ML19256A720

Amend 26 to Application for License to Mfg Floating Nuclear Plants.Amend Deals Soley W/Subj Matter of Plant Design Rept

ML19256A720
Person / Time
Site:	Atlantic Nuclear Power Plant
Issue date:	01/03/1979
From:	OFFSHORE POWER SYSTEMS (SUBS. OF WESTINGHOUSE ELECTRI
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ML19256A716	List: ML19256A715 ML19256A720
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NUDOCS 7901090257
Download: ML19256A720 (100)

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OFFSHORE POWER SYSTEMS DOCKET NO. STN 50-437 INSTRUCTIONS FOR ENTERING AMENDMENT NO. 26 IN THE PLANT DESIGN REPORT

1. Replace master Table of Contents pages iii, iv in the front of each Plant Design Report volume.

2. Remove and insert pages in the Plant Design Report in accordance with the following tabulation:

Remove Pace (s) Insert Page(s)

Chaoter 2 2.9-7 thru 2.9-8a 2.9-7, 8

'2.9-8a 2.9-15 2.9-15 Apoendix 2a 2A-51, 52 2A-51, 52 2A-137, 137a 2A-137, 137a Chapter 3 3.5-5 thru 3.5-8 3.5-5 thru 3.5-8 Appendix 3D 3D-1 thru 3D-6a 3D-1 thru 3D-6 3D-6a, 6b 3D-11 3D-11 thru 3D-11h 3D-18, 19 3D-18, 19

--- Figures 3D-1, 3D-2, 3D-3 Apoendix 3G 3G-1 thru 3G-5 ---

Amendment 26 Instructier- -1 January 3, 1979 7901096TO

Remove Page(s) Insert Page(s)

O Chapter 5 5-iii, iv 5-iii, iv 5.2.1, la 5.2-1, la 5.2-le thru 5.2-1(1) 5.2-le thru 5.2-1L Chapter 8 8.1-5, 6 8.1-5, 6 Chapter 9 9-vii, viii 9-vii, viii 9-xiii, xiv 9-xiii, xiv 9.4-2a thru 9.4-29 9.4.2a thru 9.4-2c Aopendix B

--- B.18.1-1 thru B.18.1-8

--- B.18.2-1 thru B.18.2-5

.-tror' ions - 2 Amendment 26 9

January 3, 1979

TABLE OF C0tiTEllTS (00tlT)

Chapter Title .

Page 3.11 Environmental Design of Mechanical and Electrical Equipment 3.11-1 3.12 Platform Structure 3.12-1 Appendix 3A Description of Chapter 3 Computer Programs Appendix 3B Codes and Standards Appendix 3C Guidelines for Combining Events and Loads Appendix 3D Location of Information Relating to Safety Guides Appendix 3E Turbine riissile Analysis Appendix 3F Surrary of Buckling Criteria for the Steel Containment Shell Appendix 3G (Deleted) 26 4 REACTOR 4.1 Summary Description 4.1-1 4.2 Mechanical Design 4.2-1 4.3 fluclear Design 4.3-1 4.4 Thermal and Hydraulic Design 4.4-1 5 REACTOR C00LAtlT SYSTEM 5.1 Sunnary Description 5.1-1 C.2 Integrity of the Reactor Coolant System Boundary 5.2-1 5.3 Thermal Hydraulic System Design 5.3-1 5.4 Reactor Vessel and Appurtenances 5.4-1 5.5 Component and Subsystem Design 5.5-1 5.6 Instrument Application 5.6-1 Amendment 26 jjj January 3,1979

TABLE OF C0tiTEtiTS (C0flT)

Chapter Title M 6 ENGIriEERED SAFETY FEATURES 6.1 General 6.1-1 6.2 Containment Systems 6.2-1 6.3 Emergency Core Cooling Systems 6.3-1 6.4 Other Engineered Safety Features 6.4-1 6.5 Habitability Systems 6.5-1 7 INSTRUMEtiTATION AND CONTROLS 7.1 Introduction 7.1-1 7.2 Reactor Trip S,ystem RESAR 7.3 Engineered Safety Features Actuation System 7.3-1 7.4 Systems Required for Safe Shutdown 7.4-1 7.5 Safety Related Display Instrumentation 7.5-1 7.6 All Other Systems Required for Safety RESAR 7.7 Plant Control Systems 7.7-1 7.8 Fire Detecticn and Alarms 7.8-1 7.9 Hull Leakage Surveillance 7.9-1 8 ELECTRICAL POWER 8.1 Introduction 8.1-1 8.2 Offsite Power System 8.2-1 8.3 Onsite Power System 8.3-1 8.4 Engineered Safety Feature Motors 8.4-1 22 O

Amendment 22 iv Mar-h 24, 1976

TABLE OF CCNTErlTS (CONT)

Chapter Title Pace 3.11 Environmental Design of Mechanical and Electrical Equip lent 3.11-1 3.12 Platform Structure 3.12-1 Appendix 3A Description of Chapter 3 Computer Programs Appendix 38 Codes and Standards Appendix 3C Guidelines for Combining Events and Loads Appendix 3D Location of Information Relating to Safety Guides Appendix 3E Turbine Missile Analysis Appendix 3F Summary of Buckling Criteria for the Steel Containment Shell Appendix 3G (Deleted) 26 4 REACTOR 4.1 Summary Description 4.1-1 4.2 Mechanical Design 4.2-1 4.3 Nuclear Design 4.3-1 4.4 Thermal and Hydraulic Design 4.4-1 5 REACTOR COOLANT SYSTEM 5.1 Sumary Description 5.1-1 5.2 Integrity of the Reactor Coolant System Boundary 5.2-1 5.3 Thermal Hydraulic System Design 5.3-1 5.4 Reactor Vessel and Appurtenances 5.4-1 5.5 Component and Subsystem Design 5.5-1 5.6 Instrument Application 5.6-1 Amendment 26 5jj January 3, 1979

TABLE OF CONTEt4TS (CONT)

Chapter Title Page 6 ENGINEERED SAFETY FEATURES 6.1 General 6.1-1 6.2 Containment Systems 6.2-1 6.3 Emergency Core Cooling Systems 6.3-1 6.4 Other Engineered Safety Features 6.4-1 6.5 Habitability Systems 6.5-1 7 INSTRUMENTATION AND CONTROLS 7.1 Introduction 7.1-1 7.2 Reactor Trip System RESAR 7.3 Engineered Safety Features Actuation System 7.3-1 7.4 Systems Required for Safe Shutdown 7.4-1 7.5 Safety Related Display Instrumentation 7.5-1 7.6 All Other Systems Required for Safety RESAR 7.7 Plant Centrol Systems 7.7-1 7.8 Fire Detection and Alarms 7.8-1 7.9 Hull Leakage Surveillance 7.9-1 8 ELECTRICAL POWER 8.1 Introduction 8.1-1 8.2 Offsite Power System 8.2-1 8.3 Onsite Power System 8.3-1 8.4 Engineered Safety Feature Motors 8.4-1 22 O

Amendment 22 iv March 24, 1976

TABLE OF CONTEflTS (C0flT)

Chapter Title Pace 3.11 Environmental Design of Mechanical and Electrical Equipment 3.11-1 3.12 Platform Structure 3.12-1 Appendix 3A Description of Chapter 3 Computer Programs Appendix 3B Codes and Standards Appendix 3C Guidelines for Combining Events and Loads Appendix 3D Location of Information Relating to Safety Guides Appendix 3E Turbine liissile Analysis Appendix 3F Sumary of Buckling Criteria for the Steel Containment Shell Appendix 3G (Deleted) 26 4 REACTOR 4.1 Summary Description 4.1-1 4.2 Mechanical Design 4.2-1 4.3 fluclear Design 4.3-1 4.4 Themal and Hydraulic Design 4.4-1 5 REACTOR COOLANT SYSTEM 5.1 Sumary Description 5.1-1 5.2 Integrity of the Reactor Coolant System Boundary 5.2-1 5.3 Thermal liydraulic System Design 5.3-1 5.4 Reactor Vessel and Appurtenances 5.4-1 5.5 Component and Subsystem Design 5.5-1 5.6 Instrument Application 5.6-1 Amendment 26 jjj January 3,1979

TABLE OF C0tiTEtiTS (C0 tit)

Chapter Title Pg 6 ENGINEERED SAFETY FEATURES 6.1 General 6.1-1 6.2 Containment Systems 6.2-1 6.3 Emergency Core Cooling Systems 6.3-1 6.4 Other Engineered Safety Features 6.4-1 6.5 Habitability Systems 6.5-1 7 INSTRUMENTATION AND CONTROLS 7.1 ( trocuction 7.1-1 7.2 Re ctor Trip System RESAR 7.3 Engineered Safety Features Actuation System 7.3-1 7.4 Systems Required for Safe Shutdown 7.4-1 7.5 Safety Related Display Instrumentation 7.5-1 7.6 All Other Systems Required for Safety RESAR 7.7 Plant Control Systems 7.7-1 7.8 Fire Detection and Alarms 7.8-1 7.9 Hull Leakage Surveillance 7.9-1 8 ELECTRICAL POWER 8.1 Introduction 8.1-1 8.2 Offsite Power System 8.2-1 8.3 Onsite Power System 8.3-1 8.4 Engineered Safety Feature Motors 8.4-1 22 O

Amendment 22 iv March 24, 1976

TABLE OF C0fiTEtlTS (C0flT)

Chapter _ Title Pace 3.11 Environmental Design of Mechanical and Electrical Equipment 3.11-1 3.12 Platform Structure 3.12-1 Appendix 3A Description of Chapter 3 Computer Programs Appendix 3B Codes and Standards Appendix 3C Guidelines for Combining Events and Loads Appendix 3D Location of Information Relating to Safety Guides Appendix 3E Turbine flissile Analysis Appendix 3F Summary of Buckling Criteria for the Steel Containment Shell Appendix 3G (Deleted) 26 4 REACTOR 4.1 Summary Description 4.1-1 4.2 Mechanical Design 4.2-1 4.3 tiuclear Design 4.3-1 4.4 Thermal and Hydraulic Design 4.4-1 5 REACTOR C00LAtlT SYSTEM 5.1 Summary Description 5.1-1 5.2 Integrity of the Reactor Coolant System Boundary 5.2-1 5.3 Thermal Hydraulic System Design 5.3-1 5.4 Reactor Vessel and Appurtenances 5.4-1 5.5 Component and Subsystem Design 5.5-1 5.6 Instrument Application 5.6-1 Amer.jment 26 jjj January 3, 1979

TABLE OF CONTEtiTS (CONT)

Chapter Title Pm 6 ENGINEERED SAFETY FEATURES 6.1 General 6.1-1 6.2 Containment Systems 6.2-1 6.3 Emergency Core Cooling Systems 6.3-1 6.4 Other Engineered Safety Features 6.4-1 6.5 Habitability S3, ams 6.5-1 7 INSTRUMENTATION AND CONTROLS 7.1 Introduction 7.1-1 7.2 Reactor Trip System RESAR 7.3 Engineered Safety Features Actuation Sys tem 7.3-1 7.4 Systems Required for Safe Shutdown 7.4-1 7.5 Safety Related Display Instrumentation 7.5-1 7.6 All Other Systems Required for Safety RESAR 7.7 Plant Control Systems 7.7-1 7.8 Fire Detection and Alarms 7.8-1 7.9 Hull Leakage Surveillance 7.9-1 8 ELECTRICAL POWER 8.1 Introduction 8.1-1 8.2 Offsite Power System 8.2-1 8.3 Onsite Power System 8.3-1 8.4 Engineered Safety Feature Motors 8.4-1 22 O

Amendment 22 iv March 24, 1976

TABLE OF C0flTErlTS (C0tlT)

Chapter Title Page 3.11 Environmental Design of Mechanical and Electrical Equipment 3.11-1 3.12 Platform Structure 3.12-1 Appendix 3A Description of Chapter 3 Computer Programs Appendix 3B Codes and Standards Appendix 3C Guidelines for Combining Events and Loads Appendix 3D Location of Information Relating to Safety Guides Appendix 3E Turbine itissile Analysis Appendix 3F Summary of Buckling Criteria for the Steel Containment Shell Appendix 3G (Deleted) 26 4 REACTOR 4.1 Summary Description 4.1-1 4.2 Mechanical Design 4.2-1 4.3 fluclear Design 4.3-1 4.4 Thermal and Hydraulic Design 4.4-1 5 REACTOR C00LAtlT SYSTEM 5.1 Sumary Description 5.1-1 5.2 Integrity of the Reactor Coolant System Boundary 5.2-1 5.3 Thermal Hydraulic System Design 5.3-1 5.4 Reactor Vessel and Appurtenances 5.4-1 5.5 Component and Subsystem Design 5.5-1 5.6 Instruuent Application 5.6-1 Amendment 26 iii January 3,1979

TABLE OF CONTENTS (C0tlT)

Chapter Title Pace .

6 ENGINEERED SAFETY FEATURES 6.1 General 6.1-1 6.2 Containment Systems 6.2-1 6.3 Emergency Core Cooling Systems 6.3-1 6.4 Other Engineered Safety Features 6.4-1 6.5 Habitability Systems 6.5-1 7 INSTRUMENTATION AND CONTR0 Q 7.1 Introduction 7.1-1 7.2 Reactor Trip System RESAR 7.3 Engineered Safety Features Actuation System 7.3-1 7.4 Systems Required for Safe Shutdown 7.4-1 7.5 Safety Related Display Instrumentation 7.5-1 7.6 All Other Systems Required for Safety RESAR 7.7 Plant Control Systems 7.7-1 7.8 Fire Detection and Alarms 7.8-1 7.9 Hull Leakage Surveillance 7.9-1 8 ELECTRICAL POWER 8.1 Introduction 8.1-1 8.2 Offsite Power System 8.2-1 8.3 Onsite Power System 8.3-1 8.4 Engineered Safety Feature Motors 8.4-1 22 O

Amendment 22 iv March 24,1976

TABLE OF CONTENTS (C0rlT)

Chapter Title Page 3.11 Environmental Design of Mechanical ind Electrical Equipment 3.11-1 3.12 Platform Structure 3.12-1 Appendix 3A Description of Chapter 3 Computer Programs Appendix 3B Codes and Standards Appendix 3C Guidelines for Combining Events and Loads Appendix 3D Location of Information Relating to Safe'.y Guides Appendix 3E Turbine Missile Analysis Appendix 3F Summary of Buckling Criteria for the Steel Containment Shell Appendix '3G (Deleted) 26 4 REACTOR 4.1 Summary Description 4.1-1 4.2 Mechanical Design 4.2-1 4.3 Nuclear Design 4.3-1 4.4 Thermal and Hydraulic Design 4.4-1 5 REACTOR COOLANT SYSTEM 5.1 Surrnary Description 5.1-1 5.2 Integrity of the Reactor Coolant System Boundary 5.2-1 5.3 Thermal Hydraulic System Design 5.3-1 5.4 Reactor Vessel and Appurtenances 5.4-1 5.5 Component and Subt.. tem Design 5.5-1 5.6 Instrument Application 5.6-1 Amendment 26 jjj January 3, 1979

TABLE OF C0tlTEtiTS (C0tlT)

Chapter Title M 6 ENGIfiEERED SAFETY FEATURES 6.1 General 6.1-1 6.2 Containment Systems 6.2-1 6.3 Emergency Core Cooling Systems 6.3-1 6.4 Other Engineered Safety Features 6.4-1 6.5 Habitability Systems 6.5-1 7 INSTRUMEllTATI0tl AND CONTROLS 7.1 Introduction 7.1-1 7.2 Reactor Trip System RESAR 7.3 Engineered Safety Features Actuation System 7.3-1 7.4 Systems Required for Safe Shutdown 7.4-1 7.5 Safety Related Display Instrumentation 7.5-1 7.6 All Other Systems Required for Safety RESAR 7.7 Plant Control Systems 7.7-1 7.8 Fire Detection and Alarms 7.8-1 7.9 Hull Leakage Surveillance 7.9-1 8 ELECTRICAL POWER 8.1 Introduction 8.1-1 8.2

~

Offsite Power System 8.2-1 8.3 Onsite Power System 8.3-1 8.4 Engineered Sefety Feature Motors 8.4-1 22 O

Amendment 22 iv March 24, 1976

TABLE OF CONTENTS (CONT)

Chapter Title Page 3.11 Environmental Design of Mechanical and Electrical Equipment 3.11-1 3.12 Platform Structure 3.12-1 Appendix 3A Description of Chapter 3 Computer Programs Appendix 3B Codes and Standards Appendix 3C Guidelines for Combining Events and Loads Appendix 3D Location of Information Relating to Safety Guides Appendix 3E Turbine tiissile Analysis Appendix 3F Summary of Buckling Criteria for the Steel Containment Shell Appendix 3G (Deleted) 26 4 REACTOR 4.1 Summary Description 4.1-1 4.2 Mechanical Design 4.2-1 4.3 Nuclear Design 4.3-1 4.4 Thermal and Hydraulic Design 4.4-1 5 REACTOR COOLANT SYSTEM 5.1 Summary Description 5.1-1 5.2 Integrity of the Reactor Coolant System Boundary 5.2-1 5.3 Thermal Hydraulic System Design 5.3-1 5.4 Reactor Vessel and Appurtenances 5.4-1 5.5 Component and Subsystem Design 5.5-1 5.6 Instrument Application 5.6-1 Amendment 26 jjj January 3, 1979

TABLE OF CONTEtiTS (CONT)

Chapter Title Pace 6 ENGIllEERED SAFETY FEATURES 6.1 General 6.1-1 6.2 Containment Systems 6.2-1 6.3 Emergency Core Cooling Systems 6.3-1 6.4 Other Engineered Safety Features 6.4-1 6.5 Habitability Systems 6.5-1 7 INSTRUMENTATIO*1 AND CONTROLS 7.1 Introduction 7.1-1 7.2 Reactor Trip System RESAR 7.3 Engineered Safety Features Actuation System 7.3-1 7.4 Systems Required for Safe Shutdown 7.4-1 7.5 Safety Related Display Instrumentation 7.5-1 7.6 All Other Systems Required for Safety RESAR 7.7 Plant Control Systems 7.7-1 7.8 Fire Detection and Alarms 7.8-1 7.9 Hull Leakage Surveillance 7.9-1 8 ELECTRICAL PO!!ER 8.1 Introduction 8.1-1 8.2 Offsite Power System 8.2-1 8.3 Onsite Power System 8.3-1 8.4 Engineered Safety Feature Motors 8.4-1 22 O

Amendment 22 iv March 24,1976

TABLE OF C0flTEflTS (C0flT)

Chapter Title Pace 3.11 Environmental Design of Mechanical and Electrical Equipment 3.11-1 3.12 Platform Structure 3.12-1 Appendix 3A Descriptio of Chapter 3 Computer Programs Appendix 3B Codes and Standards Appendix 3C Guidelines for Combining Events and Loads Appendix 3D Location of Information Relating to Safety Guides Appendix 3E Turbine tiissile Analysis Appendix 3F Summary of Buckling Criteria for the Steel Containment Shell Appendix 3G (Deleted) 26 4 REACTOR 4.1 Summary Description 4.1-1 4.2 Mechanical Design 4.2-1 4.3 fluclear Design 4.3-1 4.4 Thernal and Hydraulic Design 4.4-1 5 REACTOR C00LAflT SYSTEM 5.1 Sumary Description 5.1-1 5.2 Integrity of the Reactor Coolant System Boundary 5.2-1 5.3 Thermal Hydraulic System Design 5.3-1 5.4 Reactor Vessel and Appurtenances 5.4-1 5.5 Component and Subsystem Design 5.5-1 5.6 Instrument Application 5.6-1 Amendment 26 jjj January 3, 1979

TABLE OF CONTEtlTS (C0tlT)

Chapter Title Pg 6 EllGIflEERED SAFETY FEATURES 6.1 General 6.1-1 6.2 Containment Systems 6.2-1 6.3 Emergency Core Cooling Systems 6.3-1 6.4 Other Engineered Safety Features 6.4-1 6.5 Habitability Systems 6.5-1 7 INSTRUMEf1TATI0tl Afl0 C0t1TROLS 7.1 Introduction 7.1 '

7.2 Reactor Trip System RESAR 7.3 Engineered Safety Features Actuation System 7.3-1 7.4 Systems Required for Safe Shutdown 7.4-1 7.5 Safety Related Display Instrumentation 7.5-1 7.6 All Other Systems Required for Safety RESAR 7.7 Plant Control Systems 7.7-1 7.8 Fire Detection and Alarms 7.8-1 7.9 Hull Leakage Surveillance 7.9-1 8 ELECTRICAL POWER 8.1 Introduction 8.1-1 8.2 Offsite Power System 8.2-1 8.3 Onsite Power System 8.3-1 8.4 Engineered Safety Feature Motors 8.4-1 22 O

Amendment 22 iv March 24, 1976

small in comparison to the mass of the ship and the analysis is based on maximum possible elastic collision', the results apply to smaller missiles as well.

2.9.2 REQUIREMENTS IMPOSED BY P0TENTIAL AIRCRAFT CRASH ON THE PLANT Requirments The owner must provide a site where the probability of a crash producing unacceptable plant damage is of the order of 10 7 Der year or less. l8 Justi fica tion The oasis for the probability value of the order of 10~70er year is discussed in the l8 introduction to section 2.9. Example calculations for representative sites are contained in Appendix 28.

Methodoloay Two methodologies have been utilized for determining aircraft accident probabil i ty. One is for crashes associated with overflights (along flight corridors)

Amendment 8 2.9-7 May 15, 1974

nd one is for those crashes attributed to take-offs and landings. The latter were considered only for tnose sites in proximity to airfields.

These two methodologies are described in detail in Appendix 28. Briefly, they are based either on records of crash rates per year, per mile flown (for the overflight case), or crash rates per aircraft movement for the near-airfield case. Both methodologies contain a distribution factor for the spatial distribution (away from a corridor, or away from an airfield).

2.9.3 REQUIREMEflTS IMPOSED BY P0TEllTIAL EXPLOSION NEAR THE PLANT WITH RESULTING MISSILES Requirement Regulatory Guide 1.91, Revision 1 requires that nuclear facilities be able to withstand the possible effects of explosions occuring on nearby transporta-tion routes. The Guide specifies the following three options for complying with this requirement:

0

1. Demonstrate that the peak positive incident overpressure resulting from cny postulated transportation accident in the vicinity of the plant will not exceed 1 psi, or

2. Demonstrate that the probability of the above overpressure limit being exceeded is acceptably low (less than 10-7 per year; 10-6 if based on demonstrably conservative assumptions), or Amendment 26 January 3, 1979 2.9-8

3. Demonstrate that all safety relat1d structures can withstand the explosion loads, if these loads exceed 1 psi incident overpressure. 26 The site envelope (Table 2.1-1) requires that each plant owner comply with either option 1 or 2 (above) by demonstrating that at each Fi4P site the probability of a shipment of munitions or petroleum products becoming involved in an accident near the site, with a subsequent detonation which produces a reflected over-pressure of 2 psi (I) or greater at the plant's Category I structures is of l 26 the order of 10~7 per year or less. For fuel supply barges, either the size of the tank compartments must be limited to 325 barrels or the owner must utilize off-loading procedures which preclude the possibility of an explosion (2). l 26 Justification The basis for the probabilities value of the order of 10-7 per year is dis-cussed in the introduction to Section 2.9. A review of waterborne commerce of explosives shows munitions are the only comodity shipped as explosives in (I) Instead of the 1 psi peak positive incident overpressure criterion stated in the Regulatory Guide, a 2 psi maximum reflected overpressure criterion is utilized for the Floating Nuclear Plant. This criterion in no way alters the effect of the regulatory guide criterion, since both values result in virtually identical " exclusion distances". This can be seen by comparing the values of K (scaled ground distance) obtained for 2 psi 26 reflected overpressure from figure 4.12 of Reference 4, with the values of K suggested by the Regulatory Guide.

(2) Appendix 2A contains analyses of two accidents in.olving detonation of explosive cargo in the plant vicinity: (1) A fuel supply barge with 325 barrel tank compartment, at 150 feet from the plant and (2) A cargo ship carrying munitions or aertroleum products, at various distances from the plant site.

Amendment 26 January 3,1979 2.9-8a

assumed that the total cargo was discharged inside the breakwater.

References (1 ) Wash-1270, Technical Report on Anticipated Transients Without Scram for Water-Cooled Power Reactors," Regulatory Staff, U.S. Atomic Energy Comission, September,1973.

(2) Corps of Engineers, " Waterborne Commerce Statistics 1970".

(3) Report from Jones, Bardelmeier and Company LTD., Bulk Shippings Consultants, "A Study of the Offshore Trade Flows and Traffic Patterns of Nine !!azardous Liquid Chemicals on the U.S. East and Gulf Coasts for Offshore Power Systems," October,1973.

(4) Department of the Army Technical Manual TM 5-1300, " Structures 26 to Resist the Effects of Accidental Explosions," June 1969.

Amendment 25 January 3,1379 2.9-15

The relative numbers of MSC and Navy vessels loaded at Earle for 1968 through May 1973 are given in Table 2A-12. It is believed that all of the 3 MSC vessels headed south while most of the Navy vessels headed north or east (3) ,

Two further types of shipment are made from Earle. A few foreign vessels load partial loads of ammunition under military assistance programs for U. S. allies. These are not included in the number of loadings tabulated.

Also V. S. warships replenish their amunition magazines at Earle. Warships are not considered a threat as munitions carriers because the quantities of munitions are much smaller than those on the cargo vessels and the maga-zines are well protected against impact damage or fire.

Munitions shipments in transit are not regarded as hazardous cargo by the U. S. Coast Guard and the ships are considered as normal cargo vessels.

Safe distances are, however, specified for loading and unloading or while the ships are at anchor in port. The stowage of the cargo is subject to Department of Defense regulations and these have recentl" been made more stringent following the loss of the " Badger State" in December 1969. This accident is discussed below.

2A.4.3.2 Recent Munitions Ship Accidents There have been only two accidents to munitions ships in the last four years. Both were vessels on charter to the MSC. 3 Amendment 3 2.A-51 January 28, 1973

On ftarch 5,1969, the AMERICAN PRODUCER carrying miscellaneous military munitions collided with a reinforced concrete pier in San Francisco harbor.

The vessel was moving at 11 knots and turned to avoid a vessel that failed to give right of way. The horizontal deck of the pier cut a slot in the vessel from the bow to 60 feet back. The vessel was declared a total loss by the insurers. There was no explosion.

On December 26, 1969, the BADGER STATE, a C2 cargo vessel, carrying bombs, sank in the Pacific in heavy weather. The ship foundered after the bombs broke out of their stowage racks and began smashing around in the holds.

The ship was finally abandoned after a low order (i.e., slow burning) detonation blew a hole through the side of one of the holds. No one was killed as a direct result of this explosion. There was no mass detonation of the cargo. As noted above, stowage requirements have been made stricter as a result of this accident.

In the last three years, 829 shipments were made from the four major munitions terminals (Table 2A-12).

2A.4.4 Detonation Analysis 2A.4.4.1 Maximum Permissible Detonation Without Serious Plant Damage The Applicant concurs with the judgement of the NRC Staff (as expressed in Regulatory Guide 1.91) that no significant plant damage would result from an explosion producing an incident overpressure of 1 psi. As noted in Section 26 2.9.3, an incident overpressure of 1 psi is equivalent to a reflected over-pressure of 2 psi.

Amendment 26 January 3, 1979 2A-52

to the off-loading dock is the spent fuel pit missile barrier which would be about 142 feet from the barge. This structure would not be damaged 26 significantly by a pres. -a pulse of 2.0 psi (reflected). The analysis indicates that tank compartments with volumes up to 325 barrels will not produce excessive reflected overpressure at the fuel pit missile barrier.

Overpressure values are plotted versus distance and time for both analysis cases (10,000 barrel barge and 325 barrel compartments) in Figure 2A-42 and 2A-43, respectively. The TNT equivalents from the barge-tank explosion are 895 lbs and 30 lbs for the two cases. Methods that might be used for ensuring that an explosion would either not occur during off-loading or would not produce excessive overpressures on plant structures are the responsibility of the site owner. They could include inerting of the fuel barge tanks dur-ing off-loading, separation between the plant and fuel supply barge during fuel off-loading, or other appropriate measures.

The particular approach used to prevent an explosion of unacceptable magni-tude near the plant during off-load'ng of a fuel resupply barge is the re-sponsibility of the site applicant. The applicant must either limit the volume of the individual tanks on the resupply barge to 325 barrels or pro-vide off-loading procedures which preclude the possibility of an explosion during off-loading.

Amendment 26 January 3,1979 2A-137

2A.8.3.2 Missile Generation Analysis Fuel supply barge debris or fragments resulting from the fuel-air explosion are missiles and have potential for damaging the plant if the explosion were to occur near the plant. The missiles resulting from a fuel-air explosion of an integral tank will consist of steel plates, deck fittings, man-hole covers, bitts, chocks, anchors and piping. This debris will be considered to fall into three classifications:

a. Small parts weighing less than 20 pounds consisting of bolts, rivets, rail fittings, small plumbing fittings and small plate fragments.

b. Medium sized parts weighing between 20 and 200 pounds consisting of manhole covers, piping and smaller deck fittings, and plate fragments.

c. Large parts weighing more than 200 pounds and consisting of capstans, 23 bitts, chocks and large sections of steel plate structure.

The nature of the fuel-air explosion is such that failure of the supply barge structure will be similar to the failure of a pressure ven?1 rather than the shock fragmentation of materials which is experienced with the detonation Amendment 23 September 10, 1976 2A-137a

3.5.2 Postulated Sources and Protection Criteria for Missiles External to the Containment Five sources of missiles have been postulated external to the containment.

1. Tornado and waterspout borne missiles

2. Turbine-generator missiles

3. Missiles resulting from the crash of a helicopter

4. Missiles resulting from failure of compressed gas cylinders

5. Missiles resulting from ship explosions near the plant.

The following criteria will be met either by eliminating the missile threat by protective measures (as in the case of tornado missiles), or by reducing the probability of missile damage to a very low level, of the order of 10-7 per year as in the case of turbine missiles.

The FNP cesign meets the following criteria: 26

1) Protection shall be provided against excessive flooding of the Float-ing Nuclear Plant.

2) Protection shall be provided against potential missiles that could cause a LOCA.

Amendment 26 January 3,1979 3.5-5

3) Protection shall be provided against potential missiles that could jeopardize functions necessary to bring the reactor to and maintain it at a safe shutdown condition during normal or abnormal conditions assuming a single active failure in the remaining equipment required for shutdown.

4) Protection shall be provided against potential missiles that could damage the main steam line isolation valves and the main steam lines between the isolation valves and the containment shell. Similarly, the main feedwater stop valves and the main feed veter lines between the stop valves and the containment shell are piotected from potential missiles.

5) Any accident. resulting from an external missile shall not cause off-site exposures in excess of 25% of the guidelines of 10 CFR 100.

l26 fio accidents other than loss of offsite power and those caused by the impact are assumed to occur simultaneous with the missile impact. For the case of tornado and waterspout missiles, multiple missiles (which do not impact at 26 the same location) are considered as the design basis. A single active fail-ure is also assumed.

Protection agains6 a potential missile may be provided by, but not necessar-ily be limited to, any one or combination of the following protection methods:

1. Compartmentalization Enclosure of the missile source or the equipment requiring protection in compartments whose walls prevent penetration by the missile.

Amendment 26 January 3,1979 3.5-6

2. Barriers - Erection of missile barriers either at the missile source or at the equipment to be protected.

3. Separation - Sufficient separation of redundant systems in a safety network so that a potential missilp cannot damage both redundant sys-tems and prevent safe shutdown of the plant.

4. Distance - Location of equipment beyond the range of a potential missile.

u. Equipment Design - Design equipment to withstand impact of potential missile without loss of function.

10 Method 2 (barriers) will be the primary method used for protection against external missiles.

Figures 3.5-2 through 3.5-10 show the areas of the plant protected from postu-lated missiles originating outside the plant. Wi thin this protection " envelope" are all systems and components required to achieve and maintain safe shutdown assuming a coincident loss of offsite power. The barriers shown on Figures 3.5-2 through 3.5-10 will withstand any missiles originating outside the plant.

Characteristics of missiles originating outside the plant are discussed in Sections 3.5.3, 3.5.5 and 3.5.7. Missiles generated within the plant and ex-ternal to the containment are discussed in Sections 3.5.4 and 3.5.6.

Amendment 10 July 19, 1974' 3.5-7

3.5.3 Tornado and Waterspout Missiles The full spectrum of potential tornado missiles falls into two categories, penetrating type missiles and impact missiles. The penetrating-type missiles are characterized by relatively high velocity and small area of impact and have their primary effect on the target structure by penetration. Impact-type missiles are characterized by relatively large mass and have their pri-mary effect on the structure by impulse loading.

It is assumed in a tornado strike that multiple external missiles (which do not 26 impact at the same location) can strike the plant. Analyses to determine the extent to which missiles generated by tornados and waterspouts may be a hazard to the Floating Nuclear Plant are based on the maximum tornado described in subsection 3.3.2. As indicated in subsection 3.3.2, and as estimated by water-spout researchers (Appendix 20), use of the design basis tornado is highly conservative.

Potential sources of missiles which are of hazard to the plant are:

1. loose materials on and structural components of the FNP itself;

2. small boats and bargcs within the breakwater and loose material or components which may 'oe carried on these crafts;

3. pieces of flotsam or hurricane debris which may be washed up on the breakwater.

Amendment 26 January 3, 1979 3.5-8

APPENDIX 3D LOCATION OF INFORMATION RELATING TO DIVISION 1 REGULATORY GUIDES Regulatory Guide Location of Number Title Information 1.1 Net Positive Suction Head for Emergency Core 6.2.2 and RESAR Cooling and Containment Heat Removal System Appendix 3A Pumps 1.2 Thermal Shock to Reactor Pressure Vessels RE3AR Appendix 3A 1.3 Assumptions Used for Evaluating the Potential Note 1 17 (Rev. 2) Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors 1.4 Assumptions Used for Evaluating the Potential 15.4 (Rev. 2) Radiological Consequences of a loss of Coolant Accident for Pressurized Water Reactors 1.5 Assumptions Used for Evaluating the Potential Note 1 Radiological Consequences of a Steam Line Break Accident for Boiling Water Reactors 1.6 Independence Between Redundant Standby (0nsite) 8.1.4 Power Sources and Between Their Distributi n Sys tems 1.7 Control of Combustible Gas Concentrations in 6.2.3, Containment Following a Loss of Coolant Accident 6.2.5 1.8 Personnel Selection and Training Note 2 1.9 Selection of Diesel Generator Set Capacity for 8.1.4 Standby Power Supplies 1.10 Mechanical (Cadweld) Splices in Reinforcing Bars 3.8.3 (Rev. 1 ) of Concrete Containments 1.11 Instrument Lines Penetrating Primary Reactor 6.2.4.1, Containment 7.1.3 1.12 Instrumentation for Earthquakes 3.7.4,7.1.4 (Rev. 1) 1.13 Fuel Storage Facility Design Basis 9.1 Amendment 17 September 2,1975 3D-1

LOCATION OF INFORMATION RELATING TO DIVISION 1 REGULATORY GUIDES (CONT)

Regulatory Guide Location of Number Title Information

~

1.14 Reactor Coolant Pump Flywheel Integrity RESAR Appendix 3.A 1.15 Testing of Reinforcing Bars for Concrete 3.8.3 (Rev. 1) Structures 1.16 Reporting of Operating Information Note 2 (Rev. 3) 1.17 Protection Against Industrial Sabotage 13.7 (Rev. 1 )

1.18 Structural Acceptance Test for Concrete Note 1 (Rev. 1 ) Primary Reactor Containments 1.19 Nondestructive Examination of Primary Contain- Note 1 (Rev. 1) ment Liner Welds 1.20 Vibration Measurements on Reactor Internals RESAR Appendix 3A Note 14 1.21 Measuring and Reporting of Effluents from Note 2 Nuclear Power Plants 1.22 Periodic Testing of Protection System 7.1.5, 8.1.4, Actuation Functions 8.3.1.2.1 1.23 Onsite Meteorological Programs Note 2 1.24 Assumptions Used for Evaluating the Potential 15.3.5 Radiological Consequences of a Pressurized Water Reactor Gas Storage Tank Failure 1.25 Assumptions Used for Evaluating the Potential 15.4.3 Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors 1.26 Quality Group Classifications and Standards 3.2.2 Note 14 1.27 Ultimate Heat Sink Note 2 26 (Rev. 2) 1.28 Quality Assurance Program Requirements Note 4 (Design and Construction)

Amendment 26 January 3,1979 3D-2

LOCATION OF INFORMATI0fl RELATING TO DIVISION 1 REGULATORY GUIDES (CONT)

Regulatory Guide Location of Number Title Information 1.29 Seismic Design Classification 3.2.1 (Rev. 1) 1.30 Quality Assurance Requirements for the Note 4 l 25 Installation, Inspection and Testing of Instrumentation and Electric Equipment 1.31 Control of Stainless Steel Welding Note 3 (Rev. 1) 1.3E Use of IEEE Standard 308-1971, " Criteria 8.1.4 for Class lE Electric Systems for Nuclear Power Generating Stations:

1.33 Quality Assurance Program Requirements Note 2 (0peration) 1.34 Control of Electroslag Weld Properties Note 4 1.35 Inservice Surveillance of Ungrouted Tendons Note 1 (Rev. 1) in Prestressed Concrete Containment Structures 1.36 Nonmetallic Thermal Insulation for Austenitic Note 4 Stainless Steel 1.37 Quality Assurance Requirements for Cleaning Note 4 of Fluid Systems and Associated Components

, of Water-Cooled Nuclear Power Plants 1.38 Quality Assurance Requirements for Packaging, Note 4 Shipping, Receiving, Storage, and Handling of Items for Water-Cooled Nuclear Power Plants 1.39 Housekeeping Requirements for Water-Cooled Note 4 Nuclear Power Plants 1.40 Qualification Tests of Continuous-Duty 3.11.1, Motors Installed Inside the Containment of 8.1.4 Water-Cooled Nuclear Power Plants 1.41 Preoperational Testing of Redundant On-Site 8.1.4, Electric Power Systems to Verify Proper Load 7.3 Group Assignments Amendment 25 June 20, 1978 3D-3

LOCATION OF INFORMATION RELATING TO DIVISION 1 REGULATORY GUIDES (CONT)

Regulatory Guide Location of Number Ti tle In formation 1.42 Interim Licensing Policy on As Low As 12.4 (Rev.1) Practicable for Gaseous Radiciodine Releases from Light-Water-Cooled Nuclear Power Reactors 1.43 Control of Stainless Steel Weld Cladding RESAR 3 of Low-Alloy Steel Components Appendix 3A 1.44 Control of the Use of Sensitized Stainless Note 5 Steel 1.45 Reactor Coolant Pressure Coundary Leakage 5.2.4,7.1.6 Detection System 1.46 Protection Against Pipe Whip Inside 3.6 Containment 1,47 Bypassed and Inoperable Status Indication for 7. 5.1,7.1. 7 Nuclear Power Plant Safety Systems 1.48 Design Limits and Loading Combinations for 3.9 Seismic Category I Fluid System Components 1.49 Power Levels of Water-Cooled Nuclear Power 1.1.1.2 (Rev. 1) Plants 1.50 Control of Preheat Temperature for RESAR 3 Welding of Low-Alloy Steel Appendix 3A l'.51 Inservice Inspection of ASME Code Class 2 Note 15 and 3 Huclear Power Plant Components 1.52 Design, Testing, and Maintenance Criteria 9.4(Introduction) 26 (Rev. 1) for Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants 1.53 Application of the Single-Failure Criterion 8.1.4, to Nuclear Power Plant Protection Systems 7.1.8 1.54 Quality Assurance Requirements for Protective Note 4 Coatings Applied to Water-Cooled Nuclear Power Plants Amendment 26 January 3,1979 3D-4

LOCATION OF INFORMATION RELATING TO DIVISION 1 REGULATORY GUIDES (CONT)

Regulatory Guide Location of Number Title Information 1.55 Concrete Placement in Category I Structures 3.8.3 1.56 Maintenance of Water Purity in Boiling Note 1 Water Reactors 1.57 Design Limits and Loading Combinations 3.8.2.6.2 for Metal Primary Reactor Containment System Components 1.58 Qualification of Nuclear Power Plant Inspec- Note 4 tion, Examination, and Testing Personnel 1.59 Design Basis Floods for Nuclear Power Plants Note 17 (Rev. 2) I 26 1.60 Design Response Spectra for Seismic Design 3.7 (Rev. 1) of Nuclear Power Plants 1.61 Damping Values for Seismic Design of Nuclear 3.7 Power Plants 1.62 Manual Initiation of Protective Actions 7.1.9 1.63 Electric Penetration Assemblies in Containment Note 4 26 (Rev. 2) Structures for Water-Cooled Nuclear Power Plants 1,64 Quality Assurance Program Requirements for Note 4 (Rev. 1) the Design of Nuclear Power Plants 1.65 Materials and Inspections for Reactor Vessel Note 7 Closure Studs 1.66 Nondestructive Examination of Tubular Products Note 8 1.67 Installation of Overpressure Protection Devices 5.2.1, 3.9.2.5 1.68 Preoperational and Initial Start-up Test Pro- 14 grams for Water-Cooled Power Reactors 1.68.1 Preoperational and Initial Start-up Testing Note 1 of Feedwater and Condensate Systems for Boiling Water Reactor Power Plants 26 1.68.2 Initial Startup Test Program to Denonstrate Note 2 Remote Shutdown Capability for Water-Cooled Nuclear Power Plants 1.69 Concrete Radiation Shields for Nuclear Power 3.8.1.l.3 Plants Amendment 26 January 3,1979 30-5

LOCATION OF INFORMATION RELATItiG TO DIVISI0fl 1 REGULATORY GUIDES (CONT)

Regulatory Guide Location of Number Ti tle Information 1.70 and Standard Format and Content of Safety Note 16 1.70.x Analysis Reports for Nuclear Power Plants 17 Series 1.71 Welder Qualification for Areas of Limited Note 4 Accessibili ty 1.72 Spray Pond Plastic Pipe Note l 1.73 Qualification Tests of Electric Valve Operators 8.1.4 Installed Inside the Containment of Nuclear Power Plants 1.74 Quality Assurance Terms and Definitions Note a 1.75 Physical Independence of Electric Systems 8.1.4 (Rev. 1) 17 1.76 Design Basis Tornado for Nuclear Power Plants 3.3 1.77 Assumptions Used for Evaluating a Control Rod Note 12 Ejection Accident for Pressurized Water Reactors 1.78 Assumptions for Evaluating the Habitability of 6.5.4 A Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release 1.79 Preoperational Testing of Emergency Core Cooling 6.3.4.2 Systems for Pressurized Water Reactors 1.80 Preoperational Testing of Instrument Air Systems O te 13 1.81 Shared Emergency and Shutdown Electric Systems Note 1 17 (Rev. 1) for Multi-Unit Nuclear Power P1 ants 1.82 Sumps for Emergency Core Cooling and Containment 6.2.2.7 Spray Systems 1.83 Inservice Inspection of Pressurized Water Reactor Note 9 Steam Generator Tubes 1.84 Code Case Acceptability - ASME Section III Design Note 4 17 (Rev. 2) and Fabrication Amendment 17 September 2, 1975 3D-6

LOCATION OF INFORMATION RELATING TO DIVISION 1 REGULATORY GUIDES (CONT)

Regulatory Guide '

Location of Numbe r Titl e Info rmation 1.85 Code Case Acceptability - ASME Section III Note 4 (Rev. 2) Materials 1.86 Termination of Operating Licenses for Nuclear Note 2 Reactors 1.87 Construction Criteria for Class I Components in Note 1 (Rev. 1) Elevated Temperature Reactors (Supplement to ASME Section III Code Cases 1592, 1593, 1594, 1595, and 1596) 1.88 Collection, Storage and Maintenance of Nuclear Note 4 Power Plant Quality Assurance Records 1.89 Qualification of Class IE Equipment for Nuclear 8.1.4 Power Plants 1.90 Inservice Inspection of Prestressed Concrete Note 1 Containment Structures with Grouted Tendons 1.91 Evaluation of Explosions Postulated to Occur 2.9 26 (Rev. 1) on Transportation Routes near Nuclear Power Appendix 2A Plant Sites 1.92 Combination of Modes and Spatial Components in 3.7 Seismic Response Analysis 1.93 Availability of Electric Power Sources Note 14 1.94 Quality Assurance Requirements for Installation, Note 4 Inspection, and Testing of Structural Concrete and Structural tcal During the Construction Phase of Nuclear Power Plants 1.95 Protection of Nuclear Power Plant Control Room 6.4.5 Operators Against an Accidental Chlorine Release 1.96 Design of Main Steam Isolation Valve Leakage Note 1 Control Systems for Boiling Water Reactor Nuclear Power Plants Amendment 26 January 3,1979 3D-6a

LOCATION OF INFORMATION RELATING TO DIVISION 1 REGULATORY GUIDES (CONT)

Regulatory Guide Location of flumber Ti tle In formation 1.97 Instrumentation for Light-Water-Coo' led Note 18 (Rev. 1) fluclear Power Plants to Assess Plant Conditions During and Following an Accident 1.99 Effects of Residual Elements on Predicted Note 24 (Rev. 1) Radiation Damage to Reactor Vessel Materials 1.101 Emergency Planning for Nuclear Power Plants Note 2 (Rev. 1) 1.102 Flood Protection for Nuclear Power Plants Note 17 (Rev. 1) 1.105 Instrument Setpoints Note 19 (Rev. 1) 1.108 Periodic Testing of Diesel Generator Units Note 2 (Rev. 1) Used as Onsite Electric Power Systems at Nuclear Power Plants 1.114 Guidance on Being Operator at the Controls Note 2 26 (Rev. 1) of a Nuclear Power Plant 1.115 Protection Against Low-Trajectory Turbine Note 20, (Rev. 1) Missiles 3.5.4 1.117 Tornado Design Classification 3.5.2, (Rev. 1) 3.5.3 1.121 Bases for Plugging Degraded PWR Steam Note 21 Generator Tubes 1.124 Service Limits and Loading Combinations Note 22 (Rev. 1) for Class 1 Linear Type Component Supports 1.127 Inspection of Water Control Struci.ure.s Note 2 (Rev. 1) Associated with Nuclear Power Plants 1.130 Design Limits and Loading Combinations for Note 23 Class 1 Plate and Shell Type Component Supports Amendment 26 O

January 3,1979 3D-6b

APPENDIX 3D HOTES TO DIVISION 1 REGULATORY GUIDE (CONT)

14. This guide (or the latest revision to the guide, as appropriate) will be addressed during the final design approval phase.

15. Guide withdrawn by the NRC.

16. The Plant Design Report was prepared in accordance with the original version of Regulatory Guide 1.70 and was filed shortly after publica-tion of Revision 1 to Regulatory Guide 1.70. Regulatory Guides in the 1.70.x series further modify Regulatory Guide 1.70 (Revision 1) to in-corporate additional information required in safety analysis reports.

Information required by these guides, and appropriate to the manufactur-ing license application, has been provided by the Applicant in response to Staff requests for additional information.

17. Flood studies for each site will be provided in the owner's application.

The following clarification pertains to the application of Regulatory Guide 1.59 to the FNP site envelope. The FNP is inherently protected from floods because the plant floats. Only during the postulated sinking emer-gency is the plant vulnerable to damage from flood. The non-mechanistic sinking emergency is a very unlikely event; therefore, the site envelope postulates the combined occurrance of sinking and a flood-producing event of 100 year severity (See Table 2.1-1 and Section 2.3).

18. The Applicant adopts the regulatory positions with the following clarif-ications and exceptions.

Post-Accident Monitoring (PAM) instrumentatica will be provided to mon- 26 itor key NSSS & B0P system parameters, the containment conditions, and the effectiveness of the Engineered Safety Features. The instrumentation supplied will provide the operator with information to enable him to per-form manual safety functions and to determine the effect of normal safety actions taken following a Condition II, III or IV event. PAM instru-mentation will be qualified in accordance with the final NRC-staff-approved version of the Westinghouse Topical Report, WCAP-8587.

Recorders will be provided for certain accident monitoring channels. Of those parameters selected to provide transient or trend information to the operator, at least one of the redundant accident monitoring channels is recorded. The recorder is not redundant, does not meet the single-failure criterion, does not have its own isolation amplifier (the in-com1r,g signal will already be isolated from the accident monitoring channel) and may have multiple pens to permit more than one channel to be recorded. The equipment in the environmental and seismic qualification program includes these recorders. These recorders will not be qualified to function during the postulated seismic event. Following the event, the recorders will regain an operating status.

Amendment 26 January 3,1979 3D-ll

APPENDIX 3D NOTES TO DIVISION 1 REGULATORY GUIDE (CONT)

When monitored parameters that are required for both norma ~ operation and accident monitoring are displayed on ore than one indicator, the indi-cators will be identical. This will achieve the operator familiarity intended by the position.

PAM instrumentation will be prWided as stated above; however, the instrument ranges delineated ir Position C.3 extend far beyond the worst expected values of measured pasameters following a design basis event.

The applicant therefore does not propose to supply instrumentation to satisfy these extended range requirements. The applicant believes that Position C.3 should be deleted or modified to require instrument ranges that correspond to credible worst case conditions. For example, the range for containment pressure is typically 115% of the plant's contain-ment design pressure.

19. The Applicant adopts the regulatory positions with the following clarif-ications.

Sufficient margin will be provided between the allowable technical specif-ication limit for the process variable and the nominal trip setpoint to account for drift of the setpoint as measured at the instrument rack during periodic testing. This margin is documented in the technical 26 specifications. The additional margin required to allow for uncertain-ties in the calibration and environmental effects on instrument accuracy caused by limiting postulated events (if applicable) is included in the allowance between the technical specification limit and the value assumed for the accident analysis.

Instruments will be selected based on expected normal and accident en-vironmental conditions. Therefore, setpoint locking devices will not ordinarily be supplied. The need for qualification testing will be evaluated and justified on a case basis.

Instruments are considered to be secure from unauthorized setpoint changes by personnel if the setpoint adjustment mechanism is located within an enclosed, locked cabinet with access to the cabinet under administrative control. In this case the instrument cabinet serves as a common securing device for all setpoint. adjustment mechanisms contained within.

20. The Applicant reserves comittment to the NDRC formula for concrete penetration (suggested in the Regulatory Guide) until the EPRI-sponsored missile tests have been completed and evaluated.

Amendment 26 January 3,1979 3D-lla

APPENDIX 30 NOTES TO DIVISI0ft 1 REGULATORY GUIDE (CONT)

21. The Applicant adopts the following position of Westinghouse, the steam generator vendor. This position has been approved by the NRC in the RESAR 414 Safety Evaluation Report.

Position C.1 The term " unacceptable defects" is interpreted to apply to those imper-fections resulting from service-induced mechanical or chemical degrada-tion of the tube walls which have penetrated to a depth in excess of the plugging limit.

Westinghouse has documented its opinions on Regulatory Guide 1.121 by corporate letter and has identified as the major exception the margin of 3 against tube failure for normal operation. Westinghouse defines tube failure as plastic deformation of a crack to the extent that the sides of the crack open to a non-parallel, elliptical configuration.

The tubing can sustain added internal pressure beyond those values be-fore reaching a condition of gross failure. Westinghouse interprets this to apply as an operating limit for the plant and considers that it introduces a conflict to the established conditions for plant 26 operation as identified in the plant tech specs. A factor of 3 is quite often used in ASME Code Design guidelines. These Code practices apply to the design of hardware and to the analyses done on these designs.

Conditioris which occur during operation of the equipment and which may affect the equipment so that design values no longer apply, are not di-rectly addressed by the initial Code requirements. That is one reason why plant tech specs have been generated to establish safe limits of operation for power station equipment. The ASME Code is not applicable to the operational criteria of steam generator tubing. The Westinghouse tubing design and tubing in the design condition has margins in excess of 3. In summary, Westinghouse satisfies the margin of I when it is used in a Code sense as in new equipment design. Moreover, Westinghouse does not beli eve that this margin should be utilized as a limiting con-dition for normal operation.

Position C.2.b In cases where sufficient inspection data exist to establish a degrada-tion allowance, the rate used will be an average time-rate determined from the mean of the test data.

Position C.3.d(l) and C.3.d(3)

The combined effect of these requirements would be to establish a maxi-mum permissible primary-to-secondary leak rate which may be below the threshold of detection with current methods of measurement. Westinghouse has determined the maximum acceptable length of a through-wall-crack based on secondary pipe break accident loadings which are typically twice the magnitude of normal operating pressure loads. Westinghouse will use a leak rate associated with the crack size determined on the basis of accident loadings.

Amendment 26 January 3,1979 3D-llb

APPENDIX 3D NOTES TO DIVISION 1 REGULATORY GUIDE (CONT) 0 Position C.3.e(6)

Computer code names and references will be supplied rather than the actual codes.

Position C.3.f(l)

A minimum acceptable tube wall thickness (Plugging Limit) will be estab-lished based on structural requirements and consideration of loadings, measurement accuracy and, where applicable, a degradation allowance as discussed in this position and in accordance with the general intent of this guide. Analyses to determine the maximum acceptable number of tube failures during a postulated condition are normally done to entirely different bases and criteria and are not within the scope of this guide.

Position C.3.f(4)

Where requirements for minimum wall are markedly different for different areas of the tube bundle, e.g., U-bend area versus straight length in Westinghouse designs, two plugging limits may be established to address the varying requirements in a manner which will not require unnecessary plugging of tubes. 26

22. The Applicant adopts the regulatory positions with the following clarifi-cations and exceptions.

a. The Regulatory Guide states in paragraph B.l(b): " Allowable service limits for bolted connections are derived from tensile and shear stress limits and their non-linear interaction; they also change with the size of the bolt. For this reason, the increases permitted by NF-3231.1, XVII-2110(a), and F-1370(a) of Section III are not directly applicable to allowable shear stresses and allowable stresses for bolts and bolted connections",

and in paragraph C.4: "This increase of level A or B service limits does not apply to limits for bolted connections."

As noted above, the increase in bolt allowable stress under emerg-ency and faulted conditions is not permitted. The applicant be-lieves that the present ASME Code rules are adequate for bolted connections. This position is based on the following:

It is recognized after extensive experimental work by several researchers that the interaction curve between the shear and tension stress in bolts is more closely represented by an ellipse and not a line. This has been clecrly recognized by the ASME.

Code Case 1644-6 specifies stress limits for bolts and represents this tension / shear relationship as a non-linear interaction equa-tion (incorporated into ASME III Appendix XVII via the Winter 77 Amendment 26 January 3,1979 3D-llc

APPENDIX 30 NOTES TO DIVISION 1 REGULATORY GUIDE (CONT)

Addendum) and has a built-in safety factor that ranges between 2 and 3 (depending on whether the bolt load is predominantly tension or shear) based on the actual strength of the bolt as determined by test (Ref:

" Guide to Design Criteria of Bolted and Riveted Joints," Fisher and Struik, copyright 1974, John Wiley and Sons, Page 54).

Study of three interaction curves of allowable tension and shear stress based on the ASME Code (emergency condition allowables per XVII-2110 and faulted condition allowables per F-1370) and the ulti-mate tensile and shear strength of bolts (obtained from experimental work published by E. Chesson, Jr. , N. L. Faustino, and W. H. Munse, "High Strength Bolts Subjected to Tension and Shear," Journal of the Structural Division, Proceedings of the American Society of Civil Engineers, October 1965, Pages 155-180) indicates that there is ade-quate safety margin between the emergency and faulted condition allow-ables and failure of the bolts.

During their tests to determine the strength and behavior character-istics of single high strength bolts subjected to various combina-tions of tension and shear (T-S), Cnesson, et. al. used a total of 115 bolts to ASTM specification A325-61T and A354-Grade BC. The A325-61T, which is a medium carbon steel, had a yield point of 77000 26 psi to 88000 psi and ultimate strength of 105000 psi to 120000 psi, depending upon the bolt diameter. The A354-Grade BC, which is a heat treated carbon steel, had a yield point of 99000 psi to 109000 psi and ultimate strength from 115000 psi to 125000 psi, again depending upon the bolt diameter.

Figure 30-1 shows the interaction curves for T-S loads on SA325 bolts.

Curve (1) represents the interaction relation (ellipse) permitted by Code Case 1644 (ASME III Appendix XVII Winter 77 Addenda) for service levels A, B and design condition. Curve (2) represents the interaction curve which considers the Code Case 1644 allowables and the increase permitted by XVII-2110(a) for service level C. Curve (3) represents the interaction curve which considers the Code Case 1644 allowables and the increase permitted by F-1370(a) for service level D. Curve (3) is the upper limit of the allowable ttresses.

The design stress limits represented by Curves 1, 2, and .' for A325 bolts are then compared against the ultimate strength of the bolts represented by Curve 4, which is based on Chessoi:'s test results.

The area between Curve 3 and Curve 4 is the safety margin between the maximum bolt stress under service level D and minimum ultimate strength of the bolt.

Factor of safety against failure for A325 bolts for various T-S ratios is shown in Figure 30-2. The safety factor varies between a minimum of 1.36 and a maximum of 2.29 depending upon the value Amendment 26 January 3,1979 3D-11d

APPENDIX 3D NOTES TO DIVISION 1 REGULATORY GUIDE (CONT) of T-S ratio. This is based upon the ultimate strength of the bolts from Chesson's test and the allowables obtained from Code Case 1644 and the increase permitted by F-1370(a) for service level D. Figure 3D-2 demonstrates that there exists an adequate factor of safety for the complete range of T-S loadings.

From this study it is observed that:

(1) For the emergency condition, the safety factor (ratio of ulti-mate strength to allowable stress) varies between a minimum of 1.63 and a maximum of 2.73 depending upon the actual tensile stress / shear stress (T/S) ratio on the bolt.

(2) For the faulted condition, the safety factor varies between a minimum of 1.36 to a maximum of 2.29, again depending upon actual T/S ratio on the bolt.

It is thus reasonable to allow an increase in these limits for the emergency and faulted conditions.

Based on the above discussion, for the emergency and faulted con- 26 ditions, the applicant will use allowable bolt stresses specified in Code Case 1644-6, as increased according to the provisions of XVII-2110(a) and F-1370(a), rMpectively.

b. The increased de; sign limit for the stress range identified in NF-3231.1(a) shall be limited to the smaller of 2 Sy or Su unless other-wise justified by shakedown analysis.

c. In paragraphs B.5 and C.8 of the Regulatory Guide, the applicant takes exception to the requirements that systems whose safety-related function occurs during emergency or faulted plant conditions must meet level B limits. The reduction of allowable stresses to to no greater than level B limits (which in reality are design limits since design, level A and level B limits are the same for the linear supports) for support structures in those systems with safety related functions occurring during emergency or faulted plant conditions is overly conservative. The primary concern is that the system remains capable of performing its safety function. For active components, this is accomplished through the operability program as discussed in Section 3.9.2.4. In the case of Class 1 piping, maintaining the pipe stresses within level D limits assures that piping geometry is maintained and that required flow is not impeded. The selection of more restrictive stress limits for component supports is not necessary to assure the functional capability of the system.

Amendment 26 January 3,1979 3D-lle

APPENDIX 30 NOTES TO DIVISION 1 REGULATORY GUIDE (CONT)

d. Paragraph C.4 of the Regulatory Guide states: "However, all increases (i.e. , those allowed by NF-3231.l(a), XVII-2110(a),

and F-1370(a)) should always be limited by XVII-2110(b) of Section III". Paragraph XVII-2110(b) specifies that member com-pressive axial loads shall be limited to 2/3 of critical buckling.

In the design of component supports, member compressive axial loads shall be limited to 0.67 times the critical buckling strength.

If, as a result of more detailed evaluation of the supports the member compressive axial loads can be shown to safely exceed 0.67 times the critical buckling strength for the faulted condition, verification of the support functional adequacy will be documented and submitted to the NRC for review. The member compressive axial loads will not exceed 0.67 times the critical buckling strength without NRC acceptance. In no case shall the compressive load exceed 0.9 times the critical buckling strength.

e. Paragraph C.4 of the Regulatory Guide states that increases in Level A or B service limits does not apply to limits for bolted connections. The applicant's design of component supports re-stricts the use of bolting material to the following applications: 26 (1) Bolting is predominantly in tension. Oversized holes are generally provided and a mechanism other than the bolts is provided to take any shear loads. Shear or shear & tension interaction occur only in isolated locations; (2) Bolts are limited to the following material A490, SA-354, SA-325, SA-540.

(3) The diameters used range between 1/2" and 3".

These limitations on bolt usage are standard in the applicant's supports. We will limit tensile loads in the bolts to 0.7 Su, but not to exceed in any case 0.9 Sy. The allowables are taken at temperature. In those few cases where bolts are used in shear or tension and shear, ASME Code Appendix XVII - 2460 Requirements will apply with an increase factor that is defined in Regulatory Guide 1.124 or in Appendix F-1370, whichever is more restrictive. This provides an adequate margin of safety for the applicant's design.

If future revisions to the bolting criteria in ASME Section III modify the applicant's criteria listed above, we will review the criteria at the time.

f. Paragraph C.6(a) of the Regulatory guide appears confusing as to what stress limits may be increased for the emergency condition. The applicant will interpret this paragraph as follows: "The stress Amendment 26 January 3,1979 3D-ll f

APPENDIX 30 NOTES TO DIVISION 1 REGULATORY GUIDE (CONT) limits-of XVII-2000 of Section III and Regulatory Position 3 increased accordirig to the provisions of XVII-2110(a) of Section III and Regula-tory Position 4, should not be exceeded for component supports de-signed by the linear elastic analysis method."

g. The method described in Paragraph C.7(b) of the Regulatory Guide is overly conservative and inconsistent with the stress limits presented in Appendix F. The applicant will use the provisions of F-1370(d) to determine service level D allowable loads for supports designed by the load rating method. If future revisions to Appendix F modify this criteria, it will be reviewed further. If the load rating method is used, further details of its implementation will be pro-vided at that time.

23. The Applicant adopts the regulatory positions with the following clarifica-tions and exceptions,

a. The applicant will use the latest revision of Code Case 1644 as approved by Regulatory Guide 1.85.

b. Paragraph B.1 states that increases are not allowed for bolted con-nections for emergency and faulted conditions. The applicant's 26 position is that it is reasonable to allow an increase in the limits for bolted connections for these conditions. Further justification concerning this position can be found in Item a of the discussion on Regulatory Guide 1.124.

c. Paragraphs C.3, C.4(a), and C.6(a) state that the allowable buckling strength should be calculated using a design margin of 2 for flat plates and 3 for shells for normal, upset, and emergency conditions.

In the design of plate-type supports, member compressive axial loads shall be limited per the requirements of Paragrapn C.3 for normal upset, and emergency conditions.

d. In paragraph C.7, the inclusion of the upset plant condition in this load combination is inappropriate. The upset plant conditions are properly considered in paragraph C.4.

e. Paragraph C.7(a) references the criterion presented in F-1370(c),

which states: " ... loads should not exceed 0.67 times the critical buckling strength of the support...".

In the design of plate-type component supports, member compressive axial loads shall be limited to 0.67 times the critical buckling strength. If, as a result of more detailed evaluation of the supports the member compressive axial loads can be shown to safely exceed 0.67 Amendment 26 January 3,1979 3D-11g

APPENDIT 3D NOTES TO DIVISION 1 REGULATORY GUIDE (CONT) times the critical buckling strength for the faulted condition, veri-fication of the support functional adequacy will be documented and submitted to the NRC for review. The member compressive axial loads will not exceed 0.67 times the critical buckling strength without NRC acceptance. The applicant has no Class 1 shell-type supports.

f. The method described in paragraph C.7(b) of the Regulatory Guide is overly conservative and inconsistent with the stress limits presented in Appendix F. The applicant will use the provisions of F-1370(d) to determine service level D allowable loads for supports designed by the load rating method.

24. The basis as well as the scope of the Guide for predicting adjustment of reference temperature as given in Regulatory Position C.1 are not appropri-ate since the data base used was incomplete and included some data which were not applicable.

The Applicant is in agreement with the Position C.2a. However, with re-spect to Position C.2b, the Applicant believes that Figure 2 is incorrect since the upper shelf energy for six-inch thick ASTM A302B reference correla-tion monitor material reported by Hawthorne indicates essentially a constant upper shelf at fluences above s 1 x 1019 n/cm2.1/ 26 With reference to the guide Position C.3, the stresses in the vessel can be 'imited during operation in order to comply with the requirements of Appendix G to 10CFR Part 50 even though the end-of-life adjusted reference temperature may exceed 200 F. By applying the procedures of Appendix G to ASME Section III, the stress limits including appropriate Code safety margin can be met, and it is not necessary to control residual elements to levels that result in predicted adjusted reference temperature of less than 200 F at end-of-life.

As an alternative to Regulatory Guide 1.99, operating limits for the Floating Nuclear Plant will be determined by using the current radiation damage curves given in Figure 3D-3 (reproduced from RESAR-3S). Recent surveillance capsule data indicate that after an initial increase RTNDT e with increased fluence and that the Westinghouse curves NDT are conservative._2 /

does not chang REFERENCES 1/ awthorne, H J. R., " Radiation Effects Information Generated on the ASTM Reference Ccrrelation - Monitor Steels," (Unpublished study).

2_/ Letter NSTMA 1843 of 6/23/78 from T. M. Anderson to Secretary of N.R.C.

Amendment 26 January 3,1979 3D-llh

APPENDIX 3D LOCATION OF INFORfLTION RELATING TO DIVISION 8 REGULATORY GUIDES Regulatory Guide Location of Number Title Information General Note: Except for the guides listed below none of the Division 8 Regulatory Guides are applicable to the Application for License to Manufacture Floating Nuclear Plants.

8.1 Radiation Symbol Note 1 8.5 Immediate Evacuation Signal 9.1.4.3 (Note 2) 8.8 Information Relevant to Maintaining Occupational 11.6 and (Rev. 2) Radiation Exposure as Low as Practicable 26 12.1.6.5 (Nuclear Reactors) Note 3 Last guide considered: 8.13.

NOTES

1. Radiation symbols furnished by the Applicant will conform to the requirement of this guide.

2. The Applicant considers that this guide applies only to the containment evacuation alarm which is operable during refueling and sounds at a preset source range neutron count rate.

3. Regulatory positions 1, 2, and 4 are not applicable to the Application for License to Manufacture Floating Nuclear Plants.

Amendment 26 January 3,1979 3D-18

APPEllDIX 3D LOCATION OF INFORMATI0f1 RELATING TO DIVISION 9 REGULATORY GUIDES Regulatory Guides Location of Number Title Information General Note: The comission has, with the approval of the Attorney General, excepted the License to Manufacture Floating Nuclear Plants requested by the Applicant from the re-quirements of Section 105c of the Atomic Energy Act.

Reference 39 FR 10470, March 20,1974. Division 9 guides 17 are, theiefere, not applicable to the Application for License to Manufacture Floating Nuclear Plants.

O Amendment 17 September 2, 1975 3D-19

160 (1) ALLOWABLE STRESS UNDER SERVICE LEVEL A, SERVICE LEVEL B, AND DESIGN CONDITION PER ASME CODE CASE 1644.

140 -

(2) ALLOWABLE STRESS UNDER SERVICE LEVEL C PER ASME CODE CASE 1644 Atl0 APPENDIX XVII-2110.

(3) ALLOWABLE STRESS UNDER SERVICE LEVEL D PER ASME CODE CASE 1644 AND APPENDIX P-1370.

120 -

(4) ULTIMATE STRENGTH PER TEST RESULTS (l" DIA.

/// SA325 BOLTS).

100 -

ULTIMATE STRENGTH SAFETY MARGIN 80 MAXIMUM O 3 ALLOWABLE f FOR SERVICE

$ 60 -

LEVEL D E 2 m

b G 40 -

5 t

20 -

/

/

' /

/

I

'e l I h I 0

0 20 40 60 80 100 120 SHEAR STRESS (ksi)

C0fiPARIS0N OF TENSILE STRESS FOR BOLTS Amendnent 26 January 3,1979

T-S STRESS RATIO 1.0:0.2 1.0:0.67 0.67:1.0 0.2:1.0 4.0 2

l l l l d

C a

Cd 5 3.0 g

u$

m

$5 t3 d 2.0 -

50; 55 55 r<

l.0 _

0.0 l 1.0:0.0 1.0:0.42 1.0:1.0 0.42:1.0 0.0:1.0 T-S STRESS RATIO FACTOR OF SAFETY AGAINST FAILURE UNDER Amendmnt 26 SERVICE LEVEL D AS A FUNCTION OF T-S January 3,1979 RATIO FIGURE 30-2

8555-95 1000 , , , _ , w i. , i +_ , ....g,, , i . t. i., , .

= =~ r- -

EFFECT OF FAST NEUTRON FLUENCE AND COPPER

'*~ ~ ^- ^ '

700 CONTENT ON THE SHIFT OF RT NDT FOR REACTOR 9

c  ;  :. [, - _;

4ESSEL STEELS IRRADIATED AT 550 F ,, , , .. ..,i. ,

, .: , m  ! it,'8 1

s. a .:

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E  : ,4 of)ASE-o ', ' e o.ss jppVV3hy 0.10 go VEh --

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Z 100 g , . . p.,t .y , fg

$ ]- I ~ I i 0 l l .: 2#7 6 -

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10 18 io l9 10 20 FLUENCE ("/cm2 > I med EFFECT OF FLUENCE AND COPPER CONTENT ON SHIFT FOR REACTOR OFRT[DITEELSEXPOSEDT VESSE 550 F TEMPERATURE Amendment 26 January 3,1979 FIGURE 30-3

TABLE OF' CONTENTS (CONT)

Section Title Page 5.5.2.1 Design Basis (Steam Generator) 5.5-2 5.5.2.2 Design Description 5.5-3 5.5.3 Reactor Coolant Piping RESAR 5.5.4 Main Steam Line Flow Restrictor 5.5-4 5.5.5 Main Steam Line Isolation S ystem 5.5-5 5.5.6 Reactor Core Isolation Cooling System RESAR 5.5.7 Residual Heat Removal System RESAR 5.5.8 Reactor Coolant Cleanup System RESAR 5.5.9 Main Steam Line and Feedwater Piping 5.5-5 5.5.10 Pressurizer RESAR 5.5.11 Pressurizer Relief Tank RESAR 5.5.12 Valves RESAR 5.5.13 Safety and Relief Valves RESAR 5.5.14 Components Supports 5.5-4 5.5.14.1 Design Basis 5.5-5 5.5.14.2 Description 5.5-Sa 5.5.14.3 Safety Evaluation 5.5-Sa 5.5.14.4 Testing 5.5-6 5.6 Instrument Application 5.6-1 Amendment 12 5-iii November 5, 1974

LIST OF TABLES Tables for this chapter are presented in RESAR 3, Chapter 5, except as O

follows:

Table flo. Ti tle Page 5.2-2 Sumary of Reactor Coolant System 5.2-la(a)

Design Transients 5.2-4 Loading Combinations 5.2-ld 5.2-5 Allowable Stresses for ASME III, Class 1 5.2-le Components 5.2-6 (Deleted)

I 5.2-7 ASME Code Cases Applicable to Class 1 Equipment 5.2-lf 26 5.2-8 Allowable Faulted Condition Primary Stresses - 5.2-li Class 1 Components LIST OF FIGURES Figures for this chapter are presented in RESAR 3, Chapter 5, except as fullows:

Title Figure flo.

Reactor Coolant System Flow Diagram 5.1-1 (3 sheets)

Reactor Vessel Inspection Tool Details (Vessel Scanner) 5.2-1 Reactor Vessel Inspection Tool Details (Nozzle and 5.2-2 Flange Scanner)

Containment Air Particulate Activity Increase as a 5.2-3 Function of Time and for Various Leaks Amendment 26 January 3,1979 5-iv

5.2 INTEGRITY OF THE REACTOR COOLANT SYSTEM PRESSURE BOUNDARY Section 5.2.2 and 5.2.3 are presented in RESAR, information for the four loop plant without stop valves is applicable. The following information supplements section 5.2.2:

Each of the three pressurizer safety valves is set at 2485 psig and reaches full relief flow of 420,000 lbs/hr at 3;; accumulation. These valves to-gether will accomodate a maximum 50!; load change without reactor trip.

5.2.1 Design Criteria, Methods and Procedures Section 5.2.1 of RESAR is modified herein to cover the wind and wave effects for stress analyses including fatigue evaluation. The basic modification is in the areas of design condition, loading and stress combinations. The Floating nuclear Pcwer plants are subject to platform motion caused by wind and wave in addi tion to earthquake as for land plants. The combined operating basis environmental transients (C) and design basis environmental transients (Cs) are defined in 3.7.1.1 and the magnitude and occurence of each event are sumarized in Tables 3.7.1-2 and 3.7.1-4. For the Reactor Coolant System, towing will be treated conservatively as an upset condition. These environ-mental transients unique to the Floating Nuclear Plants will be included within the analyses and by the computer programs which were applied to the basic RESAR design.

Amendment 12 November 5,1974 5.2-1

Methods of stresa analysis for the faulted condition are discussed in RESAR 5.2.1.3 and Appendix F of the ASME B and PV code section III. The elastic O system analysis and elastic components analysis method will be generally used for the Floating Nuclear Plant. Other methods if used will be described and documented in the permanent plant file. Stress limits for the faulted condition are presented in Table 5.2-8 which is consistent with Appendix F.

In addition, allowable stresses for class 1 components consistent with the ASME Code section III are presented in Table 5.2-5. Class I supports and associated fasteners will be designed to ASME Section III Subsection NF and 26 Appendix F to the ASME Code, except that the provisions of USNRC Regulatory Guides 1.124 and 1.130, adopted as modified in Appendix 3D to the PDR, shall apply.

Tables 5.2-2, 5.2-4, 5.2-5, and 5.2-8 of RESAR are suitably modified for the 26 Floating Nuclear Plant and are included in this report. Table 5.2-4 includes explicitly the plant transients, the relief valve and fast valve operations, and the environmental transients in the loading combinations.

Amendment 26 January 3,1979 5.2-la

TABLE 5.2-5 ALLOWABLE STRESSES FOR ASME SECTION III CLASS 1 COMPONENTS II}

Operating Condition Classification Vessels / Tanks Piping Pump Valves Design ASME Section III ASME Section III ASME Section III ASME Section III Normal ASME Section III ASME Section III ASME Section III ASME Section III Upset ASME Section III ASME Section III ASME Section III ASME Section III

{ Emergency Not Applicable 9 k'

Faulted See Table 5.2-8 See Table 5.2-8 See Table 5.2-8 a) Calculate Pm from para.

(No active class 1 NB3545.1 with Internal pump used)

Pressure Ps = 1.50P, P < 2.4 S:,or0.7Sy Ped, P,, Ps

• Ot2, Cp , Su, S n E' Sm as defined by Section III ASME Code b) Calculate Sn from para.

NB3545.2 with g" g (1) A test of the components may be performed in lieu of analysis. C = 3.0 g( P p

s

= 1.50Ps

~,_. 5 Qt2= 0

[ Pgl.3XvalueofP ed from equations of NB3545.2 (b)(1)

Sn- # 3S m

TABLE 5.2-7 ASME CODE CASES APPLICABLE TO O

CLASS 1 EQUIPMENT Code Cases 1141 Foreign Produced Steel 1332 Requirements for Steel Forgings 1334 Requirements for Corrosion Resistant Steel Bars 1335 Requirements for Bolting ftaterial 1337 Requirements for Special Type 403 Modified Forgings or Bars (Section III) 1344 Requirements for Nickel-Chromium Age-Hardenable Alloys 1345 Requirements for Nickel-Molybdenum-Chromium-Iron Alloys 1355 Electroslag Welding l 10 1364 Ultrasonic Transducers SA-435 (Section. III) 1384 Requirements for Precipitation Hardening Alloy Bars & Forgings 1388 Requirements for Stainless Steel-Precipitation Hardening 1390 Requirements for Nickel-Chromium Age-Hardenable Alloys for Bolting 1395 SA-508, Class 2 Forgings-Modified ltanganese Content 1401 Welding Repair; to Cladding 1407 Time of Examination l10 1423-2 Plate; Wrought Type 304 & 316 with Nitrogen Added 1433 Forgings; SA-387 l10 1434 Class SN Steel Casting (Postwelt ieat Treatment for 5A-407) 1448 Use of Case Interpretations of ANSI B31 Code for Pressure Piping 1456 Substitution of Ultrasonic Examination 1461 Electron Beam Welding 110 1470 External Pressure Charts for Low Alloy Steel 1471 Vacuum Electron Beam Welding of Tube Sheet Joints 1474 Integrally Finned Tubes (Section III) 1477 B-31.7, ANSI 1970 Addenda 1484 SB-163 Nickel-Chromium-Iron Tubing at a Specified Minimum Yield Strength of 40,000 psi O

5.2-lf Amendment 26 January 3,1979

TABLE 5.2-7 (Cont'd)

Code Cases 1487 Evaluation of Nuclear Piping for Faulted Conditions 1492 Postweld Heat Treatment 1493 Postweld Heat Treatment 1494 Weld Procedure Q1alification Test l10 1498 SA-508, Class 2 Minimum Tempering Temperature l10 1508 Allowable Stresses, Design Stress Intensity and/or Yield Strength Values 11 0 1515 Ultrasonic Examination of Ring Forgings for Shell Section of Section III--Class 1 Vessels 1516 Welding of Non-Integral Seats in Valves for Section III Application 11 0 1519 Use of A-105-71 in licu of SA-105 1521 Use of H. Grades SA-240, SA-479, SA-336 and SA-358 1522 ASTM Material Specifications 1523 Plate Steel Refined by Electroslag Remelting 1524 Piping 2" NPS and Smaller 1525 Pipe Descaled by Other Than Pickling 1526 Elimination of Surface Defects 1527 Integrally Finned Tubes 1528 High Strength SA508 Class 2 and SA-541 Class 2 Forgings for Section III Construction of Class 1 Components 1529 Material for Instrument Line Fittings 1531 Electrical Penetrations, Special Alloys for Electrical Penetrations Seals 15 Overpressurization of Valves 1535 Hydrostatic Test of Class 1, Nuclear Valves 1539 Metal Bellows and Metal Diaphragm Steam Sealed Valves, Class 1, 2 & 3 1542 Requirements for Type 403 Modified Forgings or Bars for Bolting Material 1544 Radiographic Acceptance Standards for Repair Welds 1545 Test Specimens from Separate Forgings for Class 1, 2, 3 & MC 1546 Fracture Toughness Test for Weld Metal Section l10 1552 Design by Analysis of Section III Class 1 Valves 0

1567 Test Lots for Low Alloy Steel Electrodes 1568 Test Lots for Low Alloy Steel Electrodes frendment 26 5.2-19

TABLE 5.2-7 (Cont'd)

Code Cases Title 1571 Materials for Instrument Line Fittings; For SA-234 Carbon Steel Fittings 1571 Vacuum Relief Valves 1574 Hydrostatic Test Pressure for Safety Relief Valves 1616 Clarification on Ultrasonic Testing of Seamless Austenitic Steel Pipe,Section III, Class 1 Construction 1361 Socket Welds O

Amendment 26 O

5.2-lh January 3,1979

TABLE 5.2-8 ALLOWABLE FAULTED CONDITION PRIMARY STRESSES - CLASS 1 COMPONENTS (5)

System (or Subsystem) Compv.enus Stress Limits for Component Supports (1)

Analysis Analysis Components Test P P +P m m b Smaller of Smaller of (4)

Elastic 2.4 S ,& 0.70 S u 3.6 Sm & l.05 S u ELASTIC ui Plastic Larger of Larger of

~

0.70 S or 0.70 S ut r 26 7 u

[ Sy +1/3 (Su-bY ) Sy +1/3 (S ut-by )

Limit Analysis 0.9 L j 0.8 L (2)

Plastic Larger or 0.70 S u Larger of 0.70 S ut PLASTIC or or Elastic Sy +1/3 (Su-by ) by +1/3 (S ut - y}

=a a !2

.A G

W

TABLE 5.2-8 (Continued)

ALLOWABLE PRIMARY STRESSES (7) ,

Component Analysis High Strength Materials Su > 100,000 psi Bolts, Structural Fasteners (8) and Other Structural Components (1) 26 P

m P

m

+P b Test Elastic .70 S u 1.05 S u Plastic .70 S u

Larger of 0.70 S ut or Sy + 1/3 (S ut-Sy) 0.8 L T Limit L 3

(9)

Table 5.2-8 Notes 26

1. Class 1 supports and associated fasteners will be designed to ASME Section III, Subsection NF and Appendix F to the ASME Code, except that the provisions of Regulatory Guides 1.124 and 1.130 shall apply as stated in Appendix 3D of the PDR.

2. L3 = Lower bound limit load with an assumed yield point equal to 0.70 y.

3. L3 and L2 = Lower bound limit load with an assumed yield point equal to 2.3 Smand Sj (but not to exceed 0.70 Su ), respectively.

4. These limits are based on a bending shape factor of 1.5. For simple bend-ing cases with different shape factors, the limits will be changed propor-tionally.

5. The use of these stress limits on Westinghouse-supplied components will be shown to be no less conservative than the limits of Appendix F to the ASME Code.

Amendment 26 January 3,1979

Table 5.2-8 Notes (Con'd)

If plastic component analysis is used with elastic system analysis or with plastic system analysis, the deformations and displacements of the indi-vidual system members will be shown to be no larger than those which can be properly calculated by the analytical methods used for the system analysis.

6. No shape factors larger than 1.5 will be used for limit analysis.

. To be used for both elastic and plastic system (or subsystem) analysis.

8. Applies to Core Support Structure nomenclature.

9. LT= Test Load defined in RESAR Section 5.2.1.3.3.

Yield stress at temperature S,=

Su = Ultimate stress from engineering stress-strain curve at temperature S = Ultimate stress from true stress-strain curve at temperature ut Sm = Stress intensity from ASME Section III at temperature Amendment 26 January 3,1979 5.2-lK

O THIS PAGE INTENTIONALLY BLANK O

Amendment 26 January 3,1979 5.2-1L

8.1.4 pESIGilCRITERIA The design bases, criteria, safety guides, standards, and other documents that are implemented in the design of this unit are:

1. Industry Manufacturing Standards.

a. ANSI C37 Switchgear

b. ANSI C50 Rotating Electrical Machinery

c. ANSI C57 Transformers, Regulators, and Reactors

d. IPCEA P-46-426 Power Cable Ampacities

e. NEMA SG3-1965 Low Voltage Power Circuit Breakers

f. NEfM S64-1968 AC 'igh Voltage Circuit Breakers 9 NEMA SG5-1967 Power Switching Assemblies

h. NEMA SG5-1966 Power Switching Equipment

i. NEMA R12-1971 Battery Chargers - General Purpose & Communications J. NEtM IB-1-1971 Defintiions for Lead Acid Storage Ratteries 8.1-5

k. fiEMA TRI-1968 Transformers, Regulators and Reactors

1. flEMA TR-P3-1970 Guide for Preparation of Specifications for Large Power Transformers

m. flEMA MGI-1967 Motors and Generators

2. The power supply for the reactor protection system and the safety systems will be in accord with Criteria 17 and 18 of 10 CFR 50, Appendix A.

3. Additional design criteria are as follows:

a. IEEE 279-1971 - Criteria for fluclear Power Plant Protection Systems.

b. IEEE 288-1969 - Guide for Induction Motor Protection.

c. IEEE 308-1971 - Criteria for Class lE Electrical Systems for fluclear Power Generating Stations.

d. IEEE 317-1976 - Electrical Penetration Assembly in Containment 26 Structures for 11uclear Fueled Power Generating Stations.

Amendment 26 January 3,1979 0

8.1-6

TABLE OF CONTENTS (CONT)

Section Ti tle Page 9.4 Air Conditioning, Heating, Cooling and 9.4-1 Ventilation Systems 9.4.1 Control Building Ventilation 9.4-2c 26 9.4.1.1 Design Bases 9.4-2c 9.4.1.2 System Description 9.4-4

9. 4.1. 3 Design Evaluation 9.4-10b 9.4.1.4 Tests and Inspections 9.4-12a

9. 4.1. 4.1 Manufacturer Shop Testing 9.4-13 9.4.1.4.2 System Testing and Inspection 9.4-13 9.4.1.5 Instrumentation and Application 9.4-14 9.4.2 Auxiliary Area & Fuel Handling Building 9.4-15 9.4.2.1 , Design Bases 9.4-15 9.4.2.2 System Description 9.4-16 9.4.2.3 Design Evaluation 9.4-26 9.4.2.5 Instrumentation and Application 9.4-30 9.4.3 Personnel Building Ventilation 9.4-38 9.4.3.1 Design Bases 9.4-38 9.4.3.2 System Description 9.4-38 9.4.3.3 Design and Safety Evaluation 9.4-40 9.4.3.4 Tests and Inspections 9.4-42 9.03.4.1 Manufacturers Shop Testing 9.4-42 9.4.3.4.2 System Testing and Inspection 9.4-43 Amendment 26 January 3,1979 9-vii

TABLE OF CONTENTS (CONT)

Section Title Page 9.4.3.5 Instrumentation and Application 9.4-43 9.4.4 Turbine Building Ventilation Systems 9.4-48 9.4.4.1 Design Bases 9.4-48 9.4.4.2 System Description 9.4-49 9.4.4.3 Design Evaluation 9.4-50 9.4.4.4 Tests and Inspections 9.4-51a 9.4.4.4.1 Manuf acturers Snop Tasting 9.4-51a 9.4.4.4.2 System Testing and Inspection 9.4-52 9.4.4.5 Ins trumentation and Application 9.4-52 9.4.5 Containment Ventilation 9.4-55 9.4.5.1 Design Bases 9.4-55 9.4.5.2 System Description 9.4-57 9.4.5.3 Design Evaluation 9.4-61 9.4.5.4 Tests and Inspections 9.4-63 9.4.5.4.1 Manufacturers Shop Testing 9.4-63 l10 9.4.5.4.2 System Testing and Inspection 9.4-64 9.4.5.5 Instrumentation and Application 9.4-65 9.4.6 Plant Vent Systen 9.4-67 9.4.6.1 Design Basis 9.4-67 9.4.6.2 Sys tem Design and Operation 9.4-68 9.4.6.3 Design Evaluation 9.4-68 9.4.6.4 Tests and Inspections 9.4-69 Amendment 10 O

9-viii July 19, 1974

LIST OF TABLES (CONT)

Table fio. Title Page 9.2.4-1 flon Essential Service Water System Components 9.2-67 9.2.4-2 fion-Essential Cooling Water Component Design 9.2-68 thru 9.2-72 9.2.4-3 Non-Essential Service Water System Water Chemistry 9.2-73 9.3.2-1 Containment Post Accident Sampling System 9.3-20 9.3.4-1 Parameters of the Chemical Volume Control Components 9.3-32 flot Included in RESAR 3 9.3.5-1 Principal Component Data Sumary 9.3-49 9.3.6-1 Gas Supply System Design Requirements 9.3-57 9.3.7-1 Bulk Chemical Supply System Equipment Design 9.3-59 thru Parameters 9.3-60a 9.3.7-2 Chemicals Used During Normal Plant Operation 9.3-60b 9.4-1 Estimated Filter Penetration by Salt Particles 9.4-lf 9.4.1-1 Control Building Ventilation System Performance 9.4-5 Data 9.4.1-2 Control Building Air Conditioning System 9.4-6 9.4.1-3 Single Active Failure Analysis for Control Room 9.4-lla Air Conditioning System Components 9.4.1-4 Control Room Air Leakage Paths 9.4-14a 9.4-1-5 Control Room flormal and Emergency Air Flow Rates 9.14-14b and 9.14-14c 9.4-2 (Deleted) 26 9.4.2-1 Auxiliary Area Ventilation Systems 9.4-31 9.4.2-2 Auxiliary Building Ventilation Systems 9.4-36,37 9.4.2-3 Fuel Building Air Leakage Pathways 9.4-37a 9.4.3-1 Personnel Building Ventilation Systems Design Con- 9.4-45 ditions and Performance Data 9.4.3-2 Personnel Building 9.4-46 9.4.3-3 (Deleted)

Amendment 26 January 3, 1979 9-xiii

LIST OF TABLES (CONT.)

Table No. Title Page 9.4.4-1 Turbine Area Ventilation System Performance 9.4-53 Data 9.4.4-2 Turbine Building Ventilation 9.4-54 9.4.5-1 Containment Ventilation System Equipment 9.4-58

' Design Parameters 9.4.5-2 Containment Ventilation System 9.4-59 9.4.6-1 Plant Vent System Components 9.4-71 9.4.6-2 Plant Vent Systems Materials 9.4-72

~

9.5.1-1 Internal Fire Protection System Components 9.5-6 Design Data 9.5.1-2 External Fire Protection System Component 9.5-12a 12 Design Data Selected Wave Height and Duration Data 9' ~

9.5.4-1 (New Jersey)

• ~

• 9.5.4-2 Selected Wave Height and Duration Data (North Carolina) 9.5.4-3 Selected Wave Height and Duration Data (Florida)

' 9 9.5.4 Selected Wave Height and Duration Data (Texas)

~

9.5.4-5 Wind Persistence Data for Cape Hatteras, North Carolina Wind Persistence Data for Corpus Christi ,

9.5.4-6 Texas

~

9.6.1-1 Platform Drain System Components

.6-11 9.6.2-1 Platform Trim System Components

~

9.6.3-1 Specific Installations of Impressed Current Cathodic Protection (ICCP) Systems O

Amendment 12 0-xiv November 3,1974

Smoke Detection Ventilation systems are provided with smoke detectors as shown on the flow diagrams. Audible and visual alarm is given in the control room. Systems are either shutdown when smoke originates outside the space or placed on a venting mode when smoke originates inside the space except the control room and emergency relocation areas, which are placed on limited minimum outside air intake mode utilizing HEPA and charcoal filters in all cases, including isolation signal or smoke signal. See Smoke and Survival Protection in Sec-tion 9.4.1.2 and Table 9.4.1-5 for limited minimum outside air intake and system air flows.

Regulatory Guide and Branch Technical Position Compliance flon-ESF HVAC Systems are designed to be in compliance with BTP ETSB 11-2,

" Design Testing, and Maintenance Criteria for Normal Ventilation Exhaust System Air Filtration and Adsorbtion Units of Light Water Cooled Nuclear Power Reactor Plants."

The ESF Air Filtration Systems of tne Floating Nuclear Plant which come within 26 the scope of Regulatory Guide 1.52 are:

1. Control room

2. Emergency relocation area (ERA)

3. Fuel pit exhaust

4. Annulus air filtration

5. Containment purge exhaust 26 Amendment 26 January 3,1979 9.4-2a

These systems are designed to be in compliance with Regulatory Guide 1.52 with the following exceptions.

Paraaraoh 2a All systems are designed redundant with the exception of ducts. This has been accepted industry practice in the past. The ducts on the ESF Systems are designed and/or protected from all credible internal and external events.

Consequently, since ducts are passive elements, the systems are not designed to account for duct failure.

Paragraph 2b Redundant atmosphere cleanup systems are protected from externally generated missiles by placing the systems within protected areas of the plant. On 26 exception is the Fuel Building Ventilation System downstream of the filtration trains. This portion of the Fuel Building Ventilation System is not missile protected because the Design Basis Fuel Handling Accident concurrent with a tornado is not credible. Secondly, the Annulus Air Filtration System downstream of the exhaust fans is not missile protected because Post-LOCA Recovery concurrent with a tornado is not credible.

Furthermore, redundant atmosphere cleanup systems are located and/or designed to preclude local phenomena from impairing their design capabilities.

Paragraph 2f It is the Applicant's understanding that the 30,000 CFM approximate limitation exists for two reasons. One is to limit the air flow rate for testing purposes and Amendment 26 January 3,1979 9.4-2b

another is, for ease of maintenance, to limit the number of filters to 10 wide by 3 high ultilizing 1000 CFM HEPA filters.

The volumetric flow rate of a single cleanup train in the FNP design is limited to 36,000 CFM. This exception is justified for the following reasons:

From a testing point of view the 30,000 is predicated on the practical limitations of Laskin nozzle air operated D0P generators, however Gas-Thermal DOP generators are available which produce adequate D0P concentration for air flows up to 50,000 CFM. From a maintenance point of view, by utilizing 1500 CFM HEPA filters, the plant design reduces the number of filter cells from the 30 mentioned in Regula-tory Guide 1.52 to 24. When filter height is greater than three filter cells an internal maintenance platform with separate external access is provided.

Paragraph 29 The pertinent pressure drops and flow rates are instrumented to signal and alarm in the Control Room. However, recording of these pressure drops and flow rates does not provide any additional information which would aid the safe operation of the plant. Therefore, recording of these signals is not considered a design requi rement.

9.4.1 CONTROL BUILDING VENTILATION 9.4.1.1 Design Bases Control building ventilation is designed to maintain the control room, process rack room, rod control room, and emergency relocation area environments Amendment 26 January 3,1979 9.4-2c

TABLE OF C0flTEflTS (CONT)

APPEllDIX B Section Ti tle Page B.18 Request for Additional Information, October 12, 1978 26 8.18.1 Foreword and Enclosures 2 and 3 to USNRC Letter B.18.1-1 (R.S. Boyd to A.P. Zechella) dated October 12, 1978 through B.18.1-8 8.18.2 Applicant's Responses B.18.2-1 through B.18.2-5 Amendment 26 January 3,1979 B-v

B.18.1 FOREWORD AfiD Ef1 CLOSURES 2 Af1D 3 TO USflRC LETTER (R.S. BOYD TO A.P. ZECHELLA) DATED OCTOBER 12, 1978 FOREWORD In a letter from Mr. R.S. Boyd dated Ocotber 12, 1978, Applicant was asked to respond to requests for additional information summarized in five numbered attachments. Applicant responded by letter dated October 31, 1978 (P.B. Haga 26 to R.S. Boyd, FilP-tflE-865). In this lette Applicant proposed to address those issues contained in Enclosures 2 and 3 to Mr. Boyd's letter which arose prior to January 1,1978. Applicant further proposed to address other issues during final design approval to the extent applicable at that time. In a letter dated December 22,1978 (H.R. Denton to A P. Zechella) the NRC accepted the substance of the Applicant's proposals outlined above. Therefore, only En-closures 2 and 3 to Mr. Boyd's letter of October 12, 1978 are reprinted in Amendment 26.

Amendment 26 January 3,1979 B.18.1-1

ENCLOSURE 2~

CATEGORY II MATTEisAPPROVED SY RRRC 9

DOCUMCrrr NUMBER REVISION TITLE RG 1.27 2 Ultimate Heat Sink f or Nuclear Power Plants RG 1.52 1 Design, Testing, and "aintenance Criteria for Engineered-Safety-?cature Atmosphere Cleanup Systen Air Filtration and Adsorption Units of Light Water Cooled Nuclear Power Plants RR 1.59 2 Design Basis Floeds for Nuclear P.cwer Plants RG 1.63 2 'lectric E Penetration Assemblics in Containment Structures for Light Watar Cooled Nuclear Power Plants RG 1.91 1 Evaluation of Exclcsions Postulated to Occur on Transportation Routes Near fluclear Power Plant Sites RG 1.102 1 Flood Protection for Nuclear Powar Plants RG 1.105 1 Instrument Sctpoints RG 1.108 1 Periodic Testing of Diesel Generator Units Used a? Onsite Electric Power Systems at Nuclear Power Plants RG 1.115 1 Protection Agains t Low-Trajectory Turbine Missiles RG 1.117 1 Tornado Design Classification RG 1.124 1 Service Limits and Loading Combinations for Class 1 Linear Type Component Supports RG 1.130 0 Design Limits and Loading Combinations for Class 1 Plant- and Shell-Type Component Supports (continued)

O B.18.1-2 Amendment 26 January 3,1979

~ -

-l-ENCLOSURE 2 DOCUMENT

!! UMBER REVISION TITLE RG 1.137 0 Fuel Oil Systems for Standby Diesel Generators (Paracraoh C.2)

RG 8.8 2 Information Relevant to Ensuring that Occupational Radiation Exoosures at Nuclear Power Stations Will be as Low as i Reasonably Achievable (Nuclear Power Reactors)

BTP ASB Guidelines for Fire Protection for Nuclear 9.5-1 Power Plants Under Review and Construction BTP MTEB 5-7 Material Selection and Processing Guide-lines for BWR Coolant Pressure Boundary Piping RG 1.141 0 Containment Isolation Provisions for Fluid Systems B.18.1-3 Amendment 26 January 3,1979

ENCLOSURE 3, CATEGORY III MATTERS APPROVED BY RRRC DOCUMENT NUMBER REVISION TITLE RG 1.99 1 Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials (Paragraphs C.1, C.2 and C.4)

RG 1.101 1 Emergency Planning for Nuclear Power Plants RG 1.114 1 Guidance on Being Operator at the Controls of a Nuclear Power Plant RG 1.121 0 Bases for Plugging Degraded PWR Steam Generator Tubes RG 1.127 1 Inspection of Water-Control Structures Associated with Nuclear Power Plants RSB 5-1 1 Branch Technical Position:

Residual Heat Removal System RSB 5-X 0 Branch Technical Position:

Reactor Coolant System Overpressurization Protection (Draft copy attached)

RG 1.97 1 Instrumentation for Light Water Reactor Nuclear Power Plants to Assess Plant Conditions During and Following an Accident (Paragraph C.3 - with additional guidance on paragraoh C.3.d to be provided later)

RG 1.68.2 1 Initial Startup Test Program to Demonstrate Remote Shutdown Capability for Water-Cooled Nuclear Power Plants RG 1.56 0 Maintenance of Water Purity in Boiling Water Reactors O

B.18.1-4 Amendment 26 January 3,1979 s

EflCLOSURE 3 BRANCH TECHNICAL p0 SIT!CN DSB 5-2 OVERPRE55URIZATION PROTECTION OF PRESSURIZED WATER REACTORS Wil!LE OPERATIflG AT LOW TEMPERATURES A. Backnruund General Design Criterion 15 of Appendix A, 10 CFR 50, requires tha t "the Reactor Cnolant System and associa ted auxiliary, ccntrol, and protection systems shall be designed with sufficient ra. gin to assure tha t the design conditions of the reactor coolant cressure boundary are not exceeded during any condition of norma.1 operation, including anticipated operational occurrences."

Anticipated operational occurrences, as defined in Apoendix A of 10 CFR 50, are "those conditions of normal oceration wnich are expected to occur cne, or more times during the life of the nuclear pcwer unit and include but are not limited to loss of power to all recirculation pumps, tripping cf the turbine generator set, isolation of the main condenser, and loss of all of fsi te power."

Aopendix G of 10 CFR 50 provides the fracture toughness reouireirents for reactor pressure vessels under all conditions. To assure that the Aonendix G limits of the reactor coelant pressure boundary are not exceeded durino any anticipated Opera tional occurrences , Technic 31 Specification pressure-temperature limits are provided for operating the plant.

The primary concern of this position is that during startup and shutdown e.ondi tions a t low tempera ture , especially in a wa ter-solid condi tico ,

the reactor coolant system cressure might exceed the reactor ve3sel pressure-temperature limitations in the Technical Specifications established for crotection egainst brittle fracture. This inadvertent overpressuri:ation could be generated by an;. one of a variety of mal-functions or operator errors. Many incidents have occurred in operating plants as described in Reference 1.

Additicnal discussion on the background of this position is contained in Reference 1.

B .18.1 -5 Amendment 26 January 3, 1979

EfiCLOSUFE 3 B. Granch Pusi tion

1. A* system should be desiqned and installed which will prevent exceeding the acclicable Technical Scecifications and Appendix G limits for the reactor coolant system wnile operaticn ac !cw temperatures. The system should be c3pable of relieving ressure during all anticipated overcressurization events at a rate suf ficient to satisfy the Technical Specification limits, particularly while the reactor coolant system is in a water-solid condition.

2. The system must be able to perform its function assuming any single active component failure. Analyse; using accropriate calcula tional techniques mus t be provided wnich demons trate tha t the sys tem will provide the required pressure relief capacity assuming tl'e most, limiting single active failure. The cause for initiation of' the event, e.g. , operator error, ccmpenent mal function, will not be considered as the sirigle active f ailure. The analysis sFculd assunte the ecst limiting allowable operating conditions and systems configuration at the time 'of the pcs tulated cause of the overoressure event. All potential everpressuri:ation events must be censidered when establishing the worst case event. 3 cme events may be prevented by protective interlocks or by locking cut cower, these events should he reviewed on an individual basis. !f the interlock / power loc'<out is acceptable, it can be excluded frcm the analyses provided the controls to prevent the event are in the plant Technical Specifications.

3. The system must meet the design requirements of IEEE 279 (see Impl emen ta t ion) . The system may be -anually enabled, hewever, the electrical ins trumenta tion and control system must provide alarms to alert the operator to:

a. Drocerly enable the sys tem at the correct plant condition during cooldcwn,

b. indicate if a pressure transient is occurring.

4. To assure operational readiness, the overpressure proteccion system must be tested in the following manner:

a. A test must be perferred to assure operability of the system electronics prior to each shutdown,

b. A tast for valve cperability must, as a minimum be conducted as specified in the ASME Code Section x[.

c. Subsequent to system, valve, or electronics maintenance, a tes t on that portion (s) of the system must be performed prior to declaring the system operational.

O B.18.1-6 Amendment 26 January 3,1979

EtlCLOSUP.E 3

5. The system must meet the reouirements of Regulatory Guide 1.26.

" Quality Group Classifications and Standards for Water , 5 team .

and Radicactive-Waste-Containing Components of Nuclear rower Pi a.9ts" and Section !!! of the ASME Code.

6. The overpressure protection system mus t be designed to function during an Ocerating Basis Earthquake. I t must not ccmpecat se tfte design criteria of any other safety-grade system with which it would interf ace, such that the rer:yirements of Regulatory Guide 1.29, " Seismic Design Classificatien" are met.

7. The overpressure protection system tw not depend on the availability of offsite power to perform its function.

8. Overpressure protection systems which take credit for an active.

cnmponent(s) to mitigate the consequences of an overpressuri'2aticn event must include additional analyps considering inadvertent system initiation / actuation or provide justification to snow that existing analyses bound such an event.

C. Imol emen ta tion The Branch Technical Position, as specified in Section 3. will be used in the review of all Preliminary Cesign approval (PDA), Final Cesign Approval (FCA) . Manufacturing License (ML), Operating License (0L), and Construction Permit (CP) applications involving plant designs inccrporating pressurized wa ter reactors. All aspects of the position will be applicable to all acolications, including CP acolications utilizin; the rept ica tion option af the Connissicn* s standardizaticn proge?m, tha t are dockeled after March 14 1978. All aspects of the position, with the exceptiori of reasonable ind justified deviations ' rem IEEE 279 ret;uirerents , will be applicable to CP, OL, ML, PDA, and c0A analications decketed orior to March 14 1978 but for which the licensing action has not been cnmoleted as of March 14, 1978. Holders of apprcoriate PCA's will te informed by letter that all aspects of the pnsition with the exception of IEEE 279 will be apolicable to their approved standard designs m that such designs shculd be modified, as necessary, to conform to ta position. Sta ff approval of proposed modifications can be applied for either by applica tion by the POA-holder en the PDA-decke! Or by each CP applicant referencing the standard design on its docket.

The folicwing guidelines may be use<!, if necessary, to alleviate impacts on licen:ing schedules for plants involved in licensing proceedings nearing completion on March 14, 1978:

B .18.1 -7 Amendment 26 January 3,1979

ENCLOSURE 3

1. Thute applicants issued an OL during the ceriod between March 14 ,

1978 and a date 12 months thereaf ter may merely cc:n'it to meeting the position prior to OL issuance but shall, by license condit.lon, be required to install all required staf f-approved c:ccifications prior to plant startup following the first scheduled rd(ueling outage.

2. Those applicants issued an OL beyond f 9.i ch la ,1979 shall install all required staff-approved modifications prior to initial plant stJrtuD.

3. Those apolicants issued a CP, PCA, er ML during the ceriod bet.<een March 14, 1978 and a date 6 months therea f ter may 'rerely c0initi t to meeting the position but shall, by license condition, be required to amend the application, within 6 months of the date of issuance cf the CP, P0A, or ML, to incluce a description of the proposed modifications and the bases for their design, and a reques t for staf f approval .

4 Those applicants issued a CP, PDA, or ML af ter September li, 1973 shall have staff approval of proposed modifications prior to issuance of the CP, PDA, or ML.

D. References

1. NUREG-Cl38, Staf f Discussion of Fif teen Tecnnical :ssues Listed in Attachment to November 3, 1976 Memorandum from Director,llRR, to flRR Staf f.

O B.18.1-8 Amendment 26 January 3,1979

B.18.2 APPLICANTS RESPONSES NRC Document .

Number / Revision Response A. Category II Matters R.G. 1.27/2 Appendix 3D R.G. 1.52/1 Appendix 3D R.G. 1.59/2 Appendix 3D R.G. 1.63/2 Revision 2 to R.G. 1.63 was issued in July 1978, after the January 1,1978 cutoff date.

However, since the plant design is currently in compliance with the regulatory guide re-quirements, the guide revision is addressed in Amendment 26. See Appendix 3D.

R.G. l.91/1 Revision 1 to R.G. 1.91 was issued in Febru-ary 1978, after the January 1,1978 cutoff date. This guide establishes structural de-sign criteria which the Applicant wishes to adopt at this time but which require Plant 26 Design Report changes for implementation.

This guide revision is therefore addressed in Amendment 26. Appendix 3G has been deleted and the location of other relevant information may be determined by reference to Appendix 3D.

R.G. 1.102/1 Appendix 3D R.G. 1.105/1 Appendix 3D R.G. 1.108/l Appendix 3D R.G. 1.115/1 Appendix 3D R.G. 1.117/1 Revision 1 to R.G. 1.117 was issued in April 1978, after the January 1,1978 cut-off date. However, since the plant design is currently in compliance with the regula-tory guide requirements, the guide revision is addressed in Amendment 26. See Appendix 3D.

Amendment 26 January 3,1979 B.18.2-1

NRC Document Number / Revision Response R.G. 1.124/1 Revision 1 to R.G. 1.124 was issued in January 1978, after the January 1,1978 cutoff date. This guide revision is ad-drcssed in Amendment 26 because the subject matter is closely related to that of R.G.

1.130 (which the Applicant is required to address). See Appendix 3D.

R.G. 1.130/0 Appendix 3D R.G. 1.137/0 This guide was issued after the January 1, 1978 cutoff date and will be addressed during the final design approval phase.

R.G. 8.8/2 Appendix 3D BTP ASB 9.5-1/0 In a letter from R.S. Boyd dated September 30, 1976, Applicant was directed to address the fire protection guidelines contained in Appendix A to Branch Technical Position 9.5-1. Applicant's evaluation is contained in topical report RP06A30, " Floating Nuclear Power Plant Fire Protection Evaluation." 26 BTP ffTEB 5-7/0 This branch technical position does not apply to the Floating Nuclear Plant.

R.G. 1.141/0 This guide was issued after the January 1, 1978 cutoff date and will be addressed during the final design approval phase.

B. Category III Matters R.G. 1.99/1 Appendix 3D R.G. 1.101/1 Appendix 3D R.G. 1.114/1 Appendix 3D R.G. 1.121/0 Appendix 3D R.G. 1.127/1 Revision 1 to R.G.1.127 was issued in March 1978, after the cutoff date of January 1,1978. However, since the guide does not apply to activities performed by the Applicant, the guide revision is addressed in Amendment 26. See Appendix 3D.

Amendment 26 January 3, 1979 B.18.2-2

NRC Document flumber/ Revision Response BTP RSB 5-1/1 A. Functional Requirements The following changes will be made to the FNP to comply with the stated functional requirements:

1. The control system on the atmospheric power operated relief valves in the main steam system will be upgraded to Class lE.

2. a) The controls for the two pressurizer power-operated relief valves will be upgraded to Class lE, or b) The controls for the a Wiliary pres-surizer spray valve will be upgraded to Class lE. To accommodate a single failure in the auxiliary pressurizer spray line, the necessary parallel piping and valves will be added to the spray line, or_ the controls for at least one of the two pressurizer power-operated relief valves will be upgraded to Class lE. 26 With these changes, the design of the FNP will be such that the reactor can be taken from normal operating conditions to cold shutdown in conformance with the NRC Branch Position. Boration can be provided by the Safety Injection System from the Boron In-jection Tank and the Refueling Water Storage Tank.

B. RHR System Isolation Requirements

1. The suction side of the RHR system meets the stated isolation requirements. Two motor operated valves in series are pro-vided in each suction line, with opera-tion from and position indication in the control room. Pressure interlocks are provided on these valves to prevent their opening, or to close them, if the Reactor Coolant System pressure is above the RHR design pressure. No single failure will prevent isolation of the RHR suction lines.

The interlocks are redundant but not di-verse. OPS believes interlock diversity is not necessary to accomplish the in-tended isolation function.

Amendment 26 January 3,1979 B.18.2-3

NRC Document Number / Revision Response BTP RSB 5-1/1 (Cont'd) 2. The discharge side of the RHR system is isolated by two check valves in series. Design provisions are in-included to pennit periodic leak tightness testing of these valves.

These check valves are part of the Safety Injection System and are shown in Figure 6.3-1, Sheets 2 and 3.

C. Pressure Relief Requirements (in RHR System)

Each RHR System suction line incorporates a pressure relief valve in accordance with the NRC Branch Position.

D. Pump Protection Requirements The FNP design meets the stated require-ments.

E. Test Requirements The FNP design meets the stated require- 2 ments. R.G. 1.22 and R.G. 1.68 are ad-dressed in Appendix 30. We expect that the requirement for tests and analyses to confirm adequate mixing of borated water and natural circulation will be met by comparison with performance of pre-viously tested plants of similar design.

F. Operational Procedures This is in the utility-owner's scope.

R.G. 1.33 is addressed in Appendix 30.

G. Auxiliary Feedwater Supply The FNP design meets the stated require-ments. Refer to Section 10.4.6.7.

Arrendment 26 January 3,1979 B.18.2-4

NRC Document Number / Revision Response BTP RSB 5-2/0 -Administrative procedures are developed to aid the operator in controlling RCS pressure during low temperature operation. However, to provide a backup to the operator, an auto-matic system will be provided to comply with NRC Branch Technical Position RSB 5-2. The Applicant is evaluating design alternatives (including the use of pressurizer power oper-ated relief valves as described in the SNUPPS application and RESAR-414) and will submit 26 design modifications as part of the final plant design. As agreed in Mr. H.R. Denton's letter of December 22, 1978, a comitment by Offshore Power Systems on a generic issue such as this is adequate for the Manufacturing License.

R.G. 1.97/1 Appendix 3D R.G. 1.68.2/1 Revision 1 to R.G. 1.68.2 was issued in July 1978, after the January 1,1978 cutoff date.

However, since the guide does not apply to activities performed by the Applicant, the guide revision is addressed in Amendment 26.

See Appendix 3D.

R.G. 1.56/0 Appendix 3D Amendment 26 January 3,1979 B .18. 2- 5