to develop l the basic, design criteria of the plant. '

L 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 will continue to be generated, and must be shipped. Also Dry Active Waste (DAW) will l O- continue to be generated, and must be shipped. The volume of ,

both will be significantly less than that given in the USAR. j i

I I

1 11-8 )

SHOREHAM 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 i handled by the overhead crane. l l

It should be noted that waste will be generated from the Spent  !

Resin Tank, the Radwaste Filter, and the Floor Drain Filter as  !

described in Section 11.2 and transferred directly into HIC's.  !

The HIC's are then transported by the transfer carts out of their J cubicles to be handled by the overhead crane for dewatering. j Tables 11.5.1-1B and 11.5.1.-2 thru 5 of the USAR are superseded by DSAR Table 11.1-1. )

11.5.2 _ystem S Input: Source Terms The actual radwaste source terms in the plant's defueled )

condition are as follows:

The combined activity concentration in the spent resin tank, radwaste filters, and the floor drain filter is assumed to equal

() the maximum in the most recent solid weste 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 Eguipment Description 11.5.3.1 General The only equipment remaining in use in this system is-as follows:

4,700 Gallon Spent Resin Tank For the defueled condition, this receives backwashed resin and filter media from the Radwaste Demineralizer and the Fuel Pool Cleanup Demineralizer and Filters (and Phase Separators if the RWCU System remains operable). (This tank is also discussed in '

Section 11.2. It is included here since it is a direct feed to the Solidification system.)

The spent resin pump transfers the spent resin to HICs which are set on the Radwaste floor or in the pits in the floor. The HIC's are then dewatered by portable air-operated diaphragm pumps which draw suction from specially designed piping internals in the

() HIC's. When convenient, HIC's may be dewatered while in the fill U cubicles.

11-9

i SHOREHAM DSAR A Baler This equipment is furnished to compress miscellaneous dry active waste (DAW) into 55 gallon drums.

L 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.

i No other equipment in this section of the USAR is required.

11.5.3.2 Wet wastes The first paragraph of this Section of the USAR no longer applies. The second paragraph remains applicable. ,

11.5.3.3 Dry wastes This Section of the USAR is applicable, as some DAW will continue to be generated, 11.5.3.4 Irradiated Reactor Components This Section of the USAR still applies.

11.5.3.5 operating Procedures This section of the USAR no longer applies except that:

1. SRT waste can be transferred into a high integrity container (HIC) where it can be dewatered by the in-house dewatering system to Federal and burial site limits. Ultimately, this waste will be shipped to burial 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.

11-10

I i

SHOREHAM DSAR i

11.5.3.6 Instrumentation i 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 l A conservative expected estimated volume of waste in HICs and i carbon steel liners is 1,000 cubic feet per year buried volume. )

See DSAR Table 11.1-1 for activities. l This statement and Table together supersede Table 11.1-1A of the USAR.

DAW volume is conservatively estimated to be 1,000 cubic feet per l year, buried volume. The DAW activity is negligible. l l

11.5.5 Packaging 1 The description contained under this heading in the latest l revision of the USAR remains unchanged. Refer to the USAR for I O

! information on this subject.

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

11.5.7 Shipment The description contained under this heading in the latest l revision of the USAR remains unchanged. Refer to the USAR for i information on this subject.  !

11.6 OFFSITE RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM l

DISCUSSION j 1

The objectives of SNPS' Offsite Radiological Environmental l Monitoring Program [REMP) are to identify and measure plant generated radioactivity in the environment and to calculate the potential dose to the surrounding population. SNPS' REMP is designed to comply with the Plant's Offsite Dose Calculation Manual (ODCM) and NRC Regulatory Guide 4.15. REMP data is I acquired by sampling various media in the environment and then j analyzing these samples for radioisotopes; Tables 11.6.3-1 and O

11.6.3-2 detail the REMP sampling / analyzing program. Since REMP l resulta vary for each sample and location, several sampling locations were selected for each medium using available 11-11

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1 l SHOREHAM DSAR i

L 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'1ocations 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 implementsSection IV.B.2 of Appendix I to 10CFR Part 50, -

by verifying that measured concentrations of radioactive L

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 intervals: 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 O' background levels of environmental activity around SNPS.

l 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 ef fective 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 L

products were obtained from the terrestrial environment. Direct radiation was measured using thermoluminescent dosimeters (TLDs).

l The range of analyres for each sample weres. gamma spectrometry,

' Sr-89 and Sr-90; I-131; H-3, gross beta, direct radiation and l noble gases. Under Post-Defuel REMP, several of the above sampling locations and/or range of analyses are discontinued.

The current Post-Defuel REMP program is outlined in' Tables L

11.6.3-1 & 11.6.3-2.

Preoperational REMP began in February 1977 and continued through 1984, although the official Preoperational REMP period; i.e. the time frame against which the data base from Operational REMP was compared, occurred during 1983 and 1984. The Operational REMP began on February 15, 1985 when initial criticality was achieved.

Except for reactor operator training programs which required the O reactor to operate at '0.0% power' (during 1988), SNPS has not generated radioisotopes since the last 5.0% power test, completed on June 6, 1987. Comparisons between the above two phases of 11-12 ,

4  !

i SHORERAM DSAR ,

REMP were documented in each Semiannual Radiological Effluent l Release 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, virtually all iodines and gaseous effluents have

- decayed away. Consequently, the surveillance requirements for SNPS' Post-Defuel REMP were reduced to below the operational level.

Justification for Reducing REMP to Post-Defuel Surveillance Levels.

L Pursuant to Reg Guide 4.1, once the initial core of the licensee has reached the point (in time) of maximum burnup, and the t

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 may be reduced to normal

() Tech Spec requirements. Since (as of August 9, 1989) the core has been in the spent fuel pool, the initial core has " exceeded" the point of maximum burnup. i 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 the reduced surveillance ,

program will be performed in two steps. Step one reduces Operational REMP to the level mandated when SNPS was to become z operational. Step two reduces the surveillance program further, to the revised requirements corresponding to the defueled L 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 preoperational REMP.

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 O 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' license. The sampling points not required by the license are 11-13

i L

l

.O - SHOREHAM DSAR l )  ;

4) Rain Water; and l

1) Gamer

2) Aquatic Plants; 5) Noble Gases. I i 3) Aquatic Sediment; l Justification for reducing REMP to the revised requirements  !

(after the core was defueled) is given based on the above j j

information; i.e., the measured environmental impact due to-5.0%

power testing was comparable to that of preoperational REMP, and l as of August 9, 1989, the core was. removed from the reactor j l pressure vessel. SNPS' last 5.0% power test was completed on (

j' . June 6, 1987, and per Ref. 9, with the exception of I-129 and u Kr-85 (4.0 aci and 1560.0 C1, respectively), all iodines and l

gaseous effluents have since decayed away. In addition, radwaste '

i system activities are quite low (listed in DSAR Sections 11.1 &

12.2). As a result, the only remaining radioisotopes (and their release pathways) for which REMP is applicable are:

Isotopefs) Source Effluent Pathway

1) Kr-85 Spent Fuel Gaseous

2) Solubles and- Radwasta Gaseous and Liquid Particulates

() SNPS8 Post-Defuel REMP Surveillance Program Outline Reduce from 36 to 18 locations

1) DIRECT RADIATION:

Quarterly Surveillance Frequency ,

l

2) AQUATIC '

a. Aquatic Plants and - Delete, not required Beach Sediments I b. Fish, Surface Water - Retain, may be impacted and Invertebrates from liquid release path to L.I. Sound )

I Perform Semiannual surveillances as i available l l 3) AIRBORNE' I

a. Iodine - Delete, insignificant quantity l'

b. Particulates, Noble - Retain, Kr-85, particu-Gas and Gross Beta lates and solubles still exist. I l

Quarterly Surveillance Frequency 1 l

l O l

11-14

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SHOREHAM DSAR f

4) TERRESTRIAL

a. Precipitation, Soil, - Delete, not Tech Spec

' and Game required l b. Potable Water - Delete, well water not impacted by discharges to ,

L.I. Sound  :

c. Milk, Food products - Retain, long lived particulates Quarterly Surveillance for Milk, l Annually for Food

SUMMARY

/ CONCLUSION

1) Examination of the radiological impact to REMP locations which are to be eliminated -- From 1983 (preoperational RCMP) 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, SNPS' core was transferred to the spent  !

fuel pool; thus, the initial core has reached maximum burnup. l Per Regulatory Guide 4.1, if the above two conditions are 3) met, then the cperational phase of REMP may be reduced to the requirements that were written when SNPS was to be operated l as designed. i i

4) The post-operational REMP surveillance program may be reduced to the requirements as delineated in DSAR Chapter 16, developed after SNPS' core was transferred to the spent fusi pool, becaures

! a) Criticality will not be reestablished at SNPS. As of August 9, 1989, no additional fission / activation products will be f generated; b) SNPS' last 5.0% power test was completed on June 6, 1987, which means that with the exception of I-129 and Kr-85, all l remaining gaseous effluents have decayed away; and -l J

c) the only 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 l effluents are being or could be processed. l l

O l

11-15

/" SHORERAM DSAR b)

L 11.6.1 Obiectives of REMP 11.6.1.1 Preoperational REM _P l

The objectives of the Preoperational REMP were:

! 1. To identify and determine baseline radiological characteristics in the environment around SNPS (these background levels were then compared with data collected ,

6uring actual plant operation);

l 2. To assure that the n.edia being sampled and analyzed are l sensitive to fluctuations in the radiological characteristics

! of the environs af. SNPS, and to assure that REMP will be responsive to radioactive discharges from SNPS (i.e., to identify indicator locations and critical pathways);

3. To provide effective monitoring of critical pathways of radiological effluents to unrestricted areas; and l 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.

The yearF 1983 and 1984 served as the official preoperational period, as stipulated in Reference 8. All data collected during 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 revealed minute quantities of radioactive fallout from i

- the October 1980 atmospheric nuclear weapons test by' the People's l Republic of China during 1980 and 1981, in addition to  ;

radioactivity remaining from two decades of atmospheric testing. l Airborne particulate samples registered an increase in gross beta levels, along with identifying the gamma emitting isotopes Er-95, Nb-95, Ru-103 and Ce-141. Also in 1983 and 1984, REMP sampling I identified low levels of iodins-131 in Port Jefferson Harbor area i 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 normal background radiation found in this environment. The atmospheric

/~' environment was sampled for airborne particulate matter, iodine, I and noble gases. All airborne radiciodine analyses were below detectable levels. In addition, milk, potable water, game, food products, beach sediments and rain water were sampled. The 11-16

1.y _ __

s.

SHOREHAM DSAR results obtained from the analyses of these samples were typical l l of the radioactivity values usually associated with samples of

these types. All radioiodine analyses of milk were below detectable levels. Direct radiation levels were relatively low, r, 'and approximately the same at all locations. No unusual i radiological characteristics were observed in the environs of '

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. ,

l 11.6.1.2 Operational REMP The objectives of Operational REMP were:

1) Identify and measure plant-related radioactivity in the environment for the calculation of potential doses to the public.

2). Identify excessive radionuclida concentrations of limited duration, so that appropriate action may be taken.

I

3) Determine the long-term variation in radionuclide concentration, or

4) determine the effects of plant effluents on the environment. j f 5) Comply with regulatory requirements and provide records to l l

- document compliance.

6) Comply with the REMP requirements as outlined previously. l Operational REMP used the Preoperational data base to identify b plant-contributed radiation, and to evaluate the possible effects of radioactive effluents on the environment. The Preoperational and Operational phases of REMP were designed to comply with i Regulatory Guide 4.15 (5) and the associated Branch Technical I 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 I increases from the accident at the Soviet Union's Chernobyl i 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

.O during 1985, and DSAR Table 11.6.1-4 presents a comparison of preoperatonal and operational REMP data from 1983 through 1988.

11-17 l

l

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

t

.s' SHOREHAM DSAR 11.6.1.3 Post-Operational REMP The objectives of Post-Defuel and Operational REMP are identical. '

r ' Differences in the execution of Post-Defuel REMP account for both L 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 L

The exposure-pathways for liquid effluents are p 1. External exposure from radionuclides in water; end

[ 2. Ingestion of fikh. and shellfish containing radiottuclides.

The concentrations of radionuclides expected to be released to the service water are listed in Section 11.2. Dilution of-these concentrations in Long Inland Sound is discussed in Se'ction 11.2.8.

USAR Section 11.6.2.1 contains detailed discussions about the A projected doses-from various liquid pathways. With the updated

. \d source terms at. 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 11.2'.9 for dose calculations.

I ~31.6.2.2 Gaseous Effluent Pathways The exposure pathways for gaseous effluents are:

1) Submersion in a cloud of noble gas;

2) Drinxing 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 beta radiation. Doses from gaseous effluent pathways are summarized in USAR Table 11~6.2-3. Computational methods are discussed in Section 11.6.2.3.

A dairy survey is performed annually to determine the location of any milking animal within a 5-mile radius of SNPS. When a milking cow or goat is found, annual doses are calculated using either current meteorological or activity release dat-, .r

() accordance with the methods specified in the Shoreham Offsite Dose Calculation Manual.

11.6.2.3 Dose Computational Methods 11-18

iu pulgu u w t

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

USAR (Section ll.6.2.3.1)' continues to apply.

11.6.2.3.2 Gaseous Effluent Pathways The discussion contained in the latest version'of the Shoreham USAR (Section 11.6.2.3.2) continues to apply.-

11.6.3 Sampline Media, Locations, and Frequency Typical Post-Operational REMP sampling locations and frequency are given-in Table ll.6.3-1. These locations are described-in Table 11.6.3-2 and kre shown in Figures'11.6.3-1 and -2. By virtue of the liquid and gaseous effluents from the plant, REMP is divided up into four distinct categories: atmospheric,.

-terrestrial, aquatic and direct radiation. Sampling media, locations, and frequencies-are discussed in the following sections.

11.6.3.1 Sampling Media

? 11.6.3.1.1 Aquatic Environment The' aquatic environment is, examined by analyzing samples of: 1)

Surface water; 2) Fish; and' Invertebrates. Surface water samples are taken in May and' October using a Niskin Bottle. The samples are placed in new polyethylene bottles following three rinses with'the sample medium prior to collection. When available samples of Winter Flounder, Pseudopleuronectes americanus, Windowpane, Scophthalmus _aquosus,. Sea Robin, Prionotus spp, Little Skate, Raia erinacea, Blackfish, Tautog

,onitis and Summer Flounder, Paralichthys dentatus are taken by trawl, sealed in plastic bags,_ frozen, and shipped to the analytical laboratory for' analysis.

When available, invertebrate samples of American Lobster, Homarus americanus, Squid, 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, frozen and shipped to the laboratory for analysis. Blue Mussels Mytilus edulis are collected by hand along jetties and soft-shell clams,11a arenaria from Wading River are shelled and sealed in plastic bags, frozen and shipped to the analytical laboratory.

?

11-19

s SHOREHAM DSAR

,D u

'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 samples (approximately 1 cfm). The samplers used l are equipped with vacuum recorders for sample volume correction J

L and to indicate sample validity and maintenance problems when E 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 L 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 l indicate the duration of time the sample is taken, Air samples are collected quarterly at St. Joseph's Villa and Lanalyzed for noble gases (Krypton-85). The samples are collected ,

using a modified low pressure air compressor. An interim holding i tank is evacuated to 20 in. Hg. Outside air is drawn into the h

- (,/ interim holder and then transferred to a sample tank for y transport to the laboratory for analysis.

1 11.6.3.1.3 Terrestrial Environment l l

l The terrestrial environment is examined by analyzing samples of L, milk and food products. When available, milk samples are L collected quarterly, except during the pasture season (May l- through October) when the sampling is increased to monthly. Milk samples are prepared for shipment in accordance with the l instruction of the laboratory performing the analysis. Food -

products consisting of vegetables and fruit are collected from area farm stands and shipped fresh to the laboratory.

1 11.6.3.1.4 Direct Radiation 1 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 l 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 1 locations are listed in Table 11.6.3-1). The units are then l packaged and shipped to the laboratory for analysis.

l

(

l 11-20

y .

4 . 'SHOREHAM DSAR .)

A.,_)7'( : \

11.6.3.2 Sampling Locations and Frequency

' 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 11.6.1) 11'.6.5 ' Data Analysis, Presentation and Interpretation The discussion contained in the lat'est 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 i The discussion contained in the latest version of the Shoreham L

USAR (Section 11.6.6) continues to apply.

I REFERENCES In Section 11.6 l 1) Regulatory Guide 4.1 " Programs for Monitoring Radioactivity l -

in the Environs of Nuclear Power Plants"

[ 2) ~ Not Used I) 3 Not Used' 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 Technical Specifications 3/4.12 Radiological Environmental Monitoring Table 3.12.1-1 "REMP" l 3/4.12.1 Monitoring Program

7) Not Used 8)- SNPS' Operational REMP Annual Reports: January 1, to December 31, 1983, 1984, 1985, 1986, 1987, & 1988 issued by Nuclear Engineering and Environmental Engineering Departments of l LILCO.

9) C-RPD-476, Rev. O, 10/21/88, "SNPS Core Thermal Power After Shutdown" 11-21

w gV N SHOREHAM DSAR TABLE 11.1-1 Radwaste Sources Greater than LLD Spent Resin Tank, Radwaste Filter, & Floor Drain Fil'ter The. activity concentration is assumed to equal-the maximum in the most recent HIC shipment (Nov-Dec 1988) and is (From Reference 2): ,

. .y -

j Activity Isotope Concentration, uCi/cc. t of Activity

• Cr-51 9.84E-04 58.46%

• Mn ' -

'2.17E-05 1.29%

• Fe-55 4.19E-04 24.88%

L *Co-57 7.92E-07 0.054-Co-58 6.43E-06 0.38%

Co-60 1.09E-04 6.51%

, '*Fe-59 4.57E-05 2.71%

• Ni-63 6.41E-06 0.38%

• Sb-124 3.25E-06 0.19%

,y' y; *2n-65 1.89E-05 1.12%

'-: H-3 6.21E-06 0.37%

• C-14 3.94E-07 0.02%

• Sr-90 1.69E-07 0.01%  !

• Zr-95 1.52E-05 0.91%

• Nb-95 2.55E-05 1.51%

0.00%

• Tc-99 4.79E-09

• I-129 7.32E-10 0.00% ,

• Cs-137 -1.34E-06 0.08% '

l '

• Ce-144 2.95E-06' O.18%

• Pu-241 , '1.59E-05 0.95%

l>

Discharge Waste Sample Tanks  ;

The activity concentration in these tanks is assumed to equal the  ;

j -,- maximum. concentration measured in the past 12 months preceding May 1989 (from Ref. 3):

[

n Activity L Isotope Concentration, uCi/cc  % of Activity E

Co-60 7.83E-08 100.0%

b 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 1

() concentrations less than LLD.

• Calculated based on generic scaling factor.

11-22

' mate 11.6.1.-4 -

nreparison Of Gmadonal - Prsryprational REMP Data

( Gmai.icnal RIDIP ) (- Preoperational N .-) ,

SAMPIE TYPE Unit / Isotope _ _

1988 1987 1986 1985 1984 1983 Ibtable Water pCi/l (H-3) '240 - 41C 140 - 450 _ 130 - 420 150'- 290 _ 120 - 540 _70 - 220 Game pCi/Kg(Cs-137) _76.7 - 9270 35.1 - 6490 54 - 3230 992 - 4330_ 641 - 5340 34.0 -'6310 Direct neem / Mnth 2.3 - 5.2 2.8 - 6.9 1.9 - 5.7 3.0 - 6.2 2.7 - 6.9 2.3 - 5.7 RarH ation Std Mnth Qtr 2.7 - 4.8 2.9 - 5.0 2.9 - 4.9 2.8 - 5.5 3.1 - 6.2 2.8 - 5.4 Air: Gross Beta [x1.0E-31 5.0 - 44.0 4.0 - 32.0 5.0 - 360 6 - 47 4.2.- 61. 5 - 54 Particulate Sr-90 pCi/m2 3 x 1.E-3 LT 0.8 LT 0.8 0.11 - 0.27 LT 0.8 LT 0.07 1.3 - 1.4

, Iodine-131 pCi/m2 3 x 1.E-3 LT 10.0 LT 10.0 35 - 1230** LT 10.0 LT 10.0 LT 30.0 ,

Aquatic pCi/Kg (Sr-90) LT 1.0 LT 1.0 LT 1.0 6.8.- 27.

• 33. LT 20.0 Plants g/Kg (Cs-137) LT 6.0

• 85.5 ~ '* 47.9

• 45. 69.7 - 140. 36 - 55 pr.1/1 (Sr-90) 0.76 - 6.00 0.61 - 5.70 0.98 .13.0 0.86 - 4.60 0.69 5.3 0.9 - 7.7, Milk pCi/l (Cr-137) 6.00 - 14.8 5.90 - 11.5 7.0 - 8.9

• 4.4 9.6 - 14 12.9 - 14.1  !

pCi/l (I-131) LT 0.20 LT 0.20 2.1 - 4.8 LT 0.20 LT 0.20 la Food pCi/Kg (I-131) LT 4.0 LT 4.0 LT 4.0 LT 4.0 LT 3.0 1R Products (wet) (Cs-137) LT 5.0 LT 5.0

• 12.2 LT 5.0 LT 5.0

• 24.7- i i

• Ranges are not given sin only one data point contained an identified isotope.

• • Evidenoc of Clcudyl Mamt.

11-23 r

^

~

e A O- '

ThBEE 11.6.1.-4 (Cont'd)-

wrison of G :irmal - Prwn=3tional MMP Data _

)- (- Preoper=*irmal MMP -)

( Operational MMP 1985 ~1984 1983 1988 1987 1986 SAMPIE TYPE Unit / Isotope _

• 5.6 LT 1.0 LT 0.9 -

• 86 pCi/Kg (Sr-90) LT 1.0 LT 1.0 Aquatic m M' LT 5.0 34.8 - 36.2 m .E __

Inv=. e rate (wet) (Ce-137) LT 1,0

• 3.3 LT 2.0 LT 1.0 LT 1.0 LT 1.0 Beach pCi/Kg (Sr-90) LT 9.0 m LT 8.0 LT 8.0 LT 8.0 Sediment _( dry) (Cs-137)_ LT 8.0 _

LT 2.0

• 1.7 LT 3.0 LT 2.0 LT 2.0 LT 2.0 Aquatic pCi/Kg'(Sr-90) ~

m

• 21.7 LT 10.0

• 30.4 44.2 - 49.4 Sediment (dry) (Cs-137) LT 10.0 180 - 280 100 - 220 50 - 270 90 - 280 pCi/1 (n-3)

• 190 170 - 430 Surface Water __

LT 0.5 -LT 0.6 LT 0.7 LT 0.5 LT 0.5 LT 0.5 Fish pCi/Kg (Sr-90) 8.4'- 21.4 8.8 - 19.1 7.11 - 17.5 11.0 - 25.8 10.2 - 13.8_ 7.70 - 17.4 pCi/Kg (Cs-137) 90 - 270 120 - 190 140 - 320 80 - 970 pCi/l (H-3) 130 - 490 130 - 410 Rain Water E M M m M 1.40 - 12.4 pCi/1 (Cs-137) 30 18 - 49 24 - 45 21 - 48 24 - 40 Noble Gases 3

pCi/m (Kr-85) 28 - 44 LT 40.0 LT 11.0 _ LT 11.0 _ LT 34.0 pCi/m3 (Xe-133) LT 11.0 _ LT 11.0 _

• Ranges are not given since only one data point cxmtained an idesttified isotope.

'11-24

c f SIKEEHAM DShR l

SHDREHRM DSAR 'mB12 11.6.3-1 i (f if o

. Posteatiaani Nialaaical' Erwironmental Monitoring Progran (IGMP)

j l Media .

Fmling Icontions Sanoling Frequency Analysis-b Direct (1) 181,2A2,381,481,5S2, Quarterly Ganna Exposure L Radiation' 6S2,7A2,8A3,981,10A1, 11A1,12A1,1363,1482,- ,;

L E 1581,1682,*5E2,*6El l Fish ~and (2) . 3C1, 14C1, *13G2 Semi-annually. Gama-isotopic Invertebrates or when in season Sr-89/90 Fruits, (3) 8B1, 6B21, *12H1 At tine of Annual Gama-isotopic J

_and vegetables Harvest H -

I. ' Airborne (4), 6S2,2A2,381,781,*11G1 Quarterly Gross-Beta L Particulates- Gama-isotopic '

Sr-89/90 1 t

Milk (5) 13B1,*10F1, or *8G2 Quarterly. During Gamna-isotopic.

Pasture Season, Sr-89/90 Monthly Surface Water _3C1 or 14C1, and *13G2 Semiannual Gama-isotopic H-3, Sr-89/90 Grab Sanple Krypton-85 1

Noble Gamma 14S2 Monthly .

h (*) Designates Qantrol Iocations (1)' Eighteen monitoring stations (16 indicator and 2 control) are used. One indicatnr location is positioned in each meteorological sector near the site boundary. One dosinater or continuously neasuring dose rate instrunant is .

_placed at each location.

(2) At each Indicator location, one sanple of each camarcially and recreaHnnally inportant species. One sanplo of same species in control  ;

, location. i l

! (3) Sanple three different kinds of broad leafy vegetables grown nearest to two l' indicator locations - having highest predicted average ground level D/Q i

!' (when milk sanples not available). Also take one sanple of same leafy '

l. vegetatim grown nearest to Control location.

l i

(4) '1hree sanples (near SNPS), one from each of the three Meteorological sectors L having the largest annually averaged ground-Invel D/Q, are taken. One sanple (near a comunity) also having the highest calculated annually averaged ground-level D/Q is taken. Establish one Control Iocation in the  !

least prevalent wind direction.

(5) Indicator sanples fra milking animals having highest rotential dose.

Sanple within 5 km distance (preferably), within 5 to 8 km where doses are calculated to exceed 1 mren/yr (second choice) or from 8 to 17 km. Control i location is 15 to 30 km from SNPS and in the least prevalent wind direction. l 1

11-25  !

p 1 1.-

4

, .e M

k.) -

SHORERAM DEAR

+

SHORDRM DSAR ThB211.6.3-2 REMP SAMPLING I4XATIONS i

DESIGRTION ,

IDChTICE in 181 Beadi mast of intake, 0.3 mile [N]

2A2 West and of Creek Ibad, 0.2 mile [ft4E]

3C1 Fish and Invertebrates, outfall Area, 2.9 miles [NE]

381 Site hug, 0.1 mile [NE]

i 4S1 Site Boundary, 0.1 mile [I!NE]

, *5E2 Calvepon, 4.5 miles [E]

SS2 Site Boundary, 0.1 mile [E]

6B21 Condezella's Farm Stand, 1.8 miles [ESE]

• 6El LIIDO ROW, 4.8 miles [ESE]

6S2 Site Boundary, 0.1 mile [ESE]

7A2- North Country Road, 0.7 mile (SE]

7B1 Overhill Ibad,1.4 miles [SE]

BA3 North Country Road, 0.6 mile [SSE]

8B1' Iocal Farm, 1.2 miles [SSE]

• 8G2 Dairy -(Cow),10.8 miles [SSE) 9S1 Service Road SNPS, 0.2 mile [S) 10A1 North Country Road, 0.3 mile [SSW)

• 10F1 Goat Farm, 9.2 miles [SSW)~

11A1 Site Boundary, 0.3' mile [SW)

• 11G1 MacArthur Substation, 16.6 miles [SW]

12A1 Meteorological 'Ibwer, 0.9 mile [WSW)

• 12H1 Background Farm, 26 miles [WSW]

13B1 Goat Farm, 1.9 miles [W)

• 13G2 Fish and Invertebrates, Background,13.2 miles (W) 13S3 Site Boundary, 0.2 mile [W) 14C1 Fish and Invertebrates, Outfall Area, 2.1 miles [WNW]

14S2 St. Joseph's Villa, 0.4 miles (WNW]

15S1 Beach west of intake, 0.3 mile [NW]

16S2 Site Boundary, 0.3 mile [NNN]

• Designates Control Iocations e 11-26

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M f FIGtNtE 11.8.3-2 Q' OFF 58TE SAMPLING LOCATIONS RADIOLOGICAL ENVIRONMENTAL '  !

o s to MONtTORfMG PROGRAM

. SHORENAM NUCLEAR POWER STATION scaLt-wtLas '

DEFUELED SAFETY ANALYSIS REPORT

=

E SHOREHAM DSAR

.(

CHAPTER 12 RADIATION PROTECTION 12.1 ASSURING THAT OPERATIONAL RADIATION EXPOSURES ARE AS LOW AS REASONABLY ACHIEVABLE The Shoreham ALARA Program, the intent of which is to maintain operational radiation exposures (ORES) to levels as low as is reasonably achievable (ALARA), is described in the USAR, Section 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. l (Reference USAR Section 12.1.1.3.2, Review of Modification of Operations. The basis of this change is a reorganization of the Nuclear Engineering Support Organization.)

4 B) Several of the items listed under Section 12.1.2, Design Considerations, are no longer applicable or have been revised. Specifically, charcoal has been removed from the Radwaste vent filters j

(considerations 10 & 12), I the' Radiation Monitoring System is now discussed in DSAR Section-12.3 (consideration 11),

demin water hose stations are located on the radwaste building floor (consideration 7),

1 shielded radwaste shipping casks and remote handling l techniques are not generally used (considerations 15 and 16),  !

i and the discussion of gaseous radwaste sources in i consideration 15 is no longer relevant, as the offgas system is shut down. ,

1 (Reference USAR Section 12.1.2, Design Considerations. The justification for lack of charcoal is the fact that as per l DSAR Sections 11.1 and 12.2, Shoreham does not possess a meaningful quantity of radiciodines. The low levels of radioactivity in the solid radwaste do not justify using shielded casks and remote handling.)

12-1 l

,((~} SHOREHAM DSAR v

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 against radiation, prior to their entry into.the RCA.

(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 requirements for visitors to Shoreham).

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 approved dosimetry devices upon their entry to the radiologically-controlled area (see Section 12.5.2.1, Access Control. Rather, only individuals working on a Radiation

' Work Permit (RWP) may have approved-dosimetry issued. 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 dosimeters (TLDs) in general access areas of the RCA.

(Reference USAR Section 12.1.3.1.1, Use of Individual.

Personnel Monitoring Devices. This change is justified by i .the low dose rates seen presently at Shoreham, and by the  ;

very low historical man-rem data in Section 12.5 of the  !

l DSAR.)

It should be noted that the Shoreham station's original physical design for radiation protection (e.g., shield walls, 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 described in the USAR (e.g, L the precautions for high dose rate jobs -- in excess of 100  !

mrem /hr) will be maintained.

12.2 RADIATION SOURCES 12.2.1 Contained Sources Fuel Sources The Shoreham reactor core has undergone three periods of low power (0-5%) testing over the past four years. The low power t

l-tests are summarized below:

12-2

SHOREHAM DSAR Specific Burnup Power Test Period Duration MWD /MT Range %

7/7-10/7/85 ~T3'3ays 27.8 UIO - 3.3 8/5-8/30/86 26 days 13.8 0.0 - 4.0 5/26-6/6/87 12 days 6.7 0.0 - 3.5 Total 1173 The detailed profiles of the above three low power test periods have been input to the ORIGEN2 (Reference 1) burnup code, along '

with the physical characteristics of the reactor fuel and bundle structural elements. Results of this analysis (Reference 2) are given in Table 12.2-1. The activities correspond to two years decay after the last.burnup period, and reflect total core inventories for those isotopes with greater than 10 curies. ,

ThatthesourcestbengthsgiveninTable12.2.-1arereasonable is evidenced by measurements taken during the defueling of the reactor. Dose rate measurements were 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 (as a function of bundle burnup)_ were calculated from the source terms presented in Table 12.2-1,-using the point kernel code QADMOD (Reference 5),

and the results 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 average gives-assurance that the calculated source terms in Table 12.2-1 are reasonably accurate, j_ As can be seen from the Table 12.2-1 only.long-lived isotopes remain from the original actinides and fission / activation products created, along with their equilibrium daughters. By far the.most radiologically significant( 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 due to the Ba-137m photon (Reference 3). For dose assessment of accidental gaseous releases (e.g. , a postulated fuel handling accident) , only Kr-85 L is meaningful (Reference 4).

12.2.2 Airborne Radioactive Material Sources The statements below 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.

Reactor Building L There is no significant source of airborne activity assumed to

! exist in the reactor building in the plant's present defueled condition.

}

l 12-3

.r 2 244 l

ys '

(l

SHOREHAM DSAR L' W Turbine Building

^

There :Lis no source of airborne activity assumed to exist in the turbine building.

Radwaste Building ,

There_is no significant source of airborne activity assumed to exist-in the radwaste building. (

Further discussion regarding airborne activity'is provided in sections 11.1 and-12.4. ,

REFERENCES

-t ,

General r 1.

L Updated Safety Analysis Report (USAR) Shoreham Nuclear Power L Station Revision 1, December 1987. r t

1. ORIGEN2, Isotope Generation and Depletion Code, ORNL CCC-371, 7/80.

2. LILCO calculation C-RPD-476,-rev. 0, 10/21/88.

3. LILCO calculation C-RPD-530, rev. 0, 05/19/89.

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

5. QADMOD-G, Point Kernel Shielding. Code, ORNL CCC-396, 12/79.

12.3. RADIATION PROTECTION DESIGN FEATURES L

12.3.1 Facility Design Features The description contained under this heading in the latest

~

revision of the Shoreham USAR remains unchanged as it is used to develop the basic design criteria of the plant. Refer to the

-USAR for information on this subject. However, the defueled condition, with low activity levels, some design features are not necessarily utilized as described in the USAR. For example, liquid filters in the radwaste system do not usually require portable shielding or remote backwashing. Also, the radiation zone designations shown on USAR Figures 12.3.1-1 through ~35 are not applicable for the plant's present condition.

12.3.2 Shie_1 ding 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.

12-4 y

3 i) 'SHOREHAM DSAR l

-12.3.3 Iventilation' 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.

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 TheLprocess 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.

Normally, nonradioactive systems that may become significantly contaminated by leaks from radioactive systems are monitored continually to_ ensure that no condition hazardous to the l operating personnel or to the general.public develops. For

-effluent streams that discharge to the. environs, sample points are located downstream of the last point of-possible radioactive fluid addition ~to=the effluent being monitored.

All monitors in the proceso 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 release points from the station that require radiation monitors: the station ventilation exhaust, the liquid radwaste effluent, and the reactor building salt water drain tank.

Area Radiation and Airborne Radioactivity Monitoring Instrumentation This section contains a description of the area and airborne radiation monitoring systems. All channels have local readout by means of a-log-ratemeter and local audible and visual alarms.

Each channel has high radiation and fail alarms which are annunciated locally and in the main control room. The area monitors are provided with an audio and visual alert and high radiation alarms. Monitors are placed in areas where personnel 12-5

- - - - -~ . _- -

c'l +

V ,

j SHOREHAM DSAR ,

-normally have access and where there is a possibly that radiation levels could become significant.

All airborne monitors are'offline monitors and are 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 accessible area.

' Airborne radiation monitoring is provided where potentially radioactive sources exists. Each of these monitors is provided ,

with,an isokinetic nozzle which is sized to obtain a

-representative air sample at the normal flow in the ventilation duct from which the sample is taken.

Table 12.3.4B lists the airborne monitors, and Table 12.3.4C L - lists the area monitors.

[

Radiation Monitoring System, Computers The RMS is equipped with redundant computers powered from U.P.S.

i 1297-INV-005/TSC Black Battery /69 kV primary feed. These units provide continual surveillance for all airborne, area, process, p)'

I s_ and effluent radiation monitors. Communication with the computer is through keyboard equipped CRT displays in the main control room, the health physics office, the process computer room, and the technical support center.

l Inservice Inspection, Calibration, and Maintenance l

'The operabill'ty of each channel of the area and airborne RMS is l checked periodically from the main control room or manually at the monitor. Both systems are checked periodically or as specified-by the plant technical specifications.

! Calibration of all monitors iis normally conducted at an interval of 18 months. This calibration will allow indication in a low, l

mid, and high response range of each monitor.

12.4 DOSE ASSESSMENT 12.4.1 pesign Obiectives 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 reasonably achievable

( ALARA) under all conditions, including the defueled condition.

() Table 12.4-1 lists the six zone designations that were originally established, along,with the maximum allowable dose rates and estimated occupancy times for each area. With the plant in its l

present condition, with spent fuel stored underwater in the pool, there are no occupiable areas which are Zone III or higher.

ta 12-6

~

SHOREHAM DSAR

(~ - t A .

l

~12.4.2' Airborne Activity Jur area within the Shoreham' f acility is described as an " airborne i radioactivity l area" if the sum of the concentrations of all

- airborne radionuclides divided by their respective Maximum Permissible Concentrations (MPCs) (from 10CFR20, Appendix B,

" Table 1, column 1)-- escoeds 0.25. At Shoreham,-there are no 'l "airbornt t dioactivity areas" in the defueled condition. With o

'the fuel in the spent fuel pool, and insignificant quantities of radioactive material elsewhere (see Sections 11.1 and 12.2), it l is not expected that nirborne radioactivity areas will exist in l the future, unless systems which are currently anticipated to l remain closed are opened to the atmosphere. In this instance, j the radiation work permit procedure (see Section 12.5) will be applied'to assure there is no release of contamination into the L air.

With exposures reasonably expected to be much less than 2 MPCa-hrs per day and/or 10 MPCa-hrs per week, paragraph 103(a)

(3) of 10CFR20 indicates that exposure, and the-resulting internal doses, need not be included in the dose assessment to

l. With no " airborne radioactivity areas" postulated, L -individuals.

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 i 12. 5 ) '. Procedures are in place for taking appropriate actions,

• including investigation, when any positive whole body count l

L

-occurs.in excess of 1% of the maximum permissible organ burden (MPOB), or 1% of the maximum permissible body burden (MPBB).

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.

3) The recent collective station dose history at Shoreham is as l

L follows (TLD data collected in response to the requirements

.r~%. of 10CFR20.407):

L) 12-7 i l'

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

i SHOREHAM DSAR Time Period Dose, man-rem 1/1/86 - 6/30/86 0.562 7/1/86 - 12/31/86 3.123 1/1/87 - 6/30/87 0.341 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 Since February of 1987, when a change was made from R. S.

Landauer to Panasonic TLDs, doses have been insignificant, and ,

due almost entirely to small statistical fluctuations _rather than I 1

actual doses.

Based on the above statements, it is anticipated that-occupational dose at Shoreham will be essentially zero in the l'

L defueled condition. Doses will of course be measured, as indicated in the Health Physics Program, Section 12.5.

I 12.4.4 Offsite Dose Assessment There are no sources (eg, N-16) in the defueled condition which could lead to offsite direct doses, either by direct radiation or L

! "skyshine", based on the source terms presented in Sections 11.1 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 REMP, 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 materials in a manner consistent with Federal and State 1 regulations during all phases of operation, is described in l Section 12.5 of the USAR. The program is applicable in its l entirety to the defueled condition at Shoreham, with the  ;

-following exceptions:

A) Handling of new fuel is no longer applicable to Shoreham. l (Reference USAR Section 12.5.1.2, Personnel l Experience and Qualifications. The basis of this change is 1 that with the Settlement Agreement with New York State, no I new fuel will be brought onsite.) l B) _ The laundry facility does not contain an automated respirator l washer, unloading table for same, or <. respirator dryer.

Cleaning of respirators is done by hand methods when 1 i

O; 1 necessary. Respirator fitting may at some time in the future be moved from the Annex Building to another onsite location.

Protective clothing is to be cleaned either onsite or offsite, as conditions warrant.

12-8 l

SHOREHAM DSAR

?

('eference R 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) The monitoring station where personnel exiting from-the controlled area frisk themselves will no longer be in general use, unless justified by the overall workload in the controlled area. Similarly, the frisking station between the Turbine Building and the Control Room'will be removed.. In general, frisking will only be done at the exit of the specific iobsite, as controlled by the RWP.~

l l-(Reference USAR Section 12.5.2.1, Location of Equipment, .;

I_nstrumentation and Facilities, Section 12.5.3.3.1, Access Control, and Figure 12.5.3-1. - The basis of this change is the fact that there is no significant contamination within the controlled area, except perhaps at specific jobsites n (controlled by the RWP). This fact is confirmed by the l -

ongoing surveillance program.) '

_( f D) The numberc 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 t by the defueled plant's activities and number of personnel.

L (Reference USAR Section 12.5.2.2, Types of Detectors and Monitoring Instruments. Justification of this change is the reduced surveillance requirements and number of plant personnel.)

E) Radiation Work Permits are required for work under any of the l- following conditions:

l ~. Work in a posted radiation area.

2. Entry into a posted high radiation area.

l 3. Work in a posted contaminated area (see-Item F below).

4. Entry into airborne radioactivity areas.

l- 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.

{

F) Under the discussion of access control, add the definition of a contaminated area:

i 12-9

I SHOREHAM 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'

l. dpm/100 sq cm.

L

-(Reference =USAR Section 12.5.3.3.1, Access Control. The L

basis for this change is a modification to the station health physics procedures, as recommended by the Institute of l Nuclear Power Operations, in their document _INPO 85-001,

'rev.1.)

l G). Under the discussion of access control, the " secondary access l facility" no longer exists.

-(Reference USAR Section 12.5.3.3.1, Access Control. The basis for this change is that as of September 1, 1989, the l secondary access facility was taken out of service.)

H) The Corporate ALARA Review Committee (CARC) now administratively reports to the Assistant Vice President, Nuclear' Operations.

O (Reference USAR Section 12.5.3.3.4, Post-Operations Review.

See L The basis for this change is an organizational change.

L Chapter 13 of the DSAR for further details.)

l 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 ,

L to the low current source terms (Reference USAR Section l 12.5.3.3). However, the procedures and commitments remain in place in the extremely unlikely event that they should be required.

A V

12-10 l'

s.

SHOREHAM DSAR 4

TABLE 12.2-1 Fuel Source Terms ISOTOPE- CURIES' HALF-LIFE H-3 1.77E+02 1.23E+01 years Mn-54 3.36E+01 3.13E+02 days-Fe-55 8.06E+02 2.70E+00 years Co-60 5.64E+02 5.27E+00 years Ni-63 4.28E+01 1.00E+02 years Kr-85 1.56E+03 1.07E+01 years Sr-89 1.54E+01 5.05E+01 days Sr-90 , 1.37E+04 2.86E+01 years Y-90 1.37E+04 6.41E+01 hours Y-91 6.81E+01 5.85E+01 days Er-95 1.48E+02 6.40E+01 days Nb-95 3.49E+02- 3.51E+01 days Ru-106 5.98E+03 3.68E+02 days Rh-106 -5.98E+03 2.99E+01 seconds Sn-119m 3.30E+02 2.93E+02 days Eb-125 1.45E+03 2.77E+00 years Te-125m 3.53E+02 -5.80E+01 days

.Te-127 1.49E+01 9.35E+00 hours Te-127m 1.52E+01 1.09E+02 days Cs-134 1.33E+02 2.06E+00 years Cs-137 1.48E+04 3.02E+01 years Ba-137m 1.40E+04 2.55E+00 minutes Ce-144 3.55E+04 2.84E+02 days Pr-144 3.55E+04 1.73E+01 minutes '

Pr-144m 4.26E+02 7.20E+00 minutes Pm-147 2.95E+04 2.62E+00 years Sm-151 3.60E+02 9.00E+01 years Eu-154 1.18E+01 8.80E+00 years Eu-155 4.47E+01 4.96E+00 years U-234 1.02E+02 2.45E+05 years Th-234 3.38E+01 2.41E+01 days Pa-234m 3.38E+01 1.17E+00 minutes U-238 3.38E+01 4.47E+09 years Pu-239 2.77E+02 2.41E+04 years Pu-241 5.58E+01 1.44E+01 years L Total 1.76E+05

( f- Notet Only isotopes with activity greater than 10 curies are listed.

nv i

SHOREHAM DSAR

-p TABLE 12.2-2 Comparison of Measured (maximum) vs.

Calculated Spent Fuel Bundle Dose Rates

• Cell Burnup Dose' Rates, rem /hr Measured /

Date Number 'qwd/st** Measured Calculated Calculated 7/14/89 07-46 0.0045 0.4 0.4 1.00 7/19 01-32 0.0047 0.7 0.4 1.75 s 7/21' 45-34 0.0385 5.1 3.5 1.46 7/24 25-48 0.0421 5.5 3.9 1.41 7/26 45-10 0.0126 1.6 1.2 1.33 7/26 47-14 0.0082 0.9 0.8 1.13 7/26 15-44 0.0398 3.2 3.7 0.86

-7/28 43-30 0.0547 6.0 6.0 1.00 7/29 21-44 0.0516 5.3 5.6 0.95 7/31 09-34 0.0514 3.8 5.6 0.68 o 8/01 39-16 0.0607 6.0 6.6 0.91 L 8/01 39-32 0.0624- 6.0 6.8 0.88 1 1.11

-() average 1

• in water at 12 inches from fuel bundle

-** gwd/st =_ gigawatt-days /short ton' O

,' , ., t

, , .t r .-

SHOREHAM DSAR 4

TABLE 12.3.4A PROCESS AND EFFLUENT MONITORS PM 13 Liquid RW Discharge PM.21 Low Range Station Vent Monitor PM 29 Gas Reactor Building Vent PM 30 Part Reactor Building Vent PM 41 Part Station Vent PM 42 Gas Station Vent .

PM 55 Gas Radwaste Vent i PM 56 Part Radwaste Vent PM 79 Saltwater Drain Tank 1

I

1. T l

t i

y

l, '

f ( fy l my M'

+-

l.

Tis.E 12.3.4B IKrA RR milBrf AIO PROGSS RADIATEN MNIIORS y)

Rarge (2) Action Nomal Operatim Monitoring 1hw tim t_a-ati= Type of Mmitor Sensitivity Taken m Ala m Statim Vaneilarim Exhaust & nitor final r=1*aap for Dounstrema of the last Offline gas ' 10f101/cc Imestigate and all Im level gaseous point of activity Parti =1ata gases (Kr45) correct case of efflusts to the min high activity -

plant vent airstream r*1e=== rate ezr-d-

. IIE. im=*=mous tecinical ===e4H-tion dose rate lhuit.

Liquid Raduaste Monitor activity releaap Effluent pipe prior Offline Liquid 10 sci /cc Autantic closure of Effluent rate during plamed to discharge into the (Co-137) 11guld unste dischenge liquid weste discharge . cirmlating water systaa valve mh 10GR20 periods limits to unrestricted areas.

-6 1 Reactor hiilding Salt & nitor activity releaan of Downstrean of the Offline Liquid 10 tC1/cc Imestigate ad correct .

Water Drain Tank service water drained during collacri m tank (Cs-137) - c=ian of high activity maintenance r=1*aan rate hing ,

10GR20 limits to unrestricted aress Reactor hiilding Monitor all low level Effluent chet prior Offline gas IONtCi/cc L.. @ aul correct Venti 1= rim gaseous effluents in the to discharge into the Parti m1=ra gases 00M35) cause of hiah activity reactor b1dg ventilation station vent relemap rate +

duct Raduaste niilding &nitor all lw level Effluent duct prior . Offline gas 10 tCi/cc L;r4:e and correct '

Ventilatim gasecus effluents in the to disclerge into the Parti m1=ra gases (Kr451 cause of high activity raduaste b1dg. vene11= tim statim wnt relan== rate.

duct.

A F > w <gri / a'-

y e-+ - -.. __+W ___,_________.m________g _ ______,____2imew.-_ .g

SIDE!Het DSAR / v:-

7'-

11M E 12.3.48 IKTA RR DTIMNT #6 PRO SS RADIAIEN MNITURS _ . . _

mnitor(1)

R=4p (2) Action knitorira Phnetim Incatim Type ofPtmitor.. Sensitivity Talum m Alara Post-kcidmt .

Statim Ventilatim mnitor finsi releaae for . Dounstrema of the last. Offline gas 10 sci /cc N/A Exhant (Im Range) all gaseous efflu sts to point of activity Partindea gases M) :

the main plant vent air- . ,

4 stream during an accident 1

(1) One mit miess specified otherwise.

(2) Dynamic range is a mininam of 4 Apendpa above sensitivity. ,

1 .

_ _ _ - ___ _ _"_ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ = - -

___________________.___.______________.__2.2 _- -

_.__ __ _ _ _____m_. __.____e_ _ _ _ _ _ -- _

m , m h -

SMMM IEAR O: 1-0 ~

DEE 12.3.4C

.m m  :

Area muitor Alert /High Setpoints margie (area /hr) (srum/hr) i ne=H=

ID21-RE-001 n oor Drain sep Tank 5/100 0.1-1000 Reactor pids. el H

-010 Nel Pool Clearmp Ptaps 5/100 0.1-1000 Reactor Bldg. el 112 .

-012 ISel Pool Equi ==r* Area 5/100 0.1-1000 Reactor Bldg. el 150-9 l

-013 Contandnated Equip. Storage 5/100 0.1-1000 Raar+nr Bldg. el 150-9

-014 Puel Storage Pool 5/10 0.1-1000 Reactor Uldg. el 17H f

-015 Reactor Head Iru=i1=H= Storage 5/100 0.1-1000 Reactor Bldg. el 17M

-022 Omnistry Iduus=^wty Mezzanine 1/5 0.01-100 Hester Bay el 31-0

-024 Raduaste Bldg. Deccm*miination Area 5/100 0.1-10p' Radunste Bldg. el 154

-026 Storage Vaults for Costniners 10/100 1.0-10 Raa===te Bldg. el 154 i

-027 Seple Roan 20/100 0.1-1000 Ba&==ite Bldg. el 374

-028 Hoor Drain Filter Area 5/100 0.1-1000 Raa==ite Bldg. el 37-7 >

-029 Rakeste Filter and Waar=11- Area 5/100 0.1-1 Raa===te Bldg. el 374

-032 Raduaste Bldg. D==inar=14w Area 20/100 0.1-1 Raa==te Eldg. el 154  !

-033 Raduaste Bldg. Holst Area 5/100 0.1-1 Ra&==ite Eldg. el 5% ,

-035 Equipsent Drain Tank Area 5/100 0.1-1 Reactor Bldg. el H

-036 TIP Drive Roan 5/100 1.0-1 Reactor Bldg. el 78-7

-037 TIP Drive Area 5/100 0.1-1 Reactor Bldg. el 78-7 l -038 New hel Storage Area 2/10 0.1-1 Reettor_ Bldg. el 17M l -042 Radweste Bldg. Ptep Area 5/100 0.1-1 Raa==ite B1dg. el 154 l

l l

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

lSS t i

I SHORERAM DSAR o . .

l Table 12.4-1 RADIATION ZONES'  ;

(original plant design basis)

Maximum Estimated i Allowable Occupancy Eone . Dose Rate Time t D'signation e Zone Description (mrem /hr) (hr/wk) ]

I Unrestricted Area - less than Unlimited-Continuous Access 0.2 i i

II Unrestricted Area - less than 50 l Periodic Access 2 ,

> III Restricted Area - less than 20 l Controlled: Frequent- 5 ,

Access i-IV Radiation Area - less than 5 Controlled Infrequent 20 Access V Radiation Area - less than l' controlled Infrequent 100 Access VI High Radiation Area - greater than -

Not Normally Accessible 100 f

r

p i

l' SHOREHAM DSAR

>L /

CHAPTER 13 CONDUCT OF OPERATIONS 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 Figure 13.1.1-2. The Assistant Vice President, Nuclear Operations has assuraed all the' responsibilities of the Vice President, Nuclear Operations. The position of Vice-president, Nuclear Operations has been eliminated.

B) The Manager, Nuclear Quality Assurance Department (NQAD),

reports directly to the Assistant Vice President, Nuclear Operations but has maintained direct access to the President

,- of the Company as he deems necessary.

The Nuclear Engineering Department has been reorganized to

\'

C) reflect a reduced level of activity in the defueled I condition. Now entitled the Engineering and Technical l Support Organization, it reports to the Manager, Engineering

& Administrative Support. It no longer includes an Engineering Assurance function.

! D) The Safety Engineering and Reliability' organization within I

the NOAD is elimCiated. This includes the ISEG and Reliability Secticns.

E) The Director, Office of Training and the Manager, Nuclear Emergency Preparedness Division report to the Vice President, Corporate Services.

F) DSAR Figure 13.1.1-1 shows revised direction of executive ,

i responsibility.

i 13.1.2 Engineering And Administrative Support Organization .

The Engineering and Administrative Support Organization consists i of the following organizational units:

n s_-

13-1

t i

SHOREHAM-OSAR ,

t

'- Licensing, Nuclear Contracts & Material Conttols, Engineering &

Technical Support and Nuclear Financial Services. These units  ;

have as many staff specialists as required to support Shorehan.

Engineering & Administrative Support personnel provide expertise' i for supplementary support functions such as licensing and regulatory activities, including assessing evolving regulations, managing all nuclear litigation and evaluating regulatory  !

documents for impact on plant design. Other responsibilities cover cost control, estimating, budget and cost administration for-the Office of Nuclear, nuclear records management and administration of site clerical administrative personnel. The units are also responsible for nuclear contract development and administration, administration of site warehouses, spare parts and inventory control.

The Engineering & Technical Support Organization consists of the following units:

Plant Support, Engineering Administration & Services, Modification Support and Nuclear Analysis. Responsibilities include, (1) Systems, Mechanical & I&C Engineering, (2)

Procurement Support, (3) General administrative support for procedures, trainirs and document control, (4) Nuclear Analysis

[~. Support for the foliaring technical functions

Radiological

\ Engineering & Health Physics, Radiological Monitoring Program (REMP), and Engineering and Nuclear Fuels Engineering, (5)

Modification Engineering Unit which requests and coordinates  ;

implementation of station modifications, coordinates post-modification retesting, and return to service.

The Engineering & Technical Support Organization also coordinates

, work performed by off-site support Engineering which includes Corporate Engineering and outside contractors. Outside l

contractors include the original plant architect - engineer and q L the NSSS Vendor.

13.1.3 Operating Organization

The Shoreham Nuclear Power Station Organization, as shown in l Figure 13.1.1-2, consists of 4 Divisions

Operations, Maintenance, Radiological Controls & Security 13.1.3.1 Operations operations is responsible for complying with the rules and regulations of the governing regulatory agencies and the monitoring of the station performance. It is composed of l

N Operations, Operations Staff, System Engineering, Compliance, and Work Planning and Scheduling.

y 13-2

i i

i SHOREHAM DSAR 71

-- Operation activities of this unit primarily consist of the routine operation of the station systems and equipment.

The Systems Engineering Unit is part of Operations and its

! responsibilities include rapidly providing specialized, operationally oriented technical expertise in station systems and i equipment. Systems Engineering also performs support functions e for other operating organizations as appropriate. i The Operations Staff Unit provides station administrative support )

and assurance that the station is in compliance with the  ;

requirements of the Operating License. ]

The Operational Compliance Unit implements the station I surveillance programs and reviews surveillance activities to .

ensure compliance with the station's Technical Specifications. l The Work Planning and Scheduling Unit performs planning and scheduling associated with plant activities.

13.1.3.2 Maintenance .

Maintenance is responsible for maintaining the Station's mechanical, electrical, instrumentation, and computer systems.

/~' It is composed of the Instrument and Controls Maintenance, and Fire Protection and Safety.

l The Instrument and Control Unit is responsible for the calibration, maintenance, and testing of instruments and control 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.

Fire Protection and Safety is responsible for implementing the Plant Fire Protection Program and for coordinating the activities of the Fire Brigade and the Site Safety Committee. The 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.

13.1.3.3 Radiological Controls Radiological Controls is responsible for the protection of the public, station personnel, and the environment from the effects of exposure to radiation. It maintains the radiation doses of

() station personnel and the public as low as reasonably achievable 13-3 ,

l SHOREHAM DSAR (AIARA) and assures proper handling, processing, and disposal of radioactive materials. Radiological Controls consists of Health .

I Physics and Radiochemistry.

The Health Physics. Unit responsibilities include the preparation of Radiation Work Permits, performance of radiological surveillances, maintenance of personnel exposure records, ,

calibration and maintenance of fixed and portable radiation  !

detection instrumentation, and proper disposal of radioactive i material, and proper disposal of radioactive material. l The Radiochemistry Unit is responsible for activities such as detection and control of environmental releases, assessment of  ;

radiation doses to the public, and station chemical and 1 radiochemical activities. j 13.1.3.4 Security The responsibilities of Security are described in the Security  !

Plan.

13.1.4 Qualification 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 l I

l utilities in maintaining training programs that produce well-qualified, competent personnel to operate the nation's l nuclear power plants (INPO 88-001) .

In the defueled state, with the NRC operating license amended to remove operating authority, there is no requirement to maintain j

) accredited training programs since the plant is no longer  ;

licensed to operate.

l The Office of Training has non-nuclear training programs j available, developed via a " systematic approach to training" L method, which can be requested by the Shoreham plant management for training of operators, technicians, and mechanics.

The Office of Training procedures outline the methods to be used  !

O to analyze training needs, and to establish or conduct required i I

U training. The Office of Training staff will be qualified in accordance with the " Training and Qualification Program".

13-4

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SHORERAM DSAR f~/)

A  !

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 nuclear operations.

Equipment Operator: Field 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 systems required to be operational during the dafueled phase.

Control Technicians: Control technicians and computer technicians will be trained in accordance with the Control Technician training' program developed for power plant technicians. ,

Mechanics / Electricians: LILCO mechanics / electricians attend formal training as part of LILCO's maintenance training programs.

These programs qualify mechanics / electricians as apprentices with ,

journeyman qualifications available in the area of welding, rigging, machinery, electrical, and general maintenance skills.

j ,

The Shoreham maintenance force will be trained and qualified in

'! accordance with existing LILCO maintenance training programs.

• This program is not available for contract maintenance work forces; contractors would provide qualified mechanics and electricians.

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 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

(

1 No Change.

13-5 l

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,f SHORERAM DSAR

/4

( \ '

(/

i 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:

Membership; A' chairman or alternate chairman and four members or alternate members of the Plant Staff as designated by the chairman.

Alternate; only one alternate shall participate as a voting .

member in. ROC activities at any one time. {

Quorum; The Chairman or his designated alternate and two other members including alternates.

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 Assistant Vice '

President, Nuclear Operations, a mechanism for independently

/'_j') ascertaining-that activities related to the nuclear station (J are performed. safely and efficiently in accordance with company policies and-regulatory requirements.

The NRB is established and functional; its initial membership comprised LILCO and consultant persunne3.

Collectively, the membership has been selected to have the L

experience and capability to function effectively in the areas of responsibility as designated in license documents.

The objectives are to ensure that a representative decision -

is reached on each issue and that Assistant Vice President, Nuclear Operations is appropriately advised. The NRB membership is selected so that a majority of members are not directly responsibile for plant activities. All members, whether LILCO employees or consultants, are afforded equal voting status along with a defined route to advise the Assistant Vice President, Nuclear Operations, of 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.

l (""

/

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SHOREHAM DSAR  !

d("T .

The charter of the NRB reflects the consideration that NRD activities are not limited to items and functicus l that are designated as safety related. It is intended ,

L that NRL review and audit activities will also cover i 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.

2 Membership ]

The NRB will consist of the NRB Chairman and at least four permanent members. As a group, they will

! collectively have the competence required to review l L problems in the following areas: nuclear engineering, i chemistry and radiochemistry, radiological safety, l mechanical and electrical engineering, and QA practices. l The Chairman will be appointed by the Assistant Vice  ;

President, Nuclear Operations. The Chairman of the NRB i

l is responsible for appointing individuals to NRB g membership. Memberchip appointments are to be such that L t the collective membership includes the experience and q capability noted in the foregoing subsection. Membership appointments are subject to concurrence by the Assistant ,

Vice President, Nuclear Operations.

In the event a regular member is not able to' participate I

in NRB activities, designated alternates are authorized i to act in the place of the regular member. Any nominated 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 I necessary to obtain further scientific or technical assistance in carrying out its responsibility. Such individuals shall function as staff to the NRB, performing tasks and submitting reports as assigned by the action of the NRB.

Minimum qualifications of NRB 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.

h. Other members of the NRB and their designated alternates will be graduate engineers or equivalent and will have at least 3 years experience in the

! 13-7

s i

l i

SHOREHAM DSAR appropriately related scientific, technical, engineering, or power generation field. Members, or their designated alternates, may possess competence  !

in more than one specialty area. j

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 of 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 for the operation of the Shoreham Nuclear Power Station. A quorum shall be considered filled if conference telephone l communications are established with the requisite number of members or alternates at remote locations. No more than two alternates shall participate as voting members in.NRB activities at any meeting.

-.( f 3. Meeting Frequency

' l The NRB shall meet at least once per six months. l i

.Any member may request a special NRB meeting to consider I' a matter believed to involve a safety or radiological environmental problem. )

I

4. Records

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

numbered in sequence, and minutes of meetings shall be distributed to the President; the Assistant Vice l President, Nuclear Operations; and NRB members within 14 days following each meeting.

b. Reports of audits submitted to or conducted under the cognizance of the NRB, including recommendations of the NRB, shall be made in writing to the Assistant ,

Vice President Nuclear Operations; and to the l management positions responsible for the areas

! audited within 30 days after completion of the audit.

l 13-8 1

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f- t SHOREHAM 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 10CFR 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 CFR, Section 50.59. ,

c. Proposed tests or experiments that involve an i

unreviewed safety question as defined in 10 CFR, Section 50.59.

d. Proposed chat.ges to the Shoreham Technical i 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 safety significance,

f. Significant deviations from normal and expected performance of station equipment that affect nuclear safety.

g g. ALL REPORTABLE EVENTS l

l

h. All recognized indications of an unanticipated deficiency in some aspect of design or 9peration 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 encompass:

a. The conformance of station operation to provisions contained within the Shoreham Technical '

Specifications and applicable license conditions at O' least once per 12 months.

13-9

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T SHOREHAM DSAR (d

b. The performance, training, and qualifications of the entire station staff at least once per 12 months.  !

c. The results of actions taken to correct deficiences occurring in station equipment, structures, systems, or methods of operation that affect nuclear safety at .

least once per year.

d. The performance of activities required by the OA Program to meet the criteria of 10CFR50, Appendix B, at least onct 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 t i engineer (s) or an independent fire protection .

consultant. ,

f. Any other area of station operation considered appropriate by the NRB or the Assistant Vice i President, Nuclear Operations.

g g. The radiological environmental monitoring program and

' the results thereof at leest once per 12 months.

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

t.

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

j. The performance of activities required by the Quality Assurance Program for effluent and environmental monitoring at least once per 12. months.

7. Authority The NRB is organizationally responsible to the Assistant Vice President, Nuclear Operations.

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

Those items submitted to the NRB as described in Paragraphs 5(b) through 5(d) above, reviewed by, and I accepted by the NRB will be resolved as follows:

l 13-10 1

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[') SHORERAM DSAR i

\_/ l

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 NRB is of the opinion that approval of the NRC  !

is required, the Shoreham Nuclear Power Station staff, assisted by other LILCO nuclear organizations or by consultants, shall prepare a request for such i approval, including an appropriate safety analysis in l support of the request in accordance with approved l procedures. i If, in the course of any additional reviews of facility operations, the NRB determines that a variation from the l Technical Specifications or an unreviewed safety question  !

exists, the NRB shall immediately notify the Plant l Manager, who shall take the necessary steps to ensure nuclear safety.  !

() 13.4.3 Shoreham Independe,nt Safety Engineering Group The Shoreham Independent Safety Engineering Group (ISEG) is f eliminated for the DSAR Phase. The ISEG was required to be l

established by NUREG-0737, TMI Action Plan Requirements by each j applicant for an operating license. The ISEG was an independent 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 1 minor modifications will be performed. The principal function of the ISEG, to improve plant safety during operations, is no longer ,

applicable. The remaining activities are adquately covered under the LILCO Quality Assurance Program for Shoreham which will remain unchanged.

13.5 STATION PROCEDURES 13.5.1 Administrative Control

\ The description containe'd under this heading in the latest revision of the Shoreham USAR remains unchanged except that:

13-11

SHOREHAM DSAR

1. Safety-related station procedures shall be processed through the Review of Operations Committee (ROC) and Nuclear Quality i Assurance (NOA). l

2. The Plant Manager shall approve Station Administrative ,

Procedures, Security Plan Implementating Procedures, and  !

-Emergency Plan Implementing Procedures prior to ,

implementation.

3. Other Station Operating Procedures shall be approved by the appropriate Division Manager or by the Plant Manager prior to implementation.

.See DSAR Figure 13.5.1-1. Refer to the latest revision of the USAR for other information on this subject.

13.5.1.1 Normal 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.

13.5.1.2 Routine Maintenancey Repairs, and Fuel Handling l

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

I 13.5.1.3 Modifications The description contained uncer this heading in the latest revision of Shoreham USAR remains unchanged. Refer to the USAR ,

?

l for information on this subject.

13.5.2 Procedures 13.5.2.1 Operating Procedurea The description contained under this heading in the latest l

revision of the Shoreham USAR remains unchanged except that the General Operating Procedures now only describe integrated station operation. Startup cnd Shutdown are no longer pertinent.

Operating Procedures are not necessarily performed by, or under  :

the direction of, persons holding RO or SRO licenses.

13.5.2.2 Alarm Response 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.

13-12 ,

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

Sh0REHAM DSAR 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 in the latest revision of Shoreham USAR remains unchanged. Refer to the USAR for information on this subject.  ;

13.5.2.5 Instrument and control Systems 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.

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

j: 13.5.2.7 Shoreham Nuclear Power Station _ Emergency _ Preparedness

{

Plan

, The description contained under this heading in the latest

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

13.5.2.8 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.

i 13.5.2.9 Chemistry 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.

! 13.5.2.10 Reactor Engineering Procedures Procedures that describe the methods of nuclear performance and evaluation are originated, reviewed and approved in accordance i .with revised DSAR Figure 13.5.1-1.

l- 13.5.2.11 Plant Security Procedures I- The description contained under this heading in the latest l

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

13-13

i SHOREHAM DSAR f.

13.5.2.12 Radioactive Waste Management 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.

I 13.5.2.13 Temporary Procedures Tbedescriptioncontainedunderthisheadinginthelatest .

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

13.5.2.14 Temporary Changes To Approved Station Procedures

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

l

• 13.6 PLANT RECORDS The description contained under this heading in the latest

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

for information on this subject.  ;

l 13.7 INDUSTRIAL SECURITY The Security Plan, Training and Qualification Plan, and the

. Safeguards Contingency Plan for the Shoreham Nuclear Power Station have been submitted as separate documents. These documents are withheld from public. disclosure pursuant to 10CFR2.79(d), " Rules of Practice." The Security Plan and the L Safeguards Contingency Plan are also withheld from public j disclosure pursuant.to 10CFR73.21, " Requirements for the Protection of Safeguards Information."

L l

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SHOREHAM DSAR

_ TABLE 13.5.1 1 PROCEDURES PROVIDED FOR SHOREHAM NUCLEAR PCWER STATION A. . Administrative Procedures shall be provided to cover the j Iollowing types of administrative activites:  !

1. Authorities and Responsibilities for Safe Operation and Shutdown )

2. Equipment Control (e.g. , locking and tagging) l

3. Procedure Adherence and Temporary Change Method i

4. Procedure Review and Approval

5. Schedule for Surveillance Tests

6. Shift and Relief Turnover - Recall of Personnel

7. Log Entries and Record Retention

8. Bypass of Safety Functions and Jumper Control )

9. Operating Orders l

10. Special Orders I

11. Materials control l

12. Radiation Work Permits l

13. Access Control to Controlled Area l l

14. Personnel Training and Qualification ]

() B. Operating Procedures

1. General Operating Procedures have been provided to cover  !

the following Integrated Plant Operating Activities l

a. Surveillance.

l

2. System Operating Procedures chall describe Startup,

Normal Operating, and Shutdown for the designated l l system. Abnormal Operation, where required, shall be I contained in a section of the System Operating )

Procedure. 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 Station Service Transformer l L

d. Well Water System l

e. 4,160 V System l

f. 480 V System J I

g. Station Lighting Panels

h. 120 V ac Instrument Bus j

1. 120 V ac Reactor Protection System Bus

j. 120 V ac Uninterruptible Power Supply ,

k. 125 V de System t

1. Fuel Pool Cooling Reactor Building Normal Ventilation System (RBNVS)

m. I

n. Service Water

o. Radwaste (Liquid) 1 of 4 -

.f .

() SHOREHAM DSAR

_ TABLE 13.5.1-1 (Cont'd)

B. Operating Procedures (Cont'd.)

p. Radwaste (Solid)

q. Communications System

r. Condensate Transfer ,

s. Deluge _and Sprinkler System ,

t. Dominera11 red Water Transfer *

u. Equipment and Floor Drains

v. Fire Protection System

w. 'HVAC - Control Room l

x. HVAC - Turbine Building J y. HVAC - Radwaste Building  :'

z. Makeup Water Treatment '

aa. Station Air System ab. Smoke, Temperature, and Flame Detection System ac. Turbine Building Closed Loop Cooling System ad. CRAC Chilled Water >

~

4- 3. Emergency Procedures have been provided for combatting the following potential emergency conditions:

() a.

b.

Acts of Nature 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 ,

f 4. Abnormal Operation Procedures required to mitigate the

' consequences of the following abnormal conditions shall l

be contained in the appropriate System Operating Procedures (s):

a. None.

Note Procedures not designated as emergency procedures shall be incorporated in the Abnormal Performance section of the appropriate system or general operating procedures.

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()- SHOREHAM DSAR t

TABLE 13.5.1-1 (Cont'd) I<

Alarm Response Procedures (ARP)  ;

C.

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

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

2. Preventive and Corrective Maintenance of Safety Related l

l Equipment E. Instrument and Control Procedures shall be provided to cover the following instrumentation and control activities:

1. Measuring and Test Equipment

(

\

2.

3.

Protective Relaying Instrument Records

4. Surveillance Testing

5. Preventive Maintenance of Process Instrumentation ,

F. Fuel Handling Procedures shall be provided to cover the Io11owing Yuel handling activities:

1. Special Nuclear Materials Control and Accountability Procedures

2. Spent Fuel Handling and Shipment

3. Handling and Storage of Sealed and Unsealed Sources G. Health Physics Procedures shall be provided to cover the following radiation protection activities:

1. Dose Rate Radiation Surveys

2. Surface Radioactive Contamination Surveys

3. Personnel Contamination Survey

4. Personnel Decontamination

5. Areas and Equipment Decontamination

6. Monitoring for and Collecting and Recording of Occupational Radiation Exposure (ORE) data

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

8. Use of Protective Clothing and Respiratory Equipment

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TABLE 13.5.1-1 (Cont'd) i H. Defueled Energency Preparedness Implementing Procedures I TDEPIPs) 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 4

5. Surveys, Analyses, Sampling, Assessment, and Actions l

6. Personnel and Equipment Decontamination

7. Notifications

8. Re-entry and Recovery

9. Emergency Organization, Drills, and Training .

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TABLE 13.5.1-2 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 Nealth Physics Appropriate Section Head )

- Radiochemistry Appropriate Section Head  ;

Reactor Engineering Appropriate Section Head

'Ik ) Solid Radioactive Waste Appropriate Section Head Handling and Shipping Gaseous &nd Liquid Radioactive Appropriate Section Head Waste Effluent Control Fuel Handling Appropriate Section Head Surveillance Appropriate Section Head i Security Appropriate Section Head L

I l

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(' SHOREHAM DSAR

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TABLE 13.5.1-3 FORMAT FOR STATION PROCEDURES 1

• SP Number Revision Eff. Date ,

i Signature Date .

TPC NO Date Eff Date Expr S0ction Head Quality Control Div. Mgr. _ ,

Plcnt Mgr.

Signature or N/A TITLE l

( [0 PURPOSE A brief description of the purpose for which the procedure is intended j should be clearly stated. If the procedure is used to satisfy, in any

,part, a Technical Specification surveillance requirement, indicate the Technical Specification number here.

2.0 RESPONSIBILITY Indicate the person directly responsible for ensuring the proper implementation of the procedure.

3.0 DISCUSSION Provide a brief description of the applicable component, system, or task'in sufficient detail for a knowledgeable individual to perform the required function without direct supervision. Include a list of topics or a table of. contents generally describing the extent or scope of the procedure, with page location.

l

• For temporary procedures, SP Number assignment is TP XX.XXX.XX.

i O

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1 SHOREHAM DSAR TABLE 13.5.1-3 (Cont'd) 1 l

=4.0 PRECAUTIONS .

General precautions should be listed in this section before the ,

description of the actual procedure.

i Precautions should be established, as applicable, to alert the individual performing the task to those situations in which measures should be taken early or when care should be exercised to protect equipment and/or personnel. Precautionary notes applicable to specific steps in the procedure should be included prior to that step in the ,

main body of the procedure and should be clearly identified.

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.

.0 LIMITATIONS AND ACTIONS Limitations on the parameters being controlled and appropriate corrective measures to return the parameter to normal should be specified when applicable.

f 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 i

Step-by-step instructions in the degree of detail necessary for performing the required function or task should be provided. These shall be numbered sequentially.

Note 1: O?erating Procedures (Table 13.5.1-1, Sections B.2 and B.4) shall, as appropriate, be divided into two categories:

Normal Performance shall include step-by-step instructions to complete the required operation. Subcategories may include startup, routine operation at power, rotation of equipment,

l. and shutdown.

Abnormal Performance shall include instruction to recognize the existence of and to correct out-of-normal conditionr. that occur during the normal performance. Included may be a statement of the out-of-normal condition, including limits of parameters and/or alarm annunciator action.

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(f TABLE 13.5.1-3 (Cont'd)

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 i be referenced.

Note 3: Surveillance Procedures l J

The step-by-step instructions, with appropriate signoff or  :

checkoff provisions for.each step, shall'be provided to ensure the proper performance of the surveillance activity.

9.0 ACCEPTANCE CRITERIA Specific acceptance criteria against which the test results shall be judged for approval / disapproval must be stated clearly. Acceptance '

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 cppropriate for the type of device being tested. j 10.0 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, flow .

diagrams or other drawings, manufacturer's equipment manuals, other station procedures, and system descriptions.

12.0 APPENDICES Applicable appendices (in the form of checklists, data sheets, diagrams, etc.) should be included when necessary to support the proper implementation of the procedure.

O 3 of 6

s i

SHOREHAM DSAR TABLE 13.5.1-3 (Cont'd)

EVENT ORIENTED EMERGENCY PROCEDURE FORMAT ,

Submitted: SP Number (Section Head)

Approved: Revision (Operations Manager)

Effective Date TITLE *

• Should be worded to indicate the purpose of the procedurc. ,

1. SYMPTOMS: Symptoms should be included to aid in the

' identification of the emergency. This should include alarms, operating conditions, and ,

probable magnitudes of parameter changes. If a condition is peculiar only to the emergency under consideration, it should be listed

(~}

v fir.st.

2. AUTOMATIC ACTION: (Delete if not pertinent) +

3.. IMMEDIATE ACTION: These steps should specify immediate action for operation of controls or confirmation of automatic actions that are required to stop the degradation of conditions and to mitigate the consequences of degraded conditions.

l 4. SUBSEQUENT ACTION: Steps should be included to return the reactor to a normal shutdown period under abnormal or emergency conditions..

l

5. FINAL CONDITIONS: These steps should specify the documentation, authorizations, and plant conditions that must '

be completed prior to resumption of Normal ,

Operation, defined in 22.XXX.XX.

6. DISCUSSION: A brief explanation of the procedure.

This r,ection should contain background information, causes, effects, and other information that may assist in clarifying the procedure and analyzing symptoms, Or Note: Attempt to get 1, 2 and 3 on cover page of procedure to allow rapdi evaluation and action by the operator.

l 4 of 6

SHORERAM DSAR ,

TABLE 13.5.1-3'(Cont'd) ,

(-y T ,)- l SYMPTOM ORIENTED OPERATING EMERGENCY PROCEDURE FORMAT  !

• • TITLE Submitted l '. _

(Section Head) I 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 PURPOSE ,

(A brief description'of the purpose for which the procedure is intended should be clearly stated).  :

2.0 ENTRY CONDITIONS (This section should specify the plant conditions and/or plant L( ) 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).

4 O

5 of 6

~ ~

'!?

SHORERAM DSAR y

g') TABLE 13.5.1-3 (Cont'd)

'NJ A ALARM RESPONSE PROCEDURE (ARP)

FORMAT T

submitted: ARP i

(Section Head) (Windown Number)

., s - ,

Approved:

(Operations Ngr.) (Panel Number) '

a. (Panel Sub-Section)

-Effective Date Revision:

ALARM TITLE Instr. No. Set Point: Trip _

Reset U

CA_UjSE IMMEDIATE ACTION j List those conditions which List the immediate actions night have initiated the alarm; in order for each possible list the most probable cause first, cause.

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

Note la ALARM RESPONSE PROCEDURES - will include specific instructions to l mitigate the consequences of the condition indicated by the alarmed annunciator. Alarm Response Procedures should be filed in numerical sequence in Appendix I to Volume II of the Station 1 Operation Manual. '

REFERENCES l

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' v

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SELECTED REVIEWER l" COMMENTS E l 3:  :

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1 5AFETY NON 5AFETY '

RELATED APPROYAL ROUTE RELATED

$ ELECTION .

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ROC MEMBERS DIV. AND/OR .

W PLANT MANAGER  :

REVIEW AND COMMENT 2

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PLANT MANAGER '

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L FIGURE 13.5.1-1 L

PROCEDURE FLOW DI AGRAM h.

W SHOREHAM NUCLEAR POWER STATION DEFUELED SAFETY ANALYSIS REPORT

_ . . . _ _ _ _ _ _ _ _ - - _ _ . . _ - - - . , - . - . , _ . - . . . . - . _ , . . . . . , -n. . . , , . - ,

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CHAPTER 14-INITIAL TESTS AND OPERATIONS

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. This. chapter in not needed due to thedefuelsd condition of:the

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y 7V SHOREHAM DSAR CHAPTER 15

=a ACCIDENT ANALYSIS

6 15.1 GENERAL Analytical Objective

• Chapter 15 of SNPS USAR provides the results of analyses of the spectrum of' transient and accident events which are postulated to occur with the plant operating initially at up to maximum power.

The purpose of this analysis is to identify USAR transients and

-accidents that apply to the storage and handling of the low burnup fuel.

The analysis is based on the defueled condition of the plant, L 1.e., the fuel is removed from the core and is stored in the spent fuel pool. The total decay heat is approximately 550 watts, which'is small enough that it could be removed by passive cooling and would not require the fuel pool cooling system.

l- Normal and emergency makeups are discussed in Chapter 9.

As the reactor will not be operated and the fuel is not in the i

I)

' reactor, most-of the USAR Chapter 15 events cannot occur.  ;

Approach'to Safety Analysis l

E The, safety parameter evaluated for each transient of USAR Chapter l 15'is the Minimum Critical Power Ratio (MCPR) which is a measure o of. fuel cladding integrity. Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) is the safety parameter for the reactor

( LOCA-related accidents, and indicates whether the peak cladding temperature and the zirconium-water reaction is below the specified limits. As the decay power level is extremely low during spent fuel storage,-and will not increase, MCPR and MAPLHGR limits cannot be-exceeded and are not applicable.

L Those transients and accidents of USAR Chapter 15 which pose the

) potential for a radiological release outside the primary l

containment are of primary concern.

Heat Generation-Analysis-One result from the ORIGEN2 calculation is a graph of decay heat or thermal power (in watts), as a function of time. Results of this analysis are presented in Figure 15.1-1. The calculated decay heat load as of June 1989 is approximately 0.55 kw.

15-1 l i

i

5(( )

7 7, SHOREHAM~DSAR c  ?

L It must be recognized that there are some limitations in the ORIGEN2 model, and potential inaccuracies in the calculational l

. processes of the code and its supporting data' sets. For instance, ORIGEN2 is a " point. reactor" model, and cannot deal L

L conveniently with the-spatial variations in fuel enrichment and.

l burnup. 'In addition, there are uncertainties associated with l- averaging of-nuclear cross-section data within the thermal, C resonance, and fission neutron energy ranges. Nevertheless, it is not expected that large uncertainties should occur in heat

! load estimates.. See the comparison of calculated to measures

( . dose rates in DSAR Section 12.2. This gives evidence that the L decay heat.-load calculations'are reasonable, as the same analysis

1. (ORIGEN2) was used to generate both sets of data.

Analytical Categories k

L EachEUSAR Chapter'15 event is assigned to one of six analytical L

categories. .The analytical categories and the events in each analytical category are. discussed below.  ;

i 1. Decrease in Core Coolant _ Temperature

.This' analytical category of USAR Chapter 15 events includes

/~T the following events:

U Pressure Regulator Failure - Open L 15.1.5 l

~15.1.7 Feedwater-Controller Failure - Maximum Demand 15.1.8 Loss of'Feedwater Heating 15.1.9 Shutdown Cooling (RHR) Malfunction - Decreasing Temperature.

l

~in. In the spent fuel storage condition, the pressure regulator, feedwater controller, feedwater heating system and RHR system are not operating and all'four transients

! are, therefore, not applicable.

2. Increase in Reactor Pressure Since the generator, turbine, main steam isolation valve, pressure-regulator, feedwater system, condenser and RHR systems are not operating in support of nuclear fission, the following transients are not applicable:

1 15.1.1 Generator Load Rejection ]

15.1.2 Turbine Trip g s, j

>Q 15.1.3 Turbine Trip with Failure of Generator Breakers to ,

ppe_g 15-2

l. l l

l

- -= .

s.

4 L: - -

-SHOREKAM DSAR 15.1.4 Main Steam Isolation Valve Closure 15.1.6 Pressure Regulator Failure - Closed 15.1.18 Loss of Feedwater Flow 15.1.21-Loss of Condenser Vacuum 15.1.26 Core Coolant Temperature Increase L The transient of this category applicable to spent fuel l storage.is the following:

15.1.19 Loss of AC Power L A loss of AC power condition can be postulated that will affect normal support systems. However, because of the-very low heat-generation rate (550 watts) large thermal capacity of the pool active fuel pool cooling is not required. Loss of normal cooling and makeup systems will result only in a very slow evaporation 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 a0 by the spent fuel pool and low fuel decay heat eliminate

' ~

the need for active makeup requirements.. (The rate of i evaporation is discussed in Chapter'9.) l The loss of AC power will not in itself result in any l

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, j i

while the release in this case would not be monitored, the l offsite dose consequences would be incignificant.

Doses and dose rates are bounded by the " puff release" results J 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 L 15.1.20 Recirculation Pump Trip 15.1.22 Recirculation Pump Seizure

! 15.1.23 Recirculation Flow Control Failure With Decreasing Flow 15-3

SHOREHAM DSAR

^

1 4

r

4. Reactivity and-Power Distribution Anomalies Events included in this category are those which cause rapid-increase in power. Since the reactor is defueled, the following events are not applicable:

15.1.11 Continuous control Rod Withdrawal During 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 Operationa) Transient Due to inadvertent Loading of a Fuel Assembly into an Improper Location 15.1.16 Inadvertent Loading and Operation of a Fuel Assembly in Improper Location 15.1.24- Recirculation Flow Control Failure with 1

- (}

k- Increasing Flow 15.1.25 Abnormal Startup of Idle Recirculation Pump l 1

L 15.1.33 Control Rod Drop Accident

5. Increase in Reactor Coolant Inventory Since the HPCI system is not required the following transient is not applicable:

15.1.10 Inadvertent HPCI Pump Start 1

6. Decrease in Reactor Coolant Inventory j 1

l 6.A Events Not Applicable to Spent Fuel Storage The safety relief valve and the feedwater system are not i operating in the defueled condition; therefore the  !

l following events are not applicable:

15.1.17 Inadvertent Opening of a Egfety Relief Valve 15.1.37 Feedwater System Piping Break 1

( The following event is not a design basis event and is l applicable only to power operation:

I 15-4 l

~

~ ,

SHOREHAM DSAR 13.1.27L Anticipated Transient Without Scram (ATWS) b ' The single failure-proof polar crane design eliminates the following events a

'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 andstherefore the following events are not applicable:

15.1.30 Off-Design Operational Transient as a Consequence of ] strument Line Failure 15.1.34 Pipe. Breaks Inside the Primary Containment (Loss-of-Coolant Accident) 15.1.35 Pipe Breaks Outside the Primary Containment (Steam Line Break Accident) 4 6.B Events Without Fuel Damage 15.1.29 Miscellaneous Small Releases Outside Primary Containment

. Releases that could result-from piping failures outside the primary containment include the pipe breaks in the fuel '

pool cleanup system. The resulting offsite dose will be negligible and are bounded by the Radwaste Tank Rupture accident.

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 Failure 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.

l The accident scenario postulated in the USAR Sections 11.2.3.4.2 through 11.2.3.4.4 is considered here:

(}

L 15-5 l

1

s SHOREHAM DSAR M"i '.

1 1. .A conservative partition factor of 1.0E-03 is assumed for all isotopic activities listed in Section 11.1 with the ,

exception of H-3,Jfor which it is assumed all activity is evolved into the atmosphere.

A two hour release duration is assumed.

~

-2.

3. A ground-release atmospheric dispersion factor is assumed, ,

as given in Table 15.1.36-1 for the EAB. Note that the Exclusion Area Boundary (EAB) is limiting insofar as-10CFR100 dose limits are concerned, because the release +

duration is two hours.

4. The breathing' rate of adults offsite is assumed to be 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. ,

j The dose resulting from the analysis described above are as follows:

Dose, millirem 1 ) Source Whole Body Gamma

• Beta Skin Maximum Organ **

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

i Tank Radwaste 1.2E-07 1.7E-08 8.3E-06 Filters Discharge 3.1E-08 1.4E-08 7.7E06 i Sample Tanks Totals 1.8E-05, 2.8E-06 1.3E-03,

-The consequences of the above postulated accident are negligible. The whole body gamma, rkin, and thyroid doses are 7.2E-OS%,'9.3E-10%, and 4.3E-07%, respectively, of the 10CFR100 dose guidelines.

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

• • Teen is the limiting age group, and lung is the limiting organ 1 15-6 L

1

i=

41 e.

j 1

pj p s s

SHOREHAM DSAR i

15.1.38' Failure of Air Ejector Lines As the main condenser is,not operating, this accident is no L longer a design basis event.

E l

? 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 Starting 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 i for transport directly to the secondary containment. Table I 15.1.36-1 presents the parameters used in this analysis.

( .1 15.1.36.3 Accident Description l

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: 1 e  ;

Approximate Event- Elapsed Time

'l. Fuel assembly is being handled 0+

by refueling equipment._ The  !

assembly drops. l

2. Some of the fuel rods in both- 1 min.

sc; the dropped assembly and another assembly are damaged,'resulting in the release of gaseous fission ,

l 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 O alarm to alert plant personnel.

4. Operator actions begin 5 min.

15-7

c' o

7. . , , ,

4 F O SHOREHAM DSAR 15.1.36.4 Identification of Operator Actions

1. . The operator will initiate the evacuation of the

secondary' containment and securing of Secondary y Containment-doors, if necessary.

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

. 3. The fuel handling foreman will make the operator aware of the accident, 7

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

I

5. An HP technician will post the appropriate radiological i control signs at the entrance to the secondary containment.

(J' '

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 F fission-products in the fuel's gap which is-released will i not be large (2.52 Ci of Kr-85 only). Calculations .

L.

indicate that the reactor building refueling floor exhaust

. radiation monitoring system would not alarm and-consequently the RBSVS will not be actuated (i.e., the RBNVS continues to operate). As a' result, analyses were l

performed assuming either RBSVS or RBNVS system operation.

) Secondary containment discharge rates are 167 and 6580 percent / day for the RBSVS and RBNVS cases, respectively.

As a comparison case, a " puff" release over a short period of time (2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, as suggested by Regulatory Guide 1.25),

has been analyzed. Although this 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 putf release) are given in the following sections.

15.1.36.6 Analysis of Effe, cts and Consequences 15.1.36.6.1 Evaluation Methods

-(}

The analytical methods and associated assumptions used to evaluate the consequences of this accident are consistent m with Regulatory Guide 1.25. The assumptions and parameters are given in Table 15.1.35-1.

15-8

g- W. l h

T:

'h SHOREHAM DSAR 15.1.36.6.1.1 Methods, Assumptions, and Conditions

?

The assumptions'used in the analysis of this accident are listed belows o 11 . The fuel assembly is dropped from the. maximum height n allowed by the fuel handling equipment.

l

2. The entire amount of potential energy, referenced to the top of the spent fuel racks, is available for j

application to the fuel assemblies involved in the accident...This assumption neglects the dissipation of l:

i some of the mechanical energy of the falling fuel h assembly $n 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.

f 3. None of the energy associated with the dropped fuel l: assembly-is absorbed by the fuel material (uranium dioxide).

15.1.36.6.1.2 Results and Consequences

~15.1.36.6.1.2.1 Fuel Damage l

. 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 result of dropping the-fuel issembly into the reactor vessel. The same assumption is applied here.

15.1.36.6.1.2.2 Fission Product-Release From Fuel l Fission product releases for the fuel handling accident are determined from the inventory:in Table 12.2-1.

'Specifically, it is seen that only Kr-85 is of any significance with respect to gaseous releases. The only other gaseous isotope in this table is H-3, which would add, at most, 0.1% to the skin dose from Kr-85. Using the l above number of failed rods, and the assumptions given in j .. 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 C1 l

l l

l 15-9 e

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'r~ y SHOREHAM DSAR A

15.1.36.6.1.3 Radiological 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.

p Control Room l-

.Because the amount of radioactivity released is so small, the l'- 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 l

body and skin 30-day integrated doses are, at most, 9.59E-08

' and~2.08E-04 mrem, respectively, well below the 10CFR50 GDC 19 limits.

Discussion It is seen in Table 15.1.36-2 that the (0-2 hour) EAB and (0-30 day) LPZ integrated doses are many orders of magnitude  ;

~

below,10CFR100 guidelines. Results are graphically shown in.

L E

f'/)

k-

~ 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 RBNVS case, below Technical Specifications. This indicates that the HVAC system in use in the reactor building has no meaningful effect on radiological consequences to-members of the public during a fuel handling accident with the present fuel source terms.

15.1.36A Worst Case Fuel Damage Event Scenario L

L Several " worst case", extremely conservative scenarios were examined. Specifically, for the three reactor building HVAC cases-analyzed in Section 15.1.36.5 (RBSVS operating, RBNVS operating, and puff release), instead of assuming the gap activity from 125 fuel rods is released (2.52 Ci Kr-85), it is assumed that all gaseous activity from the entire-core in l

the spent fuel pool is released (1.56E+03 Ci Kr-85). This

! _can only occur if all the fuel is postulated to be mechanically damaged and there is a complete release of

gaseous isotopes. The assumption of a complete release of the gaseous inventory is also very conservative with respect to the Regulatory Guide 1.25 assumption of a 30% release fraction given the low burnup condition of Shoreham spent Doses and dose rates are thus a factor of 617 higher

(, fuel.

than for the corresponding Regulatory Guide 1.25 cases. l l

15-10 1

I

SHOREHAM.DSAR

~

All other conditions and parameters indicated in Table 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 l Figure 15.1.36A-1. Dose rates'for the postulated worst case scenario are above. current limits, but the duration of the high dose rates in the RBNVS and puff release cases is quite short (two hours or.less).

LO. l l

i 1

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1 1

l.

i l

l L

l l

15-11 I'-

l- - - -

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SHOREHAM DSAR

~

TABLE 15.1.36-1 FUEL HANDLING ACCIDENT - PARAMETERS FOR POSTULATED ACCIDENT ANALYSES Conservative (NRC)

Assumptions I. Data and assumptions used to estimate radioactive source from postulated accidents A. Power level- -See this Chapter B. Peaking factor 1.5 C. Fuel damaged 125 rods D. Release of activity from' fuel 30% Kr-65 E. Iodine. fractions (1) Organic N/A (2) Elemental N/A

-: (3) Particulate N/A

- II. Data and assumptions used to L estimate activity released I

A. Secondary contai:3- See Section ment, discharge rate '(%/ day) 15.1.36.5 B.- Adsorption and filtra- l L tion efficiencies l (1) Elemental iodine N/A i l C. Recirculation system l parameters E (1) Flow rate N/A (2)' Mixing efficiency N/A III. Dispersion data (

A. EAB and LPZ distances (meters) 311/3,220 3

L B. X/Qs (sec/m )

EAB (0-2 hr) 1.36E-03 LPZ (0-8 hr) 2.50E-05 (8-24 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 C. Doses and Dose Rates Table 15.1.36-2 1

s 1

SHORERAM DSAR TABLE 15.1.36-2 FUEL HANDLING ACCIDENT RADIOLOGICAL CONSEQUE9EFS i

Whole Body Dose, rem Skin' Dose, rem HVAC 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, mrem /hr- .

Whole Body Gamma Skin ,

Tech. Spec Tech. Spec

( EAB LPZ Limit EAB LPZ Limit i < 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 10CFRIO0 10CFR100 j-s Operates EAB LPZ Limit _ EAB LPZ Limit

• L 1.74E-06 3.22E-08 2.50E+01 1.52E-04 2.80E-06 3.00E+02 h

Maximum (t = 0) Dose Rates, mrem /hr 4

"fif

)

Whole Body Gamma ,

Skin Tech. Spec a: Tech. Spec EAB LPZ Limit EAB LPZ Limit 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 e Maximum (t = 0) Dose Rates, mrem /hr Whole Body Gamma Skin Tech. Spec Tech. Spec EAB LPZ Limit EAB LPZ Limit 8.75E-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.

f g,

i i

SHOREHAM DSAR

}{(

i TABLE 15.1.36A-1

" WORST CASE" FUEL DAMAGE ACCIDENT )

RADIOLOGICAL CONSEQUENCES Whole Body Dose, rem Skin Dose, rem HVAC 10CFR100 10CFR100-Scenario EAB 'LPZ Limit EAB LPZ Limit RBSVS 7.03E-05 7.50E-06 2.50E+01 6.11E-03 6.52E-04 3.00E+02 Operates Maximum ( t = 0 Dose Rates, mram/hr Whole Body Gamma __

Skin Tech. Spec Tech. Spec 100B LPZ Limit EAB LPZ Limit ,

3.76E-02 6.92E-04 5.70E-02 3.27E+00 6.01E-2 '3.42E RBNVS Whole Body' Dose, rem Skin Dose, rem Operates 10CFR100 10CFR100-EAB LPZ Limit EAB LPZ _

Limit 1.08E-03 1.99E-05 2.50E+01 9.35E-02 1.73E-03 3.00E+02 Maximum (t = 0) Dose Rates, mrem /hr

!. Whole Body Gamma Skin Tech. Spec Tech. Spec EAB LPZ Limit EAB LPZ Limit 2.96E-00 5.44E-02 5.70E-02 2.57E+02 4.73E+00 3.42E-01 Puff Whole Body Dose, rem Skin Dose, rem Release 10CFR100 10CFR100 EAB LPZ Limit EAB LPZ Limit 1.0BE-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 Tech. Spec Tech. Spec EAB LPZ Limit EAB _

LPZ Limit 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

O O lC1 .

DSAR FIGURE 15.1-1 ~

SNPS Spent Fuel Decay Heat Load 1200 1080

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-!-l 840 o .

3 f 720 0

800 O

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O 360 O

240 i 120

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O 10 0 DSAR Fig u re 1 5.1. 3 6~- 1 Design - Basis Fuel Handling Accident Exclusion Area Boundary . Results RBNVS HVAC System In Operation _

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1 G' ' SHOREHAM DSAR

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CHAPTER 16 .

-TECHNICAL SPECIFICATIONS The SNPS' Technical' Specifications are found'in. Appendix '. of.

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i SHOREHAM.DSAR +

l CHAPTER 17 QUALITY ASSURANCE-l17.1 -QUALITY ASSURANCE DURING DESIGN AND CONSTRUCTION The description contained under this heading'in the latest revision of the Shoreham USAR remains unchanged. Refer to USAR for information-on this subject.

17.2 QUALITY ASSURANCE DURING THE OPERATIONAL. PHASE 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 L

designated as Quality Assurance Category I.(safety related) in USAR Table 3.2.1-1 have been redesignated as Quality Assurance Category IIA in this DSAR. The applicability of the USAR Section 17.2 Operational phase Quality Assurance Program as modified in-L this DSAR to the Quality Assurance (QA) Categories in DSAR Table 3.2.1-1 are.as follows:

"\' QA Category I - The USAR Section 17.2 Quality Assurance t

Program as modified by DSAR Section 17.2 applies to the safety related structures, systems and components which meets the-intent of 10CFR50, Appendix B.

QA Category IIA - Category IIA systems or components which

.(formerly safety were previously designated as-QA Category I related)- but no longer have a safety function. These

(.

systems or components are considered non-safety related and will no longer comply with QA Appendix B or Nuclear Codes and Standards. Deviations from previously defined QA Category'I requirements will be documented and filed for retrievability.

L E QA Category II - Appropriate measures are applied to these (non safety- structures, systems, and components in related) accordance with QA corporate policy to assure that the safety significance given to them in the USAR, Technical Specifications, and Emergency Operating Procedures are maintained.

f l

'The specific modifications of the USAR Section 17.2 applicable to s the Shoreham DSAR phase are as follows:

17-1

5:

SHOREHAM DSAR 17.2.1 Organization

- A) The hesistant Vice President, Nuclear Operations has assumed '

all the responsibilities of the Vice President, Nuclear operations and the position of Vice President, Nuclear Operations was eliminated.

B) The Manager, Nuclear Quality Assurance Department (NQAD),

reports directly to the Assistant Vice President, Nuclear Operations, but has direct access to the President of the Company as he deems necessary.

C) The Safety Engineering and Reliability organization that reported to the Manager, NQAD, has been eliminated. This organization included the Independent Safety Engineering L Group (ISEG) and Reliability Section. The Nuclear Review Board continues under the responsibility of the NQA Manager as Chairman. The Quality Control and Quality Systems organizations continue to be jointly responsible for assuring full implementation of the LILCO QA Program.

D) The Quality Systems'(QS) Manager tu located within the-protected area.

O E). The title of Senior Vice President is changed to Group Vice President.

F) "The calibration services of Gas Operations and Electric L Operations are-no longer utilized by SNPS. These services l -are now performed by SNPS personnel or qualified suppliers.

1 L 'G) The Vice President,' Corporate Services, reports to the President and provides SNPS with training, emergency -i preparedness and nondestructive examination services. The primary responsibility for providing these services previously were the Assistant Vice President and Director of the Office of Training, Nuclear Operations Support Department and Nuclear Quality Assurance Department, respectively.

These. services continue to be subject to the policies and requirements of the QA Program.

H) The LILCO organization structure for the Shoreham DSAR Phase is shown in Figure 17.2.1-1.

I) The Nuclear Engineering Department (now Engineering and Technical Support) has been reorganized and reports to the Manager, Engineering and Administrative Support.

e> J) The Nuclear Operations Support Department and the Nuclear

.(_} Engineering Department have been combined in the Engineering and Administrative Support Organization.

17-2

e

() SHOREHAM DSAR 17.2.2- Quality Assurance Procram A) The Reliability and ISEG organization are eliminated from the NQAD. 4 B) The structures, systems, and components designated as QA Category I (safety related) in USAR Table 3.2.1-1 which no longer have a safety function are designated as QA Category IIA. These structures,-systems, and components will no longer require a QA Appendix B program and need not comply with Nuclear Codes and Standards. These structures, systems, and components will be afforded the requirements associated with non-safety systems. Deviations from QA Category I requirements wi,ll be documented and filed for traceability.

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 Control of Material, Parts and Components No: change.

17.2.9 Control of Special Processes No change.

17.2.10 Inspection No change.

-O-17.2.11 Test Control No change. i 17-3

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.s SHOREHAM DSAR

' 17.2.1'2~ Control of Measuring and' Test Equipment

-No change.:

!17.2.13 Inspection, Test, and operating Status No change. ,

i 17.2.14' Inspection, Test, and operating Status c No change., i L17.2.15 Nonconforming Materia _is, Parts, or Components t:

No change.

u j .. 17.2.16 Corrective Action No change.

ll, 17.2.17 Quality Assurance Records 7m- ' No change.

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17.2.18 Audits No. change.

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L~' SHOREHAM DSAR M

NUREG-0737'TMI ACTION PLAN REQUIREMENTS

- I. OPERATIONAL SAFETY

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I.A.1.1 Shift Techniepl-Advisor NRC Position Each licensee shall provide anfon-sh'ift technical advisor to the .

- shift supervisor.; The shift technical advisor (STA) may serve 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 n 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 design and layout, including the capabilities of instrumentation and controls in the control room. The licensee shall assign'

- normal duties to the STAS that pertain to the engineering aspects of assuring safe operations of the plant, including the review '

L and' evaluation of operating experience.

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Vl LILCO Position-LUnder the conditions of the LILCO and New York State Shoreham L settlement the plant is shutdown and defueled. STA staffing is therefore not required.

~

'I.A.l.2 Shift Supervisor Administrative Duties NRC Position-Review the administrative duties of the shift supervisor and

' delegate. functions that detract from or are subordinate-to the management responsibility for assuring safe operation of'the plant to other personnel not on duty in the control room.

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L 1. The highest level of corporate management of each licensee L shall issue and periodically reissue a management directive l

that emphasizes the primary management responsibility of the L

shift supervisor for safet operation of the plant under all conditions on his shift and that clearly establishes his command duties.

-2. Plant procedures shall be reviewed to assure that the duties, responsibilities, and authority of the shift supervisor and

' ~

control room operators are properly defined to effect the

't establishment of a definite line of command and clear delineation of the command decision authority of the shift supervisor in the control room relative to other plant managment personnel. Particular emphasis shall be placed on the following:

l

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L SHORENAM'DSAR 1

a. The responsibility and authority of the shift supervisor shall be to maintain the broadest perspective of i

operat.ienal conditions affecting the safety of the plant as a matter of highest priority at all times when on duty

'in the control room. The principle shall be reinforced I 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 F 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

h. specified.

i t 3. . Training programs for shift supervisors shall emphasize and u

' .- reinforce.the responsibility for safe operation and the 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 y for plant operations. Administrative functions that detract 1 I 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 I safe operation of the plant. See the USAR for additional -

L information.

A Corporate Management Directive that emphasizes the primary 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.

L

u l

SHOREHAM DSAR s

, rQ(I I.A l.3 Shift Manning NRC Position Assure that the necessary 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 qualified individuals are readily available in the event of an abnormal or emergency situation. This should consider the r recommendations on overtime in NUREG-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 operatorr (SRO), one of whom shall be designated as the shift supervistr, two licensed reactor operators (RO), and two unlicensed auy111ary operators (AO). For a multi-unit station, depending upon the station configuration, shift staffing may be adjusted to allow

,s . credit for licensed senior reactor operators and licensed reactor operators to serve as relief operators on more than one unit; however, these individuals must be properly licensed on each such unit. At all other times, for a unit loaded with fuel, the minimum shift crew shall include one shift supervisor who whall ,

be a licensed senior reactor operator (SRO), one licensed reactor operator (RO), and one unlicensed auxiliary operator (AO). -

Adjunct requirements to the shift staffing criteria stated above are as follows:

a. A shift supervisor with a senior reactor operator's license, who is also a member of the station supervisory staff, shall be onsite at all times when at least one unit is loaded with fuel,

b. A licensed senior reactor operator (SRO) shall, at all times, be in the control room from which a reactor is being operated. The shift supervisor may from time-to-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 e g. number of control rooms from which the reactors are being Q operated.

d. In addition to the licensed senior reactor operators specified in a., b., and c. above, for each reactor

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> 1 SHOREHAM.DSAR lC  !

' x-/  !

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- 1 (RO) shall be onsite at all times and available1to 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. l

q. In additioh 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

. senior reactor operator may have fuel handling' duties but shall not have other concurrent operational duties.

Licensees of operating plants and applicants for operating gs licensees shall include in their administrative procedures 1 (required by license conditions) provisions' governing required .

shift staffing and movement of key individuals about the plant.

These provisions.are required to assure that qualified plant i

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 working 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.

The staff recognizes that there are diverse opinions on the amount 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 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

('~dT IE Circular No. 80-02.

In the event that overtime must be used (excluding extended periods of shutdown for refueling, major maintenance or major

% 4 4

)

SHOREHAM DSAR  !

~x plant modifications), the following overtime restrictions should Ec To11tued: l I

(1) An-individual shall not be permitted to work more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> straight (not including shift turnover timeli (2) There shocid be a break of at least 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (which can 4 include shift turnover time) between all work periods. ]

1 (3) An individual shall not work more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any l 7-day period. l L (4) An individual shall not work more than 14 consecutive i days without having 2 consecutive days off.

l However, recognizing that circumstances may arise requiring ,

L deviation from the above restrictions, such deviation may be l l 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.

If a reactor operator or senior reactor operator has been working more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 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 I . '( ) -

12-hour break preceding such an assignment.

We encourage the development of a staffing policy that would l

! permit the licensed reactor operators and senior reactor ,

l ' operators to be periodically assigned to other duties away from L the control boapC during their normal tours of duty. l l

If a reactor operator is required to work in excess of 8 )

continuous hours, he shall be periodically rolieved of primary duties at the control board, such that periods of duty at the L >

board do not exceed about 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at a time.

The guidelinas on overtime do not apply to the shift technical advisor provided he or nhe is provided sleeping accomodations and a 10-minute availability is assured.

t L Operating license applicants shall complete these administrative l procedures before fuel loading. Development and implementation of the administrative procedures at operating plants will be reviewed by the office of Inspection and Enforcement beginning 90 days after July 31, 1980.

LILCO Position The Shoreham Station Proced.ures implement the following:

1. The minimum shift complement consists of three operators and a sufficient number of extra people in order to meet the Emergency Plan and Fire Protection Plan requirements.

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? 4 SHOREHAM DSAR l ,

J l

L 2. The shift schedule conforms to the guidelines provided in the L Shoreham Station Procedure entitled Station Operations -

Overtime Selection as it applies to the scheduling and use of

( overtime. ,

l L 3. The movement in the plant by members of the shift complement L are such that they may be easily and rapidly informed and/or contacted and dispatched by the operators in the event an emergency situation arises.  :

The above items ensure that qualified plant personnel are available to man operational shifts.

l 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 unito are operated by LILCO, the requirement 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 the portions of this Task Action Item that apply to a shutdown and defueled plant.

I.A.2.1 Immediate Upgrade of Reactor Operator and Senior Reactor Operator Training and Qualification NRC Position L Effective May 1, 1980, an applicant for a senior reactor operator (SRO) license shall have four years of responsible power plant experience of which at least two years shall be nuclear power plant 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 fulfilled by academic or related technical training, on a one-for-one time basis.

Effective December 1, 1980, an applicant for a senior reactor operator (SRO) license shall have held an operator's license for one year.

[

n L

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) l

() license shall have three months training on shift as an extra person in the control room, l-I Effective August 1, 1980, training programs shall be modified to provide:

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--_._________w* --

-- - - - + - -

l SHOREHAM DSAR l

1. Training in heat transfer, fluid flow, and thermodynamics, l

2. Training in the use of installed plant systems to control l or mitigate an accident in which the core is severely l damaged, i i

3. Increased emphasis on reactor and plant transients.

l 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 i There will not be iny applicants for SRO or RO licenses.

The SNPS operator training program will provide training for I subjects applicable to a shutdown and defueled plant and also for fuel handling operations.

J.A.2.3 Administration of Training Programs l

.ry NRC Position  !

(_) Applicants for operator licenses will be required to grant permission to the NRC to inform their facility management regarding the results of examinations.

l Contents of the licensed operator requalification program shall be modified to include instruction in heat transfer, fluid flow, thermodynamics, and mitigation of accidents involving a degraded core.

i

! The criteria for requiring a licensed individual to participate in accelerated requalification shall be modified to be consistent with the new passing grade for issuance of a license.

Requalification programs shall be modified to require specific reactivity control manipulations. Normal control manipulations, j 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.

LILCO Position l

I')

k-It is LILCO's position that permanent members of the training staff who teach systems, integrated responses, or transients be qualified or certified to teach in the appropriate subject area.

V ~

1 l

l

-~ SHOREHAM DSAR LILCO does not intend to require either guest lecturers who are experts in particular subjects (reactor theory, instrumentation, thermodynamics, health physics, chemistry, etc.) to successfully complete an NRC SRO examination; or system experts, such as an instrument and control supervisor teaching the control rod drive system to successfully complete an NRC SRO examination.

The degree of training provided will be commensurate with the 1 tasks required to be performed.

2.A 3.1 Revise Scope and Criteria for Licensing Examinations  !

NRC Position Applicants for operator licenses will be required to grant permission to the NRC to inform their facility management I regarding the results of examinations.

ll Contents of.the licenses operator requalification program shall l be modified to include instruction in heat transfer, fluid flow,  !

thermodynamics, and mitigation of accidents involving a degraded l Core.  !

(" The criteria for requiring a licensed individual to participate in accelerated requalification shall be modified to be consistent with the new passing grade for issuance of a license.

l Requalification programs shall be modified to require specific reactivity control manipulations. Normal control manipulations, such as plant or reactor startups, must be performed. Control ,

manipulations during abnormal or emergency operations shall be l walked. l l

Simulator examinations will be included as part of the licensing i examinations.

LILCO Position

{

LILCO will not have applicants for operator licenses. Licenses will be limited to fuel handling operations.

I.B l.2 Evaluation of Organization and Management NRC Position Each applicant for an operating license shall establish an onsite independent safety engineering group (ISEG) to perform l independent reviews of plant operations.

l

-( ) The principal function of the ISEG is to examine plant operating characteristics, NRC issuances, Licensing Information Service advisories, and other appropriate sources of plant design and operating experience information that may indicate areas for improving plant safety. The ISEC is to perform independent

SHOREHAM DSAR O)

\_

review and audits of plant activities including maintenance, modifications, operational problems, and operational analysis, and aid in the establishment of programmatic requirements for plant activities. Where useful improvements can be achieved, it is expected that this group will develop and present detailed recommendations to corporate mLaagement for such things as revised procedures or equipment modifications.

Another function of the ISEG is to maintain surveillance of plant operations and maintenance activities to provide independent verifications that these activities are performed correctly and that human errors are reduced as far as practicable. ISEG will #

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 sign-off functions'auch that it becomes involved in the operating organization.

l 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

! Plan, Regulatory Guide 1.33, Standard Technical Specifications).

Rather, it is an additional independent group of a minimum of l

five dedicated, full-time engineers, located onsite, but reporting offsite to a corporate official who holds a high-level,

, technically oriented position that is not in the management chain for power production. The ISEG will increase the available t 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 l contact with and knowledge of day-to-day plant operations and

provide additional expertise. However, the shift technical advisor on shift is necessarily a member of the operating staff and cannot be independent of it.

l o It is expected that the ISEG may interface with the quality l

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 l portions of the independent, safety assessment function in a L centralized location for all the utility's plants. In such I cases, an onsite group still is required, but it may be slightly smaller than would be the case if it were performing the entire independent safety assessment function. Such cases will be O$ reviewed on a case-by-case basis.

! AT this time, the requirements for establishing an ISEG is being applied only to applicants for operating licenses in accordance l

s SHOREHAM DSAR i

t with Action Plan item I.B.I.2. The staff intends to review this activity in about a year _to determine its effectiveness and to l see whether changes are required. Applicability to operating plants will be considered in implementing long-term improvements in organization and management for operating plants (Action Plan item I.B.1.1).

LILCO Position I

LILCO 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 and licensees of plants under construction to perform analyses of transients and accidents, prepare emergency procedure l guidelines, upgrade emergency procedures, including procedures  :

for operating with natural circulation conditions, and to conduct operating retraining (see also Item I.A.2.1). Emergency procedures are required to be consistent with the actions

, necessary to cope with the transients and accidents analyzed.

L Analyses of transients and accidents were to be completed in l

early 1980 and implementation of procedures and retraining were I 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, 41, 47, 55, 57).

Based on staff reviews to date, there appear to be some recurring 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 <

l 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 L alternative actions that should b performed to mitigate the event should these systems or components fail.

l

SHOREHAM DSAR 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 oeneral, the sequence of events for the transients and accidents and inadequate core cooling analyzed should postulate multiple failures such that, if the failures 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 generator;

2. Failure of main and auxiliary feedwater;

3. Failure of high-pressure reactor 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;

5. Operator errors of omission or commission.

The analyses should be carried out far enough into the event to assure that all relevant thermal / hydraulic /neutronic phenomena are identified (e.g., upper head voiding due to rapid cooldown, 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 a logical transition from the emergency procedures into the inadequate core cooling procedure including the use of instrumentation to identify inadequate core cooling conditions.

Rationale for this transition should be discussed. Additional information that should be submitted includes:

1. A detailed description of the methodology used to develop the guidelines;

2. Associated control function diagrams, sequence-of-event diagrams, or others, if used;

3. The bases for multiple and consequential failure considerations; O 4. Supporting analysis, including a description of any computer codes used;

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

- o 1 SHOREHAM DSAR f

5. A description of the applicability of any generic results to plant-speci fic applications.

Owners' group or vendor submittals may be referenced as appropriate to support this reanalysis. If owners' group or vendor submittals have already been forwarded to the staff for review, a brief description of the submittals and justification of their adequacy to support guideline development is all that is required.

Pending staff approval of the revised analyais and guidelines, the staff will continue the pilot monitoring of emergency procedures described in task action plan Item I.C.8 (NUREG-0660) .

For PWRs, this will involve review of the luss-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 l emergency procedure review. Since the analysis and guidelines  ;

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 for operating licenses pending, the interim program for BWRs will consist of trial implementation on these six plants.

.( ) Following approval of analysis and guidelines and the pilot monitoring of emergency procedures, the staff will advise all

)

i ilicensees of the adequacy of the guidelines for application to )

their plants. Consideration will be given to human factors I engineering and system operational characteristics, such as  ;

information transfer under stress, compatibility with operator j training and control room design, the time required for component 1 and system response, clarity of procedural actions, and control l room personnel interactions. When this determination has been made by the staff, a long-term plan for emergency procedure review, as described in task action plan Item I.C.8, will be made available. At that time, the reviews currently being conducted on NTOLs under Item I.C.8 will be discontinued, and the review

~

required or applicants for operating licenses will be as i described in the long-term plan. Depending on the information I submitted to support development of emergency procedures for each l I

reactor type or vendor, this transition may take place at different times. For example, if the GE guidelines are shown to I be effective on the six plants chosen for pilot monitoring, the long-term plan for EWRs may be complete in early 1981. Operating plants and applicants will then have the option of implementing the long-term plan in a manner consistent with their operating i schedule, provided they meet the final date required for j implementation. This may require a plant that was reviewed for

an operating license under Item I.C.9 to revise its emergency l

i

() procedures again prior to the final implementation date for Item I.C.8. The extent to which the long-term program will include i

! review and approval of plant-specific procedures for operating l plants has not been established. Our objective, however, is to j

! l l

l

s .

SHOREHAM DSAR minimize the amount of plant-specific procedure review and  ;

approval required. The staff believes this objective can be i acceptably accomplished by concentrating the staff review and I approval on generic guidelines. A key element in meeting this  :

objective is use of staff-approved generic guidelines and i guideline revisions by licensees to develop procedures. For this approach to be effective, it is imperative that, once the staff has issued approval of a guideline, subsequent revisions of the guideline should not be implemented by licensees until reviewed -

and approved by the staff. Any changes in plant-specific procedures based on unapproved guidelines could constitute an ,

unreviewed safety issue under 10CFR50.59. 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 case, deviations from generic guidelines should not be implemented until staff approval is formally ,

received in writing. Interim implementation of analysis and  :

procedures for small-break loss-of-coolant accident and inadquate core cooling should remain on the schedule contained in NUREG-0578, Recommendation 2.1.9.

LILCO Position The description contained under this heading in the latest

( revision of Shoreham USAR remains unchanged with the following exception: LILCO will not continue to participate in the BWR Owners' Group program to develop Emergency Procedure Guidelines for GE Boiling Water Reactors. Refer to USAR for information on this subject.

I.C.2 Shift and Relief Turnover Procedures NRC Position The licensees shall review and revise as necessary the plant procedure for shift and relief turnover to assure the following:

1. A checklist shall be provided for the oncoming and offgoing control room operators and the oncoming shift supervisor to complete and sign. The following items, as a minimum, shall be included in the checklists

a. Assurance that critical plant parameters are within allowable limits (parameters and allowable limits shall be listed on the checklist).

b. Assurance of the availabilty and proper alignment of all systems essential to the prevention and mitigation of operational transients and accidents gs by a check of the control console (what to check and criteria for acceptable states shall be included on the checklist).

J A SHOREHAM DSAR U

c. Identification of systems and components that are in a degraded mode 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 separate entry on the checklist).

2. Checklists or logs shall be provided for completion by the

' ongoing-and oncoming auxiliary operators and technicians.  !

L 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 l

transients and accidents or initiate an operational transients (what to check and criteria for acceptable status  !

L shall be included on the checklist), and j

3. A system shall be established to evaluate yhe effectiveness l of the shift and relief turnover procedure (for example, -l periodic independent verification of system alignments) . 1 LILCO Position

.i s

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

I.C.3 shift Supervisor Responsibilities NRC Position i In the letters of September 13 and 27, October 10 e:d 30, and November 9, 1979, NRC required licensees and applicants to review and revise as necessary plant procedures and directives to assure that the duties, responsibilities, and authority were properly j defined to establish a definite line of command and clear

• L delineation of the command decision authority of the supervisor l in the control room relative to other 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 I

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.

4 information on this subject.

SHOREHAM DSAR Eq' L)

I.C.4 Control Room Access NRC Pos_i, tion 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 shift supervisor, and control room operators), to technical advisors who may be requested or required to support the ,

operation, and to predesignated NRC personnel.

Provisions shall include the following:

1. Develop and implement en administrative procedure that establishes the authority and responsibility of the person in charge of the cintrol room to limit access.  ;

2. Develop and implement procedures 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 j

established and limited to persons possessing a current senior reactor operator's license. The plan shall clearly define the lines of communication and authority l for plant management personnel not in direct command of I ( operations, including those who report to stations outside of the control room.

LILCO Position A Shoreham Station Procedure on Control Room Conduct establishes the authority and responsibility of the person in charge of the control room to limit access to the control room.

The same procedure establishes the line of authority and responsibility in the control room in the event of an emergency.

I.C.5 Procedures for Feedback of ,2perating Experience to Plant Staff NRC Position l

In accordance with Task Action Plan I.C.5, Procedures for Feedback of Operating Experience to Plant Staff (NUREG-0660),

L each applicant for an operating license shall prepare procedures to assure that operating information pertinent to plant safety originating both within and outsida the utility organization is continually supplied to operators und other personnel and is incorporated into training and retraining programs. These procedures shall:

(1) Clearly identify organizational responsibilities for review of operating experience, the feedback of pertinent information to operators and other personnel, and the incorporation of such information into training and retraining-programs;

SHOREHAM DSAR (2) Identify the administrative and technical review steps 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 affected personnel become aware i of and understand information of sufficient importance that i should not wait for emphasis through routine training and j 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 contradictory information is not conveyed to operators and

(~) other personnel until resolution is reached; and, (7) Provide periodic internal audit to assure that the feedback program functions effectively at all levels.

Each utility shall carry out an operating experience assessment function that will involve utility personnel having collective 1 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 review information from a variety of sources. These include operating information from 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 operating 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 that operations management personnel would be directly involved in tne 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 generel

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

.- x 9 M SHOREHAM DSAR

\

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 bs 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.

l 2. The Nuclear Review Board and ISEG have been eliminated.

! Refer to Section 13.4.2 and 13.4.3 for justification.

1

() 3. The Training Division Manager has been changed to office of Training.

4 The responsibilities assigned to the Plant Manager, Division 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 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 humsn errors and '

improving the quality of normal operations. This will reduce the frequency of occurrence of situations that could result in or contribute to accidents. Such a verifioncion system may include I automatic system status monitoring, human verification or I

operations, and maintenance activities independent of the people l

performing the activity (see NUREG-0585, Recommendation 5) or i both.

Implementation of automatic status monitoring if required will reduce the extent of human verification of operations and i

J maintenance activities but will not eliminate the need for such verification in all instances. The procedures adopted by the l

l

M' .

SHORERAM DSAR licensees me.y consist of two phases - one before and one after installation of automatic status monitoring equipment, if required, in accordance with item I.D.3 of NUREG-0660.

An acceptable program for verification of operating activities is described below.

The American Nuclear Society has prepared a draft revision to ANS 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:

j, (1) Applicability of the guidance of Section 5.2.6 should be i

extended to cover surveillance testing in addition to maintenance.

l (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 SRO, provided provisions are made f') to ensure that the shift supervisor is kept fully informed k/ of system status.

l (3) Except in cases of significant radiation exposure, a second qualified person should verify correct implementation or equipment control measures such as tagging of equipment.

I (4) Equipment control procedures should include assurance that control-room operators are informed of changes in equipment status and the effects of such changes.

L (5) For the return-to-service of equipment important to safety, i

, a second qualified operator should verify proper systems l alignment unless functional testing can be performed without I l

compromising plant safety, and can prove that all equipment, valves, and switches Anvolved in the activity are correctly aligned.

NOTE: A licensed operator possessing knowledge of the systems i involved and the reistionship of the systems to plant ,

safety would be a qpelified person. The staff is inveetigating the level of qualification necessary for other operators to perform these functions.

l For plants that have or will have automatic system status

(

monitoring as discussed in Task Action Plan Item I.D.3, .

> 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.

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SHORERAM DSAR 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 service,

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 NSSS Vendor Review of Procedures NRC Position Applicants for near-term operating licenses will be required to O'

obtain NSSS vendor review of low-power and power-ascension test and emergency procedures (see Regulatory Guide 1.33, Appendix A, Section 6) as a further verification of the adequacy of the procedures. 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 C.1, Item 4a and Table C.3,-Item 50 of NUREG-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.B Pilot Monitoring of Selected Eme_rgency Procedures For Near-Term Operating License Applicants NRC Position Correct emergency procedures as necessary based on the NRC audit 7 of selected plant emergency operating procedures (e.g., small break locs-of-coolant accident, loss of feedwater, restart of engineered safety features following a loss of ne power and steam-line break).

l I

SHOREHAM DSAR

(

LILCO Position j l The generic guidelines prepared by the BWR Owners' Group and i l approved by the NRC for trial implementation at the Shoreham l

Nuclear Power Station have been incorporated into the Shoreham j 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 l verified procedures were submitted for NRC review. Any comments  !

I or corrections found necessary as a result of the NRC audit were evaluated and implemented, as appropriate.

l

[ This item was closed in Section 3.1.1 of Inspection Report 83-09. I 1.D.1 Control Roos Design Reviews NRC Position In accordanew with Task Action Plan I.D.1, Control Room Design t Reviews (NUREGl-0660), all licensees and applicants for operating L

licenses will be required to conduct a detailed control-room l design review to identify and correct design deficiencies. This I l detailed control-room design review is expected to take about a l l t year. Therefore, the Office of Nuclear Reactor Regulation (NRR) i l requires that those applicants for operating licenses who are i unable to complete this review prior to issuance of a license 1 make pre'liminary assessments of their control rooms to identify l significant human factors and instrumentation problems and j establish a schedule approved by NRC for correcting deficiencies. j r These applicants will be required to complete the more detailed control room reviews on the same schedulo as licensees with I

l operating plants.

NRR is presently developing human engineering guidelines to assist each licensee and applicant in performing detailed l 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 modifications implemented. Applicants for operating licenses who will be unable to complete the detailed control-room design review prior g to issuance of a license are required to perform a preliminary l .

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 Engineering Guide to control Room Evaluation," in performing the preliminary assessment. NRR will evaluate the applicants' preliminary abr,essmens including the performance by NRR of onsite review /

i SHOREHAM DSAR l

($) -

audit.- The NRR onsite review / audit will be on a schedule consistent with licensing needs and will emphasize the following aspects to 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 L 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 ,

t multiple failures of nonsafety-grade and nonseismic systems; >

7. The adequacy of operating procedures and operator training with respect to limitations of the instrumentation 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 factors consultant, to perform the preliminary assessment.

General Physics has been involved in similar audits at the following near-term oparating license BWE's: LaSalle, Susquehanna, and Zimmer.

I l

1

_ SHOREHAM DSAR-  !

d The criteria and checklists used at Shoreham considered the draft 4 NUREG/CR-1580 " Human Engineering Guide to Control Room Evaluation" and the BWR Owners' Group Control Room Human Factor i Review guidelines.

The preliminary assessment report detailing the resulting  ;

findings and a schedule for correcting deficiencies was submitted to the NRC March 12, 1981, 6NRC-536.

i NRC performed a five day onsite review / audit of the Shoreham  ;

control room beginning March 30, 1981. A final report of their findings (CRDR/A report) was issued May 10, 1981.

A Detailed Control Room Design Review (DCRDR) has been performed, j The items identified by this review have been placed on hold due i to the LILCO and New York State Shoreham settlement.

I.D.2 Plant Safety Parameter Display Console NRC Position In accordance with Task Action Plan I.D.2, Plant Safety Parameter Display Console (NUREG-0660), each applicant and licensee shall '

g-' install a safety parameter display system (SPDS) that will t display to operating personnel a minimum set of parameters which define the safety status ofthe plant. This can be attained through continuous indicction of direct and derived variables as necessary to assess plant safety statue.

l These requirements are defined in NUREG-0696 which was published J in February 1981. l l

LILCO Position l since the shoreham Nuclear Power Station will not be operating l and because its reactor is defueled, the SPDS is not needed, l 1

l I.G.1 Training Requirements Durina Low Power Testing U l

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 A low power test program will not be conducted. This item is not

() applicable to LILCo.

r

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SHOREHAM DSAR II.B.1 Reactor Coolant System Vents This system is. specific to the reactor coolant system and is not '

needed to support the storage of the fuel in the fuel pool.

II.B.2 Plan _t Shielding to Provide Access to Vital Areas and l 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 (FHA) as described in Regulatory Guide 1.25, and the Liquid Radwaste Tank Rupture Accident. As discussed in Chapters 15 and 11, respectively, +

t

(~T neither of these involve the release of large quantities of .

\~/ radioactivity, and thus there is no need for the Post Accident L Sampling and Analysis Facility (PASF). The other areas suggested '

l 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 clearly no longer required in the defueled condition, as 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 ,

the station vent, the postulated accident would have no affect on general Radwaste Building habitability. Specifically, the Radwaste Control Room would be unaffected, since it is well isolated from the cubicles where the accident could occur. For details, see Section 9.4.3 of the USAR.

l (~h For the FHA and Worst Case Fuel Damage Accidents, the

(_J-'

habitability criteria for full occupancy of the Main Control Room and Technical Support Center are clearly not challenged. The 30-day integrated doses are:

I L

SHOREHAM DSAR Control Room Dose, mrem Whole Body Skin n

Fuel Handling Accident 9.59E-05 2.08E-01 Worst case Fuel Damage 5.92E-02 1.28E+02 Accident Technical Support center Dose, mrom

',- 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'd_pse rate area in the plant will be the refueling floor. With the Reactor Building Normal Ventilation 6 System (RBNVS) operating throughout the event, airborne l concentrations and associated dose rates are quickly dissipated l .in the Reactor Building. The time-dependent dose rates are as l '()_ follows:

Fuel Handling Accident '

Dose Rate, mrem /hr Time After Accident, hrs Whole Body Skin

0. 7.53E-02 7.11E+00 2 3.16E-04 2.98E-02 8 2.32E-11 2.20E-09 Worst Case Fuel Damage Accident  !

Time After Accident, hrs Whole Body Skin l

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 occupancy for the Reactor Building refueling deck ares, are as follows:

Dose, mrem Whole Body Skin Fuel Handling Accident 2.75E-02 2.60E+00

~W6rst Case Fuel Damage 1.70E+01 1.60E+03 The above are clearly of no concern for post-accident plant habitability.

I

f 7_,

SHOREEAM DSAR

(_ i 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 hazardous 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 sampieg the reactor and containment and is not needed to support the safe storage of the fuel in the fuel pool.

1 II.B.4 Training for Mitigating Core Damage This training 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 the fuel pool because the primary containment is not required for l a defueled reactor.

II.D.1 Performance Testing 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 unpressurized.

l 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)-

l II.E.4.1 Containment Dedicated _{tpetrations This system is not needed to support the storage of the fuel in the fuel pool because the primary containment is not required.

l 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 Refer to Chapters 12 and 15 regarding instrumentation requirements for a defueled reactor.

l I

i l

SHOREHAM DSAR 1 i

II.F.2 Identification of and Recovery from Conditions Leading to i Inadequate Core Cooling This item is specific to an operating reactor and is not needed f to support the safe storage of the fuel in the fuel pool.

II.K.l.5 Safety Related Valve Position These valves are not needed to support the storage of the fuel in '

the fuel pool.

II.K.l.10 Operating Status This section applies without change.

(Plant Staff to confirm)

II.K.l.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.

II.K.l.23 RV Level, Procedures The reactor vessel level procedures are not required.

II.K.3.3

• Failure of Power Operated Relief Valve or Safety Valve j tb Close f

This system is not needed because the reactor is not pressurized.

II.K.3.13 Separation of HPCI and RCIC System Initiation Levels -

I 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 HPDI and RCIC Svstems L This system is not needed to support the storage of the fuel in the fuel pool (See item above).

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.)

SHOREHAM DSAR

(:.)

II.K.3.17 Report on Outane of ECC S" stems -Licensee Report and Proposed Technhcal Specifncation Changes

\

This system is not needed to support the storage of the fuel in i L the fuel pool. Refer to Chapters 9 and 15 for system

! requirements. l II.K.3.18 Modification of Automatic Depressurization System I Logic - Feasibility for Increased Diversity for Some Event 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 l the fuel pool.

(CS and LPCI not required for defueled reactor.]

II.K.3.22 Automatic Switchover of Reactor Core Isolation Cooling l System Suction - Verify Procedures and Modify Design This system is not needed to support the storage of the fuel in the fuel pool.

II.K.3.24 Confirm Adequacy of Space Cooling for High-Pressure Coolant Injection and Reactor Core Isolation Cooling Systems 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. ,

IThis 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.

II.K.3.28 Study and Verify Qualification of Accumulators on ADJ _

Valves O This system is not needed because the reactor is unpressurized.

l SHOREHAM DSAR II.K.3.30 Revised Small-Break LOCA Methods to Show Compliance ,

with 10CFR50.46, Apper. dix 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 E'raluation or Anticipated Transients with Single 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. 4 Chapter 15 demonstrates that there are no radiological consequences associated with this event.

II.K.3.45 Evaluation of Depressurization with Other Than Full ADS This system is specific to operating reactor conditions and is O. not needed to support the storage of the fuel in the fuel pool.

II.K.3.46 Response to List of Concerns from ACRS Consultant (Mr.

C. Michelson)

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

! the fuel pool.

l

! [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 Upgrade Emergency 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, according to the NRC/ FEMA Memorandum of Understanding (item III.B). Approval of 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.

Six NRC teams were formed in September 1979 to implement the

" Action Plan for Promptly Improving Emergency Preparedness" (SECY 79-450). That Action Plan identifies the elements required f~e 1

I . - . - - _ _ . _ ._- -. . .

)

r- SHORERAM DSAR  ;

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 criteria and to identify to licensees the areas requiring improvements. This includes an opportunity for expression of concerns by the public through an open meeting. 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 j licensee after the site visit is summarized in a safety l evaluation report. This includes an identification of areas

requiring 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 i

SECY-79-450 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 l

(

but which are adequate to meet the essential planning elements of "NRC Guide and Checklist," NUREG-75/111, and Supplement 1 i 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 reviews by the RAC as a result of State plan concurrence efforts, including critiques of emergency exercises. The results of the Office of Inspection and Enforcement (IE) special team efforts to evaluate Licensee health physics programs 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 reactors.

NRR has sent letters to operating reactors, operating license O applicants, and holders of construction permits requesting information regarding time estimates for evacuation of areas around plants to determine the difficulty of implementing protective measures for the public.

.m .- - . - . -- - - - -- -.

i i

SHOREHAM DSAR Og LILCO Position l 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.l.2 Upgrade License Emergency Response Facilities

• NRC Position Each operating nuclear power plant shall maintain an onsite '

Technical Support Genter (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.

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 emergency 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 manage. t personnel in the TSC and the EOF. The EOF structure shall e 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 provid6 Oe- information needed by Federal, State, and local authorities for implementation of offsite emergency plans in addition to a centralized meeting location for key representatives from the agencies. Recovery operations shall be managed from this facility. Press facilities also may be available at the EOF.

1 SHOREHAM DSAR

() .

LILCO Position l I

The " Permanent Emergency Response Facilities Design Criteria and Description" applicable to this DSAR can be found in the Defueled f Emergency Preparedness Plan which is being submitted separately 1 1

.via SNRC-1651.

The facilities changes principally involve: 1

1. 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 (FHA) , as described in detail in DSAR Section 15.1.36. Because the ,

accident releases are so low, it is'(conservatively) assumed that i the TSC's HVAC system is not isolated (i.e., 7000 cfm of unfiltered intake and exhaust continues throughout the accident).

l A conservative, ground-level X/Q to the TSC intake is' assumed, l () 7.86E-04 seconds per cubic meter.

Whole body gamma and beta doses are due to Kr-85. The gamma 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, Reg.

Guide 1.3. The only radiciodine determined to be in SNPS' core is I-129, with an inventory of approximately 4 millicuries.

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).

l The resulting doses, and the associated GDC 19 Criteria, are as follows:

Dose, rem Whole Body Gamma , Beta Thyroid Results 5.02E-08 1.04E-04 1.21E-07 l

GDC 19 Criteria 25 300 300 III.A.2 Improving Licensee Emergency Preparedness --Long-Term NRC Position Each nuclear facility shall upgrade its emergency plans to provide reasonable assurance that adequate protective measures

- i l

- SHOREHAM DSAR can and will be taken in the event of a radiological emergency.

Specific criteria to meet this requirement are delineated in NUREG-0654 (FEMA-REP-1). " Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and ,

Preparation in Support of Nuclear Power Plants".  !

In accordance with Task Action Plan item III.A.1.1, " Upgrade Emergency Preparedness," each nuclear power facility was required to immediately upgrade its amargency plans with criteria provided October 10, 1979, as revised by NUREG-0654 (FEMA-REP-1, issued .

for interim use and comment, January 1980) . New plans were l submitted by January 1, 1980, using the October 10, 1979 ,

criteria. Reviews were started on the upgraded plans using NUREG-0654. Concomitant to these actions, amendments, were developed to 10CFR part 50 and Appendix E to 10CFR Part 50, to I provide the long-term implementation requirements. These new rules were issued in the Federal Register on August 19, 1980, with an ef fective date of November 3,1980. The revised rules -

delineate requirements for emergency preparedness at nuclear reactor facilities.

NUREG-0654 (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 included in the upgraded emergency plans and, along with the revised rules, provides the meteorological criteria, means for 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 ,

l licensee to provide procedures implementing the upgraded l emergency plans to the NRC for review. Publication of Revision 1 to NUREG-0654 (FEMA-REP-1) , which incorporaten the many public comments received is expected in October 1980. This is the l

document that will be used by NRC and FEMA in their evaluation of emergency plans submitted in accordance with the new NRC rules.

NUREG-0654, Revision 1; NUREG-0696, " Functional 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 plans.for installation of prompt notification systems. These revised criteria and rules supersede previous Commission guidance for the upgrading of emergency preparedness at nuclear power facilities.

LILCO__ Position 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 USAR.

~ _ _ .__ _ _ _ __

e E

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 outs.de i containment that would or could contain highly radioactive fluids during a serious transient or accident to as-3 low-as-practical levels. This program shall include the 7

following:

Immediate leak reduction (a) Implement all practical isak reduction measures for all

- systems that could carry radioactive fluid outside of

%F _ containment.

h

^* (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 m

tests at intervals not to exceed each refueling cycle.

g Applicants shall provide a summary description, together with initial leak-test results, of their program to reduce leuage 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 ere as follows (any other plens system which hac similar functions or postaccident characteristics even though not specified herein, should be included):

Residual heat removal (RHR)

Containment spray recirculation High-pressure injection recirculation Containment and primary coolant sampling Reactor core isolation cooling Makeup and 1-etdown (PWR'u only)

Waste gas (includes headers and cover gas system outside of S containment in addition to decay or storage system)

Include a list of systems containing radioactive materials which are excluded frcm program and provide justification for exclusion.

,. s i

SHOREHAM DSAR

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 M

The purpose of this program is to minimize leakage of primary coolant sources outside of the primary containment. Since in the FIPS condition there is no " primary coolant" per se, the leakage

. prevention program becomes irrelevant and unnecessary.

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

-under accident conditions must include the use of portable instruments using sample media that will collect iodine

, selectively over xenon (e.g., silver zeolite) for the following reasons:

a. The physical size of the auxiliary and/or fuel handling building precludes locating stationery monitoring instrumentation 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 req? tired 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 a low-background, e

low-contamination area for further analysis. Normally, counting rooms in auziliary buildings will not have sufficiently low backgrounds for such analyses following an accident. In the low j background area, the sample should first be purged of any '

entrapped noble gases using nitrogen gas or clean air free of

3 .)

p SHOREHAM-D$AR-l noble 1 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 to' sample all~ vital areas.

LILCO Position.

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 millicuries of I-129:(total' core). There is no measurable quantity of radiciodine elsewhere in the Reactor, Radwaste, or Turbine . .I Buildings. As such, inplant measurement of radioiodine during and after an accident is unnecessary, and-no provisions are made to perform such analyses."

III.D.3.4 Control Room Habitability NRC Position In accordance with the Task Action Plan item III.D.3.4~and' l control-room habitability, licensees shall assure that control l room operators will be adequately protected against the effects- l of accidental-release of toxic and radioactive gases and that the nuclear power plant can be safely operated or shut down under .

design basis accident conditions (Criterion 19,-" Control Room,"

.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.

Licensees may establish this basis by referencing past submittale to the NRC and/or oroviding new or additional information to supplement past schmittals.

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 j f

' 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:

1. Regulatory Guide 1.78, " Assumptions for evaluating the Habitability of Regulatory Power Plant Control Room O During a Postulated Hazardous Chemical Release",

1974.

June

2. Regulatory Guide 1.95, " Protection of Nuclear. Power Plant' Control Room Operators Against an Accident Cholorine Release"; and, 4

n. -

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'h F

SHOREHAM DSAR

3. K. G. Murphy and K. 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- 1 Room Following a Postualted' Accidental Release.

1 In'providing the basis for the habitability finding, licensees may reference their past submittals. Licensees should however 3 ensure that submittals. reflect the current facility design and

'that the information requested in Attachment 1 is provided. All licensees with control rooms that do not meet the criteria of the above-listed referencas, Standard Review Plans, Regulatory Guides, and other references shall perform the necessary i

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 and direct radiation resulting from c design-basis accidents. The toxic gas accident analysis should E 'be performed for all potential hazardous chemical releases occurring either on the site or within 5 miles of the plant-site 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 should be for the loss-of-coolant accident (LOCA) containment leakage and engineered safety feature (ESP) leakage contribution outside containment as described in Appendix A and B of Standard Review Plan Chapter 15.6.5. In addition, boiling-water reactor (BWR) facility evaluations should add any leakage from the main steam isolation valves (MSIV) (i.e., valve-stem leakage, valve seat

-leakage, main steam isolation valve leakage control system release) to the containment leakage and ESF leakage followimg a LOCA. This should not be construed as altering the staff recommendations in Section D of Regulatory Guide 1.96 (Rev. 2) regarding MSIV leakage-control systems. Other DBAS should be 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

/'N by General Design Criterion 19, the licensee should submit

(_/ sufficient information needed for an independent evaluation of the adequacy of the habitability systems. Attachment i lists the information that should be provided along with the licensee's evaluation.

^' .

SHOREHAM DSAR

' ATTACHMENT 1 INFORMATION REQUIRED FOR CONTROL-ROOM HABITABILITY EUKIUATION (1) Control-room mode of operation, i.e., pressurization and 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) total amount and size of container

.(b) - closest distance from control-room air intake

. .n Yk

.v.

SHOREHAM DSAR (4) Offsite 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 emergency 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 closure times, and filter testing requirements.

L h LILCO Position 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 e

L radioactivity released during the design basis accident (the fuel L handling accident), the control room's remote intakes and standby L 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 Design

.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 standby filtration system are no longer required. Inlet duct instrumentation is no longer required as well. As such, discussion of these items in USAR Section III.D.3.4 no longer applies. .During the design basis accident, the control room HVAC system will continue to function as during normal plant

1 operations.

A.J Design Evaluation This section is as described in USAR Section III.D.3.4, except that the remote intakes and standby filtration systema are no

SHOREHAM DSAR-longer required. Also, as per DSAR-Section 9.2.9, only one chilled water system is' requi' red with the spent fuel in the pool.

. Tests and Inspections Tests land. inspections of the control room HVAC system are as described in.USAR Section III.D.3.4, except that the standby filtration systems are no longer required.

Standby Charcoal Filtration Trains This equipment is no longer required.

m i

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