ML20148K166

Amend 1 to Chapters 1-3,Advanced BWR Ssar

ML20148K166
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Site:	05000605
Issue date:	03/29/1988
From:	GENERAL ELECTRIC CO.
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NUDOCS 8803310029
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O CHAPTER 1 INTRODUCTION AND GENERAL O DESCRIPTION OF PLANT I

l 3

O RBR A888 8888810s A DCD i

ABWR usawac Standard Plant REV.A CHAPTER 1 O TAnts Or COsTEsTs Section Iith Page 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT 1.1 INTRODUCrlON 1.1 1 1.1.1 Type of License Required 1.1-1 1.1.2 Number of Plant Units 1.1-1 1.13 Description of Location 1.1-1 1.1.4 Type of Nuclear Steam Supply 1.1-1 1.1.5 Type of Containment 1.1 1 1.1.6 Core Thermal Power Levels 1.1-1 1.2 GENERAL PLANT DESCRil' TION 1.2-1 1.2.1 Principal Design Criteria 1.21 V 1.2.2 Plant Description 1.2-5 13 COMPARISON TAELES 13-1 13.1 Nuclear Steam Supply System Design Characteristics 13-1 13.2 Engineered Safety Features Design Characteristics 13-1 133 Containment Design Characteristics 13-1 13.4 Structur=1 Design Characteristics 13-1 13.5 Instrumentation and Electrical Systems Design Characteristics 13-1 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS 1.4-1 O

'%,0 1 il Amendment 1 I

ABM ux6iooac Standard Plant REV.A CHAPTER 1 4

TABLE OF CONTENTS (Continued) O Section lille Eage 1.5 REOUIREMENTS FOR FURTIIER TECIINICAL INFORMATION 1.5-1 1.6 MATERIAL INCORPORATED HY REFERENCE 1.6-1 1.7 DRAWINGS AND OTilER DETAILED INFORMATION 1.7-1 1.8 CONFORMANCE WTTH STANDARD REVIEW PLAN AND APPLICABILITY OF CODES AND STANDARDS 1.81 1.8.1 Conformance With Standard Review Plan 1.8-1 1.8.2 Applicability of Codes and Standards 1.8-1 1.9 INTERFACES 1.9 1 APPENDIX 1A UNRESOLVED SAFETY ISSUES 1A-1 O

O 1 iU Amendment I

ABWR useiwac Standard Plant REV.A SECTION 1.1 O

v CONTENTS Section Titic Eage 1.1.1 Tme of License Reautred 1.1-1 1.1.2 Number of Plant Units 1.1-1 1.1.3 Description of Location 1.1 1 1.1.4 Troe of Nuclear Steam Supolv 1.1-1 1.1.5 Troe of Containment 1.1-1 1.1.6 fnra Thermal Power inels 1.1 1 ILLUSTRATIONS Figure Iills Eage O 1.1-1 ABWR Standard Plant Nomenclature 1.1 2 1.1-2 Heat Balance at Rated Power 1.1 3 O

1.1 il Amendment 1

ABWR ua6iooAc Standard Plant REV.A

1.1 INTRODUCTION

1.1.2 Numberof Plant Units kJ The Advanced Boiling Water Reactor Standard For the purposes of this document, only a Safety Analysis Report (ABWR SSAR),is written in single standard plant will be considered.

accordance with Regulatory Guide 1.70. For consistency with NUREG 0800 the ABWR SSAR 1.1.3 Description oflAcation includes Section 15.8 which addresses anticipated transients without scram and Chapter 18 which This plant can be constructed at any location addresses human factors. In addition, response which meets the parameters identified in Chapter to TMI related matters is presented in Appendix 2.

1A.

1.1 A Type of Nuclear Steam Supply The response to the severe accident policy statement is provided in Chapter 19. Chapter 20 This plant will have a boiling water reactor is included to provide a section for a question nuclear steam supply system designed and l and response guide, supplied by GE and designated as ABWR.

1.1.1 Type of License Required 1.1.5 Type of Containment This ABWR SSAR is submitted in support of the The ABWR will have a low leakage containment 1 application for final design approval (FDA) and vessel which comprises the drywell and pressure l design certification (DC) for the ABWR, a suppression chamber. The containment vessel is l standard plant with a Nuclear Island scope. The a cylindrical steel lined reinforced concrete '

Nuclear Island consists of all buildings which structure integrated with the reactor building. )

are dedicated exclusively or primarily to housing The containment nomenclature is specified in ,

systems and the equipment related to the neclear Figure 1.1-1. j system or controls access to this equipment and

')

1 systems. Under the Nuclear Island concept 1.1.6 Core Thermal Power Levels )

presented herein, there are three buildings that  !

comprise the Nuclear Island. These are: The information presented in this ABWR SSAR pertains to one reactor unit with a rated power (1) Reactor Building (including containment); level of 3926 hfWt and a design power level of 4005 h!Wt. The station utilizes a single cycle, (2) Service Building; and forced-circulation, boiling water reactor (BWR). The heat balance for rated power is  ;

(3) Control Building. shown in Figure 1.1-2. The station is designed to operate at a gross electrical power output of For each of the buildings, the Nuclear Island approximately 1356 htWe and net electrical power scope inci' ides the design of all structures output of approximately 1300 htWc.

including tne foundation mats and everything within these structures.

I System boundaries may vary with the nature of the interf ace, but the general rule of determining interfaces is that boundaries extend to just outside of the building walls. A major factor in the design process is the determination of the exact description of the interface.

Parameters such as dimensions and orientation of the pipes, type of connections, and pressures at the interface points are established and identified in Section 1.9.

0(~%

Amendment 1 1.11

ABM 23A6iooAC Standard Plant REV.A O

1 r SECONDARY CONTAINMENT l5 BOUNDARY l ,

DRYWELL *

• CONNECTING l VENT DRYWELL l

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s PRIMARY CONTAINMENT l 1 Nw VESSEL

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! k SUPPRESSION k l

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k HORIZONTAL k

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(km m_ h-._...-..J BASE MAT

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PRIMARY '

CONTAINMENT SOUNDARY Figure 1.1-1 ABWR STANDARD PLANT NOMENCLATURE O

Amendment 1 1.12

ABWR unamae Standard Plant suW A O

ASSUMED SYSTEM LOSSES 1040P ' s THERMAL 1.1 mw LEGEND 4 1 1 MAIN STEAM FLOW 1684.3 x 104 G l

G = F LOW,Ib/hr

= TEMPER ATURE, 'F

~

11 .8 h T ,q h = ENTHALPY, Btu /lb -

985 P M = % MOISTURE (TURBINE INLET)

P = PRESSURE, psia MAIN FEED FLOW M = ISOLATION VALVES /

3926 MW, N 4

1714.3 x 10 G 1580.7 x 104 G 398.0 h n 397.6 h 420 T 420 T TOTAL CORE FLOW 8

115.1 x 10 0.15 Ahpuup

/ 415.4 h 436 T J ), b di ha/

527.6 h CLEANUP ,

RIP AND jL DEMINERALIZER '

SYSTEM CONTROL 3.62 x 104 G ROD DRIVE 61.9 h l PURGE FLOW 91 T 33.62 x 104 G I 527.6 h 533 T v

87 592 01 l

i Figure 1.1-2 HEAT BALANCE AT RATED POWER 9! l Amendinem i 1.13

ABM 2346ioaac Standard Plant REV.A SECTION 1.2 O CONTENTS Section Title Page 1.2.1 Princloal Deslan Criteria 1.2-1 1.2.1.1 General Design Criteria 1.2-1 1.2.1.1.1 Power Generation Design Criteria 1.21 1.2.1.1.2 Safety Decign Criteria 1.21 1.2.1.2 System Criteria 1.23 1.2.1.2.1 Nuclear System Criteria 1.2-3 1.2.1.2.2 Electrical Power Systems Criteria 1.2-4 1.2.1.23 AuxiliaryS 3stems Criteria 1.2-4 1.2.1.2.4 Shielding and Access Control Criteria 1.2-4 1.2.1.2.5 Process Control Systems Criteria 1.2-4 1.2.1.2.5.1 Nuclear System Process Control Criteria 1.2-4 1.2.1.2.5.2 Electrical Power System Process Control Criteria 1.2-4 1.2.2 Plant Description 1.2-5 1.2.2.1 Site Characteristics 1.2 5 1.2.2.1.1 Site Location 1.25 1.2.2.1.2 Description of Plant Environs 1.2-5 1.2.2.1.2.1 Meteorology 1.2-5 1.2.2.1.2.2 Hydrology 1.2-5 1.2.2.1.23 Gcology and Seismology 1.2-5 1.2.2.2 General Arrangement of Structures and Equipment 1.2 5 0 1.2 ..

u Amendment 1

\ 33A6100AC Standard Plant REV.A SECTION 1.2 CONTENTS (Continued) O Section Title Page 1.2.23 Nuclear Systems 1.2 5 1.2.23.1 Reactor Core and Control Rods 1.2-5 1.2.23.2 Reactor Vessel and Internals 1.2-6 1.2.233 Reactor Internal Recirculation Pumps (RIP) 1.2-6 1.2.23.4 Residual Heat Removal System 1.2-6 1.2.23.5 Reactor Water Cleanup System 1.2-6 1.2.23.6 leak Detection and Isolation System 1.2-6 1.2.2.4 Nuclear Safety Systems 1.2-7 1.2.2.4.1 Reactor Protection System 1.2-7 1.2.2.4.2 Neutron hiontoring System 1.2-7 1.2.2.43 Control Rod Drive System 1.2 7 h

1.2.2.4.4 Control Rod Drive Housing Supports 1.2-7 l

1.2.2.4.5 Control Rod Braking Mechanism 1.2 7 1.2.2.4.6 Nuclear System Pressure Relief System 1.2-7 1.2.2.4.7 Reactor Core Isolation Cooling (RCIC) System 1.2-7 1.2.2.4.8 Emergency Core Cooling Systems 1.2-8 1.2.2.4.8.1 High Pressure 1.2-8 1.2.2.4.8.1.1 High Pressure Core Spray (HPCF) System 1.2-8 1.2.2.4.8.1.2 RCIC Description 1.28 1.2.2.4.8.2 Automatic Depressurization System (ADS) 1.28 1.2 iii 9

Amendment I

ABM 23461ooac Standard Plant ' REV.A SECTION 1.2 O coureurs (contiooea) '

Section Iltle Page 1.2.2.4.83 Low Pressure Flooder Loop (LPFL) 1.2-9 1.2.2.4.9 Containmeot 1.29 1.2.2.4.9.1 Functional Design 1,2-9 1.2.2.4.9.2 Heat Removal 1.2-9 1.2.2.4.93 Emironmental Systems :1.2-9 1.224.9.4 Wetwell/Drywell Spray 1.2-9 1.2.2.4.9.5 Flammability Control 1.29 1.2.2.4.10 Containment and Reactor VesselIsolation Control System 1.2-9 1.2.2.4.11 hiain Steamline Isolation Valves 1.2-9 1.2.2.4.12 hiain Steamline Flow Restrictors 1.2 10 0 1.224.13 Radiation hionitoring and Control 1.2 10 1.2.2.4.14 Residual Heat Removal (RHR) System (Containment Cooling) 1.2-10 1.214.15 Process Radiation hionitoring System 1.2 10 1.224.16 Standby Gas Treatment System 1.2-10 1.2.2.4.17 Reactor Building Isolation Control System 1.2-10 1.2.2.4.18 Standby AC Power Supply 1.2-10 1.2.2.4.19 DC Power Supply 1.2-11 1.214.20 Standby Liquid Control System 1.2-10 1.2.2.4.21 Remote Shutdown System 1.2-11 1.214.22 Condensate Storage Facilities and Distribution System 1.2-11 1.2.2.5 Electrical Systems and Instrumentation and Control 1.2-11 0 1.2iv Amendment 1 i ..

ABWR usamac Standard Plant arv 4 SECTION 1.2 CONTENTS (Continued) O Section Title Eage 1.2.2.5.1 Electrical Power Systems 1.2 11 1.2.2.5.1.1 Unit Auxiliary AC Power System 1.2-11 1.2.2.5.1.2 Unit Class 1E AC Power System 1.2-11 1.2.2.5.13 Safety System Protection System Power System 1.2-11 1.2.2.5.1.4 Instrument Power System 1.2-11 1.2.2.5.1.5 Uninterruptible Power System 1.2 12 1.2.2.5.1.6 Unit Auxiliary DC Power System 1.2-12 1.2.2.5.1.7 Unit Class IE DC Power System 1.2-12 1.2.2.5.2 Nuclear System Process Control and Instrumentation 1.2 12 1.2.2.5.2.1 Rod Control and Information System 1.2 12 1.2.2.5.2.2 Recirculation Flow Control System 1.2-12 g 1.2.2.5.23 Neutron Monitoring System 1.2-12 1.2.2.5.2.4 Refueling Interlocks 1.2-12 1.2.2.5.2.5 Reactor VesselInstrumentation 1.2-12 1.2.2.5.2.6 Process Computer System 1.2 13 1.2.2.6 Feedwater Control System 1.2 13 1.2.2.7 Fuel Handling and Storage Systems 1.2 13 1.2.2.7.1 New and Spent Fuel Storage 1.2 13 1.2.2.7.2 Fuel Handling System 1.2-13 1.2.2.8 Cooling Water and Auxiliary System 1.2 13 1.2.2.8.1 Reactor Building Cooling Water System 1.2 13 1.2 v 0

Amendment 1

ABM ursiooac Standard Plant any.A SECTION 1.2 O coureurs (continuea)

Section .7dtle Egge 1.2.2.8.2 Fuel Pool Cooling and Cleanup System (FPCS) 1.2-13 1.2.2,83 Makeup Water (Condensate) System 1.2-14 1.2.2.8.4 Plant Chilled-Water System 1.2-14 1.2.2.8.5 Process Sampling Systems 1.2-14 1.2.2.8.6 Plant Equipment and Floor Drainage 1.2-14 1.2.2.8.7 Service and Instrument Air Systems 1.2 14 1.2.2.8.8 High Pressure Nitrogen (HPIN) Gas Supply 1.2 14 1.2.2.8.9 Diesel Generator Fuel-Oil Storage and Transfer System 1.2-14 1.2.2.8.10 Heating, Ventilating, and Air Conditioning (Emironmental) System 1.2-15 O 1.2.2.6.11 Lighting Systems 1.2-15 1.2.2.8.12 Fire Protection System 1.2-15 1.2.2.9 Radiation Monitoring and Control 1.2-15 l

1.2.2.9.1 Process Radiation Monitoring 1.2 15 1.2.2.9.2 Area Radiation Monitors 1.2 16 1.2.2.93 Dust Radiation Monitors 1.2 16 1.2.2.9.4 Containment Atmospheric Monitors 1.2-16 1.2.2.10 Shielding 1.2-16 O

1.2-si Amendment 1

23A6100AC Standard Plant _

RI?V. A SECTION 1.2 ILLUSTRATIONS O Figure 11tle Page 1.21 Site Plan 1.2-17 1.2-2 Reactor, Reactor Building, Primary Containment Arrangement Plan, Section 0/180 1.2 18 1.2-3 Reactor, Reactor Building, Primary Containment Arrangement Plan, Section 270/90 1.2-19 1.2-4 Reactor, Reactor Building, Primary Containment Arrangement Plan at Floor El(-) 0200 mm 1.2-20 1.2-5 Reactor, Reactor Building, Primary Containment Arrangement Plan at Floor El (-) 6700 mm 1.2 21 1.2-6 Reactor, Reactor Building, Arrangement Plan at Floor El (-) 6700 mm 1.2 22 1.27 Reactor, Floor Reactor Building, Primary 0

Containment Plan at Ground Floor lil 7300 mm 1.2-23 1.2-8 Reactor, Reactor Building, Primary Containment Plan at El9500 ram 1.2-24 1.2-9 Reactor, Reactor Building, Primary Containment Plan at El 11500 mm 1.2-25 1.2-10 Reactor, Reactor Building. Primary Containment Plan at El 12500 mm 1.2-26 1.2 11 Reactor, Reactor Building, Primary Containment Plan at El 16500 mm 1.2 27 1.2 12 Reactor, Reactor Building Plan at El 18700 mm 1.2 28 1.2 13 Reactor, Reactor Building Plan at El 26700 mm 1.2 29 1.2 vii 9

Amendment 1

23A6100AC Standard Plant RIIV. A SECTION 1.2 O ILLUSTRATIONS Figure Title Page 1.2 14 Control and Senice Building Plot i Plan 1.2-30 1.2-15 Control and Senice Building Typical Section 1.2-31 1.2-16 Control and Senice Building Piping interface at El (-) 3500 mm 1.2-32 1.2-17 Control and Service Building Clean Area Access at El(-) 0200 mm 1.2-33 1,2 18 Control and Senice Building E/I I Interface at El 4300 mm 1.2-34 1.2 19 Co, trol and Senice Building Main Access and Controlled Area Access at El 7300 mm 1.2 35

, 1.2-20 Control and Senice Building Steam O T ci et ei 127oo -- 1.2-36 1.2 21 Control and Senice Building T/C Rooms at El 17000 mm 1.2-37 1.2-22 Control and Senice Building - Control Room at El 23100 mm 1.2-38 O

1.2-sili Amendment 1

I 23A6100AC Standard Plant any A 1.2 GENERAL PLAN'I' DESCRIPTION (8) Reactor controle, including alarms, are arranged to allow the operator to rapidly 1.2.1 Principal Design Criteria assess the condition of the reactor system and locate system malfunctions.

The principal design criteria are presented in two ways. First, they are classified as either a (9) Interlocks or other automatic equipment are power generation function or a safety function, provided as backup to procedural control to Second, they are grouped according to system, avoid conditions requiring the functioning Although the distinctions between power genera- of nuclear safety systems or engineered tion or safety functions are not always clear cut safety features.

and are sometimes overlapping, the functional classification facilitates safety analyses, while (10) The station is designed for routine the grouping by system facilitates the under- continuous operation whereby steam standing of both the system function and design. activation prnducts, fission products, corrosion products, and coolant dis-1.2.1.1 General Design Criteria sociation products are processed to remain within acceptable limits.

1.2.1.1.1 Power Generation Design Criteria 1.2.1.1.2 Safety Design Criteria (1) The plant is designed to produce steam for direct use in turbine generator unit. (1) The station design conforms to applicable codes and regulations as described in Table (2) Heat removal systems are provided with 1.8.

sufficient capacity and operational adequacy to remove heat generated in the reactor (2) The station is designed, fabricated, core for the full range of normal erected, and operated in such a way that operational conditions and abnormal the release of radioactive material to the C, operational transients, environment dees not exceed the limits and guideline values of applicable pvernment (3) Backup heat removal systems are provided to regulations pertaining to the release of remove decay heat generated in the core radioactive materials for normal opera-under circumstances wherein the normal tions, for abnormal transients; and for operational heat removal systems become accidents.

inoperative. The capacity of such systems is adequate to prevent fuel cladding damage. (3) The reactor core is designed so its nuclear characteristics do not contribute to a l (4) The fuel cladding in conjunction with other divergent power transient. )

plac'. systems is designed to retain j integrity so that the consequences of any (4) The reactor is designed so there is no l f ailures are within acceptable limits tendency for divergent oscillation of any I throughout the range of normal operational operating characteristic considering the conditions and abnormal operatier:al interaction of the reactor with other transients for the design life of the fuel, appropriate plant systems.

(5) Control equipment is provided to allow the (5) The design provides means by which plant reactor to respond automatically to load op:rators are alerted when limits on the changes and abnormal operational transients, release of radioactive material are approached. 1 (6) Reactor power level is manually control-icble. (6) Sufficient indications are provided to allow determination that the reactor is (7) Control of the reactor is possible from a operating within the envelope of conditions single location, considered safe by plant analysis.  !

Amendment 1 1.21 l

1

ABWR m6imac Standard Plant myA (7) Radiation shiciding is provided and access allowances for natural environmental control patterns are established to allow a properly trained operating staff to control disturbances such as carthquakes, floods, and storms at the station site.

g radiation doses within the limits of applicable regulations in any mode of normal (15) Standby electrical power sources have plant operations. sufficient capacity to power all safety related systems requiring electrical power (9) Those portions of the nuclear system wt concurrently.

form part of the reactor coolant pressure boundary are designed to retain integrity as (16) Standby electrical power sources are a radioactive material containment bauier provide 1 to allow prompt reactor shutdown following abnormal operational transients and removal of decay heat under circum-and accidents. stances where normal auxiliary power is not available.

(9) Nuclear safety systems and engineered safety features function to assure that no damage (17) A containment is provided that completely to the reactor coolant pressure boundary encloses the reactor systems, drywell, and results from internal pressures caused by suppression chambers. The contain ment abnormal operational transients and employs the pressure suppression concept.

accidents.

(18) l' is possible to test primary contain.

(10) Where positive, precise action is immediate- ment integrity and leak tightness at ly required in response to abnormal opera- periodic intervals, tional transients and accidents, such action is automatic and requires no decision or (19) A secondary containment is provided that manipulation of controls by plant operations completely encloses the primary containment personnel, above the reactor building basemat. This (11) Safety related actions are provided by secondary containment provides for a controlled, monitored release of any h

equipment of sufficient redundance and inde- potential radioactive leakage from the pendence so that no single failure of active primary containment, components, or of passive amponents ;a cer-tain cases in the long term, will prevent (20) The primaa containment and secondary the required actions. For systems or compo- containment in conjunction with other nents to which IEEE-279 apply, single fail- safety related features limit radio-urcs of either active or passive electrical logical effects of accidents resulting in somponents are considered in recognition of the release of radioactive material to the the higher anticipated failure rates of containmen: volumes to less than the l passive electrical components relative to prescribed acceptable limits. l passive mechanical components.  !

(21) Provisions are made for removing energy (12) Provisions are made for control of active from the primary containn ent as necessary com;mnents of safety related systems from to maintain the integrity of the the control room. contairiment system following ac idents that l release energy to the containmcnt. I (13) Safety related systems are designed to permit demonstration of their functional (22) Piping t hat penetraics the primary  ;

performance requirements, containment and could serve as a path for i the uncontrolled release oi radioactive (14) The design of safety related systems, material to the environs is automatically components and rtructures includes isolated when necessary to limit the ol Amendment 1 112 l

l

ABM 23acimac Standard Plant ,RJea radiological impact from an uncontrolled 1.2.1.2 System Criteria release to less than acceptable limits.

q( ) The principal design criteria for particular (23) Emergency core cooling systems (ECCS) are systems are listed in the following subsections.

provided to limit fuel cladding temperature to less than the limits of 10CFR50.46 in the 1.2.1.2.1 Nuclear Sptem Criteria event of a loss of-coolant accident (LOCA).

(1) The fuel cladding is a radioactive material (24) The emergency core cooling systems provide barrier designed to retain integrity so for continuity of core cooling over the that failures do not result in dose complete range of postulated break sizes in consequences that exceed acceptable limits the reactor coolant pressure boundary. throughout the design power range.

(25) Operation of the emergency core cooling (2) The fuel cladding in conjunction with other systems is initiated automatically when plant systems is designed to retain required regardless of the availability of integrity so that the consequences of any offsite power supplies and the normal failures are within acceptable limits generating system of the station. throughout any abnormal operational transient.

(26) The control room is shielded against radia-tion so that continued occupancy under de- (3) Those portions of the nuclear system that sign basis accident conditions is possible, form part of the reactor coolant pressure boundary are designed to retain integrity (27) In the event that the control room becomes as a radioactive material barrier during inaccessible, it is pos.ible to bring the normal operation and foliowing abnormal reactor from power range operation to cold operational transients and accidents.

shutdown conditions by utilizing alternative

] controls and equipment that are available outside the control room, (4) The capacity of the heat removal systems provided to remove heat generated in the reactor core for the full range of normal (28) Backup reactor shutdown capability inde- operational transients as well as for pendent of normal reactivity control is abnormal operational transients is adequate provided. This backup system has the to prevent fuel cladding damage that capaulity to shut down the reactor from any results in dose consequences exceeding normal operating condition and subsc- acceptable limits.  !

quently to maintain the suhiown condition. '

(5) The reactor is capable of being shut down  !

(29) Fuel handling and storage facilities are automatically in sufficient time to permit designed to prevent inadvertent criticality decay heat removal systems to become and to maintain shielding and cooling of effective following loss of operation of spent fuel as necessary to meet operating normal heat removal systems. The capacity and offsite dose constraints. af such systems is adequate to prevent fuel cladding damage.

(30) Systems that have redundant or backup safety functions are physically separated, and (6) The reactor core and reactivity control arranged so that credible events causing system are designed such that control rod damage to one region of the reactor island action is capable of making the core complex has minimum prospect for subcritical and maintaining it even with compromising the functional capability of the rod of highest reactivity worth fully the redundant system. withdrawn and unavailable for insertion.

O Amendment 1 1.23

ABM MA6100AC Standard Plant anv.A (7) Backup reactor shutdown capability is radiation doses within the limits of applicable provided independent of normal reactivity regulations in any normal mode of plant control provisions. This backup system has operation, g

the capability to shut down the reactor from any operating condition and subsequently to 1.2.1.2.5 Process Control Systems Criteria maintain the shutdown condition.

The principal design criteria for the process (8) The nuclear system is designed so there is no control systems are as follows:

tendency for divergent oscillation of any operating characteristic, considering the 1.2.1.2.5.1 Nuclear System Process Control interaction of the nuclear system with other Criteria appropriate plant systems.

(1) Control equipment is provided to allow the 1.2.1.2.2 Electrical Power Systems Criteria reactor to respond automatically to load changes within design limits.

Sufficient normal auxiliary and standby sources of electrical power are provided to (2) It is possible to control the reactor power attain prompt shutdown and continued maintenance level manually, of the station in a safe condition under all credible circumstances. The power sources are (3) Nuclear systems process displays, controls adequate to accomplish all required essential and alarms are arranged to allow the safety actions under a!! postulated accident operator to rapidly asses the condition of conditions. the nuclear system and to locate process system malfunctions.

1.2.1.2.3 Auxiliary Systems Criterla 1.2.1.2.5.2 Electrical Power System Process (1) Fue! handling and storage facilities are Control Criteria designed to prevent inadvertent criticality and to maintain adequate shielding and (1) The Class 1E power systems are designed with cooling for spent fuel. three (3) divisions with any two divisions being adequate to safely place the unit in (2) Other auxiliary systems, such as service the hot shut down condition.

water, cooling water, fire protection, beating and ventilating, communications, and (2) Protective relaying is used to detect and lighting, are designed to function as needed isolate faulted equipment from the system

'during normal and/or accident conditions. with a minimum of disturbance in the event of equipment failure.

(3) Auxiliary systems that are not required to effect safe shutdown of the reactor or (3) Voltage relays are used on the emergency maintain it in a safe condition are designed equipment buses to disconnect the normal so that a failure of these systems shall not source in the event of loss of offsite power prevent the essential auxiliary systems from and to initiate starting of the standby performing their design functions. emergency power system diesel generators.

1.2.1.2.4 Shiciding and Access Control (4) The standby emergency power diesel ,

Criteria generators are started and loaded l

automatically. I Radiation shielding is provided and access control patterns are established to allc,w a (5) Safety related electrically operated break- ,

properly trained operating staff to control ers are controllable from the control room. l l

Amend e t1 1.2-4 l

ABM ua6imac Standard Plant RIN. A (6) Monitoring of essential generators, non-essential power, emergency core cooling p transformers, and circuits is provided in the main control room, systems, heating ventilating and air V conditioning and supporting systems; 1.2.2 Plant Description (2) Service Building - personnel facilities, non essential switchgear and portions of the 1.2.2.1 Site Characteristics non-essential heating and ventilating systems; 1.2.2.1.1 Site Location (3) Control Building - includes the control The plant is located on a site adjacent to or room, the computer f acility, the cable close to a body of water with sufficient capacity tunnels, some of the plant essential for either once through or recirculated cooling switchgear and the essential liVAC system.

or a combination of both methods.

The arrangement of these structures on the 1.2.2.1.2 Description of Plant Environs plant site is shown in Figure 1,2-1. Figures 1,2 2 through 1.2 22 show the equipment 1.2.2.1.2.1 Meteorology arrangement in the principal buildings.

The safety related structures and equipment 1.2.23 Nuclear Systems are designed to retain required functions for the loads resulting from any tornado with The nuclear system includes a direct-cycle characteristics not exceeding the values provided forced circulation boiling water reactor that in Table 2.0-1. produces steam for direct use in the steam turbine. A heat balance showing the major Tornado missiles are discussed in Section 3.5. parameters of the nuclear system for the rated power conditions is shown in Figure 1.12.

O V

1.2.2.1.2.2 Ilydrology 1.2.23.1 Reactor Core and Control Rods The safety design basis of the plant provides that structures of safety significance will be Fuel for the reactor core consists of unaffected by the hydrologic parameter envelope slightly enriched uranium dioxide pellets scaled defined in Chapter 2. in Zircaloy-2 tubes. These tubes (or fuel rods) are assembled into individual fuel assemblies.

1.2.2.1.2.3 Geology and Seismology Gross control of the core is achieved by movable, bottom entry control rods. The control The structures of safety significance for the rods are cruciform in shape and are dispersed plant are designed to withstand a safe shutdown throughout the lattice of fuel assemblies. The earthquake (SSE) which results in a freefield control rods are positioned by individual peak acceleration of 0.3g. control rod drives.

1.2.2.2 General Arrangement of Structures and Each fuel assembly has several fuel rods with Equipment axially zoned gadolinia Gd 02 3 mixed in solid solution with the UO 2. The Gd 023 is The principal structures located in the plant burnable poison which diminishes the reactivity nuclear island are the following: of the fresh fuel. It is depleted as the fuel reaches the end of its first cycle.

(1) Reactor Building includes the containment, drywell, and major portions of the nuclear A conservative limit of plastic strain is the steam supply system, steam tunnel, refueling design criterion used for fuel rod cladding area, diesel generators, essential power, failure. The peak linear heat generation for O

Amendment 1 1.25

ABM uA6iooAc Standard Plant anv. A steady-state operation is well below the fuel result in a leak flow that is less than the ECCS damage limit even late in life. Experience has capacity allowing full core coverage. The g shown that the control rods are not susceptible internal pumps are a wet taotor design with no w to distortion and have an average life expectancy shaft seals, thereby providing increased many times the residence time of a fuel loading. reliability, reduced maintenance requirements and decreased operational radiation exposure.

1.2.23.2 Reactor Vessel and Internals The RIP has a low rotating inertia. Coupled with the solid state adjustable speed drives The reactor vessel contains the core and (ASD) the RIP can respond quickly to load supporting structures; the steam separators and transients and operator demands.

dryers; the reactor internal pumps; the control rod guide tubes; the distribution lines for the 1.2.23.4 Residual ileat Removal Sptrm feedwater, core flooders, and standby liquid control; the in-core instrumentation; and other The residual heat removal (RHR) system is a components. The main connections to the vessel system of pumps, beat exchangers, and piping include steamlines, feedwater lines, control rod that fulfills the following functions:

drive and in core nuclear instrument housings, core flooder lines, residual heat removal lines, (1) removes decay and sensible heat during a d standby liquid control line, core differential after plant shutdown; pressure line, internal pump pressure sensing lines, and water level instrumentation. (2) injects water into the reactor vessel following a loss-of coolant accident to The reactor vessel is designed and fabricated reflood the core in conjunction with other in accordance with applicable codes for a core cooling systems (Subsection 1.2.2.4.8);

pressure of 1250 psig. The nominal operating pressure in the steem space above the separators (3) removes heat from the containment following is 1040 psia. The vessel is fabricated of low a loss of coolant accident to limit the alloy steel and is clad internally with stainless steel or Ni Cr-Fe Alloy (except for the top head, increase in containment pressure. This is accomplished by cooling and recirculating g

nozzles, and nozzle weld zones which are unclad). the suppression pool water.

The reactor core is cooled by demineralized 1.2.23.5 Reactor Water Cleanup System water that enters the lower portion of the core and boils as it flows upward around the fuel The reactor water cleanup system (RWCU) rods. The steam leaving the core is dried by recirculates a portion of tr actor coolant stear- separators and dryers located in the upper through a filter demineralizer to remove portion of the reactor vessel. The steam is then particulate and dissolved impurities from the directed to the turbine through the main reactor coolant. It also removes excess coolant steamlines. Each steamhne is provided with two from the reactor system under controlled isolation valves in sen::, one on each side of conditions and provides clean water for the the containment barrier. reactor head spray nozzle.

1.2.233 Reactor Interua! Recirculation Pumps 1.2.23.6 leak Detection an6 Isolation Spte:n The reactor internal pumps (RIPS) are internal The leak detection and isolation system pumps which provide a continuous internal circu- consists of temperature, pressure, flow, and lation path for the core coolant flow. The RIPS fission product sensors with associated are located at the bottom of the vessel. The instrumentation, starms, and isolation pump motors are enclosed in pressure housings functions. This system detects and annunciates under the vessel. A break in the housing will leakage (and closes isalation valves, as 3

Amendment 1 1.2-6

ABM usamac Standard Plant _ REV.A required)in the following systems: 1.2.2.4.3 Control Rod Drive System (1) main steamlines; When a scram is initiated by the RPS, the control rod drive (CRD) system inserts the (2) reactor water cleanup system (RWCU); negative reactivity necessary to shut down the reactor. Each control rod is normally (3) residual heat removal (RHR) system; controlled by an electric motor unit. When a scram signal is received, high. pressure water (4) reactor core isolation cooling (RCIC) system; stored in nitrogen charged accumulators forces the control rods into the core and the electric (5) feedwater system; motor drives are signalled to drive the rods '

, into the core. Thus, the hydraulic scram action (6) emergency core cooling systems (ECCS); and is backed up by an electrically energized insertion of the control rods.

(7) miscellaneous systems.

1.2.2.4.4 Control Rod Drive Housing Supports Small leaks generally are detected by monitoring the air coolers, condensate flow, Control rod drive housing supports are radiation levels, and drain sump fill up and located interna! to the reactor vessel and the pump-out rates. Large leaks are also detected by control rod drive. The supports limit the changes in reactor water level and changes in travel of a control rod in the event that a flow rates in process lines. control rod housing is ruptured. The supports prevent a nuclear excursion as a result of a 1.2.2.4 Nuclear Safety Systems housing failure and thus protects the fuel barrier.

1.2.2.4.1 Reactor Protection System 1.2.2.4.5 Control Rod Braking Mechanism The reactor protection system (RPS) initiates a rapid, automatic shutdown (scram) of the An electro mechanical braking mechanism is reactor. It acts in time to prevent fuel incorporated in each control rod to limit the cladding damage and any nuclear system process velocity at which a control rod can fall out of barrier damage following abnormal operational the core should a hydraulic line break and transients The reactor protection system failure of flatge bolts or a spool piece. This overrides all operator actions and process action limits the rate of reactivity insertion controls and is based on a fail safe design resulting from a rod drop accident.

philosophy that allows appropriate protective action even if a single failure occurs. 1.2.2.4.6 Nuclear System Pressure Relief System -

1.2.2.4.2 Neutron Monitoring System A pressure relief system consisting of Those portions of the neutron monitoring safety / relief valv?s mounted on the main system (NMS) that are part of the RPS qualify as steamlines is pros:Jed to prevent excessive a nuclear safety system. The startup range pressure inside the nuclear syst im as a result neutron monitor (SRNM) and the average power of operational transients or accidents, range monitors (APRM) which monitor neutron flux via incore detectors provide scram logic inputs 1.2.2.4.7 Reactor Core Isolation Cooling to the reactor protection system (RPS) to (RCIC) System initiate a sesam in time to prevent excessive fuel clad damage as a result of over power The RCIC sptem proddes makeup water to the transients. The SRNM system also generates a reactor vessel when the vessel is isolated and simulated thermal power signal. Both upscale is also part of the emergency core cooling neutron flux and upscale simulated thermal power network (Subsection 1.2.2.4.8). The RCIC system  !

are conditions which provide scram logic signals. uses a steam driven turbine pump unit and Amendment 1 1.27 ee - --

-n ,we

i l

l ABM 23A6100AC l Standard Plant RIV A operates automatically in time and witb minimum flow bypass line to the suppression pool sufficient coolant flow to maintain adequate and a cooling water supply line to auxiliary g I water level in the reactor vessel for events equipment. W l defined in Section 5.4. '

Follovcing a reactor scram, steam generation 1.2.2.4.8 Emergency Core Cooling Systems (ECCS) in the reactor core continues at a reduced rate due to the core fission product decay heat. The In the event of a breach in the reactor turbine bypass system diverts the steam to the coolant pressure boundary that results in a loss main condenser, and the feedwater system of reactor coolant, three independent divisions supplies the makeup water required to maintain of ECCS are provided to maintain fuel cladding reactor vessel inventory, below the temperature limit as defined by 10CFR50.46 . Each division contains one high In the event the reactor vessel is isolated, pressure and one low pressure inventory makeup and the feedwater supply is unavailable, relief system. The systems are: valves are provided to automatically (or remote manually) maintain vessel pressure within 1.2.2.4.8.1 liigh Pressure desirable limits. The water level in the reactor vessel drops due to continued steam 1.2.2.4.8.1.1 liigh. Pressure Core Flooder (IIPCF) generaiion by decay heat. Upon reaching a System predetermined low level, the RCIC system is initiated automatically. The turbine driven HPCF are provided in two divisions to maintain pump supplies water from the suppression pool or an adequate coolant inventory inside the reactor from the CSP to the reactor vessel. The turbine vessel to limit fuel cladding temperatures in the is driven with a portion of the decay heat steam event of breaks in the reactor coolant pressure from the reactor vessel, and exhausts to the boundary. The systems are initiated by either suppression pool, high pressure in the drywell or low water level in the vessel. They operate independently of all in the event there is a LOCA, the RCIC system other systems over the entire range of system in conjunction with the two HPCF systems, is g

operating pressures. The HPCF system pump motors designed to pump water into the vessel from are powered by a diesel generator if auxiliary approximately 150 >sig to full operating power is not available. The systems may also be pressure. This combination of systems provides used as a backup for the RCIC system, adequate core cooling until vessel pressure drops to the point at which the low pressure 1.2.2.4.8.1.2 RCIC Description flooder loop (LPFL) subsystems of the RHR can be placed in operation.

One division contains the RCIC system which consists of a steam driven turbine which drives a During RCIC operation, the wetwell suppres-pump assembly and the turbine and pump sion pool acts as the heat sink for steam gene-accessories. The system also includes piping, rated by reactor decay heat. This results in a valves, and instrumentation necessary to rise in pool water temperature. Heat exchangers implement several flow paths. The RCIC steam in the residual heat removal (RHR) system are supply hne branches off one of the main steam used to maintain pool water ter.2perature within lines (leaving the reactor pressure vessel) and acceptable limits by cooling the pool water goes to the RCIC turbine with drainage provision directly.

to the main condenser. The turbine exhausts to the suppression pool with vacuum breaking 1.2.2.4.8.2 Automatic Depressurization Sptem protection. Makeup w:ter is supplied from the (ADS) condensate storage pool (CSP) or the suppression pool with the preferred source being the CSP. The ADS rapidly reduces reactor vessel RCIC pump discharge lines include the main pressure in a loss-of coolant accident, enabling discharge line to the feedwater line, a the low pressure RhP. to deliver cooling water to test-return line to tne suppression pool, a the reactor vessel.

Amendment 1 1.2-8

MM 23A6100AC Standard Plant REV.A The ADS uses some of the safety relief valves (4) the reactor building which is structurally that are part of the nuclear system pressure integrated with the concrete primary con-n)

(. relief system. The safety relief valves used for tainment structure.

ADS are set to open on detection of appropriate low reactor water level and high pressure A secondary containment which surrounds the signals. The ADS will not be activated unless primary containment permits monitoring and either high pressure core flooder system (HPCF) treating all potential radioactive leakage from or RHR/ Low Pressure Flooding Loop pumps are the primary containment. Treatment consi.ts of operating. This is to ensure that adequate HEPA and activated charcoal filtration.

coolant will be available to maintain reactor water level after the dentessurization. 1.2.2.4.9.2 Heat Removal 1.2.2.4.83 Low Pressure Flooder Loop (LPFL) 'Ihe containment RHR system is summarized in Subsection 1.2.2.4.14.

Low pressure flooding is an operating rnode of each RHR System, but is discussed here because 1.2.2.4.93 Environmental Systems the LPFL mode acts in conjunction with other injection systems. LPFL uses the pump loops of The containment heating and ventilating the RHR to inject cooling water into the pressure systems are described in Subsection 1.2.2.8.10.

vessel. LPFL operation provides the capability of core flooding at low vessel pressure following 1.2.2.4.9.4 Wetwell/Drpell Spray a LOCA in time to maintain the fuel cladding below the prescribed temperature limit. A spray system is provided for wetwell/

drywell cooling in the suppression chamber and 1.2.2.4.9 Containment drywell air space. The wetwell/drywell spray can be initiated manually if a high-containment-1.2.2.4.9.1 Functional Design pressure signal is received. Each subsystem is supplied from a separate redundant RHR The primary containment design for this plant subsystem.

incorporates the drywell/pr:ssure suppression feature of previous BWR containment designs into 1.2.2.4.9.5 Flammability Control a dry-containment type structure. In fulfilling its design basis as a fission product barrier, An atmo:pheric control system is designed to the primary containment is a low-leakage establish and maintain an inert atmosphere structure even at the increased pressures that within the primary containment during all plant could follow a main steamline rupture or a fluid operating modes except during plant shutdown for system line break. refueling or maintenance.

The main features of the containment design 1.2.2.4.10 Containment and Reactor Vesse! 1 include: Isolat!aa Control I (1) the drywell, a cylindrical steel lined For the ABWR this function is performed by ,

reinforced concrete structure surrounding the the leak detection and isolation system '

reactor pressure vessel (RPV); (S ubsection 1.2.2.3.6).

(2) a suppression pool filled with water which 1.2.2.4.11 Main Steamline Isolation Valves senes as a heat sink during normal operation and accident conditions; All pipelines that both penetrate the containment and offer a potential release path (3) the air space above the suppression pool; and for radioactive material are provided with 1

O  !

l Amendment 1 1.2-9 1

l l

MWR ux6iooac Standard Plant _

REV.A redundant isolation capabilities. Automatic then discharged back to the suppression pool, to isolation valves are provided in each main steamline. Each is powered by both steam the drywell sprey header, to the suppression chamber spray header, or to the RPV.

g pressure and spring force. These valves fulfill the following objectives: 1.2.2.4.15 Process Radiation Monitoring System (1) prevent excee.,ive damage to the fuel barrier (See Subsection 1.2.2.9) by tiraiting the loss of reactor coolant from the reactor vessel resulting from either a 1.2.2.4.16 Standt,y Gas Tn atment System major leak from the steam piping outside the containment or a malfunction of the pressure The standby gas treatment system (SGTS) control system resulting in excessive steam minimizes exfiltratirn of contaminated air from flow from the reactor vessel; the secondary containment to the environment a following an accident or abnormal condition (2) limit the release of radioactive materials by which could result in abnormally high airborne isolating the reactor coolant pressure radiation in the reactor building. Because the boundary in case of the detection of high fuel storage area is also in the secondary steam line radiation, containment it also can be exhausted to the SGTS.

1.2.2.4.12 Main Stearaline Flow Restrictors All components of the SGTS are operable A venturi type flow restrictor is installed in during loss of offsite power.

the reactor pressure vessel nozzle for each steamline. These restrictors limit the flow of 1.2.2.4.17 Reactor Ilullding isolation Control steam Dom the reactor vessel before the main System steamline isolation valves are closed in case of a main steamline break outside the containment The reactor building isolation control (RBIC) and also provide high flow isolation signals. sys:em cutomatically ialtiates closure of isola-tion dampers in all ventilation ducts which are 1.2.2.4.13 Radiation Monitoring and Control potential leakage paths for radioactive material to the environs. The action is taken upon indi.

This system is part of radiation monitoring cation of a potential breach in the nuclear sys-and control system. See Subsection 1.2.2.9. tem process barrier or a fuel handling accident.

1.2.2.4.14 Residual lleat Removal (RilR) System 1.2.2.4.18 Standby AC Power Supply (Containment Cooling)

Standby ac power is supplied by three diesel The RilR is placed in operation to: (1) limit generators. Each Class IE division is supplied the temperature of the water in the suppression by a separate diesel generator. There are no pool and the atmospheres in the drywell and sup- provisions for transferring Class 1E buses pression chamber following a design basis LOCA; hetween standby ac power supplies or supplying (2) control the pool temperature during normal more than one engineered safety feature (ESP) operation of the safety / relief valves and the imm one diesel rencrator. This one to one RCIC system; and (3) reduce the pool temperature rebtionship between diesel generator and ESF following ua isolation transient. In the con- dividon ensures that a failure of one diesel tainment cooling mode of operation, the RilR main generetor can affect only one ESF division. The system pumps take suction from the suppression diesel generators are housed in the reactor pool and pump the water through the RilR heat building which is a Seismic Category I j exchangers where cooling takes place by trans. structure, to comply with applicable NRC and ferring beat to the service water. The fluid is IEEE design guides and criteria.

O Amendment 1 1.2-10

ABM ux6iooAc Standard Plant any A 1.2.2.4.19 DC Power Supply (d) instrument power system; O The vi t 8 s re=, i aere de t ci ss 1e 125 volt de power systems.

(e) eninterr#etibie newer srstem

, (f) unit auxiliary d c power sptem, and 1.2.2.4.20 Standby Liquid Control System (g) unit Class IE d c power system.

The standby liquid control system (SLCS) provides an alternate method to bring the nuclear Each of these systems is described briefly in fission reaction to suberiticality and to the following subsections.

maintain suberiticality as the reactor cools.

The system makes possible an orderly and safe 1.2.2.5.1.1 Unit Auxillary A C Power System shutdown in the event that not en'ough control rods can be inserted into the reactor core to The unit auxiliary a c power system supplies accomplish shutdown in the normal manner. The power to unit loads that are non safety related syste m is sized to counteract the positise and uses the main generator as the normal power reactivity effect from rated power to the cold source with the reserve auxiliary transformers shutdown condition. as a backup source. The unit auxiliary transformer steps down the a c power to the 1.2.2.4.21 Remote Shutdown System 6900 V and 4160-V station bus voltage.

In the event that the control room becomes la.2.5.1.2 Unit Class 1E A C Power System inaccessible, the reactor can be brought from power range operation to cold shutdown conditions The unit Class 1E a c power system supplies by the use of controls and equipment that are power to the unit Class IE loads. The offsite available outside the control room, power sources converge at the system. The system includes diesel generators that serve as 1.2.2.4.22 Condensate Storage Facilities and standby power sources, independent of any onsite

\ Distribution System or ( ffsite source. Therefo,c, the system has three sources. Furthermore, the system is The condensate storage tank receives demi- divided into three divisions, each with its own neralized water form the condensate water makeup independent distribution network, diesel system and may also receive low conductivity generator, and redundant load group, A fourth water from the condensate return of the primary division for the safety logic and conRol system loop, from the radwaste disposal system and the bus receives power from the division 1 source.

condensate system in the turbine building.

1.2.2.5.13 Safety System Protection System 1.2.2.5. Electrical Sp tems and instrumentation Power System -

and Control Four divisions of the safety system logic and 1.2.2.5.1 Electrical Power Systems control (SSLC) power system provide a Class IE source of 120-VAC single phase control power.

The following systems provide electrical power The primary power source for the SSLC power to station auxiliaries: system is the Class IE AC power systun, with a normal seeking automatic throw over switch to (a) unit auxiliary a-c power system; the Class 1E DC power system for back up.

(b) unit Class IE a-c power system; 1.2.2.5.1.4 Instrument Power Sptem (c) safety system logic and control system power The instrument power system supplies 120-VAC system; single-phase power to instrument and control -

I

,G I v

Amendment i 1.2 11

ABMt 3346iooac Standard Plant niw. A loads which do not require an uninterruptible flow rate through the core and thereby changes power source. the core power level. The system can automati- g cally adjust the reactor power output to the W 1.2.2.5.1.5 Uninterruptible Power System load demand. The solid state adjustable speed drives (ASD) provide variable voltage, variable The uninterruptible power system (UPS) frequency electrical power to the RIP motors, supplies regulated 120.VAC single phase power to in response to plant needs, the recirculation non Class 1E instrument and controlloads which flow control system adjusts the ASD pov er supply require an uninterruptible source of power. The output to vary RIP speed, core flow, and core power sources for the UPS are similar to those power.

for the SSLC, but are non Class 1E.

1.2.2.5.2.3 Neutron blonitoring System 1.2.2.5.1.6 Unit Auxiliary DC Power System The neutron monitoring system (NhtS) is a The unit auxiliary DC power system supplies system of in-core neutron detectors and power to unit DC loads that are nonsafety- out of-core electronic monitoring equipment.

related. The system consists of two battery The system provides indication of neutron flux, chargers, two batteries, two motor control which can be correlated to thermal power level centers, and two distribution panels. for the entire range of flux conditions that can exist in the core. There are fixed in core 1.2.2.5.1.7 Unit Class 1E DC Power System sensors which provide flux level indications during reactor startup and low power operation.

The unit Class 1E DC power system supplies 125 The startup range neutron monitors (SRNht) and VDC power to the unit Class 1E loads. Battery average power range monitors (APRht) allow chargers are the primary power sources. The assessment of local and overall flux conditions system, which includes storage batteries that during power range operation. Tnc automatic serve as standby power sources, is divided into traversing in core probe (ATIP) System prov%

four divisions, each with its own independent a means to calibrate the SRNht. The NNIS pros.les distribution network, battery, battery charger, inputs to the rod control and information system h

and redundant load group. to initiate rod blocks if preset flux limits or period limits for rod block are exce:ded as wt11 1.2.2.5.2 Nuclear System Process Control and as inputs to the RPS if other limits for scra:t Instrumentation are enceded.

1.2.2.5.2.1 Rod Control and Information System 1.2.2.5.2..t Refueling interh>cks The rod control and information system (RCIS) A system of intedocks that r e st ric. s provides the means by which control rods are movement of refueling equipmem ad control rods {

positioned from the control room for power when the reactor is in the refueling and starta I control. The system operates the rod drive modes is provided to prevent and inadvertent l motors to change control rod position. For criticality during refueling operation. The I operation in the normal gang movement mode, one interlocks back up procedural controls that have gang of control rods can be manipulated at a the same objective. The interlocks affect the time. The system includes the logic that refueling platform, refueling platform hoists, restricts control rod movement (rod block) under fuel grapple, and control rods.

certain conditions as a backup to procedural controts. 1.2.2.5.2.5 Reactor Vessel Instrumentation 1.2.2.5.2.2 Recirculation Flow Control System In addition to instrumentation for the nuclear safety systems and engineered safety During normal power operation, the speed of featurcs, instrumentation is provided to monitor the reactor internal purnps is adjusted to control and transmit information that can be used to flow. Adjusting RIP speed changes the coolant assess conditions existing inside the reactor g

Amendment 1 1.2 12

ABM unsioaac Standard Plant REV.A vessel and the physical condition of the vessel The reactor building crane handles the spent .

itself. This instrumentation monitors reactor fuel cask from the transport device to the spent p)

(. vessel pressure, water level, coolant fuel loading pit. The fuel handling platform temperature, reactor core differential pressure, transfers the fuel assemblies between this coolant flow rates, and reactor vessel head inner storage area, the reactor core, and the spent seal ring leakage. fuel shipping cask _ New fuel bundles are handled by the reactor building crane auxiliary 1.2.23.2.6 Process Computer System book. These bundles are stored in the new fuel vault on the reactor refueling floor. New fuel -

On line process computers are provided to is transferred from the vault to the spent fuel monitor and log process variables and make pool with the reactor building crane auxiliary certain analytical computations, book.

The turbine generator speed. load controls can The handling of the reactor head, removable initiate rapid closure of the turbine control internals, and drywell head during refueling is valves (rapid opening of the turbine bypass accomplished using the reactor building crane.

valves) to prevent turbine overspeed on loss of ,

the generator electric load. 1.2.2.8 CoolingWater and Auxillary System 1.2.2.6 Feedwater Control System 1.2.2.8.1 Reactor Building Cooling Water System The feedwater control system automatically The reactor building cooling water system contrcis the flow of feedwater into the reactor (RBCWS) provides cooling water to certain pressure vessel to maintain the water within the designated equipment located in the reactor vessel at predetermined levels. A fault- building. Adequate capacity and redundancy is tolerant tripricated, digital controller using provided in heat exchangers and pumps to ensure conventional three. dimensional control scheme is performance of the cooling system under all used to accomplish this function. modes of plant operation. In the event of loss of offsite power, emergency power for the system 1.2.2.7 Fuel Handling and Storage Systems is available from the onsite emergency diesel generators. The closed loop design provides a 1.2.2.7.1 New and Spent Fuel Storage barrier between radioactive systems and the service water discharged to the environment.

New and spent fuel storage racks are designed Heat is removed from the closed loop by the to prevent inadvertent criticality and load cooling water system. Radiation monitors are buckling. Sufficient cooling and shielding are provided to detect contaminated leakage into the provided to prevent excessive pool heatup and closed systems, personnel exposure, respectively. The design of the fuel pool provides for corrosion resistance, 1.2.2.8.2 Fuel Pool Cooling and Cleanup System adherence to Seismic Category I requirements, and (FPCS) prevention of kerr from reaching 0.95 under dry or flooded conditions. This subject is further This system maintains acceptable levels of discussed in Section 9.1. temperature and clarity and minimizes )

radioactivity levels of the water in the spent 1.2.2.7.2 Fuel liandling System fuel storage, cask pools and reactor well exchanger / separator storage pools on top of the The fuel. handling equipment includes a 150 ton containment. The FPCS also maintains the reactor building crane, fuel. handling platform, temperature and water level in the service pool l fuel inspection stand, fuel preparation machine, and equipment pool. The system includes two jib hoist, and other related tools for reactor heat exchangers, each capable or removing the j servicing. All equipment conforms to applicable decay heat generated from an average discharge i cod s and standards. of spent fuel, and two filter /demineralizers, O

Amendment 1 1.2 13

ABWR ux6ioorc Standard Plant anv.A cach unit having the capacity to process the 1.2.2.8.6 Plant Equipment and Floor Drainage g system flow or greater to maintain the desired w purity level. The equipment and floor drainage systems are designed to collect liquid waste throughout the 1.2.2.8.3 Makeup Water (Condensate) System plant and discharge the radioactive waste to the radwaste system for processing. Separate drain.

The makeup water system maintains the required age facilities are provided for nonradioactive capacity and flow of the condensate for the RCIC waste.

and ilPCF systems and maintains the required level in the condenser botwell. The system also stores The drainage system is also used to detect and transfers water during refueling and cask abnormal leakage in safety related equipment storage pool water during fuel shipping cask rooms and the fuel transfer area.

loading, receives and stores the process effluent from the liquid radwaste system, provides makeup 1.2.2.8.7 Ser31ce and Instrument Air Sptems to other plant systems where required, and provides condensate to the control rod drive The service and instrument air systems (CRD) hydraulic system. provide a continuous suppiy of compressed air of suitable quality and pressure for instrument The system consists of a condensate storage control and general plant use. The service air pool, three condensate transfer pumps, and the compressor and the instrument air compressor necessary controls and instrumentation. discharge into their respective air receivers.

The air is then distributed throughout the 1.2.2.8.4 Plant Chilled. Water Systems plant. Instrument air is additionally filtered and dried prior to distribution throughout the The plant chilled water systems consists of plant.

the 11VAC Normal Cooling Water System, and the 11VAC Emergency Cooling Water System. The IIVAC normal cooling water system provides chilled 1.2.2.8.8 liigh Pressure Nitrogen (11 PIN) Gas Supply h

water to the air supply cooling coils of the reactor building, to the heating / cooling coils in Nitrogen gas is normally supplied by the the drywell, and the control building electrical Atmospheric Control System to meet the require- '

equipment rmm. The HVAC emergency coolingwater ment on the main steam system safety relief l system provide; chilled water to the cooling valve automatic depressurization and relief '

coils in the controf building essential function accumulators, the main steam isolation i electrical equipment room, the main control room valves, instruments and pneumatic valves using and the diesel generator electrical equipment nitrogen in the reactor building. When this areas. The safety-related chilled water systems supply of pressurized nitrogen is not available, are designed to meet the requirements of the llPIN automatically maintains nitrogen Criterion 19 of 10CFR50. pressure to this equipment. The liPIN system consists of high pressure nitrogen storage 1.2.2.8.5 Process Sampling Systems bottles with piping, valves, instruments, controls and a control panel.

The process sampling system is furnished to provide p-icess information that is required to 1.2.2.8.9 Diesel Generator Fuel Oll Storage monitor plant and equipment performance and and Transfer Sptem changes to operating parameters. Representative liquid and gas samples are taken automatically The major components of this system are the and/or manually during normal plant operation for fuel oil storage tanks, pumps, and day tanks, laboratory or on.line analyses. Each diesel generator has its own individual O

Amendment 1 1.2-14

ABM 23461ooac Standard Plant ' REV,A supply components. Each storage tank is designed (6) the power block exhaust system to maintain to supply the diesel needs during the post LOCA the refueling floor at a negative pressure O period and each day tank has the capacity for two with respect to the outside atmosphere to hours of diesel generator operation. Each fuel- prevent the potential release of airborne oil pump is controlled automatically by day tank contamination; level and feeds its day tank from the storage tank. Additional fuel oil pump supply fuel to (7) the diesel generator area air exhaust system each diesel fuel manifold from the day tank. to provide cooling during operation of the diesel generators. A tempered air supply 1.2.2.8.10 Heatlog, Ventilating,and Air system controls the thermal emironment when Conditionlog (Environmental) System the diesel generators are not operating; and The plant environmental control systems (8) coolers in the steam tunnel and ECCS rooms control temperature, pressure, h'imidity, and to remove heat generated during operation of airborne contamination to ensure th: integrity of the equipment in these rooms.

plant equipment, provide acceptable working conditions for plant personnel, and limit offsite 1.2.2.8.11 Lighting Systems releases of airborne contamina.ats.

The design basis for the lighting facilities The following environmental systems are is the standard of the Illuminating Engineering provided: Society. Special attention is given to areas where proper lighting is imperative during (1) the control room air conditioning system normal and emergency operations. The system consisting of supply, recirculation / exhaust design precludes the use of mercury vapor and makeup air cleanup units to ensure the fixtures in the containment and the fuel-habitabilit. of the control room under normal handling areas. The normal lighting systems are and abnormal conditions of plant operation; fed from the unit auxiliary transformers.

O (2) the reactor building secondary containment e m erse ex re - er is se r riied 8x e=8 i#eerca s>re'x buses backed up by diesel generators. Normal HVAC system maintains a negative pressure in operation and regular simulated offsite power the secondary containment under normal and loss tests verify system integrity.

abnormal operating conditions thereby isolating the environs from potential leak 1.2.2.8.12 Fire Protection System sources. This system removes heat generated during normal plant operation, shutdown, and The fire protection system is designed to refueling periods; provide an adequate supply of water or chemicals to points throughout the plant where 're pro-(3) the drywell cooling system to remove heat tection is required. Diversified fire alarm and from the drywell generated during normal fire suppression types are selected to suit the plant operating including reactor scrams, hot particular areas or hazards being protected, standby, shutdown, and refueling periods; Chemical fire fighting systems are also provided as additions to or in lieu of the water fire (4) the power block pressure control supply and fighting systems. Appropriate instrumentation exhaust system to distribute air so that a and controls are provided for the proper negative pressure is maintained in the operation of the fire detection, annunciation emergency core cooling equipment rooms, and fire fighting systems, thereby isolating the potential airborne contamination in these rooms; 1.2.2.9 Radiation 51onitoring and Control (5) the electrical equipment supply and exhaust 1.2.2.9.1 Process Radiation hionitoring system to pressurize the electrical rooms allowing exfiltration of air to the battery Process radiation monitoring systems monitor rooms for exhaust to the outside atmosphere; and control radioactivity in process and Amendment 1 1.2 15

ABM n46ioaac 2

Standard Plant anv.A effluent streams and activate appropriate alarms and controls.

A process radiation monitoring system indicates and records radiation levels associated with selected plant process streams and effluent paths leading to the environment. All effluents from the plant which are potentially radioactive are monitored.

Process radiation monitoring is also discussed in Chapters 7,9, and 11, 1.2.2.9.2 Area Radiation hfonitors Area radiation monitoring systems alert occupants and the control room personnel of excessive gamma radiation levels at selected locations within the plant.

1.2.2.93 Dust Radiation hfonitors The Dust Radiation hionitoring System is designed to collect dust in the air, to rnonitor the concentration level of radioactive dust and iodine and to provide alarms if set points are exceeded.

1.2.2.9.4 Containment Atmospheric 51onitors The containment atmospheric monitoring system (CAhtS) measures alarms and records radiation levels and the oxygen concentration in the primary containment under post-accident conditians. It is automatically put in service upon detection of LOCA conditions. '

1.2.2.10 Shiciding Shielding is provided throughout the plant, as required to maintain radiation levels to operating personnel and to the general public within the applicable limits set forth in 10CFR20 and 10CFR100. It is also designed to protect certain plant components from radiation exposure resulting in unacceptable alterations of m rial properties or activation.

O Amendment 1 1.2-16 e

ABWR 23461ooac Standard Plant nriv. A i

l 1

TO BE PROVIDED BY DECEMBER 1988 O

1 l

\

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Figure 1.2-1 Site Plan Amendment ] 1.2-17 l

l

ABWR 2 mime Standard Plant RiiV A O

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TO BE PROVIDED BY DECEMBER 1988 O

Figure 1.2 2 Reactor, Reactor BuiMing, Primary Containment Arrangement Plan, Section 0/18d 9

Amendment 1 1.2 18

l ABWR 2mimac Standard Plant RIN. A l 1

l l

O i i

TO BE PROVIDED BY DECEMBER 1988 l

Figure 1.2 3 Reactor, Reactor Building, Primary Containment Arrangement Plan, Section 270/90 O

Amendment 1 1.2 19

ABWR ma c Standard Plant airv ^

O TO 13E PROVIDED BY DECEMf3ER 1988 O

Figure 1.2-4 Reactor, Reactor Building, Primary Containment Arrangement Plan at Floor El(-) 0200 mm O

Amendment 1 1.2 20

ABWR 23461*ac Standard Plant Rf!V. A O  !

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1 I

i 1

TO BE PROVIDED BY DECEMBER 1988 O

l I

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Figure 1.2 5 Reactor, Reactor Building, Primary Containment Atrangement Plan at Hoor El(-) 6700 mm Amendment 1 1.2 21

ABWR ^

m6imac S* . dard Plant -

lutv. A O

I i

I I

I l

I l

TO BE PROVIDED BY DECEMI3ER 1988 l l

9l' l

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Figure 1.2-6 Reactor, Reactor Building, Arrangement Plan at Floor El(-) 6700..im )

Amendment 1 1.2-22 l

1 i

ABWR 23461mac Standard Plant nrw. A O

TO BE PROVIDED BY DECEMBER 1988 O

)

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Figure 1.2-7 Reactor, Floor Reactor Building, Primary Containment Plan at Ground Roor El 7300 mm Amendment 1 1.2 23

ABWR 23A61MAC Standard Plant is,. 3 i

e l

l i

TO BE PROVIDED BY DECEMBER 1988 9'

l l

ngure u.s neacer, acactor Building, Primary Containment Plan at E19X0 mm Amendment 1 1.2 24 f

l

ABM ua6icoac Standard Plant REV A O

TO BE PROVIDED BY DECEMBER 1988 O

Figure 1.2-9 Reactor, Reactor Building, Primary Containment Plan at El 115(M mm Amendment 1 1.2 25

ABWR 2mimac Standard Plant nirv. A O

TO BE PROVIDED BY DECEMBER 1988 O

l Figure 1.210 Reactor, Reactor Building, Primary Containment Plan at El 125N mm O

Amendment 1 1226

ABWR nisiooac Standard Plant RiiV. A O

TO BE PROVIDED BY DECEMBER 1988 O

Figure 1.2-11 Reactor, Reactor Building, Primary Containment Plan at El 16500 mm Amendment 1 1227

ABWR 2n6muc Standard Plant RI?V. A O

TO BE PROVIDED BY DECEMBER 19SS O

Figure 1.212 Reactor, Reactor Building Plan at El 18700 mm Amendment 1 1.2 28

i 23A6100AC Standard Plant REV.A O i I

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1 I

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TO BE PROVIDED BY DECEMBER 19SS

'l O l i

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l Figure 1.2-13 Reactor, Reactor Building Plan at El 26700 mm Amendment 1 1.2-29

MM 21%100AC ElBDdArdfjaDt Rfi\M O

TO BE PROVIDED BY DECEMBER 1988 O

l l

Figure 1.214 Control and Senice Building Plot Plan Amendment 1 1.2 30 l

ABMR 23A6100AC Standard Plant niiv. A O

TO BE PROVIDED BY DECEMBER 1988 O

l l

Figure 1.2-15 Control and Senice Building Typical Section Amendment 1 1.2 31

l ABWR 2**c sua@xunant RIiv A  !

O 1

1 TO BE PROVIDED BY DECEMBER 1988 l

1 l

Oi 1

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i Figure 1.216 Control and Scaice Building Piping Interface at El (-) 3500 mm O

Amendment I l

ABWR 2u6ime Standard Plant RE\M i

O 1

l 1

I i

I

)

TO BE PROVIDED BY DECEMBER 19S8 O

1 Figure 1.217 Control and Senice Building Clean Area Access at El(-) 0200 mm O

Amendment 1 g 2 33

ABWR DA61NAC Standard Plant ruiv. A 9

TO BE PROVIDED BY DECEMBER 1988 O

Figure 1.2-18 Contrd and Senice Building - E/I Interface at El 4300 mm g

Amends. vet 1 1.2-14 i

MM "^',I^f Standard Plant p O

1 l

l TO BE PROVIDED BY DECEMBER 1988 O

l l

Figure 1.2-19 Control and Senice Building - Main Access and Controlled Area Access at El 7300 mm Amendment 1 1.2 15

ABWR 2miwAc ,

Standard Plant ruiv. A O

TO BE PROVIDED BY DECEMBER 1988 O

l l

Figure 1.2 20 Control and Senice Building - Steam Tunnel at El 12700 mm Amend.nent 1 1.2-36

ABWR 2neiwac Standard Plant anv. A O

l l

TO BE PROVIDED BY DECEMBER 1988 O

l l

l Figure 1.2-21 Control and Senice Building - T/C Rooms at El 17000 mm Amendment 1 1.2 37

ABM 23A61%AC l Standard Plant nitv. A O

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1 TO BE PROVIDED BY DECEMBER 1988 1

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Figt te 1.2-22 Control and Senice Building - Control Room at El 23100 mm O

Amendment 1 1.2-38

23A6100AC Standard Plant Rev. A SECTION 13 CONTENTS Section Iille Eage 1.$.1 Nuclear Steam Sunoly System Deslen Characteristics 13 1

' 3.2 Enntneered Safety Features Deslan Characteristics 13-1 133 Containment Deslan Characteristics 13-1 13A Structural Desien Characteristics . 1.3-1 13.5 Instrumentation and Electrical Systems Iksien Characteristics 13-1 TABLES Table I!11g Page 13-1 Comparison of Nuclear Steam Supply Steam e Design Characteristics 13-2

(

13-2 Comparison of Engineered Safety Features Design Characteristics 13-12 13-3 Comparison of Containment Design Characteristics 13-15 13-4 Comparison of Structural Design Characteristics 13 19 l

l O

13=ii Amendmert 1

ABWR ux61ooxc Standard Plant REV.A 1.3 COMPARISONTAI1LES O

L> This section highlights the principal design l

features of the plant and compares its major l features with those of other boiling water j reactor facilities. The design of this facility i is based on pr >ven technology obtained during the development, design, construction, and  ;

operation of boiling water reactors of similar I types. The data, performance, characteristics, and other information presented here represent a i current, firm design.

1.3.1 Nuclear Steam Supply System Design Characteristics Table 1.31 summarizes the design and operating characteristics for the nuclear steam supply systems. Parameters related to power output for a sing!c plant unless otherwise noted.

1.3.2 Engineered Safety Features Design Characteristics Table 1.3-2 compares the engineered safety features design characteristics.

O 2 3 3 contaia-eat oesiaa Characteristics Table 1.3 3 compares the containment design characteristics.

1.3.4 Structural Design Characteristics Table 1.3 4 compares the structural design characteristics.

1.3.5 Instrumentation and Electrical Systems Design Characteristics Table 7.1-2 compares the instrumentation and electrical systems design characteristics.

Amendment 1 13-1

MkN 13A6100AC Standard Plant REV.A TABLE 1.31 Il v COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS (Parameters are related to rated power out for a single plant unless otherwise noted)

His Plant GESSAR NMP2 Grand Gulf ABWR BWR/6 BWR/S BWR/6 Design 278-872 238-748 251 764 251-800 Hermal and livdraulle (See Section 4.4)

Rated power 3,926 3,579 3,323 3,833 (MWt)

Design power 4,005 3,729 3,463 4,025 (MWt) (ECCS design basis)

Steam flow 16.843 15.40 14.263 16.491 rate, Mlb/hr at 4200F (FW Temp)

Core coolant 115.1 IN.0 108.5 112.5

( flow rate (Mlb/hr)

Feedwater flow 15.807 15372 14.5M 16.455 rate (Mlb/hr)

System pressure, 1,N0 1,NO 1,020 1,040 nominal in steam dome (psia)

Average power 50.6 54.1 49.15 54.1 density (kW/ft)

Maximum linear 13.4 13.4 13.4 13.4 heat generation rate (kW/ft)

Average linear 5.6 5.9 5.40 5.93 beat generation rate (kW/ft)

Maximum heat 369,600 361,600 354,255 361,600 flux (Btu /hr/ft2)

O Amendment 1 1.12

ABM 23461ooac Standard Plant any. A TABLE 1.3 1 COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS (Continued)

His Plant GESSAR NMP2 Grand Gulf AHWR HWR/6 HWR/S BWR/6 Deslen 278-872 238 748 251 764 251-800 Dermal and livdraulle (Continued)

Average Heat (later) 159,500 144,032 160,300 flux 2

(Btu /hr-ft )

Maximum UO2 (later) 3,435 3,325 3,435 temperature (OF)

Average (later) 2,185 2,130 2,185 volumeterie fuel temperature (OF)

Average 566 565 566 565 cladding surface temperature ( F)

Minimum critical power ratio 1.16 1.20 1.24 1.20 h

(MCPR)

Coolant 527.7 527.6 527.5 527.9 enthalpy at core inlet (Blu/lb)

Core maximum 75 79 76.2 76 voids within assemblics Core average 14.5 14.7 13.1 14.6 exit quality

(% steam)

Feedwater 420 420 420 420 temperature (OF)

G Amendment 1 1.33

ABM 23461oo^c Standard Plant nrv. A TABLE 1.31 COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS (Continued)

His Plant GESSAR NMP2 Grand Gulf ABWR BWR/6 BWR/5 BWR/6 Design 278-872 238 748 251 764 251-800 Dermal and Ilydraulle (Continued)

Design power peaking factor Maximum relative 1.40 1.40 1.40 1.40 assemble power local peaking 1.25 1.13 1.24 1.13 factor Axial peaking 1.40 1.40 1.40 1.40 factor f Total peaking 2.43 2.26 2.43 2.26

( factor Nuclear (first core)

(Section 43)

Water /UO2 2.88 2.70 2.55 2.70 volume ratio (cold)

Reactidty with < 0.99 < 0.99 < 0.99 < 0.99 strongest control rod out (ke rd Dynamic void 5.40c@ -7.16 -8.57 -7.14 l coefficient 120% rated I (c/%) at core output average voids 39.2 40.95 40.54 4131

(%)(EOC rated output)

Fuel temperature -0360 -0.412 0.419 -03%

doppler coeffi.

cient (c/0C)

(EOC rated output)

O Amendment 1 1.3-4

ABM MA61NAC Standard Plant niiv. A i

TABLE 1.31 COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS (Continued)

This Plant GESSAR NMP2 Grand Gulf ABWR BWR/6 BWR/S BWR/6 Design 278-872 238-748 2517M 251-800 Nuclear (first core) (Continued)

Initial average 2.22 1.90 1.90 1.70 U-235 enrichment (%)

Initial cycle 9,950 9,138 9,200 7,500 exposure (mwd /short ton)

Core Mechanical Fuel assembly (Section 4.2)

Number of fuel 872 748 560 800 assemblies Fuel rod array 8x8 8x8 8x8 8x8 O

Overall length 176 176 176 176 (inches)

Weight of UO2 451 456 466 458 per assembly (Ib) (pellet type)

'Veight of fuel 675 697 698 697 assembly (lb)

(includes channel)

Fuel Rods (Section 4.2)

Number of fuel 62 62 63 62 rods per assembly Outside diameter 0.483 0.483 0.493 0.483 (in.)

Cladding 0.032 0.032 0.032 0.032 0.032 thickness (in.)

G Amendment 1 1.3-5

ABWR m6mc Standard Plant nrw.A TABLE 1.31

(,

d COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS (Continued)

This Plant GE3SAR NMP2 Grand Gulf ABWR BWR/6 HWR/S HWR/6 Design 278-872 238 748 251-764 251 800 Core Mechanical Diameteral gap, 0.009 0.009 0.009 0.009 pellet to-cladding (in.)

Length of gas 13.23 9.48 14 9.48 plenum (in.)

Cladding Zircaby-2 Zircaloy-2 Zircaloy-2 Zircaloy 2 material

• Fuel Pellets (Section 4.2)

Material UO2 UO2 UO2 UO2 O oesitxc*er theoretical)

• 5 95 95 95 0.411 0.410 0.416 0.410  !

Diameter (in.)

length (in.) 0.410 0.410 0.420 0.410 i

Fuel Channel 1 (Section 4.2)

Thickness 0.100 0.120 0.100 0.120 (in.) )

Cross section 5.48 x 5.48 5.45 x 5.45 5.48 x 5.48 5.45 x 5.45 dimensions (in.)

Material Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4 Core assembly (Section 4.2)

Fuel weight 392,990 341,640 265,551 365,693 as UO2 (lb)

• Free standingloaded tubes Amendment 1 1.3-6

ABWR 2mimac Standard Plant - niiv A TAlli,E 1.3 1 COMPARISON OF NUCLEAR STEAM SUPPIN SYSTEM DESIGN CilARACTERISTICS (Continued)

This Plant GESSAR NMP2 Grund Gulf AllWH llWH/6 IlWR/S IlWH/6 DnJgu 278 872 238 748 251 764 251 800 Core hicchanical (Continued)

Core diameter 203.3 185.2 1m2 191.5 (equivalent)

(in.)

Core height 146 150 146 150 (active fuel)

(in.)

Ernctor Control Sutelu (Chapters 4 and 7) hiethod of Movable hiovable hiovable hiovable variation of control control rmis control control reactor power rods and and variab!c rods and rmis and variable forced variable variable forced coolant coolant now forced coolant forced coolant h

flow flow flow Number of 205 177 185 193 movable con-trol rods Shape of Cruciform Cruciform Cruciform Cruciform movabic control rods Pitch of 12.2 12.0 12.0 12.0 movable control rods Control ll 4C ll 4C ll 4C Il 4C materialin granules granules granules granules movable rods compacted compacted compacted compacted in SS tubes in SS in SS in SS tubes tubes tubes O

Amendment i 1.L7

ABM 23h6100AC Standard Plant RIIV. A TABLE 1.31 o

O COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM ,

DESIGN CHARACTERISTICS (Continued)  ;

l I

This Plant GFJSAR N5f P 2 Grand Gulf AB%R BWR/6 BWR/5 BWR/6 Ihslgn i 278-872 238 748 251 764 251 800 Core Mechanical (Continued)

Reactor Control System (Chapters 4 and 7)

Type of Bottom Bottom Bottom Bottom control rod entry entry entry entry drives electric locking locking locking ,

hydraulic piston piston piston i fine motion  !

l Type of Burnable Burnable Burnable Burnable temporary poison; poison; poison; poison; Reactisity gadolinia- gadolinia- gadolinia- gadolinia-control for urania urania urania urania inital core fuel rods fuel rods fuel rods fuel rods In-core neutron instnimentation (Chapters 4 and 7)

Total number 208 164 172 176 of LPRM dectec-tors Number of in- 52 41 43 44 core LPRM penetrations Number of I.PRM 4 4 4 4 detectors per penetration Number of SRM

• 4 4 6 i penetrations Number of IRM 10' 8 8 8 penetrations Total nuclear 62 53 43 58 instrument penetrations

• For the AB WR, the SRM and IRM are combined as one and called SRNM.

/ i Amendment 1 1.k8 l l

I

-i

1 ABWR u^6im^c 1 Standard Plant RI!V A I 1

TAllLE 1.31 COMPARISON OF NUCLEAR STEAM SUPP13 SYSTEM .

I DESIGN CIIARACTERISTICS (Continued) l This Plant GESSAR NMP2 Grand Gulf I ABWR HWR/6 EWR/S HWR/6 Desicn 278 872 238 748 351 764 251-800 i Core Mechanleal (Continued) l In-core neutron l Instrurnentation (Continued)

Source range N/A monitor range l Intermediate * ' "

N/A range monitor 3 range Startup range "*

N/A N/A N/A l neutron monitor Power range Approximately 1% power to 125% powcr monitors range Local power 208 164 172 176 range monitors h

l Average power 4 4 6 8 )

range monitors  !

l Number and type 5 Sb Be 7 Sb-Be 7 Sb-Be 7 Sb-Be '

ofin core neutron sources Reactor Vessel (Section 53)

Material Low-alloy low-alloy Low-alloy Low alloy steel / steel / steel / steel / j stainless stainless stainless stainless i and Ni-Cr Fe clad clad clad I Alloy clad I l

1

• Shutdoan through criticality

" Prior to criticality to lowpower

"* Shutdoun through lowpower O

Amendment 1 1.3-9

23A6100AC Standard Plant anv. A TABLE 1.31 COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS (Continued)

His Plant GESSAR NMP2 Grand Gulf ABWR BWR/6 B%R/5 BWR/6 Design 278-872 238 748 251 764 251 800 Core Mechanical (Continued)

Reactor Vessel (Continued)

Design pressure 1,250 1,250 1,250 1,250 (Psi)8 Design tempera- 575 575 575 575 ture (OF)

Inside diameter 23-2 19-10 20-11 20 11 (ft in.)

Inside height 08 11 70-4 72 5 72 7 (ft in.)

Minimum base 7.50 6.0 6.19 6.19 metal thickness (cylindrical O scciie > <t )

Minknum cladding 1/8 1/8 1/8 1/8 thickeess (in.)

Reactor Coolant Rectreulation  !

(Chapter 5)

Number of recir- 0 2 2 2 culation loops Design pressure inlet leg N/A' 030 1650 1250 (psig)

ABHR design utili:es Reactor Intemal Pumps, (RlP) l O

Amendment 1 1.3 10

1 I

ABWR 23a6imac Standard Plant anv. A TAllLE 1.31 COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM h DESIGN CHARACTERISTICS (Continued)

This Plant GESSAR NMP2 Grund Gulf ABWR HWR/6 HWR/S HWR/6 Design 21H22 238 748 251 764 251-800 Reactor Coolant Recirculation (Continued)

(Chapter 5) outlet leg N/A*" 1650'; 1675'; 1650*;

(psig) 1550 " 1550 1550 "

Design temper- N/A*" 575 575 575 ture (OF)

Pipe diameter N/A"' 22/24 24 24 (in.)  !

l Pipe material N/A"' 304/316 316k 304/216  !

(ANSI)  !

Recirculation 30,516 42,000 47,200 44,600 pump flow rate / Pump (gpm) i Number of jet N/A*" 20 20 24 G;

pumps in reactor Main Steamlines (Subsection 5.4.9)

Number of 4 4 4 4 steamlines l

Design Pressure 1,250 1,250 1,250 1,250 (psig) i Design temper- 575 575 575 575 ture (0F)

Pipe diameter 28 26 26/28 28 (in.)

Pipe material Carbon Carbon Carbon Carbon I steel steel steel steel Pump and discharge piping to and including discharge block valve

' Dichargepipingfrom discharge block valve to vessel

• " ABWR design utili:cs 10 Reactor Intemal Pump, (RIP's)

Amendment 1 1.3-11

MIN 23A6100AC Standard Plaut REV.A l TABLE 1.3 2 o

O COMPARISON OF ENGINEERED SAFETY FEATURES l DESIGN CHARACTERISTICS  !

'Ihis Plant GESSAR NMP2 Grand Gulf ABWR BWR/6 BWR/5 BWR/6 System / Component 278-872 238-748 251-764 251-800 Emereeney Core Cooline System _s (sized on design power section 6.3) 14w Pressure Core Soray Systems

• Number ofloops N/A 1 1 1 flow rate N/A 6000 at 6350 at 7000 at (gpm) N/A 122 psid 128 psid 122 psid lilch Pressure Core Sorav System" Number ofloops 2 1 1 1 Flow rate 800 at 1550 at 1550 at 1650 at (gpm) 1177 psid 1147 psid 1130 psid 1147 psid 3200 at 6110 at 6350 at 7000 at 100 psid 200 psid 200 psid 200 psid Reactor Core Isolation Cooline System (Subsection 5.4.6)

Flow rate 800 at 700 at 600 at 800 at (gpm) 165 1192 165 1192 1173 165 1192 psia psia psia psia reactor reactor reactor reactor Pressure Pressure Pressure Pressure Automatic Deoressuration System Number of relief 8 8 7 8 valves ABHR design utilizes the lowpressureflooder mode of the RHR system ABHR design is aflooder system not a spray system O

Amendment 1 1.3 12

ABWR mime StalidarGant tu l a TAlli.E 1.3 2 COMPARISON OF ENGINEEREI) SAFETY FEATURES DESIGN CilARACTERISTIC (Continued)

'this Plant GESSAH NMP2 Graml Gulf AllWH llWH/6 IlWH/5 IlWH/6 Suttn1LCuiuventni 118El 2&l48 251dM 251 M Lun.l'rtuutt.Calandalts11un' Number ofloops 3 3 3 3 Number of pumps 3 3 3 3 110w rate 4200 at 71(X) at 7450 at 7450 at (gpm/ pump) 40 pal 20 psit! 26 psit! 20 pul AuAlljarISnitv13  !

linidunultat_ErniuialEnttui  !

(Sul>section 5,4.7)

Htattur311uldonittw}lognadt 1

1 i

Number of h> ops 3 2 2 2 Number of pumps" 3 flow rate 42tX) 2 71(X) 2 7450 2

7450 g

(gpm/ pump)

I Duty (MilTU/hr*" 29.0 46.9 41.t> 50.0 i heat exchanger)

J l

Number of heat 3 2 2 2 exchangers l'rimary contain- 4M) 71(X) 7450 74N) ment cooling mm!c 110w rate (gpm)

Allif *R design referred to as Low Pressure Fhunter The design of the purnps is in part based on the required capacity daring the reactor flooding rnode, The near exchanger duty for suppression pool cooHng is based on assurning they are placed in operation 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> after reactor shutdown.

O Amendment 1 i 11)

MN 23A6100AC Standard elant anv. A TABLE 1.3 2 O Couriaisox or exGixeentD sirerv ectrones DESIGN CHARACTERISTIC (Continued)

This Plant GESSAR NMP2 Grand Gulf ABWR HWR/6 HWR/5 BWR/6 Sntem/ Component 278-872 238-748 251 764 251-800 Reactor shutdown cooline mode (continued)

• 25,300 Flow rate 8000 7400 (gpm/ heat total exchanger

• 6 Number of pumps 3 loops 2 at RBCW 12,000 gpm 1 at 1,300 gpm Fuel Pool Cooline and Cleanun System (Subsection 9.1.3)

Capacity 'i.55 8.0 15.0 11.8 (MBtu/hr)

(,

l Not speci/ led.

l l

O Amendment 1 1.3 14

ABWR m61*^c Standard Plant mw. A TAllLE 1.3 3 COMPARISON OF CONTAINMENT O

DESIGN CIIARACTERISTICS nis Plant GESSAR NMP2 Grand Gulf AllWR llWR/6 IlWR/5 ilWR/6 Containment

• 278 872 233 748 25174 251 800 Priman Type Over and Mark III Over and hf ark 111 under free- under reinforced Pressure standing pressure concrete Suppres- steel with Suppres- containment sion reinforced sion with steel concrete Mark 11 liner shield building Construction Reinforced Cylindrical Reinforced Reinforced concretr free- with stccl concrete with stcel standing liner cylinder liner; steel with with hemi-steel cilipsoidal spherical structure head head; steel g lined W Dr>well Concrete Concrete" Frustum Concrete" cylinder cylinder ofconc cylinder upper portion Pressure- Concrete Free stand. Cylindri- Concrete suppression cylinder ing steel cal lower cylinder chamber with con. portion crete back-ing Containment 45 15 45 15 internal design )

pressure (psig)

Dr>well internal 45 30 45 30 design pressure I (psig)

Drywell frec 259,563 275,000 303,418 270,000 3

volume (ft )

lih rte applicable, containment parameters arc based on design ratedpower.

' Net part of containment boundary.

Amendment 1 1.1 15

l MM 23A61ooAc Standard Plant anv. a j l

TABLE 1.3 3 I O Come4alsos og CoxTiisuexT DESIGN CHARACTERISTICS (Continued) ,

'Ihis Plant GESSAR N M P-2 Grand Gulf ABWR BWR/6 BWR/5 BWR/6 Containment

• 278-872 238-748 251 764 251-800 Primary (Continued)

Pressure- 210,475 1,140,000 192,028 1,400,000 suppression chamber free volume (ft3 )(HWL)

Pressure- 126,426 129,600 154,794 136,000 suppression (uppc. pool (upper pool water dump = pool dump 3 34,200 = 72,800 volume (ft )(LWL)

Submergence of 11.8 to 20.8 7.5 11.0 max. 9.5 min.

vent pipe below pressure pool surface (ft) (HWL)

Design temper. 340 330 340 330 ature of drywell('F) f]

Downcomer vent 2.5 - 3.5 2.53.5 137 2.5-3.5 pressure loss factor Break area / 0.01 0.012 0.0108 0.008 total vent area Calculated maxi- 39 23.0 39.7 22.0 aum drywell pressure after blowdown (psig).

Hhere applicable, containmentparameters are based on design ratedpower.

O Amendment 1 1.3-16

ABWR m6imac Standard Plant nrw.A TABLE 1.3 3 COMPARISON OF CONTAINMENT O

DESIGN CilARACTERISTICS (Continued)

This Plant GESSAR NMP2 Grand Gulf ABWR HWR/6 BWR/5 HWR/6 Containment? 278-872 238 748 251 764 251-800 Primarv (Continued)

Pressure- 26 8.7 M.0 9.0 suppression chamber (psig)

Initial 50 50 50 30 pressure-suppression [w,1 temperature rise (OF) during LOCA Leakage rate 0.5 1.0 1.1 035

(% free volumc/

day)

Secondary Type Controlled Controlled Controlled Controlled leakage leakage leakage leakage elevated release Con 3truslien Lower levels Reinforced Reinforced Reinforced Concrete Concrctc Concrete Upper levels Reinforced Steel Stect Concrete super- super-structure structure and siding and siding Roof Reinforced Steel Steel concrete decking decking Hhere applicable, containment parameters are based on design ratedpower.

• Not Specified.

O Amendment 1 1.1 17

ABWR m6imc Standard Plant nity. A TABLE 1.3 3 COMPARISON OF CONTAINMENT DESIGN CHARACTERISTICS (Continued)

This Plant GESSAR NMP2 Grand Gulf ABWR BWR/6 BWR/5 BWR/6 Containment

• 278-872 238-748 251 764 251-800 Secondary (Continued)

Internal design 0.25 0.25 0.25 0.25 pressure (psig)

Design inleakage 100 100 100 100 rate (% free volume / day at 0.25 in.

H2O)

O Hhere applicable, containmentparameters are based on design ratedpower.

O Amendment 1 1.3 18

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

ABM uasimac Standard Plant REV.A TABLE 1.3-4 COMPARISON OF STRUCTURAL DESIGN CHARACTERISTICS

'Ris Plant GESSAR NMP2 Grand Gulf ALWR BWR/6 BWR/S BWR/6 278-872 238-748 251 764 251-800 Seismic Desien (Section 3.7)

Oxrating Basis Eanhquake horizontal g 0.15 0.15 0.075 0.075 vertical g 0.067 0.10 0.75 0.05 Safe Shutdown Earthquake honzontal g 03 0.30 0.15 0.* !

vertical g 0.20 0.20 0.15 0.10 Wind Desien '

(Subsection 33.2)

Translational 60 70 max. 70 60 (mph) 5 min.

Tangential (mph) 260 290 290 300 O

I

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G' Amendment 1 1.3 19  ;

.__m____. - _ . . _ _ _ _ _ _ _ _ _ _ _ . _ . _ . . . _ . _ . _ . . . . _ _ _ _ . _ _ _ . _ _ . _ . . _ . -._.__..

MN 23A6tooAc Standard Plant REV.A SECTION 1.4 O r m es Table Ikla East 1.4-1 Commercial Nuclear Reactors Completed, Under Construction, or In Design by General Electric 1.4-2 t

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MN 23A61ooAc Standard Plant REV.A 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS GE will design and deliver the direct cycle boiling water nuclear steam supply system, fabricate the first core of nuclear fuel, and provide technical direction for installation and startup of this equipment. GE has engaged in the development, design, cone.truction, and operation of boiling water reactors since 1955. Table 1.41 lists the GE teacto.rs comphted, under construction, or on order. Thus, GE has substantial experience, knowledge, and capability to design, manufacture, and furnish technical assistance for 'he installation and startup of reactors.

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Amendment 1 1,41 I

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

1 ABWR uuimac l Standard Plant arv.A l i

b Table 1.41 COMMERCIAL NUCLEAR REACTORS COMPLETED, UNDER CONSTRUCTION, OR IN DESIGN BY GENERAL ELECTRIC l l

Year of Rating Year of Low Power Station Utility (MWe) Order IJStnic Dresden 1 Commonwealth Edison 207 1955 1959 Humboldt Bay Pacific G&E 70 1958 1%2 KHAL Germany 15 1958 1%1 Garigli.mo Italy 150 1959 1964 Big Rock Point Consumers Power 72 1959 1%3 JPDR Japan 11 1960 1%3 ,

KRB Germany 237 1%2 1%7 '

Tarapur 1 India 190 1%2 1%7 Tarapur 2 India 190 1%2 1%9 GKN Holland 52 1%3 1968 Oyster Creek JCP&L 640 1%3 1%9 Nine hiile Point Niagara hichawk 610 1%3 1%9 Dresden 2 Commonwealth Edison 794 1%5 1%9 Pilgrim Boston Edison 670 1%5 1972 hiillstone 1 NUSCO 652 1%5 1970 Tsuruga Japan 340 1%5 1970 3

[d Nuclenor Fukusidma 1 Spain Japan 440 43) 1%5 1966 1971 1971 BKW KKhi Switzerland 306 1966 1972 Dresden 3 Commonwealth Edison 794 1966 1971 hionticello Northern 5tates 548 1966 1970 Quad Cities 1 Commonwealth Edison 789 1966 1972 Browns Ferry 1 TVA 1067 1966 1973 Browns Ferry 2 TVA 1067 1966 1974 Quads Cities 2 Commonwealth Edison 789 1966 1972 Vermont Yankee Vermont Yankee 515 1966 1972 Peach Bottom 2 Philadelphia Electric 1065 1966 1973 Peach Bottom 3 Philadelphia Electric 1065 1966 1974

)

Fitzpatrick PASNY 821 1968 1974 Shoreham LILCO 820 1%7 1984 Cooper Nebraska PPD 778 1%7 1974 I Browns Ferry 3 TVA 1067 1%7 1977 i Limerick 1 Philadelphia Electric 1100 1%7 1984 Hatch 1

)

Georgia Power 786 1%7 1974  ;

Fukushima 2 Japan 762 1%7 1975 i Brunswick 1 Carolina P&L 821 1968 1977 i Brunswick 2 Carolina P&L 821 1968 1974 Duane Arnold Iowa ELP 545 1968 1974 Fermi 2 Detroit Edison 1093 1968 1987 Hope Creek 1 PSE&G 1067 1%9 1984 O

V Amendmer.t 1 1 A-2 l

ABWR 2346ioarc Standard Plant REV.A Table 1.41 COMMERCIAL NUCLEAR REACTORS COMPLETED, UNDER CONSTRUCTION, OR IN DESIGN BY GENERAL ELECTRIC (Continued)

Year Of Rating Year of Low Power Statto.i Utility (MWe) Order Licenst Hope Creek 2 PSE&G 1067 1%9 1986 Chinshan 1 Taiwan 610 1%9 19M Caorso 1 Italy 822 1%9 1F Hatch 2 Georgia Power 786 1970 1978 La Salle 1 Commonwealth Edison 1078 1970 1982 La Salle 2 Commonwealth Edison 1078 1976 1983 Susquehanna1 Penns>vania P&L 1050 1%7 1982 Susqu-hanna 2 Pennsyvania P&L 1050 1968 1984 CSnshan 2 Taiwan 610 1970 1979 Hanford 2 WPPSS 1100 1971 1983 Nine Mile Point 2 Niagara Mohawk 1100 1971 1982 G-and Gr'f 1 Mississippi P&L 1250 1971 1982 Fukushiraa 6 Japan 1135 1971 1979 Tokd Japan 1135 1971 1977 Ri , bend l GulfStates 940 1972 1985 Perry 1 Cleveland Electric 1205 1972 1981 Laguna Verde 1 Leibstadt M:xico Switzerland 660 940 1972 1972 1988 1984 h

Kuv h ,ng i Tawian 992 1972 1981 Keyc.og2 Taxian 9^2 1972 D82 Chntor.1 Illinois Power 950 1973 1986 Ccnf.cr tes Spain 975 1973 1985 Laguna Veroe 2 Mexico 660 1973 1990 Alto Lazio 1 Italy 982 1974 1990 Alto Lazie 2 Italy 982 1974 1991 K.shiwazaki 6 Japan 1300 1987 194 Kashiwazaki 7 Japan 1300 1987 1998 l

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Amc' s e 1.43

ABM 23A6100AC Standard Plant REv. A 1.5 REQUIREMENTS FOR FURTHER Q TECHNICAL INFORMATION In the December 1986 technical description of the Advanced Building Water Reactor (ABWR) GE,in Section 3, provided a description of the test and  ;

development program associated with the ABWR. Of the effort described in that report all have been satisfactorly completed except the In Plant FMCRD test and the program related to the Reinforced Concrete Containment Vessel. The results of these two efforts will be available for staff review during the review of the safety analysis ,

report.

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Amendment 1 13 1

23A6100AC Standard Plant REV,A 1.6 MATERIALINCORPORATED BY Q REFERENCE This information will be provided by December 1988.

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Amendment 1 1.61

ABWR MA6100AC Standard Plant REV.A 1.7 DRAWINGS AND OTHER Q DETAILED INFORMATION This information to be supplied by December 1988.

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Amendment 1 1.71

ABM MA6100AC Standard Plant REV.A SECTION 1.8 O courEurS Section Title Page 1&1 Conformance With Standard Review Plan 1.8-1 1.82 Applicability of Codes and Standards 1.8-1 TABLES

.Tahle Tit!c Eage 1.8-1 Summary of Differences From SRI Section 1 1&2 1&2 Summary of Differences From SRP Section 2 1.8-3 1&3 Summary of Differences From SRP Section 3 1.8-4 ,

1.8-4 Summary of Differences From SRP Section 4 1.8-5 O 1&5 Semmarv ef Differences Fre- SRP Sectien s 1.8-6 1.8-6 Summary of Differences From CRP Section 6 1.8-7 1.8-7 Summary of Differences From SRP Section 7 1.8-8 1.8-8 Summary of Differences From SRP Section 8 1&9 1.8-9 Summary of Differences From SRP Section 9 1.8-10

)

1.8-10 Summary of Differences From SRP Section 10 1.8-11 1.8-11 Summary of Differences From SRP Section 11 1.8-12 1.8-12 Summary of Differences From SRP Section 12 1.8 13 i

1.8-13 Summary of Differences From SRP Section 13 1.8-14 1.8-14 Summary of Differences From SRP Section 14 1&l5 1&l5 Summary of Differences From SRP Section 15 1.8-16 O

v 1 . 8 - 11 caendment 1

ABM unamac Standard Plant _ niw. 4 SECTION 1.8 TABLES (Continued)

Inb!c Illic Eage 1.8-16 Summary of Differences From SRP Section 16 1.8 17 1.8-17 Summary of Differences From SRP Section 17 1.8-18 1.8 18 Summary of Differences From SRP Seriion 18 1.8 19 O

O 1.8 iii ,

Amendment 1

ABM ursiooac Standard Plant artv, A 1.8 CONFORMANCEWITH STANDARo l REVIEW PLAN ANo APPLICABILITY OF O ContSixo STANoARoS l

1.8.1 Conformance With Standard '

Review Plan i l

This subscetion provides the information <

requhed by 10 CFR 50. 34 (g) showing conformanes l with the Standard Review Plan (SRP). This 1 information is presented by SRP section in Tables  !

1.8-1 through 1.8-18.

1.8.2 ApplicabilityorCodes and Standards To be provided by June 1988.

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Amendment 1 1A1

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

23A6100AC Standard Plant REV.A TABLE 1.81 SUh1A1ARY OF DIFFERENCES FROh! SRP SECTION 1 AHWR Subsection Specific SRP Summary Description %here SRP Sectlan Acceptance Criteria of Difference Discussed (TO BE PROVIDED BY DECEMBER 1988)

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O Amendment I 142 1

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ABM ux6ioaac Standard Plant REV.A TABLE 1.8 2 g

SUMMARY

OF DIFFERENCES FROM SRP SECTION 2 ABWR Subsection Specific SRP Summan Description Mhere SRP Section Acceptance Criteria of Difference Discussed (TO BE PROVIDED BY DECEMBER 1988)

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ABM 234ciooac Standard Plant anv. A TABLE 1.8 3

SUMMARY

OF DIFFERENCES FROM SRP SECTION 3 AHWR Subsection Specific SRP .

Summary Description %here SRP Section Acceptance Criteria of Difference Discussed (TO BE PROVIDED BY DECEMBER 1988)

O O

Amendment 1 I 5-4

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

1 23A6100AC Standard Plant RI!V. A TABLE 1.8-4 g

SUMMARY

OF DIFFERENCES FROM SRP SECTION 4 AllWR Subsection Specific SRP Summary Descriptiou %here SRP Section Acceptance Criteria of Difference Discussed (TO BE PROVIDED HY DECEMHER 1988)

O O

Amendment 1 1A5

ABWR memc Standard Plant REV.A

,.} TABLE 1.8 5

SUMMARY

OF DIFFERENCES FROM SRP SECTION 5 ABWR Subsection Specific SRP Summary Description Mhere SRP Section Acceptance Criteria of Difference Discussed 4

(TO BE PROVIDED BY DECEMBER 1988)

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ABWR mamac Standard Plant any. A TABLE 1.8 6 g

SUMMARY

OF DIFFERENCES FROM SRP SECTION 6

/.BWR Subsection Specine SRP Summary Description Where SRP Section Acceptance Criteria of Difference Discussed (TO BE PROVIDED BY DECMIBER 1983)

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MM 23A6100AC Standard Plant REV.A i ,

J h TABLE 1.8 7

SUMMARY

OF DIFFERENCES FROM SRP SECTION 7 ABWR Subsection Specific SRP Summary Description Where SRP Section Acceptance Criteria of Difference Discussed

.I (TO BE PROVIDED BY DECEMBER 1988)

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Amendment 1 1.88

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ABM 23AG100AC Standard Plant RIN. A TABLE 1,8 8 g

SUMMARY

OF DIFFERENCES FROM SRP SECTION 8 ABWR Subsection Specific SRP Summary Description %here SRP Section Acceptance Criteria of Difference Discussed (TO BE PROVIDED BY DECEMBER 1988)

O O

AmeMment 1 1A9

ABWR us6mc Standard Plant nuv.A TABLE 1.8 9

SUMMARY

OF DIFFERENCES FROM SRP SECTION 9 AHWR Subsection Specific SRP Summary Description %here SRP Section Acceptance Criteria of Difference Discussed 4

(TO BE PROVIDED BY DECEMBER 1988)

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I Amendment 1 1.8 10

ABWR aw c Standard Plant REV.A TABLE 1.810 SU5th!ARY OF DIFFERENCES FRO 51 SRP SECTION 10 AllWR Subsection Specific SRP Summary Description Where SRP Section Acceptance Criteria of Difference Discussed (TO BE PROVIDED llY DECEMilER 1988)

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ABM 234simac Standard Plant anv. A Q TABLE 1.8-11

SUMMARY

OF DIFFERENCES FROM SRP SECTION 11 ABWR Subsection Specific SilP Summary Description . Where SRP Section Acceptance Criteria of Difference Discussed i

1 (TO BE PROVIDED BY DECEMBER 1988) <

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Amendment 1 ]L12

ABM .un61ooac Standard Plant nirv. 4 TABLE 1.812 g

SUMMARY

OF DIFFERENCES FROM SRP SECTION 12 AHWR Subsection Specific SRP Summary Description Where SRP Section Acceptance Criteria of Difference Discussed (TO BE PROVIDED HY DECEMBER 1988)

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ABM ursiooac Standard Plant nity, A Q TABLE 1.813

SUMMARY

OF DIFFERENCES FROM SRP SECTION 13 J

ABWR Subsection Specific SRP Summary Description Mhere SRP Section Acceptance Criteria of Difference Discussed (TO BE PROVIDED BY DECEMBER 1988)

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23A6100AC Standard Plant Rev.A TABLE 1.814 g

SUMMARY

OF DIFFERENCES FROh! SRP SECTION 14 ABWR Subsection Specific SRP Summary Description %here SRP Section Acceptance Criteria of Difference Discussed 1

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Amendment 1 1.8 15

ABWR DA6100AC Standard Plant anv. A r

Table'1.815 O

SUMMARY

OF DIFFERENCES FROM SRP SECTION 15 ABWR l Subsection Specific SRP Summary Descrir.tlon Where SRP Section Acceptance Criteria of Difference Discussed 1

(TO BE PROVIDED BY DECEMBER 1988)

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ABWR zwimac  ;

Slandard Plant REV.A 1

TABLE 1.816 g

SUMMARY

OF DIFFERENCES FROM SRP SECrlON 16 1 AllWR ,

Subsection l Specific SRP Summary Description Where l SRP Section Acceptance Criteria of Difference Discussed l (TO BE PROVIDED llY DECEMBER 1988)

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O Amendment 1 1.R-17

ABM 23461ooac Standard Plant any. A TABLE 1.817 O

SUMMARY

OF DIFFERENCES FROM SRP SECTION 17 AHWR Subsection Specific SRP Summary Description %here SRP Section Acceptance Criteria of Difference Discussed (TO BE PROVIDED BY DECEMBER 1988) 1 O  !

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ABM un61ooac Standard Plant REV.A TABLE 1.8-18 g

SUM 51ARY OF DIFFERENCES FROh! SRP SECTION 18 ABWR Subsection Spe cific SRP Summary Description Where SRP Section Acceptance Criteria of Difference Discussed (TO BE PROVIDED BY DECEMBER 19&M) 0 1

i 9

Amendment 1 1A19

ABWR 2346imac Standard Plant REV.A 1.9 INTERFACES O To be prosided by December 1988.

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Amendment 1 1.9-1 l

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I AFPENDIX 1A UNRESOLVED SAFETY ISSUES O

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ABWR ux6200ac Standard Plant REv. A 1A.1 UNRESOLVED SAFETY ISSUES O Will be prosided by December 31,1988.

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O Amendment 1 1 A.. ]

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0 1

CHAPTER 2 SITE CHARACTERISTICS l O l O

ABWR 2u6-n Standard Plant Rriv. A CHAPTER 2 O TAnts or CoxTesTS l

Section ]ltic Eags 1 2 SITE CHARACTERISTICS 2.0 SUh1NfARY 2.0-1 2.1 L151ITS 151 POSED ON SRP SECTIONJJ ACCEPTANCE CRITERIA BY NUCLEAE ISLAND DESIGN 2.11 l

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4 ABM 234610o4 0

Standard Plant REV.A' SECTION 2.0 i S CONTENTS l TABLES  :

Table 1111e East i^

2.0-1 Envelope of ABWR Plant Site Design Parameters 2.0-2 i

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2.0-il Amendment 1 4

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

MN 23A6100AD Standard Plant REV.A 2.0

SUMMARY

j This section defines the envelope of site related parameters which the ABWR standard plant is designed to accommodate. These parameters envelope most potential sites in the U.S. A summary of the site envelope design parameters is given in Table 2.01.

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Amendment 1 2.01 s

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M 2W100AD Standard Plant RIIV. A TABLE 2.01 O ENVELOPE OF ABWR PLANT SITE DESIGN PARAMETERS hiasimum Ground Water Level: Extreme Wind: Basic Wind Speed:

2 feet below grade 110 mph (1)/130 mph (2) hiasimum Flood (or Tsunami) 14sel:(3) Tornado:(4) 1 foot below grade hiaximum tornado wind speed: 260 mph Translational velocity: 57 mph

- Radius: 453 ft Precipitation (for Roof Design): - Maximum atm AP: 1.46 psid hiaximum rainfall rate: 10 in/hr hiissile Spectra: Per ANSI /ANS 2.3 hiaximum snow load: 50 lb/sq. ft.

Design remperatures: Soll Properties:

Ambient hiinimum Bearing Capacity (demand): 15ksf 1% Exceedance Values - hiinimum Shear Wave Velocity: 1000 fps

- hiaximum: 1000F dry bulb /770F coincident wet Liquification Potential:

bulb None at plant site resulting hiinimum:-10 F from OBE and SSE I 0% Exceedance Values (Historicallimit) hiaximum:115 Fdrybulb/82 Fcoincident wet Seismology: i bulb OBE Peak u. .,und Acceleration (PGA):

p hiinimum: 400F 0.10g(5) (6) d Emergency Cooling Water Inlet: 950F SSE PGA : 0.30g(5)

- SSE Response Spectra: per Reg. Guide 1.60 1 SSE Time History: Envelope SSE Response Spectra (1) S0 year recurrence interval; value to be utilized for design of non-safety related structures only.

(2) 100 year recurrence interval; value to be utilized for design for safety-related structures only.

(3) Probable marimum flood level (PMF), as defined in ANSI /ANS-2.8, ' Determining Design Basis Flooding at Power a ^ctor Sites."

(4) 1,000,000-year tomado recurrence interval, with associatedparameters based on ANSI /ANS-2.3.

(3) Free-field, at plant grade elevation.

(6) For conservatism, a value of 0.15g is employed to evaluate structural and component responses in Chapter 3.

Amendment 1 2.02

t ABM DA6100AD

' Standard Plant REV.A SECTION 2.1 i

h CONTENTS

! TABLES i Table 11tle P_ ass l

2.1 1 Limits imposed On SRP Section II Acceptance Criteria by Nuclear Design 2.12 i

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ABM useiman l Standard Plant arv.A 2.1 LIMITS IMPOSED ON SRP SECTION 11 ACCEITANCE CRITERIA BY NUCLEAR ISLAND DESIGN This section defines limits imposed on SRP Section II acceptance criteria 'ay (1) the envelope of ABWR plant site design parameters given in Table 2.01 and (2) the assumptions, both implicit and explicit, related to site characteristics employed in the evaluation of the Nuclear Island. These limits are presented in Table 2.1-1 for the following five SRP Section U categories of site characteristics:

(1) Geography and Demography;  !

(2) Nearby Industrial, Transportation and i Military Facilities; l 1

(3) Meteorology; I (4) Hydrology Engineering; and l (5) Geology. Seismology and Geotechnical Engineering.

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Amendment 1 2.11 l l

4

, - - _ . . _ , . - . _ _, _ _ . . _ ._ _ _ _ . - - _ __.-,- 7,._r- ---.-n,--,,,.- x

ABWR u ^6too^ o Standard Plant _ RM TABLE 2.1 1 n

U LIMITS IMPOSED ON SRP SECTION II ACCEPTANCE CRITERIA BY NUCLEAR ISLAND DESIGN SRP SECTION SUBJECT LI51tTS GEOGRAPIIY AND DE310GRAPIIY 2.1.1 Site Location and Description None.

2.1.2 Exclusion Area Authorityand None.

Control 2.13 Population Distribution None.

NEARBY INDUSTRIAL, TRANSPORTATION AND hilLITARY FACILITIES 2.2.1 Identification of Potential Identify potential hazards in the site vicinity 2.2.2 Hazards in Site Vicinity that have a probability of occurrence >10 7 per year which produce: (1) missiles more energetic than the AISI/ANS 23 missile spectra, or (2) pressure affects in excess of the design basis tornado.

2.23 Evaluation of Potential Accidents Evaluate only those potential hazards identified above.

I 1

blETEOROLOGY l

23.1 Regional Climatology Per Table 2.0-1. 1 23.2 local hteteorology None 233 Offsite hieterological hicasure- None.

ment Programs 23.4 Short Term Dispersion Estimates Show that the site meteorological dispersion for Accidental Atmospheric values as calculated in accordance with Releases Regulatory Guide 1.145, and compared to dose values given in Chapter 15, result in doses less than stipulated in 10CFR100 and the applicable portions of SRP Sections 11 and 15, 23.5 1.ong Term Diffusion Estimates None.

O Amendment 1 2.12

ABWR meimn Standard I'lant iam A TAllLE 2.1 1 LIMITS IMPOSED ON SRP SECTION II ACCEL'TANCE CRITERI A O IlY NUCLEAR ISLAND DESIGN (Continued)

SRP SECTION SUll. LECT L1511T5 IlYDROI.OGY ENGINEERING 2.4.1 liydraulie Description Per Table 2.0-1.

2.4.2 Fhxxis Per Table 2.0-1.

2.43 Probable hf aximum Flood on None.

Streams and Rivers 2.4.4 Potential Dam Failures Demonstrate that failure of existing and Scismically Induced potential upstream or downstream water control structures will not exceed ikxxling I foot below grade.

2.4.5 Probable blaximura Serge and Probable maximum serge and sicche flooding 11 Sicche Flomling foot below grade.

2.4 r3 Probable hiaximum Tsunami Probable maximum tsunami Ikxxling s I foot below grade.

2.4.7 !ce Effects None.

2.4.8 Cooling Water Channels Demonstrate that cooling water chahnels and and Reservoirs reservoirs supplying emergency cooling water has a failure probability of s 10 7 pc year.

2.4.9 Channel Diversion None.

2.4.10 Flooding Protection Requirements None.

2.4.11 Cooling Water Supply Failure rate of emergency cooling water supply 1 10'7 per year.

2.4 12 Groundwater Per Table 2.01.

2.4.13 Accidental Releases of Liquid None.

Ef0uents in Ground and Surface Waters 2.4.14 Technical Specifications and None.

Emergency Operation Requirement e

AmenJment 1 2 1..)

i

ABWR zwiovan Standard Plant anv.A TABLE 2.1 1 LIMITS IMPOSED ON SRP SECTION II ACCEPTANCE CRITERIA BY NUCLEAR ISLAND DESIGN (Continued)

SRP SECTION SURIECT 1.15flTS GEOLOGY, SEIShlOLOGY AND GEOTECllNICAL ENGINEERING 2.5.1 Basie Geology and Seismic None.

Information 2.5.2 Vibratory Ground Motion Per Table 2.0-1.

2.53 Surface Faulting None.

2.5.4 Stability of SuNurface Per Table 2.0-1.

Materials and Foundations 2.5.5 Stability of Slopes None.

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Amendment 1 2.1 -4

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APPENDIX 3A SEISMIC SOIL-STRUCTURE IMEMI N ANAIJSIS O

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ABM nasioast Standud Plant REV A APPENDIX 3A O

TABLE OF CONTENTS Section IWs Eage 3A SEISMIC SOIL-STRUCTUREINTERACTION ANALYSIS 3A.1 INTRODUCTION 3A.1 1 3A.2 ABWR STANDARD PLANT SITE PLAN 3A.21 3A.3 GENERIC SITE CONDITIONS 3A3-1 3A3.1 Soil Deposit Depth 3A3-1 3A3.2 Soil Profile and Properties 3A3-1 3A33 Ground Water Table 3A3-2 3A3.4 Summary of Site Conditions 3A3-3 3A..t INPUT MOTION AND DAMPING VALUES 3A 4-1 3A.4.1 Input Motion 3A.4-1 3A.4.2 Damping Values 3A.41 3A.5 SOll-STRUCTURE INTERACTION ANALYSIS METilOD 3A.5-1 3A.S.1 Introduction 3A.5-1 3A.S.2 Complex Response Method 3A.5-1 l 3A.53 Methodology and Analysis Procedure 3A.5-1 3A.6 FREE FIELD SITE RESPONSE ANALYSIS 3A.6-1 3A.7 SOll STRUCTURE INTERACTION ANALYSIS CASES 3A.71 l

3A.8 ANALYSIS MODELS 3A.81 l 3A.8.1 Structural Models 3A.81 3A.S.2 Foundation Models .3A.8-1 0

3A ii Amendment 1

ABWR m6imu Standard Plant REY.A APPENDIX 3A g TABLE OF CONTENTS (Continued)

Section 11 tic Pagt 3A.9 ANALYSIS RESULTS 3A.91 3A.10 REFERENCES 3A.101 3AA ATI'ACilMENT A TO APPENDIX 3A:

SSI ANALYSIS USING ALTERNATE APPROACil 3AA.1 INTRODUCTION 3AA1 3AA.2 METilODOI.OGY 3AA 2 3A.O ANALYSIS CONDITIONS 3AA5 3AA.4 REFERENCES 3AA-6 9

9 3A iii Amendment 1

l 23A6100AE Standard Plant Rev. A i l

i O 3^ 1 israo ov erio x d - tr i a i= s ciie 3.8 =>i 8 8 sei -ic design loads presented in Appendix 3G, which This appendix presents soil structure are obtained as a result of application of the interaction (SSI) analyses performed for the SSI on the methodology described in this generic site conditions adopted for establishing appendix. The RB at any site would be seismic design loads and seismic adequacy of the considered seismically qualified provided that reactor building (RB) complex of the ABWR in addition to the acceptable geotechnical data:

standard plant for a 0.15g' operating basis earthquake (OBE) and a 0.3g safe shutdown (1) The peak ground acceleration is less than or earthquake (SSE) excitation. The free field equal to 03g SSE and 0.15g OBE.

design spectra of OBE and SSE are described in Subsection 3.7.1 and are defined per Regulatory (2) Th.: site design response spectra are less Guide 1.60. The SSI analysis results in the form than or equal to those given in Regulatory of site enveloped seismic design loads at key Guide 1.60 normalized to the peak ground locations in the RB complex are presented in accelerations in item (1).

Appendix 3G.

(3) There is no potential for liquefaction at in order to ensure the seismic adequacy of the the plant site due to OBE and SSE as RB structures and the associated reactor and reviewed and concurred with by the NRC staff other equipment, an extensive seismic analysis is (the liquef aetion poteatial of tbe required. For a standard plant design, the foundation and site soils will be analysis roust performed over a range of site investigated and reported for a long parameters. The site parameters considered and duration, New Madrid type earthquake).

their ranges together form the generic site conditions. The generic site conditions are (4) There is no potential for fault displacement selected to provide an adequate seismic design at the plant site as reviewed and concurred O margin for the standard reactor building facility located at any sites with site parameters within with by the NRC staff, the range of this study. For sites to be located (5) The embedment depth of the reactor building with these facilities site-specific geotechnical is 25.7 m (85 ft). The excavation tolerance data will be developed and submitted to the NRC is 115 cm (10.5 ft).

! demonstrating compatibility with the design analyses assumptions (see Subsection 2.5.4). (6) The average shear wave velocity for the top 9 m (30 ft) of soil is 305 m/sec (1000 This appendix details the basis for selecting it/sec) minimum. The upper bound shear wave the site conditions and anal, sis cases, and the velocity is 3048 m/sec (10000 ft/sec).

method of the seismic soil structure interaction analysis. A description of the input motion and (7) For layered soll sites with parameters which damping values, the structural model, and the have very abrupt variations with depth, soil model are included. analysis with site unique properties will be performed to confirm the applicability of Seismic adequacy of the reactor building is the generic analysis.

l l

4 1

The OBE given in Chapter 2 is one third of l the SSE, i.e., 0.10g. However, as discussed in Chapter 2, a more conservative value of on e half of the SSE, i.e., 0.15g, was employed to evaluate the structural and component response.

V Amendment 1 3A.1 1

ABWR 2mimit Standard Plant _ REV, A (8) The soil bearing capacity at the site is adequate to accommodate plant design loads.

g To demonstrate the seismic adequacy of the standard ABWR reactor building design, a total of 42 SSI cases are analyzed for 14 generic site conditions using the finite-element method for the OBE condition. The cnveloped results reported in Appendix 3G form the design OBE loads in which the effect of a different SSI analysis l method is also included. l The design SSE load magnitude is taken as two times the design OBE load magnitude. No separate SSI analysis is performed for the SSE input with higher damping values. The design SSE loads are conservatively taken as two times the design OBE loads.

l 1

l O

O Amendment 1 3A.I.2

MM 23A6100AE '  !

4 Standard Plant REV A f SECTION 3A.2 -

h ILLUSTRATIONS i

Figure .T111t East l, 3A.21 ABWR TypicalSite Plan 3A.2 2 i i

l 3A.2-2 001800 Section View 3A.2 3 4

i l 4 .

t A

s I

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i i

1 1 1

! l l

i IO I

I l l

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I 4

.a 1

1 i

i J

i J

1 i i i l 1

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3A.2 il 1

J Amendment 1 i

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

M 23A6100AE Standard Plant RN A 3A.2 ABWR STANDARD PLANT SITE PLAN C)

The typical site plan of the ABWR standard plant is shown in Figure 3A.21. The plan orientations are identified by 00 1800 and 900 2700 directions. The cross section along the 001800 is shown in Figure 3A.2 2. The reactor building is nearly square in plan with dimensions of 56 m x 59 m (184 ft .,194 ft). The building is deeply embedded with embedmunt depth of 25.7 m (85 ft). Adjacent to the /eactor building along the 00 direction it. the control building. This building is rectangular in plan having a plan dimensions of 16 m x 45 m (52 ft x 147 ft) and the embedment depth of 12,2 m (40 ft). Adjacent to the control building in the 00 direction farther away from the reactor building is the turbine building wh;eh is 106 m (348 ft) long, 58.5 rn (192 ft) wide, and 18.3 m (60 ft) embedded. Adjacent to the control building in the 900 direction is the service building which is 16 m x 27.5 m (53 ft x 90 ft) with side extension toward the reactor building side of 10 m x 12.5 m (32 ft x 40 ft). Tbc service buildings, like other structures shown in Figure 3A.21, are considered minor for their

\O effects on SSI of RB, and are not considered in this appendix. All buildings are supported on separated basemats. The separation distance between buildings is 2 m (6.6 ft).

The primary objective of t'ae analysis  !

performed in this appendix is to obtain seismic response for the reactor building complex. The consideration of other buildings in the seismic analysis is only regt. ired when they are expected to affect the seismic response of the reactor building.

In the modeling of the buildings, the 00 1800 and 900 2700 directions are designated as X and Y axes, respectively. As evident from the site plan, the consideration of structure to structure interaction effe:t in the I SSI analysis for the reactor building is only required for X direction. For the analysis in the Y direction, only the reactor building by itself needs to be considered since the structure to structure interaction effect in this direction is expected to be insignificant.

O Amendment 1 3A.21

ABMR 23xaoort Standard Plant REVJ O

1I r

y St.1 m ,;

!.'........... ............{

I  :

I TURBINE BUILDING ::

., . ,t CONTROL BUILDING O- ,

em REACTOR SUILDING Y

270* M' I

l m

1H'

, Mm Figure 3A.2-1 ABWR TYPICAL SITE PLAN Amendment 1 3A22

~

3k E

3 em c2. #

5  %#

2$

E 2m 2m

--*1 l+- -+{ l+--

1, 59 m 33, 16 m _ , 106 m i g- -l CONTROL BUILDING GRAD E L 7.O m REACTOR l -

TURBINE BUILDING a m BUILDING 18.3 m 7///////H/3 12.2 m  ?///////l///////////////////////////l///l ///////////// '

- wass/mzan NOTE: ELEVATION IS RELATIVE TO THE RPV BOTTOM HE AD 88 027-50 h

7 ES v" sg Figure 3A.2-2 0*-180* SECTION VIEW >m O O 9

ABM 23A6100AE REV.A Standard Plant O SECTION 3A3 CONTENTS Section M Eagt 3A.3.1 Soll Deoosit Denth 3A3-1 3A3.2 Soll Pronic and Pronerties 3A3-1 3A33 Ground Water Table 3A3-2 3A3.4 Surnmary of Site Conditions 3A3 3 TABLES ,

Table Illie Eage  !

3A3-1 Average Shear Wave Velocities in Layers  !

(Water Tab!c at 0.61 m (2 ft) Below Grade 3A3-5 3A3-2 Strain Dependent Sheat Modulus 3A3-6 3A3-3 Strain Dependent Soil Damping 3A3-7 3A3-4 Average Shear Wave Velocities in Layers For UB Profile With Other Water Tables 3A3-8 3AS-5 Site Conditions Considered (Water Table at 0.61 m (2 ft) depth except noted) 3A3-9 ILLUSTRATIONS Figure Iille Eags i

3A3-1 Shear Wave Velocity Profiles Considered for SSI Analyses 3A3-10 3A3-2 Range of Shear Wave Velocities for Nuclear Power Plant Sites in High Seismic Areas 3A3-11 3AS-3 Variation of Shear Modulus and Damping Ratio with Shear Strain Used in Analyses 3A312

,l

)O 3m Amendment 1 ev

ABM u s6ioo w Standard Plant RW A q

u, 3A.3 GENERIC SITE CONDITIONS 61 m (200 ft).

This section describes the generic site In sumtnary, the variations of soil deposit conditions and their design parameters based on a thickness are accounted for by considering the range of soil properties used in the following four representative soil deposit soil structure interaction analysis described in depths.

this appendix.

Minimum (embedment depth) 25.7 m (85 ft)

Tbc site conditions are varied to cover a Shallow soil deposit 45.7 m (150 ft) range of expected site conditions in relatively Ictermediate soil deposit 61.0 m (200 ft) high seismic areas where a nuclear power plant Deep soil deposit 91.5 m (300 ft) may be constructed.

3A.3.2 Soll Profile and Propcrties From the SSI analysis point of view, site conditions can be characterized in terms of: (a) The range of soil profiles considered in soll deposit depth above bedrock, (b) soil this appendix is based on the velocity profiles profile and properties, and (c) ground water used in GESSAR (Refs. I and 2). A total of level. Parameter variations in each of these eight velocity profiles are selected and shown three areas for establishing generic site design in Figure 3A.31. These velocity profiles are envelopes are presented as follows. designated with the abbreviations: UB, VP2 through VP6, HR, and EH.

3A.3.1 Soll Deposit Depth The profile UB represents a soil profile of To encompass most of the potential site con- which the shear wave velocity at a depth y below ditions, a broad range of soil deposit depth is the ground surface is obtained using the modulus considered. The minimum depth is the embedment parameter K2 max and the following equations:

O depth for which the building is supported direct.

V ly on rock. For this case, the soil depth is Gmax(y)= 1000 K2 max / am(7) (3^*l) 25.7 m (85 ft) which is the reactor building __

embedment depth. Vs = / Gmax/r (3A 2)

The other soil depths are determined using the where GESSAR (Refs.1 and 2) as the guide. By maintaining the ratio of soil depth below the Gmax = maximum shear modulus in psf basemat to the basemat width to be approximately am = cffective mean pressure (in psf) at the same as for the GESSAR design, the shallow depth y; it is assumed equal to 0.7 depth is chosen to be 45.7 m (150 ft). times the effective overbutden pressure, which corresponds to the The deepest soil deposit considered in GESSAR use of an st rest coefficient of is 91.5 m (300 ft). This depth is selected on latet al pressure equal to 0.55.

the basis of a survey of sixty two U. S nuclear K2 max = modulus parameter power plant sites, which indicates that the rock p = mass density level in the majority of sites is located at a Vs = shear wave velocity depth less than 61 m (200 ft) from the ground surface. A 91.5 m (300 ft) depth is, thus, a This soit profile is assumed to consist of reasonable upper bound and, therefore, is seven horizontallayers. The K2 max value, total selected to be the deep soil deposit case for unit weight, and Poisson's ratio for each layer  ;

this appendix. Between the shallow and deep soil are as shown in Table 3A.31.

cases, an intermediate depth is chosen to be at

O l 1

Amendment 1 3.G1 I

l

ABWR 22^6 = ^n Standard Plant nrN. A Note that for submerged layers the Poisson's velocities in layers for all soil profiles are ratio is to be adjusted such that the minimum tabulated in Table 3A.3 2.

P wave velocity is not below the P wave velocity g

in water. The values of average shear wave The shear modulus and material damping of velocity, shown in Figure 3A.31 for UB, are soil are strain dependent. Figure 3A.3-3 shows computed using Eqs. 3A 1 nnd 3A 2 at the the variation of shear modulus and damping mid depth of each layer for the ground water ration with shear strain for various soit pro-level at a depth of 0.61 m (2 ft) below grade, files considered. On the basis of the recommen-dations made in Ref. 3 which indicates that the Tbc profiles VP2 through VP6 are selected best estimate values of shear modulus should not based on three generalized soil zones shown in be less that 40% of their low strain values and Figure 3A.3 2: a soil zone (sands, silts, clays, values of internal soil damping of a hysteretic and gravelly soils), a transition zone, and a nature should be limited to a maximum of 15% of soft rock and well ccmented soil zone. Velocity critical, the shear modulus reduction factors profile VP2 represents an average profile of the and damping ratios at various strain levels are soil zone; VP3 and VP4 bound the transition zone, shown in Tables 3A.3-3 and 3A.3 4. For the HR and VP4 and VP6 bound the soft rock and profile the free field site response analysis well cemented soil zone. Those velocity profiles results show that the strain compatible shear are smooth curves representative of the average modulus and material damping are essentially variation of shear modulus with depth that can be unchanged from their initial values. Therefore, expected within each of the soil zones. the SSI analyses for the liR profile and harder EH profile are performed using the initial modu-The profile HR represents a hard rock site lus and a conservative material damping of 0.1%

with a uniform shear wave velocity of 1524 m/sec (undamped). The effect of ground water on soil (5000 ft/sec). The profile EH represents an properties is addressed in Subsection 3A.3.3.

extra hard rock site with a uniform shear wave velicity of 3048 m/sec (10000 ft/sec) which is a 3A.3.3 Ground Water Table realistic uper bound velocity of rock sites a T

suitable for nuclear power plants. For soils below the ground water table, its unit we:3 ht is reduced due to submergence based The lower bound shear wave velocity for the on the following equation:

top 9 m (30 ft) of soil among all profiles consi-dered is 303 m/sec (994 ft/sec). The upper bound y' = y 79 (3A3 3 velocity is 3048 m/sec (10000 ft/sec). This con-stitutes a reasonable range of potential site conditions that are suitable for nuclear power where plants.

y' = the submerged unit weight The general soil layer properties (layer thickness, total unit weight, and Poisson's y = the total unit weight ratio) defined for the UB profile are also adopted for all other profiles except HR and EH. yu = the unit weight of water (1.0 The profiles HR and EH are considered to be t/m3 or 62,4 pcf) uniform elastic half space with a constant Poisson's ratio of 0.3 and unit weight density of 2.25 t/m3 (140 pcf). These values are also The effective (mean) confining pressure at used to be the half space bedrock properties in depth y within a soil deposit, assuming a the SSI analyses for all soil cases considered, horizontal water table, can be caiculated as:

The shear wave velocity of the bedrock is considered to be 3R8 m/sec (10000 ft/sec) for EH om = 1 + 2Ko (7 hi + y' h2) profile and to be 1524 m/sec (5000 ft/sec) for 3 (3A 4) all other profiles. The average shear wave O

Amer 4 ment 1 3A.3-2

ABM u^6ioore Saandard Plant arv.A where In order to evaluate the effects of water table location variation (assuming no soil

( Ko = coefficient of horizontal carth pressure failure) on structural response, three water at rest at depth y table locations are considered, namely, the bi = height of the column of unsaturated soil high, intermediate, and low water tables. The above depth y high water table is the base case which is h2

= height of the column of saturated soil located at 0.61 m (2 ft) below grade. The low above depth y water table is taken to be 25.7 m (85 ft) which is at the base of the reactor building Using 0.55 for Ko. Eq. (3A-4) becomes foundation basemat. The intermediate water table is assumed at 12.2 m (40 ft) below grade om = 0.7 ( y h1 + y'h2 ) (3A 5) which is at about the midheight of the reactor building embedment. Since the ground water may Since the shear modulus is a function of the have more pronounced effects on soft sites, the effective confining pressure which is, in turn a UB profile with 45.7 m (150 ft) depth of soil function of the soll unit weight, the location of deposit is investigated for all three water ,

the ground water table affects the shear modulus tables defined above. The average shear wave and the wave propagation velocity in the soil. velocities of the UB profile 'naving the water Since the water is essentially incompressible and table located at 12.2 m (40 ft) and 25.7 m (85 cannot resist shear, the presence of the water ft) below grade are shown in Table 3A.3 5. The has an important influence on the compressional water table level for other site conditions is wave velocity but produces only a minor effect on based on the basic high water table case which the shear wave velocity. is at 0.61 m (2 ft) below grade.

The effect of ground water on the soit 3A.3.4 Summary of Slie Conditions properties is considered using the following procedure: The above discussions cover the range of site O parameters in terms of soil deposit depth, soil (a) Perform one dimensional deconvolution profile and properties, and water table analysis for the horizontal excitation location. Based on the GESSAR experience (Refs.

component to obtain strain compatible shest 1 and 2) that the shallower soil depth in modulus, G, corresponding to the induced general resulted in higher structural response, strain level. The corresponding shear wave all velocity pofiles, except the HR and Eli velocity, Vs, is then computed as profiles, are considered for the 45.7 m (150 ft) shallow soil case, and only a limited soil Vs = / G/p (3A-6) velocity profiles need be considered for other l depths. For the deeper deposits of 61 m (200 r

(b) Compute the corresponding compressional wave f t) and 91.5 m (300 f t) in depth, it is I velocity, V p, using the following sufficient to consider only the UB profile since equation: its shear wave velocity profile has the largest variation with depth. Consequently, the effects Yp of variation in soil properties with depth is

=

V /g#

s (3A.7) accounted for more representatively. The minimum soil depth of 25.7 m (85 f t) is where p is a Poisson's ratio. The lower bound considered for the UB, VP4, HR and EH profiles of V p or f the submer3d soil below water table which adequately cover the range of profiles l is the compressional wave velocity of water taken considered. As mentioned before, the water to be 1463 m/sec (4800 ft/sec). When the table variation is taken into account by the UB computed V p of soil is smaller than 1463 m/sec, profile for the shallow deposit case. A total adjustment of soil Poisson's ratio is required so of 14 site conditions are selected for seismic that the Vp of soil is equal to 1463 m/sec. analysis and are summarized in Table 3A.3 6.

O j Amendment t 3A3 3

' j l

l

23A6100AE Sagdard Plant nV. A l Q Table 3.31 Soll Properties for UB Profile d

layer Depth Total Unit Weight

• Poisson's

• M(nominal) D K2rnas gn' M tatig 0-9 0 30 140 1.92 120 038

. 9-15 30 50 140 2.00 125 035 15 22.5 50 75 140 2.00 125 035 22.5 30 75 100 140 2.00 125 035 30-45 100-150 160 2.09 130 032 45-60 150-200 160 2.09 130 032 4

60 90 200 300 240 2.17 135 030 1 O i Total unit weight and Poisson's ratio for each layer are also i

used for other soll prof.*les, except for the two homogeneous rock profiles HR and EH of which the total unit weight is 2.25 t/m3 (140 pcf) and Poisson's ratio is 0.3.

i i

i

. O Amendment 1 14.14 l

1 1

ABM ux61oore Standard Plant Rev. A Table 3A.3 2 Average Shear Wave Velocities in Layers OVater Tab!c at 0.61 m (2 ft) Below Grade)

Laytr VELOCin' PROFILES Depth (M/SEC)

M (nominal) LIB m YPJ m YPJ XIM llB Ell 0-9 303 375 549 732 887 1067 1524 3048 9 15 374 412 576 738 893 1067 1524 3048 15-22.5 Alf 585 756 912 1067 1524 3048 22.5 30 45) ;1 610 762 915 1067 1524 3048 30-45 524 463 622 768 918 1067 1524 304S 45-60 573 488 M3 793 930 1067 1524 3048 60-90 758 521 674 817 945 #~ 1524 3n48 Layer VELOCIn' PROFILES Depth (FT/SEC)

EI 1.!I1 m YP2 m YPl YP4 IIB Ell 0 30 994 1230 1800 2400 2910 3500 5000 10000 30 50 1226 1350 1890 2420 2930 3500 5000 10000 50-75 1372 1430 1920 2480 2990 3500 5000 10000 l

75-100 1493 1480 2000 2500 3000 3500 5000 10000 100-150 1718 1520 2040 2520 3010 3500 5000 10000 l l

150-200 1878 1600 2110 2600 30'0 3500 5000 1C000 l l

200 300 2485 1710 2210 2680 3100 3500 SOCX; 10000 l

I l

9l Amendment 1 M.k5 I

ABM 2346 ooie Standard Plant REV.A O Table 3A.3-3 Strein Dependent Shear Modulus EFFECTIVE MODULUS REDUCTION FACTOR

• SHEAR STRAIN VP3, VP4 VP6 &

7eg(%) LOGi gyeg UB & VP2 & VPS HR 1.00E-4 -4.0 1.00 1.00 1.00 3.16E-4 3.5 0.99 0.99 1.00 1.00E-3 -3.0 0.94 0.% 0.99 3.16E-3 2.5 0.86 0.905 0.%

1.00E-2 2.0 0.74 0.813 0.90 3.16E 2 -1.5 0.52 0.66 0.82 1.00E 1 1.0 0.40 0.462 0.72 0 0.316 -0.5 0.40 0.40 0.72 1.00 0.0 0.40 0.40 0.72 316 0.5 0.40 0.40 0.72 10.00 1.0 0.40 0.40 0.72

• This ] 'Thas t to be o;glied to the shear modulus at low shear 'ain amplitudes here defined as 1E 4 % to obtian the modulus at higher strain levels.

i l

O Amendment 1 3A.3 6

1 ABWR. us.4e Standard Plant REV.A l l

Table 3A.3-4

)

Strain Dependent Soll Damping I I

EFFECTIVE  % OF CRITICAL I SilEAR STRAIN UB VP6 IIR verg(%) LOG io7eg VP1 TO VPS 1.00E-4 -4.0 0.60 0.60 0.60 l

3.16E-4 -3.5 0.75 0.75 0.65 1.00E-3 3.0 130 1.00 0.80 3.16E-3 2.5 2.59 1.60 1.10 1.00E-2 2.0 4.70 2.60 1.50 1

3.16E-2 1.5 8.40 4.60 2.10 i 1.00E 1 1.0 13.50 7.50 3.00 0316 -0.5 15.00 11.50 3.00 1.00 0.0 15.00 14.60 3.00 l

l 3.16 0.5 15.00 15.00 3.00 l l

10.00 1.0 15 00 15.00 3.00  !

O Amendment 1 3A.37

ABWR 23A62mae Standard Plant REV.A l

Table 3A.3 5 (qJ Average Shear Wave Velocities in Layers For UB Profile with Other Water Tables LAYER WATER TABLE LOCATION BELOW GRADE DEf7H (M/SEC)

M (nominal) 1231 2611 0-9 352 352 9-15 442 442 15-22.5 471 495 22.5-30 498 538 30-45 559 594 45-60 601 629 60-90 783 811 O

LAYER WATER TABLE LOCATION BELOW GRADE DEITil (FT/SEC)

EI BEI 81EI 0-30 1155 1155 30 50 1450 1450 50-75 1545 1625 75-100 16M 1764 100-150 1834 1948 150-200 1970' 2063 200-300 2569 2659 O

Amendment 1 3A.38

l l

23A6100AE Standard Plant anv. A Table 3A.3 6 g

Site Conditions Considered j (Water Table at .61 m (2 ft) depth except noted)  ;

1 l

SOIL '

PROFILE SIIEAR WAVE VFLOCflY PROFILE DEPTil hi mn LIB m m M M M IIB Eli 25.7 (85) UBIDSS -- - -----

VP4D85 --- ----

liRD85 EllD85 45.7(150) UB1D150 VP2D150 VP3D150 VP4D150 VP5D150 VP6D150 - - - - - - - - 1 UB2D150* l UB3D150" 61 (200) UB1D200 -- --- -- - -- - - - - ---- --

l 91.5 (300) UB1D300 -- ---- ---- --- --- --- ---

l l

• Water Table at 12.2 m (40 f6

" Water Table at 25.7 m (85 ft) l Naming System:

mDann where l 1

m : Velocity profile identifier l

Dann Soil deposit depth in feet l

O Amendmerit 1 3A3-9 1

ABWR 2wi=^e Standard Plant REV A l

O l

l l

SHE AR WAVE VELOCITY (ft/sec) 0 1000 2000 3000 4000 , 5000 , 10,000 i i i ri r g

l I I I U B L ~1 - 10 1

L1 50 -

l I I - 20 1

L-- - 30 100 -

l t

i I

E g 40 3 E

{ 150 -

h E o 8 s 3

50 g e

i I

I -- - --

60 200 -

l 70 l

I 250 -

g g

80 l I l VP2 VP3  ! VP4'VPS VP6 HR EH ~

300 0

300 l '

600 I l '

900 4

1200 1500

$'3000 SHEAR WAVE VELOCITY (meter /sec) 88 027 51 O Ngure 3A.3-1 SHEAR WAVE VELOCITY PROFILES CONSIDERED FOR SSI ANALYSES Amendment 1 3A.3-10

• 9>

g llllllllllllllllllllllllllllllllllllll . COMPETENT ROCK j '

~

7 _

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{ i 3!

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i ii_SOFTWOdiAND d55i U kU SdILSli 3000 4!:!

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{ TRANSITION ZONE O ..

• = - - -

2000 . __

m - .-

< SOIL SITES (SANDS, CLAYS, SILTS AND GR AVELLY SOILS) 3:

W 1000 LOOSE TO MEDIUM DENSE SAND; SOFT AND MEDIUM STIFF CLAY I I I I I 0

O 50 100 150 200 250 300 DEPTH (ft) @g2:

f= Figure 3A.3-2 RANGE OF SHEAR WAVE VELOCITIES FOR NUCLEAR POWER PLANT SITES IN HIGH SEISMIC AREAS 88 027-52

]

O O O

M 23A6100AE i Standard Plant REV.A O l 1.0 1

• m 3 3 0.8 -

< J E ROCK (PROFILE HR)

$ o@ PROFILE VP6 H 2 o'* ~

< e PROFILES

$ $ VP3, VP 4, VP5

$ 5 o 2 0.4 -

2 2 .

$ R PROFILES 4, y 0.2 -

yp2, UB 0

d 10 10-3 10-2 10'1 1 10 SHE AR STR AIN,7 (%)

30 (D

L> 25 -

O

% 20 -

6 P

I 15 - PROFILES VP2 e THROUGH VP5.

3 UB

< 10 -

o PROFILE VP6 5 -

ROCK (PROFILE HR) 0 10 d

10 4 10-2 10-1 1 10 SHEAR STRAIN,7 (%)88-027 53 Figure 3A.3-3 VARIATION OF SHEAR MODULUS AND DAMPING RATIO WITH O SHEAR STRAIN USED IN ANALYSES Amendment 1 3A.k12

ABM 23461004e Standard Plant REV.A SECTION 3A,4 O

CONTENTS Section TRlt East 3A.4.1 Innut Motion 3A.4-1 3A.4.2 Damnian Values 3A.4-1 ILLUSTRATIONS Figure Iltle Eagt 3A.4-1 Control Motion- H1 3A.4-2 3A.4 2 Control hiotion--H2 3A.4 3 3A.4-3 Contiol biotion--Vertical (V) 3A.4-4 3A.4-4 TH (Velocity)- H1--Horizontal Control hiotion (0.15g) 3A.4-5 O 3^ 4-5 T" <v iecitx) "2 -"erize t i ce trei"e<> co 258 ) 3^ 4 6 3A.4-6 TH (Velocity) -Vertical Control Motion (0.15g) 3A.4 7 i

3A.4-7 TH (Displacement) H1 Horizontal Control Motion (0.15g) 3A.4-8 l 4

3A.4 8 TH (Displacement).H2 Horizontal Control Motion (0.15g) 3A.4 9 3A.4-9 TH (Displacement) Vertical Control Motion (0.15g) 3A.410 0 3A.4-il Amendment 1

1 1

ABM 23A61MAE Standard Plant REV.A J l

3A.41NPUT MOTION AND DAMPING soil damping values resulting from the  ;

q V

VALUES free field site response analysis as described in Subsection 3A.6. The strain dependent soil 3A.4.1 Input Motion damping versus strain is described in Subsection 3A.3.2 The time-history method is used in performing the seismic soil structure interaction analysis.

~ Earthquake (input or control) motion in the form of synthetic acceleration time histories is generated r.s described in Subsection 3.7.1.2 for all three components designated as Hi , H2.

and V. The Hi and H2 are the two horizontal components mutually perpendicular to each other.

In the SSI analyses, Hi and H2 are used in the horizontal Y (90 ) and X (00) directions, respectively. The V is the vertical component.

Each time history component has a duration of 22 seconds and is digitized at 0.01 see time intervals. The acceleration time histories which are developed from the design response spectra per Regulatory Guide 1.60 and normalized to 0.15g OBE are shown in Figures 3A.3-1 through 3A.3 3.

The corresponding velocity and displacement time Ms;ories are abown in Figures 3A.4 4 through 3A.4.9. Subsution 3.7.1.2 provides' the response e esi n response spe tra.

The input motion (or control motion) is defined at the finished grade in the free field.

The motion is assumed to be generated by vertically propagating plane seismic shear waves for the horizontal components and compression for the vertica.1 component waves.

3A.4.2 DampingValues The structural components damping values used '

in the seismic analysis are in accordance with those specified in Regulatory Guide 1.61. These values for both the OBE and the SSE are summarized in Table 3.71. When developing seismic system design data for the SSE, the higher damping values of Regulatory Guide 1.61 are not used; the data is obtained by doubling the OBE values which are based on the lower damping values.

Thp soil material damping values used for the SSI analysis are based on the strain compatible  ;

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ABM BA61MAE Standard Plant REV.A Q SECTION 3A.5 CONTENTS j l

i Section M East 3A.5.1 Introduction 3A.51 3A.5.2 Comnlex Response Method 3A.5-1 3A.53 Methodolony and Analysis Procedure 3A.5-3 ILLUSTRATIONS Figure Title Eage  ;

3A.51 Substructuring ofInteraction Model 3A.5-6 i

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ABM 23461oaxe Standard Plant REV.A 3A.5 SOII STRUCTUREINTERACTION soil property variations, but also include the ANALYSIS METHOD effects of different SSI approaches in (3j predicting soil structural response.

3A.5.1 Introduction No separate SSI analysis is performed for the Nuclear Island structures are massive SSE input; the design SSE loads are taken as two structures typically embedded at a considerable times the design OBE loads since the SSE input depth in a soil deposit. An important aspect in is twice the OBE ir put and the higher structural the seismic design of these structures is the damping ratios allowed for the SSE are not used.

evaluation of the dynamic ideraction between the structures and the soil. Such interaction In performing SSI analysis using the finite effects may significantly affect the response of element method, the detailed structural models the structures as wei'. as the equipment systems, described in Subsection 3.7.2 are coupled with the soil model. Structural responses in terms The problem of soil structure interaction has of accelerations, forces, moments, and stresses been the subject of considerable research in are computed directly. Floor response spectra recent years and a number of interaction models are obtained from the calculated response have been developed and are in use in the acceleration time histories (see Subsection industry. The distinction between two generally 3.7.2.5). This effectively eliminates the need recognized approaches, namely, the finite element for a second step structural response analysis (direct solution) and half-space (substructur- in which the fixed base structural model is ing), becomes blurred in using the current SSI subjected to the base motions resulting from the analysis approaches which include both aspects of first step SSI analysis. The direct solution the modeling techniques. also has an added advantage that the structural response to all components of base motion The approach used to obtain seismic design including rocking motion components for an envelopes in this appendix is based on the embedded foundation is automatically accounted finite-element snethod using substructuring for in the solution.

(V~T technique. The newly developed, two and three dimensional (2 D and 3 D), linear finite The SSI analyses for the three directional element computer program SASSI (Ref. 4) developed earthquake components are performed separately, at the University of California, Berkeley, by a The maximum co directional responses to each of team working under the technical direction of the three earthquake components are combined Prof. J Lysmer is used. The program uses finite using the SRSS method to obtain the combined elements with complex moduli for modeling the maximum structural response in each selected structure and foundation and adopts the flexible degree of freedom of interest.

volume method of substructuring and the frequency domain cornplex response method of analysis In compliance with the requirements in SRP procedure. Detailed descriptions of the complex 3.7.2, an alternate approach based on continuem response method and SSI methodology are presented mechanics is also used for SSI analyses for 3 in Subsections 3A.5.2 and 3A.S.3. representative sites. Attachment A to this appendix presents the methodology of this The SASSI program is used to ar.alyze all 14 approach and the analysis conditions consid-site conditions defined in Subsection 3A.3.4 for cred. Effects of this alternate approach are 42 SSI cases as described in Subsection 3A.7 for included in enveloping the results of finite-the OBE condition. The enveloped results which element method for developing the design loads, also include consid: rations given to the effect of a different SSI approach in forming the design 3A.5.2 Complex Response Method OBE loads. The resulting design loads therefore not only constitute an adequate consideration of For the dynamic analysis of a linear damped G

b Amendment 1 1A.51 I

MM 23A6100AE Standard Plant RER A system, the equation of motion in the time domain with complex coefficient matrix which can be at time t may be written as in the following solved for the complex response vector M' for &

matrix form: each frequency. The solution for U.* for a unit N

,, , applied force, i.e., E' = 1, is the transfer MM+DM+EM=E (3A-8) function.

where d, $. and M. are respectively the For systems with material damping, the first acceleration, velocity and displacement vector at two terms of Eq. 3A 12 can be combined into a time t; E and M are the assembled total stiffness complex stiffness matrix K' and mass matrices; Q is the damping matrix and E is the dynamic load vector acting on the system E' = E + i wD (3A-13) at time t. For an SSI system for which the soil material damping in characterized by the constant For systems with constant hysteresis material hysteresis damping model and the foundation damping, E' can be formed directly using complex impedances are frequency dependent, the equation shear modul,i, E*3, and complex constrained of motion is most conveniently formed and solved moduli, E in the frequency domain using the complex equations. p, defined by the following response method. The application of the complex response method involves two solution steps: E,3 steady state response analysis to obtain the = E3 (12$3 2+2iS3 /1. g23) frequency response functions, and 2) Fourier (3A 14) analysis of the input and convolution to determine the transient response. E,

= Ep(12S p2+ 2iSp/1 - $2 p)

For the steady state harmonic excitation at (3A 15) the circular frequency w, the exciting force vec:or is written in the form: Where E3 and E are the real numbers E (t) = E* eIWI (3A 9) corresponding to the skear and constrained moduli of the material, respectively, and S3 and g

p are the critical damping ratios associated where E* is a complex constant vector with S waves and P-waves, respectively. For representing the applied force amplitude and the praetiea1 applications, $3 and p p are phase angle. usually taken to be equal, i.e, 3

=

Sp =

S, in which case the Poisson's ratio is a real The steady state displacement response is also number, harmonic with the frequency o :nd can accordingly be written Using the transfer function solution obtained from the complex response method, the transient M (t) = M* eiet (3A-10) response of the system due to transient excitations such as carthquake motions or applied Substitution of Eqs. 3A 9 and 3A-10 into Eq. dynamic loads, can be analyzed by decomposing the 3A 8 yields the matrix equation input excitation into Fourier components with period N At, where at is the time C' M* = E' (3A 11) interval of digitization of excitation and N is the number of digitized points. Using the where C' is the frequency dependent complex discrete representation, the discretized load dynamic stiffness matrix in the form function E (t) is defined by the series C' = E + i wD w2 M (3A 12) Eg = E(K at), K = 1,2,...,N (3A 16)

Equation 3A-11 is a system of linear equations O

Amendment 1 3A.5-2

1 ABM 2346ioars Standard Plant REV.A which can be expanded into a discrete Fourier the inverse Fast Fourier Transform algorithm. In-p series in the form principle, the system of complex linear equations

() E(t) = Re E N/2 Ej e wt3 defined by Eq. 3A-11 has to be solved for each of the 1+N/2 frequencies shown in Eq. 3A-18. For the s=0 (3A-17) excitation defined by a real time function, the ,

where e s are the discrete circular highest frequency corresponds to 1/(2at). This '

frequeneies defined by: is called the Nyquist (or folding) frequency. The choice of at usually results in the Nyquist us = 223 frequency which is much higher than the highest N at (3A-18) frequency of interest. The series in Eq. 3A 18 can thus be truncated at some lower frequency and fj is complex amplitude defined by number, scut. The corresponding cutoff N1 frequency Ej = 1 I Ek exP( ios kat);

N K=0 s = 0, s = E fcut = scut, s = 0,..., scut <d 2 N at 2 (3A-19a) (3A-21)

N-1 Ej = 2 E Ek exp(-iw3 kat); and only scut solution to Eq. 3A-11 are N K=0 lis ands E required. For seismic analysis, scut is still a 2 large number (typically of the order of 500). In (3A 19b) practice Eq. 3A-11 is solved for only a limited number of frequencies in the frequency range of The numerical operations in Eqs 3A-19 are interest. The solutions for the remaining performed using the efficient Fast Fourier frequencies are obtained by interpolation in the Transform (FFT) algorithm. This technique frequency domain, requires that N be a power of 2 which can always p

V be achieved by adding trailing zeros to the 3A.5.3 Methodology and Analysis Procedure discrete excitation time series. For damped systems these trailing zeros also serve as a The flexible volume substructuring method used quiet zone which allows the transient response in SASSI considers the structure (s) and the motions to die out at the end of the duration to foundation (s) to be partitioned as shown in Figure avoid cyclic overlapping error in the discrete 3A.5 1. In this partitioning, the structure Fourier transform procedure. (Figure 3A.51) consists of the superstructure plus the basement minus the excavated soil. The Once the El values are obtained, the Fourier foundation (Figure. 3A.51), consists of the transform of the transient response can be original site, i.e., the site without soil obtained by convolution of the transfer function excavation for the structure basement.

and the Fourier coefficient, Ei, giving the Interaction between the structure and the complex frequency response function Mi for each foundation occurs at all basement nodes. For discrete frequency w3. seismic analysis, the equation of motion for substructure C in Fig. 3A.5-1 can be written in Finally the transient response in the time the frequency domain as follows:

domain U(t) can be obtained by the Fourier .

synthesis, i.e., Cj3 .Cli P

r O '

U(t)=Re N/2 I UjeiWt s

( Q3[

< g [,

s=0 (3A 20) ) '

Cis CkCff+5*f ( 1 *t s XI YU Numerically, this op: ration is performed using -

(3A 22) 1 Amendment 1 3A 5 3 l

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ABM 23A6100AE Stupdard Plant REV A where the submatrices .C* are the complex dynamic soil column deconvolution analysis to obtain stiffness matrices and defined by the following the equivalent linear soil properties com- A relation: patible with seismic strains induced in the W free-field. These properties are then used C* = E' w2 M (3A-23) for the SASSI site response analysis. The free field site response analysis is where E' is the complex stiffness matrix and M is described in Subsection 3A.5.

the mass matrix, and w is the excitation circular frequency. The subscripts used in Eq. (2) Solve the impedance problem. This step 3A 22 are defined in Figure 3A 15. The matrix involves determining the matrix Xt which Xt is the impedance computed for all the can be obtained efficiantly from a series of interacting nodes and .U's and Et are the 2 D plane strain or 3 D axisymmetric complex total displacements of the superstructure solution of substructure (b) in Figure ,

and foundation nodes, respectively. 3 A .5 1. The halfspace simulation, if l requested in Step 1, will be automatically l The complex modulus method of material included in this step of analysis as well.

representation described in the previous subsection is used to compute the complex (3) Form the load vector. The seismic excita-stiffness of the elements allowing for variation tion load vector in Eq. 3A 22 can be comput-of material damping in each element. ed from the product of solutions in Steps 1 and 2.

Based on the above equation of motion, the '

solution to the harmonic soil-structure (4) Form the complex dynamic stiffness matrix.

interaction problem can be achieved in the This step involves forming the coefficient I following five steps: matrix C' in the left-hand side of Eq. 3A-

22. The superstructure is modeled using fb (1) Solve the site response problem. This step nite elements from the finite element libra-involves determining the free field dis-placement amplitude Uf within the exca-ry of SASSI. The foundation is modeled us-ing two sets of elements. The first set is g

vated volume. The site is modeled by hori- used to represent excavated soil in the zontal soil layers resting on rigid base or basement volume. Usually brick elements are clastic halfspace. The halfspace at the used in 3 D analysis and four node plane bottom boundary is modeled using the vari- strain elements are used in 2-D analysis, able depth method in which an extra layer of The second set of elements are used to model soil with a total depth of 1.5 A (where A the embedded portion of the supe structrue is the shear wave length in the halfspace) is such as the basemat or side walls of the added to the soil profile. Thus, the added structural foundation below the grade, soil layer depth varies with frequency. Similarly, brick, and plane strain elements Furthermore, the added soil column in and/or plate elements are used for this subdivided into n sublayers. A value of n=10 purpose.

is adequate and is used ir the analyses.

In the finite element discretization of the In addition to variable depth soil layer, the basement volume, the element side dimensions halfspace simulation also uses viscous are selected to be smaller than or equal to dashpots in the horizontal and vertical A/S where A is the shortest shear wave directions at the base of the added soil 1 ngth of interest in the soil layer. Thus, layer. the rnaximum side dimension of the elements in the foundation model are governed by the SASSI is a linear analysis program. The shear wave velocity of the soil layer and iteration on strain dependent soil properties highest frequency of interest in the cannot be directly incorporated in the analysis.

analysis. It is therefore, necessary first to perform the one dimensional free field in selecting the highest frequency of g

Amendment 1 3AM

I MN 23A6100AE Standard Plant anv,A j l

analysis, the frequency content of input motion and fixed base modal frequencies and

{3, modal mass participation factor of the  !

superstructure are considered. Generally, j it is adequate to solve the SSI problem for l frequencies up to the fixed base modal '

frequency whose fixed base cumulative modal mass amounts to 90% of the total mas .

Furthermore, if the nodal points for which responses are to be computed are located on the subcomponents of the superstructure model, then the local modal frequencies corresponding to these subcomponents should be considered in selecting the highest frequesey of analysis.

(5) Solve the structural problem. This step involves solving the linear equation of Eq.

3A 22 for each frequency.

For seismic excitations, the above steps are performed at several selected frequencies. These frequencies are usually selected to be close to major fixed base modal frequencies of the struc-ture and those of its subcomponents that are of interest. An effective interpolation scheme is e used to compute the response at all frequencies h required by the Fourier Transform Techniques to compute the response in time domain.

O Amendment 1 3A.5-5

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MM 23A6100AE Standard Plant nev. A i i

1 SECTION 3A.6 TABLES-l Table 1111e East 3A.6-1 OBE Free-Field Site Response Results for All i Soil Profiles (Average Properties) 3A.6-2 l I

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ABM 22^61oo^n Standard Plant REV.A 3A.6 Free-Field Site Response for all soil layers of the soit deposit, and the Analysis natural frequency of the soil column. These

) frequencies vary from 1.5 Hz to 15 Hz reflecting The behavior of soil is nonlinear under the wide range of dynamic soil properties and seismic excitation. The soil nenlinearity can be solid deposit configurations considered in this conveniently separated into primary and secondary appendix.

nonlinearities. The primary nonlinearity is associated with the state of deformations induced For the vertical earthquake component, the by the free field ground motion. The secondary P-wave velocity corresponding to the calculated nonlinearity is attributed to the SSI effects. strain compatible shear modulus (or shear wave This secondary effect on structural response is velocity) for each layer is determined following usually not significant and is neglected in the the steps described in Subsection 3A.3.3. The appendix. soil damping associated with the P wave is considered to be the same as the final iterated The soil primary nonlinearity is approximately damping associated with the shear wave, accounted for by an iterative process to obtain equivalent linear soil properties of shear modulus and material damping corresponding to effective strains developed in the free field.

The strain compatible properties are then assigned to be the linear soil properties in the SSI iialysis. These properties also determine the maximum element size of the foundation model using the finite-element approach.

The free field site response analysis is performed using computer program SHAKE (Ref. 5) q which uses the principle of one dimensional b propagation of shear waves in the vertical direction for a system of homogeneous, visco clastic soil iayers. An equivalent linear method is used to compute strain compatible shear modulus and material damping for each soil layer. All sites except the EH profile defined in Table 3A 6 are analyzed using the H t horizontal component input motion at the ground surface. The analyses are performed up to maximum frequency of 20 Hz which is considered to be adequate for the purpose of obtaining strain-compatible properties. The free field site response analysis is not performed for the i EH profile since the next hardest profile HR already shows that the shear modulus and material damping corresponding to the induced strain level are essentialy the same as their initial values  !

at low strains. Therefore, the SSI analyses for i the HR and EH profiles are performed using the l initial modulus and a conservative material damping of 0.1% (undamped).

Table 3A.61 summarizes the results of OBE free field response analyses in terms of the strain compatible shear wave velocity, averaged Amendment 1 3A.61 l

l l

23A6100AE S(ADd ard Plant Rev.A Table 3A.61 OBE i ree ,"' eld Site Response Results for All Su ! cofiles (Average Properties)

SOII. AVERAGE SHEAR PROFILE WAVE VELOCTIY FREQUENCY m m/sec (n/sec1 tcts1 UB1D85 323 (1059) 3.37 UB1D150 366 (1201) 2.22 UB1D200 392 (1286) 1.80 UB1D300 482 (1581) 1.49 UB2D150 414 (1359) 2.47 UB3D150 437 (1434) 2.65 VP2D150 365 (11 % ) 2.05 VP3D150 552 (1811) 3.11 O ve4o85 216 (2>49) 6 98 VP4D150 715 (2345) 3.99 VP5D150 870 (2855) 4.81 VP6D150 1055 (3459) 5.82 HRD85 1521 (4989) 14.67 EHD85 3G48 (10000)' 29.4

• Initial velocity corresponding to low strain level.

O Amendment 1 3A.6 2

23A6100AE Rinndard Plant REV.A I

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i 3A.7-1 Summary of SSI Cases Considered 3A.7 2 e

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ABM 2346imic Standard Plant arv.A 3A.7 SOII STRUCTUREINTERACTION vertical direction. For HRD85 site only p ANALYSIS CASES horizontal analysis is performed. Altogether a V total of 7 cases are considered for this group To establish seismic design envelopes of the of 2 D SSI analyses, reactor building situated at the fourteen generic sites defined in Subsection 3A.3, the SSI To account for the three dimensional (3 D) analyses are performed for a total of forty two SSI effectt, the reactor building by itself is cases as summarized in Table 3A.71. The further analyzed for four representative sites analysis cases considered can be categorized into ranging from soft soil to hard rock using the three groups. 3-D option of the SASSI program. Three cases in each of the two horizontal directions and two Group AmMis Conditions cases in the vertical direction are selected among sites UB1D150, VP3D150, VP5D150, and A 2 D cases for the reactor HRD85 as shown in Table 3A.71. This amounts to building alone a total of eight 3 D analpis cases.

B 2 D cases for the reactor building with the control It is evident from the above that a large and turbine buildings number of analysis cases are considered for a C 3-D cases for the reactor broad range of site conditions. All forty two building alone cases (27 for Group A,7 for Group B, and 8 for Group C) are analyzed for the OBE excitation.

The group A cases, 27 case in total, are Their analysis results are enveloped to analyzed using the 2 D option of the SASSI establish the OBE design loads for the reactor program for all fourteen site conditions. Since building of the ABWR standard plant. The SSE the dynamic characteristics of the reactor design loads are taken to be twice of the building in the two horizontal principal enseloped OBE loads for the seismic system of directions are very similar, all fourteen sites the reactor building structure.

O V

are analyzed in the 001800 (X) direction and only selected sites are needed in the 900 2700 (Y) direction. The sites analyzed in the Y direction are chosen to be UB1D150, VP3D150, VP5D150, HRD85, and EHDSS which are 1 the representative soft, medium, stiff, and hard soil conditions. These five sites are also '

analyzed in the vertwal direction so that the range of soil stiffness effect on the reactor building vertical response is adequately j accounted for. In addition, three more sites 1 (UB1D200, UB2D150, and UB3D150) are I investigated in the vertical direction to 1 evaluate the impact of soil deposit depth and ground water table variations.

As shown in Table 3A.71, the sites UB1D150, VP3D150, VP5D150, and HRD85 are also analyzed in 2 D considering the control and turbine buildings. The purpose of these analyses (group B cases) is to evaluate the effects of structure to structure interaction on the reactor building response for representative soft, medium, stiff, and hard sites. For each site except the HRD'S site, the anahses are performed in the horizontal 00 1800 direction and

\

Amendment 1 3A.71

1 ABM u^62 min Standard Plant REV.A ,

Table 3A.71 0- Summary of SSI Caws Considered A. 2 D for R/B Alone Case ID SITE ID 00 1800 900 2700 Vertical UB1D150 C2X C2Y C2V VP2D150 C3X - ---

VP3D150 C4X C4Y C4V VP4D150 CSX - -

VP5D150 C6X C6Y C6V VP6D150 C7X -- ---

UD1 DSS C8X --- --.

VP4D85 C9X --- -

HRD85 C10X C10Y C10V UB1D200 C11X --- C11V UB1D300 C13X -- -

UB2D150 C14X --- C14V l UB2D150 C15X -- C15V EHD85 C16X C16Y C16V O B. 2 D For R/B + C/B + T/B Case ID SITE ID 00 1800 900-2700 Vertical l

US1D150 C21X - C21V VP3D150 C4-1X - C4-1V VPSD150 C6-1X -- C61V l HRD85 C10-1X --- --

C. 3-D For R/B Alone Case ID SITE ID 00 1800 900-2700 Vertical UB1D150 T2X --- -.-

VP3D150 - T4Y T4V i

VPSD150 T6X T6Y --

l HRD85 T10X T10Y T10V NOTE: See Table 3A 3-6 for Site ID Descriptions i Amendment 1 3A 7-2 ,

1

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ABM 22461oaxe i

Standard Plant REV,A Q SECTION 3A.8 CONTENTS Section M Eage  ;

3A.8.1 Structural Models 3A.81 1

3A.8.2 Foundation Models 3A.8-1 3A.8.2.1 2 Dimensional Models 3AE-1 3A.8.2.2 3-Dimensional Models 3A.8-2 ILLUSTRATIONS Figure 1111e Eage 3A.81 Reactor Buuding Model 3A.8 3 3AE 2 RPV and Internals Model 3A.8-4 3A.8-3 Control Building Model 3A.8-5 3AE-4 Turbine Building Model 3A.8-6 3AE 5 The Excavated Soil Elements of Reactor Buuding Foundation Model(X-Direction) 3A.8-7 3A.8-6 Reactor Buuding Foundation Modelin X Direction 3A.8-8 3AE 7 Excavation Soil Model--Control Building 3A.8-9 3A.8-8 Foundation Model of Control Building 3A_8-10 3A.8 9 Excavated Soil Model-Turbine Buuding 3AS-11 3A.810 Reactor Building Foundation Modelin Y Direction 3A.812 3A.811 3-D SASSI Quarter Model of Reactor Building 3A.813 3A.8-12 Nodal Points: EL 15.86 m 3A.814 O

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Figure Tuls Eagt 3A.8-13 Nodal Points: EL 113 m 3A.8-15 3A.8-14 Nodal Points: EL 8.25 m 3A.816 3A.8-15 Nodal Points: EL 5.2 m 3A.8-17 3A.816 Nodal Points: EL -2,15 m 3A.8-18 3A.8 Nodal Points: EL 0.90 m 3A.8-19 3A.8-18 Nodal Points: EL 3.95 m 3A.8-20 3A.8-19 Nodal Points: EL 7.0 m 3A.8 21 3A.8-20 Plate Elements of the Base Slab 3A.8-22 3A.8 21 Plate Elements of the Side Wall (X-27.25 m) 3A.8-23 3A.8 22 Plate Elements of the Side Wall (Y=28.749 m) 3A.8-24 g

3A.8 23 Connection of the Main Stick to the Side Walls 3A.8-25 i

O 3A.8 iii Amendment 1

ABM 2miooAn Standard Plant RFV A s

3A.8 Analysis Models 3A.8.2 Foundation Models V 3A.8.1 Structural Models 3A12.1 2 D Foundation Models The structural model of the reactor building This set of rnodels consists of the models complex including the reactor pressure vessel and for reactor building foundation for X and internal components as described in Subsection Y directions. The model developed for 3.7.2.1.5.1.1 is used in SSI analysis. The model X-direction is also applicable for the vertical Figures 3.7-28 and 3.7 29 are also shown in this analysis. The model for X direction consists of appendix as Figures 3A.81 and 3A.8 2 for easy the model for reactor building by itself and the reference, models for the control and turbine buildings.

The control building model is described in The finite element mesh in the X Z plane Subsection 3.7.2.1.5.1.2.1. Figure 3A.8 3 shows representing the excavated soil elements for the the model used in SSI analysis. The model has 8 reactor building foundation is shown in Figure lumped masses connected by massless beam 3A.8 5. This model consists of 162 four node elements, plane strain elements. The strain. compatible equivalent linear soil properties obtained form The turbine building model considered is shown the free field analysis are used for the soil in Figure 3A.8-4 It consists of two concentric finite elements. ,

lumped mass beam sticks representing the building structures and the turbine generator pedestal. The foundation model of the reactor building This simple representation is sufficient since below grade is shown in Figure 3A.8 6. This the turbine building response is of no concern in model consists of 50 four node plane strain this a p p e n dix but the effeet cf elements with the properties of concrete to structure to structure interaction on the reactor represent the base slab and the side walls of building response is. This effect can be the reactor building. The mass of the basemat O adequately accounted for by maintaining the turbine building foundation configuration and the is considered by using the concrete weight density for the base sIab cIements.

building total mass as well as the building Consequently, the lumped mass at the base of fundamental frequency in the simplified model structural model is modified to subtract the representation, basemat mass. The side wall elements are massless since the mass of all structural In 2 D SSI analyses the structural model components is already included in the structural properties (stiffness and mass) are converted model. The structural model is connected to the into values corresponding to per unit foundation base slab by a rigid beam connecting Node 88 to depth (dimension in the third direction) to the Node 229 at the middle of the slab (see ,

maintain compatibility with the 2 D foundation Figure 3A.8 6). l model. l To properly transfer the rotation of the in 3 D SSI analyses for the reactor building, stick model to the base slab (and vice versa), a the properties of the outer walls represented by set of rigid beams are placed at the middle of the outer stick below the grade in Figure 3A.81 the slab connecting Nodes 220 through 238. The are adjusted to subtract the properties of the stick representing the outer walls of the four embedded outer walls. These walls, as reactor building is conneued to the side walls described in the following subsection, are by a set of rigid springs in horizontal modeled as part of the finite element foundation directions to reflect the direct contact model. Because of two planes of symmetry, the condition of the outside wall with the soil, model properties and boundary conditions are The finite element mesh size selected for the adjusted to correspond to the quarter model, model is capable of transmitting seismic waves O

Amendment 1 3A B 1 l

ABM uAatman Standard Plant nry A with frequencies up to 17 cps for UB soil profile 3A.8.2.2 3 D Foundation Model and beyond 25 cps for the stiffer soil sit:s for horizontal analysis, in the vertical direction By taking advantage of two planes of g

this model is adequate for frequencies above 33 symmetry, the quarter model of reactor building cps for all soil sites considered, foundation is constructed as shown in Figure 3A.811. The nodal points of the finite element For the purpose of evaluating the mesh at different elevations are shown in structure to structure effect on the reactor Figures 3A.812 through 3A.819. A total of 343 building response, the foundation of the control brick elements are used for modeling the building is assumed to be rigid in this excavated soil in the embedded foundation appendix. volume. One column of the soil is modeled (see Figure 3A.812) along toe height of the wall The finite element mesh for the excavated soil outside of the excavated soit volume for the for the control building foundation is shown in purpose of computing the dynamic soil pressure Figure 3A.8 7. In this model, the embedment on the wall. The base slab is modeled with 49 depth is 12.2 m (40 ft) and the width of the plate elements with the properties of concrete foundation is 16 m (52.5 ft). The foundation (see Figure 3A.8 20). The exterior walls of the model and the connection of the building stick building are also modeled with plate elements as model to the foundation are shown in Figure shown in Figures 3A.8 21 and 3A.8 22.

3A.8 8. As shown in the figure, the base of the stick model (Figure 3A.8-3) is connected to the The outer stick of the reactor building model base slab by a set of rigid beams to maintain (Figure 3A.81) is connected to the base slab rotation compatibility. and the side walls with rigid beam as shown in Figure 3A.8 23. Note that the layout of those The turbine building foundation is 106 to (348 rigid beams as shown is for the X direction ft) wide with embedment depth of 18.3 m (60 ft). analysis. This is consistent with the beam The foundation is assumed to be rigid for the stick model assumptions that only the walls purpose of evaluating the structure to structure parallel to excitation direction provide shear interaction effect on the reactor building, rigidity and portion of cross walls only Figure 3A.8 9 shows the finite element mesh for contributes additional flexural rigidity, the excavated soil eleme;nts. Different layouts of rigid beams are used for the Y and vertical directions of analysis. The The finite element :nesh for tr e foundation is foundation:model developed is adequate for the same as that for she excavated soilahown in frequencies of analysis up to 17 cps in the Figure 3A.8 9 except that the nodes on the horizontal direction and for the softest soil foundation base are further linked together by case (UB). This frequency increases beyond 25 rigid beams and the Node 508 is linked to the eps for the medium soil cases VP3. For stiffer structural model shown in Figure 3A 8 4. sites (VP5 and HR) a foundation model with coarser element size is used to reduce the The 2-D reactor building foundation model for computer cost. The element size is chose to l Y direction is similar to the X direction model. ensure that its frequency transmissibility is l The foundation width is 56 m (184 ft) and the above 25 cps for the horizontal excitation. The foundation depth in the third dimension is 59 a coarser model is capable of transmitting l (194 ft) for the Y direction. The connection of frequenaies above 33 cps for the vertical l the stick model to the base and side walls are excitation for all site conditions and, thus, is similar to those of the X direction model. The used for all vertical 3 D SSI analyses.

Y direction foundation model is shown in Figure 3 A.8 10. For this direction, no consideration for structure to structure interaction effect is l

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• H3 He 957 956 959 960 Mt M2 857 858 859 860 841 862 843 864 757 758 759 760 781 782 763 764 657 654 659 660 641 662 663 664 557 558 559 560 541 54 2 563 564 457 454 459 460 461 442 463 464 357 358 359 360 341 342 343 364 15J6 m 257 258 259 260 261 262 263 2 64 4.107 m SJ14m 12.321 m 16.428 m 20.538 m 24442 m 28.748 m 88-027 85 Figure 3A.8-22 PLATE ELEMFNTS OF THE SIDE WALL (Y = 28.749 m)

Amendment 1 3A 8-24

l ABWR ux6w0au Standard Plant REV.A

- . . /y '

AsBR 5:

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Amendment 1 3A.8 25

23A6100AE l Standard Plant REV.A 3A.9 ANALYSIS RESULTS The analysis results of the cases defined in Subsection 3A 7 are enveloped to establish all site envelope loads. These enveloped loads ,

obtained using the finite element method are I further evaluated against the enveloped results of the cases analyzed by an alternate approach based on continuum mechanics. Conservative design consideration is then given to establish  !

the seismic design loads which include the effects of a different SSI approach on the structural response predicted by the finite element method. ~

j The analysis results in the form of I site enveloped seismic design loads at key l locations in the reactor building complex are presented in Appendix 3G.

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Amendment 1 3A.91

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3A.10 REFERENCES

1. General Electric Company BWR/6 238 Standard Safety Analysis Report (GESSAR), Docket No.

STN 50-447, November 7,1975.

2. General Electric Company GESSAR II BWR/6 Nuclear Island Design (2M7007), March 1980.

3. NUREG/CR 1161, Recommended Revisions to Nuclear Regulatory Commission Seismic Design Criteria, May 1980.

4. Lysmer, J., Tabatabaie Raissi, M., Tajirian, F., Vabdani, S., and Ostandan, F., SASSI--A .

System for Analysl2 of Soll. Structure l Interaction, Report No. UCB/GT/8102, Geotechnical Engineering, University of California, Berkeley, CA, April,1981.

5. Schnabel, P.B., Lysmer, J., and Seed, H.B.,

SHAKE -A Computer Program for Earthquake Response Analysis of Horizontally Layered Sites, Report No. EERC 7212, Earthquake Engineering Research Center, University of California, Berkeley, CA,1972.

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O Amendment 1 3A.101

MM 23A6100AE Standard Plant anv A g ATTACHMENTA ILLUSTRATIONS Figure Illlt East 3AA 1 Schematic Representation of the Elements of the Substructure Approach in Analyzing Soil Structure Interaction 3AA 7 3AA 2 ConceptualIllustration of Hybrid Method in CLASSI/ASD Using Substructure Deletion and Boundary Element Method 3AA 8 3AA-3 Definition of Surfaces and Volumes in Hybrid Method CLASS 1/ASD 3AA 9 3AA-4 Top Foundation Surface Subregion Mesh with Dimensions Normalized to Corresponding X and Y Side Dimensions 3AA 10 3AA 5 Embedded Foundation Surface Subregion Mesh with Dimensions Normalized to Corresponding X and Y Side Dimensions 3AA 11 3AA 6 Boundary Element Representation of a Typical Embedded Rectangular Foundation 3AA 12 l

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23A6100AE Standard Plant any. A l

ATTACHMENT A O To APPENDIX 3A SSI ANALYSIS USING ALTERNATE APPROACH i

3AA.1 INTRODUCTION l 1

The Standard Review Plan Section 3.7.2 requires that two approaches be considered in soil structure interaction analysis. As described in Appendix 3A, the finite-element method is used to establish site enveloped seismic design loads which include conservative design considerations of effects of an alternate approach. The alternate approach considered is based on continuum mechanics using the computer program CLASSI/ASD (Ref.1) developed by ASD International, Inc., San Francisco, CA. This is 1 a linear analysis program capable of handling '

three-dimensional soil / multiple embedded foundation interaction problem. This attachment to appendix 3A presents the detailed methodology descriptions and analysis conditions considered.

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Amendment 1 3AA1

-4 ABM , 23A6100AE Standard Plant REV.A 3AA.2 METHODOLOGY In what follows the step by step analysis O procedures in the calculations of impedance V Cl. ASSI/ASD uses the frequency dependent functions, scattering matrix, soil-structure three dimensional Continuum Impedance Approach to interactions, and the interactions between model the essentially semi infinite soil medium, adjacent foundations through foundation soil are and a substructure approach which subdivides described.

complicated soil structure interaction problems into more manageable parts. Figure 3AA 1 shows (1) Calculation ofImpedance Matrix in a schematic way the three major steps of the substructure approach. The impedance matrix describes the frequency dependent force displacement The half space is first analyzed using ao relationship of the foundation (s). Figure 3AA-3 innovative Hybrid Integral Equation Formulation shows the foundation boundary surfaces of a (or sometimes referred as Substructure Deletion typical embedded rectangular foundation, in Formulation) which combines the source CLASSI/ASD the foundation boundary surfaces of distribution and the mixed distribution of Sg (top surface) and Se (embeddeJ surfaces) singularities respectively over the top are discretized by equal numbers of elements, foundation surface and the embedded foundation Nt and N e The unknowns are assumed to be surface. Briefly, on the top foundation surface at the center of each element, anJ to have a where the geometry is simpler, the free surface constant value over the element. Integrations Green's function which satisfies the free surface associated with the Green's functions over the boundary condition of en uniformly layered elements areas are carried out by using the visco clastic half space is used in constructing 4 point Gaussian quadra,ture formula. The the source integral equation relating the surface harmonic time factor eNt will be omitted displacements and tractions. hereafter for clarity. The resulting sets of linear algebraic equations for the top surface On the soil volume occupied by the embedded and the embedded surfaces of the foundation can foundation where the geometry can be represented be written respectively, in matrix form, as by rectangular blocks, the Stoke's fundamental solutions (the full space 3 D Green's functions) ft = kh ut' of Elastodynamics in an infinite clastic space (3AA-1) are used in constructing the mixed integral and equation relating displacements and tractions along the embedded foundation soilinterface. A Jft i ,k [t k[e , fut I

? L matching conditionand the force equilibrium is then followed displacement (fe by,,kimposing t kee, q ue )

d ]4:

compatibility conditions over the top foundation surface. From that the force displacement (3AA 2) relationship (the impedance matrix) of the deleted half space can be easily deduced. where Figure 3AA 2 presents a conceptual {fle(I}

t c = nodal forces acting on illustration of the Hybrid Method implemented in surfaees S t and S e, CIASSI/ASD. To summarize, the impedances for the respeatiye1y, embedded foundation in a layered viscoelastic half space is derived as the sum of two {k;j} = impedance matrix of surface p rts -(1) Impedances of a layered viscoelastic i due to forces acting on half space before the soil is excavated, and (2) surfacej.

Impedances of the volume of soil occupied by the embedded foundation. {ut},{u}e = nodal displacements over surfaces S t and S e, respectively.

Amendmerit 1 3AA 2 l

ABWR uA6ioaan i Standard Plant nuv.A l superscripts h and f are used to differentiate unit amplitude input motion can be wTitten as between the half space and the embedded &

foundation soil interface, {f} = kje {ug} T (3AA-7)

The matching conditions are then satisfied by substituting Equation 3AA-1 into Equation 3AA 2. where The force displacement relationship on surfaces Seis obtained subsequently as {ut} =(3Nex1) free field motion vector along the embedment depth due to fe = unit amplitude control motion applied at grade level f'g f ' h' ll 'I

,ket, k k le Assuming that the foundation is rigid, (t , it i +

'r9 massless, and perfectly bonded to the soil along k ge l u:

e the foundation soil interface Se,the (3AA 3) resulting generalized forces acting on the The impedance matrix for the embedded foundation reference point can be written as foundation at surfaces Se is therefore given by (F) n[a]T {r}

kje =

(3AA-8)

'f'h' 'f11 f Knowing (F), and the rigid body foundation k k gg k le k(t.

e it -

d i + impedance matrix, [K), as described in Equation r 3AA-5, the normalized frequency dependent k ee scattering vector, {S}, can be easily found by (3AA-4) solving The total rigid body foundation impedance matrix about a foundation reference point is then obtained as

{F) =[K) {S} (3AA 9) h

[K(w)) = [o]T ge e [a]

(3AA-5) The fundamental calculation in soil structure where [o] = (3Ne x 6) rigid body motion interaction analysis using substructure method influence matrix representing the geometry of the in the frequency domain is the calculation of foundation soil interface w.r.t. the foundation the foundation motion by means of the Fourier reference point synthesis given by:

r 1

10 0 0 z y u(t)= 1 U(w) eIWt L

=

q 010 z 0 x 2n M (3AA 10)

I, 0 01 y -x 0 j

. s

~

In this equation , u(t) denotes the response (2) Calculation of Scattering Matrix in the time domain, while U(w) represents the frequency response of the foundation Scattering, as defined by CLASSI/ASD, is the displacement. The latter is calculated massless foundation response under the action of according to the governing equations of motion a unit harmonic seismic incident wave in the given by:

absence of the superstructure, r 1 -w Using matrix notation, the resulting force 4 2 ([h!o) + [h!B (w)l) +

3 acting on the foundation soil interface due to [K(w)) l (U (w)) = {F'(w)) I

/

(3AA 11) g !

Amendreent 1 3AA 3 l

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ABM MA6100AE Standard Plant nev. A where For structures that are not connected to each n

[Mo] =

(6 x 6) mass matrix of the other and rest on separate foundation mats with N foundations, the mass matrices, [Mo] and foundation [MB (w) ], consist of a total of N block diagonal matrices, each of which is (6 x 6) in

[M B(W)] =

frequency dependent mass matrix size. The coupling effects between the of the superstructure subjected foundations appear in the (6N x 6N) foundation to base excitation impedance matrix, which has the following form:

{F' (w)} = (6 x 1) generalized foundation (K ill(K 12}"' (K1Nl ~

driving force vector due to seismic input motion [K21}{K22]"'(K2N]

[Kj = . . .

The different interaction effects are clearly ' '

• separated in Equation 3AA 11. The interaction .[KNillKN2] IKNNl-effects between the foundation, superstructure, and soil (impedance) are represented by the (3AA 12) corresponding bracketed terms on the left haud side of the equation. The effects of scattering The foundation motion, { U}, is now a of the input motion is included in the term on 6N vector composed as the right hand side. The total foundation ,

motion, {U (w)}, is therefore representative of {U} = q{U1 }T {U2 }T... {Ug}TLT cach of the appropriate soil structure '

(3AA 13) interaction effects.

The driving force vector is also expanded to After the unknown, {U} is determined for each (6N X 1) and has the form frequency, the complete deformation and/or stress ,

O" analysis of the superstructure may be {F*} = l{F1 '}T {p2*}T,,{pN')

accomplished by simply applying {U} as the base (3AA-14) excitation parameter, for each of the wave form considered.

Once all required responses have been calculated in the frequency domain, the Inverse Fourier Transform technique is used to calculate the equivalent response in the time domain.

(4) Interactions Through the Soll Betweea Adjacent Structures The procedures described in the previous sections can readily be extended to analyze the interactions through the soil between adjacent structures. The governing equations of motion, in this case, is still given by Equation 3AA 11 except that the bracketed terms on both sides of the equation must be redefined.

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Amendment 1 3AA-4 1

ABM 23A6ioore Standard Plant REV.A 3AA 3 ANALYSIS CONDITIONS Therefore, a total of 50 elements (150

/' degrees of freedom is used in the numerical The input motic.n in the form of synthetic discretization for each foundation as shown in acceleration time histories for three earthquake Figure 3AA 6. It is worth mentioning that components as defined in Subsection 3A.4.1 is Dasgupta (Ref. 3) demonstrated that the used as the control motion at the finished grade wellposedness of the substructure deletion in the free field. The input motion is assumed formulation can be guaranteed when the number of to be vertically propagating plane seismic degrees of freedom associated with the top waves. The maximum ground motion acceleration foundation surface is equal to those l for the OBE is 0.15g. corresponding to the foundation soil interfaces.

The structural damping values are the same as Three sites representative of soft, medium, those defined in Subsection 3A.4.2. The and stiff soil conditions are considered. They structural modal damping ratios are calculated correspon ' to sites UB1D150, VP3D150, and according to the procedure described in VPSD150 2escribed in Subsection 3A 3.4. All Subsection 3.7.2.15. three soil profiles (UB, VP3, and VPS) have the ,

soil deposit depth of 45.7 m (150 ft) with the  !

The soil m a t e rial damping is ground water table located at 0.61 m (2 ft) strain dependent. CLASSI/ASD is a linear below grade. Rock sites are not considered in

]

analysis program. The soil nonlinearity cannot this alternate SSI analysis since the SSI ,

be directly incorporated in the analysis, effects on response of structures founded on l Therefore, it is necessary first to perform the rock are expected to be less significant than one-dimensional free field soil column those founded on soil. The SSI analyses deconvolution analysis (Ref. 2) to obtain the perfortned for the rock sites using the finite equivalent linear properties for shear modulus element approach as described in Appendix 3A and material damping compatible with seismic should be sufficient.

strains induced in the free-field. The i O strain. compatible properties are then assigned to Each of the three sites is analyzed for two be linear soil properties in the SSI analysis. conditions, the reactor building by itself, and I

The strain dependent soil properties are reactor building with adjacent control and described in Subsection 3A.3.2. turbine buildings. This constitutes a total of six (6) SSI analysis cases (cr.ch case is The building structural models used in SSI analyzed with simultaneous excitation of three  ;

analysis are the same as those described in statistically independent earthqua' e ;ornponents I Subsection 3A.8.1 except that no adjustment is covering a reasonable range of soil site rnade to the beam stick representing the embedded conditions defined in Subsection 3A.3 and, thus, outer walls of the reactor building which is adequately evaluating the effects of SSI and needed in modeling the building foundation by structure to structure interaction on the

, finite elements, reactor building response using an approach different from the finite element approach. The The foundations for the three buildings enveloped results of these 6 cases are evaluated (reactor, control, and turbine building) are against the all site enveloped loads obtained modeled separately according to the from the finite element approach to establish configurations shown in Figures 3A.11 and conservative design loads presented in Appendix 3A.12. Each foundation is assumed to be rigid. 3G.

Using the boundary element method, the top surface of the foundation is discretized into 25 rectangular subregions as shown in Figure 3AA-4, while each of the foundation soil interfaces is discretized into 5 rectangular subregions as shown in Figure 3AA 5.

O Amendment 1 3AA.5

23A6100AE Standard Plant REV.A 3AA.4 REFERENCES

1. CLASSI/ASD Computer Program for Three-Dimensional Soil / Multiple Foundation  ;

interaction Analysis, User's Manual, l Version 2.1, ASD International, Inc., San i Francisco, CA, November,1987. ,

1

2. Schnabel, P.B., Lysmer, J., and Seed, H.B., .

SHAKE--A Computer Program for Eanhquake j Response Analysis of Horizontally Layered Sites, Report No. EERC 7212, Earthquake Engineering Research Center, University of California, Berkeley, CA,1972. l 1

3. D a s g u p t a, G ., Wellpos edn ess of l Substructure Deletion Formulations, Proceedings, Sixteen Midwestern Mechanics Conference, Vol 10, Manhattan, KS, S ept e mber,1979.

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1 ABWR 22^6ioo^e Standard Plant REV.A 1

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BEFORE SOIL IS EXCAVATED FOUNDATION (BOUNDARY VISCOELASTIC HALF SPACE (CONTINUUM SOLUTION) ELEMENT METHOD) 8842748 me 4, Figure 3AA-2 CONCEPTUAL ILLUSTR ATION OF HYBRID METHOD IN CLASSI/ASD USING SUBSTRUCTURE DELETION AND BOUNDARY ELEMENT METHOD (h

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LAYERED, VISCOELASTIC HALF SPACE VOLUME. Vh (NO EXCAVATION) i j

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o ms DEFINITION OF SURFACES AND VOLUMES IN HYBRID METHOD CLASSI/ASD Figure 3AA-3

ABWR na6im.e REV A Standard Plant e

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ABWR m awAu RFV A Standard Plant O

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NORM AllZED TOTAL SUBREGION ARE A = 1.0125 88 027 90 Figure 3AA-5 EMBEDDED FOUNDATION SURFACE SUBREGION MESH WITH O DIMENSIONS NORMAll2ED TO CORRESPONDING X AND Y SIDE DIMENSIONS Amendment 1 3AA 11

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O O O

O APPENDIX 3B CONTAINMENT LOADS O

(TO BE PROVIDED BY DECEMBER 31,1988) l O

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APPENDIX 3C COMPUTER PROGRAMS USED IN THE DESIGN AND ANALYSIS OF O SEISMIC CATEGORYI STRUCTURES l I

O

ABWR uuiwan Standard Plant any. A APPENDIX 3C O TAllLE OF CONTENTS Section T_ille

. Eage 3C COMPUTER PROGRAMS USED IN THE DESIGN AND ANALYSIS OF SEISMIC CATEGORYI STRUCTURES 3C.1 INTRODUCTION 3C.1-1 3C.2 STATIC AND DYNA 511C STRUCTURAL ANALYSIS SYSTESIS (STARDYNE) 3C.2-1 3C.2.1 Description 3C.2-1 3C.2.2 Validation 3C.21 3C.23 Extent of Application 3C.2 1 3C3 CONCRETE ELE 5fENT CRACKING ANALYSIS PROGRA51 (CECAP) 3C3-1 3C3.1 Description 3C3-1 3C3.2 Validation 3C3-1 3C33 Extent of Application 3CJ 1 0 3C-il Amendment 1

4

. 23A6100AE Standard Plant REV.A 3C.1 INTRODUCTION

- The following Seismic Category 1 Structures and their foundations of the nuclear island are analyzed and/or designed using the computer programs described in this appendix.

(1) Reactor Building (2) Concrete Containment Structure I

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

23A6100AE Standard Plant any. A l

3C.2 STATIC AND DYNAMIC STRUCTURAL l Q ANALYSIS SYSTEMS (STARDYNE) 3C.2.1 Description 1 STARDYNE is a large-scale, finite element program that has a broad range of analysis types and many different structural elements. The STARDYNE system of structural analysis programs is segmented into individual programs. A variety of static or dynamic analyses may be performed by using one or more of the individual programs in a coordinated series of computer runs.

3C.2.2 Validation STARDYNE is written and maintained by the System Development Corporation of Santa Monica, California, and is available on the Control Data Corporation (CDC) system. Program validation l documentation is available at CDC in San Francisco, California.

3C.2.3 Extent of Application This program is used for the static analysis of the reactor building and containment O structure.

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Amendment 1 3C2-1

23A6100AE Standard Plant anv. A 3C.3 CONCRETE ELEMENT CRACKING n ANALYSIS PROGRAM (CECAP)

V 3C.3.1 Description CECAP computes stresses in a concrete element under thermal and/or nonthermal (real) loads, considering effects of concrete cracking. The element represents a section of a concrete shell or slab, and may include two layers of reinforcing, transverse reinforcing, prestressing tendons, and a liner plate.

The program outputs stresses and strains along the element in the concrete, reinforcement, and liner plate, and resultant forces and moments for l the composite concrete element.

CECAP assumes linear stress-strain relationships for steel and for concrete in compression. Concrete is assumed to have no tensile strength. The solution is an iterative process, whereby tensile stresses found initit'l; in concrete are relieved (by cracking) ano redistributed in the element. The equilibrium of nonthermal loads is preserved. For thermal 1 effects, the element is assumed free to expand l

(] inplane, but is fixed against rotation. The capability for expansion and cracking generally results in a reduction in thermal forces and moments from the initial condition.

3C.3.2 Validation CECAP is written and maintained by Bechtel Power Corporation (BPC) of San Francisco, California. Program validation documentation is available at BPC in San Francisco.

3C.3.3 Extent of' Application This program is used for the analysis of the reactor building and containment structure.

O Amendment 1 X'.3 1

O APPENDIX 3D COMPUTER PROGRAMS USED IN O THE DESIGN OF COMPONENTS, EQUIPMENT AND STRUCTURES I

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ABWR DA61%AE Standard Plant RIN. A APPENDIX 3D O' TABLE OF CONTENTS Section Iltle Eage 3D COMPUTER PROGRAMS USED IN THE DESIGN OF COMPONENTS, EQUIPMENT, AND STRUCTURES 3D1 INTRODUCTION 3D.1 1 3D.2 FINE MOTION CONTROL ROD DRIVE 3D.2-1 3D.2.1 Fine Motion Control Rod Drive--FMCRD01 3D.2-1 3D.2.2 Structure Analysis Programs 3D.2-1 3DJ REACTOR VESSEL AND INTERNALS 3D.3-1 3D.4 PIPING 3D.41 3D.4.1 Piping Analysis Program PISYS 3D.4 1 3D.4.2 Component Analysis--ANSI 7 3D.4-1 3D.4.3 Area Reinforcement--NOZAR 3D.4-1 3D.4.4 Dynamic Forcing Functions 3D.4-1 3D.4.4.1 Relief Valve Discharge Pipe Forces Computer Program RVFOR 3D.4-1 3D.4.4.2 Turbine Stop Valve Closure--TSFOR 3D.41 3D.4.5 Integral Attachment- LUGST 3D.4-1 3D.4.6 Piping Dynamic Analysis Program--PDA 3D.41 3D.4.7 Piping Analysis Program--EZPYP 3D.4-2 3D.4.8 Thermal Transient Program--LION 3D.4-2 3D.4.9 Differential Displact, ment Program- DISPL 3D.4-2 3D.5 PUMPS AND MOTORS 3D.5-1 3D.S.1 Structural Analysis Program -SAP 4G07 3D.51 3D.S.2 Effects of Flange Joint Connections FTFLG01 3D.5-1 0

3D il Amendment I

ABWR us6imae Standard Plant nrw. 4 APPENDIX 3D TABLE OF CONTENTS (Continued)

Section Htle Eagt 3D.6 IIEAT EXCilANGERS 3D.6-1 3D.6.1 Structural Analysis Program -SAPAG07 3D.61 3D.6.2 Calculation of Shell Attachment Parameters and Coefficients BILDR01 3D.6-1 3D.7 SOll STRUCTURE INTERACTION 3D.71 3D.7.1 A System For Analysis of Soil-Structure Interaction- SASSIO1S 3D.71 3D.7.2 Continuum Linear Analysis of Soil-Structure Interaction CLASSI/ASD 3D.71 3D.73 Free-Field Reponse Analysis- SHAKE 3D.71 O

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ABM 2346ioorn Standard Plant Rev.A 3D.1 INTRODUCTION O As discussed in Subsection 3.9.1.2, this appendix describes the major computer programs used in the analysis of the safety related components, equipment and structures. The quality of the programs and the computed results are controlled. The programs are verified for.

their application by appropriate methods, such as hand calculations, or comparison with results from similar programs, experimental tests, or published literature including analytical results or numerical results to the benchmark problems.

The updates to Appendix 3D will be provided to indicate any additional programs used by GE and especially by vendors of components and equipment, or the later version of the described programs, and the method of their verification.

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Amendment 1 3D.l.1

ABWR

• g Standard Plant 4

SECTION 3D.2 O CONTENTS 4

Section. Title PASc 3.D.2.1 Fine Motion Control Rod Drive-FMCRD01 3D.21 3.D.2.2 Stru.ture Analysis Prwrams 3D.21 t

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ABWR ux6iman Standard Plant anv x 3D.2 FINE MOTION CONTROL ROD O onive 3D.2.1 Fine Motion Control Rod Drive- FMC3D01 The program FMCRD01 is used to obtain scram performance data for various inputs to the fine motion control rod drive (FMCRD) stress analysis for both code and non code parts. The use of this program is addressed in Subsection 3.9.1.3.2. Experimental data ou pressure drops,  !

friction factors, effects of misalignment, etc.,

are used in the setting up and perfecting of this i code. Internal drive pressures and temperatures l used in the stress analysis are also determined during actual test'og of the prototype FMCRD.

3D.2.2 Structural Analysis Programs Structural analysis programs, such as NASTRO4V l and ANSYS, that are mentioned in Subsections 3D.3 1 and 3D.5 are used in the analysis of the FMCRD.

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MN MA6100AB Standard Plant any. A  ;

I 3D.3 REACTOR PRESSURE VESSEL O AND INTERNALS The following computer programs are used in the analysis of the reactor pressure vessel, core support structures, and other safety class reactorinternals: NASTR04V, SAP 4G07, HEATER, USAGE 01, ANSYS, CLAPS, ASSIST, SEISM 03AND SASS 101. These programs are described in Subsection 4.1.4.

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ABWR ux63oore Standard Plant nw A secTioN 30.4 O CONTENTS Section Title Eagt 3 D.4.1 Pinine Analvsls Procram-PISYS 3D.4-1 3 D.4.2 Component Analysis-ANSI 7 3D.4-1 3 D.4.3 Area Reinforcement-NOZAR 3D.4-1 3 D.4.4 Dmamle Forcine Functions 3D.4-1 3 D.4.4.1 Relief Valve Discharge Pipe Forces Computer Program RVFOR 3D.41 3D 4.4.2 Turbine Stop Valve Closure- TSFOR 3D41 3 D.4.5 Intecral Attachment-LUGST 3D.41 3 D.4.6 Ploine Dmamic Analysis Procram-PDA 3D.41 3 D.4.7 Pinine Analysis Procram-EZPYP 3D.4 2 3 D.4.8 Thermal Transient Procram-LION 3D.4-2 3 D.4.9 Differential Dholacement Procram-DISPL 3D.4-2 i

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ABWR mme Standard Plant RIV A 3D.4 PIPING and cumulative usage factors in accordance with p NB 3600 of the AShlE Code, Section 111. The V 3D.4.1 Piping Analysis Program PISYS program was written to perform stress analysis in accordance with the AShfE sample problem and PISYS is a computer code for analyzing piping has been verified by reproducing the results of systems subjected to both static and dynamic the sample problem analysis.

piping loads. Stiffness matrices representing standard piping components are assembled by the 3D.4.3 Area Reinforcement NOZAR program to form a finite element model of a piping system. The piping elements are connected The computer program NOZAR (Nozzle Area to each other via nodes called pipe joints. It reinforcement Program) performs an analysis of is through these joints that the modelinteracts the required reinforcement area for openings.

with the environment, and loading of the piping The calculations performed by NOZAR are in system becomes possible. PISYS is based on the accordance with the rules of the AShiE Code, linear clastic analysis in which the resultant Section !!I,1974 edition.

deformations, forces, moments and accelerations at each joint are proportional to the loading and 3D.4.4 Dynamic Forcing Functions the superposition of loading is valid.

3D.4.4.1 Relief Valse Discharge Pipe Forces PISYS has a full range of static dynamic load Computer Program RVFOR analysis options. Static analysis includes dead weight, uniformly-distributed weight, thermal The relief valve discharge pipe connects the expansion, externally applied forces, moments, pressure relief valve to the. suppression pool.

imposed displacements and differential support When the valve is opened, the transient fluid movement (pseudo-static load case). Dynamic ana- flow causes time dependent forces to develop on lysis includes mode shape extraction, response the pipe wall. This computer program computes g spectrum analysis, and time history analysis by the transients fluid mechanics and the resultant Q modal combination or direct integration. In the pipe forces using the method of characteristics.

response spectrum analysis, i.e. uniform support motion response spectrum analysis (UShiA) or inde- 3D.4.4.2 Turbine Stop Valse Closure TSFOR pendent support motion response spectrum analysis (IShfA), the user may request modal respons: com- TSFOR program computes the time history bination in accordance with NRC Regulatory Guide forcing function in the main steam piping due to 1.92. In the ground motion (uniform motion) or turbine stop valve closure. The program independent support time history analysis, the utilizes the method of characteristics to normal mode solution procedure is selected. In compute fluid momentum and pressure loads at analysis involving time varying nodal loads, the each change in pipe section or direction.

step by step direct integration method is used.

3D.4.5 Integral Attachment LUGST The PISYS program has been benchmarked against Nuclear Regulatory Commission piping models. The The computer program LUGST evaluates the results are documented in a report to the stress in the pipe wall that is produced by Commission, *PISYS Analysis of NRC Benchmark loads applied to the integral attachments. The Problems", NEDO 24210, August 1979, for mode program is based on Welding Research Council shapes and UShiA options. The IShtA option has Bulletin 198.

been validated against NUREG/CR 1677,

• Piping Benchmark Problems Dynamic Analysis Independent 3D.4.6 Piping Dynamic Analysis Support hiotion Response Spectrum hiethod,' Program PDA published in August 1985.

The pipe whip dynamic analysis is performed 3D.4.2 Component Analysis--ANSI 7 using the PDA computer program, as described in Subsection 3.6.2.2.2. PDA is a computer program The ANSI 7 computer program determines stress used to determine the response of a pipe Amendment 1 3D 41

ABWR 23461oo^u REV A Standard Plant subjected to the thrust force occurring after a 3D.4.9 Differential Displacement pipe break. Program DISPL $

The program treats the situation in terms of The DISPL program provides differential generic pipe break configuration, which involves movements at each piping attachment point based a straight, uniform pipe fixed at one end and on building modal displacements.

subjected to a titne dependent thrust force at the other end. A typical restraint used to reduce the resulting deformation is also included at a location between the two ends. Nonlinear and time-independent stress strain relations are used to model the pipe and the restraint. Using a plastic hinge concept, bending of the pipe is assumed to occur only at the fixed end and at the location supported by the restraint.

Effects of pipe shear deflection are consi-dered negligible. The pipe-bending moment-deflection (or rotation) relation used for these locations is obtained from a static nonlinear cantilever beam analysis. Using moment-angular rotation relatians, nonlinear equations of motion are formulated using energy considerations and the equations are numerically integrated in small time steps to yield the time-history of the pipe motion.

The comprehensive verification work for the PDA program is described in detail in Subsection 3.6.2.2.2.

3D.4.7 Piping Analysis Program-EZPYP EZPYP links the ANSI 7 and PISYS program together. The EZPYP program can be used to run several PISYS cases by making user-specified changes to a basic PISYS pipe model. B,3 control-ling files and PISYS runs, the EZPYP program gives the analyst the capability to perform a complete pipicg analysis in one computer run.

3D.4.8 ThermalTransient Program-LION The LION program is used to compute radial and axial thermal gradients in piping. The program calculates a time history of AT 1, AT 2.

Ta, and Tb (defined in the ASME Code,Section III, Subsection NB) for uniform and tapered pipe wall thickness.

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3DJ.1 Structural Analysis Pronram-SAP 4G07 3D.S.1 3DJ.2 Effects of Flanne .loint Connections-FITLG01 3D.51 i I J

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MM 23A6100An Standard Plant anv.A 3D.5 PUMPS AND MOTORS Pc!!owing are the computer programs used in the dynamic analysis to assure the structural and functional integrity of the pump and motor assemblies, such as those used in the ABWR ECCS systems.

3D.5.1 Structural Analysis Program -SAP 4G07 SAP 4G07 is used to analyze the structural and functionalintegrity of the pump / motor systems.

This program is also identified in Subsections 4.1.4.1.2 , 3D.3 and 3D.6. This is a gent.ral structural analysis program foe static and dynamic analysis of linear elastic complex structures. The finite element displacement method is used to solve the displacement and stresses of each element of the structure. The structure can be composed of unlimited number of three dimensional truss, beam, plate, shell, solid, plane strain plane stress and spring elements that are axisymmetric. The program can treat thermal and various forms of mechanical loading. The dynamic analysis includes mode n superposition, time history, and response (j spectrum analysis. Seismic loading and time dependent pressure can be treated. The program is versatile and efficient in analyzing large and complex structural systems. The output contains displacement of each nodal point as well as stresses at the surface of each element.

3D.S.2 Effects of Flange Joint Connections- FTFLG01 The flange joints connecting the pump bowl casings are analyzed using the FTFLG01 program.

This program uses the local forces and moments j determined by SAP 4G07 to perform flat flange j calculations in accordance with the rules set  !

forth in the ASME Code, Section 111, Appendices I XI and L k l l

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ABM 23A61%AE Standard Plant REV A SECTION 3D.6 O coureurs Section Iltle P. age 3D.6 Heat Exchangers 3D.61 3 D.6.1 Structural Analysis Pronram-SAPAG07 3D.61 3D.6.2 Calculation of Shell Attachment Parameters and Coemclents-BILDR01 3D.6-1 4

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1 Mkb 23A6100AE Standard Plant RIN. A 3D.6 HEATEXCHANGERS O The following computer programs are used in dynamic and static analysis to determine structural and functional integrity of the heat exchangers, such as those used in the ABWR RHR system.

3D.6.1 Structural Analysis Program -SAP 4G07 The structural integrity of the heat ,

exchanger is evaluated using SAP 4G07. This program is described in Subsection 3D.S.1, 3D 6.2 Calculation of Shell Attachment Parameters and Coeffl.-lents- BILDR01 BILDR01 is used to calculate the shell attachment parameters and coefficients used in the stress analysis of the support to shell junction. The method per Welding Research Council Bulletin 107 is implemented in BILDR01 to calculate local membrane stress due to the support reaction loads on the heat exchanger shell.

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MM MA6100AE Standard Plant ,

nev. A SECTION 3D.7 O CONTENTS Section Igig Eagt 3D.7.1 A System For Analysis of Soll Structure i Interaction-SASS 10lS 3D.71 3D.7.2 Continuum Linear Analysis of Soll. Structure Interaction-CLASSI/ASD 3D.71 3 D.7.3 Free Field Response Analysis-SHAKE 3D.71 O

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3D.7 SOIL-STRUCTURE INTERACTION dimensional soil structure interaction response O

V 3D.7.1 A System For Analysis of of surface founded structures using a frequency.

dependent continuum impedance approach. The Soll Structure Interaction SASS 101S basic version of the CLASSI family of computer programs was developed by Professor J.E. Luco of This program consists of a number of the University of California at San Diego, and interrelated computer program modules which can Professor H.L. Wong of the University of be used to solve a wide range of dynamic Southern California. Additional development soil structure interaction (SSI) problems in two effort was contributed by Dr. R.J. Apsel of the or three dimensions. This program is used to University of California at San Diego.

obtain enveloped seismic design loads based on the finite element method using substructuring In the CLASSI methodology, the continuum me-technique, as described in Section 3A.5 of chanics approach is used to characterize the Appendix 3A af this document. A description of site foundation system and the incident seismic this program is included in Subsection 4.1.4.1.9. waves in terms of complex, frequency dependent impedance matrices and driving force vectors.

The computer program SASSI was developed by The superstructure is represented in terms of the University of California, Berkeley, under the its fixed force vectors. The superstructure is technical direction of Prof. John Lysmer. The represented in terms of its fixed base mass ma-Bechtel version of the program was obtained from trix, mode shapes, and frequencies, and its mo-the University of California, Berkeley, under a dal damping coefficients. These structural dy-license agreement with the University. During namic properties can be calculated using any the course of installation, testing, and valida- standard finite element formulation. Compatibi-tion of the Bechtel version of the program on the lity and dynamic equilibrium requirements at the CDC CRAY System, some modifications and enhance- superstructure foundation interface are then ments were made to the program to improve the used to determine the three dimensional response performance. These include correcting the motion of the complete superstructure-foundation I' phases in Rayleigh wave calculation, replacing system.

the plate element, modifying the spring element to include damping capability, and providing the The program CLASSI/ASD is an improved version option for local end release condition in beam of the ClASSI family of computer codes, which is element. The CRAY version provided to GE,iden- developed by ASD International, Inc. This tified as GE ECP SASS 101S, contains the same mo- version is verified in accordance with the ASD's difications and enhancements made to the Bechtel Quality Assurance Program and requirements of CRAY version to date. The program was verified 10CFR50, Appendix B. Results from the program against benchmark results reported by various are verified by benchmark results obtained by investigators in the technical literature. various investigators and published in the technical literature.

3D.7.2 Continuum Linear Analysis of Soll Structure Interaction-- 3D.7.3 Free Field Response Analysis -

CLASSl/ASD SHAKE This computer program is used in analyzing This program is used to perform the limited comparative cases to comply with the dual free field site response analysis required in (finite element and half space) soil structure the seismic SSI analysis (see Subsection 3A.6).

analysis requirements, as described in Attachment A to Appendix 3A of this document. The program SHAKE is a computer program developed at the is a linear analysis program using the University of California, Berkeley, by Schnable, substructure approach based upon continuum Lysmer and Seed. (See Reference 5 of Subsection mechanics for half gace. 3A.10).The program uses the principle of one-dimensional propagation of shear waves in the The program CLASSI is comprised of a series of vertical direction for a system of horizontal, computer codes developed to calculate the three- visco clastic soil layers to compute soil Amendment 1 3D,71

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i responses in the free field. The nonlinearities in soil shear modulus and damping are accounted '

for by the use of equivalent linear soil pro-perties using an iterative procedure to obtain values for modulus and damping compatible with tne effective shear strains in each layer. The finaliterated, strain compatible properties are used as equivalent linear soil properties in seismic SSI analysis.

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APPENDIX 3E FRACTURE MECHANICS, LEAK RATE CALCULATIONS AND LEAK Q DETECTION METHODS O

ABWR ummte Standard Plant REV A APPENDIX 3E O TABLE OF CONTENTS Section 31tle Eagt 3E FRACTURE MECHANICS LEAK RATE 3 CALCULATION AND LEAK DETECTION 1 METHODS 3E.1 INTRODUCTION 3E.1 1 3E.2 MATERIAL FRACTURE TOUGIINESS CilARACTERIZATION 3E.21 j l

3E.2.1 Fracture Toughness Characterization 3E.21 l 1

3E.2.2 Carbon Steels and Associated Welds 3E.2 2 3E.23 References 3E.2-5 3E3 FRACTURE MECil ANIC MET 110DS 3E31 3E3.1 Elastic Plastic Fracture Mechanics or (J/T) Methodology 3E3-1 0 3E3.2 Application of (Jfr) Methodology to Carbon Steel Piping 3E3-3 3E33 References 3E3 3 3E.4 LE AK RATE CALCULATION METilODS 3E.4-1 3E.4.1 Leak Rate Estimation for Pipes Carrytag i Water 3E.41 i 3E.4.2 Flow Rate Estimation for Saturated Steam 3E.4 2 3E.43 References 3E.4-4 3E.5 LEAK DETECTION CAPABILITIES 3E.51 3E il Amendment 1

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APPENDIX 3E  ;

O FRACTURE MECHANICS LEAK RATE CALCULATION AND LEAK DETECTION METHODS  ;

3E.1 INTRODUCTION As discussed in Subsection 3.6.3, this appendix characterizes the fracture mechanics properties of ABWR piping materials and analysis methods, including the leak rate calculation methods. In Appendix 3F, these properties and methods are applied to specific piping systems to demonstrate their cligibility for the LBB qualification.

Piping qualified by LBB would be excluded from the non mechanistic postulation requirements of double ended guillotine break (DEGB) specified in Subsection 3.6.3. The LBB qualification means that the through wall flaw lengths that are detectable by leakage monitoring systems (see Subsection 5.2.5) are significantly smaller than the flaw lengths that could lead to pipe rupture or instability.

O Section 3E.2 addresses the fracture mechanics O properties aspects required for evaluation in accordance with Subsection 3.6.3. Section 3E.3 describes the fracture mechanics techniques and methods for the determination of critical flaw lengths and evaluation of flaw stability.

Explained in Section 3E.4 is the determination of flaw lengths for detectable leakages wi.a margin. Finally, a brief discussion on the leak detection capabilities is presented in Section 3 E.5.

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ABWR usaman Standard Plant REv. A SECTION 3E.2 O CONTENTS Section M EASC 3E.2.1 Fracture Tourhness Characterization 3E.21 3E.2.2 Carbon Steels and Associated Welds 3E.2 2 3E.2.2.1 Fracture Toughness Test Program 3E.2 2 3E.2.2.1.1 Charpy Tests 3E.2-3 3E.2.2.1.2 Stress-Strain Tests 3E.2-3 3E.2 2.1.3 J R Curve Tests 3E.2-4 3E.2.2.2 Material (J/T) Curve Selection 3E.2-4 3E.2.2.2.1 Material (Jfr) Curve for 550'F 3E.2 4 3E.2.2.2.2 Material (J/r) Curve for 420'F 3E.2 5 3E.2.3 References 3E.2 5 O

TABLES Table 11tle Eagt 3E.21 Electrodes and Filler Metal Requirements for Carbon Steel Welds 3E.2-7 3E.2-2 Supplier Provided Chemical Composition and Mechanical Properties Information 3E.2-8 3E.2 3 Standard Tension Test Data At Temperature 3E.2 9 3E.2-4 Summary of Carbon Steel J.R Curve Tests 3E.210 O

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T ABWR ux6ioore Standard Plant any A SECTION 3E.2 ILLUSTRATIONS Figure 1111e Eage 3E.2 la Schematic Representation of StaterialJ lategral R Curve 3E.2-11 3E.2-1b Schematic Representation of hiaterialJ T Cune 3E.211 3E.2-2 Carbon Steel Test Specimen Orientation Code 3E.2-12 3E.2 3 Toughness Anisotropy of ASThi 106 Pipe (6 in Sch. 80) 3E.213 3E.2-4a Charpy Energies for Pipe Test hiaterial as a Function of Orientation and Temperature 3E.214 3E.2-4b Charpy Energies for Plate Test hiaterial as a Function of Orientation and Temperature 3E.215 3E.2-5 Comparison of Base hietal, Weld and HAZ Charpy Energies for SA333 GR. 6 3E.2-16 3E.2-6a Plot of $50'F True Stress True Strain Cunes for SA333 GR. 6 Carbon Steel 3E.217 3E.2-6b Plot of 550'F True Stress True Strain Cunes for SA516 GR. 70 Carbon Steel 3E.218 3E.2-6c Plot of 350 F True Stress True Strain Cun'es for SA333 Gr. 6 Carbon Steel 3E.2 19 3E.2-6d Plot of 350 F True Stress True Strain Cunts for SA516 Gr. 70 Carbon Steel 3E.2 20 3E.2-7 Plot of 550'F Test J R Cune for Pipe Weld 3E.2-21 3E.2-8 Plot of $50*F Jmod, Tmod Data From Test J R Curve .

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23A6100AE Standard Plant REV.A 3E.2 MATERIALFRACTURETOUGH- The crack growth invariably involves some O NESS CHARACTERIZATION elastic unloading and distinctly nonproportional plastic deformation near the crack tip. J.

This subsection describes the fracture integralis based on the deformation theory of toughness properties and flow stress evaluation plasticity [4,5] which inadequately models both for the ferritic materials used in ABWR plant of these aspects of plastic behavior. In order I piping, as required for evaluation according to to use J integral to characterize crack growth Section 3.6.3. (i.e. to assure J controlled crack growth), the following sufficiency condition in terms of a i 3E.2.1 FractureToughness nondimensional parameter proposed by Hutchinson Characterization and Paris [6), is used:

When the elastic plastic fracture mechanics u. h . M > >1 (E.2-2)

(EPFM) methodology or the J T methodology is used J da to evaluate the leak before break conditions with postulated through wall flaws, the material Where b is the remaining ligament. Reference toughness property is characterized in the form 7 suggests that w>10 would satisfy the of J integral resistance curve (or J R curve) [1, J controlled growth requirements. However,if 2,3]. The J R curve, schematically shown in the requirements of this criteria are strictly Figure 3E.2 la, represents the material's followed, the amount of crack growth allowed resistance to crack extension. The onset of would be very small in most test specimen crack extension is assumed to occur at a critical geometries. Use of such a material J.R curve in value of J. Where the plane strain conditions J/T evaluation would result in grossly are satisfied, initiation J is denoted by JIC. underpredicting the instability loads for large Plane strain crack conditions, achieved in test diameter pipes where considerable stable crack specimen by side grooving, generally provide a growth is expected to occur before reaching the lower bound behavior for material resistance to instability point. To overcome this difficulty, O i di r ci arewis. ern i i8i erere a a mediri a 3 ini ai.

Jmod, which was shown to be effective even Once the crack begins to extend, the increase when limits on w were grossly violated. The of J with crack growth is measured in terms of Ernst correction essentially factors in the slope or the nondimensional tearing modulus, effect of crack extension in the calculated T, expressed as: value of J. This correction can be determined experimentally by measuring the usual T=E,# (E.21) parameters: load, displacement and crack length, 2 da af The definition of Jmod i5:

The flow stress, of, is a function of the yield and ultimate strength, and E is the elastic Jmod " J + da modulus. Generally, of si assumed as the f*aoEllfi}l da 6pl  ;

average of the yield and ultimate strength. The (E.2 3) slopeM of the material J R curve i function of cra'cl extension an. Generally, decrea. Where ses with crack extension thereby giving a convex J is based on deformation theory of upward appearance to the material J R curve in plasticity Figure 3E.2 la.

G is the linear clastic Griffith To evaluate the stability of crack growth, it energy release rate or clastic J, is convenient to represent the material J R curve Jet.

in the J.T space as shown in Figure 3E.21b. The ,

resulting curve is labeled as J T material. 6pt is the nonlinear part of the Crack Instability is predicted at the intersec. load-point displacement, (or tion point of the J/T material and J/T applied simply the total minus the elastic

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l MN 23A6100AE Standard Plant REV.A 1 displacement). treatment which refines the grain structure and, (2) a charpy test at 50*F with a specified '

a,a o are the initial and current crack minimum absorbed energy of 13 ft lbs.

lengths respectively.

The electrodes and filler metal requirements For the particular case of the compact tension for welding carbon steel to carbon or low alloy specimen geometry, the preceding Equation and the steel are as specified in Table 3E.21. A corresponding rate take the form comprehensive test program was undertaken to characterize the carbon steel base and weld Jmod 7 material tougheness properties. The next

" J + fa *o . bJyl .da (E.2 4) section describes the scope and the results of this program.

where Jp1 is the nonlinear part of the j deformation theory J, b is the remaining ligament 3E 2.2.1 Fracture Toughness Test Program and y is i The test program consisted of generating true 7 =

(1 + 0.76 b/W) (E.2 5) stress true strain curves, J-Resistance curves and the charpy V. notch tests. Two materials Consequently the modified material tearing were selected : (1) SA333 Gr. 6,16 inch modulus Tmod can be defined as: diameter Schedule 80 pipe and (2) SA516, Gr. 70, 11/4 inch thickness plate. Table 3E.2 2 shows i Tmod = Tmat + E 2 .Jpl the chemical composition and mechanical property ,

or2 b (E.2-6) test information provided by the material j supplier. The materials were purchased to the Since in most of the test J R curves the same specifications as those to be used in the w>10 limit was violated, all of the material ABWR applications.  ;

J T data were recalculated in the Jmod, Tmod format. The Jmod, Tmod calculations were To produce a circumferential butt weld, the g

performed up to crack extension of aa=10% of pipe was cut in two pieces along a the original ligament in the test specimen. The circumferential plane and welded back using the J T curves were then extrapolated to larger J shielded metal are process. The weld prep was values using the method recommended in NUREG of tingle V design with a backing ring. The 1061, Vol. 3 (9]. pebrat ternperature was 200'F.

3E.2.2 Carbon Steels and Associatef The plate material was cut along the Welds longitudinal axis and welded back using the SAW )

process. The weld prep was of a single V type i The carbon steels used in the ABWR reactor with one side as vertical and the other side at coolant pressure boundary piping are: SA 106 Gr 45 . A backing plate was used during the B, SA 333 Gr. 6 and SA 672, Gr. C70. The first welding with a clearance of 1/4 inch at the specification covers seamless pipe and the second bottom of the V. The intergass temperature was one pertains to both seamless and seam welded maintained at less than 500 F.

pipe. The last one pertains to seam welded pipe for which plate stock is specified as SA 516. Gr. Both the plate and the pipe welds were

70. The corresponding material specifications X rayed according to Code [11] requirements and for carbon steel flanges, fittings and forgings were found to be satisfactory.

are given by a GE specification [10].

It is well known that carbon steel base l While the chemical composition requirements materials show considerable anisotropy in I for a pipe per SA 106 Gr. B and SA 333 Gr. 6 are fracture toughness properties. The toughness identical, the latter is subjected to two depends on the orientation and direction of additional requirements: (1) a normalizing heat propagation of the crack in relation to the principal direction of mechanical working or  !

Amendment 1 3E.2 2

l ABM 23A6100AE Rev. A Standard Plant gain flow. Thus, the selection of proper orien- Figures 3E.2 4a and b it is clear that even at 3 tation of charpy and J R curve test specimen is room temperature the upper shelf conditions have (d important. Figure 3E.2-2 shows the orientation been reached for both the materials.

code for rolled plate and pipe specimen as given in ASTM Standard E399 p2]. Since a through. wall No such anisotropy is expected in the weld  :

circumferential crack configuration is of most metal since it does not undergo any mechanical  !

interest from the DEGB point of view, the L T working after its deposition. This conclusion i specimen in a plate and the L C specimen in a is also supported by the available data in the pipe provide the appropriate toughness properties technical literature. The weld metal charpy for that case. On the other hand, T L and C L specimen in this test program were oriented the i specimen are appropriate for the axial flaw case, same way as the LC or LT orientations in Figure  !

3E.2 2. The HAZ charpy specimens were also Charpy test data are reviewed first since they oriented similarly, provide a qualitative measure of the fracture toughness. Figure 3E.2 5 shows a comparison of the charpy energies from the 333 Gr. 6 base metal, 3E.2.2.1.1 Charpy Tests the weld metal and the HAZ. In most cases two specimens were used. Considerable scatter in The absorbed energy or its complement, the the weld and HAZ charpy energy values is seen, lateral expansion measured during a Charpy V- Nevertheless, the average energies fro the weld notch test provides a qualitative measure of the metal and the HAZ seem to fall at or above the material toughness. For example, in the case of aserage base metal values. This indicates that, austenitic stainless steel flux weldments, the unlike the stainless steel flux weldments, the observed lower Charpy energy relative to the base fracture toughness of carbon steel weld and HAZ, metal was consistent with the similar trend as measured by the charpy tests, is at least observed in the J Resistance curves. The Charpy equal to the carbon steel base metal.

tests in this program were used as preliminary O indicators of relative toughness of welds, HAZs and the base metal.

The preceding results and the results of the stress strain tests discussed in the next section were used as a basis to choose between The carbon steel base materials exhibit the base and the weld metal properties for use considerable anisotropy in the Charpy energy as in the J T methodology evaluation, illustrated by Figure 3E.2 3 from Reference 13.

This anisotropy is associated with development of 3E.2.2.1.2 Stress Strain Tests grain flow due to mechanical working. The Charpy orientation C in Figure 3E.2 3 (orientations LC The stress strain tests were performed at and LT in Figure 3E.2 2) is the appropriate one three temperatures: Room temperature,350 F for evaluating the fracture resistance to the and 550'F. Base and weld metal from both extension of a through wall circumferential the pipe and the plate were tested. The weld flaw. The upper shelf Charpy energy associated specimens were in the as welded condition. The with axial flaw extension (orientation A in standard test data obtained from these tests are Figure 3E.2 3) is considerably lower than that summarized in Table 3E.2 3.

for the circumfereratial crack extension.

An examination of Table 3E.2 3 shows that the i A similar trend in the base metal charpy measured yield strength of the weld metal, as energies was also noted in this test program, expected, is considerably higher than that of Figures 3E.2 4a and b show the pipe and plate the base metal. For example, the 550'F material Charpy energies for the two orientations yield strength of the weld metal in Table 3E.2 3 as a function of temperature. The tests were ranges from 53 to 59 ksi, whereas the base metal conducted at six temperatures ranging from room yield strength is only 34 ksi. The impact of temperature to 550 F. From the trend of the this observation in the selection of appropriate Charpy energies as a function of temperature in material (J/T) curve is discussed in later Amendment 1 3E23

l ABM 234620o4n Rev.A Standard Plant sections. shows that 5 tests were conducted at 550'F.

Two tests were on the weld metal, two were on Figures 3E.2-6 a through d show the plots of the base metal and one was on the heat affected h

the 550'F and 350* F stress stain curves zone. Figure 3E.2 8 shows the plot of material for both the pipe and the plate used in the Jmod, Tmod values calculated from the J Aa test. As expected, the weld metal stress strain values obtained from the 550'F tests. The curve in every casc is higher than the correspon- value of flow stress, af, used in the ding base metal curve. The Ramberg Osgood format tearing modulus calculation (Equation E.21) was characterization of these stress strain curve is 52.0 ksi based on data shown in Table 3E.2 3.

given in Section 3E.3.2 where appropriate values To convert the deformation J and h values of a and n is rolso provided, obtaiaed from the J.R curve into Jmod. Tmodi Equations E.0 4 and E.2 6 were used. Only the 3E 2.2.1.3 J R Curve Tests data from the pipe weld (Specimen ID OWLC A) and the plate base metal (Specimen ID BMLI 12) are The test temperatures selected for the J R shown in Figure 3E.2 8. Due to a malfunction in curve tests were: room temperature,350 F and the instrumentation, only few unreliable data 550'F. Both the weld and the base metal were points were obtained in the pipe base metal included. Due to the curvature, only the IT plan (Specimen ID OBLC 2) J R curve test. Therefore, compact tension (CT) specimens were obtained from the data from this test were not included in the the 16 inch diameter test pipe Both IT and 2T evaluation. The J.R curves from the other two plan test specimens were prepared from the test 550 F evaluated as described in the next plate. All of the CT specimens were side grooved paragraph. For comparison purposes, Figure to produce plane strain conditions. 3E.2 8 also shows the SA106 carbon steel J T data obtained from the J.R curve reported by Table 3E.2 4 shows some details of the J R Gudas [14]. Based on the data shown, the J T curve tests performed in this test program. The curve for 550 F evaluations. The curve also J R curve in the LC orientation of the pipe base includes extrapolation to higher J values based a metal and in the LT orientation of the plate base on the method recommended in NUREG 1061, Vol.3 W metal represent the material's resistance to [9),

crack extension in the circumferential direc-tion. Thus, the test results of these orienta. The Jmod Tmod data for the plate weld tions were used in the LBB evaluations. The metal and the plate HAZ were evaluated. A orientation effects are not present in the weld comparison shows that these data fall slightly metal. As an example of the J R curve obtained below those for the plate base metal shown in in the test program, Figure 3E.2 7 shows the plot Figure 3E.2 8. On the other hand, as noted in of J R curve obtained from specimen OWLC A. Subsection 3E.2.2.1.2, the yield strength of the weld metal and the HAZ is considerably higher 3E 2.2.2 Material (JfD Curve Selection than that of the base metal. The material stress strain and J T curves are f ac two key The normal operating temperatures for most of inputs in determining the instability load and the carbon steel piping in the reactor coolant flaw values by the (J/T) methodology, pressure boundary in the ABWR generally fallinto Calculations performed for representative two categories: 528 550*F and 420 F. The through wall flaw sizes showed that the higher '

latter temperature corresponds to the operating yield strength of the weld metal more than com-temperature of the feedwater piping system. The pensates for the slightly lower J R curve and, selections of the appropriate material (J/T) consequently, the instability load and flaw curves for these two categories are discussed predictions based on base metal properties are next. smaller (i.e., conservative). Accordingly, it was concluded that the material (J T) curve 3EJ.2.2.1 Material J/T curve for 550'F shown in Figure 3E.2 8 is the appropriate one to use in the LBB evaluations for carbon steel A review of the test matrix in Table 3E.2 4 piping at 550'F.

O Amendment t 3E2-4

i 21A6100AE !

Standard Plant REV.A 3E.2.2.2.2 Material Jfr Curve For 420'F 4. Rice, J.R., 'A Path Independent integral and the Approximate Analysis of Strain Since the test temperature of 350,F can be Concentration by Notches and Cracks,'J.

considered reasonably close to the 420'F, the Appl. Mech., 35, 379 386 (1968),

test J.R curves for 350'F were used in this case. A review of the test matrix in Table 5. Begley, J.A., and Landes, J.D., 'The 1 3E.2 4 shows that three tests were conducted at integral as a Fracture Criterion,' Fracture 350'F. The Jmod, Tmod data for all three Toughness, Proceedings of the 1971 National tests were reviewed. The flow stress value used Symposium on Fracture Mechanics, Part II, in the tearing modulus calculation was 54 ksi ASTM STP 514, American Society for Testing based on Table 3E.2 3. Also reviewed were the Materials, pp. 1 20 (1972),

data on SA106 carbon steel at 300'F reported by Gudas (14). 6. H utchinson, J.W., and Pa ris, P.C.,

"Stability Analysis of J. Controlled Crack ,

Consistent with the trend of the 550'F Growth,* Elastic Plastic Fracture, ATSM STP 1 data, the 350 F weld metal (J T) data fell 668, J.D Landes, J.A. Begley, and G.A. l below the plate and pipe base metal data. This Clarke, Eds., American Society for Testing probably reflects the slightly lower toughness of and Materials,1979, pp. 37 64.

the SAW weld in the plate. The (J/T) data for the pipe base metal fell between the flate base 7. Kumar, V., German, M.D., and Shib, C.F.,

metal and the plate weld metal. Ba :d on the 'An Engineering Approach for Elastic. ,

considerations aimilar to those presented in the Plastic Fracture Analysis,* EPRI Topcal l previous section, the pipe base metal J.T data, Report NP 1831, Electric Power Research although they may lie above the weld J T data, Institute, Palo Alto, C' Jul" '081. j were used for selecting the appropriate (J T)  ;

curve. Accordingly, the curve shown in Figure 8. Er n s t, H. A., "Alate, .a. .,esistan ce an d j 3E.2 9 was developed for using the (J T) Instability Beyond J. Controlled Crack methodology in evaluations at 420 F. Growth," Elastic Plastic Fracture: Second Symposium, Volume I Inlastie Crack l 3E 2 3 References Analysis, ASTM STP S03, C.F. Shih and J.P. l Gudas, Eds., American Society for Testing l

1. Paris, P.C., Tada, H., Zahoor, A., and Ernst, and Materials,1983, pp.11911213.

H., 'Th e Th eory of instability of the Tearing Afode of Elastic Plastic Crack 9. Report of the U.S. Nuclear Regulatory Growth,' Elastic Pla., tic Fracture, ASTM STP Commission Piping Review Committee, 66S, J.D Landes, J.A. Begley, and G.A Clarke, NUREG 1061, Vol.3, November 1904.

Eds., American Society for Testing Materials, 1979, pp.5 36. 10. *hiaterials and Process Specification -

ABHR,' General Electric Report No. 22A7014,

2. ' Resolution of the Task A 11 Reactor Vessel Rev.B, Sept.1982.

Afaterials Toughness Safety issue,'

NUREG 0744, Rev.1 October 1982. 11. ASME Boiler & Pressure Vessel Code,Section III, Division 1, Nuclear Power Plant

3. Paris, P.C., and Johnson, R.E., "A blethod of Components, American Society of Mechanical Application of Elastic Plastic Fracture Engineers,1980.

Afechanics to Nuclear Vessel Analysis,'

Elastic Plastic Fracture, Second Symposium, 12. ASTM Standard E399, ' Plane Strain Fracture Volume II Fracture Resistance Curves and Toughness of Aletallic Afaterials.*

Engineering Application, ASTM STP 803, C.F Shih and J.P. Gudas, Eds., American Society 13. Reynolds, M.B., "Failure Behavior in ASTAf tot Testing and Materials,1983, pp. A106B Pipes Containing Axial Through H'all 1151140. Flaws,' General Electric Report No.

O Amendment 1 3E25

23xaaorn ABWR RW A Standard Plant GEAP 5620, April 1968.

O

14. Gudas, J.P., and Anderson, D.R., '/I.R Curve Charateristics of Piping Material and Welds,' NUREG/CP-0024, Vol. 3, March 1982.

O l

O 3E24 Amendment 1

M 23A6100AE Standard Plant n,w. 4 O

TABLE 3E.21 ELECTRODES AND FILLER METAL REQUIREMENTS FOR CARBON STEEL WELDS Electrode or Filler Metal Base Material P No. Process Specification Classification Carbon Steel to P 1to SMAW SFA5.1 E7018 Carbon Steel or P-1, P-3 1.ow Alloy Steel P 4 or GTAW SFA 5.18 E70S 2, E70S 3 P5 PAW 4

GMAW SFA 5.18 E70S 2,E70S 3,E70S-6 SFA 5.20 E70T.1 SAW SFA 5.17 F72EM12K, F72EL12 O

i 4

'l O

Amendment 1 3E.2 7

ABWR 2weioare Standard Plant REV.A O

TABLE 3E.2 2 SUPPLIER PROVIDED CHEMICAL COMPOSITION AND MECHANICAL PROPERTIES INFORMATION Material Product Chemical Composillon Mech. Property Form C Mn P S SI Sy(ksi) Su(ksi) Elongation

(%)

SA 333 Gr.6 16 In. 0.12 L18 .01 .026 0.27 44.0 67.5 42.0 Heat #52339 Sch.80 Pipe SA 516 Gr.70 1.0 In. 0.18 0.98 0.017 0.0022 0.25 46.5 70.5 31.0 Heat #E18767 Plate Note: (1) Pipe was normalized at 16500F. Held for 2 hrs, and air cooled.

(2) Plate was normalized at 17000F for one hour and still air cooled.

I I

l O

Amendment 1 3E.24 1

ABM 2 mime arv.A Standard Plant ,

O l TABLE 3E.2 3 STANDARD TENSION TEST DATA AT TEMPERATURE SPEC. MATERIAL 1EST 0.2% YS UTS Elong. RA NO. TEMP Ihall ikall fil 1 OW1 PIPE WELD RT 66.1 81.6 32 77.2 OW2 PIPE WELD 550F $9.0 93.9 24 56.7 ITWL2 PLATE WELD 550F 53.0 9L4 34 513 IBL1 PLATE BASE RT 44.9 73.7 38 513 IBL2 PLATE BASE 350F 37.9 64.2 M 68.9 IBL3 PLATE BASE 550F 34.1 69.9 29 59.4 OB1 PIPE BASE RT 43.6 68.6 41 67.8 OB2 PIPE BASE 350F 42.2 74.9 21 55.4 OB3 PIPE BASE 550F 34.6 78.2 31 55.4 O

l j

O Amendment 1 3E29 1

ABM 2mioaxs Standard Plant RR A TA3LE 3E.2 4 O

SUMMARY

OF CARBON STEEL J R CURVE TESTS b'2 SDedmeD ID Shg Descriotlog Itagg, (1) OWLC A 1T Pipe Weld 5500F (2) OBCle) 1T Pipe Base C L Orientation RT (3) OBLC2 1T Pipe Base 1-C Orientation $500F (4) OBLC3-B IT Pipe Base L-C Orientation 3500F (5) BM14 1T Plate Base Met 4 L T Orientation RT (6) BML4-14 2T Plate Base Meta 1 T Orientation RT (7) BML2-6 2T Plate Base Meta l-T Orientation 3500F (8) BML112 2T Plate Base Meta 1-T Orientation 5500F (9) WM3-9 2T Plate Weld Metal RT O (10) XWM111 2T Plate Weld Metal 350 F (11) WM2 5 2T Plate Weld Metal 5500F (12) HAZ (Non- Heat Affected Zone, Plate RT standard)

Width = 2.793*

(13) OWLC 7 1T Pipe Weld RT Notes:

1. Pipe base meta SA333 Gr.6

2. Plate base meta SA516 Gr.70

3. Pipe weld made by shielded metal are welding.

4. Plate weld made by submerged are welding.

O Amendment i 3IL210 l

YS Standard Plant "[

O l I

l J J SLOPE da O

)

As T 87 592 02 87 592 03

. l

\

l a

Figure 3E.2-1a SCHEMATIC Figure 3E.2-1b SCHEMATIC REPRESENTATION REPRESENTATION i Q OF M ATERI AL J-INTEGRAL R CURVE OF MATERIAL J-T CURVE

^=wmai 2 su.n

I

ABWR 23462m4e Standard Plant REV.A 9

V

+P

-s uY t

W

, L-L-C CRACK PLANE ORIENTATION CODE FOR BAR AND HOLLOW CYLINDER H L

'TNIAON G

NDI T ON LE TU ECC TI I IRE G NGG DD IRGIN LOLI NNOR SIOF OL UOF RR

\ GB- T EXXIS A

\ ~~

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\ \-

\ <y A O%

'A

$ c,$ \

R)t E _

_S S V NS ERSE s, se' S SHOR ICKNE THT TRA v

/

CRACK PLANE ORIENTATION CODE FOR RECTANGULAR SECTIONS 87 592-04 Figure 3E.2-2 CARBON STEEL TEST SPECIMEN ORIENTATION CODE Amendment 1 3E212

ABWR uuime Standard Plant REV A O-250 -

1. ./, .- #

,,, C 1

A

& D 200 -

SPECIMEN ORIENTATION b

m 150 -

z O  :

u 1M -

C 50 -

f B I I l l

-200 -100 0 100 200 300 TEMPERATUR E (* F) ,

O Figure 3E.2-3 TOUGHNESS ANISOTROPY OF ASTM 106 PIPE (6 in.Sch.80)

Amendment 1 3E.213

ABWR n46imit Standard Plant REV.A O

250 -

SA 516 GR. 70 BASE MATERIAL 6 ORIENTATION LT O ORIENTATION TL 200 -

E h

I g 150 -

W E O E A A

^ T 5 300 _

3 f O

50 -

O O O O g O

O 100 200 300 400 500 TEST TEMPERATURE (F) 87 592 06 Figure 3E.2-4a CHARPY ENERGIES FOR PIPE TEST MATERIAL AS A FUNCTION OF 3 ORIENTATION AND TEMPERATURE W

Amendment 1 3E 214

ABM 23A6100AE Standard Plant arv.A

,_3 250 200 SA 333 GR. 6 BASE MATERIAL A ORIENTATION LC

_ O ORIENTATION CL 4

s 150 -

U 5

5 S

g 100 O O e

A A A g A A 0 0 0 0 0 0

0 100 200 300 400 500 600 TEST TEMPER ATU RE (* F) 87 592 07 Figure 3E.2-4b O CHARPY ENERGIES FOR PLATE TEST MATERIAL AS A FUNCTION OF ORIENTATION AND TEMPERATURE Amendm5nt 1 3fL215

33A6100AE Standard Plant REV A 9

A BASE METAL g 250 - O WELD METAL O HAZ 200 -

O

$ 150 5

5  % ..

[

100 -

50 - l MATE RI AL: SA 333 GR.6 SPECIMEN ORIENTATION LC l

0 500 600 0 100 200 300 400 TEST TEMPE R ATURE ('F) 87 592 08 Figure 3E.2-5 COMPARISON OF BASE METAL, WELD AND HAZ CHARPY ENERGIES g' FOR SA 333 GR. 6 3E.216 Amendment 1

l Mkb 23A6100AE Standard Plant REV A Q,

SA 333 GR. 6 PIPE WELD METAL as*

1. s#p 80 -

j

1. p\ BASE METAL T l

5 ,/

8 / TEMPERATURE 550*F 60 /

j 0 0 e /

/

e /

/

/

40 -#

1

,4 20 -

o l I I I I I I I I 0.02 0.04 0.06 0.08 0.10 0.12 0.14 0.16 0.18 0.20 TRUE STR AIN 87 592 09 Figure 3E.2-6a PLOT OF 550*F TRUE STRESS-TRUE STRAIN CURVES FOR SA 333 GR. 6 CARBON STEEL Amendment 1 3E.217

ABM 23xerooxe Standard Plant REV.A 9

SA 516 GR. 70 TEMPER ATURE 550*F 100 -

WELD METAL l

80 -

y*  ;

/  :

h 3

gw j

1. /N BASE METAL l

m

/ a y i

/

$ /

? /

/ i

/

40 A i

1 i r

i l 20 -

l I I I I I I I I o

0.02 0.04 0.06 0.08 0.10 0.12 0.14 0.16 0.18 0.20 i

TRUE STR AIN 87 592 10 Figure 3E.2-6b PLOT OF 550* F TRUE STRESS-TRUE STRAIN CURVES 9

FOR SA 516 GR,70 CARBON STEEL Arnendment 1 3E.2-18

ABWR 23A61 NAE Standard Plant REV.A O

100 -

80 -

T a

12 w

E 60 -

O E E

r 40 -

I 20- -

I I  !  !  !  ! I I 0

0.02 0.04 0.06 0.08 0.10 0.12 0.14 0.16 0.18 TRUE STRAIN 87 592 11 Figure 3E.2-6c PLOT OF 350* F TRUE STRESS-TRUE STRAIN CURVES FOR SA 333 GR,6 CARBON STEEL Amendment 1 3gg,39 i

ABWR 23462004E Standard Plant REV.A O

100 80 -

s 80 -

E B

n 9

% 40 20 -

o l l l l l l l l 0.02 0.04 0.06 0.08 0.10 0.12 0.14 0.16 0.18 TRUE STRAIN 87 592 12 Figure 3E.2-6d PLOT OF 350* F TRUE STRESS-TRUE STRAIN CURVES FOR SA 516 GR. 70 CARBON STEEL Amendment 1 3E.2 20 l

l

• l

MN nA6100AE Standard Plant arv. 3 O

I 1

4000 ,

e a

3000 -

o a

0 I E o a i

Os,2 i

! g C

= \

~

F OWLC A p WELD 550'F x a,o, . ,,

O Jmod 0

O.00 0.04 0.08 0.12 0.16 0.20 0.24 0.28 i l

DE LTA - a (in.)

87 592 13 l

O Figure 3E.2-7 PLOT OF 550*F TEST J-R CURVE FOR PIPE WELD Amcodment 1 3E.2 21

i ABWR 23461ooin Standard Plant REV.A 5000 0

A GUDAS DATA 9 CARBON STEEL WELD DATA 4000 - l A I A

A

- A A

e A 3000 -

A 5 e A u -

' A

, 4 2000 -

A A

1@ -

LOWER BOUND MATERIAL J-T CURVE

! FOR CARBON STEEL AT 550'F 0

0 100 200 300 400 500 600 700 TEARING MODULUS, T 87 592 14 Figure 3E.2-8 PLOT OF 550'F Jmod Imod DATA FROM TEST J-R CURVE h Amendment 1 3E2 22

ABM 2346iooxs Standard Plant REV A O l 1

5000 -

4000 -

c \

1. 3000 O i 2000 -

1000 -

1 I I l 0

100 200 300 400 TEARING MODULUS, T

- 87 592 15 l

l 1

O Figure 3E.2-9 CARBON STEEL J-T CURVE FOR 420*F 3E 2-23 Amendment 1 l

ABWR MA6100AE I Standard Plant REV A l

SECTION 3E3 l O courEurs  !

Section M Eggg )

3E.3.1 Elastic Plastle Fracture Mechanics or U/T)

Methodclogy 3E3-1 3E3.L1 Basic (1/F) Methodology 3E3.1

)

3E3.1.2 J Estimation Scheme Procedure 3E3-1 3E3.13 Tearing Irstability Evaluation Considering Both Tee Membrane and Bending Stress 3E3-2 3E.3.2 AppJigtl,.qn_q[Qfr) Methodoloey to Cart,on SteelPiulng 3E3-3 3ES.2.1 Deterrsination of Ramberg Osgood Parameters I For 550 F Evaluation 3E3-3 1

3E3.2.2 Determination of Ramberg Osgood Parameters For 420 F Evaluation 3E3-3 3E33 References 3E3-4 O I i

1

)

3ES-il O

Amendment 1 l

ABWR m6me Standard Plant REv.A SECTION 3E3 ILLUSTRATIONS Figure Title Eage 3E3-1 Schematic Blustration of Tearing StaM11ty Evaluation 3E3-6 3E3-2 A Schematic Representation of Instability Tension and Bending Stresses as a Function of Flaw length 3E3-7 3ES-3 SA 333 Gr 6 Stress Strain Data at550* Fin the Ramberg-Osgood Format 3E3-8  ;

1 3E3-4 Carbon Steel Stress-Strain Data at 350'F in the '

Ramberg-Osgood Format 3E3-9 1

i l

O l

l l

l l

i i

9 l 3ES-iii Amendment 1

1 1

\

23A6100AE Standard Plant REV.A l

3E3 FRACTURE MECHANICMETHODS Intersection point of the material and l

{s} applied (J/T) curves denotes the instability This subsection deals with the fracture point. This is mathematically stated as I

mechanics techniques and methods for the follows:

determination of critical flaw lengths and instability loads for materials used in ABWR. Jappled (a,P) = Jmat (a) (3E3-2)

These techniques and methods comply with Criteria l (5) through (11) described in Section 3.63. l Tapplied <Tmat (stable) (3ES-3) l 3EJ.1 Elastic Plastic Fracture Tapplied >Tmat (unstable)

Mechanics or (J/T) Methodology The load at instability is determined from Failure in dur.ile materials such as highly the J versus load plot also shown schematically  !

tough ferritic materials is characterized by in Figure 3E31. Thus, the three key curves in considerable plastic deformation and significant the tearing stability evaluation are: Japplied l amount of stable crack growth. The EPFM approach versus Tapplied Jmat versus Tmat and i outlined in this subsection considers these Japplied versus load. The determination of aspects. Two key concepts in this approach are: appropriate Jmat versus Tmat or the material (1) J integral [1,2] which characterizes the (J/T) curve has been already discussed in intensity of the plastic stress strain field subsection 3E.2.1. The Japplied Tapplied surrounding the crack tip and (2) the tearing or the (J/T) applied curve can be casily instability theory [3,4] which examines the generated through perturbation in the crack stability of ductile crack growth. A key length once the J a tied versus load advantage of this approach is that the material information is availabN for different crack fracture toughness characteristic is explicitly lengths. Therefore, only the methodology for factored into the evaluation. the generation of Japplied versus load information is discussed in detail.

O 382.>.1 BasictJ/T> uetaedelear 3EJ.1.2 J Estimation Scheme Procedure Figure 3E.31 schematically illustrates the J/T methodology for stability evaluation. The The Japplied or J as a function of load was i material (J/T) curve in Figure 3E.3-1 repre- calculated using the GE/EPRI estimation scheme  !

sents the material's resistance to ductile crack procedure [5, 6]. The J in this scheme is  ;

extension. Any value of J falling on the mate- obtained as sum of the clastic and fully plastic '

rial R. curve is denoted as Jmat and is a func- contributions:

tion solely of the lucrease in crack lengthaa.

Also defined in Figure 3E3-1 is the ' applied' J, J=J+J p e (3E3-4) which for given stress strain properties and overall component geometry, is a function of the The material true stress strain curve in the applied load P and the current crack length, a. estimation scheme is assumed to be in the Hutchinson and Paris [4] also define the Ramberg Osgood format:

following two nondimensional pararneters:

n (3E3-5)

E . BJ applied +o IE i Tapplied "af 2 g, h,j * [(E t

e o,j o,j (3E3-1) where,'o is the material yield stress,

.E_,. $ mat ' =

fo , and o and n are obtained Tmat "or2 da by fitting the preceding equation to the rnaterial true stress strain curve. -

where E is Young' modulus and a r is an appropriate flow stress. The estimation scheme formulas to evaluate -

Amendraent 1 31D-1

23A6100AE Standard Plant REV.A the J integral for a pipe with a through wall This aspect is addressed next.

circumferential flaw subjected to pure tension or pure bending are as follows 3E.3.1.3 Tearing Instability Evaluation Considering Both the Membrane and Bending Tension Stresses (3ES-6)

J = f1(ae , tR) EE2 + Based on the estimation scheme formulas and n+1 the tearing instability methodology just 8

o e 'o c (A) h 1(.A, n, B) 'E ' outlined, the instability bending and tension b b tP.o stresses can be calculated for various through wall circumferential flaw lengths, where, Figure 3E.3 2 shows a schematic plot of the instability stresses as a function of flaw f 1(A, n, B) , a F2 A, n, E) length. For the same stress level, the b t t allowable flaw length for the bending is in R' t' expected to be larger than the tension case.

Po-20o Rt [w . 7 - 2 arc sin When the applied stress is a combination of (1 sin y)] the tension and bending, a linear interaction 2 rule is used to determine the instability stress or conversely the critical flaw length. The Bending application of linear interaction rule is (3E3-7) certainly conservative when the instability load ir, close to the limit load.

J = fie (a , tE) EM*+ n+1 o "o ' o c (3) 1h g, n,t E)

M The interaction forrnulas are following: (See

.M o. Figure 3E.3 2) g where, Critical Flaw 12neth (3ES-8)

II(3, n,8) . xa (g)2 p 2 ,c , ( gl) 8 ,tc + ( ab ) a c,t b t I at + #b St + #b (A, n, B) b t where:

M , , M [cos (p . i sin (7)] og = applied membrane stress ob - applied bending stress The nondimensional functions F and b are given in Reference 6 ac,t = critical flaw length for a tension stress of (#t+0b)

While the calculation of J for given o, o,

1. o and load type is reasonably straight- ac,b = critical flaw length for a bending forward, one issue that needs to be addressed is stress of (ot+ 0b) the tearing instability evaluation when the loading includes both the membrane and the Instability Bendine Stress bending stresses. The estimation scheme is , (3E3-9a) capable of evaluating only one type of stress at S a tirne, b " (1 at -ft) o'b O

Amendment 1 3E32 l

l ABM 2346ioo4E REV.A Standard Plant p

V where: that a limit load approach is feasible.

However, test data at high temperatures Sb = instability bending stress for flaw specially involving large diameter pipes are length, a,in the presence of membrane currently not available. Therefore, a (J/T) stress, at. based approach is used in the evaluation.

= applied membrane stress 3E.3.2.1 Determination of Ramberg Osgood ,

't I Parameters for 550'F Evaluation

= instability tension stress for flaw l o{

length, a. Figure 3E.2 6a shows the true stress true strain curves for the carbon steels at a'b -instability bending stress for flaw 550*F. The same data is plotted here in length, a. Figure 3E.3-3 in the Ramberg Osgood format. It is seen that, unlike the stainless steel case, Once the instability bending stress, Sb , in each set for stress strain data (i.e. data the presence of membrane stress, at, is derived from one stress strain curve) follow l determined, the instability load margin approximately a single slope line. Based on the corresponding to the detectable leak-size crack visual observation, a line representing a = 2, (as required by LBB criterion in Section 3.6.3) ri = 5 in Figure 3E3-3 was dran as representing can be calculated as follows: a reasonable upper bound to the data shown.

Instabilitylead Margin _o + S (3E.3-9b) The third parameter in the Ramberg Osgood

- t format stress stain curve is o o, the yield stress. Based on the severalinternal GE It is assumed in the preceding equation that data on carbon steels such as SA 333 Gr.6, and the uncertainty in the calculated applied stress SA 106 Gr.B a reasonable value of 550*F O is essentiai>x asseciated with the sires , due te vieid strensih was judsed as 34600 esi. Te applied bending loads and that the membrane surnmarize, the following values were used in stress, which is generally due to the pressure this report loading, is known with greater certainty. This of carbon steels forasthe 550(J/T}F: methodology evaluation method of calculating the margin against loads is also consistent with the definition of load margin employed in Paragraph IWB 3640 of Section a =2.0 XI [7].

n =5.0 3E.3.2 Application of(J/T)

Methodology to Carbon Steel Piping oo = 34600 psi From Figure 3E.2 3,it is evident that carbon E = 26x10' psi steels exhibit transition temperature behavior marked by three distinct stages: lower shelf, 3E.3.2.2 Determination of Ramberg Osgood transition and upper shelf. The carbon steels Parameters for 420 F Evaluation generally exhibit ductile failure mode at or above upper shelf temperatures. This would Figure 3E.3-4 shows the Ramberg Osgood (R 0) suggest that a net section collapse approach may format plot of the 350*F true stress stain be feasible for the evaluation of postulated data on the carbon steel base metal. Also shown flaws in carbon steel piping. Such a suggestion in Figure 3E.3-4 are the CE data a SA 106 Grade was also made in a review report prepared by the B at 400* F. Since the difference between Naval Research Lab [8). Low temperature (i.e. the ASME Code Specified minimum yield strength less than 125 F) pipe tests conducted by GE at 35 0

• F a n d 420

• F is s m all, th e

[9] and by Vassilaros (10) which involved 350 F stress strain data were considered circumferentially cracked pipes subjected to applicable in the determination of R O bending and/or pressure loading, also indicate parameters for evaluation at 420'F.

Amendment 1 3E.3 3

ABWR mamae Standard Plant any A A review of Figure 3E.3 4 indicates that the 668, J.D La ndes, J. A. Begley, a n d majority of the data associated with any one test G.A. Clarke, Eds., American Society for can be approximated by one straight line. Testing and Materials,1979, pp. 37-64.

It is seen that some of the data points 5. Kumar, V., German, bl.D., and Shib, C.F.,

associated with the yield point behavior fall "An Engin eering Approach for along the y-axis. However, these data points at Elastic Plastic Fracture Analysis,' EPRI low stain level were not considered significant Topcal Report NP-1831, Electric Power and, therefore, were not included in the R O fit. Research Institute, Palo Alto, CA July 1981.

The 350*F yield stress for the base 6. ' Advances in Elastic-Plastic Fracture material is given in Table 3E.2-3 as 37.9 ksi. Analysis,' EPRI Report No. NP 3607, August Since the difference between the ASME Code 1984.

specified minimum yield strengths of pipe and plate carbon steels at 420 F and 350 F is 7. ASME Boiler and Pressure Vessel Code, roughly 0.9 ksi, the a o value for use at Section XI, Rules for In-service Inspection 420 F are chosen as (37.9 0.9) or 37 ksi, of Nuclear Power Plant Components, ASME, In summary, the following values of R O 1986 Edition.

parameters are used for evaluation of 420 F:

8. Ch a n g, C.I.,e t al,

• Piping in ciastic O

o = 37,000 psi Fracture Me c h a n ic s A n alysis,'

N U R E G / C R-1119, J u n e 1980. l a = 5.0 l

9. "Reactor Primary Coolant System Rupture l n = 4.0 Study Quarterly Progress Report No.14, ,

July September,1968,* GEAP-5716, AEC 3E.3.4 References Research and Development Report, December 1968.

g l

1. Rice, J.R., *A Path Independent Integral and 1 the Approximate Analysis of Strain 10. V a s sila r o s, M .G., e t al, *J-In tegral Concentration by Notches and Cracks,' Tearing instability Analyses for 8 inch J. Appl. Mech., 35, 379 386 (1968). Diameter A S TM A J06 Steel Pipe,' l NUREG/CR 3740, April 1984 I

2. Begley, J.A., and Landes, J.D., 'The J Integral as a Fracture Criterion,' Fracture 11. Harris, D.O., Lim, EY., and Dedhia, D.D.,

Toughness, Proceedings of the 1971 National "Probability of Pipe Fracture in the Symposium on Fracture Mechanics, Part II, Pn' mary Coolant Loop of a PWR Plant, Volume ASTM STP 514, American Society for Testing 5, Probabilistic Fracture Mechanics Materials, pp. 1 20 (1972). Analysis,' U.S. Nuclear Regulatory CommissionReportNUREG/CR 2189, Volume 5

3. Paris, P.C., Tada, H., Zahoor, A., and Ernst, Wathington,DC,1981.

H., Th e Th eory of Instability of th e Tearing Mode of Elastic Plastic Crack 12. Buchalet, C.B., and Bamford, W.H., ' Stress Growth,' Elastic Plastic Fracture, ASTM STP Intensity Factor Solutions for Continuous 668, J.D Landes, J.A. Begley, and G.A Clarke, Surface Flaws in the Reactor Pressure Eds., American Society for Testing Materials, Vessels," Mechanics of Crack Gromb, ASTM 1979, pp.5-36. STP 590. American Society for Testing Materials,1976, pp. 385 402.

4. H ut chin son, J.W., a n d Paris, P.C.,

' Stability Analysis of J Controlled Crack 13. Hale, D.A., J.L. Yuen and T.L. Gerber, Growth,' Elastic Plastic Fracture, ATSM STP ' Fatigue Crack Growth in Piping and RPV O

Amendment 1 3E.3-4

MM 23A6100AE i Standard Plant REV.A l Steels in Simulated BWR Water Environmnent,' l O General Electric Report No. GEAP.24098, January 1978.

I

14. Hale, D.A., C.W. Jewett and J.N. Kass,

' Fatigue Crack Growth Behavior of Four Structural Alloys in High Temperature High Purity Oxygenated Water,' Journal of Engineering Materials and Technology, Vol.

101, July 1979.

15. Hale, D.A., et al,

• Fatigue Crack Growth in ,

Piping and RPVSteels in Simulated BWR Water l Environment - Update 1981,' General Electric Proprietary Report NEDE-24351, July 1981.

l 1

l 0 l l

l l

l l

l l

O Amendment 1 3E35

MM uA61ooAs Standard Plant arv. 4 O

i i

l l

J J INSTA81LITY POINT (Jmat, Tmat)

INSTABILITY li STRESS OR LOAD

{

l '

I I I (J,pp,T.pp) 8 4

STRESS OR LOAD T 87 592 16 i

l Figure 3E.3-1 SCHEMATIC ILLUSTRATION OF TEARING STABILITY EVALUATION Amendment 1 3E M.

ABWR 23461ooxe Standard Plant REV.A O

I a

R

, b=nR

\

I

, l TENSION BENDING v> l

'"o __. _ _

l ,

i I l

I

. I

! l I I I [

i  ! I

• s Sc,t Scb F LAW LENGTH 87 592 17 Figure 3E.3-2 A SCHEMATIC REPRESENTATION OF INSTABILITY TENSION AND BENDING STRESSES AS A FUNCTION OF FLAW STRENGTH g

Amendment 1 3E3-7

l l

23A6100AE Standard Plant arv A I l

O 1% -

l a - 2, n - 5 0 2

O b O O O A g 10 l

C 9 o l O 8 A. O A .*  ; O 6

.. O 5 b'O 4

., 3 2

.g 13 I I I I I I I I I I I O 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.80.9 1.0 1.1 LOGE 87 592 18 O Fipre 3E.3-3 SA 333 GR,6 STRESS-STRAIN DATA AT 550*F IN THE RAMBERG-OSGOOD FORMAT Amendment 1 3E.3-8

ABM a346ioarc Slandard Plant REV.A l l

STRESS-STRAIN DATA AT 350*F PlPE SA 333 GR6 PLATE SA 516 GR 70 PlPE SA 106 GR B (CE DATA) )

100 -

n l a = 4.5, n = 4.7 a = 5.0, n = 4.0 12 l

10 9

8 10 -

6 a = 2.2, n = 3.5 5 4

3 2

1 l l 1 1 I I I I I I I I 0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 1.1 LOGE l 87 592 19 Figure 3E.3-4 CARBON STEEL STRESS-STRAIN DATA AT 350*F IN THE R AMBERG-OSGOOD FORMAT g

Amendment 1 3E.59 I t

1

ABM ux6iman Standard Plant REV.A SECTION 3E.4 Q

l 1

Y CONTENTS I Section M Eggg 1 3E.4.1 Leak Rate Estimation for Ploes Caminn Water 3E.41 3E.4.1.1 Description of Basis for Flow Rate Calculation 3E.4-1 3E.4.1.2 Basis for Crack Opening Area Calculation 3E.4-1 3E.4.13 Comparison VeriScation With Experimental Data 3E.4-2 3E.4.2 Flow Rate Estimation for Saturated Steam 3E.4-2 3E.4.2.1 Evaluation Method 3E.4 2 1

3E.4.2.2 Selection of Appropriate Friction Factor 3E.4-2 3E.4.23 Crack Opening Area Formulation 3E.4-3 3E.43 References 3E.4-4 1

O TABLES Table 1111e East 3E.4-1 Mass Flow Rate for Several ft/Db Values 3E.4 5 O

3E.4-il Amendment 1

23A6100AE Standard Plant REV.A SECTION 3E.4 ILLUSTRATIONS Figure 11tle Eagt 3E.4-1 Comparison of PICEP Predictions with Measured Leak Rates 3E.44 3E.4-2 Pipe Flow Model 3E.4-7 3E.4-3 Mass Flow Rates for Steam / Water Mixtures 3E.4-8 3E.4-4 Friction Factors for Pipes 3E.4-9 O

O 3E.4-iii Amendment 1

ABM 23462mie Standard Plant REV.A 3E.4 LEAK RATE CALCULATION For given stagnation conditions and crack n

U METHODS geometries, the leak rate and exit pressure are calculated using an iterative search for the Leak rates of high pressure fluids through exit pressure starting from the saturation cracks in pipes are a complex function of crack pressure corresponding to the upstream geometry, crack surface roughness, applied temperature and allowing for friction, stresses, and 'alet fluid thermodynamic state. gravitational, acceleration and area change Analytical predictions of leak rates essentially pressure drops.. The inertial flow calculation consist of two separate tasks: calculation of the is performed when the critical pressure is crack opening area, and the estimation of the lowered to the back pressure without finding a fluid flow rate per unit area. The first task solution for the critical mass flux.

requires the fracture mechanics evaluations based on the piping system stress state. The second A conservative methodology was developed to task involves the fluid mechanics considerations handle the flow of two phase mixture or in addition to the crack geometry and its surface superheated steam through a crack. To make the roughness information. Each of these tasks are model continuous, a correction factor was now discussed separately considering the type of applied to adjust the mass flow rate of a fluid state in BWR piping. saturated mixture to be equal to that of a slightly subcooled liquid. Similarly, a 3E.4.1 IAak Rate Estirnation for correction factor was developed to ensure Pipes Carrying Water continuity as the steam became superheated. The superheated model was developed by applying EPRI developed computer code PICEP [1] was thermodynamic principles to an isentropic used in the leak rate calculations. The basis expansion of the single phase steam.

for this code and comparison of its Icak rate predictions with the experimental data is The code can calculate flow rates through described in References 2 and 3. This code was fatigue or IGSCC cracks and has been verified also recently used in the successful application against data from both types. The crack surface of LBB to primary piping system of a PWR. The roughness and the number of bends account for basis for flow rate and crack opening area the difference in geometry of the two types of calculations in PICEP is briefly described cracks. The guideline for predicting leak rates I

first. A comparison with experimental data is through IGSCCs when using this model was based I shown next. on obtaining the number of turns that give the best agreement for Battelle Phase 11 test data 3E.4.1.1 Description of Basis for Flow Rate of Collier et al. l4). For fatigue cracks, it Calculation is assumed that the crack path has no bends.

The thermodynamic model implemented in PICEP 3E.4.1.2 Basis for Crack Opening Arta computer program assumes the leakage flow through Calculation '

pipe cracks to be isenthalpic and homogeneous, but it accounts for non equilibrium "flashing' The crack opening area in PICEP code is transfer process between the liquid and vapor calculated using the estimation scheme phases. formulas. The plastle contribution to the displacent is computed by summing the Fluid friction due to surface roughness of the contributions of bending and tension alone, a walls and curved flow paths has been incorporated procedure that underestimates the displacent in the model. Flows through both parallel and from combined tension and bending. However, the convergent cracks can be treated. Due to the plastic contribution is expected to be complicated geometry within the flow path, the insignificant because the applied stresses at model uses some approximations and empirical normal operation are generally such that they do factors which were confirmed by comparison not produce significant plasticity at the against test data, cracked location.

Amendment 1 3E.4-1

23A6100AE Standard Plant RW.A 3E.4.1.3 Compard son Verification with effective surface protrusion height to hydraulic Experimental Data diameter, were relied upon in this case. Figure g 3E.4-4, from Reference 6, graphically shows such W Figure 3E.41 from Reference 3 shows a a relationship for pipes. The e/Dh ratio comparison PICEP prediction with measured leak for pipes generally ranges from 0 to 0.50.

rate data. It is seen that PICEP predictions are However, for a fatigue crack consisting of rough virtually always conservative ('..c., the leak fracture surfaces represented by a few mils, the flow rate is underpredicted). roughness height e at some location may be almost as much as 6. In such cases, c/Db 3E.4.2 Flow Rate Estimation for would seem to approach 1/2. There are no data Saturated Steam or any analytical rnodel for such cases, but a crude estimate based on the extrapolation of the 3E.4.2.1 Evaluation Method results in Figure 3E.4 4 would indicate that f may be of the order of 0.1 to 0.2. For this The calculations for this case were based on evaluation an average value of 0.15 was used j the maximum two-phase flow model developed by with the modification as discussed next.

Moody [ Reference 5]. This model predicts the flow rate of steam water mixtures in vessel For blowdown of saturated vapor, with no ,

blowdown from pipes (see Figure 3E.4 2). A key liquid present, Moody states that the friction i parameter that characterized the flow passage in factor should be modified according to the Moody analysis is fL/Dh, where, f is the )

1 coefficient of friction, L, the length of the (3E.4-1) I flow passage and Dh , the hydraulic diameter. . 1/3 The hydraulic diameter for the case of flow fg- fosp d through a crack is 26 where 6 is the crack "8 opening displacement and the length of the flow where passage is t, the thickness of the pipe. Thus, fg = modified friction factor the parameter fL/D b ni the Moody analysis was interpreted as ft/26 for the purpose of this fGSP = factor for single phase gl )

evaluation.

Ei = liquid / vapor specific volume ,

Figure 3E.4 3 shows the predicted mass flow "S ratio evaluated at an average  !

rates by Moody for fL/Dh of 0 and 1. Similar static pressure in the flow path I plots are given in Reference 5 for additional fL/Db values of 2 through 100. Since the steam This correction is necessary because the ,

in the ABWR main steam lines would be essentially absence of a liquid film on the walls of the I saturated, the mass flow rate corresponding to flow channel at high quality makes the two phase the upper saturation envelope line is the flow model invalid as it stands. The average appropriate one to use. Table 3E.41 shows the static pressure in the flow path is going to be mass flow rates for a range of (L/Dh values for something in excess of 500 psia if the initial a stagnation pressure of 1000 psi which is pressure is 1000 psla; this depends on the roughly equal to the pressure in an ABWR piping amount of flow choking and can be determined system carrying steam. from Reference 5. However, a fair estimate of l

(vf/vg) 1/3 is = 0.3, so the friction l A major uncertainty in calculating the leakage factor for :sturated steam blowdown may be taken rate is the value of f. This is discussed next, as 0.3 of that for mixed flow.

3E.4.2.2 Selection of ApproprLate Friction Based on this discussion, a coefficient of ,

Factor friction of 0.15 x 0.3 = 0.045 was used in the l

flow rate estimation. Currently experimental l Typical relationships between Reynolds' Number data are unavailable to validate this assumed I and relative roughness e/D h, the ratio of value of coefficient of friction.

O Amendment 1 3fL4 2

MkN 23A6100AE Standard Plant REV.A q "

• L3 Crack Opening Area Formulation k./ (#) (3E.4-4) ,

Tt. . rack opening areas were calculated using Ab " #!! . w . R 2,(3 + cosp) gt

' EFM procedures with the customary plastic zone

, E 4 correction. The loadings included in the crack opening area calculations were: pressure, weight where, and thermal expansion. i ab = bending stress due to weight and l The mathematical expressions given by Paris thermal expansion loads 1 and Tada [7] are used in this case. The crack I opening areas for pressure (Ap) and bending e is half crack angle l stresses (Ab ) were separately calculated and I then added together to obtain the total area, (3E.4-5)

A.c I(e) t = 20 2

1+ l */*

For simplicity, the calculated membrane ,

stresses from weight and thermal expansion loads l8.6 - 133 + 24 L )* ,4 were combined with the axial membrane stress, "Ij l a p, due to the pressure. ,

l 122.5 7 + 205.

The formulas are summarized below: I

- 247.5 + 242 1" Ap=# (2xRt) Gp (A) (3E.4-2) l (0 < $ < 100 )

where, The plastic zone correction was incorporated op = axial membrane stress due to by replacing a and e in these formulas by r.e pressure, weight and thermal and dewhich are given by expansion loads.

E = Young's modulus 2 deff = 8 + (3E.4-6) j R = pipe radius a oy' t = pipe thickness se=de.R A = shell parameter = a//Rt The yield stress, 'y, was conservatively assumed as the average of the code specified a = balf erack length yield and ultimate strecgth. The stress intensity f actor, K t otal, in clu d e s (3E.44) contribution d.e to both the membrane and Gp(A) = A2 + 0.16 A4(osA11)

= 0.02 + 0.81 A2 + 030 A3 4

+ 0.031 (l a A 5 5) Ktotal = Km+Kb (3E.4 7)

Amendment 1 3g43

ABWR uom l Standard Plant am'. A where, 6. Daughterly, R.L. and Franzini, J.B., "Fluid Mechanics with Engineering Applicationd,' 1 Km"#P [a". Fp(A) McGraw-Hill Book Company, New York 1%5.

Fp(A) = (1 + 03225 A' )$ 7. P.C. Paris aand H. Tada, 'The Application of Fracture Proof Design Postulating

= 0.9 + 0.25 A (oSXS1) Circumferential Through Wall Cracks,' U.S (1 SASS) Nuclear Regulatory Commission Report NUREG/CR4464, Washington, DC, April 1983.

Fb (e) = 1 + 6.8 '/'

- 13.6 s/2+20 /*

(0$e1100')

The steam mass flow rate, M, shown in Table 3E.4-1 is a function of parameter, ft/26. Once the mass flow rate is determined corresponding to the calculated value of this parameter, the leak rate in gpm can then be calculated.

3E.43 References

1. Norris, D., B. Cheral, T. Griesbach. 1987.

PICEP: Pipe Crack Evaluation Program, NP 3596 SR, Special Report, Rev.1, Electric h

Power Research Institute, Palo Alto, CA.

2. Chexal, B. & J. Horrmitz. A Critical Flow Model for Flow Through Cracks in Pipes, to be presented at the 24th ASME/AICHE National Heat Transfer Conference, Pittsburgh, Pennsyvania, August 9 12, 1987.

3. B. Chexal & J. Horowitz, "A Crack Flow Rate Model for Leak - Before - Break Applications,' SM;RT 9 Transachoir Vol. G, pp. 281285 (1987).

4. Collier, R.P., et al, 'Two Phase Flow Through Intergranular Stress Corrosion Cracks and Resulting Acoustic Emmision,' EPRI Report No. NP 3540 LD, April 1984.

5. Moody, F.3., ' Maximum Two-Phase Vessel Blowdown from Pipes,' 3. Heat Transfer, Vol.

88, No. 3,1966, pp. 285 295.

O Amendment 1 3EM

ABWR m3-Standard Plant any.A O TABLE 3E.41 MASS FLOW RATE FOR SEVERAL f1/Dh VALUES I

f1/Db MASS FLOW NATE, Ibm /sec ft.' l M l 0 3800 1 2200 2 1600 3 1150 4 920 0 5 .00 10 580 20 400 ,

l 50 260 l l

100 185 4

f 1

I i

O Arnendenent 1 3E43 l

ABWR zweioore Standard Plant REV A O

20 . . . . . . . - - . . . .

il g l .

BCL MACHINE BCLIGSCC * -

BEREKLEY SLIT " -

- CANADIAN - MAX. y * -

[o 15 . CANADIAN - AVG. / .

C Q

. CREC DATA

. WYLE DATA

[

u gn .

E IHIDATA a

10 -

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/ z

/u nN o

O . .f. M C

w -

5 ~ '

,w n _

~

~

N  :

+"' l

' ' ' ' ' I ' ' ' ' l ' ' ' '

0 O 5 10 15 20 MEASURED LEAK RATE (GPM) 87 592 21 l

Figure 3E.4-1 COMPARISON OF PICEP PREDICTIONS WITH O MEASURED LEAK RATES Amendment 1 3E.44 1

l

I l

i MkN 23A6100AE Standard Plant REV A l

l 0

ENTRANCE PROPERTIES EXIT p, PROPERTIES g Xi P2 S1 X2

• z

(%, G G Gu

.\  :%

ho LIQUID ho g

8 g

4gg#e:.1 / VAPOR STAGNATION r i.  :

PROPERTIES 8 i Po ho ' ' *

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,,,g;;f ,-

O p:

< l

!i f: ISENTROPIC  :  : L -g ENTRANCE i i e 1  :

STATION 1 STATION 2 87 592 22 Figure 3E.4-2 PIPE FLOW MODEL Amendment 1 3E47

1 MNN 2M6100AE REV.A i

Standard Plant {

l l

fLD=0

- SATURATION ENVELOPE j $?g8h = - - ~ . I i ,000 1 # !# fp ' -

M000 s

( -- @ .go >

500 - l 5  ?/

2 i h MAX FLOW RATE Gu (Ibu see ft 3

-C ---INLET STATIC PRESSURE P (psia) t 0

O 1000 2000 3000 INLET STAGNATION PRESSURE Po (psla)

(a)

O _

fLD=1

~8 ((- SATURATION ENVELOPE 3 -

1- - / 7

$ .., l -

E 1000 ---

I #

+

[ g d'#@@ o N ),

D .

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\

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M8 A 9 500 -

~

h f $

?

$ ly MAX FLOW RATE Gu (Ibu see-ft )

$ ---lNLET STATIC PRESSURE P (psia) i n.

I I I I I I I ' ' ' ' '

O O 1000 2000 3000 INLET STAGNATION PRESSURE Po (psia)

(b) 87 592 23 Figure 3E.4-3 MASS FLOW RATES FOR STEAM / WATER MIXTURES Amendment 1 3E44

1 ABWR

• I Standard Plant l

h VALUES OF (D"V) FOR WATER AT 60'F (DIAM IW Inches x VELOCITY IN fps) = Na/6,839 0.10.20.4' 1 2 4 10 20 40 100 200 400 1000 2000 4000 10,000 VALUES OF (D"V) FOR ATMOSPHERIC AIR AT 60'F 2 46 10 20 40 60 100 200 400 1000 2000 4000 10,000 40,000 100,000 09 "

0.08

_f,.pfm ~ ;

,,y 3 7, g . g. = _ _._.

g

..cou ,Ls l

": Tokoutswer noucs pines, . _ _ __  !

q ,

0.07

M -. i_ ,

4 0.05

\ h$! w._ --~

N ,

0.04

.c" >

[ 0.06 0 05 --

g 0.02 Q*

,\ g _

s' 0.015 b 0.04 - --Q 5 4' -

4, q  ;._ b .

O.01  !

c w 0.008 2 5

e 0 03 -- -- ---

fA . s

\

l h N . s

\'

g ',

i Q , ., O.002 g 2

9 0.02 - -

i M1

'rb A s

0.001 O.0008 %

P b

g ivooTH PiPti ,

% L%5 '-

4_

s

,4 0.0006 L 0.0004 c-l

, 0.015  % s s

N .%

~

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-N ,,,,,

^

( , 0.0001 A -

0.00005 0.01 ,3, .,o g q , , ,s, i il . l .I 18 7,0 *?! , l. %W " __

o,oooog 4 6 10 3 2346 30 4 2 34 6 go s 2346 10 6 2346 10 7 2 34 6 jos ,

REYNOLDS NUMBER N, = DVN (D ft, V, fps; y, ft 8/sec) 87 592 24 1

l l

l l

l Figure 3E.4-4 FRICTION FACTORS FOR PlPES Amendment 1 3g4,9 I

ABM 23461mse arv.A Standard Plant O 3E.5 LEAK DETECTION CAPABILITIES A complete description of various leak detection systems is provided in Subsection 5.2.5. The leakage detection system gives separate considerations to: leakage within the drywell and leakage external to the drywell. The limits for reactor coolant leakage are described in Subsection 5.2.5.4.

The total leakage consists of the identified leakage and the unirentified leakage. The identified leakage is that from pumps, valve stem packings, reactor vessel head seal and other seals, which all discharge to the equipment drain sump. The technical specification limit on the identified leak rate is expected to be 25 gpm.

The unidentified leak rate is the portion of the totalleakage received in the drywell sumps that is not identified as previously described.

The licensing (technical specification) limit on unidentified leak rate is expected to vary from 1 to 5 gpm. Therefore, leak rates of 5 and 10 gpm will be used in the leakage flaw size calculations performed in Appendix 3F to evaluate

("N v the margins against unstable flaw size and tearing instability load.

l l

l l

O 3E.51 Amendment 1

_ , , _ , a _ a - e a<. - m- w - -- - - w ----- -

O APPENDIX 3F APPLICATION OF LEAK-BEFORE-BREAK APPROACH TO O ABWR PIPING SYSTEMS

O

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

ABWR m6iman Standard Plant nrw A O

APPENDIX 3F TABLE OF CONTENTS Ses11an Iltle Eage 3F APPLICATION OF LEAK-BEFORE BREAK APPROACH TO ABWR PIPING SYSTEMS 3F.1 INTRODUCTION 3F.1 1 3F.2 MAIN STEAM PIPING 3F.2 1 3 F.2.1 System Description 3F.2-1 3F.2.2 Susceptibility to Water Hammer 3F.21 3F.23 Thermal Fatigue 3F.2-2 3F.2.4 Piping, Fitting, and Safe End Materials 3F.2 2 3F.2.5 LBB Margin Evaluation 3F.2-2 3F.2.6 Conclusion 3F.2-3 3FJ FEEDWATER SYSTF31 PIPING 3F3-1 3F.3.1 System Description 3F3-1 j 3F3.2 Susceptibility to Water Hammer 3F31 l 3F33 Thermal Fatigue 3F3-1 3F3.4 Piping, Fitting and Safe End Material 3F3-1 1

3F3.5 Piping Sizes, Geometries and Representative Stress 3F3.1 3F3.6 LBB Margin Evaluation 3F31 0 3F-il Amendment 1

NN 23A6100AE Standard Plant av 3 -

3F.1 INTRODUCTION As explained in Subsection 3.6.3, this appendix applies the methods of Appendix 3E to specific high en ,rgy piping systems in an ABWR Standard Nucicar Island plant to demonstrate their qualification for the leak before break (LBB) approach. These qualified piping systems meet the LBB criteria of Subsection 3.6.3 and are excluded from requirements for evaluation of dynamic effects (as identified in introduction to Section 3.6) resulting from a postulated break.

The LBB qualification has been demonstrated by showing that the through wall flaw lengths that are detectable by the leakage monitoring systems (See Subsection 5.2.5) are significantly smaller than the flaw lengths that could result into a pipe rupture or could become unstable.

f O

a

l 4

l j

! l 4

lO Amendment 1 3F.1 1

ABM 23A6100AE Standard Plant RIN. A -

O secTios 3r.2  :

CONTENTS Section Title East 3 F.2.1 System Descrintion 3F.21 i 3F.2.2 Suscentibility to Water Hammer 3F.2-1 3F.2.2.1 Safety Relief Valve Lift Transient Description 3F.21 L

3F2.2.2 Turbine Stop Valve Closure Transient Description 3F.2-1 l 3F.2.23 Basic Fluid Concept 3F.2 2 3F.23 Therraal Fatigue 3F.2-2 3F.2. l Pinina. Fittina. and Safe End Materials 3F.2-2 3F.2.5 LHH Marnin Evaluation 3F.2 2 3F.2.6 Conclusion 3F.2 3 0

, i I

i I

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ABWR memu Standard Plant arty. A SECTION 3F.2 g TABLES Table Title Eage 3F.2-1 Representative Stresses in Main Steam Line 3F.2-4 3F.2 2 Critical Crack Length and Instability tee 1 Margin Evaluations For h1ain Steam laae 3F.2.4 ILLUSTRATIONS Elgure Title Eage 3F.2 1 Leak Rate as a Function of Cr cl '.cagtn in Main Steam Pipe 3F.2 5 1

3F.2-iii 9' >

Y Amendment I

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ABM ux61oose ,

Standard Plant REV.A 1 3F.2 MAIN STEAM PIPING end of the discharge piping submerged in the suppression pool. Pressure waves traveling  !

O

, 3F.2.1 System Description through the discharge piping following the relatively rapid opening of the SRVs causes the l

(

The four main steam (hf S) lines carry steam discharge piping to vibrate. This in turn from the reactor to the turbine and auxiliary produces time dependent forces that act on the systems. Th: reactor coolant pressure boundary main steam piping segments.

portion of each line being evaluated in this section includes a flow restrictor is designed to There are a number of events / transients /

the limit rate of escaping steam from the postulated accidents that result in SRV lift:

postulated break in the downstream stream line.

The restrictor is also used for flow measurements a. Automatic opening signal when main steam during plant operation. The safety relief valves system pressure exceeds the set point (SRVs) discharge into the pressure suppression for a given valve (there are different pool through SRV discharge piping. The SRV set points for different valves in a safety function includes protection against over given plant).

pressure of the reactor primary system. The main steam line A has a branch connection to supply b. Automatic opening signal for all valves steam to the reactor core isolation cooling a s ,ig n e d to the automatic (RCIC) system turbine, depressurization system function on receipt of proper actuation signal.

This section addresses the MS piping system in the nuclear island which is designed and c. hianual opening signal to valve selected constructed to the requirements of the AShfE Code, by plant operator.

Section III, Class 1 piping (within cuarmost isolation valve) and Clau 2 piping. It is The SRVs close when the main steam system classified as Seismic Category I. It is pressure or when the plant operator manually inspected according to AShf E Code Section XI. releases the opening signals.

3F.2.2 Susceptibility to Water Hammer it is assumed (for conservatism) that all SRVs are activated by the same time, which Significant pressure pulsation of water hammer produces simultaneous forces on the main steam effect in the pipe may occur as a result of piping system.

opening of SRVs or closing of the turbine stop valve. A brief description of these phenomena 3F.2.2.2 Turbine Stop Vahe Closure Transient follows. These two transients are considered in Description the main steam piping system design and fatigue analysis. These events are more severe than the Prior to turbine stop valve cic" . ire, opening or closing of a main steam isolation saturated steam flows through each ma in steam valve or water carry over through main stream and line at nuclear boiler rated pressure and mass l SRV piping. hforeover, the probability of water flow rate. Upon signal, the turbine stop valves j carry over during core flooding in case of an close rapidly and flow stops at the upstream '

accident is low. side of these valves at the instant valve is I achieved. A pressure wave is created and 3F.2.2.1 Safety Relief Vahe Lift Transient travels c.t sonic VMocity toward the reactor Description vesrel t'prough each main stream line. Flow of steam into each main steam line from the reactor i SP.V ;produc9 inomentary unbalanced forces vessel nonsinues until the fluid compression a6. ting en the discharge piping system for the wave reaches the reactor vessel nozzle.

period frro the opedag et the SRV until a steady Repeated reflection of the pressure wave at the discharge now imm de reactor pressure vessel reactor vessel and stop valve ends of the main tr the suNtesuon punt is, established. This steam lines produce time varying pressures and period includes dcaring of the water slug the velocities at each point along the main steam a

Amendment t 3FS1 e

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ABWR asiaman Standard Plant anv.A lines. The combination of fluid momentum Evaluat?on of the ensuing effects are changes, shear forces, and pressure differences considered as a normal design process for the cause forcing functions which vary with position main steam piping system. The peak pressure and time, to act on the main steam piping pulses are within the design capability of a h

system. The fluid transient loads due to turbine typical piping design and the piping stresses stop valve closure is considered as design load and support loads remain within the ASME Code for upset condition. allowables.

3F.2.2.4 Basic Fluid Transient Concept It is concluded that, during these water hammer type events, the peak pressures and Despite the fact that the SRV and the turbine segment loads would not cause overstressed stop valve closure are flow starting and conditions for the main steam piping system.

flow stopping processes, respectively, the concepts of mass, momentum, and energy 3F.23 Thermal Fatigue conservation and the differential equations which represent these concepts are similar for both No thermal stratification and thermal fatigue problems. The particular solution for either of are expected in the main steam piping since the problems is obtained by incorporating the there is no large source of cold water in these appropriate initial conditions and boundary loops. A small amount of water may collect in conditions into the basic equations. Thus, the near horizontal leg of the main steam line relief valve discharge and turbine stop valve due to steam condensation. However, a slope of closure are seen to be specific solutions of the 1/8 inch per foot of main steam piping is more general problem of compressible, non. steady provided in each main steam line. Water drain fluid flow in a pipe. lines are provided at the end of slope to drain out the condensate. Thus, in this case no The basic fluid dynamic equations which are significant thermal cycling effects on the main applicable to both relief valve discharge and steam piping are expected.

turbine stop valve closure are used with a g particular fluid boundary conditions of these 3F.2.4 Piping, Fittings and Safe End W occurrences. Step wise solution of these Materials equations generates a time history of fluid properties at numerous locations along the pipe. The material specified for the 28 inch main Simultaneously, reaction loads on the pipe are steam pipe is SA155 KCF70. The corresponding determined at each location corresponding to the specification for the piping fittings and position of an cibow, forgings are given as SA420, WPL6 and SA350, LF2, respectively. The material for the safe The computer programs RVFOR and TSFOR end forging welded between the main steam piping described in Appendix 3D are used to calculate and the steam cozzle is SA508 Class 3.

the fluid transient forces on the piping system due to safety relief valve discharge and turbine 3F.2.5 Lilll Margin Evaluation stop valve closure. Both of the programs use method of characteristics to calculate the fluid The Code stress analysis of the piping will transients. be reviewed to obtain representative stress magnitudes- For the LBB evaluation purpose only The results from RVFOR program have been one pipe size, i.e. 28 inch, will be verified with various inplant test measurements considered. Table 3F.21 shows the example such as Monticello tests and Caoroso tests and stress magnitude due to pressure, weight, the test sponsored by BWR oner for NUREG 0737 at thermal expansion and SSE loads.

Wyle test facilities, Huntsville, Alabama. l Various data from the strain gages on the pipes The leak rate calculations were performed l and the load cells on the supports were compared assuming saturated steam conditions at 1050 with the analytical data and found to be in good psi. The leak rate model for saturated steam correlation, developed in Section 3E.4.2 was used in this O l Amendment 1 E22 l l

l 1

1 a

l M 23A6100Ali l Standard Plant RI!v. A l evaluation. Pressure, weight and thermal l expansion stresses were included in calculating l O1-the crack opening area. A plot of leak rate as a function of crack size was developed and is shown in Figure 3F.21. Leakage flaw lengths corresponding to 5 and 10 gpm were determined from this figure.

The calculations for the critical flaw size and instability load corresponding to leakage size cracks were performed using the J T methodology. Specifically, the $500F J.R curve shown in Figure 3E.2-8 and the Ramberg Osgood parameters given in Subsection 3E.3.2.1 were used. A plot of instability tension and bending stresses as a function of crack length was developed. Table 3F.2 2 shows the calculated critical crack size and the margin along with the instability load margin for the leakage size cracks. It is noted that the critical crack size margin is greater than 2 and the instability load margin also exceeds [2.

3F.2.6 Conclusion i For all four loops of the main steam system, ,

leakage rates of 5 and 10 gpm are used in the LBB  :

O v

evaluation based on the limit of satisfactory detection of the associated unidentified leakage. Based upon these leakage rates and representative stress magnitudes, leakage flaw lengths are calculated for 26 inch pipe and compared against the critical flaw length. The ,

margin is shown to be greater than 2 for both '

leakage rates. Also, the leak size crack stability evaluation showed a margin of at least 15.

It is also shown that other LBB criteria of  ;

Section 3.6.3 includir.g immunity to failure from effects of IGSCC, water hammer and thermal fatigue, and capability for leak detection are satisfied. Therefore, all four loops of the main ,

steem piping qualify for the leak before-break '

postulation approach.

O Amendment 1 3r3 3  ;

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ABWR m6ime Standard Plant RIN. A O TABoe 3e.21 REPRESENTATIVE STRESSES IN THE MAIN STEAM LINE Long. Weight +

Nominal Pipe Nominal Pressure Thermal SSE Pipe O.D Thickness Stress Expansion Stress Size (in) (in) (ksi) Stress (ksi)

(in) (ksi) 28 28.0 132 5.17 3.0 5.0 TABLE 3F.2 2 CRITICAL CRACK LENGTil AND anSTABILITY LOAD MARGIN EVALUATIONS FOR MAIN STEAM LINE O

Margins on Reference leakage Critical Instability l Pipe Reference Crack Crack Bending Load2 at Size Leak Rate length length Stress,Sb Critical leakage (in) (gpm) (in) (in) (ksi) Crack Crack 28 5 11.1 30.7 27.4 2.8 2.5 10 13.45 30.7 24.4 23 2.2 NOTES:

1. Based on Equation 3E3-9a.

2. Based on Equation 3E3-9b.

O Amendment 1 3F.2 4

23A6100AE Signdard Plant nnv A O

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0 0 2 4 6 8 10 12 14 16 CR ACK LENGTH (in.)

LEAK RATE AS A FUNCTION OF CRACK LENGTH O Figure 3F.2-1 IN MAIN STEAM PIPE Amendment 1 3F.2 5

23A6100AE Standard Plant anv. A l

l Q SECTION 3F3 CONTENTS l

Section Iltle Page l l

3F3.1 System Description 3F3-1 i l

3F.3.2 Susceptibility to Water Hammer 3F31 )

3FJJ ibermal Fatigue 3F3-1 l

3F3.4 Pining. Fittine and Safe End Material 3F31 3F3.5 Pininn Sizes. Geometries and Reoresentatise Stress 3F3.1 3 F.3.6 LHB Maruin Evaluation 3F3-1 3F3.7 Conclusion 3F3-1 TABLES O radie riti, rose 3F31 Data for Feedwater System Representation Pipe Size 3F3-3 3F3 2 Representative Stresses in Feedwater Line 3F3-3 3F3-3 Critical Crack 1.cngth and Instability Load Margin Evaluations for Feedwater line 3F3-4 l

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\ 3F3-il Amendment 1 1

ABW1 nA6im^tt Standard Plant REV.A 3F.3 FEEDWATER SYSTEM PIPING temperatures, pressures and thickness for representative pipe sizes in the feedwater Q 3F.3.1 System Description system. The nominal thickness for both pipe sizes correspond to schedule 80. Table 3F.3 2 The function of the feedwater (FW) system is shows the representative stress magnitudes for to conduct water to the reactor vessel over the each pipe size due to pressure, weight, thermal full range of the reactor power operation. The' expansion and SSE loads. Only the pressure feedwater piping consists of two 22 inch diameter weight and thermal expansion stresses are used lines from the high pressure feedwater heaters in the leak rate evaluation, where an sum of all connecting to the reactor vessel through three stresses is used in the instability load and 13 inch risers on each line. Each line has one critical flaw evaluation.

check valve inside the containment drywell and one positive closing check valve outside 3F.3.6 LHB Margin Evaluation containment. During shutdown cooling mode, reactor water pumped through the RHR heat The incoming water of the feedwater system is exchanger in one loop is returned to the vessel in a subcooled state. Accordingly, the leakage by way of one feedwater line, flaw length calculations were based on the procedure outlined in Section 3E.4.1. The This section addresses the feedwater piping in saturation pressure, Psat, for each pipe size the nuclear island, extending from the vessel out was calculated from the normal operation to the outboard isolation valve (ASME Class 1) temperatures given in Table 3F.31. 'Ihe leak and further through the shutoff valve to and rates were calculated as a function of crack including the seismic interface restraint (ASME length. The leakage flaw lengths corresponding Class 2). This section of the feedwater piping to 5 and 10 gpm leak rates were then determined.

is classified as Seismic Category 1.

The calculations for the critical flaw size 3F.3.2 Susceptibility to Water Hammer and the instability load corresponding to leak.

q age size cracks were performed using the J T C/ There is no record of feedwater piping failure methodology. Specifically, the J T curve shown due to water hactmer. Although there are several in Figure 3E.2-9 and the Ramberg Osgood para-check valves in the feedwater system, operating meters given in Subsection 3E.3.2.2 were used.

procedure and the control systems have been Table 3F.3 3 shows the calculated critical crack designed to limit the magnitude of water hammer sizes, and the margins along with the instabi-load to the extent that a formal design is not lity load margins for the leakage size cracks.

regt. ire d. Results are shown for both the 22-inch and '

12-inch lines, it is noted that for the two 3F.3.3 Thermal Fatigue reference leak rates, the critical crack size margin is greater than 2 and the instability Thermal fatigue is not a concern in ABWR feed- load margin also exceeds M.

water piping. The ASME Code evaluation includes 1 operating temperature transients, cold and hot 3F.3.7 Conclusion  ;

water mixing and thermal stratification.

LBB cvaluation has been conducted using two i 3F.3.4 Pipings, Fittings and Safe End values of reference leak rates: 5 and 10 gpm. l Material Based upon these leakage rates and representa-live stress magnitudes, leakage flaw lengths The material for piping is either SA333, Gr. were calculated for 22-inch and 12 inch lines.

6, or SA 672, Gr. C70. Comparison with critical crack lengths showed margin to be greater than 2. Leakage size crack 3F.3.5 Piping Sizes, Geometries and stability evaluation showed a margin of at least Representative Stress 4 Table 3F.3-1 shows the normal operating It has been also demonstrated in the Amendment 1 F.3-1

ABWR m6-u Standard Plant Rl!V. A preceding subscetions that the feedwater line meets other LBB Criteria of Section 3.6.3 including immunity to failure form effects of IGSCC, water hammer and thermal fatigue.

O O

Amendment 1 3F.12

ABM 23461oose Standard Plant REY.A O

TABLE 3F.31 DATA FOR FEEDWATER SYSTEM REPRESENTATION PIPE SIZE l

Nominal Pipe Nominal Nominal Operating Pipe O.D Rickness Temperature Pressure Size (in) (in) (OF) (psig)

(in) 12 12.75 0.687 420 1100 22 22.0 1.031 420 1100 1

TABLE 3F.3 2 REPRESENTATIVE STRESSES IN FEEDWATER LINE O

Nominal Longitudinal Weight + Safe Shut-down Pipe Pressure Hermal Earthquake (SSE)

Size Stress Expansion Str ess (ksi) Stress (ksi)

(ksi)

I 12 5.1 4.0 5.0 j 22 5.4 4.0 5.0 l

I I

O Amendment 1 3F.3-3 1

ABWR M ONAE S.gndard Plant REV.A 9

TABLE 3F.3 3 CRITICAL CRACK LENGTH AND INSTABILITY LOAD MARGIN EVALUATIONS FOR FEEDWATEP LINE Man; ins on l Reference Leakage Critical Instability l i Pipe Reference Crack Crack Bending Load2 at l Size leak Rate length Length Stress 4b Critical leakage l (in) (gpm) (in) (in) (ksi) Crack Crack 12 5 4.5 13.1 27.2 2.9 23 10 5.7 13.1 24.0 23 2.1 l l

22 5 5.2 20.4 27.9 3.9 23 1 10 6.7 20.4 25.6 3.1 2.2 NOTES:

1. Based on Equation 3E3-9a.

2. Based on Equation 3E3-9b.

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O APPENDIX 3G SEISMIC SYSTEM ANALYSIS RESULTS (TO BE PROVIDED BY DECEMBER 31,1988)

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O APPENDIX 3H DESIGN DETAILS AND EVALUATION RESULTS F SEISMIC CATEGORY I STRUCTURES O

(TO BE PROVIDED BY DECEMBER 31,1988)

O

O APPENDIX 3I EQUIPMENT QUALIFICATION ENVIRONMENTAL ESIGN CRITERIA O

(TO BE PROVIDED BY DECEMBER 31,1988) 1 0

O l

i CHAPTER 3 DESIGN OF STRUCTURES, COMPONENTS, O EQUIPMENT, AND SYSTEMS l

l i

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MkN 23A6100At!

Standard Plant niw. A CHAPTER 3  !

O V TABLE OF CONTENTS Section 'lltic Page 3

DESIGN EQUIPMENT,OF STRUCTURES, AND SYSTEMSCOMPONENTS, 3.1 CONFORNIANCE WITil NRC GENERAL DESIGN CRITERIA 3.11 3.1.1 Summary Description 3.1-1 3.1.2 Evaluation Against Criteria 3.1-1 3.2 CLASSIFICATION OF STRUCTURES. C051PONENTS.

AND SYSTENIS 3.21 3.2.1 Seismic Classification 3.2-1 3.2.2 System Quality Group Classifications 3.2-1 3.23 Safety Classifications 3.21 3.2.4 Correlation of Safety Classes with Industry Codes 3.23 3.2.5 Non Safety Related Structures, Systems, and Components 3.2-3 3.2.6 Ouality Assurance 3.2-5 33 WIND AND TORNADO I.OADINGS 33.1 33.1 Wind Loadings 33-1 33.2 Tornado Loadings 33-1 333 Interfaces 33-2 33.4 References 33-2 3.4 WATER LEVEL, (FLOOD) DESIGN 3.41 l

3.0-li l

Amendment 1

ABWR 334.ioois Standard Plant REY.A CHAPTER 3 TABLE OF CONTENTS (Continued) O Section Illic Eggs 3.5 hilSSil.E PROTECTION 3.51 3.5.1 hiissile Selection and Description 3.5-1 3.5.2 Structures, Systems, and Components to be Protected from Externally Generated hiissiles 357 3.53 Barrier Design Procedures 3.5-7 3.5.4 Interfaces 357 3.5.5 References 358 3.6 PROTICTION AGAlNST DYNA 511C EITECIS ASSOCIATED WITil Tile POSTULATED RUITURE OF PIPING 3.61 3.6.1 Postulated Pipping Failures in Fluid Systems Inside and Outside of Containment 3.6-1 3.6.2 Determination of Break 1.ocations and 9 Dynamic E.Tects Associated with the Postulated Rupture of Piping 3.6-6 3.63 leak Before. Break Evaluation Procedures 3.6 22 3.6.4 References 3.6-27 3.7 SEISS11C DESIGN 3.71 3.7.1 Seismic Input 3.7 1 3.7.2 Seismic System Analysis 3.7-4 3.0.iii h Amendment I

,- MM 23A6100A!!

Standard Plant artv. A CHAPTER 3 O TABLE OF CONTENTS (Continued)

Section Title Page 3.7.3 Seismic Sabsystem Analysis 3.7-14 3.7.4 Seismic Instrumentation 3.7 24

(

3.7.5 Interfaces 3.7 26 3.7.6 References 3.7 26 3.8 DESIGN OF SEISN11C CATEGORY I STRUCTURES 3.8-1 3.8.1 Concrete Containment 3.8 1 3.8.^ Steel Components of the Reinforced Concrete Containment 3.8-11 3.8.3 Concrete and Steel Internal Structures of the Concrete Containment 3.8-16 3.8.4 Other Seismic Category I Structures 3.8-20 3.8.5 Foundations 3.8 27 3.9 51ECIIANICAL SYSTENTS AND CON 1PONENTS 3.9-1 3.9.1 Special Topics for hfechanical Components 3.9-1 l 1

3.9.2 Dynamic Testing and Analysis 3.9-3 3.93 AShiE Code Class 1,2 and 3 Components, Component Supports, And Core Support Structures 3.9 18 3.9.4 Control Rod Drive System (CRDS) 3.9 35 i l

Reactor Pressure Vessels Internals 3.9.5 3.9-36 3.9.6 Insenice Testing of Pumps and Valves 3.9-44 3.9.7 References 3.9-45 O 3.0-iv Amendment 1

ABM DA6100AH Standard Plant IlliV. A CHAPTER 3 TAIILE OF CONTENTS (Continued)

Section Title Enge 3.10 SEISN11C OUALIFICATION OF SEIShilC CATEGORY I INSTRUh1ENTATION AND ELECTRICAL EOUIPh1ENT (INCLUDING OTilER DYNAN11C LOADS) 3.10-1 3.10.1 Seismic Qualification Criteria (Including Other Dynamic Loads) 3.10-1 3.10.2 Methods and Procedures for Qualifying Electrical Equipment and Instrumentation 3.10 2 3.10 3 Metliods and Procedures of Analysis or Testing of Supports of Electrical Equipment and Instrumentation 3.10-5 3.10.4 Operating License Review (Tests and Analyses Results) 3.10-8 3.10.5 Interfaces 3.10-8 3.11 ENVIRON 51 ENTAL DESIGN OF SAFE 7Y RELATED h1ECilANICAL AND ELECTRICAL EOUIP51ENT 3.11 1 3.11.1 Equipment IdentiGcation and Emironmental Conditions 3.11 1 3.11.2 Oualification Tests and Analyses 3.11-1 3.11.3 OucliDeation Test Results 3.11-2 3.11.4 Loss of IIcating, Ventilating, and Air Conditioning 3.11 2 3.11.5 Estimated Chemical and Radiation Emironment 3.11-2 3.11.6 Interfaces 3.11 3 3.11.7 References 3.11 3 l 3.0-v Amendment 1 l

23A6100Ali Standard Plant tuiv. A CHAPTER 3 l TABLE OF CONTENTS (Continued)

Section Title ,

1 APPENDIX 3A SEISS11C SOIL STRULTURE INTERACTION ANALYSIS APPENDIX 3H CONTAINh1EST LOADS )

APPENDIX 3C COh!PUTER PROGRAh!S USED IN Tile DESIGN AND ANALYSIS OF SEISh11C CATEGORY I STRUCTURE j APPEND:X 3D C051PUTER PROGRAh1S USED IN TIIE DESIGN OF CO51PONENTS, EQUIPA1ENT AND STRUCTURES APPENDIX 3E FRACTURE 51ECilANICS, LEAK RATE CALCULATION AND LEAK DETELTION h1ETilODS APPENDIX 3F APPLICATION OF LEAK BEFORE.HREAK APPROACll TO ABWR PIPING SYSTEhtS APPENDIX 3G SEIS511C SYSTEh! ANALYSIS RESULTS APPENDJLI DESIGN DETAILS AND EVALUATION RESULTS OF SEISN11C CATEGORY I STRUCTURES APPENDIX 31 EQUIPhlENT QUALIFICATION ENVIRONh1 ENTAL DESIGN CRITERIA O 3.0-si Amendment I

1 23A6100AE Standard Plant anv. A SECTION 3.1 O coureurs  ; Section Title Eage 3.1.1 Summary Descrintion 3.11 3.1.2 Evaluation Analnst Criteria 3.1 1 3.1.2.1 Group I - Overall Reqt.irements 3.1 1 3.1.2.1.1 Criterion 1 - Quality Stand irds and Records 11 1 3.1.2.1.1.1 Criterion 1 Statement 3.1-1 3.1.2.1.1.2 Evaluation Against Criterion 1 3.1 1 3.1.2.1.2 Criterion 2 - Design Bases for Protection Against Natural Phenomena 3.1-2 3.1.2.1.2.1 Criterion 2 Statement 3.1 2 3.1.2.1.2.2 Evaluation Against Criterion 2 3.1 2 0 31212 criterie 3 rire eret ti - 31-2 3.1.2.1.3.1 Criterion 3 Statemeat 3.13 3.1.2.1.3.2 Evaluation Against Criterion 3 3.13 3.1.2.1.4 Criterion 4 - Emironmental and Missiles Design Bases 3.1 3 l 3.1.2.1.4.1 Criterion 4 Statement 3.1 3 3.1.2.1.4.2 Evaluation Against Criterion 4 3.1-4 3.1.2.1.5 Criterion 5 Sharing of Structures, Systems, and Components 3.1-4 3.1.2.1.5.1 Criterion 5 Statement 3.1-4 3.1.2.1.5.2 Evaluation Against Criterion 5 3.1-4 3.1.2.2 Group II - Protection by Multiple Fission Product Barriers 3.1-4 l l O 3.1.ii l Amendment 1

1 l l ABWR usaman l Standard Plant iu?V. A l l SECTION 3.1 l CONTENTS (Continued) l Section Title lhige l I l 3.1.2.2.1 Criterion 10 Reactor Design 3.1 4 3.1.2.2.1.1 Criterion 10 Statement 3.1 4 3.1.2.2.1.2 Evaluation Against Criterion 10 3.15 3.1.2.2.2 Criterion 11 Reactor inherent Protection 3.15 l 3.1.2.2.2.1 Criterion 11 Statement 3.15 3.1.2.2.2.2 Evaluation Against Criterion 11 3.15 3.1.2.23 Criterion 12 Suppression of Reactor Power Oscillations 3.1 6 3.1.2.2.3.1 Criterion 12 Statement 3.1-6 3.1.2.2.3.2 Evaluation Against Criterion 12 3.16 3.1.2.2.4 Criterion 13 Instrumentation and Controi 3.1 7 h 3.1.2.2.4.1 Criterion 13 Statement 3.1-7 3.1.2.2A ? Evaluation Against Criterion 13 3.17 3.1.2.2.5 Criterion 14 - Reador Coolant Pressure Boundary 3.1-8 3.1.2.2.5.1 Criterion 14 Statement 3.1 8 3.1.2.2.5.2 Evaluation Against Criterion 14 3.1 8 3.1.2.2.6 Criterion 15 Reactor Coolant System Design 3.1 9 3.1.2.26.1 Criterion 15 Statement 3.1-9 3.1.2.2.6.2 Evaluation Against Criterion 15 3.1-9 3.1.2,2.7 Criterion 16 Containment Design 3.1 10 0 3.1 iii Amendment 1

ABWR m6ime Standard Plant REV.A SECTION 3.1 O cowrENTs (continued) Section Title Page 3.1.2.2.7.1 Criterion 16 Statement 3.1 10 3.1.2.2.7.2 Evaluation Against Criterion 16 3.1 10 3.1.2.2.8 Criterion 17 - Electric Power Systems 3.1 10 3.1.2.2.8.1 Criterion 17 Statement 3.1 10 3.1.2.28.2 Evaluation Against Criterion 17 3.1 11 3.1.2.2.8.2.1 Onsite Electric Power System 3.1 11 3.1.2.2.8.2.2 Offsite Electric Power System 3.1 11 3.1.2.2.9 Criterion 18 - Inspection and Testing of Electric Power Systems 3.1 12 3.1.2.2.9.1 Criterion 18 Statement 3.1 12 3.1.2.2.9.2 Evaluation Against Criterion 18 3.1 12 3.1.2.2.9.2.1 Onsite Electric Power System 3.1 12 3.1.2.2.9.2.2 Offsite Electric Power System 3.1-12 3.1.2.2.10 Criterion 19 - Control Room 3.1 12 3.1.2.2.10.1 Criterion 19 Statement 3.1-12 3.1.2.2.10.2 Evaluation Against Criterion 19 3.1-12 I l 3.1.23 Group Ill - Protection and Reactivity Controf System 3.1-13 3.1.23.1 Criterion 20 Protection System Functions 3.1 13 3.1.23.1.1 Criterion 20 Statement 3.1 13 3.1.23.1.2 Evaluation Against Criterion 20 3.1-13 0 3.1 iv Amendment i

ABWR m6=u Standard Plant su!v. A SECTION 3.1 CONTENTS (Continued) Section Title P. age 3.1.23.2 Criterion 21 Protection System Reliability and Testability 3.1 14 3.1.23.2.1 Criterion 21 Statement 3.1 14 3.1.23.2.2 Evaluation Against Criterion 21 3.1 14 3.1.233 Criterion 22 Protection System Independence 3.1 16 3.1.233.1 Criterion 22 Statement 3.1 16 3.1.233.2 Evaluation Against Criterion 22 3.1 16 3.1.23.4 Criterion 23 Protection System Failure blodes 3.1 17 3.1.23.4.1 Criterion 23 Statement 3.1-17 3.1.23.4.1 Evaluation Against Criterion 23 3.1 17 3.1.23.5 Criterion 24 - Separation of 9 Protection and Control Systems 3.1 17 3.1.23.5.1 Criterion 24 Statement 3.1 17 3.1.23.5.2 Evaluation Against Criterion 24 3.1 17 3.1.23.6 Criterion 25 Protection System Requirements for Reactivity Control hlalfunctions 3.1 18 3.1.23.6.1 Criterion 25 Statement 3.1 18 3.1.23.6.2 Evaluation Against Criterion 25 3.1-18 3.1.23.7 Criterion 26 Reactivity Control System Redundancy and Capability 3.1 19 3.1.23.7.1 Criterion 26 Statement 3.1 19 3.1.23.7.2 Evaluation Against Criterion 26 3.1 19 l O1 3.1v Amendment 1 l

23A6100A!! Standard Plant artv. A SECTION 3.1 O CONTENTS (Continued) Section Title Page 3.1.23.8 Criterion 27 Combined Reactivity ControlSystems Capability 3.1 20 3.1.23.8.1 Criterion 27 Statement 3.1 20 3.1.23.8.2 Evaluation Against Criterion 27 3.1 20 3.1.23.9 Criterion 28 Reactivity Limits 3.1 21 3.1.23.9.1 Criterion 28 Statement 3.1-21 3.1.23.9.2 Evaluation Against Criterion 28 3.1 21 3.1.2 3.10 Criterion 29 - Protection Against Anticipated Operational Occurrences 3.1 22 3.1.2 3.10.1 Criterion 29 Statement 3.1-22 3.1.2 3.10.2 Evaluation Against Criterion 29 3.1 22 3.1.2.4 Group IV Fluid Systems 3.1-22 3.1.2.4.1 Criterion 30 - Quality of Reactor , Coolant Pressure Boundary 3.1 22 l 3.1.2.4.1.1 Criterien 30 Statement 3.1-22  ! 3.1.2.4.1.2 Evaluation Against Criterion 30 3.1 23 3.1.2.4.2 Criterion 31 - Fracture Prevention of Reactor Coolant Pressure Boundary 3.1 23 3.1.2.4.2.1 Criterion 31 Statement 3.1-23 3.1.2.4.2.2 Evaluation Against Criterion 31 3.1 24 3.1.2.43 Criterion 32 Inspection of Reactor , Coolant Pressure Boundary 3.1 24 l 3.1.2.43.1 Criterion 32 Statement 3.1 24 3.1.2.43.2 Evaluation Against Criterion 32 3.1 24 0 3.1si Amendment 1

ABWR memm Standard Plant imv. A SECTION 3.1 CONTENTS (Continued) O Sectin.n Hils Eage 3.1.2.4.4 Criterion 33 - Reactor Coolant Makcup 3.1 25 3.1.2.4.4.1 Criterion 33 Statement 3.1 25 3.1.2.4.4.2 Response to Criterion 33 3.1 25 3.1.2.4.5 Criterion M Residuall{ cat Removal 3.1 25 3.1.2.4.5.1 Criterion 34 Statement 3.1 25 3.1.2.4.5.2 Evaluation Against Criterion 34 3.1 26 3.1.2.4.6 Criterion 35 Emergency Core Cooling 3.1 27 3.1.2.4.6.1 Criterion 35 Statement 3.1 27 3.1.2.4.6.2 Evaluation Against Criterion 35 3.1 27 3.1.2.4.7 Criterion 36 Inspection of Emergency Core Cooling System 3.1-28 3.1.2.4.7.1 Criterion 36 Statement 3.1 28 0 3.1.2.4.7.2 Evaluation Against Criterion 36 3.1 28 3.1.2.4.8 Criterion 37 Testing of Emergency Core Cooling System 3.) 29 3.1.2.4.8.1 Criterion 37 Statement 3.1 29 3.1.2.4.8.2 Evaluation Against Criterion 37 3.1 29 3.1.2.4.9 Criterion 38 Containment Ileat Removal 3.1 29 3.1.2.4.9.1 Critcrion 38 Statement 3.1 30 3.1.2.4.9.2 Evaluation Against Criterion 38 3.1 30 3.1.2.4.10 Criterion 39 Inspection of Containment lleat Removal System 3.1 30 0 3.1 vii l Amendment I

                       =                                                      .-

ABM ' ua6iooai Standard Plant RiiV. A SECTION 3.1 O CONTENTS (continued) Section Title East 3.1.2.4.10.1 Criterion 39 Statement 3.1-30  ; l 3.1.2.4.10.2 Evaluation Against Criterion 39 3.1 30  : 3.1.2.4.11 Criterion 40 - Testing of Containment Heat Removal 3.1 31 1 3.1.2.4.11.1 Criterion 40 Statement 3.1 31 3.1.2.4.11.2 Evaluation Against Criterion 40 3.1-31 3.1.2.4.12 Criterion 41 - Containment Atmosphere Cleanup 3.1 31 3.1.2.4.12.1 Criterion 41 Statement 3.1 31 3.1.2.4.12.2 Evaluation Against Criterion 41 3.1-31 3.1.2.4.13 Criterion 42 Inspection of Containment Atmosphere Cleanup System 3.1 32 3.1.2.4.13.1 i 42 Statement 3.1 32 3.1.2.4.13.2 e un Against Criterion 42 3.1 32 3.1.2.4.14 Criterion 43 - Testing of Containment Atmosphere Cleanup System 3.1 32 3.1.2.4.14.1 Criterion 43 St.atement 3.1-32 3.1.2.4.14.2 Evaluation Against Criterion 43 3.1 32 3.1.2.4.15 Criterion 44 Cooling Water 3.1 33 l 3.1.2.4.15.1 Criterion 44 Statement 3.1 33 l 3.1.2.4.15.2 Evaluation Against Criterion 44 3.1-33 3.1.2.4.16 Criterion 45 Inspection of Cooling Water System 3.1 33 3.1.2.4.16.1 Criterion 45 Statement 3.1 33 l O 3.1 viii Amendment 1

ABWR m6iun Standard Plant niw. 6 SECTION 3.1 CONTENTS (Continued) O Section Title l' age 3.1.2.4.16.2 Evaluation Against Criterion 45 3.1 33 3.1.2.4.17 Criterion 46 - Testing of Cooling Water System 3.1-34 3.1.2.4.17.1 Criterion 46 Statement 3.1 34 3.1.2.4.17.2 Evaluation Against Criterion 46 3.1 34 3.1.2.5 Group V Reactor Containment 3.1 34 3.1.2.5.1 Criterion 50 Containment Design Basis 3.1 34 3.1.25.1.1 Criterion 50 Statement 3.1 34 3.1.2.5.1.2 Evaluation Against Criterion 50 3.1 34 3.1.2.5.2 Criterion 51 - Fracture Prevention of Containment Pressure Boundary 3.1 35 3.1.2.5.2.1 Criterion 51 statement 3.1 35 3.1.2.5.2.2 Evaluation Against Criterion 51 3.1-35 3.1.2.5.3 Criterion 52 Capability for Contain-ment Leakage Rate Testing 3.1-35 3.1.2.5.3.1 Criterion 52 Statement 3.1 35 3.1.25.3.2 Evaluation Against Criterion 52 3.135 3.1.2.5.4 Criterion 53 - Provisions for Containment Testing and Inspection 3.1 36 3.1.2.5.4.1 Criterion 53 Statement 3.1 36 3.1.2.5.4.2 Evaluation Against Criterion 53 3.1 36 3.1.2.5.5 Criterion 54 Piping Systems Penetrating Containment 3.1-36 O 3.1-ix Amendment 1

23A6100AE Standard Plant REV.A SECTION 3.1 O courturs (contiooea> Section Title Page 3.1.2.5.5.1 Criterion 54 Statement 3.1 36 3.1.2.5.5.2 Evaluation Against Criterion 54 3.1 36 3.1.2.5.6 Criterion 55 Reactor Coolant Pressure Boundary Penetrating Containment 3.1 36 3.1.2.5.6.1 Criterion 55 Statement 3.1-36 3.1.2.5.6.2 Evaluation Against Criterion 55 3.1-37 3.1.2.5.7 Criterion 56 - Primary Containment Isolation 3.1-37 3.1.2.5.7.1 Criterion 56 Statement 3.1-37 3.1.2.5.7.2 Evaluation Against Criterion 56 3.1-38 3.1.2.5.8 Criterion 57 Closed System Isolation Valves 3.1 38 O 3.1.2.5.8.1 Criterion 57 Statement 3.1-38 3.1.2.5.8.2 Evaluation Against Criterion 57 3.1 38 3.1.2.6 Group VI - Fuel and Reactivity Contiol 3.1 38 3.1.2.6.1 Criterion 60 - Control of Releases of l Radioactive Materials to the Emironment 3.1-38 l i 3.1.2.6.1.1 Criterion 60 Statement 3.1 38 ) 3.1.2.6.1.2 Evaluation Against Criterion 60 3.1-38 3.1.2.6.1.2.1 Effluent Monitoring System 3.1 38 3.1.2.6.1.2.2 Radioactive Waste System 3.1-39 2.1.2.6.2 Criterion 61 Fuel Storage and Handling and Radioactivity Control 3.1-39 3.1.2.6.2.1 Criterion 61 Statement 3.1 39 O 3.1 x Amendment 1

ABWR m6iman 1 S_tandard Plant niw. A j SECTION 31 CONTENTS (Continued) Secupp Sic Bige 3.1.2.6.2.2 Evaluat.an Against Criterion 61 3.1-39 i 312.6.2.2.1 Fuel Storage and llandling System 3.1-39 3.1.2.6.2.2.2 Radioactive Waste System 3.1-40 l 3.1.2.63 Criterion 62. Prevention of Criticality , in Fuel Storage and llandling 3.1 40 l 3.1.2.63.1 Criterion 62 Statement 3.1-40 l 3.1.2.63.2 Evaluation Against Criterion 62 3.1 40 3.1.2.6.4 Criterion 63 - hionitoring Fuel and Waste Storage 3.1-40 3.1.2.6.4.1 Criterion 63 Statement 3.1-40 3.1.2.6.4.2 Evaluation Against Criterion 63 3.141 3.1.2.6.4.2.1 Fuel Storage System 3.1-41 3.1.2.6.4.2.2 Radioactive Waste System 3.1-41 3.1.2.6.5 Criterion 64 Ffonitormg Radioactivity Releases 3.1-41 3.1.2.6.5.1 Criterion 64 Ststement 3.1-41 31.2.6.5.2 Evaluation Against Cliterion 64 3.1 41 0 3.1 xi Amendment I

23A6100AE I Sandard Plant REV.A r 3.1 CONFORMANCEWITH NRC to safety shall be designed, fabricated, I \ GENERAL DESIGN CRITERIA erected, and tested to quality standards commensurate with the importance of the safety 3.1.1

SUMMARY

DESCRIPTION functions to be performed. Where generally l recognized codes and standards are used, they This section contains an evaluation of the shall be identified and evaluated to determine principal design criteria of the plant (ABWR their applicability, adequacy, and sufficiency Standard Nuclear Island) as measured against the and shall be supplemented or modified as NRC General Design Criteria for Nuclear Power necessary to assure a quality product in keeping Plants,10CFR50 Appendix A. The general design with the required safety function. A quality criteria, which are divided into six groups with assurance program shall be established and the last criterion numbered 64, are intended to implemented in order to provide adequate establish minimum rauirements for the principal assurance that these structures, systems, and design criteria for nuclear power plants. components will satisfactorily perform their safety functions. Appropriate records of the The NRC General Design Criteria were intended design, fabrication, erection, and testing of to guide the design of all water-cooled nuclear structures, systems, and components important to power plans; separate BWR specific criteria are safety shall be maintained by or under the not addressed. As a result, the criteria are control of the nuclear power unit licensee subject to a variety of interpretations. For throughout the life of the unit. this reason, there are some cases where conformance to a particular criterion is not 3.1.2.1.1.2 Evaluation Against Criterion I directly measurable. In these cases, the conformance of the ABWR Nuclear Island to the Structures, systems, and components important interpretation of the criteria is discussed. For to safety (i.e., that are safety related* as each criterion, a specific assessment of the defined in Section 3.2) are identified on Table plant design is made and a complete list of 3.2 1. The total quality assurance program is references is included to identify where detailed described in Chapter 17 and is applied to these design information pertinent to that criterion is items. The intent of the quality assurance treated in this safety analysis report (SAR). program is to assure sound engineering in all phases of design and construction through Based on the content herein, the design of the conformity to regulatory requirements and design ABWR Nuclear Island fully satisfies and is in bases described in the license application. In compliance with the NRC General Design Criteria, addition, the program assures adherence to specified standards of workmanship and 3.1.2 EVALUATION AGAINST CRITERIA implementation of recognized codes and standards in fabrication and construction. It also  ; 3.1.2.1 Group I Overall Requirements includes the observance of proper preoperational and operational testing and maintenance 3.1.2.1.1 Criterion 1 - Quality Standards and procedures as well as the documentation of the Records foregoing by keeping appropriate records. The l total quality assurance program is responsive to i 3.1.2.1.1.1 Criterion 1 Statement and in conformance with the intent of the I quality related requirements of 10CFR50 Appendix Structures, systems, and components important B. l The definition of important to-safety is an NRC unresolved generic licensing issue. In accordance with past industry practice "important-to safety" and "safety related" are considered equivalent in the evalu-O ations in this section and in this SAR. Amendment 1 3.11

ABWR ==n Sjandard Plant nitv. A j l Structures, systems, and components are (1) Appropriate consideration of the most severe identified in Section 3.2 with respect to their of the natural phenomena that have been his- g , location, service and their relationship to the torically reported for the site and sur- W safety related or non safety related function to rounding area, with sufficient margin for be rerformed. Recognized coJes and standards are the limited accuracy, quantity, and period applied to the equipment per the safety of time in which the historical data have classifications to assure meeting the required been accumulated; safety related function. (2) Appropriate combinations of the effects of Documents are maintlined which demonstrate normal and accident conditions with the that all the requirements of the quality effects of the natural phenomena; and assurance program are being satisfied. This documentation shows that appropriate codes, (3) The importance of the safety functions to be standards, and regulatory requirements are performed, observed, specified materials are used, correct procedures are utilized, qualified personnel are 3.1.2.1.2.2 Evaluation Against Criterion 2 provided, and the finished parts and components meet the applicable specifications for safe and Since the ABWR Nuclear Island is designated re:iable operation. These records are available as a standard plant, the design bases for so that any desired item of information is safety related (See Subsection 3.1.2.1.1.2) retrievable for reference. These records will be structures, systems, and components, cannot maintained during the life of the operating accurately reflect the most severe of the natu-licenses, ral phenomena that have been historically re-ported for each possible site and their sur-The detailed quality assurance program is in rounding areas. Ilowever, the envelope of site-conformance with the requirements of Criterion 1. related parameters which blanket the majority of potential sites in the conterminous United For further discussion, see the following States is defined in Chapter 2. The design sections: baserfor these structures, systems, and compo-g nents reflect this envelope of natural pher.cmena Chapter / including appropriate combinations of the ci-Section Iilk fccts of normal and accident conditions with this cavelope. The design bases meet the (1) 1.2 General Plant Description requirements of Criterion 2. (2) 3.2 Classification of Structures, Detailed discussion of the various phenomena Components, and Systems considered and design criteria developed are presented in the following sections: 3.1.2.1.2 Criterion 2 Design liases for Protection Against Natural Phenomena Chapter / bection liik 3.1.2.1.2.1 Criterion 2 Statement (1) 2.0 Summary of Site Characteristics Structures, systems, and components important to safety shall be designed to withstand the (2) 3.2 Classification of Structures, effects of natural phenomena such as carthquakes, Components, and Systems tornadoes, hurricanes, floods, tsunami, and seiches without loss of capability to perform (3) 3.3 Wind and Tornado loadings their safety functions. The design bases for these structure systems and components shall (4) 3.4 Water Level (Flood) Design reflect: O Arrendment 1 312

ABM 22^6 00^n Standard Plant any.A (5) 3.5 hiissile Protection of protective cabling if a fire occurs. The N arrangement of equipment in reactor protection (d (6) 3.7 Seismic Design channels provides physical separation to limit the effects of a fire. (7) 3.8 Design of Seismic CategoryI Structures Combustible supplies, such as logs, records, manuals, etc, are limited in such areas as the (8) 3.9 hiechanical Systems and control room to amounts required for current Components operation thus limiting the effect of a fire or explosion. (9) 3.10 Seismic Qualifications of Seismic Category I Instrumen- The plant fire protection system includes the tation and Electrical Equipment following provisions: (10) 3.11 Emironmental Qualification of (1) automatic fire detection equipment in those Safety Related hiechanical and areas where fire danger is greatest, and Electrical Equipment (2) extinguishing services which include auto-3.1.2.13 Criterion 3 Fire Protection matic actuation with manual override as well as manually operated fire extinguishers. 3.1.2.1.3.1 Criterion 3 Statement The design of the fire protection system Structures, systems, and components important meets the requirements of Criterion 3. For m safety shall be designed and located to mini- further discussion, see the following sections: mize, consistent with other safety requirements, the probability and effect of fires and explo- Chapter / sions. Noncombustible and heat-resistant mate- Ssstiq!1 Iilla I l rials shall be used whenever practical throughout the unit, pai:icularly in locations such as the (1) 3.8.2.6 Materials, Quality Control and . containment acd control room. Fire detection and Special Construction Techniques fighting systems of appropriate capacity and cap- 1 l ability shall be provided and designed to mini- (2) 7 Instrumentation and Control mize the adverse effects of fires on structures, Systems systems, and components important to safety. Fire fighting systems shall be desigced to assure (3) 8 Electric Power j that their rupture or inadvertent operatien does not significantly impair the safety capability of (4) 9.5 Fire Protection System these structures, systems, and components. (5) Appendix Fire Hazard Analysis 3.1.2.1.3.2 Evaluation Against Criterion 3 9A Fires in the plant are prevented or mitigated (6) 13 Conduct of Operations by the use of non combustible and beat resistant materials such as metal cabinets, metal wireways, 3.1.2.1.4 Criterion 4 Environmental and high melting point insulation, and flame- hilssiles Design Bases resistant markers for identification wherever practicable. 3.1.2.1.4.1 Criterion 4 Statement Cabling is suitably rated and cable tray Structures, systems, and components important loading is designed to avoid objectionable to safety shall be designed to accommodate the internal heat buildup. Cable trays are suitably effects of and to be compatible with the separated to avoid the loss of redundant channels environmental conditions associated with normal O Amendment 1 3.1-3

ABWR umman Standard Plant REV.A operation, maintenance, testing, and postulated (3) 3.4 Water Level (Flood) Design accidents including loss.of coolant accidents. These structures, systems, and components shall be appropriately protected against dynamic (4) 3.5 Missile Protection h effects, including the effects of missiles, pipe (5) 3.6 Protection Against Dynamic whipping, and discharging and conditions outside Effects Associated with the the nuclear power unit. Ilowever, dynamic effects Postulated Rupture of Piping associated with postulated pipe ruptures in nuclear power units may be excluded from the (6) 3.8 Design of Seismic Category l design basis when analyses reviewed and approved Structures by the Commission demonstrate that the probability of fluid system piping rupture is (7) 3.11 En ironmental Oualification of extremely low under conditions consistent with Safety-Related Mechanical and the design basis for the piping. Electrical Equipment 3.1.2.1.4.2 Evaluation Against Criterion 4 (8) 5.2 Integrity of Reactor Coolant Pressure Boundary Essential (see introduction to Section 3.6) structures, systems, and components are designed (9) 6 Enginected Safety Features to accommodate the dynamic effects of and to be compatible with the environmental conditions (10) 7 Instrumentation and Control associated with normal operation, maintenance, Systems and postulated pipe failure accidents including loss of coolant accident (LOCA). (11) 8 Electric Power These structures, systems, and components are 3.1.2.1.5 Criterion 5 Sharing of Structures, appropriately protected against dynamic effects Systems, and Components including the effects of missiles, pipe whipping, and discharging fluids that may result from 3.1.2.1.$.1 Criterlon 5 Statement g equipment failure. T he effects of missiles from sources external to the Nuclear Island are also Structures, systems, and camponents important considered. Design requirements specify (Le time to safety shall not be shared among nuclear po. which each must survise the extreme environn ental wer units unlee it can be shown that such shar-conditions following a LOCA. The design of these ing will not significantly impair the ability to structures, systems, and components meets the perform the safety functions including, in the requirements of Criterion 4. event of an accident in one unit, an orderly shutdown and cooldown of the remaining units. Subsection 3.6.3 identifies the requirements for the piping that is to be excluded from postu- 3.1.2.1.5.2 Evaluation Against Criterion 5 lation of pipe ruptures for design of the plant against dynamic effects from the associated pipe Since the ABWR Nuclear Island is for a ruptures. single unit station, this criterion is not applicable. For further discussion, see the following seetions: 3.1.2.2 Group II Protection by Multiple Fission Product llarriers Chapter / Section Iills 3.1.2.2.1 Criterion 10 Reactor Design (1) 2.0 Summary of Site Characteristics 3.1.2.2.1.1 Criterion 10 Statement (2) 3.3 Wind and Tornado Loadings The reactor core and associated coolant, O Amendment 1 3.1 -4

    'AB M                                                                                          u^6too^u Standard Plant                                                                                   anv. A v  control, and protection systems shall be designed       For further discussion, see the following with appropriate margin to assure that specified sections:

acceptable fuel design limits are not exceeded during any condition of normal operation including Chapter / the effects of anticipated operational Section Iitis oCCurrenCCs. (1) 1.2 General Plant Description 3.1.2.2.1.2 Evaluation Against Criterion 10 (2) 4.2 Fuel Design System The reacter core components consist of fuel assemblies, control rods, incore ion chambers, (3) 43 Nuclear Design neutron sources, and related items. The mechani-cal design is based on conservative application of (4) 4.4 Thermal and Hydraulic Design stress limits, operating experience, and experi-mental test results. The fuel is designed to pro- (5) 5.4.1 Reactor Recirculation System vide integrity over a complete range of power levels including transient conditions. The core (6) 5.4.6 Reactor Core Isolation Cooling is sized with sufficient heat transfer area and System coolant flow to ensure that fuel design limits are not exceeded under normal conditions or (7) 5.4.7 Residual Heat Removal System anticipated operational occurrences. (8) 7.2 Reactor Protection System The reactor protection system is designed to monitor certain reactor parameters, sense (9) 73 Emergency Core Cooling System abnormalities, and to scram the reactor thereby (] preventing fuel design limits from being exceeded (10) 15 when trip points are exceeded. Scram trip Accident Analyses setpoints are selected on operating experience sad 3.1.2.2.2 Criterion 11 Reactor Inherent by the safety design basis. There is no case in Protection which the scram trip setpoints allow the core to exceed the thermal hydraulic safety limits Power 3.1.2.2.2.1 Criterion 11 Statement for the reactor protection system is supplied by four independent uninterruptible AC power The reactor core and associated coolant sys-supplies. An alternate power source and battery tems shall be designed so that in the power ope. I are available for each bus. The reactor will rating range the net effect of the prompt inhe- l scram on loss of power or hydraulic pressure, rent nuclear feedback characteristics tends to j compensate for a rapid increase in reactivity. An analysis and evaluation has been made of the  ; effects upon core fuel following adverse plant 3.1.2.2.2.2 Evaluation Against Criterion 11 1 operating conditions. The results of abnormal  ! operational transients are presented in Chapter 15 The reactor core is designed to have a reac-and show that the minimum critical power ratio tivity response that regulates or damps changes (MCPR) does not fall below the transient MCPR in power level and spatial distributions of limit, thereby satisfying the transient design power production to a level consistent with safe basis, and efficient operation. I The reactor core and associated coolant, The inherent dynamic behavior of the core is control, and protection systems are designed to characterized in terms of: assure that the specified fuel design limits are not exceeded during conditions of normal or (1) fuel temperature or Doppler coefficient; abnormal operation and, therefore, ricet the O requirements of Criterion 10. (2) moderator void coefficient; and Amendment 1 3.15

ABWR m6iman Standard Plant imv. A (3) moderator temperature coefficient. For further discussion, see the following sections: g The combined effect of these coefficients in the power range is termed the power coefficient. Chapter / Sraioll Illk Doppler reactivity feedback occurs simultane-ously with a change in fuel temperature and op- (1) 1.2.1 Principal Design Criteria poses the power change that caused it; it contri-butes to system stability. Since the Doppler re- (2) 43 Nuclear Design activity opposes load changes, it is desirable to maintain a large ratio of moderator void coeffi- (3) 4.4 Thermal and Ilydraulic Design cient to Doppler coefficient for optimum load-following capability. The boiling water reactor 3.1.2.2.3 Criterion 12 Suppression of has an inherently large moderator to Doppler Reactor Power oscillations, coefficient ratio which permits use of coolant flow rate for load following. 3.1.2.23.1 Criterion 12 Statement In a boiling water reactor, the moderator void The reactor core and associated coolant, coefficient is of importance during operation at control, and protection systems shall be powe r. Nuclear design requires the void coeffi- designed to assure that power oscillations which cient inside the fuel channel to be negative, can result in conditions exceeding specified The negative void reactivity coefficient provides acceptable fuel design limits are not possible an inherent negative feedback during power or can be reliably and readily detected and transients. Because of the large negative suppressed, moderator coefficient of reactivity, the BWR has

c. number of inherent advantages, such as: 3.1.2.23.2 Evaluation Against Criterion 12 (1) The use of coolant flow as opposed to The reactor core is designed to ensure that O

coritrol rods for load following; no power oscillation will cause fuel design limits to be execeded. The power reactivity co. (2) the inherent self flattening of the radial efficient is the composite simultaneous effect power distribution; of the fuel temperature or Doppler coefficient, moderator void coefficient, and moderator tem-(3) the ev.e of control; and perature coefficient to the change in power level. It is negative and well within the range (4) the spatial xenon stability, required for adequate damping of power and spa-tial xenon disturbances. Analytical studies in-The reactor is designed so that the moderator dicate that for large boiling water reactors, temperature coefficient is small and positive and under damped, unacceptable power distribution in the cold condition; however, the overall power behavior could only be expected to occur with reactivity coefficient is negative. Typically, power coefficients more positive than about the power coefficient at full power is about 0.01 ok/k/AP/P. Operating experience has 0.04 ok/k/ AP/P at the beginning of life and shown large boiling water reactors to be inhe-about -0.03 ak/k/ AP/P at 10,000 mwd /T. rently stable against xenon induced power These values are well within the range required instability. The large negative coefficients for adequate damping of power and spatial xenon provide: disturbances. (1) good load following with well damped The reactor core and associated coolant system behavior and little undershoot or overshoot are designed so that in the power operating in the heat transfer response; range, prompt inherent dynamic behavior tends to compensate for any rapid increase in reactivity in accord with Criterion 11, (2) load following with recirculation flow control; and g Amendment i 11-6

ABM MA6100AE Standard Plant anv. A (3) strong damping of spatial pcwer variables and systems over their anticipated disturbances. ranges for normal operation, for anticipated O The reactor protection system design provides operational occurrences, and for accident conditions as appropriate to assure adequate protection from excessive fuel cladding tempera- safety; including those variables and systems tures and protects the reactor coolant pressure that can affect the fission process, the boundary from excessive pressures which threaten integrity of the reactor core, the reactor the integrity of the system. Local abnormalities coolant pressure boundary, and the containment are sensed, and if protection system limits are and its associated systems. Appropriate reached, corrective action is initiated through controls shall be provided to maintain these an automatic scram, liigh integrity of the pro- variables and systems within prescribed tection system is achieved through the combina- operating ranges, tion of logic arrangement, trip channel redun-dance, power supply redundancy, and physical 3.1.2.2.4.2 Evaluation Against Criterion 13 separation. The neutron flux in the reactor core is moni-The reactor core and associated coolant, tored by five subsystems. The startup range control, and protection systems are designed to neutron monitor (SRNhi) subsystem measures the suppress any power oscillations which could flux from startup through 15% power (into the result in exceeding fuel design limits. These power range). The power range is monitored by systems assure that Criterion 12 is met. many detectors which make up the local power range monitor (LPRhi) subsystem. The output from For further discussions, see the following these detectors is used in many ways. The sections: output of selected core-wide sets of detectors is averaged to provide a core average neutron Chapter / flux. This output is called the average power Section T_ith range monitor (APRhi) subsystem. The flow rate subsystem (FRS) provides the control and d 1.2.1 Principal Design Criteria reference signal for the APRh1 core flow rate (1) dependent trips. The automated traversing (2) 4.3 Nuclear Design incore probe (ATIP) subsystem provides a means for calibrating the LPRhi subsystem. Both the (3) 4.4 Thermal and flydraulic Design SRNhi and APRhi subsystems generate scram trips to the reactor protection system. They also (4) 7.2 Reacter Protection System - generate rod-block trips. Instrt mentation and Control The reactor protection system protects the (5) 7.3 Emergency Core Cooling System - fuel barriers and the nuclear process barrier by Instrumentation and Control monitoring plant parameters and causing a reactor scram when predetermined setpoints are (6) 7.7 Rod Control and Information exceeded. Separation of the scram and normal System Instrumentation and rod control function prevents failures in the Control reactor manual control circuity from affecting j the scram circuitry. I (7) 15 Accident Analyses j 3.1.2.2.4 Criterion 13 Instrumentation and ces of accidents involving the release of radio. Control active materials from the fuel and reactor cool-ant pressure boundary, the leak detection and 3.1.2.2.4.1 Criterion 13 Statement isolation system initiates automatic isolation of appropriate pipelines whenever monitored Instrumentation shall be provided to monitor variables exceed preselected operational limits. O Amendment 1 3.17

ABWR men Standard Plant e. A Tbc leakage limits for the reactor coolant by Section 50.2 of 10CFR50) are designed, fabri-system (See Subsection 3.1.2.2.6.2) are esta. cated, erected, and tested in accordance with blished so that appropriate action can be taken 10CFR50.55a to provide a high degree of inte-to ensure the integrity of the reactor coolant grity throughout the plant lifetime. Section pressure boundary. The monitored leakage rates 3.2 classifies systems and components within the are classified as identified and unidentified reactor coolant pressure boundary as Quality which corresponds respectively to the flow to the Group A. The design requirements and codes and equipment drain and floor drain sumps. The standards applied to this quality group ensure permissible total leakage rate limit to these high integrity in keeping with the safety-sumps is based upon the taakeup capabilities of related function. various reactor component systems, liigh pump fill up rate and pump out rate are alarmed in the In order to minimize the possibility of brit-main control room. The unidentified leakage rate tie fracture within the reactor coolant pres-as established in Chapter 5 is less than the sure boundary, the fracture toughness properties value that has been conservatively calculated to and the operating temperature of ferritic mate-be a minimum leakage from a crack large enough to rials are controlled Io ensure adequate tough-propagate rapidly but which still allows time for ness. Section 5.2 describes the methods utiliz-identification and corrective action before ed to control toughness properties of the reac-integrity of the process barrier is threatened. tor coolant pressure boundary materials. Mate-rials are to be impact tested in accordance with The process radiation monitoring system moni- ASME Boiler and Pressure Vessel Code Section tors radiation levels of various processes and Ill, where applicable. Where reactor coolant provides trip signals to the reactor protection pressure boundary piping penetrate the contain-system and leak detection and isolation system ment, the fracture toughness temperature whenever pre established limits are exceeded. requirements of the reactor coolant pressure boundary materials apply. Adequate instrumentation has been provided to monitor system variables in the reactor core, re. actor coolant pressure boundary, and reactor con-Piping and equipment pressure parts of the reactor coolant pressure boundary are asserabled h tainment. Appropriate controls have been provi- and erected by welding unicss apphcable codes ded to maintain the variables in the operating permit flanged or screwed joints Welding pro-range and to initiate the necessary corrective cedures are employed which produce welds of action in the event of abnormal operational complete fusion free of unacceptable defects. occurrence or accident. All welding procedures, welders, and welding machine operators used in producing press-Additional information on the instrumentation containing welds are qualified in accordance and controls is given in Chapter 7. with the requirements of the ASME Boiler and Pressure Vessel Code Section IX for the mate. 3.1.2.2.5 Criterion 14 Reactor Coolant rials to be welded. Qualification records Pressure Houndary including the results of procedure and perfor-mance qualification tests and identification 3.1.2.2.5.1 Criterion 14 Statement symbols assigned to each welder are maintained. The reactor coolant pressure boundary shall be Section 5.2 contains the detailed material designed, fabricated, erected, and tested so as and examination requirements for the piping and to have an extremely low probability of abnormal equipment of the reactor coolant pressure boun-leakage, rapidly propagating failure, and gross dary prior to and after its assembly and erec. rupture. tion. Leakage testing and surveillance is ac-complished as described in the evaluation 3.1.2.2.5.2 Evaluation AgainM C.;;erion 14 against Criterion 30 of the General Design Criteria. The piping and equipment pressure parts within the reactor coolant pre.ssure boundary (as defined The design, fabrication, erection, and g Amendment 1 31-8

ABM zwtoo^e Standard Plant REV.A testing of the reactor coolant pressure boundary The auxiliary, control, and protection assure an extremely low probability of failure of systems associated with the reactor coolant p) t abnormal leakage, thus satisfying the require- system act to provide sufficient margin to ments of Criterion 14. assure that the design conditions of the reactor i coolant pressure boundary (RCPB) are not For further discussion, see the following exceeded during any condition of normal sections: operation including anticipated operational occurrences. As described in the evaluation of Chapter / Criteric n 13, instrumentation is provided to bection Iitig monito: essential variables to ensure that they are within prescribed operating limits. If the (1) 1.2.1 Principal Design Criteria monitored variables exceed their predetermined settings, the auxiliary, control, and protection (2) 3 Design of Structures, systems automatically respond to maintain the Components, Equipment, and variables and systems within allowable design Systems limits. (3) 5.2 Integrity of Reactor Coolant An example of the integrated protective ac-Pressure Boundary tion scheme which provides sufficient margin to assure that the design conditions of the reactor (4) 53 Reactor Vessel coolant pressure boundary are not exceeded is the automatic initiation of the pressure relief (5) 5.4.1 Reactor Recirculation System system of the nuclear boiler system upon receipt of an overpressure signal. To accomplish over-(6) 15 Accident Analyses pressure protection of the reactor pressure ves-sel system and reactor coolant pressure bounda. (7) 17 Quality Assurance ry, a number of pressure operated relief valves are provided that can discharge steam from the 3.1.2.2.6 Criterion 15 Reactor Coolant System main steamlines Io the suppression pool. The Design pressure relief system also provides for auto-matic depressurization of the RCS in the event 3.1.2.2.6.1 Criterion 15 Statement of an LOCA in which the vessel is not depressu- , rized by the accident. The depressurization of The reactor coolant system and associated the RCS in this situation allows operation of auxiliary, control, and protection systems shall the low-pressure emergency core cooling systems be designed with sufficient margin to assure that to supply r.nough cooling water to adequately the design conditions of the reactor coolant cool the core. In a similar manner, other pressure boundary are not exceeded during any auxiliary, control, and protection systems condition of normal operation including provide assurance that the design conditions of anticipated operational occurrences, the RCPB are not exceeded during any conditions of normal operation including anticipated 3.1.2.2.6.2 Evaluation Against Criterion 15 operational occurrences. The reactor coolant system (RCS), as The application of appropriate codes and identified in Section 5.1, consists mainly of the standards and high quality requirements to the nuclear steam supply systems (NSSS) comprised of reactor coolant pressure boundary and the design the reactor vessel and appurtenances, the reactor features of its associated auxiliary, control, recirculation system and the nuclear boiler and protection systems assure that the system including the main steamlines, feedwater requirements of Criterion 15 are satisfied, lines and pressure relief discharge system; the reactor core isolation cooling (RCIC) systems; For further discussion, see the following the residual heat removal (RHR) system; and the sections: l reactor water cleanup system (RWCS). Amendment 1 3.19

ABM za^6ioo^e Standard Plant REV.A Chapter / (3) Associated containment penetrations and Section Illh isolation devices. (1) 1.2.1 Principal Design Criteria The drywell and wetwell zones condense the 9 steam and contain fission product releases from (2) 3 Design of Structure, Components, the postulated design bases accident (i.e., the Equipment, and Systems double ended rupture of the largest pipe in the primary coolant system). The leaktight primary (3) 5.2.2 Overpressurization Protection containment vessel prevents the release of fission products to the environment. (4) 5.2.5 Reactor Coolant Pressure Boundary and Core Cooling The secondary containment boundary of the Systems Leakage Detection reactor building, which completely encloses and integrates structurally the PCV, provides addi-(5) 5.3 Reactor Vessel tional radiation shielding to protect operating personnel and the public and also protects the (6) 5.4.1 Reactor Recirculation System PCV from weather and external missiles. (7) 15 Accident Analyses Temperature and pressure in the PCV are limited following an accident by using the RilR 3.1.2.2.7 Criterion 16 Containment Design sys'.cm to condense steam in the containment at-mosphere and to cool the suppression pool water. 3.1.2.2.7.1 Criterion 16 Statement The design of the containment systems meets Reactor containment and associated systems the requirements of Criterion 16. shall be provided to establish an essentially leaktight barrier against the uncontrolled For further discussion, see the following release of radioactivity to the environment and sections. g to assure that the containment design conditions W important to safety are not exceeded for as long Chapter / as postulated accident conditions require. Section ljih 3.1.2.2.7.2 Evaluation Against Criterion 16 (1) 1.2 General Plant Description l l The primary containment system consists of the (2) 3.8.2 Steel Containment l following major structures and components: 1 (3) 6.2 Containment Systems  ; (1) A leaktight primary containment vessel (PCV) enclosing the reactor pressure vessel, the (4) 15 Accident Analyses reactor coolant pressure boundary, and other branch :onnections of Ihe reactor primary 3.1.2.2.8 Criterion 17 Electric Power coolant system. The PCV is a cylindrical Systems steel lined reinforced concrete structure with a removable steel head and has upper and 3.1.2.2.8.1 Criterion 17 Statement lower drywell zones, diaphragm floor (D/F) and annular suppression chamber (or wetwell An onsite electric power system and an off. zone) under upper drywell separated by the site electric power system shall be provided to D/F. permit functioning of structures, systems, and components impoitant to safety. The safety (2) A suppression pool containing a large amount function for each system (assuming the other of water used to rapidly condense steam from system is not functioning) shall be to provide a reactor vessel blowdown or from a break in sufficient capacity and capability to assure a major pipe, that: O Amendment 1 3.1 10

ABM ni6toaan Standard Plant REY.A (1) specified acceptable fuel design limits and There are three independent AC load groups fq design conditions of the reactor coolant provided to assure independence and redundancy V pressure boundary are not exceeded as a of equipment function. These meet the safety resu1t of antieipated operationai requirements, assuming a single failure, since: occurrences, and (1) each load group is independently capable of (2) the core is cooled and containment integrity isolation from the offsite power sources, and other vital functions are maintained in and the event of postulated accidents. (2) each load group has separate circuits to The onsite electric power supplies including independent power sources. the batteries, and the onsite electric dis-tribution system shall have sufficient For each of the three AC load groups there independence, redundancy, and testability to are independent batteries which furnish DC load perform their safety functions assuming a single and control power for the corresponding divi-failure. sions. An additional battery furnishes DC load and control power for the safety system logic Electric power from the transmission network and control (SSLC) Division IV bus, to the onsite electric distribution system shall be supplied by two physically independent cir- The reactor protection instrumentation is cuits (not necessarily on separate rights of way) powered from four independent AC/DC power designed and located to minimize to the extent sources, practical the likelihood of simultaneous failure under operating and postulated accident and envi- The onsite electric power systems are ronmental conditions. A switchyard common to designed to meet the requirements of Criterion both circuits is acceptable. Each of these cir- 17. For further discussion, see the following cuits shall be designed to be available in suffi- sections: cient time following a loss of all onsite alter-nating current power supplies and the other off- Chapter / site electric power circuit to assure that speci- Section Illk fied acceptable fuel design limits and design conditions of the reactor coolant pressure boun- (1) 1.2 General Plant Description dary are not exceeded. One of these circuits shall be designed to be available within a few (2) 3.10 Seismic Qualification of Seismic seconds following a loss-of coolant accident to Category l lnstrumentation and I assure that the core cooling, containment inte- Electrical Equipment grity, and other vital safety functions are maintained. (3) 3.11 Environmental Qualification of Safety-Related Mechanical and Provisions shall be included to minimize the Electrical Equipment probability of losing electric power from any of the remaining supplies as a result of or (4) 8.3 Onsite Power Systems coincident with, the loss of power generated by the nuclear power unit, the loss of power from 3.1.2.2.8.2.2 Offsite Electric Power System the transmission network, or the loss of power , from the onsite electric power supplies. The design of the offsite power systems is j out of the Nuclear Island scope. The Nuclear l 3.1.2.2.8.2 Evaluation Against Criterion 17 Island / remainder of plant power system interfaces are addressed in Subsection 8.2.3. 3.1.2.2.8.2.1 Onsite Electric Power System O . Amendment 1 3.1 11

ABWR 2346iooan Standard Plant nrn/. 6 3.1.2.2.9 Criterion 18 Inspection and Chapter Testing of Electric Power Systems Section Titls g 3.1.2.2.9.1 Criterion 18 Statement (1) 83 Onsite Power Systems Electric power systems important to safety (2) 14 InitialTest Program shall be designed to permit appropriate periodic inspection and testing of important areas and 3.1.2.2.9.2.2 Offsite Electric Power System features, such as wiring, insulation, connections, and switchboards, to assess the The evaluation against the testability of the continuity of the systems and the condition of full operational sequence is out of the Nuclear their components. The systems shall be designed Island scope. These operations include the with a capability to test periodically: automatic transfer of power from preferred offsite to alternate offsite power sources and (1) the operability and functional performance of manual selection of any power source. The the components of the systems such as onsite Nuclear Island / remainder of plant power system power sources, relays, switches, and buses; interfaces are addressed in Subsection 8.2.3. and 3.1.2.2.10 Criterion 19 Control Room (2) the operability of the systems as a whole and, under conditions as close to design as 3.1.2.2.10.1 Criterion 19 Statement practical, the full operational sequence that brings the systems into operation including A control room shall be prosided from which operation of applicable portions of the actions can be taken to operate the nuclear po-protection system and the transfer of power wer unit safely under normal conditions and to among the nuclear power unit, the offsite maintain it in a safe condition under accident power system, and the onsite power system, conditions, including loss of-coolant acci-3.1.2.2.9.2 Evaluation Against Criterion 18 dents. Adequate radiation protection shall be provided to permit access and occupancy of the g control room under accident conditions without 3.1.2.2.9.2.1 Onsite Electric Power System personnel receiving radiation exposures in ex-cess of 5 rem whole body, or its equivalent to The important power supply buses and associ- any part of the body, for the duration of the ated normal preferred, alternate, and standby AC accident. power supplies are arranged for periodic inspec-tion and testing of each load group independent. Equipment at appropriate locations outside ly. The testing procedure includes a bus trans- the control room shall be provided: fer from normal preferred power supply to alter-nate preferred power supply, simulates a loss of (1) with a design capability for prompt hot preferred power (LOPP) signal or a LOCA signal to shutdown of the reactor including necessary start the diesel generator bringing it to operat- instrumentation and controls to maintain the ing condition. Full load testing of the diesel unit in a safe condition during hot generator can be performed by manually synchro- shutdown, and nizing the generator to the normal preferred power supply. These tests are performed (2) with a potential capability for subsequent periodically together with other equipment not cold shutdown of the reactor through the use included in the Nuclear Island which proves the of suitable procedures. engineered safety system operability. 3.1.2.2.10.2 Evaluation Against Criterion 19 Design of the standby power systems provides testability in accordance with the requirements The control room contains the following of Criterion 18. equipment: controls and necessary surveillance O Amendment 1 3.1 12

ABM 2346100,a Standard Plant REY.A equipment for operation of the plant functions (6) 9.4.1 Control Room Area Ventilation such as the reactor and its acxiliary systems, System \p

._j engineered safety features, turbine generator, steam and power conversion systems, and station    (7) 9.5.1         Fire l'rotection System electrical distribution boards.

(8) 123.2 Shielding The control room is located in a Seismic Category I control building. Safe occupancy of (9) 1233 Ventilation the control room during abnormal conditions is provided for in the design. Adequate shielding 3.1.23 Group III Protection and Reactivity is provided to maintain tolerable radiation C,,ntrol System levels in the control room in the event of a design basis accident for the duration of the 3.1.23.1 Criterion 20 Protection System accident. Functions The control building ventilation system has 3.1.23.1.1 Criterion 20 Statement redundant equipment and prosides radiation detec-tors and smoke detectors with appropriate alarms The protection system shall be designed and interlocks. The control room intake air can be filtered through high-efficiency particulate (1) to initiate automatically the operation of air / absolute (HEPA) and charcoal filters. appropriate systems including the reactivity control systems, to assure that specified The control room is continuously occupied by acceptable fuel design limits are not ex-qualified operating personnel under all operating ceeded as a result of anticipated opera-and accident conditions. In the unlikely event tional occurrences, and that the control room must be vacated and access is restricted, instrumentation and controls are (2) to sense accident conditions and initiate (] N provided outside the control room which can be utilized to safely perform a hot shutdown and a the operation of systems and components important to safety, subsequent cold shutdown of the reactor. 3.1.23.1.2 Evaluation Against Criterion 20 The control room design meets the requirements of Criterion 19. The reactor protection system is designed to provide timely protection against the onset and For further discussion, see the following consequences of conditions that threaten the  ; sections: integrity of the fuel barrier and the reactor l coolant pressure boundary barrier. Fuel damsge  ! Chapter / is prevented by initiation of an autoraatic Section Illk reactor shutdown if monitored variables of nuclear steam supply systems (see Subsection (1) 1.2 General Plant Description 3.1.2.2.6.2) exceed pre established limits of anticipated operational occurrences. Scram trip (2) 3.8.4 Other Seismic Category l settings are selected and verified to be far , Structures enough above or below operating levels to pro-vide proper protection but not be subject to

                                                                                                                    )

(3) 7 Instrumentation and Centrcl spurious scrams. The reactor protection system Systems includes the ride Grough power sources, sen-sors, transmitters, bypass circuity, and (4) 7.4.1.4 Remote Shutdown System - switches that signal the control rod system to , and lostrumentation and Controls scram and shut down the reactor. The scrams i 7.4.2.4 initiated by neutron monitoring system vari-ables, nuclear steam npply systems (NSSS) high (5) 6.4 Habitability Systems pressure, turbine stop valve closure, turbine l Amendment 1 3.1-13 I

ABM 23A6100AE Standard Plant REV.A control valve fast closure, and reactor vessel (5) 63 Emergency Core Cooling Systerr low water level prevent fuel damage following 3 abnormal operational transients. Specifically, (6) 7.2 Reactor Protection S> stem W these process parameters initiate a ser m it time to prevent the core from exceeding thermal by- (7) 73.1.1 Emergency Core Cooling Systems - draulic safety limits during abnormal operational and Instrumentation and Control transients. Response by the reactor protection 73.2.1 system is prompt and the total scram time is short. Control rod scram mot: rts in about (8) 73.1.2 Leak Detection and Isolation 290 milliseconds after the hi . set point is and System Instrumentation exceeded. 73.2.2 and Controls A fully withdrawn control rod traverses 60% of (9) 7.6.1.2 Process Radiation Monitoring its full stroke in sufficient time to assure that and System - Instrumentation and acceptable fuel design limits are not exceeded. 7.6.2.2 Controls In addition to the reactor protection system, (10) 15 Accident Analyses which provides for automatic shutdown of the re-actor to prevent fuel damage, protection systems 3.1.23.2 Criterion 21 Protection System are provided to sense accident conditions and Reliability and Testability initiate automatically the operation of other

• ystems and components important to safety.

,                                                     3.1.23.2.1 Cdterion 21 Statement Systems such as the emergency core cooling system (ECCS) are initiated automatically to limit the            The protection system shall be designed for extent of fuel damage following a loss of-coolant      functional reliability and inservice testability acrident (LOCA). Other systems automatically commensurate with the safety functions to be isolate the reactor vessel or the containment to       performed. Redundancy and independence designed prevent the release of significant amounts of into the protection system shall be sufficient radioactive materials from the fuel and the re-        to assure that:

g actor coolant pressure boundary. The controls and instrumentation for the ECCS and the isola- (1) no single failure results in loss of the tion systems are initiated automatically when protection function, and monitored variables exceed pre selected operational limits. (2) removal from service of any component or channel does not result in loss of the re. The design of the protection system satisfies quired minimum redundancy unless the accept-the functional requirements as specified in able reliability of operation of the protec-Criterion 20. tion system can be otherwise demonstrated. For further discussion, see the following The protection system shall be designed to sections: permit periodic testing of its functioning when the reactor is in operation including a capabi-Chapter lity to test channels independently to determine Section Titls _ failures and losses of redundancy that may have occurred. (1) 1.2.1 Principal Design Criteria i 3.1.23.2.2 Evaluation Against Critenion 21 1 (2) 4.6 Functional Design of Reactisity Control Systems Reactor protection system design provides , assurance that, through redundancy, each channel I (3) 5.2.2 Overpressurization Protecti, has sufficient reliability to fulfill the  ! single-failure criterion. No single component (4) 5.4.5 Main Steamline Isolation System failure, intentional bypass maintenance Amendment 1 3.1 14 l

23A6100AE Standard Plant REV.A operation, calibration operation, or test to observed on control room instrumentation. More verify operational availability impairs the importantly, the hydraulic control unit scram O diiitx e, i8e sxsie- ie verrer- its i i aca safety function. Additionally, the system design cc= - i ter i vei is ce ti e six - e iterea. assures that when a scram trip point is exceeded, The main steamline isolation valves may be there is a high scram probability. However, tested during full reactor operation. Indivi-should a scram not occur, other monitored compo- dually, they can be closed to 90% of full open nents scram the reactor if their trip points are position without affecting the reactor opera-exceeded. There is sufficient electrical and tion. If reactor power is reduced sufficiently, physical separation between channels and between the isolation valves may be fully closed, logics monitoring the same variable to prevent During refueling operation, valve leakage rates environmental factors, elec*rical transients, and can be determined. physical events from impairing the ability of the system to respond correctly. The RHR system testing can be performed during normal operation. Main system pumps can The reacto protection system includes design be evaluated by taking suction from the suppres-features that permit inservice testing. This sion pool and discharging through test lines ensures the functional reliability of the system back to the suppression pool. System design and should the reacio: variable exceed the corrective operating procedures also permit testing the action setpoint. supply valves of the three RHR lines. The lower pressure flooder mod: can be tested after reac-The reactor protection system initiates an f or shutdown. Eai active component of the ECCS automatic reactor shutdown if the monitored plant provided to operate in a design bas!s accident variables exceed preestablished limits. This is designed to be operable for test purposes system is arranged as four separately powered during normal operation of the nuclear system. divisions. The high fune.ional reliability, redundancy, I O v Each division has a logic which can produce an automatic trip signal. The logic scheme is a and inservice testability of the protection syste.n satisfy the requirements specified in two out of four arrangement. The reactor protec- Criterion 21. tion system can be tested during reactor opera- ) tion. Manual scram testing is performed by For further discussion, see the following operating one of the four manual scram controls; sections: this tests one division. The total test verifies the ability to de energize the scram pilot valve Chapter / solenoids. Indicating lights verify that the SGim Iith actuator contacts have opened. This capability for a thorough testing program significantly (1) 1.2.1 Prireipal Design Criteria increases reliability. I (2) 4.6 Functional Design of Reactivity i Control rod drive operability can be tested Control Systems l during normal reactor operation. Rod position indicators and in core neutron detectors are used (3) 5.4.5 Main Steamline Isolation Sygem to verify control rod movement. Each control rod can be withdrawn one step and then reinserted to (4) 5.4.7 Residual Heat RemovalSystem the original position without significantly perturbing the nuclear steam supply systems at (5) 6.2 Containment Systems most power levels. One control rod is tested at a time. Control rod mechanism overdrive (6) 6.3 Egnergency Core Cooling Systems demonstrates rod to-drive coupling integrity. Hydraulic supply subsystem pressures can t e (7) 'L2 Reactor Protection System O Amendment 1 3.1 15

MM 23A6100AE Standard riant any A (8) 73.1.1 Emergency Core Cooling Systems - the protection system allows operational system and Instrumentation and Control testing by the use of an independent inpu for 73.2.1 each actuator logic. When an ersential moni-tored variable exceeds its scram trip point, it (9) 73.1.2 Leak Detection and Isolation is sensed by four independent sensors. An and Control system -Instrumentation intentional bypass, maintenance operation, 73.2.2 and Controls calibration operation, or test results in a single channel trip. This leaves three channels (10) 7.6.1.2 Process Radiation hionitoring - per monitored variable, each of which is capable and Instrumentation and Controls of initiating a scram. Only two actuator logics 7.6.2.2 must trip to initiate a scram. Thus, the two-out of four arrangement assures that a scram oc-(11) 15 Accident Analyses curs as a monitored variable exceeds its scram l setting. l 3.1.233 Criterion 22 Protection System Independence The protection system meets the design requirements for functional and physical 3.1.233.1 Criterion 22 Statement independence as specified in Criterion 22. The protection system shall be designed to For further discussion, see the following assure that the effects of natural phenomena and sections: of normal aperating, maintenance, testing, and i postulated accident conditions on redundant Chapter / channels do not result in loss of the protection Section Iith I function, or shall be demonstrated to be l acceptable on some other defined basis. Design (1) 1.2.1 Principal Design Criteria techniques, such as functional diversity or diversity in component design and principles of (2) operation, shall be used to the extent practical 4.6 Functional Design of Reactivity Control Systems g to prevent loss of the protection function. (3) 5.4.5 hiain Steamline Isolation System j 3.1.233.2 Evaluation Agalnat Criterion 22 l (4) 5.4.7 Residual Heat Removal System Components of protection systems are designed j so that the mechanical, thermal and radiological (5) 63 Emergency Core Cooling Systems 1 environment resulting from any accident situation , in which the components are required to function (6) 7.2 Reactor Protection System  ! do not interfere with the operation of that funetion. (7) 73.1.1 Emergency Core Cooling System - and Instrumentation and Controls The redundant sensors are electrically and 73.2.1 physically separated. Only circuits of the same division are run in the same raceway. hi ulti- (8) 73.1.2 1cak Detection and Isolation plexed signals are carried by fiber optic medium and System to assure control sigeal isolation. 73.2.2 The reactor protection system is designed to (9) 7.6.1.2 Process Radiation hionitoring - permit maintenance and diagnostic work while the and Instrumentation and Controls reactor is operating without restricting the 7.6.2.2 plant operation or hindering the output of safety functions. The flexibility in design afforded (10) 15 Accident Analyses O Am" iment 1 1.lc

MM 23A6100AE Standard Plant REV.A l l 3.1.23.4 Criterion 23 Protection System (2) 7.2 Reactor Protection System q U Failure Mades 3.1.23.5 Criterion 24 Separation of 1 3.1.2.3.4.1 Criterion 23 Statement Protection and Control Systems The protection system shall be designed to 3.1.23.5.1 Criterion 24 Statement

• I fail into a safe state or into a state demon-strated to be acceptable on some other defined The protection system shall be separated from basis if conditions such as disconnection of the control systems to the extent that failure of system, loss of energy (e.g., electric power, in- any single control system component or channel strument air), or postulated adverse environments or failure or removal from service of any single (e.g., extreme heat or cold, fire, pressure, protection system component or channel which is steam, water, and radiation) are experienced, common to the control ar.d protection systems leaves intact a system satisfying all reliabili-3.1.23.4.2 Evaluation Against Criterion 23 ty, redundancy, and independence requirements of the protection system. Interconnection of the The reactor protection (trip) system is protection and control systems shall be limited designed to fail into a safe state. Use of an to assure that safety is not significantly independent channel for each actuator logic impaired, allows the system to sustain any logic channel failure without preventing other sensors 3.1.23.5.2 Evaluation Against Criterion 24 monitoring the same variable from initiating a scram. Any two out of four logic channel trips There is separation between the reactor pro-initiate a scram. Intentional bypass for tection system and the process control systems.

maintenance or testing causes the scram logic to Logic channels and actuator logics of the reac-revert to two out of three. A failure of any one tor protection system are not used directly for reactor protection system input or subsystem automatic control of process systems. Sensor O comp nent pr duces a trip in one channel. This outputs may be shared, but each signal is opti-condition is insufficient to produce a reactor cally isolated before intering a redundant or scram, but the system is ready to perform its non safety channel interface. Therefore, fail-protective function upon trip of another channel, ure in the controls and instrumentation of pro-cess systems cannot induce failure of any por-The environmental conditions in which the tion of the protection system. Scram reliabi-instrumentation and equipment of the reactor lity is designed into the reactor protection protection system must operate were considered in system and hydraulic control unit for the con-establishing the component specifications. trol rod drive. The scram signal and mode of Instrumentation specifications are based on the operation override all other signals, worst expected ambient conditions in which the instruments must operate. The systemt that isolation containment and reactor pressure vessel are designed so that any The tailure modes of the reactor protection one failu;e, maintenance operation, calibration (trip) system are such that it fails into a cafe operation, or test to verify operational state as required by Criterion 23. availability does not impair the functional ability of the ie station systems to respond to For further discussion, see the fol;owing ssential variab.es. sections. Process radiation monitoring is provided on i Chapter / process liquid and gas lines that may serve as l Section Iilk discharge routm for radioactive materials. Four instrume .aion channels are used to pre-(1) 1.2.1 Principal Design Criteria vent an inadvmelt scram and isolation as a re-O Amendment 1 3.1 17

ABWR m iman Standard Plant RE'!. A sult of instrumentation malfunctions. The out- 3.1.23.6.2 Evaluation Against Criterion 25 put trip signals from each channel are combined in such a way that two channels must signal high The reactor protection system provides radiation to initiate scram and main steam protection against the onset and consequences of isolation. conditions that threaten the integrity of the fuel barrier and the reactor coolant pressure The protection system is separated from boundary. Any n onitored variable which exceeds control systems as required in Criterion 24. the scram setpoint will initiate an automntic scram and not impair the remaining variables For further discussion, see the following from being monitored, and if one channel fails, se r'.io n s: the remaining portions of the reactor protection system shall function. Chapter / Section Iitk The rod control and information system is d:- signed so that no single failure can negate the (1) 1.2.1 Principal Design Criteria effectiveness of a reactor scram. The circuitry for the rod control and information system is (2) 4.6 Functional Evaluation of completely independent of the circuitry Reactivity Control Systems controlling the scram valves. This separation of the scram and normal rod control functions (3) 63 Emergency Core Cooling Systems prevents failures in the reactor manual c ntrol circuitry from affecting the scram circuitry. (4) 7.2 Reactor Trip System Because only two control rods are controlled by an individual hydraulic control unit (IICU), a (5) 73.1.1 Emergency Core Cooling System - failure that results in continued energiring of and Instrumentation and Controts an insert solenoid valve on an llCU can affect 13.2.1 only two control rods. The effectiveness of a (6) 73.1.2 Leak Detection and Isolation reactor scram is not impaired by the malfunc. tioning of any one IICU or two control rods. g and System Instrumentation 73.2.2 and Controls The design of the protection system assures that specified acceptable fuel limits are not (7) 7.6.1.2 Process Radiation htonitoring - exceeded for any single malfunction of the and Instrumentation and Controls reactivity control systems as specified in 7.6.2.2 Criterion 25. (8) 7.7.1.2 Rod Control and information For further discussion, see the following and System Instrumentation and sections: 7.7.2.2 Controls Chapter / 3.1.2J.6 Criterion 25 Protection Sptem Section Iitk Requirements for Reacthity Control Slalfunctions (1) 1.2.1 Principal Design Criteria 3.1.2.3.6.1 Criterion 25 Statement (2) 43 Nuclear Design The protection system shall be designed to as-sure that specified acceptable fuel design limits (3) 4.4 Thermal and Ilydraulic Design are not exceeded for any single malfunction of the reactivity control system such as accidental (4) 4.6 Functional Design of Reactivity withdrawal (not ejection or dropout) of control utrol Systems rods. O Amendment 1 3118 i

I ABM 23A6100AE l Standard Plant REV.A  ! (5) 7.2 Reactor Trip Sptem The circuitry for manual insertion or with-drawal of contro: rods is completely independent O 1 (6) 7.7.1.2 Rod Control and Information of the circuitry for reactor scram. This l and System - Instrumentation and separation of the gram and normal rod control 7.7.2.2 Controls functions prevents failures in the reactor manual control circuitry from affecting the (7) 15 AccMent Analyses scram circuitry. Two sources of energy (accumulator pressure and electrical power to 3.1.23.7 Criterion 26. Reacthity Control the motors of the fine motion control rod System Redundancy and Capability drives, FMCRDs) provide needed control rod insertion performance over the entire range of 3.1.23.7.1 Criterion 26 Statement reactor pressure (i.e., from operating conditions to cold shutdown). The design of the Two independent reactivity control systems control rod system includes appropriate margin ofdifferent design principles shall be provided, for malfunctions such as stuck rods in the One of the systems s'aall use control rods, pre- unlikely event that they do occur. Control rod ferably including a positive means for inserting withdrawal sequences and patterns are selected the rods, and shall be capable of reliably con- prior to operation to achieve optimum core trolling reactivity changes to assure that under performance and, simultaneously, low individual conditions of normal operation, including rod worths. The operating procedures to anticipated operational occurrences, and with ' accomplish such patterns are supplemented by the appropriate margin for malfunctions such as stuck rod pattern control system, which prevents rod rods, specified acceptable fuel design limits are withdrawals yielding a rod worth greater than not exceeded. The second reactivity control permitted by the preselected rod withdrawal system shall be capable of reliably controlling pattern. Because of the carefully planned and the rate of reactivity changes resulting from regulated rod withdrawal sequence, prompt planaed, normal power changes (including xenon shutdown of the reactor can be achieved with the burraut) to assure that acceptable fuel design insertion of a small number of the many limits are not exceeded. One of the systems independent control rods. In the event that a shall be capable of holding the reactor core reactor scram is necessary, the unlikely suberitical under cold conditions. occurrence of a limited number of stuck rods will not hinder the capability of the control 3.1.2.3.7.2 Eysluation Against Criterion 26 rod system to render the core suberitical. Two independent reactivity control systems The second independent reactivity control utilizing difference design principles are pro- system is provided by the reactor recirculation vided. The normal method of reactivity control system. By varying reactor coolant flow, it is employs control rod assemblies which contain possible to affect the type of reactivity boron carbide (B4C) powder. Positive insertion changes necessary for planned, normal power of these control rods is provided by means of the changes (including xenon burnout). In the control rod drive electrical and hydraulic sys- unlikely event that reactor flow is suddenly tem. The control rods are capable of reliably increased to its maximum value (pump runout), controlling reactivity changes during normal the core will not exceed fuel design liraits operation (e.g., power changes, power shaping, because the power flow map defines the allowable xenon burnout, normal startup and shutdown) via initial operating states so that the pump runout operator controlled insertions and withdrawals. will not violate these limits. l The control rods are also capable of maintaining i the core within acceptable fuel design limits du- The control rod system is capable of holding  ! ring anticipated operational occurrences via the the reactor core suberitical under cold I automatic scrarn function. The unlikely occur- conditions, even when the pair of control rods rence of a limited number of stuck rods during a of highest worth controlled by an hydraulic scram will not adversely affect the capability to control unit is assumed to be stuck in the fully maintain the core within fuel design limits. withdrawn position. This shutdown capability of O. 1 l Amendment 1 3.1 19 l

ABM 2346iooxu Standard Plant REV.A the control rod system is made possible by There is no credible event applicable to the designing the fuel with burnable poison (Gd2 ABWR which requires combined capability of the

03) to control the high reactivity of fresh control rod system and poison additions. The h

fuel. In addition, the standby liquid control ABWR design is capable of inaintaining the system is available to add soluble boron to the reactor core suberitical, including allowance core and render it suberitical as discussed under for a pair of stuck rods controlled by an the evaluation against Criterion 27. hydraulic control unit (HCU), without addition of any poison to the reactor coolant. The The redundancy and capabilities of the primary reactivity control system for the ABWR reactivity control systems for the BWR satisfy during postulated accident conditions is the the requirements of Criterion 26. control rod system. Abnormalities are sensed, and, if protection system limits are reached, For further discussion, see the following corrective action is initiated through automatic sections: insertion of control rods. High integrity of the protection system is achieved through the Chapter / combination of logic arrangement, actuator section Iills redundancy, power supply redundancy, and physical separation. High reliability of (1) 1.2.1 Principal Design Criteria reactor scram is further achieved by separation of scram and manual control circuitry, (2) 16 Functional Design of Reactivity individual HCU controlling a pair of control Control Systems rods, and fail safe design features built into the rod drive system. Response by the reactor (3) 7.3 Enginected Safety Feature protection system is prompt and the total scram Systems time is short. (4) 7.4.1.2 and Standby Liquid Control System - Instrumentation and Controls In the very unlikely event that more than one control rod fails to insert and the core cannot g 7.4.2.2 be maintained in a subcritical condition by con. trol rods alone as the reactor is cooled down (5) 7.7.1.2 Rod Control and Information subsequent to initial shutdown, the standby li- j and System - Instrumentation and quid control system (SLCS) can be actuated to i 7.7.2.2 Controls insert solul'c boron into the reactor core. The l SLCS has sufficient capacity to ensure that the 3.1.2.3.8 Criterion 27 Combined Reactivity reactor can always be maintained suberitical; Control Systems Capability and, hence, only decay heat will be generated by the core which can be removed by the RHR System, 3.1.2.3.8.1 Criterion 27 Statement thereby ensuring that the core will always bc , coolable. The reactivity control systems shall be de-signed to have a combined capability in conjunc- The design of the reactivity control sptems tion with poison addition by the emergency core assures reliable control of reactivity under  ; cooling systems of reliably controlling reacti- postulated accident conditions with appropriate vity changes to assure that, under postulated margin for stuck rods. The capability to cool I accident conditions and with appropriate margin the core is mairtained under all postulated for stuck rods, the capability to cool the core accident conditions; thus, Criterion 27 is is maintained. satisfied. 3.1.2.3.8.2 Evaluation Against Criterion 27 For further discussion, see the following sections: O Amendment 1 3.1 20 e

ABM11 22^6100^s Standard Plant REV.A Chapter / the preselected rod withdrawal pattern. The Section Tilk RPCS function assists the operator with an (qj effective backup control rod monitoring routine (1) 1.2.1 Principal Design Criteria that enforces adherence to established startup, shutdown, and low power level operations control (2) 43 Nuclear Design rod procedures. (3) 4.4 Thermal and Hydraulic Design The control rod mechanical design incor-porates a brake system which prevents rapid rod (4) 4.6 Functional Design of Reactivity ejection. This engineered safeguard protects Control system against a high reactivity insertion rate from a potential control rod ejection. Normal rod (5) 7.2 Reactor Trip System movement and the rod withdrawal rate is limited through the fine motion control motor. (6) 7.4.1.2 Standby Liquid Control System - and Instrumentation and Controls The accident analysis (Chapter 15) evaluates 7.4.2.2 the postulated reactivity accidents, as well as abnormal operational transients in detail. (7) 15 Accident Analyses Analyses are included for rod dropout, steamline rupture, changes in reactor coolant temperature 3.1.23.9 Criterion 28 Reactivity Limits and pressure, and cold water addition. The initial conditions, assumptions, calculational 3.1.2.3.9.1 Criterion 28 Statement models, sequences of events, and anticipated results of each postulated occurrence are The reactivity control systems shall be covered in detail. The results of these designed with appropriate limits on the potential analyses indicate that none of the postulated amount and rate of reactivity increase to assure reactivity transients or accidents results in / that the effects of postulated reactivity damage to the reactor coolant pressure accidents can neither: boundary. In addition, the integrity of the core, its support structures or other reactor (1) result in damage to the reactor coolant pres- pressure vessel internals is maintained so that sure boundary greater than limited local the capability to cool the core is not impaired yielding, nor for any of the postulated reactivity accidents described in the accident analysis. (2) sufficiently disturb the core, its support structures or other reactor pressure vessel The design features of the reactivity control internal to impair significantly the capabi- system which limit the potential amount and rate lity to cool the core. These postulated of reactivity increase ensure that Criterion 28 reactivity accidents shall include considera- is satisfied for all postulated reactivity tion of rod ejection (unless prevented by accidents. positive means), rod dropout, steamline rup-ture, changes in reactor coolant temperature For further discussion, see the following and pressure, and cold water addition. sections: 3.1.23.9.2 Evaluation Against Criterion 28 Chapter / Section Iilk The control rod system design incorporates appropriate limits on the potential amount and (1) 1.2.1 Principal Design Criteria rate of reactivity increase. Control rod with-drawal sequences and patterns are selected to (2) 3.9.4 Control Rod Drive System achieve optimum core performance and low individual rod worths. The rod pattern control (3) 3.9.5 Reactor Pressure Vessel system (RPCS) prevents withdrawal other than by Internals C Amendment 1 3.1 21

ABWR mame Standard Plant nrw.A (4) 43 Nuclear Design rod drives, main steam isolation valves, residual heat removal pumps, etc., are testable g (5) 4.53 Control Rod Drive flousing during normal reactor operation. Functional w Supports testing and calibration schedules are developed using available failure rate data, reliability (6) 4.6 Functional Design of Reactivity analysis, and operating experience. These Control Systems schedules represent an optimization of protection and reactivity control system (7) 5.2.2 Outpressurization Protection reliability by considering the f ailure probabilities of individual components and the (8) 53 Reactor Vessel reliability effects during individual compo. nent testing on the portion of the system not (9) 5.4.4 Main Steamline Flow Restrictors undergoing test. The capability for inservice testing ensures the high functional reliability (10) 5.4.5 Main Steamline Isolation Valves of protection and reactivity control systems if a reactor variable exceeds the corrective action  ; (11) 7.7.1.2 Rod Control and Information setpoint. l and System - Instrumentation and 7.7.2.2 Controls The capabilities of the protection and reactivity control systems to perform their (12) 15 Accident Analyses safety functions in the event of anticipated operational occurrences satisfy the requirements 3.1.23.10 Criterion 29 Protection Against of Criterion 29. l Anticipated Operational Occurrences l For further discussion, see the following 3.1.2 3.10.1 Criterion 29 Statement seetions: The protection and reactivity control systems shall be designed to assure an extremely high Chapter / Section Iilk g , probability of accomplishing their safety l functions in the event of anticipated operational (1) 1.2.1 Principal Design Criteria ' occurrences. i (2) 5.4.5 Main Steamline Isolation System i 3.1.2 3.10.2 Daluation Against Criterion 29 (3) 5.4.7 Residuallleat Removal System The high functional reliability of the reactor protection (trip) system and reactivity control (4) 6.2 Containment Systems system is achieved through the combination of logic arrangement, redundancy, physical and (5) 63 Emergency Core Cooling Systems electrical independence, functional separation, fail-safe design, and inservice testability. (6) 7.2 Reactor Trip System These design features are discussed in detail in criteria 21, 22, 23, 24, and 26. (7) 73 Engineered Safety Feature Systems An extremely .gh reliability of timely re-sponse to anticipated operational occurrences is (8) 15 Accident Analyses maintained by a thorough program of inservice testing and surveillance. Active components can 3.1.2.4 Group IV Fluid Systems be tested or removed from service for maintenance during reactor operation without compromising the 3.1.2.4.1 Criterion 30 - Quality of Reactor protection or reactivity control functions even Coolant Pressure lloundary in the event of a subsequent single failure. Components important to safety, such as control 3.1.2.4.1.1 Criterion 30 Statement Amendment 1 3122

ABWR m6twAn Standard Plant REV.A Components which are part of the reactor vided by the RCIC system. While the leak detec-coolant pressure boundary shall be designed, tion system provides protection from small O_ fabricated, erected, and tested to the highest leaks, the ECCS network provides protection for quality standards practical. Means shall be the complete range of discharges from ruptured provided for detecting and, to the extent pipes. Thus, protection is provided for the practical, identifying the location of the source full spectrum of possible discharges. of reactor coolant leakage. The reactor coolant pressure boundary and the 3.1.2.4.1.2 Evaluation Against Criterion 30 leak detection and isolation system are designed to meet requirements of Criterion 30. By utilizing conservative design practices and detailed quality control procedures, the For further discussion, see the following pressure retaining components of the reactor sections: coolant pressure boundary (RCPB) are designed and fabricated to retain their integrity during Chapter / normal and postulated accident conditions (See Section Tilk Subsection 3.1.2.2.5.2). Accordingly, compo-nents which comprise the RCPB are designed, (1) 1.2.1 Principal Design Criteria fabricated, erected, and tested in accordance with recognized industry codes and standards (2) 3.2 Classification of Structures, listed in Chapter 5 and Table 3.21. Further, Components, and Systems product and process quality planning is provided as described in Chapter 17 to assure conformance (3) 5.2.2 Overr.essurization Protection with the applicable codes and standards, and to retain appropriate documented evidence verifying (4) 5.2.5 Detectbn of Reactor Coolant compliance. Because the subject matter of this 1.cakage Through Reactor Coolant criterion deals with aspects of the RCPB further Pressure Boundary discussion on this subject is treated in the O response to Criterion 14. (5) 5.3 Reactor Vessel Means are provided for detecting leakage in (6) 5.4.1 Reactor Recirculation Pumps the reactor coolant system (RCS). The leak de. tection and isolation system consists of sensors (7) 7.3.1.2 Leak Detection and Isolation and instruments to detect, annunciate, and, in and System Instrumentation and some cases, isolate the reactor coolant pressure 73.2.2 Controls boundary from potential hazardous leaks before predetermined limits are exceeded. Small leaks (8) 7.7.1.1 Reactor Vessel Instrumentation are detected by temperature and pressure changes, increased frequency of sump pump operation, and (9) 17 Quality Control System by measuring fission product concentration, in addition to these means of detection, large leaks 3.1.2.4.2 Criterion 31. Fracture Presention I are detected by changes in flow rates in process of Reactor Coolant Pressure Boundary lines, and changes in reactor water level. The allowable leakage rates have been based on the 3.1.2.4.2.1 Criterion 31 Statement I predicted and experimentally determined behavior ) of cracks in pipes, the ability to make up the The reactor coolant pressure boundary shall i RCS the normally expected background leakage due be designed with sufficient margin to assure to equipment design, and the detection capability that when stressed under operating, maintenance, of the various sensors and instruments. The testing, and postulated accident conditions: total leakage rate limit is established so that, in the absence of normal ac power with loss of (1) the boundary behaves in a nonbrittle manner, feedwater supply, makeup capabilities are pro- and O l Amendrent 1 3.1-23

ABM ux61oore Standard Plant REV.A (2) the probability of rapidly propagating are accounted for in the reactor operation. fracture is minimized. g The reactor coolant pressure boundary is de- w The design shall reflect consideration of signed, maintained, and tested to provide ade-service temperatures and other conditions of the quate assurance that the boundary will behave in boundary material under operating, maintenance, a non brittle manner throughout the life of the testing, and postulated accident conditions and plant. Therefore, the reactor coolant pressure the uncertainties in determining: boundary is in conformance with Criterion 31. (1) material properties; For further discussion, see the following sections: (2) the effects of irradiation on material properties; Chapter / Sectioti liik (3) residual, steady state, and transient stresses; and (1) 3 Design of Structures,Compon-ents, Equipment and Systems (4) size of flaws. (2) 5.2 Integrity of Reactor Coolant 3.1.2.4.2.2 Evaluation Against Criterion 31 Pressure Boundary Brittie fracture control of pressure- 3.1.2.43 Criterion 32 -Inspection of Reactor retaining ferritic materials is provided to Coolant Pressure Boundary ensure protection against nonductile fracture. To minimize the possibility of brittle fracture 3.1.2.43.1 Criterion 32 Statement failure of the reactor pressure vessel, the reactor pressure vessel is designed to meet the Compor ats which are part of the reactor requirements of ASME Code Section 111. coolant pressure boundary shall be designed to permit: (1) periodic inspection and testing of g The nil ductility transition (NDT) temperature important areas and features to assess their is defined as the temperature below which structural and leaktight integrity and, (2) an ferritic steel breaks in a brittle rather than appropriate material surveillance program for ductile manner. The NDT temperature increases as the reactor pressure vessel. a function of neutron exposure at integrated neutron exposures greater than about 1x1017 nyt 3.1.2.43.1 Evaluation Against Criterion 32 with neutron of energies in excess of 1 MeV. The reactor pressure vessel design and engi. The reactor assembly design provides an annu- neering effort include provisions for inservice tar space from the outermost fuel assemblics to inspection. Removable plugs in the reactor the inner surface of the reactor vessel that shield wall and/or removable panels in the insu-serves to attenuate the fast neutron flux inci- lation provide access for examination of the dent upon the reactor vessel wall. This annual vessel and its appurtenances. Also, removable voluna contains the core shroud, and reactor insulation is provided on the reactor coolant coolant. Assuming plant operation at rated power system safety / relief valves, and on the main and availability of 100% for the plant life time, steam and feedwater systems extending out to and the neutron fluence at the inner surface of the including the first isolation valve outside con-vessel causes a slight shift in the transition tainment. Inspection of the reactor coolant temperatures. Expected shifts in transition pressure boundary is in accordance with the ASME temperature during design life as a result of Boiler and Pressure Vessel Code Section XI. environmental conditions, such as neutron flux, Secion 5.2 defines the Inservice Inspection are considered in the design. Operational Plan, access provisions, and areas of restricted limitations assume that NDT temperature shifts access. O Amendment 1 3144

MM 23A6100AE Standard Plant REV.A Vessel material surveillance samples will be rature and pressure changes, increased frequency p located within the reactor pressure vessel. The of sump pump operation, and by measuring fission d program willinclude specimens of the base metal, product concentration. In addition to these weld metal, and beat affected zone metal. means of detection, large leaks are detected by changes it. riow rates in process lines and chan-The plant testing and inspection program ges in reactor water level. The allowable leak-ensure that the requirements of Criterion 32 will age rates have been based on predicted and expe-be met. rimentally determined behavior of cracks in pipes, the ability to make up reactor coolant For further discussion, see the following leakage, the normally expected background sections: leakage due to equipment design, and the detection capability of the various sensors and Chapter / instruments. The total leakage rate limit is Section Iith established so that,in the absence of normal AC power containment with a loss of feedwater (1) 3.9 Mechanical Systems and supply, makeup capabilities are provided by the Components RCIC system. (2) 5.2 Integrity of Reactor Coolant The plant is designed to provide ample Pressure Boundary reactor coolant makeup for protection against small leaks in the reactor coolant pressure 3.1.2.4.4 Criterion 33 Reactor Coolant boundary for anticipated operational occurrences Makeup and postulated accident conditions. The design of these systems meets the requirements of 3.1.2.4.4.1 Criterlon 33 Statement Criterion 33. A system to supply reactor coolant makeup for For further discussion, see the following

 ) protection against small breaks in the reactor    sections:

molant pressure boundary shall be provided. The system safety function shall assure that Chapter / specified acceptable fuel design limits are not Section Iilk exceeded as a result of reactor coolant loss due to leakage from the reactor coolant pressure (1) 5.2.5 Detection of Reactor Coolant boundary and rupture of small piping or other Leakage Through Reactor Coolant small components which are part of the boun- Pressure Boundary. dary. The system shall be designed to assure that, for onsite electric power system operation (2) 5.4.6 Reactor Core Isolation Cooling (assuming offsite power is not available) and for System offsite electric power system operation (assuming onsite power is not available), the system safety (3) 63 Emergency Core Cooling Systems function can be accomplished using the piping, pumps, and valves used to maintain coolant (4) 7.3.1.2 Leak Detection and lsolation inventory during normal reactor operation, and System - Instrumentation and 7.3.2.2 Controls i 3.1.2.4.4.2 Response to Criterion 33 l 3.1.2.4.5 Criterion 34 Residual Heat Removal Means are provided for detecting reactor cool-l ant leakage. The leak detection and isolation 3.12.4.5.1 Criterion 34 Statement , system consists of sensors and instruments to detect, annunciate, and, in some cases, isolate A system to remove residual heat shall be the reactor coolant pressure boundary from poten- provided. The safety function shall be to tial hazardous leaks before predetennined limits transfer fission product decay heat and other are exceeded. Small leaks are detected by tempe- residual heat from the reactor core at a rate O Amendmert 1 3.1 25

ABWR m6mc Standard Plant anv. A such that specified acceptable fuel design limits Loads important to plant operation and safety and the design conditions of the reactor coolant are split and diversified between switchgear \a W pressure boundary are not exceeded, sections, and means are provided for detection and isolation of system faults. Suitable redundancy in components and features and suitable interconnections, leak detection, The plant layout is designed to effect and isolation capabilities shall be provided to physical separation of essential bus sections, assure that, for onsite electric power system standby generators, switchgear, interconnec-operation (assuming offsite power is not tions, feeders, power centers, motor control available) and for offsite electric power system centers, and other system components, operation (assuming onsite power is not available), the system safety function can be Full capacity standby diesel generators are accomplished, assuming a single failure. provided to supply a source of electrical power which is self-contained within the nuclear 3.1.2.4.5.2 Esaluation Against Criterion 34 island and is not dependent on external sources of supply. The standby generators produce AC The RHR System provides the means to remove power at a voltage and frequency compatib.'e with decay heat and residual heat from the nuclear the normal bus requirements for essential steam supply systems (NSSS) so that refueling and equipment within the plant. Each of the diesel servicing of NSSS can be performed. generators has sufficient capacity to start and carry the essential loads it is expected to The major equipment of the RHR system consists drive, of heat exchangers, main system pumps, and 1 service water pumps. The equipment is connected The RHR system is adequate to remove residual by associated valves and piping, and the controls heat from the reactor core to assure fuel and acd instrumentation are provided for proper reactor coolant pressure boundary design limits system operation. are not expected. Redundant reactor coolant circulation paths are available to and from the g Three independent loops are located in vessel and RHR system. Redundant onsite separate protected areas, electric power systems are provided. The design of the RHR system, including its power supply, Both normal AC power and the auxiliary onsite meets the requirements of Criterion 34. power system provide power adequate to operate all the auxiliary loads necessary for plant ope- For further discussion, see the following ration. The power sources for the plant auxi. sections: i liary power system are sufficient in number and I of such electrical and physical independence that Chapter / l no single failure will prevent auxiliary systems Section Illk from supporting two of the three RHR divisions. (1) 5.4.7 Residual Heat Removal System The plant auxiliary buses supplying power to engineered safety features and reactor protection (2) 6.3 Emergency Core Cooling System systems and those auxiliaries required for safe shutdown are connected by appropriate switching (3) 7.3.1.1 Emergency Core Cooling System - to standby diesel. driven generators located in and Instrumentation and Controls the plant. Each power source, up to the point of 7.3.2.1 its connection to the auxiliary power buses, is capable of complete and rapid isolation from any (4) 8.2 Onsite Power Systems other source. O Amendment 1 3.1 26

23A6100AE Standard Plant REV A pressure boundary including a complete and (5) 9.2 Water Systems sudden circumferential rupture of the larger V pipe connected to the reactor vessel. (6) 15 Accident Analyses The HPCF system coasists of two subsystems, 3.1.2A.6 Criterion 35 Emergency Core each having a single motor driven pump, system Cooling piping, valves, controls, and instrumentation. The RCIC system consists of similar equipment 3.1.2A.6.1 Criterion 35 Statement except that it is a single system and the pump delivering high pressure flow is driven by steam A system to provide abundant emergency core turbine. The HPCF and RCIC systems assure that cooling shall be provided. The system safety the reactor core is adequately cooled to prevent function shall be to transfer heat from the excessive fuel clad temperatures for breaks in reactor core following any LOCA at a rate such the nuclear steam supply systems (NSSS) which do that: not result in rapid depressurization of the reactor vessel. The HPCF or RCIC system (1) fuel and clad damage that could interfere continues to operate when reactor vessel with continued effective core cooling is pressure is below the pressure at which the prevented, and RHR/LPFL system operation maintains core cooling. A source of water is available from (2) clad metal water reaction is limited to either the condensate storage pool or the negligible amounts, suppression pool. Suitable redundancy in components and features The ADS functions to reduce the reactor pres-and suitable interconnections, leak detection, sure so that flow from RHR/LPFL enters the reac-isolation, and containment capabilities shall be tor vessel in time to cool the core and prevent provided to assure that, for onsite electric excessive fuel clad temperature. The ADS uses power system operation (assuming offsi e power is several of the Nuclear Boiler System safety / not available) and for offsite electric power relief valves to relieve the high pressure steam system operation (assuming onsite power is not to the suppression pool. available), the system safety function can be accomplished, assuming a single failure. The HPCF system consists of a centrifugal pump that can be powered by normal auxiliary 3.1.2A.6.2 Evaluation Against Criterion 35 power of the standby AC power system, a core flooder sparger in the reactor vessel above the The Emergency Core Cooling System (ECCS) core, piping and valves to convey water from the consists of the following: condensate storage pool or the suppression pool to the sparger, and associated controls and (1) High Pressure Core Flooder (HPCF) System; instrumentation. In case of low water level in the reactor vessel or high pressure in the (2) Reactor Core Isolation Cooling (RCIC) drywell, the HPCF system automatically injects System; water into the vessel in time and at a sufficient flow rate to cool the core and , (3) Low Pressure Flooder (LPFL) mode of the prevent excessive fuel temperature. I Residual Heat Removal System (RHR); and In case of low water level in the reactor or (4) Automatic Depressurization System (ADS). high pressure in the drywell, the LPFL mode of operation of the RHR system pumps and injects The ECCS is designed to limit fuel claddirg water into the reactor vessel in time to flood temperature over the complete spectrum of the core and prevent excessive fuel tempera-possible break sizes in the reactor coolant ture. The RHR system is described in Subsection O Amendment 1 3.1 27

ABWR DAMMAE Standard Plant REV A 3.1.2.4.5.2. Protection provided by RiiR/LPFL (3) 63 Emergency Core Cooling System extends to a small break where the ADS has operated to lower the reacter vessel pressurt. (4) 73.1.1 Emergency Core Cooling System - and Instrumentation and Controls Results of the performance of the ECCS for the 73.2.1 entire spectrum of liquid line breaks are discussed in Subsection 6.3.3. Peak cladding (5) 83 Onsite Power Systems temperatures are well below the 22000F design basis. (6) 9.2 Water Systems Also provided in Subsection 6.3.3 is an (7) 15 Accident Analyses analysis to show that the ECCS conforms to the 10CFR50 Appendix K. This analysis shows complete 3.1.2.4.7 Criterion 36 Inspection of complian:e with the Criterion 35 with the Emergency Core Coollag System following results: 3.1.2.4.7.1 Criterion 36 Statement (1) Peak clad temperatures are well below the 22000F NRC acceptability limit; The emergency core cooling system shall be designed to permit appropriate periodic (2) The amount of fuel cladding reacting with inspection of important components, such as steam is nearly an order of magnitude below spray rings in the reactor pressure vessel, the 1% acceptability limit; water injection nozzles, and piping, to assure the integrity and capability of the system. (3) The clad temperature iransient is terminated while core geometry is still amenable to 3.1.2.4.7.2 Evaluation Against Criterion 36 cooling; and The ECCS discussed in Criterion 35 includes (4) The core temperature is reduced and the decay inservice inspection considerations. The spray heat can be removed for an extended period of spargers within the vessel are accessible for g time, inspection during each refueling outage. Removable plugs in the reactor shield and/or The redundancy and capability of the onsite panels in the insulation is prosided on the ECCS electrical power systems for the ECCS are repre- piping up to and including the first isolation sented in the evaluation against Criterion 34. valve outside the drywell. Inspection of the ECCS is in accordance with the intent of ASME The ECCS is adequate to prevent fuel and clad Code Section XI. Subsection 5.2.4 defines the damage which could interfere with effective core Inservice Inspection Plan, access provisions, cooling and to limit clad metal water reaction to and areas of restricted access, a negligible amount. The design of the ECCS, including the power supply, meets the During plant operations, the pumps, valves, requirements of Criterion 35. piping, instrumentation, wiring, and other components outside the containment can be For further discussion, see the following visually inspected at any time. Components sections: inside the containment can be inspected when the containment is open for access. When the Chapter / reactor vessel is open for refueling or other Section liik purposes, the spargers and other internals can be inspected. Portions of the ECCS which are (1) 5.4.7 Residual licat Removal System part of the reactor coolant pressure boundary are designed to specifications for inservice (2) 5.4.6 Reactor Core isolation Cooling inspection to detect defects which might affect System the cooling performance. Particular attention 9 Amendment 1 3128

ABM 23461oorn Standard Plant REV.A (~* will be given to the reactor nozzles, and core power sources, and the operation of the asso-i flooder spargers. The design of the reactor ves- ciated cooling water system, set and internals for inservice inspection and the plant testing and inspection program ensures 3.1.2.4.8.2 Evaluation Against Criterion 37 that the requirements of Criterion 36 will be met. The ECCS consists of the HPCF system, RCIC system, LPFL mode of the RHR system, and the ADS For further discussion, see the following system. Each of these systems is provided with sections: sufficient test connections and isolation valves to permit appropriate periodic pressure testing Chapter / to assure the structural and leaktight integrity Section Iitic of its components. (1) 3.9.5 Reactor Pressure Vessel Each of the ECCS systems is designed to Internals permit periodic testing to assure the operability and performance of the active (2) 5.2.4 Inservice Inspection and Testing components of each system. of Reactor Coolant Pressure Boundary The pumps and valves of these systems will be tested periodically to verify operability. Flow (3) 53 Reactor Vessel rate tests will be conducted on the HPCF system, RCIC system, and RHR/LPFL system. l (4) 63 Emergency Core Cooling Systems I The ECCS will be subjected to tests to verify ) (5) 6.6 Inservice Inspection of Class 2 the performance of the full operational sequence and 3 Components that brings each system into operation. The operation of the associated cooling water sys-3.1.2.4.8 Criterion 37 Testing of Emergency tems is discussed in the evaluation of Criterion Core Cooling System 46. It is concluded that the requirements of Criterion 37 are met. 3.1.2.4.8.1 Criterion 37 Statement For further discussion, see the following The emergency core cooling system shall be sections: designed to permit appropriate periodic pressure and functional testing to assure, Chapter / Section Iitic (1) the structural and leaktight integrity of its components; (1) 5.2.2 Overpressurization Protection (2) the operability and performance of the active (2) 63 Emergency Core Cooling Systems components of the system; and (3) 73.1.1 Emergency Core Cooling Systems - (3) the operability of the system as a whole and, Instrumentation and Controls under conditions as close to design as practical, the performance of the full (4) 83.1 AC Power Systems operational sequence that brings the system into operation including operation of (5) 16 Technical Specifications applicable portions of the protection system, the transfer between normal and emergency 3.1.2.4.9 Criterion 38 Containment lleat Removal O Amendment 1 3.129

M 23A6100AE Standard Plant REV.A 3.1.2.4.9.1 Criterion 38 Statement For further discussion, see the following sections: - A system to remove heat from :he reactor containment shall be prosided. The system safety Chapter / function shall be to reduce rapidly, consistent Section Iitic with the functioning of other associated sys-tems, the containment pressure and temperature (1) 5.4.7 Residual Heat Removal System following any LOCA and maintain them at acceptable low levels. (2) 6.2.2 Containment Heat Removal Systems Suitable redundancy in components and features (3) 8.3.1 AC Power Systems and suitable interconnections, leak detection, isolation, and containment capabilities shall be (4) 9.2 Water Systems provided to assure that, for onsite electric power system operation (assuming offsite power is (5) 13 Accident Analyses not available) and for offsite electric power system operation (assuming onsite power is not 3.1.2.4.10 Criterion 39 Inspection of available), the system safety function can be Containment ileat Removal System accomplished, assuming a single failure. 3.1.2.4.10.1 Criterion 39 Statement 3.1.2.4.9.2 Evaluation Against Criterion 38 The containment heat removal system shall be The containment heat removal function is designed to permit appropriate periodic inspec-accomplished by the suppression pool cooling mode tion of important components, such as the torus, of the RilR. Following a LOCA, suppression pool sumps, spray nozzles, and piping, to assure the cooling mode limits the temperature within the integrity and capability of the system, wetwell by recirculating the suppression pool water and removing heat via the RHR system heat exchangers. Any or all redundant RilR system heat 3.1.2.4.10.2 Evaluation Against Criterion 39 g exchangers can be manually activated. This Provisions are made to facilitate periodic subsystem is initiated manually following inspections of active components and other indiction of high suppression pool temperature. important equipment of the containment heat If a LOCA signal is present, the RilR will removal systems. During plant operations, the function in the core cooling (LPFL) mode. pumps, valves, piping, instrumentation, wiring, and other components outside the containment can Following a LOCA, wetwell and drywell spray be visually inspected at any time and will be mode of the RilR condenses steam within the inspected periodically. Such components inside drywell and wetwell zones of the containment by the containment will be tested and inspected spraying suppression pool water cooled through during periodic outages. The testing the heat exchangers. Wetwell/drywell spray is frequencies of most components will be started manually. The drywell spray mode is correlated with the component inspection. initiated by operator action post-LOCA in the presence of high drywell pressure. The wetwell The suppression pool is designed to permit spray mode can be manually initiated in the appropriate periodic inspection. Space is control room, unless an overriding LOCA signal provided outside the containment for inspection for the LPFL is present. The wetwell spray mode and maintenance. does not depend on the operation of the suppres-sion pool cooling mode. The containment beat removal system is designed to permit periodic inspection of major The redundancy and capability of the offsite componenh This design meets the requirements and ensite electrical power systems for the RilR of Criterion 39. system is presented in the evaluation against Criterion 34. For further discussion, see the following sections: hl l Amendment 1 3,1 30 I

ABM 22^6ioo^n Standard Plant REV.A Chapter / Design Criterion 46. It is concluded that the s Section Ilth requirements of Criterion 40 are met. (1) 5.4.7 Residual Heat Removal System For further discussion see Subsection 6.2.2. (2) 6.2 Containment Systems 3.1.2.4.12 Criterion 41 Containment Atmosphere Cleanup (3) 63 Emergency Core Cooling Systems 3.1.2.4.12.1 Criterion 41 Statement (4) 9.2 Water Systems Systems to control fission products, 3.1.2.4.11 Criterion 40 Testing of hydrogen, oxygen, and other substances which may Containment Heat Removal System be released into the reactor containment shall be provided as necessary to reduce, consistent 3.1.2.4.11.1 Criterion 40 Statement with the functioning of other associated systems, the concentration and quantity of The containment heat removal system shall be fission products released to the environment signed to permit appropriate periodic pressure following postulated accidents and to control nd functional testing to assure: the concentration of hydrogen or oxygen and other substances in the containment atmosphere (1) the structural and leaktight integrity of its following postulated accidents to assure that components; containment integrity is maintained. (2) the operability and performance of the active Each system shall have suitable redundancy in components of the system; and components and features and suitable interconnections, leak detection, isolation, and (3) the operability of the system as a whole, containment capabilities to assure that, for o and, under conditions as close to the design onsite electric power system operation (assuming C as practical, the performance of the full offsite power is not available) and for offsite operational sequence that brings the system electric power system operation (assuming onsite into operation including operation of power is not available), its safety function can applicable portions of the protection system, be accomplished, assuming a single failure. the transfer between normal and emergency power sources, and the operation of the 3.1.2.4.12.2 Evaluation Against Criterion 41 associated cooling water system. I The quantity of fission products releaseo 3.1.2.4.11.2 Evaluation Against Criterion 40 into the environment following postulated acci. I dents is controlled by the standby gas treatment  ; The containment heat removal function is systun (SGTS) that has the redundancy and capa. ' accomplished by a suppression pool cooling mode bility to filter the gaseous effluent from the of the RHR system. primary and the secondary containment. The SGTS l filters are classified as passive components. l The RHR system is provided with sufficient I test connections and isolation valves to permit Hydrogen is controlled to below cornbustible l periodic pressure and flow rate testing. limits by the atmospheric control system (ACS) that maintains an inert atmosphere. The pumps and valves of the RHR system will be l operated periodically to verify operability. The These systems have design provisions to ) cooling mode is not au:omatically initiated, but ensure that safety function is accomplished,  ; operation of the components is periodically assuming a single failure. These systems meet i verified. The operation of associated cooling the requirements of Criterion 41. For further I water systems is discussed in the response to discussion, see the following sections: O Amendment 1 3.1 31

M 23A6100AE Standard Plant any.A Chapter / (2) 6.2.5 Combustible Gas Controlin Section Iille Containment (1) 1.2 General Plant Description (3) 6.5.1 Engineered Safety Features Filter System (2) 6.2.5 Combustib!c Gas Controlin Containment (4) 6.5.3 Fission Product Control Systems (3) 6.5.1 Engineered Safety Features (5) 6.6 Inservie: Inspection of Class 2 Filter System and 3 Components (4) 6.53 Fission Product Control Systems (6) 7 Instrumentation and Controls (5) 7 Instrumentation and Controls (7) 8 Electric Power (6) 8 Electric Power (8) 9.5.9 Suppression Pool Cleanup System (7) 9.5.9 Suppression Pool Cleanup System 3.1.2.4.14 Criterion 43 Testing of Containment Atmosphere Cleanup Systems (8) 15 Accident Analyses 3.1.2.4.14.1 Criterion 43 Statement 3.1.2.4.13 Criterion 42. Inspection of Containment Atmosphere Cleanap System The containment atmosphere cleanup systems shall be designed to permit appropriate periodic 3.1.2.4.13.1 Criterion 42 Statement pressure and functional testing to assure: The containment atmosphere cleanup systems (1) the structural and leaktight integrity of shall be designed to permit appropriate periodic inspection of important components, such as its components; g filter frames, ducts, and piping, to assure the (2) the operability and perfortnance of the integrity and capability of the systems, active components of the systems such as fans, filters, dampers, pumps, and valves; 3.1.2.4.13.2 Evaluation Against Criterion 42 and Except for components located in the contain- (3) the operability of the systems as a whole ment and steam tunnel, all components of the and, under conditions as close to design as fission product control system can be inspected practical, the performance of the full during normal plant operation at power. The operational sequence that brings the systems l components within (be containment and steam into operation including operation of l tunnel may be inspected during refueling and applicable portions of the protection maintenance outages, system, the transfer between normal and emergency power sources, and the operation The design of the system, therefore, meets the of associated systems. requirements of Criterion 42. For further discussion, see tbe following sections: 3.1.2.4.14.2 Evaluation Against Criterion 43 Chapter / All active components of the fission product Section lille control system can be tested during normal plan operation at power. The SGTS filters are (1) 1.2 General Plant Description classified as passive components. O Amendment 1 3.1-32

23A6100AE Standard Plant REV.A 1 Complete system operation can be tested during designed with suitable redund tacy, isolation ,, reactor shutdown, capability, and separation suct. that no single (j failure prevents a safe plant shutdown. , The design of the system, therefore, meets the i requirements of Criterion 43. For further The design of this system meets the j discussion, see the following sections: requirements of Criterion 44. Chapter / For further discussion, see the following Section Iith sections: (1) 1.2 General Plant Description Chapter / Section Ilth (2) 6.2.5 Combustible Gas Controlin Containment (1) 1.2 General Plant Description (3) 6.5.3 Fission Product Control System (2) 9.2 Water Systems (4) 7 Instrumentation and Control 3.1.2.4.16 Critedon 45. Inspection of Cooling Water System (5) 8 Electric Power 3.1.2.4.16.1 Criterion 45 Statement 3.1.2.4.15 Criterion 44 Cooling Water The cooling water system shall be designed to 3.1.2.4.15.1 Criterion 44 Statement permit appropriate periodic inspection of impor-tant components such as heat exchangers and pip. A system to transfer heat from structures, ing to assure the integrity and capability of systems, and components important to safety to an the system, ultimate heat sink shall be provided. The system pV safety function shall be to transfer the combined 3.1.2.4.16.2 Evaluation Against Criterion 45 heat load of these structures, systems, and components under normal operating and accident All important components in the Nuclear conditions. Island scope are locate <1 in accessible locations to facilitate periodic inspection during normal Suitable redundancy in components and features plant operation. Suitable manholes, handholes, and suitable interconnections, leak detection, inspection ports, or other design and layout and isolation capabilities shall be provided to features are provided for this purpose. assure that for onsite electric power system operation (assuming offsite power is not These features meet the requirements of available) and for offsite electric power systems Criterion 45. operation (assuming onsite power is not available), the system safety function c. n be For further discussion, see the following accomplished, assuming a single failure, sections: 3.1.2.4.15.2 Evalnation Against Criterion 44 Chapter / Section Ilth The system provided to transfer heat from safety related equipment to the ultimate heat (1) 1.2 General Plant Description sink is the reactor building cooling water system. (2) 9.2 Water Systems This system is operable either from offsite (3) 14 Initial Test Program power or from onsite emergency power and is Amendment 1 3.133

l ABM 23A61 corn Standard Plant REV.A 3.1 2.4.17 Criterion 46 Testing of Cooling (4) 16 Technical Specifications Water System 3.1.2.4.17.1 Criterion 46 Statement 3.1.2.5 Group V Reactor Containment h 3.1.2.5.1 Criterion 50. Containment Design The cooling water system shall be designed to Basis permit appropriate periodic pressure and functional testing to assure: 3.1.2.5.1.1 Criterion 50 Statement (1) the structural and leaktight integrity of its The reactor containment structure, including components; access openings, penetrations, and the contain-ment heat removal system shall be designed so (2) the operability and the performance of the that the containment structure and its internal active components of the system; and compartments can accommodate, without exceeding the design leakage rate and with sufficient mar-(3) the operability of the system as a whole and, gin, the calculated pressure and temperature under conditions as close to design as conditions resulting from any loss of coolant practical, the performance of the full accident. This margin shall reflect consider-operational sequence that brings the system ation of: into operation for reactor shutdown and for loss of coolant accidents, including (1) the effects of potential energy sources operation of applicable portions of the which have not been included in the determi-protection system and the transfer between nation of the peak conditions, such as normal and emergency power sources, energy in steam generators and as required by Section 50.44 energy from metal water and 3.1.2.4.17.2 Evaluation Against Criterion 46 other chemical reactions that may result from degraded emergency core cooling Redundancy and isolation are prosided to allow periodic pressure and functional testing of the functioning; g system as a whole including the functional (2) the limited experience and experimental data sequence that initiates system operation. This available for defining accident phenomena also includes transfer between the offsite power and containment responses; and supply and the onsite emergency diesel. generator power supply. At least one of the redundant (3) the conservatism of the calculational model systems is in service during normal plant and input parameters. operations. 3.1.2.5.1.2 Evaluation Against Criterion 50 The system design thus meets the requirement < of Criterion 46. Design of the containment is based on the safe shutdown earthquake (SSE) postulated to For further discussion, see the following occur at the site simultaneously with the design sections: basis accident (DBA) which is defined as the worst loss of coolant accident (LOCA) pipe break Chapter / having the consequences of maximum containment Section Ilt and drywell pressure and/or temperature. These conditions are coupled with the loss of offsite (1) 1.2 General Plant Descriptien power. (2) 9.2 Water Systems The maximum pressure and temperature reached in the drywell and containment during this (3) 14 Initial Test Program worst case accident are shown in Chapter 6 to be O Amendment 1 3.1-M

ABM nA6too^n Standard Plant REV.A well below the design pressure and temperature of 3.1.23.2.2 Evaluation Against Criterion 51 p the structures. This provides an adequate margin d for uncertainties in potential energy sources. The primary containment vessel (PCV) is a re-inforced concrete structure with ferritic parts, The design of the containment system thus such as removable head, which are made of mate-meets the requirements of Criterion 50. rit.1 that has a nil ductility transition tem. perature of at least 30 F below the minimum For further discussion, see the following service temperature. sections: The PCV is enclosed by and is integrated with Chapter / the reinforced concrete reactor building. The Section Iit]g pre operational test program and the quality assurance program ensure the integrity of the (1) 3.7 Seismic Design containment and its ability to meet all normal operating and accident requirements. (2) 3.8 Design of Seismic CategoryI Structures The containment design thus meets the requirements of Criterion 51. (3) 6.2.1 Containment Functional Design For further discussion, see the following (4) 6.2.2 Containment Heat P.emoval System seetions: (5) 15 Accident Analyses Chapter / Section Iills 3.1.2.5.2 Criterion 51 Fracture Prevention of Containment Pressure Houndari (1) 3.8 Design of Seismic Category I Structures 3.1.2.5.2.1 Criterion 51 Statement (2)17 Quality Assurance The reactor containment boundary shall be designed with sufficient margin to assure that 3.1.2.53 Criterion 52. Capability for under operating, maintenance, testing, and Containment leakage Rate Testing postulated accident conditions: 3.1.2.53.1 Criterion 52 Statement (1) its ferritic materials behave in a nonbrittle manner; and The reactor containment and other equipment which may be subjected to containment test (2) the probability of rapidly propagating conditions shall be designed so that periodic fracture is minimized. integrated leakage rate testing can be conducted at containment design pressure. The design shall reflect consideration of service temperatures and other conditions of the 3.1.2.53.2 Evaluation Against Criterion 52  ; containment boundary material during operation, maintenance, testing, and postulated accident The containment system is designed and conditions and the uncertainties in determining: constructed and the necessary equipment is I provided to permit periodic integrated leak rate (1) materialproperties; tests during the plant lifetime. The testing program is conducted in accordance with 10CFR50 (2) residual, steady-state, and transient Appendix J. stresses; and The testing provisions provided and the test (3) size of flaws. program meet the requirements of Criterion 52. O Amendment 1 3.1 35

ABM 2mioore Standard Plant REV.A For further discussion, see the following (2) 6.2.6 Containment leakage Testing sections: Chapter / 3.1.23.5 Criterion 54 Piping Systems Penetrating Containment h Section Ilth 3.1J.5.5.1 Criterion 54 Statement (1) 3.8.2.7 Testing and Inservice Inspection Requirements Piping systems penetrating primary reactor containment shall be provided with leak detec. (2) 6.2.6 Containment Leakage Testing tion, isolation, and containment capabilitics having redundancy, reliability, and performance 3.1.2.5.4 Criterion 53 Provisions for capabilities which reflect the importance to Containment Testing and Inspection safety of isolating these piping systems. Such piping systems shall be designed with a capabi. 3.1.2.5.4.1 Criterion 53 Statement lity to periodically test the operability of the isolation valves and associnted apparatus and to determine if valve leakage is within acceptable The reactor containment shall be designed to limits. permit: 3.1.2.5.5.2 Evaluation Against Criterion 54 (1) appropriate periodic inspection of all important areas such as penetrations; Piping systems penetrating containment are designed to provide the required isolation and (2) an appropriate surveillance program; and testing capabilities. These piping systems are provided with test connections to allow periodic (3) periodic testing at containment design pres- leak detection tests as necessary to determine sure of the leaktightness of penetrations if valve leakage is within acceptable limits, which have resilient seals and expansion 3 bellows. The actuation test circuitry provides the W means for testing isolation valve operability as 3.1.2.5.4.2 Evaluation Against Criterion 53 necessary to determine if operability is within acceptable limits. There are special provisions for conducting individual leakage rate tests on applicable The design and provisions made for piping penetrations. Penetrations are visually systems penetrating containment meet the i inspected and pressure tested for leaktightness requirements of Criterion 54. at periodic intervals in accordance with 10CFR50 Appendix J. 3.1.2.5.6 Criterion 55 Reactor Coolant Pressure Boundary Penetrating Containment The provisions made for protection testing meet the requirements of Criterion 53. 3.1.2.5.6.1 Criterion 55 Statement For further discussion, see the following Each line that is part of the reactor coolant sections: pressure boundary and that penetrates primary reactor containment shall be provided with Chapter / containment isolation valves as follows, unless Section Iith it can be demonstrated that the containment isolation provisions for a specific class of (1) 3.8 Design of Seismic Category l lines such as instrument lines, are acceptable Structure on some other defined basis: O Amendment 1 3.1 36

ABM ux61oaAs Standard Plant anv. A (1) One locked. closed isolation valve inside and significant release of radioactivity Similar-one locked closed isolation valve outside ly, for lines which do not penetrate the O ceniain- nii er ceniain- ni hnt which rer- a eeriien ef the reactor coolant pressure boundary, the design (2) One automatic isolation valve inside and one ensures that suitable isolation from the reactor locked closed isolation valve outside coolant pressure boundary can be achieved, containment; or The design of the isolation systems detailed (3) One locked-closed isolation valve inside and in the following sections meets the requirements one automatic isolation valve outside of Criterion 55, containment. A simple check valve may not be used as the automatic isolation valve outside For further discussion, see the following containment; or sections: (4) One automatic isolation valve inside and one Chapter / automatic isolation valve outside Section Iith containment. A simple check valve may not be used as the automatic isolation va'se outside (1) 5.2 Integrity of Reactor Coolant containment. Pressure Boundary Isolation valves r atside containment shall be (2) 6.2.4 Containment Isolation Systems located as close i , the containment as practical and upon loss c' actuating power, automatic (3) 7 Instrumentation and Controls isolation valve shall be designed to take the position that provides greater safety. (4) 15 Accident Analyses Other appropriate requirements to minimize the (5) 16 Technical Specificatiens p probability or consequences of en accidental rupture of these lines or of lines connected to 3.1.2.5.7 Criterion 56 - Primary Containment them shall be provided as necessary to assure Isolation adequate safety. Determination of the appropri-ateness of these requirements, such as higher 3.1.2.5.7.1 Criterion 56 Statement quality in design, fabrication, and testing, additional provisions for inservice inspection, Each line that connects directly to the con-protection against more severe natural phenomena, tainment atmosphere and penetrates primary reac-and additional isolation valves and containment, tor containment shall be prosided with contain-shall include consideration of the population ment isolation valves as follows, unless it can density, use characteristics, and physical be demonstrated that the containment isolation characteristicc of the site environs, provisions for a specific class of lines, such as instrument lines, are acceptable on some 3.1.2.5.6.2 Evaluation Against Criterion 55 other defined basis: The reactor coolant pressure boundary as (1) One locked closed isolation valve inside and defined in 10CFR50, Section 50.2, consists of the one locked closed isolation valve outside reactor pressure vessel, pressure retEining containment; or appurtenances attached to the vessel, valves, and pipes which extend from the reactor pressure (2) One automatic isolation valve inside and one vessel up to and including the outermost isola. locked-closed isolation valve outside tion valves. The lines of the reactor coolant containment; or pressure boundary which penetrate the contain-ment have suitable isolation valves capable of (3) One locked. closed isolation valve inside and isolating the containment, thereby precluding any 4 O Amendusent 1 3.1 37 i

ABM 23xaoarn Standard Plant RIN. A one automatic isolation valve outside 3.1.2.5.8.2 Evaluation Against Criterion 57 containment. A simple check valve may not be used as the automatic isolation valve outside containment; or Each line that penetrates containment and is not connected to the containment atmosphere and hl is not part of the rcactor coolant pressure (4) One automatic isolation valve inside and one boundary has at least one isolation valve automatic isolation valve outside contain- located outside containmect, ment. A simple check valve may not be used as the automatic isolation valve outside Details demonstrating conformance with Crite-containment. rion 57 are prosided in the following section: Isolation valves outside containment shall be Chapter / located as clow to the containment as practical Section Iith and upon loss of actuating power, automatic isolation valves shall be designed to take the (1) 6.2.4 Containment Isolation Systems position that provides greater safety. 3.1.2.6 Group VI Fuel and Reactivity Control 3.1.2,5.7.2 Evaluation Against Criterion 56 3.1.2.6.1 Criterion 60 Control of Releases of The manner in which the containment isolation Radioactive Af aterials to the Environmer.t system meets this requirement is detailed in the following sections: 3.1.2.6.1.1 Criterion 60 Statement Chapter / 1'he nuclear power unit design shallinclude Section Ilik means to control suitably the release of radio-active materials in gaseous and liquid effluents (1) 6.2.4 Containment Isolation Systems and to handle radioactive solid wastes produced (2) 7 Instrumentation and Controls during normal reactor operation including anticipated operational occurrences. Sufficient g holdup capacity shall be provided for retention (3) 15 Accident Analyses of gaseous and liquid effluents containing radioactive materials particularly where (4)16 Technical Specifications unfavorable site emironmental conditions can be expected to impose unusual operational 3.1.2.5.8 Criterion 57 Closed System limitations upon the release of such effluents isolation Yahes to the environment. I 3.1.2.5.8.1 Criterion 57 Statement 3.1.2.6.1.2 Evaluation Against Criterion 60 i l Each line that penetrates primary reactor 3.1.2.6.1.2.1 Emuent hf onitoring System l containment and is neither part of the reactor  ! coolant pressure boundary nor connected directly The Nuclear Island effluent monitoring j to the containment atmosphere shall have at least system is designed with provisions to prevent l one containment isolation valve which shall be radioactivity releases during accidents from either automatic, or locked closed, or capable of exceeding the limits of 10CFR100. remote manual operation. This valve shall be outside the containment and located as close to Ventilation air from the various Nuclear Is-the containment as practical. A simple check land areas is continuously monitored, and con-valve may not be used as the automatic isolation trolled ventilation air for containment areas is valve. exhausted through HEPA and charcoal filters if 9 Amendment 1 3.1 38

ABM 23 moore Standard Plant REV.A radioactive material release rate limits are ponents important to safety; reached. O In the event of an accident inside contain-(2) wiis protection; it sie siciai 8 rer ai tie-ment, noncondensible gases are contained within the leaktight containment vessel. (3) with appropriate containment, confinement, and filtering systems; The design of the effluent monitoring system meets the requirements of Criterion 60. (4) with a residual heat removal capability having reliability and testability that For further discussion, see the following reflects the importance to safety of decay sections: heat and other residual heat removal; and Chapter / (5) to prevent significant reduction in fuel Section Iith storage coolant inventory under accident conditions. (1) 1.2 General Plant Description 3.1.2.6.2.2 Evaluation Against Criterion 61 (2) 5.2.5 Detection of Leakage through Reactor Coolant Pressure 3.1.2.6.2.2.1 Fuel Storage and Handling System Boundary Fuel storage pools have adequate water (3) 6.2 Containment Systems shielding for stored spent fuel. Adequate shielding for transporting fuel is also (4) 11.5 Process and Effluent provided. Liquid level sensors are installed to Radiological Monitoring and detect low poolwater level. Buildings are Sampling Systems designed to meet Regulatory Guide 1.13 criteria. The fuel storage pools are designed (5) 15 Accident Analyses with no penetrations below the water level that is needed for niaintenance of adequate water 3.1.2.6.1.2.2 Radioacthe Waste System shielding and cooling. Also, check valves are used in pool circulation lines to prevent The design of the radioactive waste system siphoning. (RWS) is out of the Nuclear Island scope. The Nuclear Island / remainder of plant interfaces for New fuel storage racks are located in the the RWS are addressed in Subsection 11.1.8. concretc fuel storage vault. No cooling or air filtering system is required. These storage 3.1.2.6.2 Criterion 61 Fuet Storage and racks preclude accidental criticality (see llandling and Radioacthity Control evaluation against Criterion 62). The new fuel storage racks do not require any special 3.1.2.6.2.1 Criterion 61 Statement inspection and testing for nuclear safety purposes. The fuel storage and handling, radioac-tive waste, and other systems which may contain The fuel storage and handling system is de. radioactivity shall be designed to assure signed to assure adequate safety under normal adequate safety under normal and postulated and postulated accident conditions. The design accident conditions. These systems shall be of these systems meets the requirements of designed: Criterion 61. (1) with a capability to permit appropriate For further discussion, see the following periodic inspection and testing of com. sections:

  )

Amendment 1 3.1 39

ABWR mima Standard Plant REV.A Chapter / The spent fuel is stored under water in the Section Iitic spent fuel pool. A full array of loaded spent g fuel racks is designed to be suberitical, by at W (1) 5.4.7 Residual Heat Removal System least 5% ok. Neutron absorbing material, as an integral part of the design, is employed to (2) 6.2 Containment Systems assure that the calculated ke rr, including biases and uncertainties, will not exceed 0.95 (3) 9.1 Fuel Storage and Handling under all normal and abnormal conditions. The abnormal conditions accounted for are an (4) 12 Radiation Protection earthquake, accidental dropping of equipment, or impact caused by the horizontal movement of fuel 3.1.2.6.2.2.2 Radioactive Waste System handling equipment without first disengaging the fuel from the hoisting equipment. The design of the radioactive waste system (RWS) is out of the Nuclear Island scope. Tne Refueling interlocks include circuitry which Nuclear Island / remainder of plant interfaces for senses conditions of the refueling equipment and the RWS are addressed in Subsection 11.1.8. the control rods. These interlocks reinforce operational procedures that prohibit making the 3.1J.6.3 Criterion 62 Prevention or reactor critical. The fuel handling system is Criticality in Fuel Storage and ilandling designed to provide a safe, effective means of transporting and handling fuel and is designed 3.1.2.6.3.1 Criterion 62 Statement to minimize the possibility of mishandling or maloperation. Criticality in the fuel storage and handling system shall be prevented by physical systems or The prer.ence of fixed neutron absorbing mate-processes, preferably by use of geometrically rial in the new and spent fuel storage and the safe configurations. design of fuel handling systems precludes acci-3.1.2.6.3.2 Evaluation Against Criterion 62 dental criticality in accordance with Criterion 62. g Appropriate plant fuel handling and storage For further discussion, see the following facilities are provided to preclude accidental section: crit i nlits for new and spent fuel. Criticality in new and spent fuel storage is prevented by Chapter / presence of fixed neutron absorbing material. &ct. ion Lilt Fuel elements are limited by rack design to only top-loaded fuel assembly positions. The new and (1) 9.1 Fuel Storage and Handling spent fuel racks are Seismic Category I components. 3.1.2.6.4 Criterion 63 51onttoring Fuel and Waste Storage New fuelis placed in dry storage in the top-loaded new fuel storage vault. This vault con. 3.1.2.6.4.1 Criterion 63 Statement tains a drain to prevent the accumulation of water. Neutron absorbing material in the new Appropriate systems shall be provided in fuel fuel storage vault racks prevents an accidental storage and radioactive waste systems and critical array, even in the event the vault be- associated handling areas to: comes flooded or subjected to seismic loadings. (1) detect conditions that may result in loss of The center to-center new fuel assembly spacing residual heat removal capability and limits the k ert of the array to not more than excessive radiation levels, and 0.95 for new dry fuel. (2) initiate appropriate safety actions. O Amendment 1 3.1-40

ABM 234s 004e 1 Standard Plant REY.A 3.1.2.6.4.2 Evaluation Against Criterion 63 anticipated operational occurrences. The following releases are monitored: O 3.1.2.6.4.2.i ruei ster s>siem (1) gaseous releases, and Appropriate systems have been provided to meet the requirements of this criterion. A (2) liquid discharge. malfunction of the fuel pool cooling and cleanup system which could result in loss of residual In addition, the containment atmosphere is heat removal capability and excessive radiation monitored, levels is alarmed in the control room. Alarm conditions include low fuel pool cooliag water For further discussion of the same means and pump discharge pressure and high/ low levelin the equipment used for monitoring reactivity re-fuel storage pool and skimmer surge tanks. leases, see the following sections: System temperature is also continuously monitored and alarmed in the control room. The area Chapter / radiation monitors indicate radioactivity in this Section Titic area and initiate an alarm in the control room on abnormal radiation. (1) 5.2.5 Detection of Leakap through Reactor Coolant Pressure T!.e fuel storage system satisfies the Boundary requirements of Criterion 63. For further discussion, see the following sections: Chapter / Section Illh ,1 v (1) 9.1 Fuel Storage and Handling 3.1.2.6.4.2.2 Radioactlie Waste System The design of the radioactive waste system (RWS) is out of the Nuclear Island scope. The Nuclear Island / remainder of plant interfaces for the RWS are addressed in Subsection 11.1.8. 3.1.2.6.5 Criterion 64 Monitoring Radioactisity Releases 3.1.2.6.5.1 Criterion 64 Statement Means shall be provided for monitoring the reactor containment atrnospherc, spaces containing components for recirculation of loss.of coolant accident fluids, effluent discharge parts, and the plant environs for radioactivity that may be released from normal operations including anticipated operational occurrences and from postulated accidents. 4 3.1.2.6.5.2 Evaluation Against Criterion 64 Means have been provided for monitoring radioactivity releases resulting from normal and Amendment 1 3,1-41

MN 23A6100AE Standard Plant REV.A SECTION 3.2 O coursurs Settlon M g 3.2.1 Seismic Classincaden 3.21 3.2.2 Ouality Groun Classifications 3.2-1 3.23 Safety Classincationa 3.2-1 3.23.1 Safety Class 1 3.22 3.23.2 Safety Class 2 3.2-2 3.233 Safety Class 3 3.22 3.2.4 Correlation of Safety classes With Industry Codes 3.23 3.23 Non Safety Related Structures. Systems, and Comnonents 3.2-4 3.2.5.1 Definition of Non Nuclear Safety (NNS) Category 3.2-4 O 3.2.5.2 Design Requirements for NNS Structures, Systems V and Components 3.2-4 3.2.6 Oualliv Assurance 3.25 4 I O 3.2il 1 _ _ . - ,----.______y - _

l 4 ABWR meimn i Standard Plant REV,A SECTION 3.2 g TAllLES l Table Title Eage i l 1 3.21 Classification Summary 3.2-6 l 3.22 Minimum Design Requirements for an Assigned Safety Designation 3.2 7  ; l 3.23 Ouality Group Designations Codes and ) Industry Standards 3.2-8 l i

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l 9 1 1 l l l I I 1 3.2 iii 9

ABM 23461ooan Standard Plant nrn/a 3.2 CLASSIFICATION OF STRUCTURES, (as discusseo in Section 3.7) in combination (-) COMPONENTS, AND SYSTEMS with other appropriate loads. Nuclear Island structures, systems and The seismic classifications indicated in components are categorized as nuclear safety- Table 3.2-1 meet the requirements of Regulatory related or non nuclear safety-related (see Table Guide 1.29 except as otherwise noted in the 3.2-1). The safety-related structures, compo- table, nents, and systems perform nuclear safety-related functions as defined here, and are clas- 3.2.2 Quality Group Classifications sified in accordance with Subsection 3.2.3. In addition, specific design requirements are iden- Quality group classifications as defined in tified for the safety related equipment commen- NRC Regulatory Guide 1.26 are shown in Table surate with their safety classification (see 3.2-1 for all components under the heading, Table 3.2 2 and 3.2 3). 'Ouality Group Classification". Although not within the scope of Regulatory Guide 1.26 A safety related function is a direct or definitions, component supports, core support support function that is necessary to assure: structures and primary containment boundary that are within the scope of ASME Code, Section III,

1. the integrity of the reactor coolant are assigned p:r Tables 3.2-2 and 3.2-3, a pressure boundary; or quality group classification as identified in Table 3.2-1.

2. the capability to shut down the reactor and maintain it in a safe condition; or Quality group classifications and design and fabrication requirements, as indicated in Tables

3. the capability to prevent or mitigate the 3.2-1 and 3.2-3, respectively, meet the consequences of accidents which could result requiremento of Regulatory Guide 1.26. Figure in potential offsite exposures comparable to 6.2 38 depicts quality group classifications of the guidelines exposures of 10CFR100. the components in major systems.

3.2.1 Seismic Classification 3.2.3 Safety Classifications Nuclear Island structures, systems, and Safety related structures, systems, and components, including their foundations and components of the Nuclear Island are classified supports, that must remain functional in the for design requirements as Safety Class 1, event of a safe shutdown earthquake (SSE) are Safety riass 2, or Safety Class 3 in accordance designated as Seismic Category 1. with thir nuclear safety importance. These safety classifications are idedfied on Table All safety-related Nuclear Island structures, 3.21 for principal structures, systems, and components, equipment, and systems are classified components. Components within a system are as Seismic Category I, except those (e.g., pipe assigned different safety classes depending upon whip restraints), as noted on Table 3.2-1, which their differing safety importance; a system may need not remain functional in the event of an thus have components in more than one safety SSE. Also some non safety-related structures, class. Safety classification for supports systems, and components are classified as Seismic within the scope of ASME Code, Section ill, Category I as noted on Table 3.21. depends upon that of the supported component. The Sebmic Category I structures, systems and The definitions of the safety classes in this components are designed to withstand, without section are based on Section 3.3 of ANS Standard loss of function, the appropriate seismic loads 52.1 1983, and examples of their broad applica-O V Amendment 1 3.21

MM 23A6100AE Standard Plant RiiV A tion are given. Because of specific design (4) Ensure emergency core cooling where the considerations, these general definitions are equipment provides coolant directly to the & subject to ioterpretation and exceptions. Table rete (e.g., emergency core cooling systems); W 3.21 identifies component classifications on a component by component basis. (5) Provide or maintain sufficient reactor coolant inventory for emergency core cooling hiinimum design requirements for various (e.g., suppression pool). safety-related classes are delineated in Tables 3.2 2 and 3.2 3. Where possible, reference is Safety Class 2 includes the pressure-made to accepted industry codes and standards retaining portions of the following: which define design requirements commensurate with the safety-related function (s) to be (1) Those components of the control rod system performed in cases where industry codes and which are necessary for emergency negative standards have no specific design requirements, reactisity insertion; the sections that summarize the requirements to be implemented in the design are indicated. (2) Emergency core cooling systems; 3.2.3.1 Safety Class 1 (3) Pnmary containment vessel; Safety Class 1 (SC-1) applies to all (4) Post-accident containment heat removal components of the reactor coolant pressure systems; boundary (as defined in 10CFR50.2), and their supports, whose failure could cause a loss of (5) Pipes hasing a nominal pipe size of one inch reactor coolant at a rate in excess of the normal or smaller that are connected to the reactor makeup system, and which are within the scope of coolant pressure boundary, the AShf E Code, Section Ill. Safety Class 2 structures, systems, and Safety Class I components are identified in Table 3.2-1. components are identified in Table 3.2-1. 3.233 Safety Class 3 3.2.3.2 Safety Class 2 Safety Class 3, (SC-3) applies to those Safety Class 2 (SC-2) applies to pressure- structures, systems, and coraponents, not retaining portions, and their supports, of pri- included in SC-1 or -2, that are designed and nary containment and to other mechanical equip- relied upon to accomplish the following nuclear ment, requirements for which are with the scope safety related functions: of the AShiE Code, Section 111, ti at are not included in SC-1 and are designed and relied upon (1) Provide for functions defined in SC 2 by to accomplish the following nuclear safety- means of equipment, or portions thereof, related functions: that is not within tne scope of the ash 1E Code, Section Ill; (1) Provide primary containment radioactive material holdup or isolation; (2) Provide secondary containment radioactive material holdup, isolation, or heat removal; (2) Provide emergency heat removal for the primary containment atmosphere to an (3) Except for primary containment boundary btermediate heat sink, or emergency removal extension functions, ensure hydrogen of radioactive material from the primary concentration control of the pFmary containment atmosphere; containment atmosphere to acceptable limits; (3) Introduce emergency negativc reactivity to (4) Remove radioactive material from the atmos-make the reactor subcritical; phere of confined spaces outside primary O Amendment 1 3.22

MM 23A6100AE Standard Plant niu containment (e.g., control room or secondary (2) Electr. cal and instrumentation auxiliaries containment) containing SC 1, 2, or 3 necessary for operation of the d equipment; safety related systems and components; (5) Maintain geometry within the reactor to (3) Systems or components which restric: the ensure core reactivity control or core rate of insertion of positive reactivity; cooling capability; (4) Secondarycontainment; (6) Structurally bear the load or protect SC-1, 2, or 3 equipment in accordance with the (5) Service water systems required for the requirements; purpose of: (7) Provide radiation shielding for the control (a) Removal of heat from SC-1, SC-2 er SC 3 room or offsite personnel; equipment (8) Ensure nuclear safety-related functions (b) E nergency core cooling, provided by SC 1, -2, or -3 equipment (e.g., provide heat removal for SC-1, -2, or -3 (c) Post .'ccident heat removal from the heat exchangers, provide lubrication of SC-2 suppression pool, and or 3 pumps, provide fuel oil to the emergency diesel engine); (d) Providing cooling water needs for the functioning of emergency system; (9) Provide actu3 tion or motive power for SC-1,

            -2, or -3 equipment;                           (6) Initiating systems required to accomplish emergency core cooling, containment isola.

(10) Provide information or controls to ensure tion and other safety-related functions; cr pability for manual or autoretic actua-tion of nuclear safety-rele ] L required of SC 1, -2, or -3 , unctions (7) Spent fuel pool; ipment; (8) Fuel supply for the onsite emergency (11) Supply or process signals or supply power electrical system; required for SC-1, -2, or -3 equipment to perform their required nuclear (9) Emergency equipment area cooling; and safety-related functions; (10) Compressed gas or hydraulic systems (12) Provide a manual or automatic interlock required to provide control or operation function to ensure or maintain proper per- of safety related systems. formance of nuclear safety-related functions required of SC-1, -2, or 3 equipment; Safety Class 3 structures, systems and components of the Nuclear Island are identified (13) Provide acceptable environments for SC-1, in Table, .>. 1.

            -2, or 3 equipment and operating personnel; or                                             3.2.4 Correlation of Safety Classes with Industry Codes (14) Monitor plant variables that are identified requiring Category 1 e1eetrica1                   The design of plant safety related equipment instrumentation in Table 1 of Regulatory is commensurate with the safety importance of Guide 1.97.                                    the equipment. Hence, the various safety classes have a gradation of design Safety Class 3 includes the following:              requirements. The correlation of safety classes with other design requirements is summarized in (1) Reactor protection system;                        Tables 3.2-2 and 3.2 3.

O Amendrnent 1 3.23

1 ABWR maman Standard Plant suW A 3.2.5 Non Safety Related Structures, the health and safety of the public; Systems, and Components (9) Following a control room evacuation, g 3.2.5.1 Definition of Non Nuclear Safety provide an acceptable environment for (NNS) Category equipment required to achieve or maintain a safe shutdown condition; Structures, systems, and components that are not SC-1, -2, or 3, are non nuclear safety. (10) llandle spent fuel, the failure of which related (NNS) and are identified with *N' in the could result in fuel damage such that Safety Class column of Table 3.2-1. sigrificant quantities of radioactive material could be released from the fuel; Some NNS structures, systems and components have one or more selected but limited, require- (11) Ensure reactivity control of stored fuel; ments that are specified to ensure acceptable performance of specific NNS functions. The (12) Protect safety-related equipment necessary selected requirements are established on a to attain or maintain safe shutdown case by-case basis commensurate with the specific following fire; and NNS function performed (see Table 3.2 2). The functions performed by this subset of NNS (13) Monitor variables to: structures, systems, and components are: (a) Verify that plant operating conditions (1) Process, extract, encase, or store are within technical specification radioactive waste; limits (e.g., emergency core cooling water storage tank level, safety-(2) Ensure required cooling for the stored fuel related cooling water temperature); (e.g., spent fuel pool cooling system); (b) Indicate the status of protection (3) Provide cleanup of radioactive material from the reactor coolant system or the fuel system bypasses that are not automatically removed as a part of the h storage cooling system; protection system operation; (4) Monitor radioactive effluents to ensure Ibat (c) Indicate status of safety-related release rates or total releases are within equipment; or limits established for normal operations and transient events; (d) Aid in determining the cause or consequences of events for (5) Resist failure that could prevent any SC 1, post-aecider t invcstigation.

       -2, or 3 equipment from performing its nuclear safety related function (see Table   3.2.5.2 Design Requirements for NNS 3.2-2);                                       Structures, Systems and Components (6) Structurally bear the load or protect NNS         The design requirements for NNS equipment are equipment providing any of the functions      specified by the designer with appropriate listed in this Subsection 3.2.5.1;            consideration of the intended service of the equipment and expected plant and environmental (7) Provide permanent shielding for protection     conditions under which it will cperate.                  ,

of SC-1, -2, or -3 equipment or of onsite l personnel; Where appropriate, the Seismic Category 1,  ; ASME Code Section 111, and IEEE Class 1E (8) Provide operational, maintenance or requirements are specified on a limited basis, post accident recovery functions involving Generally, da. sign requirements are based on radioactive materials without undue risk to applicable indusuv codes and standards. Where O' Amendment 1 32-8

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

ABM 23A6100AE Standard Plant - - - nnv. A l l l l these are not available, accepted industry or I engineering practice is followed. ) 3.2.6 Quality Assurance Structures, systems, and components that perform nuclear safety related functions conform , to the quality assurance requirement of 10CFR50 ' Appendix B as shown in Table 3.21 under the  ! heading, 'Ouality Assurance Requirements," and in ) Table 3.2-2. Some NNS structures, systems, and components meet the same requirements as noted on Table 3.21. The Quality Assurance Program is I described in Chapter 17. ) i i l l l J O Amendment 1 3.25

1BWR ux6iooti: Standard Plant nitv. A TABLE 3.21 O CLASSIFICATION

SUMMARY

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l l

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I (TO BE PROVIDED BY JUNE 30,1988)

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l l i

~T (J

O . Amendment 1 3.2-6

ABWR ua6iman Standard Plant REV.A TABLE 3.2 2 MINIMUM DESIGN REQUIREMENTS FOR AN ASSIGNED SAFE'IT DESIGNATION h (TO BE PROVIDED HYJUNE 30,1988) l I l O 1 l l i 1 l l l l O Amendment 1 3.2-7 _ _ ~ . . . _ _ _ _ . _ _ _

ABWR 23461ooan standard Plant anv.A TABLE 3.2-3 O ouitirv caour ossioxirioxs . couss iso isuusrav srisoians l (TO BE PROVIDED BY JUNE 30,1988) l l l O l l O Amendment 1 3.2.g

i l -AB M M A6100AH Standard Plant any.A SECTION 3.3 O Cosreurs Section Illie Page 33.1 Wind leadinns 33-1 33.1.1 Design Wind Velocity 33 1 33.1.2 Determination of Applied Forces 33-1 33.2 Tornado Loadinns 33 1 33.2.1 Applicable Design Parameters 33-1 33.2.2 Determination of Forces on Structures 33-1 33.23 Effect of Failure of Structures or Components Not Designed for Tornado Loads 33-2 333 Interfaces 33-2 333.1 Site Specific Design Basis Wind 33 2 333.2 Site-Specific Design Basis Tornado 33-2 0 3333 Remainder of Plant Structures, Systems and Components Not Designed for Tornado Loads 33-2 33.4 References 33-2 0 33-il Arn<ndment 1

ABWR ua6imae t Slandard Plant REva 1 SECTION 33 $ TABLES Table Title Page 33-1 Importance Factor (1) for Wind Loads 33-3 33-2 Velocity Pressure Distribution and Gust Factors at Various lleights 33-4 l t O l l l l i O 33111 Amendment I

ABMR u^6ioo^u Standard Plant anv ^ 3.3 WIND AND TORNADO LOADINGS Reference 1. Reference 2 is used to obtain the p effective wind pressures for cases which Nuclear Island structures which are Seismic Reference 1 does not cover. Since the Seismic Category I are designed for tornado and extreme Category I structures are not slender or wind phenomena, flexible, vortex-shedding analysis is not required and the above wind loading is applied 3.3.1 Wind Loadings as a static load. 33.1.1 Design Wind Velocity 3.3.2 Tornado Imadings Seismic Category I structures are designed to 33.2.1 Applicable Design Parameters withstand a design wind velocity of 130 mph at an elevation of 33 feet above grade with a The design basis tornado is described by the recurrence interval of 100 years, following parameters: 33.1.2 Determination of Applied Forces (1) A maximum tornado wind speed of 260 mph at a radius of 453 feet from the center of the The design wind velocity is converted to tornado; velocity pressure in accordance with Reference 1 using the formula: (2) A maximum translational velocity of 57 mph; qz = 0.00256 K z (IV)2 (3) A maximum tangential velocity of 203 mph, calculated as defined in Section 3.3 of where K z =the velocity pressure exposure Reference 3, based on the translatio'al coefficient which depends upon the velocity of 57 mph; type of exposure and height (z) , above ground per Table 6 of (4) A maximum atmospheric pressure drop of 1.46 R efe rence 1. psi with a rate of the pressure change of 0.27 psi per second in accordance with I = the importance factor which depends Reference 3; and on the type of exposure; appropriate values of I are listed in Table (5) The spectrum of tornado-generated missiles 3.31, and their pertinent characteristics as given in Subsection 3.5.1.4. V = design wind velocity of 130 mph, and 33.2.2 Determination of Forces un Structures q2 = velocity pressure in psf The procedures of transforming the tornado The velocity pressure (qz) distribution with loading into effective loads and the distri-height for exposure types C and D of Reference 1 bution across the structures are in accordance are given in Table 33 2. with Reference 4. The procedure for trans-forming the tornado generated missile impact The design wind pressures and forces for into an effective or equivalent static load on buildings, components and cladding, and other structures is given in Subsection 3.5.3.1. The i structures at various heights above the ground loading combinations of the individual tornado I are obtained, in accordance with Table 4 of loading components and the load factors are in I Reference 1 by multiplying the velocity pressure accordance with Reference 4. by the appropriate pressure coefficients and gust factors. Gust factors are in accordance with The reactor building and control building are Table 8 of Reference 1. Appropriate pressure not vented structores. The exposed exterior coefficients are in accordance with Figures 2, roofs and walls of these structures are designed 3a, 3b, 4, and Tables 9 and 11 through 16 of for the 1.46 psi pressure drop. Tornado dampers O Amendment i 33-1 l

1

                                                                                                  )
                                                                                      " 3"^U      i Standard Plant                                                                           Rl!V. A   i I

are provided on all air intake and exhaust Standards Committee Working Group ANS- 23, openings. These dampers are designed to withstand a negative 1.46 psi pressure. American Nuclear Society.

4. Bechtel Topical Report BC-TOP-3-A, Revision 33.23 Effect of Vallure of Structures or 3, Tornado and Extreme Wind Design Criteria Components Not Designed for Tornado Loads for Nuclear Power Plants.

All Nuclear Island structures are designed for tornado loads. 3.3.3 Interfaces 3.33.1 Site-Specific Design Basis Wind The site-specific design basis wind shall not exceed the design basis wind given in subsection 33.1. 3.33.2 Site-Specific Design Basis Tornado The site specific design basis tornado shall not exceed the design basis tornado given in subsection 3.3.2. 3.33.3 Remainder of Plant Structures, Systems, and Components Not Designed for Tornado Loads All remainder of plant structures, systems, and components not designed for tornado loads shall be analyzed for the site specific loadings to ensure that their mode of failure will not affect the ability of the Seismic Category 1 Nuclear Island structures, systems, and components to perform their intended safety functions. , 3.3.4 References

1. ANSI Standard A58.1, Minimum Design Loads for Buildings and Other Structures, Committee A. 58.1, American National Standards Institute.

2. ASCE Paper No. 3269, Wind Forces on Structures, Transactions of the American Society of Civil Engineers, Vol.126, Part II.

3. ANSI /ANS 2.3, American National Standard, Estimating Tornado and Extreme Wind Characteristics at Nuclear Power Sites, O

Amendment 1 33-2

___m-___ _ .._ ___._-._..-. _ _ . _ _ _ _ _ . _ _ .. _--_ __ . . ____ . . . ._ _ -.__ 1 ABM 23A6100AE Standard Plant REV.A O Table 3.31 Importance Factor (I) for Wind IAads EXPOSURE C EXPOSURE D 1.00 1.04 .

                                                                                                                                              'l i

l 1 i i l NOTES: l

1. These values of (I) are based on Table 5 of ,

Reference 1 but are modified to reflect the ' 100 year return period of the design wind O velocity versus the 50 year return period basis 'of Reference 1. 1

2. Exposure categories are as defined in l Section 6.5.3 of Reference 1.

I 1 O Amendment 1 3.33 1 w,,-w, - -*,v. ~ .,-- . ,w,-,w--,,,w.e., v, w

ABWR uuiman Standard Plant REV.A Table 3.3 2 VELOCITY PRESSURE DISTRIBUTION AND GUST FACTORS AT VARIOUS HEIGHTS EXPOSURE TYPE C Height qz Gh Windward Side Wall Roof Leeward Zone (ft/sec) Wall Pressure Suction Suction Wall Suction z (ft) 0.8G qz 0.7G gh 0.7G gh 0.5G gh 4 (Ib/ft') (Ib/ft ) (Ib/ft ) (Ib/ft ) 0-15 34.6 1.32 37 58 58 42 20 37.6 1.29 39 57 57 41 25 40.2 1.27 41 56 56 40 30 42.4 1.26 43 56 56 40 40 45.9 1.23 45 54 54 39 50 48.9 1.21 47 54 54 38 60 51.5 1.20 49 53 53 38 70 53.6 1.19 51 53 53 38 80 55.8 1.18 53 52 52 37 90 58.0 1.17 54 52 52 37 100 59.7 1.16 55 51 51 37 120 62.7 1.15 58 51 51 36 140 65.8 1.14 60 50 50 36 160 68.4 1.13 62 50 50 36 EXPOSURE TYPE D O Height qz Ga Windward Side Wall Roof Leeward Zone (ft/sec) Wall Pressure Suction Suction Wall Suction z (ft) 0.8G qz 0.7G gh 0.5G gh 0.7Ggqh (lb/ft ) (Ib/ft ) (Ib/ft ) (lb/ft ) , l 0-15 56.2 1.15 52 58 58 42 20 59.4 1.14 54 58 58 41 25 61.8 1.13 56 57 57 41 30 M.1 1.12 57 57 57 41 40 68 3 1.11 61 56 56 40 50 71.1 1.10 63 56 56 40 60 73.9 1.09 M 55 55 39 70 76 3 1.08 66 55 55 39 80 78.1 1.08 67 55 55 39 90 80.0 1.07 68 54 54 39 100 81.9 1.07 70 54 54 39 i 120 84.7 1.06 72 54 54 38 1 140 87.5 1.05 74 53 53 38 160 89.8 1.05 75 53 53 38 gh

• 9z @ z = 123.69 ft (roof height above grade) for the reactor building O

Amendment 1 3.34

23A6100AE Standard Plant REV.A 3.4 WATER LEVEL (FLOOD) DESIGN O (To be provided by December 31,1988) O I O Amendment 1 3.43 l

ABM 2n63oore Standard Plant REV.A SECTION 3.5 CONTENTS Section Illh Page 3.5.1 Missile Selection and Descriotion 3.5-1 3.5.1.1 Internally Generated Missiles (Outside Containment) 3.5-3 3.5.1.1.1 Rotating Equipmeot 3.5-3 3.5.1.1.1.1 Missile Characterization 3.53 3.5.1.1.1.2 RCIC Steam Turbine 3.5-3 3.5.1.1.13 Main Steam Turbine 3.5-3 3.5.1.1.1.4 Other Missue Analysis , 3.53 3.5.1.1.2 Pressurized Components 3.53 3.5.1.1.2.1 Missile Characterization 3.5-3 4 3.5.1.1.2.2 Missile Analyses 3.5-3 3.5.1.13 Missile Barriers and Loadings 3.5-5 3.5.1.2 Internally Generated Missiles (Inside Containment) 3.5-6 3.5.1.2.1 Rotating Equipment 3.5-5 i 3.5.1.2.2 Pressurized Components 3.5-6 j 3.5.1.23 Missile Barriers and Loadings 3.5-6 l l 3.5.1.2.4 Evaluation of Potential Gravitational l Missiles inside Containment 3.5-6 l 3.5.13 Turbine Missiles 3.5-6 3.5.1.4 Missiles Generated by Natural Phenomena 3.5-6 3.5.1.5 Site Proximity Missiles (Except Aircraft) 3.5-7 3.5.1.6 Aircraft Hazards 3.5-7 O 3.5-il I Amendment 1 l

l I MM 23A6100AE Standard Plant REV.A SECTION 3.5 9; CONTENTS (Continucd) Section Tills Eage 3.5.2 Stnictures. Systems. and Components to be Protected from Externally Generated Missiles 3.5-7 l 3.53 Barrier Deslan Procedures 3.5-7 3.53.1 Local Damage Prediction 3.5-7 l 1 3.53.1.1 Concrete Structures and Barriers 3.5-7 3.53.1.2 Steel Structures and Barriers 3.5-7 3.53.2 Overall Damage Prediction 3.57  ; 3.5.4 Interfaces 3.5-% 3.5.4.1 Protection Against Turbine Missiles 3.5-8 j 1 3.5.4.2 Missile Generated by Natural Phenomena from Remainder of Plant Structures, Systems and Components 3.5 8 - 3.5.43 Other Missiles 3.5-8 3.5.5 References 3.5-8 ILLUSTRATIONS I Figure Title Ears i 1 1 3.5-1 Missile Velocity and Displacement Characteristics Resulting from Saturated Steam and Water Blowdowns (1050 psia Stagnation Pressure) 3.5-9 9 3.5 iii Amendment 1

ABM 2346iooan Standard Plant ni"L.a 3.5 MISSILE PROTECTION intent of 10CFR50 Appendix A, General Design O The missile protection design basis for Criteria for Nuclear Power Plants. Seismic Category I structures, systems and Potential missiles that have been identified components is described in this section. A are listed and discussed in later subsections. tabulation of safety related structures, systems, and components (both inside and outside After a potential missile has been containment), their location, seismic category, identified, its statistical significance is and quality group classification is given in determined. A statistically significant missile Table 3.2-1. General arrangement drawings is defined as a missile which could cause showing locations of the structures, systems, and unacceptable plant consequences or violation of components are presented in Section 1.2. the guidelines of 10CFR100. Missiles considered are those that could The examination of potential missiles and result from a plant related failure or incident their consequences is done in the following including failures within and outside of manner to determine statistically significant containment, environmental generated missiles and missiles: site proximity missiles. The structures, shields, and barriers that have been designed to (1) If the probability of occurrence of the withstand minile effects, the possible missile missile (P1 ) is determined to be less than loadings, and the procedures to which each 10-7 per year, the missile is dismissed barrier has been designed to resist missile from further consideration because it is impact are described in detail. considered not to be statistically significant. 3.5.1 Missile Selection and Description (2) If (P 1) is found to be greater than 10 7 per year, it is examined for its probability Cor,ponents and equipment are designed to have of impacting a design target (P2)- a low potential for generation of missiles as a basic safety precaution. In general, the design (3) If the product of (P1 ) and (P2) is less ) that results in reduction of missile generation than 10 7 per year, the missile is i potential promotes the long life and usability of dismissed from further consideration. a component and is well within permissible limits of accepted codes and standards. (4) If t h e p r o d u c t of (P ) and (P 2) IS 1 greater than 10-7 per year, the missile is Seismic Category I structures have been examined for its damage probability (P3 ). analyzed and designed to be protected against a If the combined probability (i.e., P,ix wide spectrum of missiles. For example, failure P2xP 3 = P4) is less than 10-' per of certain rotating or pressurized components of. year, the missile is dismissed, equipment is considered to be of sufficiently high probability and to presumably lead to (5) Finally, measures are taken to design generation of missiles. However, the generation acceptable protection against missiles with of missiles from other equipment is considered to (P4) greater than 10 7 per year to be of low enough probability and is dismissed reduce (P 1), (P 2 ), and/or (P 3), so from further consideration. Tornado generated that (P 4) is less than 10 7 per year. missiles and missiles resulting from activities particular to the site are also discussed in this Protection of essential structures, systems section. The missile protection criteria to and components is afforded by one or more of the which the plant has been analyzed comply with the following practices: O Amendment 1 3.51

MM 33A6100AE Standard Plant REV.A (1) Location of the system or component in an (6) Ultimate heat sink; individual missile proof structure; (7) Automatic depressurization system relief h (2) Physical separation of redundant systems or valves; components of the system for the missile trajectory path or calculated range; (8) Standby dieselgenerator system; (3) Provision of localized protection shields or (9) CRD seram system (hydraulic and barriers for systems or components; electrical); (4) Design of the particular strueture or (10) Fuel pool cooling and cleanup system; component to withstand the impact of the most damaging missile; (11) Remote shutdown panel; (5) Provision of design features on the (12) Reactor protection system; potential missile source to prevent missile generation; and/or (13) All containment isolation valves; (6) Orientation of the potential missile source (14) IIVAC systems required during operation of to prevent unacceptable consequences due to items (1) through (13); and missile generation. (15) Electrical and control systems and wiring The following criteria have been adopted in required for ooeration of items (1) through provide an acceptable design basis for the (14). plant's capability to withstand the statistically significant missiles postulated inside the The following general criteria are used in , reactor building. the design, manufacture, and inspection of 3 , equipment: W l (1) No loss of containrr.ent function as a result I of missiles generated internal to (1) All pressurized equipment and sections of ' containment. piping that may periodically become isolated under pressure are provided with  ! (2) Reasonable assurance that a safe plant shut- pressure relief vr.lves acceptable under l down condition can be achieved and ASME Code Section Ill. The valves ensure ) maintained. that no pressure buildup in equipment or ' piping sections exceeds the design limits (3) Offsite exposure within the 10CFR100 of the materials involved, guidelines for those potential missile damage events resulting in radiation (2) Components and equipment of the various activity release, systems are designed and built to the stan-dards established by the ASME Code or other  ; The systems requiring protection are: equivalent industrial standard. A strin- ) gent quality control program is also en-(1) Reactor coolant pressure boundary; forced during manufacture, testing, and in-stallation. (2) Residual heat removal system; (3) Volumetric and ultrasonic testing where re. (3) liigh pressure core flooder system; quired by code coupled with periodic inser-vice inspections of materials used in com-(4) Reactor core isolation cooling system; ponents and equipment add further assurance that any material flaws that could permit (5) Essential senice water system; the generation of missiles are detected. O Amendment 1 3.52

I MM 23A6100All l Standard Plant REV.A i 3.5.1.1 laternally Generated Missiles (Outside (2) Sufficient physical separation (barriers hq Containment) and/or distance) of safety.related and redundant equipment exists so that the These missiles are considered to be those combined probability (P 1 x P2 ) is less missiles resulting internally from plant than 10-7 per year, equipment failures within the Nuclear Island but outside containment. These conclusions are arrived at by noting that pumps, fans, and the like have synchronous 3.5.1.1.1 Rotating Equipment motors and thus cannot achieve an overspeed condition. At rated speed, if a piece such as a 3.5.1.1.1 Missile Characterization fan blade breaks off, it will not penetrate the casing. As an example, a containment high purge Equipment within the general categories of exhaust fan has been analyzed for a thrown blade pumps, fans, blowers, diesel generators, compres- at rated speed conditions using an analytical sors, and turbines and, in particular, components expression from Reference 1. It is determined, in systems normally functioning during power re- based on maximum thickness this blade could actor operation, has been examined for any possi- penetrate, that the blade would not escape the ble source of credible and significant missiles. fan casing and consequently (P1 ) is less than 10-7 per year. 3.5.1.1.1.2 RCIC Steam Turbine 3.5.1.1.2 Pressurized Components The RCIC steam turbine driving the pump is not a credible source of missiles. It is provided 3.5.1.1.2.1 Missile Characterization with mechanical overspeed protection as well as automatic governing; very extensive industrial Potential missiles which could result from and nuclear experience with this model of turbine the failure of pressurized components are ' has never resulted in a missile which penetrated analyzed in this subsection. These potential the turbine casing. missiles may be categorized as contained fluid-energy missiles or stored strain-energy 3.5.1.1.1.3 Main Steam Turbine (clastic) missiles. These potential missiles have been conservatively evaluated against the The onh identifiable source of credible and design criteria in Subsection 3.5.1. significant, internally-generated missiles from plant equipment failures outside containment is Examples of potential contained fluid.cnergy due to a possible gross structural failure of the missiles are valse bonnets, valve stems, and main steam turbine. Protection of safety related retaining bolts. Valve bonnets are considered systems from potential missiles generated by this jet propelled missiles and have been analyzed as source is discussed in Subsection 3.5.1.3. such. Valve stems have been analyzed as piston type missiles, while retaining bolts are 3.5.1.1.1.4 Other Missile Analysis examples of stored strain energy missiles. No remaining credible missiles meet the 3.5.1.1.2.2 Missile Analyses significance criteria of having a probability

  * (P )4 greater than 10-7 per year for rotating              Pressurized components outside the contain-or pressurized equipment, because either:              ment capable of producing missiles have been reviewed. Although piping failures could re, ult (1) The equipment design and manufacturing in significant dynamic effects if permitted to criteria mentioned previously result in whip, they do not form missiles as such because               1 (P 1) being less than 10-7 per year; or          the whipping section remains attached to the O

Amendment 1 3.M

MM 33A6100AE Standard Plant RE5!. A remainder of the pipe. Since Section 3.6 on this, the probability of a valve bonnet addresses the dynamic effects associated with missile striking both Division 1 and 2 vital & W pipe breaks, pipes are not included here as targets for safe shutdown is extremely low potential internal missiles, making the resultant probability much less than 10-7 per year. All pressurized equipment and sections of pip-ing that may periodically become isolated under (2) Valve Sten,s All the isolation valves pressure are provided with pressure relief valves installed in the reactor coolant systems acceptable under the AShfE Code, Section III. have stems with a back seat which eliminates the possibility of ejecting valve stems even The only remaining pressurized components if the stem threads fail. Since a double considered to be potentially capable of producing failure of highly reliable components would missiles are: be required to produce a valve stem missile, the overall probability of occurrence is (1) valve bonnets (large and small); less than 10 7 per year. IIence valve stems can be dismissed as a source of (2) valve stems; missiles. (3) pressure vessels; (3) Pressure Vessels hioderate energy vessels less than 275 psig are not credible missile (4) thermowells; sources. The pneumatic system air bottles are designed for 2500 psig to the AShf E Code, (5) retaining bolts; and Section III requirements. The bottles are not considered a credible source of missiles (6) blowout panels. for the following reasons. These are analyzed as follows: (a) The bottles are fabricated from (1) Valve Honnets - Valves of ANSI rating 900 heavy wall rolled steel; h psig and above and constructed in accordance (b) The operating orientation is vertical with the AShf E Code, Section 111 are with the ends facing concrete slabs. pressure scal bonnet-type valves. hiost The bottles are topped with steel covers valves of ANSI rating 600 psig and below are thick enough to preclude penetration by valves with bolted bonacts. Valve bonnets a missile. are prevented from becoming missiles by limiting stresses in the bolting to those (c) The fill connection is protected by a defined by the AShiE Code and by designing permanent steel collar, flanges in accordance with applicable code requirements. (d) The bottles are strapped in a rack to prevent them from toppling over. The Failure rates (P1) for salve bonnets are rack is seismically designed to the AShf E in the range of 10-4 to 10 5 per year. Code, Section 111, Subsection NF The target strike probability (P 2) has not requirements. been calculated. Rather, the physical separation requirements incorporated for Considering this, the bottles, as an safety related divisional equipment makes assembly with rack and cover, are incapable the design acceptable. All safe shutdown of producing jet propelled missiles with a functions in the Nuclear Island design have frequency greater than 10 7 per year. redundant backups and these redundant items are separated either by considerable (4) Thermowells Thermowells are welded to distance or a missile proof barrier. Based sockolet connections which in turn are O Arnendment 1 3.54

ABM 2346 moan l Standard Plant anv. A l l welded to the wall of the pipe. An analysis 73 j p V of a postulated failure of tuis weld has (Uj l been performed. The following expression (D/p = equivalent loss coefficient relates the missile displacement and between the. broken velocity following the postulated failure: pressurized component and fluid reservoir, dimension-Y = v. fn / 1 j.~V less; (W/A) \1 V/um iu. where AE = area of break, ft2; and y = distance traveled by the missile Ap = area of pressurized from the break (ft) component between break and fluid r e se rvoir, ft 2 W = missile weight (Ib) (assumesAp 2., A E)- A = 2 As illustrated in Figure 3.51, the effect frontal ares of missile (ft ) of friction on the velocity displacement

u. = asymptotic velocity of jet relationship is reasonably small. It can be (ft/sec) conservatively assumed that the most extreme friction condition persists with f' = 100

v. = asymptotic specific volume of for the case of saturated water blowdown and jet (ft 3/lb) f' = 0 for the case of saturated steam blowdown.

V = velocity of missile (ft/sec) A typical thermowell weights about 2 lb. . Inherently, the water and steam velocities Based on ejection by steam at 1050 psig, the h, are equal (i.e., a unity velocity ratio) in ejection velocity could reach 200 ft/ sec a saturated water blowdown. The jet asymp- which is not sufficient to inflict totic velocity (u.) and the jet asympto- significant damage to critical systems, tic specific volume are determined by the (P4) is therefore less than 10-7 per methods described by Reference 2. The year. corresponding velocity-displacement rela-tionships for missiles resulting from satu- (5) Retaining Bolts Nuts, bolts, nut and bolt rated water and saturated steam blowdowns combinations, and nut and stud combinations are presented in Figure 3.5-1. The ordinate have only a smcIl amount of stored energy is the missile velocity, V, and the abscissa and thus are of no concern as potential is the displacernent parameter, Y', given by missiles. Y' = Y (6) Blowout Panels - Blowout panels are hinged (W/A) to prevent them from becoming missiles. Guard rails for personnel protection have Included in Figure 3.5-1 is the influence of been provided where required by the swing

,               different values of the friction parameter,           pattern. Thus by design, (P 2) is less f*, defined by                                        than 10-7 per year.                                  1 3.5.1.1.3 Missile Barriers and Loadings f* ,                        2

[U} (Ap/ (Djp Certain cases of rotating and pressurized components generating missiles described in where Subsection 3.5.1.1.2 give credit for potential O  ; I l Arrendment 1 33 5 i l

ABM asA6ioo^n Slandard Plant anv. A missile-consequence mitigation by structural 3.5.1.2.4 Evaluation of Potential walls and slabs. These walls and slabs are Gravitational h11ssiles Inside Containment designed to withstand internal missile effects; the applicable seismic category and quality group Gravitational missiles inside the containment h classification are listed in Section 3.2. have been considered as follows: Penetration of the structural walls by internally generated missiles is not considered credible. Seismic Category I systems, components, and structures are not potential gravitational 3.5.1.2 Internally Generated hiissiles (Inside missite sources. Containment) Non-Seismic Category I items and systems Internal missiles are those resulting from inside containment are considered as follows: plant equipment failures within the contain-ment. Potential missile r.ources from both (1) CableTray rotating equipment and pressurized components are considered. All cable trays for both Class 1E and non. Class 1E circuits are seismically supported 3.5.1.2.1 Rotating Equipment whether or not a hazard potential is evident. By an analysis similar to that in Subsection 3.5.1.1.1, i t i s c o n c l u d e d t h a t n o i t e m s o f (2) Conduit and Non Safety Pipe rotating equipment inside the containment have the capability of becoming potential missiles. Non-Class 1E conduit is seismically All reactor internal pumps are incapable of supported if it is identified as a potential achicving an overspeed condition and the motors hazard to safety-related equipment. All and impellers are incapable of escaping the Nuclear Island non safety related piping casing and the reactor vessel wall, respectively, that is identified as a potential hazard is seismically analyzed per Subsection g 3.5.1.2.2 Pressurized Components 3.7.3.13. W Identification of potential rnissiles and their (3) Equipment for hiaintenance consequences outside containment are specified in Subsection 3.5.1.1.2. The same conclusions are All other equipment, such as hoists, that is drawn for pressurized components inside of required during maintenance will either be containment. For example, the ADS accumulators removed during operation, moved to a are designed to the AShf E Code, Section III, location where it is not a potential hazard requirements and are therefore not considered a to safety related equipment, or seismically credible missile source. One additional item is restrained to prevent it from becoming a fine motion control rod drives (FhfCRD) under the missile. reactor vessel. The Fhi' RD mechanisms are not credible missiles. The PhiCRD housings are 3.5.1.3 Turbine 511ssiles designed (Section 4.6) to prevent any significant nuclear transient in the event of a drive housing See Subsection 3.5.4. break. 3.5.1.4 511ssiles Generated by Natural 3.5.1.2.3 bilssile Barriers and Loadings Phenomena Credit is taken in some cases of rotating and Tornado generated missiles have been deter-  ; pressurized componente generating missiles for mined to be the limiting natural phenomena ' missile consequence mitigation by structural hazard in the design of all structures required walls and slabs. Penetration for the containment for safe shutdown of the nuclear power plant. walls, floors and slabs by potential inissiles is Since tornado missiles are used in the design , not considered credible. basis, it is not necessary to consider missiles I O Amendment 1 316

ABM ux61ooan Standard Plant REV.A generated from other natural phenomena. The 3.53.1 Local Damage Predletion p missiles considered in the design are per V Reference 8. The prediction of local damage in the impact area depends on the basic material of construc-3.5.1.5 Site Peoximity Missiles Except tion of the structure or barrier (i.e., concrete Aircraft or steel). The corresponding procedures are presented separately. Composite barriers re External missiles other than those generated not utilized in the Nuclear Island for missile by tornados are not considered as a design basis protection. (i.e.1 10-7 per year). 3.53.1.1 Concrete Structures and Barriers 3.5.1.6 Aircraft Hazards The modified Petry formula (Reference 3) is Aircraft hazards are not a des applied analytically for missile penetration in for the Nuclear Island (i.e.110 perjiyear). n basis event concrete. To prevent perforation, a minimurr concrete thickness of 2.2 times the penetratio 3.5.2 Structures, Systems, and thickness determined for an infinitely thie Components to be Protected from concrete slab is employed. In the event that Externally Generated Missiles spalling or scabbing is unacceptable, a minimum concrete thickness of 3 times the penetration The sources of external missiles which could thickness determined for an infinitely thick affect the safety of the plant are identified in concrete slab is provided. These design Subsection 3.5.1. Certain items in the Nuclear procedures have been substantiated by full scale Island are required to safely shut down the impact :ests in which reinforced concrete panels reactor and maintain it in a safe condition (12 to 24 inches thick, 3000-psi design assuming an additional single fallare. These strength) were impacted by poles, pipes, and items, whether they be structures, systems, or rods simulating tornado borne debris (Reference h' components, must therefore all be protected from externally generated missiles. 4). 3.53.1.2 Steel Structures and Harriers These items are the safety-related items listed in Table 3.2-1. Appropriate safety The Stanford equation (Reference 5) is classes and equipment locations are given in this applied for steel structures and barriers. table. All of the safety related systems listed I are located in buildings which are designed as 3.53.2 Overall Damage Prediction tornado resistant. Since the tornado missiles are the design basis missiles, the systems, The overall response of a structure or I structures, and components listed are considered barrier to missile impac; depends largely upon to be adequately protected. Provisions are made the location of impact (e.g., near mid span or ] to protect the charcoal delay tanks against near a support), dynamic properties of the i tornado missiles. structure / barrier and missile, and on the ki-netic energy of the missile. In general, it has , 3.5.3 Barrier Design Procedures been assumed that the impact is plastic with all l of the initial momentum of the missile t ans-The procedures by which structures and ferred to the structure or barrier and only a l barriers are designed to resist the missiles portion of the kinetic energy absorbed as strain described in Subsection 3.5.1 are presented in energy within the structure or barrier. this section. The following procedures are in accordance with Section 3.5.3 of NUREG 0800 After demonstrating that the missile does not (Standard Review Plan), perforate the structure or barrier, an equi-O Amendment 1 3.5-7

ABWR 33A6100AE Standard Plant ansf. A valent static load concentrated at the impact 3. A. A mirik a n, Design of Protective g area is determined. The structural response to Structures, Bureau of Yards and Docks, W this load, in conjunction with other appropriate Publication No. NAVDOCKS P-51, Department design loads, is evaluated using an analysis of the Navy, Washington, D.C., August 1960. procedure similar to that in Rr'"ence 6 for rigid missiles, and the procedure Reference 7 4. A. E. S t e p h e a s o n, F u11 S c aie for deformable missiles. Tornado Missile Impact Tests, EPRI NP-440, July 1977, prepared for Electric Power 3.5.4 Interfaces Research Institute by Sandia Laboratories. 3.5.4.1 Protection Against Turbine Missiles 5. W. B. Cottrell and A. W. Savolainen, U. S. Reactor Containment Te c h n ology, Protection of the Nuclear Island shall be ORNL-NSIC-5, Vol.1, chapter 6, Oak Ridge prosided for by: (1) placement and orientation of National Laboratory. the turbine generator in accordance with the guidelines presented in Regulatory Guide 1.115, 6. R. A. Williamson and R. R. Alvy, Impact or (2) providing local structures that prevent Effect of Fragments Striking Structural the turbine missiles from impacting on Nuclear Elements, liolmes and Narver, Inc., Revised Island buildings or that limit their penetrating November 1973. capabilities to less than the capability of the structures. 7. J. D. Riera, On the Stress Analysis of Structures Subjected to Aircraft impact 3.5.4.2 Missiles Generated by Natural Phenomena Forces, Nuclear Engineering and Design, from Remainder of Plant Structures, Systems and North liolland Publishing Co., Vol. 8,1968. Components

8. ANS 2.3.

The remainder of plant structures, systems, and components shall be analytically checked to h ensure that during a site specific tornado they will not generate mksiles exceeding the missile = considered under Subsection 3.5.1.4 3.5.4.3 Other Missiles Analyses shall be provided that demonstrate that the probability of site proximity missiles (including aircraft) impacting the Nuclear Island and causing consequences greater than 10CFR Part 100 exposure guidelines is 5.10-7 per year. 3.5.5 References

1. C. V. Moore, The Design of Barricades for Hazardous Pressure Systems, Nuc' ear 1 Engineering and Design, Vol. 5.1967. j

2. F. J. Moody, Prediction of Blowdown Thrust l and let Forces, ASME Publication 6911T-31, August 1969.

l 9l Amendment 1 33-8

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 ,               Figure 3.5-1 MISSILE VELOCITY ANO DISPLACEMENT CHARACTERISTICS RESULTING FROM                                                                             h y                            SATURATED STEAM AND WATER BLOWDOWNS (1050 psia STAGN ATION PRESSURE)                                                                         >m

ABM 23461oote Standard Plant anv. A SECTION 3.6 O cosresrs Section Title Eagt 3.6.1 Postulated Ploinn Failures in Fluid Systems Inside and Outside of Containment 3.6-1 3.6.1.1 Design Bases 3.6-1 3.6.1.1.1 Criteria 3.6-1 3.6.1.1.2 Objectives 3.6-2 3.6.1.13 Assumptions 3.6-2 3.6.1.1.4 Approach 3.6-3 3.6.1.2 Description 3.6 3 3.6.13 Safety Evalutdon 3.6-3 3.6.13.1 General 3.6-3 3.6.13.2 Protectico Methods 3.6-4 3.6.13.2.1 Geoeral 3.6-4 3.6.13.2.2 Separation 3.6-4 3.6.13.23 Barriers, Shields, and Enclosures 3.6-5 3.6.13.2.4 Piping Restraints 3.65 l 3.6.133 Specific Protection Measures 3.6-5 l 3.6.2 Determination of Break locations and Ihnamic Effects Associated with the i Postulated Ruoture of Ploine 3.6-6 l 3.6.2.1 Criteria Used to Define Break and Crack Location and Configuration 3.6-6 ) l 3.6.2.1.1 Definition of High Energy Fluid Systems 3.6-6 I 3.6.2.1.2 Definition of Moderate Energy Fluid Systems 3.6-6 O O 3.6-il Amendment 1

b 23A6100AI! Standard Plant ne/. A SECTION 3.6 CONTENTS (Continued) $ Section Title Bige 3.6.2.13 Postulated Pipe Breaks and Cracks 3.6-6 3.6.2.1.4 Locations of Postulated Pipe Breaks 3.6-7 3.6.2.1.4.1 Piping Meeting Separation Requirements 3.6-7 , 3.6.2.1.4.2 Pipingin Containment Penetration Areas 3.6-7 3.6.2.1.43 ASME Code Section III Class 1 Piping in Areas Other Than Containment Penetration 3.6-9 3.6.2.1.4.4 For ASME Code Section III Class 2 and 3 Piping in Areas Other Than Containment Penetration 3.6-9 3.6.2.1.4.5 Non ASME Class Piping 3.6-10 3.6.2.1.4.6 Separating Structure with liigh. Energy Lines 3.6-10 3.6.2.1.4.7 Piping in the Steam Tunnel and Seismic Interface Restraint 3.6 10 3.6.2.1.5 locations of Postulated Pipe Cracks 3.6-10 3.6.2.1.5.1 Piping Meeting Separation Requirements 3.6-10 3.6.2.1.5.2 liigh Energy Piping 3.6-10 3.6.2.1.53 Modcrate Energy Piping 3.6-10 3.6.2.1.53.1 Pipingin Containment Penetration Areas 3.6-10 3.6.2.1.53.2 Piping in Areas Other Than Containment Penetration 3.6 10 3.6.2.1.5.4 Moderate Energy Piping in Proximity to liigh Energy Piping 3.6-11 3.6.2.1.6 Types of Breaks and Cracks to be Postulated 3.6-11 3.6.2.1.6.1 Pipe Breaks 3.6-11 3.6.2.1.6.2 Pipe Cracks 3.6-12 3.6-iii 0 Amendment 1

23A6100AE Standard Plant anv. A I I SECTION 3.6 l n l V CONTENTS (Continued) ' Section Title Page 3.6.2.2 Analytic hiethods to Define Blowdown Forcing Functions and Response hiodels 3.6-13 3.6.2.2.1 Analytical hiethods to Define Blowdown Forcing Functions 3.6-13 3.6.2.2.2 Pipe Whip Dynamic Response Analyses 3.6-14 3.6.23 Dynamic Analysis hfethods to Verify Integrity and Operability 3.6-15 3.6.23.1 Jet Imp'.ngement Analyses and Effects on Safety-Related Components 3.6-15 3.6.23.2 Pipe Whip Effects on Essential Components 3.6-18 3.6.23.2.1 Pipe Displacement Effects on Componentr, in the Same Piping Run 3.6-18 3.6.23.2.2 Pipe Displacement Effects on Safety Related Structures, Other Systems, and Components 3.6-19 3.6.233 Loading Combinations and Design Criteria for Pipe Whip Restraints 3.6 19 3.6.2.4 Guard Pipe Assembly Design 3.6-22 3.6.2.5 hiaterial to be Supplied for the Operating License Review 3.6 22 3.63 Isak-Before Break Evaluation Prccedures 3.6-22 3.(31 General Evaluation 3.6-23 3.63.1 Det-rministic Evaluation P2 ocedure 3.6-24 3.6.4 Interfaces 3.6 27 3.6.4.1 Summary of Pipe Break Analysis Resuits 3.6-27 1 3.6.4.2 Leak Before Break Analysis Repo" 3.6-27 3.6.5 References 3.6-27 1 3.6-iv i Amendinent 1

ABWR 3346iman Standard Plant REV.A SECTION 3.6 TABLES O Table Illic Eage 3.6-1 Essential Systems, Components, and Equipment for Postulated Pipe Failures inside Containment 3.6-28 3.6-2 Essential Systems, Components, and Equipment for Postulated Pipe Failures Outside Containment 3.6-30 3.63 liigh Energy PipingInside Containment 3.6-31 3.6-4 liigh Energy Piping Outside Containment 3.6 32 3.6-5 Cornparison of PDA and NSC Code 3.5-33 ILLUSTRATIONS Figure Illic Eage 3.6-1 Typical Restraint Force Deflection Curve, PDA Verification Program 3.6-34 3.6-2 Piping configuration (BWR5 Recirculation System) Used in Verification of PDA g Computer Program 3.6-35 3.6-3 Jet Characteristics 3.6-36 3.6-4 liomogeneous Jet Asymptotic Area - Saturated Water and Steam Blowdown 3.6-37 3.6-Sa fl./D Versus Ration of Width of Jet at Asymptotic Plane to Width of Jet at l Break Plane Steaa 3.6-38  : 3.6-5b IL/D Versus Ration of Width of Jet at Asymptotic Plane to Width of Je ' t Break Plane - Saturated Water 3.6 39 3.6-6 Typical Pipe Whip Restraint Configuration 3.6-40 l 3.6-v 9 Amendment 1

l ABM 234aooxe l Standard Plant REV A 3.6 PROTECTION AGAINST DYNAMIC Subsection 3.6.3 describes the implementation l n EFFECTS ASSOCIATED WITil TIIE of the leak.before-break (LBB) evaluation proce- I t.' y POSTULATED RUPTURE OF PIPING dures as permitted by the broad scope amendment to General Design Criterion 4 (GDC 4) published This Section deals with the structures, in Reference 1. The piping systems that are systems, components and equipment in the ABWR demonstrated by these procedures to qualify for Standard Nuclear Island. the LBB behavior (See Appendices 3E and 3F) are not postulated to break in the d sign and evalu-Subsections 3.6.1 and 3.6.2 describe the ation that are required to be performed, in design bases and protective measures which ensure accordance with Subsections 3.6.1 and 3.6.2, for that the containment; essential systems, compo- the potential dynamic effects from postulated nents and equipment; and other essential struc- piping breaks. However, such piping systems are tures are adequately protected from the conse- evaluated for pipe crack effects in accordance quences associated with a postulated rupture of with Subsections 3.6.2.1.5 and 3.6.2.1.6.2. high energy piping or crack of moderate-energy piping both inside and outside the containment. 3.6.1 Postulated Piping Failures In Fluid Systems Inside and Before delineating the criteria and assump- Outside of Containment tions used to evaluate the consequences of pip-ing failures inside and outside of containment, This subsection sets forth the design bases, it is necessary to define a pipe break event and description, and safety evaluation for determin-a postulated piping failure: ing the effects of postulated piping failures in fluid systems both inside and outside the con-Pipe break event: Any single postulated tainment, and for including necessary protective piping failure occurring during normal plant measures. operatior and any subsequent piping failure and/or eqmpment failure that occurs as a direct 3.6.1.1 Design Bases consequence of the postulated piping failure. (G) 3.6.1.1.1 Criteria Postulated Piping Failure: Longitudinal or circumferential break or rupture postulated in Pipe break event protection conforms to 10CFR50 high energy fluid system piping or throughwall Appendix A, General Design Criterion 4, Environ-leakage crack postulated in moderate energy fluid mental and Missile Design Bases. The overall system piping. The terms used in this definition design for this protection is in general compli-are explained in Subsection 3.6.2. aqce with NRC Branch Technical Positions (BTP) ASE 31 and MEB 31 included in Subsections Structures, systems, components and equipment 3.6.1 and 3.6.2, respectively, of NUREG-0800 that are required to shut down the reactor and (Standard Review Plan), mitigate the consequences of a postulated piping failure, without offsite power, are defined as MEB 3-1 describes an acceptable basis for essential and are designed to Seismic Category I selecting the design locations and orientations requirements. of postulated breaks and cracks in fluid systems piping. Standard Review Plan Sections 3.6.1 and The dynamic effects that may result from a 3.6.2 describe acceptable measures that could be postulated rupture of high. energy piping include taken for protection against the breaks and iaissile generation; pipe whipping; pipe break cracks and for restraint against pipe whip that reaction forces; jet impingement forces; compart- may result from breaks. ment, subcompartment and cavity pressurizations; decompression waves within the ruptured pipes and The design of the containment structure, com- l seven typet of loads identified with loss of ponent arrangement, pipe runs, pipe whip re- I coolant accident (LOCA) on Table 3.9 2. straints and compartmentalization are done in l O Amendment 1 361

ABWR m6umn Standard Plans REV.A consonance with the acknawledgment of protection in item (4) below. A SACF is malfunction or against dynamic effects associated with a pipe break event. Analytically sized and positioned loss of function of a component of electric-al or fluid systems. The failure of an ac. g d pipe whip restraints are engineered to preclude tive component of a fluid system is consi-damage based on the pipe break evaluation. dered to be a loss of component function as a result of mechanical, hydraulic, or elec. 3.6.1.1.2 Objecti es trical malfunction but not the loss of com-ponent structural integrity. The direct Protection against pipe break event dynamic consequences of a SACF are considered to be effects is provided to fulfill the following a part of the single active failure. The objectives: single active component failure is assumed to occur in addition to the postulated (1) Assure that the reactor can be shut down piping failure and any direct consequences safely and maintained in a safe cold shut- of the piping failure, down condition and that the consequences of the postulated piping failure are mitigated (4) Where the postulated piping failure is to acceptable limits without offsite power. assumed to occur in one of two or more re-dundant trains of a dual purpose moderate-(2) Assure that containment integrity is energy essential system (i.e., one required maintained. to operate during normal plant conditions as well as to shut down the reactor and (3) Assure that the radiological doses of a mitigate the consequences .a the piping postulated piping failure remain below the f ailure), single active f ailure of limits of 10CFR100. components in the other train or trains of that system only are not assumed, provided 3.6.1.1.3 Assumptions the system is designed to Seismic Category I standards, is powered from both offsite and The following assumptions are used to determine the protection requirements, onsite sources, and is constr ucted, operated, and inspected to quality $ assurance, testing and inservice inspection (1) Pipe break events may occur during normal standards appropriate for nuclear safety-plant conditions (i.e., reactor startup, related systems. Residual heat removal operation at power, normal hot standby

• or system is an example of such a system.

reactor cooldown to a cold shutdown conditions but excluding test modes) (5) If a pipe break event involves a failure of non Seismic Category I piping, the pipe (2) A pipe break event will not occur break event most not result in failure of simultaneously with a seismic event; or a essential systems, components and equipment seismic event will not initiate a pipe break to shut down the reactor and mitigate the event. This applies to Seismic Cateery I consequences of the pipe break event and non Seismic Category I piping. considering a SACF in accordance with items (3) and (4) above. (3) A single active component failure (SACF) is assumed in systems used to mitigate conse- (6) If loss of offsite power is a direct con-quences of the postulated piping failure and sequence of th. pipe break event (e.g., trip to shut down the reactor, except as noted of the turbine generator producing a power Normal hot standby is a normally attained zero power plant operating state (as opposed to a hot standby initiated by a plant upset condition) where both feedwater and main condenser are available and in use. Amendment 1 362

ABM MA6100AE I Standard Plant REV.A surge which in turn trips the main breaker), 3.6.1.1.4 Approach i then a loss of offsite power occurs in a i O* mechanistic time se2uence with a S^CF. To comrix with the objeciives previoesix Otherwise, offsite power is assumed described, the essential systems, components, available with a SACF. and equipment are identified. The essential systems, components, and equipment, or portions (7) A wh:pping pipe is not capable of rupturing thereof, are identified in Table 3.61 for pip-impacted pipes of equal or greater nominal ing failures postulated inside the containment pipe diameter, but may develop throughwall and in Table 3.6 2 for outside the containtr :nt. cracks in equal or larger nominal pipe sizes with thinner wall thickness. 3.6.1.2 Descriptlen (8) All available systems, including those ac- The lines identified as high energy per tuated by operator actions, are available to Subsection 3.6.2.1.1 are listed in Table 3.6-3 mitigate the consequences of a postulated for inside the containment and in Table 3.6 4 piping failure. In judging the availability for outside the containment. Moderate energy of systems, account is taken of the postu- piping defined in Subsection 3.6.2.1.2 is not lated failure and its direct consequences specifically listed. Pressure response anslyses such as unit trip and loss of offsite power, are performed for the subcompartments conta?ing and of the assumed SACF and its direct con- high-energy piping. A detailed discussion Of sequences. The feasibility of carrying out the line breaks selected, vent paths, room operator actions are judged on the basis of volumes, analytical methods, pressure results, ample time and adequate access to equipment etc., is provided in Section 6.2 for primary being available for the proposed actions. containment subcompartments. Although a pipe break event outside the The effects of pipe whip, jet impingement, containment may require a cold shutdown, up spraying, and flooding on required function of q to eight hours in hot standby is ellowed in essential systems, components, and equipment, or V order for plant personnel to assess the portions thereof, inside and outside the situation and make repairs. containment are considered. (10) Pipe whip occurs in the plane defined by the In particular, there are no high energy lines piping geometry and causes movement in the near the control room. As such, there are no direction of the jet reaction. If unre- effects upon the habitability of the control strained, a whipping pipe with a constant room by a piping failure in the control building energy source forms a plastic hinge and or elsewhere either from pipe whip, jei .mpinge- i rotates about the nearest rigid restraint, ment, or transport of steam. Further discussion i anchor, or wall penetration. If unre- on control room habitability system < i ;.mided ' strained, a whipping pipe without a constant in Section 6.4. energy source (i.e., a break at a closed valve with only one side s abject to 3.6.1.3 Safety Evaluation pressure) is not capable of forming a plastic hinge and rotating prc.vided its 3.6.1.3.1 General  ; movement can be defined and evaluated. 1 An analysis of pipe break events is performed (11) The fluid internal energy associated with to identify those essential systems, components, the pipe break reaction can take 'nto and equipment that provide protective actions account any line restrictions (e.g., flow required to mitigate, to acceptable limits, the limiter) between the pressure source and consequences of the pipe break event. 1 break location and absence of energy  ; reservoirs, as applicable. Pipe break events involving high energy fluid Amendment 1 3.6-3

ABM 2346iooxe Standard Plant REV.A systems are evaluated for the effects of pipe therefore, is the basic protective measure whip, jet impingement, flooding, room pressuri- incorporated in the design to protect against a zation, and other environmental effects such as the dynamic effects of postulated pipe failures. W ' temperature. Pipe break events involving moderate energy fluid systems are evaluated for Due to the complexities of several divisions wetting from spray, flooding, and other environ- being adjacent to high-energy lines in the dry-mental effects, well and reactor building steam tunnel, speci-fic break locations are determined in accordance By means of the design features such as with Subsection 3.6.2.1.4.3 for possible spatial separation, barriers, and pipe whip restraints, a separation. Care is taken to avoid concentra-discussion of which follows, adequate protection ting essential equipment in the break exclusion is provided against the effects of pipe break zone allowed per Subsection 3.6.2.1.4.2. If events for essential items to an extent that spatial separation requirements (distance and/or their atility to shut down the plant safely or arrangement to prevent damage) cannot be met mitigate the consequences of the postulated pipe based on the postulation of specific breaks, failure would not be impaired. barriers, enclosures, shields, or restraints are prosided. These methods of protection are dis-3.6.1.3.2 Protection Methods eussed on Subsections 3.6.1.3.2.3 a o d 3.6.1.3.2.4. 3.6.1.3.2.1 General For other areas where physical separation is The direct effects associated with e particu- not practical, the following high-energy line-lar postulated break or crack must be mechanis- separation analysis (IIELSA) evaluation is done tically consistent with the failure. Thus, actu- to determine which high energy lines meet the al pipe dimensions, piping layouts, material pro- spatial separation requirement and which lines l pcrties, and equipment arrangements are consider- require further protection: ) ed in defining the following specific measure for protection against actual pipe movement and other associated consequences of postulated failuret (1) For the IIELSA evaluation, no particular break poitits are identified. Cubicles or g, areas through which the high ener<,y lines (1) Protection against the dynamic effects of pass are examined in total. Breaks i re pos-pipe failures is provided in the form of tulated at any point in the piping syste: t. pipe whip restraints, equipment shields, and l physical separation of piping, equipment, (2) Essential systems, cinponents, and equipmt,t  ! and instrumentation. at a distance greater than thirty feet from ) any high energy piping are considered as ) (2) The precise method chosen depends largely meeting spatial separation requirements. No l apon limitations placed on the designer such damage is assumed to occur during to jet im-as accessibility, maintenance, and proximity pingement since the i,apingement force be-to other pipes. comes negligible beyond 30 feet. Likewise, a 30 ft evaluation zone is established for 3.6.1.3.2.2 Separation pipe breaks to assure protection against potential damage from a whipping pipe. As. The plant arrangement provides physical surance that 30 feet upresents the maximum separation to the extent practicable to maintain free length is me.de in the piping layout. the independence of redundant essential systems (including their auxiliaries) in order to prevent (3) Essential systems, components, and equipmeat the loss of safety function due to any single at a distance less than 30 feet from any postulated event. Redundant trains (e.g., A and high energy piping are eva:uated to see if B trains) and divisions are located in separate damage could occur to more than one compartments to the extent possible. Physical essential division, preventing safe shutdown separation between redundant essential systems of the plant. If damage occurred to on ly with their related auxiliary supporting features, one division of a red"ndant system, the Amendment i 36-4

23A6100AE Standard Plant anv. a requirement for redundant separation is The design criteria for restraints is given in p met. Other redundant divisions are Subsection 3.6.2.3.3. V available for safe shutdown of the plant and no further evaluation is performed. 3.6.13.3 Specific Protection Measures (4) If damage could occur to more than one (1) Nonessential systems and system components division of a redundant essential system are not required for the safe shutdown of within 30 ft of any high energy piping, the reactor, nor are they required for the other protection in the form of barriers, limitation of the offsite release in the shields, or enclosures is used. These event of a pipe rupture. Ilowever, while method of protection are discussed in Sub- none of this equipment is needed during or section 3 6.1.3.2.3. Pipe whip restraints following a pipe break event, pipe whip as discussed in Subsection 3.6.1.3.2.4 are protection is considered where a resulting used if protection from whipping pipe is not failure of a nonessential system or possible by barriers and shields. component could initiate or escalate the pipe break event in an essential system or 3.6.1.3.2J Barriers, Shields, and Enclosures component, or in another nonessential system whose failure could affect an essential Protection requirements are met through the system. protection afforded by the walls, floors, columns, abutments, and foundations in many (2) For high energy piping systems penetrating cases. Where adequate protection is not already through the containment, isolation valves present due to spatial separation or existing are located as close to the containment as plant features, additional barriers, deflectors, possible, or shields are identified as necessary to meet the functional protection requirements. (3) The pressure, watu level, and flow sensor instrumentation for those essereial systems, Q D Barriers or shields that are identified as necessary by the use of specific break locations which are required to funnon following a pipe rupture, are protected. in the drywell and steam tunnel are designed for the specific loads associated with the particular (4) liigh energy fluid system pipe whip break location. restraints and protective rr easures are designed so that a postulated 1 reak in one Barrien or shields that are identified as pipe could not, in turn, lead to a rupte.e necessary by the HELSA evaluation (i.e., based on of other nearby pipes or compt sents if the no specific break locations), are designed for secondary rupture coule result in worst case loads. The closest high energy pipe consequences that would be considered location and resultant loads are used to size the unacceptaih for the initin postulated barriers. break. 3.6.1.3.2.4 Pipe Whip Restraints (5) For any postulated pipe rup.u e, the structural integrity of the cor.r aioment Fipe whip restraints are used where pipe break structure :s maintained. In add'tior , for protection requirements could not be satisfied those postulated ruptures classMed as a using spatial separation, barriers, shields, or loss of reactor coolant, the design leak enc!osures alone. Restraints are located based tightness of the containment fission product on the specific break locationi determined in ac- barrier is maintained. cordance with Subsections 3.6.2.1.4.3 and 3.6.2.- 1.4.4. After the restraints are located, the (6) Safety / relief valves (SRV) and the Reactor piping and essential systems are evaluated for Core Isolation Cooling (RCIC) System steam-jet impingement and pipe whip. For those cases line are located and restrained so that a where jet imningement damage could still occur, pipe failure would not prevent depressuri-o barriers, shields, or enclosures are utilized. zation. l .l l Amendment t 3.65

ABWR m6mu Standard Plant REV.A (7) Separation is provided to preserve the those systems or portioni of systems that, independence of the Low Pressure Flooder during normal plant condi..ons (as defined in (LPFL) Systems. Subsection 3.6.1.1.3(1)),are either in operation or are maintained pressurized under conditions h (8) Protection for the FMCRD scram insert lines where either or Loth of the following are met: is not required since the motor operation of the FMCRD can adequately insert the control (1) maximum operating temperature exceeds rods even with a complete loss of insert 200 F, or lines (See Subsection 3.6.2.1.6.1). (2) maximum operating pressure exceeds 275 psig. (9) The escaps of steam, water, combustible or corrosive fluids, gases, and heat in the 3.6.2.1.2 Definition of Moderate Energy Fluid event of a pipe rupture do not preclude: Systems. (a) Accessibility to any areas required to Moderate energy fluid systems are defined to cope with the postulated pipe rupture; be those systems or portions of systems that, during normal plant conditions (as defined in (b) Habitability of the control room; or Subsection 3. 6.1.1. 3. ( 1 )) , are either in operation or are maintained pressurized (above (c) T b e a biii t y of essentia1 atmospheric pressure) under conditions where instrumentation, electric power both of the following ar,: met: supplies, components, and controls to perform their safety related function. (1) maximum operating temperature is 200'F or less, and 3.6.2 Determination of Break Locations and Dynamic Effects (2) maximum opsrating pressure is 275 psig or Associated with the Postulated less. Rupture of Piping g Piping systems are classified as W Information concerning break and crack moderate-energy systems when they operate as location criteria and methods of analysis for high energy piping for only short operational dynamic effects is presented in this Subsection. periods in performing their system function but, The location criteria and methods of analysis are for the major operational period, qualify as needed to evalcate the dynamic effects associated moderate energy fluid systems. An operational with postulated breaks and cracks in high and period is ennsidered short if the total fraction moderate energy fluid system piping inside and of time that the system operates within the outside of primary containment. This information pressure temperature conditions specified for provides the basis for the requirements for the high energy fluid systems is less than two protection of essential structures, systems, and percent of the total time that the system components defined in introduction of Section operates as a moderate energy fluid system. 3.6. 3.6.2.1.3 Postulated Pipe Breaks and Cracks 3.6.2.1 Criteria Used to Define Break and Crack Location and Configuration A postulated pipe break is defined as a

                                                  .udden gross failure of the pressure boundary The following subsections establish the         either in the form of a complete circumferential criteria for the location and configuration of       tverance (guillotine break) or a sudden postulated breaks and cracks,                      longitudinal split without pipe severance, and is postulated for high energy fluid systems 3.6.2.1.1 Definition of liigh. Energy Fluid        only. For moderate energy fluid system, pipe Systems                                            failures are limited to postulation of cracks in piping and branch runs. These cracks affect the High.cnergy fluid systems are defined to be     surrounding environmental conditions only and do O

Amendment t 366

ABM 23461004n l Standard Plant REV.A l not result in whipping of the cracked pipe, are generally not identified with particular High energy fluid systems are also postulated to break points. Breaks are postulated at all p) ( have cracks for conservative environmental possible points in such high energy piping l conditions in a confined area where high- and systems. However, in some systems break points j moderate energy fluid systems are located. are particularly specified per the following l subsections if special protection devices such  ! The following high energy piping systems (or as barriers or restraints are provided. portions of systems) are considered as potential i candidates for a postulated pipe break during 3.6.2.1.4.2 Piping in Containment Penetration  ! normal plant conditions and are analyzed for Areas potential damage resulting from dynamic effects: l No pipe breaks or cracks are postulated in (1) All piping which is part of the reactor those portions f piping from containment wall coolant pressure boundary and subject to to and inclueng the inboard or outboard reactor pressure continuously during station isolation valves which meet the following operation; requirement in addition to the requirement of the ASME Code, Section III, Subarticle NE 1120: j (2) All piping which is beyond the second  ; isolation valve but subject to reactor (1) The following design stress and fatigue , pressure continuously during station limits are not exceeded: l operation; and  ! For ASME Code. Section III. Class 1 Pitiinn  ; (3) All other piping systems or portions of l piping systems considered high-energy (a) The maximum stress range between any two  ! systems. loads sets (including the zero load set) does not exceed 2.4 Sm, and is Portions of piping systems that are isolated calculated

• by Eq. (10) in NB 3653, ASME O frota the source of the high energy fluid during normal plant conditions are exempted from Code, Section III. I consideration of postulated pipe breaks. This if the calculated maximum stress range includes portions of piping systems beyond a of Eq. (10) exceeds 2.4 Sm, the stress normally closed valve. Pump and valve bodies are ranges calculated by both Eq. (12) and also exempted from consideration of pipe break Eq. (13)in Paragraph NB 3653 meet the becrt e of their greater wall thickness, limit of 2.4 S m-3.6.2.1.4 Locations of Postulated Pipe Breaks (b) The cumulative usage factor is less than 0.1 Postulated pipe break locations are selected as follows: (cj The maximum stress, as calculated by Eq.

(9) in NB 3652 under the loadings 3.6.2.1.4.1 Piping Meeting Separation resulting from a postulated piping Requirements failure beyond these portions of piping does not exceed the lesser of 2.25 S m Based on the HEISA evaluation described in a n d 1.8 Sy except that following a Subsection 3.6.1.3.2.2, the high energy lines failure outside containment, the pipe which meet the spatial separation requirements between the outboard isolation vdve and For those loads and conditions in which Level A and Level B stress limits have been specified in the Design Specification. O Amendtnent 1 3.67

ABM 2346ioore Standard Plant any.A the first restraint may be permitted analyses, or tests, are performed to higher stresses provided a plastic hinge demonstrate compliance with the lirrits of is not formed and operability of the item (1). valves with such stresses is assured in accordance with the requirement (3) The number of circumferential and longi-identified in Section 3.9.3. Primary tudinal piping welds and branch connections loads include those which are deflection are minimized. Where guard pipes or pene-limited by whip restraints. tration sleeves are used, the enclosed por-tion of fluid system piping is seamless con-For ASME Code. Section III. Class 2 Pinine struction and without circumferential welds unless specific access provisions are made (d) The maximum stress as calculated by the to permit inservice volumetric examination sum of Eqs (9) and (10) in Paragraph of longitudinal and circumferential welds. NC 3652, ASME Code, Section III, considering those loads and conditions (4) The length of these portions of piping are thereof for which level A and level B reduced to the minimum Icogth practical. stress liraits are specified in the system's Design Specification (i.e., (5) The design of pipe anchors or restraints sust.ined loads, occasional loads, and (e.g., connections to containment thermal expansion) including an OBE penetrations and pipe whip restraints) do event does not exceed 0.8(1.8 Sh+ not require welding directly to the outer SA). The Sh and SA are allowable surface of the piping (e.g., flued integ-stresses at maximum (hot) temperature rally forged pipe fittings may be used) and allowable stress range for thermal except where such welds are 100 percent expansion, respectively, as defined in volumetrically examinable in service and a Article NC 3600 of the ASME Code, detailed stress analysis is performed to Section III. demonstrate compliance with the limits of (e) The maximum stress, as calculated by Eq. item (1). g (9) la NC-3653 under the loadings (6) Guard pipes or s!ceves provided for those resulting from a postulated piping portions of piping (See Subsection 3.6.2.4) failure of fluid system piping beyond in the containment penetration areas are these portions of piping does not exceed constructed in accordance with the rules of the lesser of 2.25 Sh and 1.8 Sy. Class MC, Subsection NE of the ASME Code, Section III, where the guard pipe or sleeve Primary loads include those which are is part of the containment boundary. In deflection limited by whip restraints. addition, the entire guard pipe or sleeve The exceptions permitted in (c) aboyc assembly is designed to meet the following may also be applied provided that when requirements and tests: the piping between the outboard isola-tion valve and the restraint is con- (a) The design pressure and temperature are structed in accordance with the Power not less than the maximum operating Piping Code ANSI B31.1, the piping is pressure and temperature of the onclosed either of seamless construction with pipe under normal plant conditions. full radiography of all circumferential welds, or all longitudinal and circum- (b) The Level C stress limits in NE-3220, ferential welds are fully radiographed. ASME Code, Section III, are not exceeded under the loadings associated with (2) Welded attachments, for pipe supports or containment design pressure and other purposes, to these portions of piping temperature in combination with the safe are avoided except where detailed stress shutdown earthquake. O Amendment 1 364

o ABM 234620o4e Standard Plant REV.A (c) The assemblies are subjected to a single stress or cumulative usage factor locations q pressure test at a pressure not less may be shifted; however, the initially (j than its design pressure, determined intermediate break locations need not be changed unless one of the following  : (d) The assemblics do not prevent the access conditions exists: t required to conduct the inservice examination specified in item (7). (i) The dynamic effects from the new (as built) intermediate break locations (7) A 100% volumetric inservice examination of are not mitigated by the original pipe all pipe welds would be conducted during whip restraints and jet shields, each inspection interval as defined in IWA 2400, ASME Code, Section XI. (ii) A change is required in pipe parameters such as major differences in pipe size, 3.6.2.1.4.3 ASME Class 1 Piping In Areas Other wall thickness, and routing. Han Containment Penitration 3.6.2.1.4.4 ASME Class 2 and 3 Pipingin With the exception of those portions of piping Areas Other %an Containment Penetration identified in Subsection 3.6.2.1.4.2, breaks in ASME Code, Section 111, Clars 1 piping are With the exceptions of those portions of postulated at the following locations in each piping identified in Subsection 3.6.2.1.4.2, piping and branch run: breaks in ASME Codes, Section III, Class 2 and 3 piping are postulated at the following locations (a) At terminalends' in those portions of each piping and branch run: (b) At intermediate locations where the (a) At terminal ends (see Subsection maximum stress range (see Subsection 3.6.2.1.4.3, Paragraph (a)) 3.6.2.1.4.2, Paragraph (1)(a)) as calculated by Eq. (10) and either Eq. (b) At intermediate locations selected by one of (12) or Eq. (13) exceeds 2.4 S m. the following criteria: (c) At intermediate locations where the (i) At each pipe fitting (e.g., elbow, tee, cumulative usage factor exceeds 0.1. cross, flange, and nonr,tandard fitting), welded attachment, and As a result of piping re analysis due to valve. Where the piping contains no differences between the design configuration fittings, welded attachments, or and the as built coufiguration, the highest valves, at one location at each extreme l of the piping run adjacent to the l Extremities of piping runs that connect to protective structure. structures, components (e.g., vessels, pumps, valves), or pipe anchors that act as (ii) At each location where stresses calcu. rigid constraints to piping motion and lated (see Subsection 3.6.2.1.4.2, thermal expansion. A branch connection to a Paragraph (1)(d)) by the sum of Eqs. l main piping run is a terminal end of the (9) and (10) in NC/ND 3653, ASME Code, 1 branch run, except where the branch run is Section Ill, exceed 0.8 times the sum classified as part of a main run in the of the stress limits given in NC/ND. 1 stress analysis and , shown to have a 3653. , significant effect 0- the main run l b eh a vio r. In piping runs which are As a result of piping re analysis due maintained pressurized during normal plar.t to diffeteaces between the desiga conditions for only a portion of the run configuration and the as built (i.e., up to the first norms fly closed configuration, the highest stress valve) a terminal end of such runs os the locations may be shifted; however, the piping connection to this closed valve. initially determined intermediate break L) Amendment i 3.6-9

ABM 33A01M Ali Standard Plant nin/. A locations may be used unless a redesign identified in Subsection 3.6.2.1.4.2, leakage of the piping resulting in a change in the pipe parameters (diameter, wall cracks are postulated for the most severe environmental effects as follows: g thickness, routing) is required, or the dynamic effects from the new (as built) (1) For AShfE Code, Section til Class 1 piping, intermediate break location are not at axial locations where the calculated mitigated by the original pipe whip stress range (see Subsection 3.6.2.1.4.2, restraints and jet shields. Paragrat. . (1)(a)) by Eq. (10) and either Eq. (11) or Eq. (13) in NB-3653 exceeds 1.2 3.6.2.1.4.5 Non ASSIE Class Piping Sm-Breaks in seismically analyzed non AShlE Class (2) For ASME Code, Section 111 Class 2 and 3 or (not ASME Class 1,2 or 3) piping are postulated non ASME class piping, at axial locations according to the same requirements for ASME Class where the calculated stress (see Subsection 2 and 3 piping above. Separation and interaction 3.6.2.1.4.4, Paragraph (b)(ii)) by the sum requirements between Seismically analyzed and of Eqs. (9) and (10) in NC/ND-3653 exceeds non-seismically analyzed piping are met as 0.4 ilmes the sum of the stress limits given described in Subsection 3.7.3.13. in NC/ND-3653. 3.6.2.1.4.6 Separating Structure With liigh- (3) Non ASME class piping which has not been Energy IJoes evaluated to obtain stress information have leakage cracks postulated at axial locations if a structure separates a high energy line that produce the most severe environmental from an essential component, the separating effects. structure is designed to withstand the consequen-ces of the pipe break in the high energy line at 3.6.2.1.53 Moderate. Energy Piping locations that the aforementioned criteria require to be postulated. However, as noted in 3.6.2.1.5.3.1 Piping in Containment Penetration Subsection 3.6.1.3.2.3, some structures that are Areas h identified as necessary by the HELSA evaluation (i.e., based on no specific break locations), are Leakage cracks are not postulated in those designed for worst case loads, portions of piping from containment wall to and including the inboard or outboard isolation 3.6.2.1.5 Locations of Postulated Pipe Cracks valves provided they meet the requirements of the ASME Code, Section 111, NE-1120, and the Postulated pipe crack locations are selected stresses calculated (See Subsection 3.6.2.1.4.4, as follows: Paragraph (b)(ii)) by the sum of Eqs. (9) and (10) in ASME Code, Section Ill, NC-3653 do not 3.6.2.1.5.1 Piping Meeting Separation exceed 0.4 times the sum of the stress limits Requirements given in NC 3653. Based on the HELSA evaluation described in 3.6.2.1.5.3.2 Piping in Arras Other'Ihan Subsection 3.6.1.3.2.2, the high or moderate- Containment Penetration energy lines which meet the separation require-ments are not identified with particular crack (1) L.cakage cracks are postulated in piping

ocations. Cracks are postulated at all possible located adjacent to essential structures, points that are necessary to demonstrate adequacy systems or components, except

of separation or other means of protections pro-vided for essential structures, systems and (a) Where exempted by Subsections components. 3.6.2.1.5.3.1 a n d 3.6.2.1.5.4, 3.6.2.1.5.2 liigh Energy Piplog (b) For ASME Code, Section 111, Class 1 pip-ing the stress range calculated (see g With the exception of those portions of piping Subsection 3.6.2.1.4.2, Paragraph (1) w Amendment 1 3610

MM 23A6100AE Standard Plant REV.A (a)) by Eq. (10) and either Eq. (12) or Eq. Table 3.2-1). Additionally, the 11/4 inch p (13) in NB 3653 is less than 1.2 S m. hydraulic control unit fast scram lines do Q (c) For ash 1E Code, Section III, Class 2 or 3 and not require special protection measure because of the following reasons: non-AShiE class piping, the stresses calcu-lated (see Subsection 3.6.2.1.4.4, Paragraph (a) The piping to the control rod drives (b)(ii)) by the sum of Eqs. (9) and (10) in from the hydraulic control units (HCUs) NC/ND 3653 are less than 0.4 times the sum are located in the containment under of the stress limits given in NC/ND 3653. reactor vessel, and in the reactor building away from other safety-related (2) Leakage cracks, unless the piping system is equipment; therefore should a line fail, exempted by item (1) above, are postulated it would not affect any safety related at axial and circumferential locations that equipment but only impact on other HCU result in the most severe environmental lines. As discussed in Subsection 3.6. consequences. 1.1.3, Paragraph (7), a whipping pipe will only rupture an impacted pipe of (3) Leakage cracks are postulated in fluid ,maller nominal pipe size or cause a system piping designed to nonseismic through wall crack in the same nominal standards as necessary to meet the pipe size but with thinner wall environmental protection requirements of thickness. Subsection 3.6.1.1.3. (b) The total amount of energy contained in 3.6.2.1.5.4 Afoderate Energy Piping in Proximity the 11/4" piping between normally to High. Energy Piping closed scram insert valve on the HCU module and the ball check valve in the bloderate energy fluid system piping or control rod housing is small. In the portions thereof that are located within a event of a rupture of this line, the l ( compartment of confined area involving considerations for a postulated break in ball check valve will close to prevent reactor vessel flow out of the break. high energy fluid system piping are acceptable  ; without postulation of throughwall leakage cracks (c) Even if a number of the HCU lines rup-except where a postulated leakage crack in the tured, the control rod insertion fune-moderate-energy fluid systes piping results in tion would not be impaired since the more severe environmental conditions than the electrical motor of the fine motion con-break in the proximate high energy fluid system trol drive would drive in the control piping, in which case the provisions of rods. Subsection 3.6.2.1.5.3 are applied. , (2) Longitudinal breaks are postulated only in j 3.6.2.1.6 Types of Breaks and Cracks to be piping having a nominal diameter equal to or  ! Postulated greater than four inches. 3.6.2.1.6.1 Pipe Breaks (3) Circumferential breaks are only assumed at all terminal ends. j The following types of breaks are postulated in high energy fluid system piping at the (4) At each of the intermediate postulated break locations identified by the criteria specified in locations identified to ext.eed the stress Subsection 3.6.2.1.4. and usage factor limits of the criteria in Subsections 3.6.2.1.4.3 a n d 3.6.2.1.4.4, (1) No breaks are postulated in piping having a considerations is given to the occurrence of nominal diameter less than or equal to one either a longitudinal or circumferential inch. Instrument lines one inch and less break. Examination of the state of stress nominal pipe or tubing size meet the in the vicinity of the postulated break provision of regulatory Guide 1.11 (See location is used to identify the most Amendment 1 3.6 11

ABWR U M MAE Standard Plant anv. A probably type of break. If the maximum in the center of the piping at two stress range in the longitudinal direction diametrically opposed points (but not a is greater than 1.5 times the maximum stress range in the circumferential direction, only concurrently) located so that the reaction W force is perpendicular to the plane of the the circumferential break is postulated. piping configuration and produces out of-Conversely, if the maximum stress range in plane bending. Alternatively, a single the circumferential direction is greater split is assumed at the section of highest than 1.5 times the stress range in the tensile stress as determined by detailed longitudinal direction, only the longitudi- stress analysis (e.g., finite element nal break is postulated. If no significant analysis), difference between the circumferential and longitudinal stresses is determined, then (9) The dynamic force of the fluid jet discharge both types of breaks are considered. is based on a circular or elliptical (2D x 1/2D) break area equal to the effective (5) Where breaks are postulated to occur at each cross sectional flow area of the pipe at the intermediate pipe fitting, weld attachment, break location and on a calculated fluid or valve without the benefit of stress pressure modified by an analytically or calculations, only circumf erential breaks experimentally determined thrust coefficient are postulated. as determined for a circumferential break at the same location. Line restrictions, flow (6) For both longitudinal and circumferential limiters, positive pump-controlled flow, and breaks, after assessing the contribution of the absence of energy reservoirs may be upstream piping flexibility, pipe whip is taken into account as applicable in the assumed to occur in the plane defined by the reduction of jet discharge. piping geometry and configuration for circumferential breaks and out of plane for 3.6.2.1.6.2 Pipe Cracks longitudinal breaks and to cause piping movement in the direction of the jet reac-tions. Structura1 members, piping The following criteria are used to postulate throughwall leakage cracks in high- or moderate-g restraints, or piping stiffness as demon. energy fluid systems or portions of systems. strated by inelastic limit analysis are considered in determining the piping (1) Cracks are postulated in moderate energy movement limit (alternatively, circumfer- fluid system piping and branch runs ential breaks are assumed to result in pipe exceeding a nominal pipe size of one inch. severance and separation amounting to at least a one diameter lateral displacement of (2) At axial locations determined per Subsection the ruptured piping sections). 3.6.2.1.5, t h e post uIat e d c r ac k s a r e oriented circumferentially to result in the (7) For a circumferential break, the dynamic most severe environmental consequences. force of the jet discharged at the break location is based upon the effective (3) Crack openings are assumed as a circular  ; cross-sectional flow area of the pipe and on orifice of area equal to that of a rectangle  ! a calculated fluid pressure as modified by having dimensions one half pipe diameter in an analytically or experimentally determined length and one half pipe-wall thickness in  ; thrust coefficient. Limited pipe width, displacement at the break location, line restrictions, flow limiters, positive (4) The flow form the crack opening is assumed pump controlled flow, and the absence of to result in an environment that wets all  ; energy reservoirs are used, as applicable, unprotected components within the compart-in the reduction of the jet discharge. ment, with consequent flooding in the co.n. i partment and communicating compartments, I (8) Longitudinal breaks in the form of axial based on a conservatively estimated time , split without pipe severance are postulated period to effect corrective actions. j l Amendment i 16-12 i

1 MM 23A6100AE Standard Plant RIN. A 3.6.2.2 Analytic Methods to Denne Blowdown and steady state forcing functions. Simply, Forcing Functions and Response Models. they are as follows: 3.6.2.2.1 Analytic Methods to Denne Blowdown (a) The transient forcing functions at Forcing Functions, points along the pipe result from the propagatica of waves (wave thrust) along The rupture of a pressurized pipe causes the the pipe and from the reaction force due flow characteristics of the system to change to the rnomentum of the fluid leaving the creating reaction forces which can dynamically end of the pipe (blowdown thrust), excite the piping system. The reaction forces are a function of time and space and depend upon (b) The waves cause various sections of the fluid state within the pipe prior to rupture, pipe to be loaded with time dependent break flow area, frictional losses, plant system forces. It is assumed that the pipe is characteristics, piping system, and other one dimensional in that there is no factors. The methods used to calculate the attenuation or reflection of the reaction forces for various piping systems are pressure waves at bends, elbows, and the presented in the following subsections, like. Following the rupt u re, a decompression wave is assumed to travel The criteria that are used for calculation of from the break at a speed equal to the fluid blowdown forcing functions include: local speed of sound within the fluid. Wave reflections may occur at the break (1) Circumferential breaks are assumed to result end. Vessel and free space conditions in pipe severance and separation amounting are used as boundary conditions. The to at least a one diameter lateral blowdown thrust causes a reaction force displacement of the ruptured piping sections perpendicular to the pipe break. unless physically limited by piping restraints, structural members, or piping (c) The initial blowdown force on the pipe 8 stiffness as may be demonstrated by is taken as the sum of the wave and inelastic limit analysis (e.g., a plastic blowdown thrusts and is equal to the hinge in the piping is not developed under vessel pressure (Po ) times the break loading). area (A). Af ter the initial decompression period (i.e., the time it (2) The dynamic force of the jet discharge at takes for a wave to reach the first the break location is based on the change in direction), the force is cross sectional flow area of the pipe and on assumed to drop off to the value of the a calculated fluid pressure as modified by blowdown thrust (i.e.,0.7 PoA). analytically- or experimentally determined thrust coefficient. Line restrictions, flow (d) Time histories of transient pressure, limiters, positive pump-controlled flow, and flow rate, and other thermodynamic the absence of energy reservoirs are taken properties of the fluid are used to into accounts, as applicable, in the calculate the blowdown force on the pipe reduetion of jet discharge. using the following equation: (3) All breaks are assumed to attain full area instantaneously (a rise time not exceeding 2 one millisecond is used for the initial F= (P Pa ) + SI- A pulse). Ec . Blowdown forcing functions are determined by where the following method: F = blowdown force The predicted blowdown forces on pipes fed by a pressure vessel are described by transient P = pressure at exit plane O Amendment 1 3.6 13

MkN 23A6100All Standard Plant swv. A Pa = ambient pressure (3) The analytical model adequately represents the mass / inertia and stiffness properties of u = velocity at exit plane the system. h p = density at exit plane (4) Pipe whipping is assumed to occur in the plane defined by the piping geometry and A = area of break configuration and to cause pipe movement in the direction of the jet reaction. ge = Newton's constant. (5) Piping within the broken loop is no longer (c) Following the transient period, a considered part of the RCPB. Plastic steady state period is assumed to deformation in the pipe is considered as a exist. Steady-state blowdown forces are potential energy absorber. Limits of strain calculated including frictional are imposed which are similar to strain effects. For saturated steam, these levels allowed in restraint plastic effects reduce the blowdown forces from members. Piping systems are designed so the theoretical maximum of 1.26 Po A. that plastic instability does not occur in For subcooled water, a reduction from the pipe at the design dynamic and static the theoretical maximum of 2.0 Po A is loads unless damage studies are performed found through the use of Bernoulli's and which show the consequences do not result in standard equations such as Darcy's direct damage to any essential system or equation which accounts for friction. component. 3.6.2.2.2 Pipe Whlp Dpamic Response (6) Components such as vessel safe ends and val-Analyses ves which are attached to the broken piping system, do not serve a safety related func. The prediction of time dependent and steady- tion, or failure of which would not further thrust reaction loads caused by blowdown of sub- escalate the consequences of the accident cooled, saturated, and two phase fluid from rup- are not designed to meet ASME Code-imposed tured pipe is used in design and evaluation of limits for essential components under fault-dynamic effects of pipe breaks. A detailed dis- ed loading. Ilowever, if these components cussion of the analytical methods employed to are required for safe shutdown or serve a to compute these blowdown loads is given in Sub- protect the structural integrity of an es-section 3.6.2.2.1. Following is a detailed dis- sential component, limits to meet the Code cuuion of analytical methods used to account for requirements for faulted conditions and li-this loading. mits to ensure required operability will be met. T'ac criteria used for performing the pipe whip dynamic response analyses include: The pipe whip analysis is performed using the PDA computer program, a program as described in (1) A pipe whip analysis is performed for each Appendix 3D, predicts the response of a pipe postulated pipe break, liowever, a given subjected to the thrust force occurring after a analysis can be used for more than one post- pipe break. The program treats the situation in ulated break location if the blowdown forc- terms of generic pipe break configuration which ing function, piping and restraint system involves a straight, uniform pipe fixed at one geometry, and piping and restraint system end and subjected to a time dependent thrust properties are conservative for other break force at the other end. A typical restraint locations, used to reduce the resulting deformation is also included at a location between the two ends. (2) The analysis includes the dynamic response Nonlinear and time independent stress. strain of the pipe in question and the pipe whip relationships are used to model the pipe and the rertraints which transmit loading to tbc restraint. Using a plastic hinge concept, support structores, bending of the pipe is assumed to occur only at O Amendment 1 3614

ABM1 nA6ioo^n Standard Plant RIM A the fixed end and at the location supported by comparison demonstrates that the simplified e the restraint, modeling system used in PDA is adequate for pipe h Effects of pipe shear deflection are consider-rupture loading, restraint performance, and pipe movement predictions with the meaningful ed negligible. The pipe bending moment deflec. design requirements for these low probability tion (or rotation) relation used for these loca. postulated accidents. tions is obtained from a static nonlinear cantilever beam analysis. Using the moment ro- 3.6.23 Dynamic Analysis Methods to Verify tation relation, nonlinear equations of motion of Integrity and Operability the pipe are formulated using e'nergy considera-tions and the equations are numerically integrat-ed in small time steps to yield time history of 3.6.23.1 Jet Impingement Analpes and the pipe motion. Effects on Safety Related Components A comprehensive verification of the PDA The methods used to evaluate the jet effects program has demonstrated cons;rvatism inherent in resulting from the postulated breaks of high-the program and the analytical methods utilized, energy piping are presented in this subsection. Part of this verification work includes an inde-pendent analysis by Nuclear Services Corporation The criteria used for evaluating the effects (NSC), under contract to the General Electric of fluid jets on essential structures, systems, Company of the recirculation piping system for a and components are as follows: pre ABWR standard plant design. The recircula-tion piping system was chosen for study due to (1) Essential structures, systems, and compo-its complex piping arrangement and assorted pipe nents are not impaired so as to preclude es-sizes. The NSC analysis included clastic / plastic sential functions. For any given postulat-pipe properties, clastic / plastic restraint pro- ed pipe break and consequent jet, those es. perties, and gaps between the restraint and sential structures, systems, and components pipe. The piping / restraint system geometry and need to safely shut down the plant are properties and fluid blowdown forces were the identified, same in both analyses. However, a linear approx-imation was made by NSC for the restraint load- (2) Essential structures, systems, and compo-deflection curve. This approximation is demons- nents which are not necessary to safely shut trated in Figure 3.61. The effect of this down the plant for a given break are not approximation is to give lower energy absorption protected form the consequences of the fluid  ; of a given restraint deflection. Typically, this jet. , yields higher restraint deflections and lower ' restraint to structure loads that the GE (3) Safe shutdowm of the plant due to postulated analysis. The deflection limit used by NSC is pipe ruptures within the RCPB is not the design deflection at one half of the ultimate aggravated by sequential failures of uniform strain for the GE restraint design, safety related piping and the required emergency cooling system performance is A comparison of the NSC analysis with the PDA maintained, analysis presented in Table 3.6 5 shows that the PDA predicts higher loads in 15 of the 18 (4) Offsite dose limits specified in 10CFR100 restraints identified on figure 3.6.2. This is are complied with, due to the NSC modelincluding energy absorbing effects in secondary pipe elements and structural (5) Postulated breaks resulting in jet members. However, PDA predicts higher restraint impingement loads are assumed to occur in deflections in 50% of the restraints. The higher high energy lines at full (100%) power deflections predicted by NSC for the lower loads operation of the p ant, are caused by the linear approximation used for the force deflection curve rather than by (6) Throughwall leakage cracks are postulated in differences in computer techniques. This moderate energy lints and are assumed to O Amendment 1 3.6 13

ABM 234620are Standard Plant REV.A result in wetting and spraying of essential (7) The distance of jet travel is divided into structures, systems, and components, two or three regions. Region 1 (Figure 3.6-3) extends from the break to the (7) Reflected jets are considered only when asymptotic area. Within this region the there is an obvious reflecting surface (such discharging fluid flashes and undergoes as a flat plate) which directs the jet onto expansion from the break area pressure to an essential equipment, Only the first the atmospheric pressure. In Region 2 the reflection is considered in evaluating jet remains at a constant diameter. For potential targets. partial separation circumferential breaks, the area increases as the jet expands; (8) Potential targets in the jet path are con- therefore, it is assumed that Region 3 sidered at the calculated final position of never occurs. In Region 3 (except the broken end of the ruptured pipe. This partial separation circumferential breaks), selection of potential targets is considered interaction with the surrounding adequate due to the large number of breaks environment is assumed to start and the gj t analyzed and the protection provided from expands at a ha1f angie of 10 the effects of these postulated breaks. (Dotted lines, Figures 3.6 3a and c.) The analytical methods used to determine which (8) The analytical model for estimating the targets will be impinged upon by a fluid jet and asymptotic jet area for subcooled water and the corresponding jet impingement load include: saturated water assumes a constant jet area. For fluids discharging from a break (1) The direction of the fluid jet is based on which are below the saturation temperature the arrested position of the pipe during at the corresponding room pressure or have steady-state blowdown. a pressure at the break area equal to the room pressure, the free expansion does not (2) The impinging jet proceeds along a straight occur. (9) The distance downstream from the break (3) The total impingement force acting on any where the asymptotic area is reached cross-sectional area of the jet is time and (Region 1) has been found (for distance insariant with a total magnitude circumferential and longitudinal breaks) to equivalent to the steady state fluid be approximately equal to five pipe blowdown force given in Subsection 3.6.2.2.1 diameters (Reference 3). Assuming a linear and with jet characteristics shown in Figure expansion from the break area to the 3.6-3. asymptotic area, the jet shape can be defined for Region 1 as well as Regions 2 , (4) The jet impingement force is uniformly and 3. Figure 3.6-4 is used to determine l distributed across the cross sectional area the asymptotic area. , of the jet and only the portion intercepted l by the target is considered. (10) Both longitudinal and fully separated  ! circumferential breaks are treated (5) The break opening is assumed to be a circu- similarly. The value of fL/D used in the I lar orifice of cross sectional flow area blowdown calculation is used for jet l equal to the effective flow area of the impingement also.  ; break. (11) Circumferential breaks with partial (i.e., (6) The jet impingement force is equal to the h<D/2) separation between the two ends of steady state value of the fluid blowdown the broken pipe not significantly offset force calculated by the methods described in (i.e., no more than one pipe wall thickness Subsection 3.6.2.2.1. lateral displacement) are more difficult to O Amendment 1 3 6-16

ABM 23461 man Standard Plant REV.A quantify. For these cases, the following where assumptions are made. O (a) The jet is uniformly distributed around

                                                                    ^Ae^eistremFisurea.6-4 the periphery.                                     rA= SR as explained in item (11) (f)

(b) The jet cross section at any cut through the pipe axis has the configuration (12) Target loads are determined using the depicted in Figure 3.6 3b and the jet following procedures and assumptions. regions are as therein delineated. (a) F o r b o t h t h e f u lly s e p a r a t e d (c) Thejet force Fj = totalblowdown F. circumferential break and the longitudinal break, the jet is assumed (d) The pressure at any point intersected by to reach its asymptotic at five pipe the jet is: diameters from the break (Region 1, Figure 3.6 3a). For design purposes Pj = Fj within this region, the jet is assumed to have linear expansion half angle $: AR where 0.5(DA-D) DA-D half tan d = = AR= the total 3600 area of the jet at a SD 10D radius equal to the distance from the pipe centerline to the target, where (e) The pressure of the jet is then A multiplied by the area of the target DA is the diameter of the fully submerged within the jet. expanded jet or (f) Region 1 in Figure 3.6 3b, is assumed to be 2.5 pipe diameters. 1/2  : ( '4AA' j (g) The area (AR) of the jet at target DA " intersection rT from pipe centerline .x is calculated by using Figure 3.6-Sa and 3.6-5b to determine H (width of jet at the asymptotic plane) and the where  ; J relationship i AA si determined from Figure 3.6-4. AR = 2x rTH. l Figure 3.6 Sa and 3.6 5b were developed (b) The area within Region 2 can be assumed using Figure 3.6 4 and the following to be constant out to the beginning of equations: Region 3, which starts at the intersection of a line drawn at 100 AA = asymptotic area = 2x RAH half angle dotted line (figure 3.6 3a j and c) and the boundary of the jet, in i AB = break area = 2x Rh Region 3, the area expands at a constant i 100 half angle. I rH i AA/AB" ^ l

  .o                                                                                                                l Amendment 1                                                                                          3.6 17 4

W

ABWR ux6ioaxe Standard Plant nry A (c) After determination of the total area of D = pipe OD of target pipe for a the jet at the target, the jet pressure fully submerged pipe. 3 is calculated by: W When the target (pipe) is larger than the area Fj of the jet, the effective target area equals the Pi = _ expanded jet area Ax Ate " Ax where (3) For all cases, the je; area (Ax) is as. P=i incident pressure .sumed to be uniform and the load is uniformly distributed on the impinged Ax= area of the expanded jet at the target area Ate-target intersection. (13) For the partial separation circumferential If the effective target area (Ate) is less than break, the target loads are calculated expanded jet area (A t e 5 Ax ), the target is similarly except that the jet cross section fully submerged in the jet and the impingement appears as in Figure 3.6-3b and AR " Ax load is equal to (P 1) ( A te). If the and DA = H and is calculated per item effective target area is greater than expanded (11) (g). jet area (Ate > A x), the target intercepts the entire jet and the impingement load is equal 3.6.23.2 Pipe Whip Effects on Essential to(P 1 ) ( Ax ) = Fj. The effective target Components area (Ate) for various geometries follows: This subsection provides the criteria and (1) Flat surface For a case where a target methods used to evaluate the effects of pipe with physical area At is oriented at angle displacements on essential structures, systems, 4 with respect to the jet axis and with no and components following a postulated pipe flow reversal, the effective target area rupture. g Ate i5: Pipe whip (displacement) effects on essential structures, systems, and components can be Age = (A t) (sin 4). placed in two categories: (1) pipe displacement effects on components (nozzles, valves, tees, etc.) which are in the same piping run that the (2) Pipe Surface As the jet hits the convex break occurs in; and (2) pipe whip or controlled surface of the pipe,its forward momentum is displacements onto external components such as decreased rather than stopped; therefore, building structure, other piping systems, cable the jet impingement load on the impacted trays, and conduits, etc. area is expected to be reduced. For conservatism, r:o credit is taken for this 3.6.23.2.1 Pipe Displacement EfTects on reduction and the pipe is assumed to be Companents in the Same Piping Run impacted with the full impingement load. However, where shape f actors are The criteria for determining the effects of justifiable, they may be used. The pipe displacements on inline components are as effective target area Ate I5: I0ll0*5: Ate = (DA)(D) (1) Components such as vessel safe ends and valves which are attached to the broken where piping system and do not serve a safety function or failure of which would not DA = diameter of the jet at the further escalate the consequences of the target interf ace, and accident need not be designed to meet ASME Amendment 1 3 618

MM 23A6100AE Standard Plant any. A Code Section llI. imposed limits for piping supports, are designed to function and essential components under faulted loading. carry load for an extremely low probability y gross failure in a piping system carrying (2) If these components are required for safe high energy fluid. In the ABWR plant, the shutdown or serve to protect the structural piping integrity does not depend on t'ue pipe integrity of an essential component, limits whip restraints for any piping design loading to meet the ASME Code requirements for combination including earthquake (See Subsection faulted canditions and limits to ensure 3.2.1). When the piping integrity is lost required operability are met, because of a postulated break, the pipe whip restraint acts to limit the movement of the The methods used to calculate the pipe whip broken pipe to an acceptable distance. The pipe loads on piping components in the same run as the whip restraints (i.n., those devices which serve postulated break are described in Section only to control the movement of a ruptured pipe 3.6.2.2.2. following gross failure) will be subjected to once in a lifetime loading. For the purpose of 3.6.2.3.2.2 Pipe Displacement Effects on the pipe whip restraint design, the pipe break Essential Structures, Other Systems, and is considered to be a faulted condition (See Components Subsection 3.9.3.1.1.4) and the structure to which the restraint is attached is also analyzed The criteria and methods used to calculate the and designed accordingly. The pipe whip effects of pipe whip on external components restraints are non ASME Code components; consists of the following: however, the ASME Code requirements are specified in the design selectively to assure (1) The effects on essential structures and its safety related function if ever needed, barriers are evaluated in accordance with the barrier design procedures given in The pipe whip restraints utilize energy Subsection 3.5.3 absorbing U rods to attenuate the kinetic energy of a ruptured pipe. A typical pipe whip

 ,_) (2) If the whipping pipe impacts a pipe of equal   restraint is shown in Figure 3.6 6. The or greater nominal pipe diameter and equal principal feature of these restraints is that or greater wall thickness, the whipping pipe they are installed with several inches of does not rupture the impacted pipe, annular clearance between them and the process Otherwise, Ge impacted pipe is assumed to pipe. This allows for installation of normal be ruptured.                                 piping insulation and for unrestricted pipe thermal movements during plant operation.

(3) If the whipping pipe impacts other Select critical locations inside primary components (valve actuators, cable trays, containment are also monitored during hot conduits, etc.), it is assumed that the functional testing to provide verification of l impacted component is unavailable to adequate clearances prior to plant operation. mitigate the consequences of the pipe break The specific design objectives for the event. restraints are: I (4) Damage of unrestrained whipping pipe on (1) The restraints shall in no way increase the 1 essential structures, components, and reactor coolant pressure boundary stresses systems other than the ruptured one is by their presence during any normal mode of prevented by either separating high energy reactor operation or condition, systems from the essential systems or l providing pipe whip restraints. (2) The restraint syste.n shall function to stop i the movement of a pipe failure (gross loss 3.6.2.3.3 Ieading Combinations and Design of piping integrity) without allowing damage Criterla for Pipe Whip Restraint to critical components or missile development; and Pipe whip restraints, as differentiated from Amendment 1 3619

ABWR MA6100Ah Standard Plant nrw. A (3) The restraints should provide minimum Each of these componerts i. typically bindrance to inservice inspection of the constructed of a different ma:crial with a process piping. different design objective in order to perform the overall design functiont therefore, the For the purpose of design, the pipe whip material and inspection requirements and design restraints are designed for the following dynamic limits for each are somewhat different. The loads: requirements for each type component are as follows: (1) Blowdov.n thrust of the pipe section that impacts the restraint; (1) Type I restraints (e.g., U rods) (2) Dynamic inertia loads of the moving pipe (a) hiaterials section which is accelerated by the blowdown All materials which are used to absorb thrust and subsequent impact on the energy through significant plastic restraint; deformation conform to: (3) Design characteristics of the pipe whip 1. AShiE Code Section lli, Subsection NB, restraints are included and verified by the for Class 1 Components, and Section pipe whip dynamic analysis described in 11; or Subsection 3.6.2.2.2; and

2. A S T hi S p e c i fi c a t i o n s wit h (4) Since the pipe whip restraints are not consideration for brittle fracture contacted during normal plant operation, the control; or postulated pipe rupture event is the only design loading condition. 3. AShiE Code Section III, Subsection NF if applicable.

The pipe whip restraints are composed of several components, each of which perform a different function. These components are (b) Inspection Inspection and identification of g categorized as Types I,11,111, and IV: material conform to: Type 1 - Restraint energy absorption members 1. AShiE Code Section Ill, Subsectkm NB that, under the influence of for Class 1 Components (Section V impacting pipes (pipe whip), will Non Destructive Examinations absorb energy by significant plastic hiethods); or deformation (e.g.. U rods).

2. ASThi Specifications procedures Type 11 - Restraint connecting members which including volumetric and surface form a direct link between the inspection; or restraint plastic members and the structure (e.g., clevises, brackets, 3. AShiE Code Section III, Subsection NF, pins), if applicable.

Type 111- Restraint connecting member struc- (c) Design Limits tural attachments which provide the method of securing the restraint- 1. Design local strain The perm anent connecting members to be structure strain in metallic ductile materials (e.g., weld attachments bolts), is limited to 50% of the minimum ac-tual tiltimate uniform strain (at the Type lV - Structural and civil components are rnaximum stress on an engineering steel and concrete structures which stress-strain curve) based on re-ultimately must carry the restraint straint material tests or, when de-load (e.g., shield wall , trusses). monstrated to be as or more Amendment 1 3620

ABWR 23A61 Mali Standard Plant REV.A conservative,1/2 of minimum percent 2. Recommended stress limits per ash 1E j^ elongation as specified in the Code Section 111, Subsection NF for C applicable AShfE Code Section 111 or faulted conditions if applicable ASThi Specifications. (3) Type III restraints (fasteners)

2. Design steady state load The maximum restraint load is limited to (a) hiaterials 80% of the minimum calculated static Fastener material conforms to ASThi, ultimate restraint strength at the AShiE, or hill requirements, drywell design teniperature. This strain is less than 50% of the (b) Inspection ultimate uniform strain for all All fasteners are inspected or certified roaterials used for Type I componeuts. per applicable ASTH!, ash 1E, or hill specifications.

(2) Type II restraints (e.g., clevises, brackets, pins) (c) Design Limits Design limits are the same as for Type (a) hiaterials Il restraints. hiaterials selection conforms to: (4) Type III restraints (welds)

1. ASThi Specifications including (a) Materials consideration for brittle fracture control, or Weld materials for attachments to carbon steel structures are limited to low

2. ASPF Code Section !!I, Subsection NF, hydrogen type, if app- _

(] (b) Inspection V (b) Inspection Liquid penetrant surface inspection are inspection conforms to: performed per ASThi Specification E165 or AWS Structural Welding Codes, AWS.D1.1.

1. AShfE/ ASTM requirements or process qualification and finished part (c) Design Limits surface inspection per ASTM methods, or The maximum primary weld stress intensity (two times maximum shear

2. ASME Code Section III, Subsection NF, stress) is limited to three times AWS or if applicable. AISC building allowable weld shear stress.

(c) Design Limits (d) Pucedures Design limits are based on the following , stress limits: Procedures and welders will be qualified I per the latest AWS Code for welding in

1. Primary stresses (in accordance with building structures.

definitions in ASME Code Section III) are limited to the higher of 70% of (5) Type IV restraints (structural and civil Su where Su = minimum ultimate components strength by tests or ASTM specification or Sy + 1/3 (Su-Sy) M a t e rial, inspe ction, a n d d e sign where Sy = minimum yield strength by requirements for the structural and civil n test or ASTM specification, or components are provided by industry k) Amendment 1 3621

ABWR m6imm Standard Plant RIN. A standards such as AISC, ACl, and ASME Code 3.6.2.4 Guard Pipe Assembly Design Section III, Division 11, along with appropriate requirements imposed for similar The ABWR single primary containment does not h loading events. These components are also require guard pipes to avoid steam bypass of the designed for other operational and accident suppression pool in case of a pipe break or loadings, seismic loadings, wind loadings, crack at the containment penetration. All fluid and tornado loadings. system piping that may require postulation of a pipe break or crack (See introduction to Section The design basis approach of categorizing 3.6 and subsection 3.6.2.1.6) in the containment components is consistent in allowing less penetration areas meet the break exclusion stringent inspection requirements for those requirements of Subsection 3.6.2.1.4.2 o r components subject to lower stresses. 3.6.2.1.5.3.1, as applicabic. Considerable strength margins exist in Type 11 through IV components up to the limit of load 3.6.2.5 Material to be Supplied for the capacity (fracture) of a Type I component. Operating Ucense Review Impact properties in all components are considered since brittle type failures could See Subsection 3.6.4. reduce the restraint system effectiveness. 3.6.3 Leak Before Break In addition to the design considerations, Evaluation Procedures strain rate effects and other material property variations have been considered ir. the design of Per Regulatory Guide 1.70, Revision 3, the the pipe whip restraints. The material safety analysis Section 3.6 has traditionally properties utilized in the design have included addressed the protection measures against one or more of the following methods: dynamic effects associated with the non-mechanistic or postulated ruptures of piping. (1) Code minimum or specification yield and ultimate strength values for the affected The dynamic effects are defined in introduction to Section 3.6. Three forms of piping failure g components and structures are used for both (full flow area circumferential and longitudinal the dynamic and steady state events; breaks, and throughwall leakage crack) are postulated in accordance with Subsection 3.6.2 (2) Not more than a 10% increase in minimum code and Branch Technical Position MEB 3 l of NUR EG - or specification strength values is used 0800(Standard Resiew Plan), when designing components or structures for the dynamic event, and code minimum or flowever, in accordance with the revised specification yield and ultimate strength General Design Criterion 4 (GDC 4) the values are used for the steady state loads: mechanistic leak.before break approach (LBB), justified by appropriate fracture mechanics (3) Representative or actual test data values techniques, is now (Reference 1) an acceptable are used in the design of components and procedure to exclude design against the dynamic structures including justifiably elevated effects from the postulation of breaks in high-strain rate afft cted stress limits in excess energy piping. Described in this Subsection are of 10%; or the criteria and procedures for the LBB approach which are utilized to qualify piping for (4) Representative or actual test data are used exclusion from postulation of breaks. This for any affected component (s) and the Subsection is based on proposed (Reference 4) minimum code or specification values are Section 3.6.3 of NUREG 0800. used for the structures for the dynamic and the steady state events The LBB approach is not used to exclude postulation of cracks and associated effects in l Amendment 1 3622 O 4

ABWR 2346iooxu Standard Plant nry A l accordance with Subsections 3.6.2.1.5 and portion thereof) is evaluated with the following a 3.6.2.1.6.2. considerations in addition to the deterministic L) LBB evaluation procedure of Subsection 3.63 2 The LBB approach is not applicable to piping systems where operating experience has indicated (1) Degradation by erosion, erosion /cotrosion l particular susceptibility to failure from the and emsion/ cavitation due to unfavorable l effects of intergranular stress corrosion flow conditions and water chemistry is l cracking (IGSCC), water hammer, thermal fatigues, examined. The evaluation is based on ti e or erosion. industry experience and guidelines. Addi-tionally, fabrication wall thinning of el-The LBB approach is not used in tNs safety bows and other fittings is considered in the analysis report to replace existing regulations purchase specification to assure that the or criteria pertaining to the design bases of code minimum wall requirements are rut. Emergency Core Cooling System (Subsection 6.3), These evaluations demonstrate that these me-Containment System (Subsection 6.2) or equipment chanisms are not potential sources of pipe qualification (Subsection 3.11). However, bene- rupture fits of tl.e LBB procedures to these areas will be taken and the subsections will be revised as the (2) The ABWR plant design involves operation regulations will be relaxed by the NRC. For below 7000F in ferritic steel piping and clarity, it is noted that the LBB approach is not below 8000F in austenitic steel piping, used to relax the design requirements of the pri- This assures that creep and creep fatigue mary containment system that includes the primary are not potential sources of pipe rupture. containment vessel (PCV), vent systems (vertical flow channels and horizontal vent discharges), (3) The design also assures that the piping drywell zones, suppression chamber (wetwell), material is not susceptible to brittle vacuum breakers, PCV penetrations, and drywell cleavage-type failure over the full range of head. However,in designing for loads per Table system operating temperatures (that is, the 3.9 2, which does not apply to these PCV subsys- material is on the upper shelf). (] ' tems, the seven types of design loads identified with LOCA induced dynamics of suppression pool or (4) The ABWR plant design specifies use of shield wall annulus pressurization are excluded austenitic stainless steel piping made of if they are a result of LOCA postulated in those material (e.g., nuclear grade or low carbon piping that meet the LBB criteria. type) that is recognized as resistant to IGSCC. The material of piping in reactor Appendix 3E characterizes fracture mechanics coolant pressure boundary is ferritic steel. properties of piping materials and analysis me-thods including leakage calculation methods, as (5) A systems evaluation of potential water required by the criteria of this Subsection. hammer is made to assure that pipe rupture Following NRC's review and approval, this appen- due to this mechanism is unlikely. Water dix will become approved LBB methodology for app- hammer is a generic term including various lication to ABWR Standard Nuclear Island piping. unanticipated high frequency hydrodynamic Appendix 3F applies these properties and metheds events such as steam hammer and water l to specific piping to demonstrate their eligibi- slugging. To demonstrate that water hammer lity for exclusion under the LBB approach. As is not a significant contributor to pipe j noted in Subsection 3.6.2.5, updates of these rupture, reliance on historical frequency of l analyses for specific plants and application of water hammer ;. vents in specific piping l the approved methodolog; of Appendix 3E methods systems coupled with a review of operating i to other piping systems will be submitted as part procedures and conditions is used for this i of the operating licensing resiew, evaluation. The ABWR design includes I features such as vacuum breakers and jockey 3.6.3.1 General E5aluation pumps coupled with improved operational i procedures to reduce or eliminate the p The high energy piping system (or analyzable potential or water hammer identified by past U Amendment 1 3623

ABM DA6 M W Standard Plant niiv. A experience. Certain anticipated water (1) Use the fracture mechanics and the leak hammer events, such as a closure of a valve, rate computational methods that are accept-are accounted for in the Code design and ed by the NRC staff, or are demonstrated analysis of the piping, accurate with respect to other acceptable computational procedures or with (6) The systems evaluation also addresses a po- experimental data. tential for fatigue cracking or failure from thermal and mechanical induced fatigue. (2) Identify the types of materials and ma-Based on past experience, the piping design terials specifications used for base metal, avoids potential for significant mixing of weldments and safe ends, and provide the high and low temperature fluids or materials properties including toughness mechanical vibration. The startup and and tensi!c data, long term effects such as preoperational monitoring assures avoidance thermal aging, and other limitations. of detrimental mechanical vibration. (3) Specify the type and magnitude of the loads (7) Based on experience and studies by Lawrence applied (forces, bending and torsional Livermore Laboratory, potential indirect moments), their source (s) and method of sources of indirect pipe rupture are remote combination. For each pipe size in the causes of pipe rupture. Compliance with the functional system, identify the location (s) snubber surveillance requirements of the which have the least favorable combination technical specifications assures that of stress and material properties for base snubber failure rates are acceptably low, metal, weldments and safe ends. (8) Initial LBB evaluation is based on the (4) Postulate a throughwall flaw at the design configuration and stress levels that location (s) specified in (3) above. The are acceptably higher than those identified size of the flaw should be large enough so by the initial analysis. This evaluation is that the leakage is assured detection with reconciled when the as-built configuration is documented and the Code stress evaluation sufficient raargin using the installed leak detection capability when the pipes are g is reconciled. It is assured that the subjected to normal operating loads. If as-built configuration does not deviate auxiliary leak detection systems are relied significantly shm the design con'iguration on, they should be described. For the to invalidate the initial LBB evaluation, or estimation of leakage, the normal operating a new evaluation coupled with necessary loads (i.e., deadweight, thermal expansion, configuration modifications is made to and pressure) are to be combiacd based on assure applicability of the LBB procedure. the algebraic sum of individual values. (9) Sufficiently reliable, redundant, diverse Using fracture mechanics stability analysis and sensitive leak detection systems are or limit load analysis based on (11) below, provided for monitoring of leak. The system and normal plus SSE loads, determine the that is relied upon to predict the through- critical crack size for the postulated wall flaw used in the deterministic fracture throughwall crack. Determine crack size mechanics evaluation is sufficiently margin by comparing the selected leakage reliable and sensitive to justify a margin size crack to the critical crack size. of 2 on the leakage prediction. Demonstrate that there is a margin of 2 between the leakage and critical crack 3.6.3.2 Deterministic Evaluation Procedure sizes. The same load combination method selected in (5) below is used to determine The following deterministic analysis and the critical crack size, evaluation are performed as an NF.C approved method for the ABWR Standard Nuclear Island to (5) Determine margin in terms of applied loads justify applicability of the LBB concept. by a crack stability analysis. Demonstrate O Amendmeni 1 3624

ABWR us6ioose 4 Higdard Mant REV.A that the leakage size cracks will not expe- adequate, a determination is made to demon-q ricoce unstable crack growth if 1.4 times strate that the generic data base represents b the mermal pins SSE loads are applied. De- the range of plant materials to be evalu-ruonstrate that crack growth is stable and ated. This determination is based on a com-the finol crack is limited such that a parison of the plant material properties doubfocaded pips break will not occur. The identified in (2) above with those of the dead weight, thermal expansion, pressure, materials used to develop the generic data SSE (inertiat), and seismic anchor motion base. The number of material heats and weld (SAM) loads are combined based on the same procedures tested are adequate to cover the method used for the primary stress evalu- strength and toughness range of the actual atica by the ASME Code. The SSE (inertial) plant materials. Reasonable lower bound and SAM loads are combined by square root- tensile and toughness properties from the of the sum of-the squares (SRSS) method. plant specific generic data base are to be used for the stability analysis of indivi-(6) The piping material toughness (J-R curves) dual materials, unless otherwise justified. and tensile (stress-strain curves) properties are determined at temperatures Industry generic data bases are reviewed to near the upper range of normal plant provide a reasonable lower bound for the operation, population of material tensile and toughness properties associated with any individual (7) The specimen used to generate J R curves is specification (e.g., A106, Grade B), material assured large enough to provide crack type (e.g., austenitic stect) or welding extensions up to an amount consistent with procedures. J/T condition determined by analysis for the application. Because practical specimen The number of material heats and weld proce. size limitations exist, the abihty to dures tested should be adequate to cover the obtain the desired amount of experimental range of the strength and tensile properties crack extension may be restricted. In this expected for specific material specifica-case, extrapolation techniqua is uscd as tions or types. Reasonable lower bound described in NUREG 1061, Volume 3, or in tensile and toughness properties from the NUREG/CR-4575.Othertechnique canbeused industry generic data base are used for the ifadequatelyjustified. stability analysis of individual materials. (8) The stress strain curves are obtained over if the data are being developed from an  : the range from the proportional limit to archival heat of material, three stress- l maximum load. strain curves and three J resistance curves l from that one heat of materir.1 is sufficient. l (9) Preferably, the materials tests should be The tests should be conducted at temperatures conducted using archivil materials for the near the upper range of normal plant pipe being evaluated, if archival material operation. Tests should also be conducted at is not available, plant specific or industry a lower temperature, which may represent a wide generic material data bases are plant condition (e.g., hot standby) where pipe assembled and used to define the required break would present safety concerns similar to material tensile and toughness properties. normal operation. These tests are intended Test material includes base and weld metals, only to determine if there is any significant dependence of toughness on temperature over l (10) To provide an acceptable level of reli- the temperature range of interest. The lower l ability, generic data bases are reasonable toughness should be used in the fracture I lower bounds for compatible sets of material mechanies evaluation. One J R curu ud ont tensile and toughness properties associated stress strain curve for one base mehi and  ! with materials at the plant. To assure that weld metal are considered adequats to the plant specific generic data base is determine temperature dependence. O Amendment 1 3.6-25 J

ABWR 23461oasu Standard Plant _ nev. A (11) There are certain limitations that currently When the master curve is constructed using preclude generic use of limit load analyses Eqs. (1), (2), and (3) or (5), the allowable g to evaluate leak before break conditions circumferential throughwall Daw length can be W deterministically. liowever, a modified determined by entering the master curve at a limit load analysis can be used for stress index (SI) value determined from the austenitic steel ;.iping to demonstrate loads and austenitic steel piping material of acceptable margins as indicated below: interest. The allowable flaw size determined from the master curve at the appropriate SI Construct a master Curve where a stress index, value can then be used to determine if the SI, given by required margins are met. Allowable values of 0 are those that result in S being greater Si =S+MP m (1) than zero from Eqs. (3) and (5). The flow is plotted as a function of postulated total stress used to construct the master curve and circumferential th cughwall flaw length, L, the definition of SI used to enter the master defined by curve are defined for each material category as follows: L =20R (2) Base Metal and TIG Welds: where The flow stress used to construct the master S = hr [ 2 sins sin 0], (3) cutve is n of = 0.5 (oy+o) u

  # = 0.5 [(n - 0) n (Pm/of)]                (4) when the yield strength, ay , and the ulti-0        = half angle in radians of the postu-  mate strength, ou, at temperature are lated throughwall circumferential    known.

flaw. if the yield and ultimate strengths at temper. g R = pipe mean radius, that is, the aver- ature are not known, then Code minimum values age between the inner and outer at temperature can be used, or alternatively radius, if Pm = the combined membrane stress, CD < 2.5, then including pressure, deadweight, and 17M seismic components, or = $1 ksi, or M = 1.4, the margin associated with the load combination method selected for if the analysis, per item (5). ) (SD > 2.5, then i at = flow stress for austenitic steel 17M  ; pipe material categories. I at = 45 ksi. If 8 + $ from Eqs. (2) and (4) is greater than x,then The value of SI used to enter the master cur /c for base metal and TIG welds is S = hr [ sins] (5) x Si = M (Pm+P) (7) b where where p= x(Pm/of). (6) Pb = the combined primary bending stress, Amendment 1 3 6-26

ABM 23461ooxe Standard Plant niw. A including deadweight and SSE (inertial) 3.6.4.2 laak Before Break Analysis Report components.

     /                                                        As required by Reference 1, an LBB analysis Shielded hietal Arc (ShfAW) and Submerced Arc report shal! be prepared for the piping systems (SAW) Welds:                                   excluded from the analyses for the dynamic effects due to their failure. The report shall The flow stress used to construct the master   include only the piping stress analysis results curve is 51 ksi                                for the piping systems analyzed and reported for LBB in Appendix 3F in order to show that the The value of SI used to enter the master stress levels assumed in Appendix 3F are not curve for ShiAW and SAW is                     exceeded by the reported results.

SI = hi (Pm+Pb+P)Z e (8) 3.6.5 References where 1. Modification of General Design Criterion 4 Requirements for Protection Against Dynamic Pb = the combined primary bending stress, Effects of Postulated Pipe Rupture, Federal including deadweight and seismic Renister. Volume 52, No. 207, Rules and components. Regulations, Pages 41288 to 41295, October 27,1987 Pe = combined expansion stress at normal operation. 2. RELAP 3, A Computer Program for Reactor Blowdown Analysis, IN-1321, issued Junc Z = 1.15 [1.0 + 0.013 (OD-4)] for ShiAW, (9) 1970, Reactor Technolony TID 4500. Z = 130 [1.0 + 0.010 (OD-4)) for SAW, (10) 3. hioody, F. J., Pluid Reactor and Impingement Lmds, Vol.1, ASCE Specialty Conference on and Structural Design of Nuclear Plant {)g

s. Facilities pp. 219 262, December 1973.

OD = pipe outer diameter in inches.

4. Standard Review Plan; Public Comments When the allowable flaw length is determined Solicited, Federal Recister. Volume 52, No.

from the master curve at the appropriate SI 167, Notices, Pages 32626 to 32633, August value, it can be used to determine in the 28, 1987. required margins on load and flaw size are { , met using the following procedure, l 1 7 For the method of load combination described in item (5), let hi = 1.4, and if the allowable flaw length from the master curve is at least equal to the leakage size flaw, , then the margin on load is met. I l 3.6.4 Interfaces  ! 3.5.4.1 Summary of Pipe Break ' ilysis Results A summary of the dynamic analyses applicable to high energy piping systems shall be prepared in accordance wi'.h Subsectic' 3.6.2.5 of Regulatory Guide 1.70, Revis,on 3. O At endment 1 3.6-27

23A6100AE Standard Plant any. A Table 3.61 O ESSENTIAL SYSTEMS, COMPONENTS, AND EQUIPMENT

• FOR POSTULATED PIPE FAILURES INSIDE CONTAINMENT

1. Reactor Coolant Pressure Boundary (up to and including the outboard isolation valves)

2. Containment Isolation system and Containment Boundary (including liner plate) l

3. Reactor Protection system (SCRAM SIGN.1LS) l

4. Emergency Core Cooling Systems" (For LOCA events only)  ;

4 One of the following combinations is available (see Table 6.3-3): l 1 (a) HPCF(B and C) + RCIC + RHR LPFL(B and C) + ADS (b) HPCF (B and C) + RHR LPFL (A and B and C) + ADS (c) HPCF (B or C) + RCIC + RHR LPFL (A and either of B or C) + ADS

5. Core Cooling Systems (other than LOCA events) l (a) HPCF (B or C) or RCIC (b) RHR LPFL(A or B or C) + ADS (c) RHR shutdown Cooling Mode (two loops) l O

V (d) RHR Suppression Pool Cooling Mode (two loops)

6. Control rod drive (scram / rod insertion)

7. Flow restrictors (passive)

8. Atmospheric control (for LOCA event only)

9. Standby gas treatment *" (for LOCA event only)

10. Control Room Emironmental"'

11. The following equipment / systems or portions thereof required to assure the proper operation of those essential items listed in items 1 through 10.

(a) Class 1E electrical systems, ac and de (including diesel generator system"*,6900,480 and 120V ac, and 125V de emergency buses, motor control centers * * *, switchgear"*, batteries"* and distribution systems) O Amendment 1 3.6 28

ABWR 334atoaan Standard Plant anv. A Table 3.6-1 ESSENTIAL SYSTEMS, COMPONENTS, AND EQUIPMENT

• FOR POSTULATED PIPE FAILURES INSIDE CONTAINMENT (Continued) h (b) Reactor Building Cooling Water"' to the following:

1. Room coolers

2. Pump Coolers

3. Diesel generator jacket coolers

4. Electrical switchgear coolers (c) Emironmental Systems"* (HVAC)

(d) Instrumentation (including post LOCA monitoring) l l l O l l NOTE l The essential items listed in this table are protected in l accordance with Subsection 3.6.1 consistent with the particular l pipe break evaluated. Reference Section 6.3 for detailed discussion of emerg,ncy core cooling capabilities.

              "* Located outside containment but listed for completeness of essential shutdown requirements.

O Amendment 1 3.6-29

23A6t00AE standard Plant anv A Table 3.6 2 C' essExTi4LsysTEus,COueOxtxTs.ixo EouiemEsT rOa POSTULATED PIPE FAILURES OUTsIDE CONTAINMENT

1. Containment Isolation System and centainment boundary,

2. Reactor Protection System (SCRAM signals) i

3. Core Cooling systems ]

(a) HPCF(B or C) or RCIC (b) RHR-LPFL (A or B or C) + ADS (c) RHR shutdown cooling mode (two loops) (d) RHR suppression pool cooling mode (two loops) 1

4. Flow restrictors

                                                                                                            )

5. Control room habitability l

6. Spent fuel pool cooling j

7. Standby gas treatment

8. The following equipment / systems or portions thereof required to assure the proper operation of those essential items listed in items 1 through 7.

(a) Class 1E electrical systems, ac and dc (including diesel generator j system,6900,480 and 120V ac, and 125V de emergency buses, motor ' control centers, switchgear, batteries, auxiliary shutdown control panel, and distribution systems). (b) Reactor Building Cooling water to the following: (1) Roorn coolers (2) Pump coolers (motors and seals) (3) Dieselgeneratorjacket coolers (4) Electricalswitchgear coolers (c) HVAC  ; 1 (d) Instrumentation (including post accident monitoring) The essential items listed in this table are protected in accordance with Subsection 3.6.1 consistent with the particular pipe break O evaluated. Amendment 1 3.6-30

ABWR ua6iman Standard Plant nity. A Table 3.6 3 HIGil ENERGY PIPING INSIDE CONTAINMENT h Piping System Main Steam Main Steam drains Steam Supply to RCIC Feedwater Recirculation Motor Cooling IIPCS (RPV to first check vahr) RilR LPFL (RPV to first check vahr) RilR (Suction from RPV to first normally closed gate vahr) Reactor Water Cleanup (from RilR and RPV drain) RPV IIcad Spray (RPV to first check vahc) RPV Vent (RPV to first closed valve) Standby Liquid Control (from IIPCS to lirst check vahr) CRD (Scram / rod insertion) RPV Bottom licad Drain Lines (RPV to first closed vahrs) Miscellaneous 3 inch and smaller piping

                                                                                                )

O Amendment 1 3631 t

l 23A6100A[I Standard Plant anv. A Table 3.6 4 . O nion exsaov rieisc oursine coxTrismexT Piping System 1 I Main Steam Main Steam Drains Steam supply to RCfC Turbine ] Feedwater CRD (to and from HCU) RHR (injection to feedwater from nearest check vahrs in the RHR lines) Reactor Water Cleanup (to Feedwater sia RHR and to first inlet vahe to RPV bead spray) Reactor Water Cleanup (pumps suction and discharge) - 0 l l l l 1 O Amendment 1 3.6-32

ABVM 23A6100AE Standard Plant any. A ) l Table 3.6-5 , COMPARISON OF PDA AND NSC CODE i Break Restraint Restraint  % of Design Pipe  ; indenti- Indenti- No. Deflection Restraint Deflection i fication fication of Bars Load (Klps) (In.) Deflection (in ) (Fle. 3.6-2) (Fic. 3.6-2) Ep.A ESC PD.A NS.C PD.A NS_C EJ12A NS.C fJ).A NSC RCly RCR1 5 5 803.2 788 3 6.57 7.926 79.93  %.4 17.72 15.58 RC2LL RCR1 5 5 766.4 458.4 14.99 7.495 125 62.6 35.83 24.52 RC3tt RCR2 6 6 747.0 639.7 2.27 3.73 27.65 4535 17.16 20.11 RC3Lt RCR2 6 6 7%.6 7803 10.22 10.54 57.0 59.6 41.48 43.0 RC41 ,L RCR3 5 5 846.0 838.4 7.64 8.05 92.95 97.98 18.87 16.43 RC4LL RCR3 8 8 1019.0 1073.9 5.43 4.21 99.23 76.85 2338 17.25 RC4CV RCR3 8 8 1260.7 1275.0 4.49 5.58 8037 99.89 22.56 18.73 RC6AV .7 CR 3 8 8 928.5 722.5 1.22 1.77 22.46 31.7 23.68 9539 RC7j RCi' 7 6 6 9533 801.6 6.20 5.75 76.4 79.12 16.46 21.63 g RC8LL RCR6 4 4 599.0 0 8.28 0 112.46 0 26.75 839 RC8LL RCR7 6 6 895.0 0 8.16 0 110.76 0 29316 839 RC9CV RCR6 4 4 575.8 520.16 4.16 5.53 50.63 6733 13.2 14.56 RC9LL RCR8 6 6 830.2 546.8 11.400 6.815 95.29 56.9 36.612 26.24 RCR8 6 6 8183 493.6 10.98 5.99 91.72 50.07 31.404 23.71 l RC11A l RC13 RCR10 4 4 668.4 478.0 5.87 3.66 03.5 5839 1337 10.44  ! RC16 RCR11 4 4 438 687.4 518.4 6.59 105 69.86 1537 10.22 ) i RC14CV RCR20 8 8 285.0 309.6 2.83 5.88 46 3 95.92 15.45 13.% I RC14LL RCR20 8 8 1163 129.9 0.% 336 10.5 37.1 22.13 23.56 O Amendment 1 3.6-33

AB%R MA6100AE Standard Plant REV A O % TYPICAL FORCE DEFLECTION CURVE FC A A SIX B AR,24 in. PIPE RESTR AINT LOADED AT O' 1100 F = 6 (102 228 (4 - 5 96410 24j

                                                                                                  /
                      ,x   -
                                                                                          /
                                                                                      /
                  -                                                              /

700 - DEFLECTION LIMIT l 0 /

                                                          /                                                                   !

W w

                                               /                   UNEAR IDE All2ATION 400   -

r 300 - 1

• D--- TOT A L C LE AR ANC E = 5 98 n 200 -

100 - I I I I I Og O 6 I 7 8 9 I 10 11 12 13 14 15 PIPE DE FLECTION.6 hn l Ul Ul I

                                                               +                    +       l TOTAL CLE ARANCE =                           p               (   '
                                                                                                          )

0 Y a a INITI A L - EF FECTivE - CLEARANCE CLE AR ANCE (4 in.) (198 in] 8842722 O rio#,e s.e-, Tveica' assTa^i"T ronce-oer'scrion curve. PDA VERIFICATION PROGR AM Amendment 1 36 M

l 33A6100AE Standard Plant REY 3 1 O e 5 h AC14CV A C14 L L ACR1 / v - R CR 7C 0

                                                   %M ACA10 e

ACA9 l M( 43 san I O AC11A o O 1 AC9LL l l ACA3 (1 AC9CV j

                                                                  /

ACA7 AC4LL - 5 AC4CV KEY: RC 7J

• TYPICAL BRE AK LOCAllON A C6 A V ACA3A = TYPICAL RESTR AINT DESIGN ATION 8842723 Figure 3.6-2 PIPING CONFIGURATION (BWR 5 REClRCULATION SYSTEM)

USED IN VERIFICATION OF PDA COMPUTER PROGR AM g Amendment 1 36M

ABWR 2mmt l Standard Plant arv 3 l l 1 ASYMPTOTIC AREA A A

  .                BREAK ARE A
                                                                   -.         f r,        ~ -
                                                                                              /

f- i r

                                                   %is                        /               %

R E GION REGION REGION

                                                    *--- 1              :            2     ;  '

3  : :

a. CIRCUYF E RENTI AL BRE AK - F ULL SE DAR ATION

                          ^R m N                        REGION 2 a

l ASYMPTOTIC ARE A A g

                                      \                            o                       'T l

t a j

                                    \                 /        REGION 1             ,g                                                                                    f 0       - (9     t                 1                                    v   v e

f-3 z y. l h / ~' R / n v JET CROSS SECTION FOR SEP AR ATION h6 O/2 1 H

b. CIRCUYFERENTIAL 8RE AK PARTI AL SEPAR ATION REGION 3 9

                          }                     l                         Ah f                       REGION 2                                     {           !

{ i I d g / g\ lI \\ l/ I

                                              /                       REGION 1
                                                                                                                           //
                             \              /                             Y                                          s
              --O      _

V _

                                                               )

{ O

c. LONC;TUDINAL 8RE AK 88-027 24 O F;gurE 3.6-3 JET CH ARACTERISTICS Amendment 1 3 6-Y,

ABWR ammae Standard Plant RFV A O 10$o pse STAGN ATION PRES $UME HOMOGENEOUS JET 100 UNITY VELOCITY R ATIO Ag/A8

                                                                              $ ATUR ATED WATE R 8 LOWDOWN 10  -

5 - i E 5 'AI ^8 E $ATURATED STE AM 8 LOWDOWN U E. t> 7 1.0 L 1 l I I I I 0 2o 4o so 80 100 fL/D 88 027 25 Figure 3.6-4 HOMOGENEOUS JET ASYMPTOTIC AREA - SATURATED WATER AND STEAM BLOWDOWN g Amendment 1 3637

l l l ABWR 23uixa Standard Plant nfv4 100 A$YMPTOTIC _ H - PLA N E

               ~

b 2.5D ,

                                                                                                                .j u             ,
                                                                                                                  )
               -                                                   D            p Hh 5 0/2 I

l l i Q d 10 - O l l i i i i 1.0 0 0.5 1.0 1.5 2.0 2.5 3.0 3.5 H/h 88 027 26 O Figure 3.6-Sa f L/D VERSUS RATIO OF WIDTH OF JET AT ASYMPTOTIC PLANE TO WIDTH OF JET AT BREAK PLANE - STEAM Amendment 1 343g

MN 33ASINAE Standard Plant arv 4 O

                ~

H ASYMPTOTIC _- PLANE

                                         \     c 2.50 h ( D/;
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E g 10 - E o M M N M 0 5 to 15 H/m 8d427 27 Figure 3.6-5b fL/D VERSUS RATIO OF WIDTH OF JET AT ASYMPTOTIC PLANE TO WlDTH OF JET AT BREAK PLANE - SATURATED WATER Amendment 1 3 539 i

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ABM ua61ooxu Standard Plant REV.A SECTION 3.7 O coxresrs Section Title Page j 3.7.1 Seismic inout 3.7-1 3.7.1.1 Design Response Spectra 3.7 1 3.7.1.2 Design Time History 3.7-2 3.7.13 Critical Damping Values 3.7-3 3.7.1.4 Supporting hiedia for Seismic Category I Structures 3.73 3.7.1.4.1 Soil-Structure Interaction 3.7-4 3.7.2 Seismic System Analysis 3.7-4 3.7.2.1 Seismic Analysis Methods 3.7-4 3.7.2.1.1 The Equations of Dynamic Equilibrium for Base Support Excitation 3.7-4 3.7.2.1.2 Solution of the Equations of Motion by Modal Superposition 3.7-5 3.7.2.13 Analysis by Response Spectrum Method 3.75 3.7.2.1.4 Support Displacements in Multi Supported Structures 3.7-6 3.7.2.1.5 Dynamic Analysis of Buildings 3.7 7 3.7.2.1.5.1 Description of Mathematical Models 3.7 7 3.7.2.1.5.1.1 Reactor Building and Reactor Pressure Vessel 3.77 3.7.2.1.5.1.2 Control Bullding 3.7-8 3.7.2.1.5.2 Rocking and Torsional Effects 3.78 3.7.2.1.53 Hydrodynamic Effects 3.7 8 3.7.2.2 Natural Frequencies and Response Loads 3.7-9 l i O 3.7il Amendment 1

ABWR uuiman Standard Plant anv. A SECTION 3,7 CONTFNTS (Continued) ) i Section Title Pane l l 3.7.23 Procedure Used for blodeling 3.7-9

                  .                                                                  l 3.7.23.1 hiodeling Techniques for Systents Other Than Reactor Pressure Vessel                            3.7-9 3.7.23.2 hiodeling of Reactor Pressure Vessel and Internals 3.7-9 l

3.7.2.4 Soil-Structure Interaction 3.7-10 3.7.2.5 Development of Floor Response Spectra 3.7-10 3.7.2.6 Three Components of Earthquake hiotion 3.7-10 1 3.7.27 Combination of h1odal Responses 3.7-11 l I 3.7.2.8 Interaction of Non. Category I Structures with l 1 Seismic Category I Structures 3.7 11 3.7.2.9 Effects of Parameter Variations on Floor , Response Spectra 3.7-11 I 3.7.2.10 Use of Constant Vertical Static Factors 3.7-12 O1 i 3.7.2.11 hfethods Used to Account for Torsional Effects 3.7 12 l 3.7.2.12 Comparison of Resporues 3.7 12 i l 3.7.2.13 hiethods for Seismic Analysis of Category I Dam 3.7-12 j l 3.7.2.14 Determination of Seismic Category 1 Structure Overturning hfoments 3.7 12 3.7.2.15 Analysis Procedure for Damping 3.7 13 3.73 Seismie Subsystem Analysis 3.7 14 3.73.1 Seismic Acalysis hiethods 3.7 14 3.73.2 Determination of Number of Earthquake Cycles 3.7-15 9 3.7 iii Amendment 1

i l 23A6100AE Standard Plant REY.A SECTION 3.7 O courtsrs (continuea) Section Titic Page 3.73.2.1 Piping 3.7 15 3.73.2.2 Other Equipment and Components 3.7-15 3.733 Procedure Used for Modeling 3.7-15 3.733.1 Modeling of Piping Systems 3.7 15 3.733.1.1 Summary 3.7 15 3.733.1.2 Selection of Mass Points 3.7 16 3.733.13 Selection of Spectrum Curves 3.7-16 3.733.2 Modeling of Equipment 3.7-16 3.7333 Field Location of Supports and Restraints 3.7-17 3.73.4 Basis of Selection of Frequencies 3.7-17 O 3.73.5 Use of Equivalent Static Load Methods of Analvsis 3.7 17 3.73.5.1 Subsystem Other Than NSSS 3.7 17 3.73.5.2 NSSS Subsystems 3.7-17 3.73.6 Three Components of Earthquake Motion 3.7 17 3.73.7 Combination of Modal Responses 3.7 18 3.73.7.1 Subsystems Ol'n er Than NSSS 3.7 18 3.73.7.2 NSSS Subsystems 3.7-18 3.73.7.2.1 Square-Root of the-Sum-of the-Squares Method 3.7 18 3.73.7.2.2 Double Sum Method 3.7 19 I 3.73.8 Analytical Procedure for Piping 3.7 19 3.73.8.1 Piping Subsystems Other Than NSSS 3.7-19 O 3.7iv Amendment 1

ABWR msma Standard Plant nev. A SECTION 3.7 CONTENTS (Continued) h Section Title Eags 3.73.S.1.1 Qualification by Analysis 3.7 19 3.73.8.1.2 Rigid Subsystems with Rigid Supports 3.7 19 3.73.8.13 Rigid Subsystems with Flexible Supports 3.7 19 3.73.8.1.4 Flexible Subsystems 3.7 20 3.73.8.1.5 SIatic Analysis 3.7 20 3.73.8.1.6 Dynamic Analysis 3.7 21 3.73.8.1.7 Damping Ratic 3.7 22 3.73.8.1.8 Effect of Differential Building hiovements 3.7-22 3.73.8.2 NSSS Piping Subsystems 3.7-22 3.73.8.2.1 Dynamic Analysis 3.7 22 3.73.8.2.2 Effect of Differential Building hiovements 3.7-23 h 3.73.9 h1ultiple Supported Equipment Components With Distinct Inputs 3.7 23 3.7 3.10 Use of Constant Vertical Static Factors 3.7 23 3.73.11 Torsional Effects of Eccentric h1 asses 3.7-23 3.7 3.12 Buried Seismic Category i Piping and Tunnels 3.7 23 1 3.7 3.13 Interaction of Other Piping with Seismic Category i i Piping 3.7 23 l l 3.73.14 Seismic Analysis for Reactor Internals 3.7-24  ! l 3.7 3.15 Analysis Procedures for Damping 3.7 24 3.7.4 Seismie Instrumentation 3.7 24 3.7.4.1 Compari<on with NRC Regulatory Guide 1.12 3.7 24 3.7 v 9 Amendment I

. ~ . - - . . . - . . - . . - . - . - _ - - - . -__ ABM uasiooxu Standard Plant REV.A SECTION 3.7 O CONTENTS (continued) Section Title East 3.7.4.2 Location and Description ofInstrumentation 3.7 24 3.7.4.2.1 Time-History Accelerographs 3.7-24 3.7.4.2.2 Peak Recording Accelerographs 3.7-25 3.7.4.23 Seismic Switches 3.7-25 3.7.4.2.4 Response Spectrum Recorders 3.7 25 3.7.4.2.5 Recording and Playback Equipment 3.7-25 3.7.43 Control Room Operator Notification 3.7 25 3.7.4.4 Comparison of Measured and Predicted Responses 3.7-26 3.7.4.5 In-sersice Surveillance 3.7 26 3.7.5 Interfaces 3.7 26  ! O 3.7.5.1 seis m ic e r - e'ers 3.7-26 3.7.6 References 3.7-26 1

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ABWR MMINAE i Standard Plant anv.A SECTION 3.7 TABLES h Table Title Eage 3.71 Damping for Different Materials 3.7-27 3.7 2 Natural Frequencies of the Reactor Building Complex in X Direction (0 - 180 Axis)- Fixed Base Condition 3.7 28 3.7-3 Natural Frequencies of the Reactor Building Complex in Y Direction (90 - 270 Axis) - Fixed Base Condition 3.7-29 3.7-4 Natural Frequencies of the Reactor Building Complex in Z Direction (Vertical) Fixe'd Base Condition 3.7 30 3.7 5 Natural Frequencies of the Control Building - Fixed Base Condition 3.7-30 3.7-6 Number of Dynamic Response Cycles Expected During a Seismic Event for Systems & Components 3.7 31 3.7 7 Description of Seismic Instrumentation 3.7 32 h 3.7-8 Set Points for Active Response Spectrum Recorders 3.7 33 3.7-9 Scismic Monitoring Instrumentation Surveillance Requirements 3.7 34 3.7 vii Amendmect I

l ABM ua6 ooan Standard Plant anv. 4 sECTION 3.7 O ittusrairious Fipre Title Page I l 3.7-1 Horizontal Safe Shutdown Earthquake Design i Spectra 3.7-35 3.7-2 Vertical Safe Shutdown Earthquake Design Spectra 3.7 36 3.73 Synthetic Time History, First Horizontal Direction, Damping Ratio 0.01 3.7-37 3.74 Spthetic Time History, First Horizontal Direction, Damping Ratio 0.02 3.7 38 3.7-5 Synthetic Time History, First Horizontal Direction, Damping Ratio 0.03 3.7-39 3.7-6 Synthetic Time History, First Horizontal Direction, Damping Ratio 0.04 3.7 40 3.7-7 Synthetic Time History, First Horizontal Direction, Damping Ratio 0.07 3.7 41 O 3.7 8 Synthetic Time History, First Horizontal Direction, Damping Ratio 0.10 3.7-42 I 3.79 Synthetic Time History, Second Horizontal Direction, Damping Ratio 0.01 3.7-43 3.7 10 Synthetic Time History, Second Horizontal Direction, Damping Ratio 0.02 3.7-44 . 3.7-11 Spthetic Time History, Second Horizontal Direction, Damping Ratio 0.03 3.7-45 3.7 12 Synthetic Time History, Second Horizontal l Direction, Damping Ratio 0.04 3.7-46 I l 3.7 13 Synthetic Time History, Second Horizontal I Direction, Damping Ratio 0.07 3.7-47 l 3.7 14 Synthetic Time History, Second Horizontal Direction, Damping Ratio 0.10 3.7-48 O 3.7-sili Amendment 1

ABM ax6imau  ! Standard Plant any. 6 SECTION 3.7 ILLUSTRATIONS (Continued)  : Figure Title Eage , i I 3.7-15 Synthetic Time History, Vertical Direction, Damping Ratio 0.01 3.7-49 l 3.7-16 Synthetic Time Ilistory, Vertical Direction, Damping Ratio 0.02 3.7-50 l 1 3.7-17 Synthetic Time History, Vertical Direction, l Damping Ratio 0.03 3.7 51 l l 3.7-18 Synthetic Time History, Vertical Direction, l Damping Ratio 0.04 3.7-52 l l 3.7-19 Synthetic Time History, Vertical Direction, i Damping Ratio 0.07 3.7-53 l l 3.7 20 Synthetic Time Ilistory, Vertical Direction, l Damping Ratio 0.10 3.7 54 3.7-21 Coherence Function C 12 for Earthquake . Components H1 and 112 3.7 55 3.7-22 Coherence Function C,3 for Earthquake Ol Components H1 and V 3.7 56 , 1 3.7-23 Coherence Function C23 for Earthquake Components H2 and V 3.7-57 1 3.7-24 Power Spectral Density Function of Synthetic Hi Time History 3.7 58 1 3.7-25 Power Spectral Density Function of Synthetic l H2 Time History 3.7-59 l 3.7-26 Cun1ulative Power Spectral Density Function of Synthetic H1 Time History 3.7-60 3.7-27 Cumulative Power Spectral Density Function of Synthetic H2 Time History 3.7-61 3.7-28 Seismic System Analytical Model 3.7-62 l l O 3.7ix Amendment 1

23A6100AE Standard Plant 'REV A SECTION 3.7 l h ILLUSTRATIONS (Continued) Figure Ilt[g Page 3.7 29 Reactor Building Elevation (0* - 180* Section) 3.7-63 3.7-30 Reactor BuiMing Elevation (90 - 270* Section) 3.7-64 l 3.7 31 Reactor Building Model 3.7-65 1 1 3.7 32 Reactor Pressure Vessel and Internals Model 3.7-66 I 3.7-33 Control Building Section View (0 180") land Dynamic Model 3,7-67 o  ; l O 3.7x Amendment 1

y ABWR ux6toore Standard Plant Rev A 3.7 SEISMIC DESIGN During the OBE loading condition, the safety-T related systems are designed to be capable of All structures, systems, and equipment of the continued safe operation. Therefore, for this facility are defined as either Seismic Category I loading condition, safety related structures, and or non Seismic Category I. The requirements for equipment are required to operate within design Seismic Category I identification are given in limits. Section 3.2 along with a list of systems, compo-nents, and equipment which are so identified. The seismic design for the SSE is intended to provide a margin in design that assures capabili- 1 All structures, systems, components, and equip- ty to shut down and maintain the nuclear facility l ment that are safety related, as defined in Sec. in a safe condition. In this case, it is only  ! tion 3.2, are designed to withstand earthquakes necessary to ensure that the required systems and as defined herein and other dynamic loads includ- components do not lose their capability to per-ing those due to reactor building Sibration (RBV) form their safety related function. This is caused by suppression pool dynamics. Although referred to as the no loss of function criterion this section addresses seismic aspects of design and the loading condition as the SSE loading and analysis in accordance with Regulatory Guide condition. 1.70, Revision 3, the methods of this Section are also applicable to other dynamic loading aspects. Not all safety related components have the same functional requirements. Foi example, the The safe shutdown earthquake (SSE) is that reactor containment must retain capability to earthquake which is based upon an evaluation of restrict leakage to an acceptable level. the maximum earthquake potential considering the Therefore, based on present practice, clastic regional and local geology, seismology, and behavior of this structure under the SSE loading specific characteristics of local subsurface condition is ensured. On the other hand, there material. It is that earthquake which produces are certain structures, components, and systems the maximum vibratory ground motion for which that can suffer permanent deformation without O seis-ie ceteserr 's> stems a ee-re e te re designed to remain functional. These systems and iess er r==etie . eiei 8 a vesseis ere examples of the latter where the principal components are those necessary to ensure: requirement is that they retain contents and allow fluid flow. (1) the integrity of the reactor coolant pressure boundary; Table 3.21 identifies the equipment in various systems as Seismic Category I or non-(2) the capability to shut down the reactor and Seismic Category 1. maintain it in a safe shutdown condition; and 3.7.1 Seismic Input (3) the capability to prevent or mitigate the consequences of accidents that could result 3.7.1.1 Design Response Spectra in potential offsite exposures comparable to the guideline exposures of 10CFR100. The design earthquake loading is specified in j terms of a set of idealized, smooth curves called The operating basis earthquake (OBE) is that the design response spectra in accordance with earthquake which, considering the regional and Regulatory Guide 1.60. local geology, seismology, and specific charac-teristics of local subsurface material, could Figure 3.71 shows the standard ABWR design j reasonably be expected to affect the plant site values of the horizontal SSE spectra applied at during the operating life of the plant. It is the ground surface in the free field for damping that earthquake which produce vibratory ground ratios of 2.0, 5.0, 7.0 and 10.0% of critical motion for which those features of the nuclear damping where the maximum horizontal ground power plant necessary for continued operation acceleration is 0.30g. without undue risk to the health and safety of the public are designed to remain functional. Figure 3.7-2 shows the standard ABWR design Amendment 1 33 1  ;

ABWR nuimAu Standard Plant anv. A valt.a of the vertical SSE spectra applied at the The magnitude of the SSE design time history ground surface in the free field for damping is equal to twice the magnitude of the design g ratios of 2.0, 5.0, 7.0, ar.d 10.0% of critical OBE time history. The OBE time histeries and W damping where the maximum vertical ground response spectra are used for dynamic analysis acceleration is 0.30 g at 3311z, same as the and evaluation of the structural Seismic System; maximum horizontal ground acceleration. the OBE results are doubled for evaluating the structural adequacy for SSE. For development of The design values of the OBE response spectra floor response spectra for Seismic Subsystem are one half' of the spectra shown in Figures analysis and evaluation, see Subsection 3.7.2.5. 3.71 and 3.7 2. These spectra are shown in Figures 3.7 3 through 3.7 20. The response spectra produced from the OBE design time histories are shown in Figures 3.7 3 The design spectra are constructed in through 3.7 20 along with the design OBE accordance with Regulatory Guide 1.60. The response spectra. The closeness of the two normalization factors for the maximum values in spectra in all cases indicates that the two horizontal directions are 1.0 and 1.0 as synthetic time histories are acceptable. applied to Figure 3.7-1. For vertical direction, the normalization factor is 1.0 as applied to The response spectra from the synthetic time Figure 3.7-2. histories for the damping values of 1,2,3 and 4 percent conforrs so the requirement for an 3.7.1.2 Design T1me llistory enveloping procedure provided in Item II.1.b of Section 3.7.1 of NUREG 0800 (Standard Revi:w The design time histories are synthetic Plan, SRP). Ilowever, the response spectra for acceleration time histories generated to match the higher damping values of 7 and 10 percent the design response spectra defined in Subsection show that there are some deviations from the SRP 3.7.1.1. requirement. This deviation is considered inconsequential, because (1) generating an The design time histories considered in GESSAR artificial time history whose response spectra (Reference 1) are used. They are developed based would envelop design spectra for five different g on the method proposed by Vanmarcke and Cornell damping values would result in very conservative (Reference 2) because of its intrinsic capability time histories for use as design basis input, of imposing statistical independence among the and (2) the response spectra from the synthetic synthesized acceleration time history time histories do envelop the design spectra for components. The carthquake acceleration time the lower damping values. This is very history components are identified as III,112, and important because the loads due to SSE on l V. The III and 112 are the two horizontal structures should use 7 percent damping for l components mutually perpendicular to each other. concrete components, but are obtained by I Both 111 and 112 are based on the design horizontal ratioing up the response from the OBE analysis l ground spectra shown in Figurc 3.71. The V is involving the lower damping. The OBE analysis i the vertical component and it is based on the uses only the lower damping values (up to 4%), l design vertical ground spectra shown in Figure which are consistent with the SRP requirements l 3.7 2. (See Subsection 3.7.1.3). The OBE given in Chapter 2 is one third of the SSE, i.e., 0.10 g, for the ABWR Standard Nuclear Island design. However, as discussed in Chapter 2, a more conservative value of on e-h alf of th e SSE, i.e., 0.1.S g, was employed to evaluate the structural and component response. O Amendment 1 3.7-2

23A6100AE Standard Plant nrw. A The frequency range used in generating the for the ABWR design, /3 response spectra from synthetic histories is 0.2 V to 33 Hz. The frequency range intervals used in To aid in further comparisons of PSDFs, the generating those spectra is the same as given in cumulative PSDFs of the two time histories, Table 3.7.1 1 of SRP Section 3.7.1. which represent the cumulative energy contribu-tions as a function of frequency, are computed The coherence function for the three earthquake as the cumulative area under the PSDF curves. acceleration time history components H1, H2, and V The cumulative PSDFs versus frequency as are generated to check the statistical indepen- computed for H1 and H2 time histories are shown dence among them. The coherence function for H1 in Figures 3.7-26 and 3.7-27, respectively. For and H2 is given in Figure 3.7 21; for H1 and V in comparison, the corresponding cumulative target Figure 3.7-22; and for H2 and V in Figure 3.7 23. PSDFs are also shown in these figures. As can All values within the frequency range between 0 to be seen from these figures, the calculated time 50 Hz are calculated at a frequency increment of history cumulative PSDFs envelop the cumulative 0.1 H z. The small values of these coherence target PSDF with a wide margin in the frequency functions indicate that the three components are range between 0.2 Hz and 34 Hz. sufficiently statistically independent. 3.7.13 Critical Damping Values To assess the energy content of the synthetic time history, the power spectral density functions The damping values for OBE and SSE analyses (PSDFs) are generated from the two horizontal are presented in Table 3.7-1 for various components H1 and H2. The PSDFs are computed at a structures and components. They are in frequency increment of 0.024 Hz, and are smoothed compliance with Regulatory Guides 1.61 and 1.84 using the 3-point moving average method as recommended in Reference 3. For Seismic System evaluation of the SSE, the larger SSE damping values shown in Table 3.7-1 The stationary duration used in the calculation are not used. The SSE loads are obtained by is taken to be 22 seconds which is the total doubling the OBE loads that result from the OBE l duration of the synthetic time history. The Seismic System analysis based on the lower OBE calculated PSDFs for the H1 and H2 time histories damping values (see Subsection 3.7.1.2). j are shown in Figures 3.7 24 and 3. 7-25, respec- 1 lively, for frequencies ranging from 0.2 to 34 Hz. For analysis and evaluation of Seismic Subsystems (piping, components and equipment), The target PSDFs in the form of Kanai-Tajimi the floor response spectra are obtained from the PSDF as proposed in Reference 3 are also plotted OBE time-history response of the Seismic System, on these figures for comparison. The Kanal- that supports the subsystems. The floor Tajimi PSDF is given by response spectra are computed (see Subsection 3.7.2.5) for damping values that are applicable So(w) = S o (3.7-1) to the subsystems under OBE as well as SSE; and further the OBE spectra are doubled to obtain 1+ 4 (j (w/wg)2 the SSE floor response spectra for input to the SSE analysis in design of the subsystems. 1-(w/wg)2 2+4g(wjeg)2 g 3.7.1.4 Supporting Media for Seismic Category in which So = 1,100 in /2s e c3 (this value I Structures corresponds to a peak acceleration of I g), wg = 10.66 rad /sec and (g = 0.9793. As can be seen The following ABWR Nuclear Island Seismic from Figures 3.7 24 and 3.7-25, the calculated Category I structures have concrete mat PSDFs generally envelop target PSDFs in the foundations supported on soil, rock or compacted frequency range below 10 Hz; but they fall below backfill. The maximum value of the embedment the target PSDFs in the frequency range above 10 depth below plant grade to the bottom of the Hz. However,it should be noted that conformance base mat is given below for each structure. to the target P3DF is currently not a requirement Amendment 1 3.73

ABM uA6iooan Standard Plant REV.A (1) Reactor Building (including the enclosed mode shapes, and appropriate damping factors of primary containment vessel and reactor the particular system toward the solution of the g pedestal) - 25.7 m (84 f t, 4 in.). equations of dynamic equilibrium. The time- W history approach may alternately utilize the (2) Control Building - 12.2 m (40 ft). direct integration method of solution. When the structural response is computed directly from (3) Serdee Building Surface founded. the coupled structure soil system, the time-history approach solved in the frequency domain All of the above buildings have independent is used. The frequency domain analysis method foundations. In all cases the inaximum value of is described in Appendix 3A. embedment is used for t' e dynamic analyr'h to determine seismic soil-structure interaction 3.7.2.1.1 'Ihe Equations of Dynamic Equilibrium effeets. The foundation support materials for Base Support Excitation withstand the pressures imposed by appropriate loadirig combinations without failure. The total Assuming velocity proportional damping, the structural height of each building is described in dynamic equilibrium equations for a lumped mass, Subsection 3.8.2 through 3.8.4. For details of distributed stiffness system are expressed in a the structural foundations refer to Subsection matrix form as: 3.8.5. The Nuclear Island is designed for a range of soil conditions given in Appendix 3A. (3.7-2) [M] { U(t) } + [c] { u (t) } +[K] { u (t) } = 3.7.1.4.1 Soll Structure Interaction {P(t)} When a structure is supported on a flexible where foundation, the soil structure interaction is taken into account by coupling the structural {u(t)} = time dependent displacement model with the soil medium. The finite element vector of non support points representation is used for a broad range of supporting medium conditions. A different relative to the supports (ut(t) - u(t) + u s(t)) g representation based on the continuum impedance approach is also used for selected site (0(t)} = time-dependent velocity vector conditions. Detailed methodology and results of of non support points relative the soil-structure interaction analysis are to the supports provided in Appendices 3A and 3G, respectively. (U(t)} = time dependent acceleration 3.7.2 Scismic System Analysis vector of non-support points relative to the supports This subsection applies to the design of Seismic Category I structures and the reactor [MJ = mass matrix pressure vessel (RPV). Subsection 3.7.3 applies to all Seismic Category I piping systems and [C] = damping matrix equipment. [K] = stiffness matrix 3.7.11 Seismic Analysis Methods (P(t)} = time-dependent inertia force Analysis of Seismic Category I structures and vector (-[M] (u s(t)} acting the RPV is accomplished using the response at non support points spectrum or time-history approach. The time-history approach is made either in the time doraain The manner in which a distributed mass, or in the frequency domain. distributed stiffness system is idealized into a lumped mass, distributed stiffness system of Either approach utilizes the natural period, Seismic Category I structures and the RPV is O' Amendment 1 374

ABM 23A6100AE Standard Plant REY.A shown in Figure 3.7 28 along with a schematic The mode shape vectors are also orthogonal l representation of relative acceleration; 'd (t), with respect to the mass matrix [M]. O. support acceleration; u s (t) and total a c c e l e r a t i o n ; *u*t ( t ) . The orthogonality of the mode shapes can be used to effect a coordinate transformation of the 1 3.7.2.1.2 Solution of the Equations of Motion displacements, velocities and accelerations such l by Modal Superposition that the response in each mode is independent of I the response of the system in any other mode The technique used for the solution of the Thus, the problem becomes one of solving n equations of motion is the method of modal independent differential equations rather than o superposition, simultaneous differential equations; and, since the system is linear, the principle of superposi. The set of homogeneous equations represented by tion holds and the total response of the system the undamped free vibration of the system is: oscillating simultaneously in n modes may be determined by direct addition of the responses in [M] {*d (t)} + [K] {u (t)} = {0}. (3.7-3) the individual modes. Since the free oscillations are assumed to be 3.7.2.1.3 Analysis by Response Spectrum Method harmonic, the displacements can be written as: The response spectrum method is based on the {u (t)} = {4} e,wt. (3.7-4) fact that the modal response can be expressed as a set of convolution integrals which satisfy the governing differential equations. The advantage where of this form of solution is that, for a given ground motion, the only variables under the in- * {p} = column matrix of the amplitude of tegral are the damping factor and the frequency. displacements {u} Thus, for a specified damping factor it is possi-Q w = circular frequency of oscillation ble to construct a curve which gives a maximum value of the integral as a function of frequency, t = time. Using the calculated natural frequencies of vibration of the system, the maximum values of Substituting Equation 3.7 4 and its d,erivatives the modal responses are determined directly from in Equation 3.7 3 and noting that eiwt is not the appropriate response spectrum. The modal necessarily zero for all values of et yields: maxima are then combined as discus.ed in Subsection 3.7.2.7. [-w2 [M] + [K)) {d} = {0}. (3.7-5) I When the equipment is supported at more than Equation 3.7-5 is the classic dynamic two points located at different elevations in the characteristic equation, with solution involving building, the response spectrum analysis is ) the eigenvalues of the frequencies of vibrations performed using the envelope response spectrum of wi and the eigenvalues mode shapes, {p};, all attachment points. Alternatively, the (i = 1, 2, .. . , n ) . multiple support excitation analysis methods may be used where acceleration time histories or For eaeh f requency wi, there is a response spectra are applied to all the equip'nent corresponding solution vector {d}; determined attachment points. in some cases, the worst , to within arbitrary scalar factor Y known i as single floor response spectcum selected from a l the normal coordinate. It can be shown that the set of floor response spectra obtained at various ' mode shape vectors are orthogonal with respect to floors may be applied identically to all floors l the weighting matrix [K] in the n.d.imensional provided there is no significant shift in fre- l vector space, quencies of the spectra peaks. O Amendment 1 3.75

ABM ux62 mas Standard Plant anv. A 3.7.2.1.4 Support Displacements in hiulti-Supprted Structures Cas and K as = damping and stiffness matrices denoting the g in the preceding sections, analysis procc- coupling forces developed in dures for forces and displacements induced by the active degrees of time dependent support displacement were dis- freedom by the motion of ibe cussed. In a multi supported structure there supports and vice versa; are, in addition, time dependent support dis-placements which produce additional displace. -Fa = pr eseribc d externai ments at nonsupport points and pseudo static time-dependsnt forces forces at both support and nonsupport points. applied or. Ihe actiye degrees of freedom; and 7, governing equation of motion of a structural system which is supported at rnore than Fs = reaction forces at the one point and has different excitations applied system support points. at each may be expressed in the following concise matrix form: Total differentiation with respect to time is 'hi aO a Caa Cas ont butions ithe ix d degrees of freedom a}

 --          '          -~           <

r J have been removed in the equation. The O hi s, [G U 3 )+ ,CasC 33,dU? s procedure utilized to construct the damping matrix is discussed in Subsection 3.7.2.15. The

     .K     Kas. U a      =

mass and clastic stiffness matrices are r, F] a formulated by using standard procedures.

  +                           7 K

_ as K ss. '[U3 J FJ s (3.7-6) Equation 3.7-6 can be separated into two sets where of equations. The first set of equations can be Ua = displacement of the active written as: h (unsupported) degrees of 2 . _ (3.7-7a) freedom; [hi ] {U3} + [C33](M }3 + [Kss] {Us } s Us - Specified displacements of + [Ca s) (ba ) + (Kas] {Ua } = {F3 }; support points; and the second set as: hia and hi s = lumped diagonal mass ,, , (3.7 7b) matrices associated with the [hia l (Ua} + [Ca al(U 3} + I Kaal {Ua} active degrees of freedom ._ _ _ and the support points; + [Casl (U s } + [K 33 ] (U3 ] = { F a i; Caa and Kaa = dampiig matrix and elastic The timcwise solution of Equation 3.7 'Tb can stiffness matrix, be obtained easily by using the standard normal respectively, expressing the mode solution technique. After obtaining the forces developed in the displacement response of the active degrees of active degrees of freedom freedom (U a), Equation 3.7-7a can then be used du to the motion of the to salve the support point reaci:on forces ac' - iegrees of freedom; (F3 ). Cy,< - .i .  :

• ces due to unit Afodal superposition is used to determine the and displacement solutions of the uncoupled form of Equation iorts; 3.7 7a. The procedure is identical to that described in Subsection 3.7.2.1.2.

O endment 1 3.74

ABWR m mae Standard Plant RIN. A 3.7.2.1.5 Dyna'nic Analysis of Buildirgs (a) the reinforced concrete containment vessel p (RCCV) that includes the reactor shield wall (' The time history method either in the time (RSW), the reactor pedestal, and the reactor domain or in the frequency domain is used in the pressure vessel (RPV) and its internal dynamic analysis of buildings. As for the components (b) the secondary containment zone modeling, both finite element and lumped mass having many equipment compartmente and (c) the methods are used. clean zone. The building basemat is assumed to be rigid. Building elevations along the 0 - 3.7.2.1.5.1 Description of Mathematical Models 180 and 90 270' sections are shown in Figures 3.7-29 and 3.7-30, respectively. The A mat'lematical model reflects the stiffness, mathematical model is shown in Figure 3.7-31. mass, and damping characteristics of the actual Model elevations are with respect to the RPV structural systems. One important censideration botton. head. The model X and Y axes correspond is the information required fiom the analysis. to the RB 0 -180 aad 90 270 Consideration of maximum relative displacements directions, respectiveiy. The Z axis is along among supports of Seismic Category I structures, the vertical direction. The combined RB model systems, and components require that enough as shown in Figure 3.7-31 basically consists of points on the structure be used. Locations of two uncoupled 2 D models in the X-Z and Y-Z Seismic Category I equipment are taken into plancs since the building is essentially of a consideration. Buildings are mathematically symmetric design with respect to its two modeled as a system of lumped masses located at principal directbos in the horizontal plane. elevations of mass concentrations such as floors. The coupling effects of the lateral and torsional motions on the building natural In general three dimensional models are used frequencies in the horizontal directions are for seismic analysis. In all structures, six found to be negligible. Therefore, the degrees of freedom exist for all mass points t.ncoupled 2-D models which omit the torsional (i.e., three traas1ationa1 and three degrees of freedom are used for seismic dynamic O rotational). However, in most structures, ome analysis. The methods used to account for U of the dynamic degrees of freedom can be torsional effects to define design loads are neglected or can be uncoupled form each other so given in Subsection 3.7.2.11. that separate analyses can be performed fer different types of motions. The model shown in Figure 3.7-31 corresponds to the X-Z plane. The only differences in terms Coupling between the two horizontal motiens of schematic represeraation between the X Z and l occurs when the center of mass, the centroid, and Y Z plane models are that (1) the two building l the center of rigidity do not coincide. The walls represented above EL.18.5 m (60.7ft) in degree of coupling depends on the amount of the X-Z plane by two sticks combine into one  ; eccentricity and the ratio of the uncoupled stick in the Y-Z plane, and (2) the rotational ' torsional frequency to the uncoupled lateral spring between the RCCV top slab (node 90) and frequency. Since lateral /torsiona: coupling and the basemat top (node 88) is preser':d only in tortional response can significantly influence the X-Z plane. floor accelerations, structures are in general designed to keep minimum eccentricities. Each structure in the reactor building However, for analysis of s*ructures that possess complex is idealized by a center-lined stick msual eccentricities, a model of the support model of a series of massless beam elements. building is developed to i.iclude the effect of Axial, flexural, and shear deformation effects lateral / torsional coupling, are included in formulating beam stiffness terms. Coupling between individual structures 3.7.2.1.5.1.1 Reactor Building and Reactor is modeled by linear spring elements. Masses Pressure Vessel including dead weights of structural elements, equipment weights and piping w:ights are lumped The reactor building (RB) cornplex includes: to nodal points. The weights of water in the p U Amendment 1 37-7

ABWR m6iooxit Standard Plant uv. A spent fuel storage pool and the suppression pool reactor pedestal is a cylindrical structure of a are also considered and lumped to appropriate locations. composite steel-concrete design. The total stiffness of the pedestal includes the full g strength of the concrete core. Mass points are The portions of the reactor building outside selected at equipment interface locations and the RCCV are box type shear wall systems of geometrical discontinuities, in addition, reinforced concrete construction. The major intermediate mass points are chosen to result in walls between floor slabs are represented by beam more uniform mass distribution. The pedestal clernents of a bor cross section. The shear supports the reactor pressure vessel and it also rigidity in the direction of excitation is provides lateral restraint to the reactor provided by the parallel walls. The bending control rod drive housings below the vessel. rigidity includes the cross walls contribution. The top of the RSW is connected to the RPV by The reactor building is fully integrated with the the RPV stabilizers which are modeled as spring RCCV through floor slabs at various elevations. elements. Spring elements are used to represent the slab in-plane shear stiffness in the horizontal The model of the RPV and its internal direction. The outer and inner walls between EL. components is described in Subsection 44.7 m (146.6ft) and 18.5 m (60.7ft) along the X 3.7.2.3.2. This model as shown in Figure 3.7-32 direction are also coupled rigidly in rotation is coupled with the above-described RB model for about the Y axis at the connecting slab the seismic analysis. locations. In the vertical direction a single mass point is used for each slab and it is 3.7.2.1.5.1.2 Control Building connected to the walls and R CCV by spring clements. The spring stiffness is determined so The control building section view and dynamic that the fundamental frequency of the slab in the model is shown in Figure 3.7 33. It is a vertical direction is maintained. centerlined beam model consisting of eight lumped mass points located at floor elevations. The RCCV is a cylindrical structure with a A series of beam elements is used representing flat top slab with the drywell opening, which, the building stiffness, h along with upper pool girders and reactor building walls, form the upper pool. Mass points 3.7.2.1.5.2 Rocking and Torsional Effects are selected at the RB floor slab locations. Stiffnesses are represented by a series of beam Rocking effects due to horizontal ground elements. In the X Z plane, a rotational spring movement are considered in the soil structure element connecting the top slab and the basemtt interaction analysis as described in AppenMx is used to account for the additional rotational 3A. Whenever building response is calculat ' rigidity providec' b., the integrated RCCV-pool from a second step structural analysis, rocking girder-building walls system. The RCCV is also effects are included as input simultaneously coupled to the RPV through ;he refueling bellows, applied with the horizor.tal translational motion to the RSW through the RSW stabilirsrs, and to at the basemat. 'i nc iorsional effect considered the reactor pedestal through the diaphragm is described in Subsection 3.7.2.11. floor. Spring elements are used to account for these interactions. The lower drywell access 3.7.2.1.5.3 Ilydrodynamic Effects tunnels spanning between the RCCV ud the reactor pedestal are not modeled since flexible rings are For a dynamic system in which a liquid such provided which are designed to reduce the as water is involved, the hydrodynamic effects coupling effects, on adjacent structures due to horizontal excita-tion are taken into consideration by including Tbt RSW consists of two steel ring plates with hydrodyna:aic mass coupling terms in the mass concrete fill in between for shielding purposes. matrix. The basic formulas used for caputing l Concrete in the RSW does not contribute to these terms are in Reference 4. In the vertical ' stiffness; but its weight is included. The excitation, the hydrodynamic coupling effects l Amendment 1 378 l l

ABM utsicorn Standard Plant REV.A are assumed to be negligible and the water mass R=f Fundamental frv,iency of the supported is lumped to appropriate structural locations. subsystem / frequency of the dominant O support motion 3.7.2.2 Natural Frequencies and Response Loads If the subsystem is comparatively rigid in The natural frequencies up to 33 Hz for the relation to toe supporting system, and also is reactor and control buildings are presented in rigidly connected to the supporting system, it Tables 3.7 2 through 3.7 5 for the fixed base is sufficient to include only the mass of the condition, subsystem at the support point in the primary system model. On the other hend, in case of a Enveloped response loads at key locations in subsystem supported by very flexible the reactor building complex due to OBE for the connections, e.g., pipe supported by hangers, range of site conditions considered in Appendix the subsystem need not be included in the 3A are presented in Appendix 3G , Response primary model. In most cases the equipment and spectra at the major equipment elevations and compoeients, which come under the definition of support points are also given in Appendix 3G. subsystems, are analyzed (or tested) as a dect,upled system from the primary structure and The SSE loads are two times the OBE loads as the seismic input for the former is obtained by explained in Subsection 3.7.1.2. the analysis of the latter. One important exception to this procedure is the reactor 3.7.23 Procedure Used for hiodeling coolant system, which is considered a subsystem but is usually analyzed using a coupled model of 3.7.2.3.1 hiodeling Techniques for Systems the reaetor coolant system and primary Other Han Reactor Pressure Vessel strueture. An importan' step in the seismic analysis of In the second method of modeling, the systems other than the reactor pressure vessel is structure of the system is represented as a two-O the procedure used for modeling. The techniques or three-dimensional finite element model using center around two methods. The first method, the combinations of beam, plate, shell, and solid system is represented by lumped masses and a set elemt nts. The details of the mathematical of spring dashpots idealizing both the inertial models are determined by the complexi:y of the and stiffness properties of the system. The actual structures and the information required details of the mathematical models are determined for the analysis. by the complexity of the actual structures and the information required for the analysis. For 3.7.23.2 hlodeling of Reactor Pressure Vessel the decoupling of the subsystem and the and Internals supporting system, the following criteria equivalent to the SRP requirements are used: The seismic loads on the RPV and reactor internals are based on coupled dynamic analysis (1) If Rm 10.01, decoupling can be done for with the reactor building. The mathematical any R t. model of the RPV and internals is shown in Figure 3.7-32. This model is coupled with the (2) If 0.011 Rm 10.1, decoupling can be donc reactor building model for this analysis. if R(10.8 or Rr .> 1.25. The RPV and internals mathematical model (3)IfRm > 0.1, an approximate model of 2e consists of lumped masses connected by clastic subsystem should be included in the prim:re beam element members. Using the clastic proper-system model. ties of the structural components, the stiffness properties of the model are dein:nmed and 1:se Where R m and Rf are defined 4. effects of axial bending and shear are included. Rm= Total mass of the supported system / Mass points are located at all points of hlass that supports the subsystem critical interest such as anchors, supports, Amendment 1 3.79

AMM 23A6100AE Standard Plant anv.A points of discontinuity, etc. In addition, mass then obtaining its natural frequencies and mode points are chosen so that the mass distribution shapes. The dynamic response at the mass points g in various zories is uniform as practicable and is subsequently obtained by using a time history w the full range of frequency of response of inte- approach. rest is adequately represented. Further, in order to facilitate hydrodynamic mass calcula. Using the acceleration time history response tions, several mass points (fuel, shroud, vessel) of a particular mass point, a spectrum response are selected at the same elevation. Tbc RPV and curve is developed and incorporated into a internals are quite stiff in the vertical direc- design acceleration spectrum to be utilized for tion. Vertical modes in the frequency range of the seismic analysis of equipment located at the interest are adequately obtained with few dyrsmic mass point. Horizontal and vertical response degrees of freedom. Therefore, vertical masses spectra are computed for various damping values are distributed to a few key nodal points. The applicable for OBE and SSE evaluation of various length of control rod drive housing are equipment. Two orthogonal horizontal and one grouped in to the two representative lengths vertical earthquakt component are input shown in Figure 3.7-37. These lengths represent separately. Response spectra at selected the longest and shortest housing in order to locations are then generated for each earthquake adequately represent the full range of frequency component separately. They are combined using response of the housing 3. the square root of the-sum of the squares (SRSS) method to predict the total co-directional floor Not included in the mathematical model are the response spectrum for that particular stiffness properties of light components, such as frequency. This procedure is carried out for in core guide tuber and housings, sparger, and each site soil case used in the soil structure their supply headers. This is done to reduce the interaction analysis. Response spectra for all complexity of the dynamic model. For the seismic site-soil cases arc finally combined to arrive I responses of these components, floor response at one set of final response spectra. l spectra generated from system analysis is used. An alternate approach to obtain co direc. The presence of a fluid and other structural tional floor response spectra is to perform g' components (e.g., fuel within the RPV) introduces dynamic analysis with simultaneous input of a dynamic coupling effect. Dynamic effects of various earthquake components if those water enclosed by the RPV are accounted for by components are statistically independent to eacl- i introduction of a hydrodynamic mass matrix which other. l will serve to link the acceleration terms of the I equations of motion of points at the same The C*c. floor response spectra are obtained i elevation in concentric cylinders with a fluid by doubling the OBE response spectra as entrapped in the annulus. The details of the explained in Subsection 3.7.1.3. hydrodynamic mass derivation are given in Reference 4. The response spectra values are computed as a minimum either at frequency intervals as 3.7.2A Soll Structure Interaction specified in Table 3.7.1-1 of SRP 3.7.1 or at a set of frequencies in which each frequency is The soil model and soil structure interaction within 10% of the previous coe. analysis are described in Appendix 3A. 3.7.2.6 nree Components of Earthquake Motion 3.7.2.5 Development of Floor Response Spectra The three components of earthquake motion are in order to predict the seismic effects on considered in the building seismic analyses. To equipment located at various elevations within a properly account for the responses of systems structure, floor response spectra are developed subjected to the three-directional excitation, a using a time-history analysis technique. statistical combination is used to o; tain the net response according to the SRSS criterion of The procedure entails fitst developing the Regulato y Guide 1.92. The SRSS method accounts mathematical model assuming a linear system and for the randomness of magnitude and direction of Amendment 1 3.7-10

l MN 23A6100AE Standard Plant REV,A earthquake motion. The SRSS criterion, applied the maximum acceleration range having the to the responses associated with the three same amplification factor as the most components of ground earthquake motion, is used strongly amplified. for seismic stress computation for steel structural design as well as for resultant (2) The time history used to calculate the floor seismic member force computations for reinforced response spectra produces a ground response concrete structural design, which envelopes the design ground response spectra. In order to do this, it has 3.7.2.7 Combination of Modal Response spectral peaks which are substantially higher than the design spectra. Since only the time history method is used for seismic system analysis, the response spectrum (3) The bui! ding and soil damping values used in combination of modal responses is not applied, the analpis a'e near the lower bound of the available damping data. The actual values 3.7.2.8 Interaction of Non Category I of damping are expected to be much higher Structures with Seismic Category I Structures than the values used in the analysis. The interfaces between Seismic Category I and (4) The yield strengths used in the analysis are non Seismic Category I structures and plant based on the minimum values and are equipment are designed for the dyn mic loads and considerably lower than expected values, displacements produced by both the Category I and non Category I structures and plant equipment. (5) The additional strength and damping that is All non Category I structures will meet any one available when materials are stressed beyond of the following requirements: yield are neglected when using linear clastic analytical methods. (1) The collapse of any non Category I structure will not cause the non Category I structure (6) The working stresses for most equipment are ( to strike a Seismic Category I structure usually considerably below the yield component. stresses. 1 (2) The collapse of any non Category I structure (7) The calculated natural frequencies of ' will not impair the integrity of Seismic equipment are usually lower than actual l Category I structures or components because of conservative modeling l assumptions. l (3) The non-Category I structures will be analyzed and designed to prevent their These elements of conservatism are in series 1 failure under SSE conditions in manner such (i.e., they are compounded), which results in an j that the margin of safety of these structures extremely conservative design. The only reasor. l is equivalent to that of Seismic Category I for broadening the spectra at all is to account structures. for the unlikely possibility that a particular piece of equipment might have a na' ural 3.7.2.9 Effects of Parameter Variations on frequency which is not on the calculated Floor Response Spectra spectral peak but is on the real peak. The following conservative assumptions are Since the peaks characteristic of the low included in the calculation of the floor response damping response are narrow, such an occurrence spectra: is extremely improbable. Even if this eventuality does occur, the extreme conservatism (1) The expected actual earthquake time histories described above ensures seismic adequacy of are enveloped by a smooth ground response equipment design. Further, the floor response spectrum for design use. The smooth curve spectra obtained from the time history analysis leads to conservative effects on modal of the. building are broadened plus and minus 10% analysis because it treats all the modes in in frequency. Alternatively, peak shif ting Amendment 1 3.7 11

ABWR ux6iooxit Standard Plant mma method of ASME Code Case N 397, as permitted by 3.7.2.13 Methods for Seismic Analysis of Regulatory Guide 1.84, Revision 24, is used. Category I Dams The broadening method of accounting for The analysis of all Category I dams, if variations causes modes having frequencies near applicable for the site, tsking into the spectral peaks to be calculated as though consideration the dynamic nature of forces (due they experience the peak acceleration. This is to both horizontal and vertical earthquake quite conservative because the spectra for the loadings), the behavior of the dam material actual structure have only one utrow peak under carthquake loadings, soif structure somewhere in the 20% broadened range. interaction effects, and nonlinear stress strain relations for the soil, will be used. Analysis 3.7.2.10 Use of Constant Vertical Static of earth filled dsms, if applicable, includes an Factors evaluation of deformations. Since all Seismic Category I structures and 3.7.2.14 Determination of Seismic Category I the RPV are subjected to a vertical dynamic Structure Overturning Moments analysis with a time hi~. tory defining the input, no constant vertical sta:ic factors are utilized. Seismic loads are dynamic in nature. The method of calculating seismic loads with dynamic 3.7.2.11 Methods Used to Account for Torsional analysis and then treating them as static loads Effects to evaluate the overturning of structures and foundation f ailures while treating the Torsional effects for two-dimensional analyt- foundation materials as linear clastic is ical models are accounted for in the following conservative. Overturning of the structure, manner. The locations of the center of mass are assuming no soil slip failure occurs, can be calculated for each floor. The centers of rigid- caused only by the center of gravity of the ity and rotational stiffness are determined for structure moving far enough horizontally to each story. Torsion effects are introduced in cause instability, each story by applying a rotational moment about g its center of rigidity. The rotational moment is Furthermore, when the combined effect of calculated as the sum of the products of the in- carthquake ground motion and structural response ertial force applied at the center of mass of is strong enough. the structure undergoes a eaci. fbor above and a moment arm equal to the rocking motion pivoting about either edge of the distance from the center of mass of the floor to base. When the amplitude of rocking motion the center of rigidity of the story plus five becomes so large that the center of structural percent of the maximum building dimension at the mass reaches a position right above either edge level under consideration. To be conservative, of the base, the structure becomes unstable and the absolute values of the moments are used in may tip over. The mechanism of the rocking the sum. The torsional moment and story shear motion is like an inverted pendulum and its are distributed to the resisting structural ele- natural period is long compared with the linear, ments in proportion to each individual stiffness. elastic struc tural response. Thus with regard to overturning, the structure is treated as a The RPV model is axisymmetric with no built in rigid body. eccentricity, llence, the torsional effects for the RPV are only those associated with the The maximum kinetic energy can be conserva-reactor building model, tively estimated to be: 3.7.2.12 Comparison of Responses E=1 s E mj (v}{) 2 + (vy) 2 ' 2 , (3.7-8) Since only the time history method is used for structural analysis, the responses obtained from where (vit) and (vy)are the maximum values of response spectrum and time-history methods are the total lateral velocity and total vertical not compared. velocity, respectively, of mass m;. Arnendment t 3.7-12

ABM ux61oorn Standard Plant any.A Values f or (vn)j and (vy)j are (2) An eigenvalue analysis of the linear system p computed as f o 11 o w s : model is performed. This results is the v eigenvector matrices (p() which are (vH) 2 = (vx ) 2 + (vH) 2 normalized and satisfy the orthogonality 8 8 8 (3.7-9) conditions: (.t.712) (vy) 2 = (vz ) 2 + (vy) 2 2 8 4[8 Kpj = w , and 4T g4; 8 8 (3.7 10) 8 8

                                                                                            = 0 for i$j where (vH)g and (vy)g are the peak                     where horizontal and vertical ground velocity, respectively, and (vx )[ and (vz); are the             K         = stiffness matrix; maximum values of the relative lateral and vertical velocity of mass m .                          N         = circular natural frequency asso-ciated with mode i; and Letting mobe total mass of the structure and base mat, the energy required to overturn the      T 4I structure is equal to                                            = transpose of ith mode eigen-vectordi Eo=mogh                             (3.7 11)

Matrix p contains all translational and where h is the height to which the center of mass rotational coordinates. of the structure must be lifted to reach the overturning position. Because the structure may (3) Using the strain energy of the individual act be a symmetrical one, the value of h is components as a weighting function, the computed with respect to the edge that is nearer following equation is derived to obtain a to the center of mass. The structure is defined suitable damping ratio ($;) for mode i. '

   ]

t as stable against overturning when the ratio Eo to E sexceeds 1.5. N (3.7 13) Si = 4 1 Cj ([8 Kpi)j These cakulations assume the structure rests v on the ground surface, hence, are conservative i j=1 becau:e the structure is actually embedded to a considerable depth. The embedded effect is where considered only when the rptio Eoto Es is less than 1.5. Sj = modal damping coefficient for ith mode; 3.7.2.15 Analysis Procedure for Damping N I

                                                                      = total number of structural In a linear dynamic analysis using a modal                       elements;                                ,

superposition approach, the procedure to be used  ! to properly account for damping in different 4; =compoaeat of ith mode  ! elements of a coupled system modelis as follows: eigenvector corresponding to jth , element; I (1) The structural percent critical damping of l the various structural elements of the model T = Transpose ofpi elined d above; is first specified. Each value is referred 4 8 to ularascomponent the damping ratio which (Cj)ibutes to theof a partic-contr = percent critical damping Cj , complete stiffness of the system. associated with element j; f k Amendment 1 3.7-13

ABWR 33umn Standard Plant RIM A K = stiffness matrix of element j; and described in Subsection 3.7.2.1.1 generates timchistories at various support elevations for g w = circular natural frequency of mode use in the analysis of subsystems and W

i. equipment. The structural response spectra curves are subsequently generated from the time 3.7.3 Seismic Subsystem Analysis history accelerations.

3.7.3.1 Seismic Analpis Methods At each level of the structure where vital components are located, three orthogonal This subsection discusses the methods by which components of floor response spectra, two Seismic Category I subsystems and components are horizontal and one vertical, are developed. The qualified to ensure the functional integrity of floor response spectrum is smoothed and the specific operating requirements which envelopes all calculated response spectra from characterize their Seismic Category I different site soil conditions. The response designation. spectra are peak broadened plus or minus 10% When coraponents are supported at two or more in general, one of the following five methods elevations, the response spectra of each of seismically qualifying the equipment is chosen elevation are superimposed and the resulting based upon the characteristics and complexities spectrum is the upper bound envelope of all the of the subsystem: individual spectrum curvcs considered. (1) dynamic analysis; For vibrating systems and their supports, multi degree of.frect*om models are used in (2) testing procedures; accordance with the lumped. parameter modeling techniques and normal mode theory described in (3) equivalent static load method of analysis; S u b s e c t io n 3.7.2.1.1. Piping analysis is d e scribe d in S ubse ction 3.7.3.3.1. (4) a combination of (1) and (2); or g (5) a combination of (2) and (3). Category I subsystems and components, all the loads normally acting on the equipment are Equivalent static load method of subsystem simulated during the test. The actual mounting analysis is described it. Subsection 3.7.1.5. of the equipment is also simulated or duplicated. Tests are p;rform:d by supplying Appropriate design response spectra (OBE and input accelerations to the shake table to such SSE) are furnished to the manufacturer of the an extent that generated test response spectra equipment for seismic qualification purposes. (TRS) ervelope the required response spectra. Additional information such as input time history is also supplied only when necessary. For certain Seismic Category I equipment and components where dynamic testing is necestary to When analysis is used to qualify Seismic ensure functional integrity, test performance Category I subsystems and components, the data and results reflect the following: analytical techniques must conservatively account for the dynamic nature of the subsystems or (1) performance data of equipment which has been components. Both the SSE and OBE, with their subjected to dynamic loads equal to or difference in damping values, are considered in greater than those experienced under th-the dynamic analytis as explained in Subsection specified seismic conditions; 3.7.1.3. (2) test data frorr. previously tested comparable The general approach employed in the dynamic equipment which has been subjected under analysis of Seismic Category I equipment and similar conditions to dynamic loads equal to component design is based on the response or greater than those specified; and spectrum technique. The time. history technique Amendment 1 3.7 14

MM 2 M 6100AE Standard Plant any. A (3) actual testing of equipment in accordance with one of the methods described in (1) the fundamental frequency and peak seismic Os Subsection 3.9.2.2 and Section 3.10. loads are found by a standard seismic analysis (i.e., from eigen extraction and 3.7.3.2 Determination of Number of Earthquake forced response analysis); Cycles (2) the number of cycles which the component 3.73.2.1 Piping experiences are found from Table 3.7 6 according to the frequency range within Fifty (50) peak OBE cycles are postulated for which the fundamental frequency lies; and fatigue evaluation. , (3) for fatigue evaluation, one half percent 3.7.3.2.2 Other Equipment and Components (0.005) of these cycles is conservatively assumed to be at the peak load, and 4.5% Criterion ll.2.b of SRP Section 3 7.3 rccom- (0.045) at the three quarter peak. The mends that at le=t one sale shutdown earthquake remainder of the cycles have negligible (S5E) and five operating basis carthquakes (OBEs) contribution to fatigue usage, should be assumed during the plant life. It also recommends that a minimum of 10 maximum stress The SSE has the highest level of response, cycles per earthquake should be assumed (i.e.,10 However, the encounter probability of the SSE is cycles for SSE and 50 cycles for OBE). For so small that it is not necessary to postulate equipment and components other than piping,10 the possibility of more than one SSE during the peak OBE stress cycles are postulated for fatigue 60 year life of a plant. Fatigue evaluation due evaluation based on the following justification. to the SSE is not necessary since it is a faulted condition and thus not required by AShiE To evaluate the number of cycles engendered by Code Section III. a given earthquake, a typical Boiling Water Reac-p tor Building reactor dynamic model was excited by The CBE is an upset condition and is included d three different recorded time histories: hiay 18, in fatigue evaluations according to AShiE Code 1940, El Centro NS component,29.4 sec; 1952, Section Ill. Investigation of seismic histories Taft N69 W component,30 sec; and hiarch for many plants show that during a 60 year life 1957, Golden Gates 89'E component,13.2 sec. it is probable that five earthquakes with The modal response was truncated so that the intensities one tenth of the SSE intensity, and respon e of three different frequer.cy bandwidths one earthquake approximately 20% of the proposed could be studied,0+ to 10 Hz,10 to-20 Hz, and SSE intensity, will occur. The 60 year life 20 to-50 Hz. This was done to give a good corresponds to 40 years of actual plant approximation to the cyclic behavior expected operation divided by a 67% usage ftctor. To from structures with different frequency content. cover the combined effects of these earthquakes and the cumulative effects of even lesser Enveloping the results from the three earth- earthquakes,10 peak OBE stress cycles are quakes and averaging the results from several postulated for fatigue evaluation. different points of the dynamic model, the cyclic behavior given in Table 3.7-6 was formed. 3.733 Ptocedure Used for Afodeling Independent of earthquake or component 3.7.33.1 hiodeling of Piping Systems frequency,99.5% of the stress reversals occur below 75% of the maximum stress level, and 95% of 3.7.33.1.1 Summary the reversals lie below 50% of the maximum stress level. To predict the dynamic response of a piping system to the specified forcing function, the in summary, the cyclic behavior number of dynamic model must adequately account for all fatigue cycles of a component during a earthquake significant modes. Careful selection must be is found in the following manner: e,ade of the proper response spectrum curves and Amendaient 1 3.7 15

1 1 ABWR m6iom Standard Plant anv2 I proper location of anchors in order to separate Penetration assemblies (head fittings) are Seismic Category I from non. Category I piping

systems, also very stiff compared to the piping system and are assumed to act as an anchor. The g

stiffness matrix at the attachment location 3.7J3.1.2 Selection of Mass Points of the process pipe (i.e., main steam, RilR supply or RilR return) head fitting is When performing a dynamic analysis, a piping sufficiently high to decouple the penetration system is idealized either as a mathematicsl assembly from the process pipe. Previous model consisting of lumped masses connected by analysis indicates that a satisf ar nry weightless clastic members or as a consistent minimum stiffness for this attachment point mass model. The clastic members are given the is equal to the stiffness in bending and properties of the piping system being analyzed. torsion of a cantilevered pipe section of the The mass points are carefully located to same size as the process pipe and equal in adequately represent the dynamic properties of length to three times the process pipe outer the piping system. A mass point is located at diameter, the beginning and end of every elbow or valve, at the extended valve operator, and at the For a piping system supported at more than intersection of every tee. On straight runs, two points located at different elevations in mass points are located at spacings no greater the building, the response spectrum analysis is than the span length corresponding to 33 lin A performed using the envelope response spectrum mass point is located at every extended mass to of all attachment points. Alternatively, the account for torsional effects on the piping multiple support excitation analysis methods may system. In addition, the increased stiffness and be used where acceleration time histories or mass of valves are considered in the modeling of response spectra are applied at all the piping a piping system. . attachment points. Finally, the worst single floor response spectrum selected from a set of 3.7.33.13 Selection of Spectrum Curses floor response spectra obtained at various in selecting the spectrum curve to be used for floers may be applied identically to all floors provided it envelops the other floor response h dynamic analysis of a particular piping system, a spectra in the set, curve is chosen which most closely describes the accelerations existing at the end points and 3.733.2 Modeling of Equipment restraints of the system. The procedure employed for decoupling the main steam and feedwater For dynamic analysis, Seismic Category I piping systems when establishing the analytical equipment is reprennted by lumped mass systems models to perform seismic analysis are as w hich consist of discrete masses connected by follaws: weightless springs. The criteria used to lump masses are: (1) The small branch lines (6. inch diameter and less) are decoupled from the main steam and (1) The number of modes of a dynamic system is j feedwater piping systems and analyzed controlled by the number of masses used; separately. therefore, the number of masses is chosen so that all significant modes are included. (2) The stiffness of all the anche's and its The modes are considered as significant if upporting steel is large enough to the corresponding naturai frequencies nre j effectively decouple the piping on either less than 33 Itz and the stresses calculated ' side of the anchor for analytic and code from tiese mooes are greater than 10% of the jurisdictional boundary purposes. The RPV is total stresses obtained from lower modes. l very stiff compared to the piping system and This approach is acceptable provided at thus, during normal operating conditions, the least 90% of the loading / inertia is RPV is also assumed to act ss an anchor. contained in the modes used. Alternately,  ; O' Amendnwnt 1 1716

MM 23A610M.E Standard Plant REV A the number o1 degrees of freedom are taken engineer. An additional examination of these s more than twice the number of modes with supports and restraining devices is made to s frequencies less than 33 liz. assure that their location and characteristics are consistent with the dynamic and statie (2) Mass is lumped at any point where a analyses of the system. significant concentrated weight is located (e.g., the motor in the analysis of pump 3.73.4 Basis of Selection of Frequencies motor stand, tne impeller in the analysis of pump shaft, etc). All frequencies in the range of 0.25 to 3311z are considered in the analysis and testing of (3) If the equipment has free end overhang span st uctures, systems, and componentc. These with flexibility significant compared to the frequencies are excited under the seismic center span, a mass is lumped at the overhang excitation, span. If the fundamental frequency of a component (4) When a mass is lumped between two supports, is greater than or equal to 33 Hz, .t is treated it is located at a point where the maximum as seismically rigid and analyted accordingly, displacement is expected to occur. This Frequencies less than 0.25 liz are not considered tends to lower the natural frequencies of the as they represent very flexible structures and equipment because the equipment frequencies are not encountered in this plant, are in the higher spectral range of the response spectra. Similarly, in the case of The frequency range between 0.25 liz and 33 Hz live loads (mobile) and a variable support covers the range of the broad band response stiffness, the location of the load and the spectrum used in the design, magnitude of support stiffness are chosen to yield the lowest frequency content for the 3.7.3.5 Use of Equivalent Static Load Methods system. This ensures conservative dynamic of Analysis loads since the equipment frequencies are such that the floor spectra peak is in the 3.7.3.5.1 Subsystems Other1han NSSS lower frequency range. if not, the modelis adjusted to give more conservative results. See Subsection 3.7.3.8.1.5 for equivalent static load analysis method. 3.7333 Field tecs'Jon of Supports and Restraints 3.73.5.2 NSSS Subsystems The field location of seismic supports and When the natural frequency of a structure of restraints for Seismic Category I piping and component is unknown, it may be analyzed by piping systems components is selected to satisfy applying a static force at the center of mass, the following two conditions: In order to conservatively account for the possibility of more than one significant dynamic (1) the location selected must furnish the mode, the static force is calculated as 1.5 required response to control strain within tims s the mass times the maximum spectral allowable limits; and acceleration from the floor response spectra of the point of attachments of multispan (2) adequate building strength and stiffness for structures. The factor of 1.5 is adequate for attachment of the component supports must be simple beam type structures. For other more dvailable. Complicated structures, the factor used is justified. The finallocation of seismic supports and re-straints for Seismic Category I piping, piping 3.73.6 Three Components of Earthquake Motion system components, and equipment, including the placement of snubbers, is checked against the The total seismic response is predicted by drawings and instructions issued by the combining the response calculated from the two O Amendmeet 1 3.7 17

ABM MA61MAE Standard Plant REv.A horizontal and the vertical analysis. N = number of modes considered in the analysis, g When the response spectrum method is used, the W method for combining the responses due to the Closely spaced modes are combined by taking three orthogonal components of seismic excitation the absolute sum of the such modes. is given as follows:

            ~               '

An alternate to the absolute sum method 3 1/2 presented in Regulatory Guide 1.92 is the R; = R3. I0IIO* IDS: E 8J (3.7-14) j=1 N 1/2 R= R2 + 22 lRI Rm! where 1 8 (3.7 16)

                                                                     .i= 1                     -

Rj = maximum, coaxial seismic response of interest (e.g., displacement, where the second summation is to be done on all moment, shear, stress, strain) in Iand m modes whose frequencies are closely directions i due to earthquake spaced to cach other. excitation in direction j, (j = 1, 2, 3). 3.73.7.2 NSSS Subsystems R; = seismic response of interest in i in a response spectrum modal dynamic direction for design (e.g., analysis, if the modes are not closely spaced displacement, moment, shear, (i.e., if the frequencies differ from each other stress, strain) obtained by the by more than 10% of the lower frequency), the SRSS ule to account fc,r the modal responses r.re combined by the nonsimultaneous occurrence of the square root-of-the sum of- the-squares (SRSS) R;j's. method as described in Subsection 3.7.3.7.1 and Regulatory Guide 1.92. g 3.73.7 Combination of Modal Response if some or all of the modes are closely 3.7.3.7.1 Subsystems Otherihan NSSS spaced, a double sum method, as described in Subsection 3.7.3.7.2.2, is used to evaluate the When the response spectrum method of modal combined response, in a time history method of analysis is used, contributions from all modes, dynamic analysis, the vector sum of every step except the closely spaced modes (i.e., the is used to calculate the combined response. The difference between any two natural frequencies is use of the time history analysis method equal to or less than 10%) are combined by the precludes the need to consider closely spaced square root-of the sum of-the-squares (SRSS) modes, combination of modal responses. This is defined mathemrtically as: 3.7.3.7.2.1 Square Root-of the Sum-of the-Squares Method N R= ( R;) 2 Mathematically, this SRSS method is expressed E (3.7 15) as follows: i=1 where R= fN ( R;) 2)1/2 R = combined response; i E ' (3.7 17) i-i / R; = response to the ith mode; and 9 Amenoment 1 3.7 18

ABM u^62oo^n Standard Plant __ REV.A where where wk and #k are the modal frequency q and the damping ratio in the kth mode, V R = combined response; respectively, and ta is the duration of the Rj = response to the ith mode; and 3.7.3.8 Analytical Procedure for Piping N = number of modes considered in the analysis. 3.73.8.1 Piping Subsystems Other Than NSSS 3.7J.7.2.2 Double Sum Methol. 3.73J.1.1 Qualineation by Analysis This method, la defined in Regulatory Guide The methods used in seismic analysis vary 1.97, is mathematically: according to the type of subsystems and supportit.g structure involved. The following possible cases are defined along with the rN N 1 /2 1 associated analytical methods used. R= l E E lRk Rs ! (ks/I ik=1 s=1 (3.7 18) 3.73J.1.2 Rlgid Subsystems with Rigid Supports where If all natural frequencies of the subsystem R = representative maximum value of a are greater than 33 Hz, the subsystem is particular response of a given considered rigid and analyzed statically as element to a given component of such. In the static analysis, the seismic excitation; forces on each component of the subsystem are - obtained by concentrating the mass at the center Rk = Peak value of the response of the of gravity and multiplying the mass by the element due to the kt mode; appropriate maximum floor acceleration. N = n u m b e r o f sig n i fie a n t m o d e s 3.73.8.1J Rigid Subsystems with Flexible considered in the modal response Supports combination; and if it can be shown that the subsystem itself R3 = peak value of the response of the is a rigid body (e.g., piping supported at only element attributed to sth mode two points) while its supports are flexible, the overall subsystem is modeled as a single degree. where of freedom subsystem consisting of an effective mass and spring. (J k- W's) L1 2

 'ks = 1+                                                        The natural frequency of the subsystem is         ;

(#k Wk + #s Ws )/ computed and the acceleration determined from (3.7 19) the floor response spectrum curve using the appropriate damping value. A static analysis is in which performed using 1.5 times the acceleration value. In lieu of calculating the natural 1/2 frequency, the peak acceleration from the wk

• Wk 1-4 2 spectrum curve may be used.

k 1 If the subsystem has no definite orientation, i 2 Sk " Sk + the excitation along each of three ruutually I td Wk perpendicular axes is aligned with respect to l the system to produce maximum loading. The l O 4 l l Amendment 1 3.7 19

ABM ursioorn Standard Plant arv. A excitation in each of the three axes is considered to act simultaneously. The N 1/2 excitations are combined by the SRSS method. R= I R2 + 2Z l Rf R m I b 8 (3.7 20) 3.7.3.8.1.4 Flexible Subsystems \i=1 I If the piping subsystem has more than two where the t.econd summation is to be done on all supports, it cannot be considered a rigid body 1 and m modes whose frequencies are closely and must be modeled as a multi-degree-of freedom spaced to each other, subsystem. and where The subsystem is modeled as discussed in Subsection 3.7.3.3.1 in sufficient detail (i.e., Rj = response to the ith mode number of mass points) to ensure that the lowest natural frequency between mass points is greater N = number of significant modes than 33 Hz. The mathematical modelis analynd considered in the modal response using a time history analysis technique or a combinations. response spectrum analysis approach. After the natural frequencies of the subsystem are The excitation in each of the three reajor obtained, a stress analysis is performed using orthogonal directions is considered to act the inertia forces and equiva ent static loads simultaneously with their effect combined by the obtained from the dynamic analysis for each mode. SRSS method. For a response spectrum analysis based on a 3.7.3.8.1.5 Static Analysis modal superposition method, the modal response accelerations are taken directly from the A static analysis is performed in lieu of a spectrum. The total seismic stress is normally dynamic analysis by applying the following obtained by combining the modal stress using the forces at the concentrated mass locations SRSS method. The seismic stress of closely (nodes) of the analytical model of the piping g spaced modes (i.e., within 107c of the adjacent system: mode) are combined by absolute summation. The resulting totalis treated as a pseudomode and is (1) horizontal static load, P h = C hW, in one then combined with the remaining modal stresses of the horizontal principal directions; by the SRSS method. (2) equal static load, F h, in the other The approach is simple and straightforward in horizontal principal direction; and all cases where the group of modes with closely spaced frequencies is tightly bundled (i.e., the (3) vertical static load, Fy = CvW; lowest and the highest modes of the group are within 10% of each other). However, when the where group of closely spaced modes is spaced widely over the frequency rangc of interest while the Ch, Cy = multipliers of the gravity frequencies of the adjacent modes are closely acceleration, g, determined spaced, the absolute sum method of combining from the horizootal and ver- i response tends to yield over conservative tical floor response spectrum results. To prevent this problem, a general curves, respectively. (They l approech applicable to all modes is considered are functions of the period and j appropriate. The following equation is merely a the appropriate damping of the - mathematical representation of this approach. piping system); and The most probable system response, R, is given W = weight at node points of the by: analytical model. 9 - - , n l I

ABM 23461oorn Standard Plant nw A For special case analyses, Ch and Cy may N be taken as: M;pij (1) 1.0 times the zero-period acceleration of the i=1 response spectrum of subsystems described in sj = Subsection 3.7.3.8.1.2; N Mg2 i , (3.7 21) 8) (2) 1.5 times the value of the response spectrum at the determined frequency for subsystems [ i=1 , described in Subsection 3.7.3.8.1.3 a nd 1.7.3.8.1.4; and where (3) 1.5 times the peak of the response spectrum for subsystems described in Subsections Mi = ith mass 3.7.3.8.1.3 an d 3.7.3.8.1.4. gij = component of An alternate method of static analysis which e a r t h q u a k ection dir e$ ij, in t h e allows for simpler technique with added conserva-tism is acceptable . No determination of natural g'ij = ith characteristic displacement frequencies is made, but rather the response of in the jth mode the subsystem is assumed to be the peak of the appropriate response spectrum at a conservative sj = modal participation factor for and justifiable value of damping. The response the jth mode is then multiplied by a static coefficient of 1.5 to take into account the effects of both N = number of masses. multifrequency excitation and multimodel l response. (5) Using the appropriate response spectrum curve the spectral acceleration, ra , for 3.7.3.8.1.6 Dpamic Analysis the jl mode as a function of the jth mode natural frequency and the damping of The dynamic analysis procedure using the the system is determined. response spectrum method is provided as follows: (6) The maximum modal acceleration at each mass (1) The number of node points and members is point, i, in the model is computed as indicated. If a computer program is follows: utilized, use the same order of number in the computer program input. The mass at each aij = sj ar j$ij (3.7 22) node point, the length of each member, clastic constants, and geometric properties are determined, where (2) The dynamic degrees of freedom according to aij = acceleration of the ith mass the boundary conditions are determined. point in the jth mode. 1 (3) The dynamic properties of the subsystem (7) The maximum modal inertia force at the ith ) (i.e., natural frequencies and mode shapes) mass point for the jth mode is calculated  ! are computed, from the equation: (4) Using a given direction of earthquake motion, Fi j = Mi aij (3.7 23) j the modal participation factors, sj, for each mode are calculated: (8) For each rnode, the maximum inertia forces i O Amendment t 3.7 21

ABWR -- Standard Plant REV.A are applied to the subsystem model, and the modal into the piping system. The stress thus pro-forces, shears, moments, stresses, and duced is a secondary stress. It is justifiable g W deflections are determined, to place this stress, which results from l restraint of free-end displacement of the piping (9) The modal forces, shears, moments, stresses, system,in the secondary stress category becaua and deflections for a given direction are the stresses are self limiting and, when the combined in accordance with Subsection stresses exceed yield strength, minor 3.7.3.8.1.4. distortions or deformations within the piping system satisfy the condition which caused the (10) Steps (5) through (9) are performed for each stress to ocent. of the three earthquake directions. The earthquake thus produces a stress-(11) The seismic force, shear, moment, and stress exhibiting property much like a thermal resulting from the simultaneous application expansion stress and a static analysis can be of the three components of earthquake used to obtain actual stresses. The I loading are obtained in the following differential displacements are obtained from the i manner: dynamic analysis of the building. The displacements are applied to the piping anchors j R R2+R2+R2 (3.7 24) and restraints corresponding to the maximum x y z differential displacements which could occur. The static analysis is made three times: once R = Lquivalent seismig for one of the horizontal differential response quantity (force, displacements, once for the other horizontal shear, moment, stress, differential displacement, and once for the etc.) vertical. RxRyRz = c o1in e a r response 3.7.3.8.2 NSSS Piping Subsystems q u a n t itie s due to earthquake motion in the 3.7.3.8.2.1 Dynamic Analysis g x, y, and z directions, respective' . As described in Subsection 3.7.3.3.1, pipe line is idealized as a mathematical model 3.7J.8.1.7 Damping Ratio consisting of lumped masses connected by clastic members. The stiffness matrix for the piping The damping ratio percentage of critical damp- subsystem is determined using the clastic ing of piping subsystems corresponds to Regula- properties of the pipe. Thi. includes the tory Guide 1.61 or 1.84 (ASME Code Cate N-4111). effects of torsional, bending, shear, and axial The damping ratio is specified in Table 3.71. deformations as well as changes in stiffness due to cuned members. ) 3.7.3.8.1.8 Effect of Differential Building Mosements Next, the mode shapes and the undamped natural frequencies are obtained. The dynamic In most cases, piping subsystems are anchored response of the subsystem is usually calculated and restrained to floors and walls of building, by using the response spectrum method of analy-that may have differential movements during a sis. When the connected equipment is supported seismic event. The movements may range from at more than two points located at different insignificant differential displacements between elevations in the building, the response spec-rigid walls of a common building at low eleva- trum analysis is performed using the envelope ) tions to relatively large displacements between response spectrum of all attachment points. separate buildings at a high seismicity site. Alternatively, the multiple excitation analysis q methods may be used where accele<ation time  ; Differential endpoint or restraint deflee- histories or response spectra are applied at all tions cause forces and moments to be induced the equipment and piping attachment points. Amendn, cat 1 3.7-22 1 i

r n W ABM 23A6100AE Standard Plant REV.A l 3.7.3.8.2.2 Effect of Differential Bullding adequately accounted for in the analysis. Mo ements In case of buried systems sufficiently flex. , Os ible relative to the surrounding or under- l The relative displacement between anchors is lying soil, it is assumed that the systems  ! determined from the dynamic analysis of the will follow essentially the displacements structures. The results of the relative anchor- and deformations that the soil would have if point displacement are used in a static analysis the systems were absent. When applicable, to determine the additional stresses due. to procedures, which take into account the relative anchor-point displacements. Further phenomena of wave travel and wave reflection details are given in Subsection 3.7.3.8.1.8. in compacting soil displacements from the ground displacements, are employed. 3.7.3.9 Multiple Supported Equipment Components With Distinct inputs (2) The effects of static resistance of the surrounding soil on piping deformations or The procedure and criteria for analysis are displacements, differential movements of described in Subsections 3.7.2.1.3 and piping anchors, bent geometry and curvature 3.7.3.3.1.3. changes, etc., are considered. When applicable, procedures utilizing the 3.7.3.10 Use of Constant Vertical Static principles of the theory of structures on Factors clastic foundations are used. All Seismic Category I subsystems and compo- (3) When applicable, the effects due to local nents are subjected to a vertical dynamic soil settlements, soil arebing, etc., are analysis with the vertical floor spectra or time also considered in the analysis. histories defining the input. A static analysis is performed in lieu of dynamic analysis if the 3.7.3.13 Interaction of 0ther Piping with peak value of the floor spectra is used in the Seismic Category I Piping p v analysis. Th vertical ground design response spectrum is used for equipment vertical seismic In certain instances, non Scismic Category I load determination if it can be shown that the piping may be connected to Seismic Category I structures supporting the equipment are rigid or piping at locations other than a piece er equip. quasi rigid in vertical ditection. A factor of ruent which, for purposes of analysis, could bc

• 1.5 it applied to the peak acceleration of the represented as an anchor. The transition points applicable floor response spectrum when the typically occur at Seismic Category I valves static analysis is used. which may or may not be physically anchored.

Since a dynamic analysis must be modeled from 3.7.3.11 Torsional Effects of Eccentric Masses pipe anchor point to anchor point, two options exist: Torsional effects of eccentric masses are included for Seismic Category I subsysterns (1) specify and design a structural anchor at similar to that for the piping systems disc'sssed the Seismic Category I valve and analyze the in Subsection 3.7.3.3.1.2. Seismic Category I subsystem; or, if I impractical to design an anchor, 3.7.3.12 Buried Seismic Category 1 Pipir.g and Tunnels (2) analyze the subsystem from the anchor point in the Seismic Category I subsystem through For buried Category I buried piping systems the valve to either the first anchor point and tunnels the following items are considered in in the non Seismic Category I subsystem; or the analysis: to sufficient distance in the non Seismic Category 1 Subsystem so as not to (1) The inertial effects due io an earthquake significantly degrade the accuracy of upon buried systems and tunnels will be analysis of the Seismic Category I piping. A V Amendment 1 3.7 23

ABWR = = = Standard J3rnt nrw. n Where small, non Scismic category piping is (6) recording and playback equipment; and directly attached to Seismic Category I piping, 3 its effect on the Seismic Category I piping is (7) annunciators. W accounted for by lumping a portion of its mass with the Seismic Category I piping at the point The location of seismic instrumentation is of attachment. outlined in Table 3.7 7. Furthermore, non Seismic Category I piping 3.7.4.2.1 Time.llistory Accelerographs (particularly high energy piping as defined in Section 3.6) is designed to withstand the SSE to Time-history accelerographs produce a record avoid jeopardizing adjacent Seismic Category I of the time varying acceleration at the sensor piping if it is not feasible or practical to location. This data is used directly for analy-isolate these two piping systems. sis and comparison with reference information trnd may be, by calculational methods, converted to 3.7.3.14 Seismic Analysis for Reactor response spectra form for spectra comparisons Internals with design parameters The modeling of RPV internals is discussed in Each triaxial acceleration sensor unit con-Subsection 3.7.2.3.2. The damping values are tains three accelerometers mounted in an ortho-given in Table 3.7-1. The seismic model of the gonal array (two horizontal and one vertical). RPV and internal is shown in Figure 3.7 32. All acceleration ut.its have their principal axes oriented identically. The mounted units are 3.7.3.15 Analysis Procedures for Damping oriented so that their axes are aligned with the building major axes used in development of the Analysis procedures for damping are discussed mathematical models for seismic analysis. in Subsection 3.7.2.15. One TilA is located on the reactor building 3.7.4 Seismic Instrumentation (RB) foundation mat, El (-) 13.2 M. at the base of an RB clean zone for the purpose of measuring g 3.7.4.1 Comparison with NRC Regulatory Guide the input vibratory motion of the foundation 1.12 mat. A second TIIA is located in an RB clean zone at El (+ ) 26.7 M on the same azimuth as the The seismic instrumentation program is foundation mat TilA. They provide date on the consistent with Regulatory Guide 1.12. frequency, amplitude, and phase relationship of the seismic response of the reactor building 3.7.4.2 Location and Description of structure. A third TilA is located in the free Instrumentation field at the finished grade approximately 160 M from any ststion structures with axes oriented in The following instrumentation and associated the same direction as the reactor building equipment are used to measure plant response to accelerometers. earthquake motion: Two seismic triggers, connected to form redun-(1) three triaxial time history accelerographs dant triggering, are provided to start the TilA (TilA); recording system. They are located in the free field at the finished grade 160 M from the reac-(2) three peak recording accelerographs (PRA); tor building. The trigger unit consists of or-thogonally mounted acceleration sensors that act. l (3) two triaxial seismic triggers; uate relays whenever a threshold acceleration is exceeded for any of the three axes. The trigger (4) one scismic switch (SS); is engineered to discriminate against false starts from other operating inputs such as traf-(5) fcor response spectrum recorders; fic, elevators, people, and rotating equipment. O Amendment 1 3 7-24

ABM 23461oosu Standard Plant RLV A I Magnetic tape recording and playback units are structures and equipment. Response spectra are l provided fro multiple channel recording and play- recorded for three mutually orthogonal  ! O back of the THA accelerometer signals. The data directions at the sensor location by inscribing recordings include an additional recorded channel steel reed deflections upon record plates. One l j for the timing reference signal generated in the recorder is located on the reactor building 1 control unit. The recording and playback systems foundation in a clean zone. Another recorder is  ! have a special cabinet furnished for those located on the control building foundation. If l instruments and devices necessary for system the OBE design response spectra values for l testing, annunciating, calibration, and control, specific frequencies are exceeded during an l This cabinet is located in the control equipment earthquake, specific switches mounted in the l room. recorders annunciate the specific frequencies in i the control equipment room. j 3.7.4.2.2 Peak Recording Accelerographs Two other recorders do not contain alarm Each seraor unit contains three peak recording contacts. One is mounted in the reactor accelerographs mounted in a mutually orthogonal building pipe tunnel on a 20-inch RHR line and j array. The units are unpowered and record peak another is on a FMCRD control panel support. I accelerations triaxially by proportional scratches on record plates. The PRAs that are 3.7.4.2.5 Recording and Playback Equipment mounted directly on equipment have one axis coincident with the principal equipment axis. A cabinet located in the control equipment All other PRAs have their principal axes oriented room houses the recording, playback, and identically with one horizontal axis parallel to calibration units that are used in conjunction the major horizontal axis assumed in the seismic with the THA sensors to produce a time history analysis. record of the carthquake. It also contains audible and visual annunciators wired to display One PRA is located on a reactor water cleanup initiation of the THA recorder and the power Q k/ unit (RWCU) regenerative heat exchanger support. supply components for all equipment contained A second PRA is located on an RHR pipe support, within the cabinet. A third PRA is located on a diesel generator support. 3.7.43 Control Room Operator Notification Data from PRAs must be manually retrieved Activation of the seismic triggers causes an following an earthquake and is used in the audible and visual annunciation in the main I detailed iuvestigations for particular control room to alert the plant operator that an i structures, systems, and equipment. earthquake has occurred. The annunciation is 1 set to occur at 0.01g vertical acceleration on ) 3.7.4.23 Seismic Switches the free field. l One triaxial seismic switch (SS) is installed The triggers cause initiation of the THA on the reactor building foundation. This device recording system at horizontal or vertical actuates a visual and audible annunciator in the acceleration levels slightly higher than the main control room when the OBE acceleration on at expected background level including induced least one of the axes has been exceeded. When vibrations from sources such as traffic, the threshold acceleration is sensed, the relay elevators, people, and machinery. The initial closes and remains closed for an adjustable set points may be changed once significant plant period after the threshold is no longer exceeded, operating data have been obtained which indicate that a different setpoint would provide better  ! 3.7.4.2.4 Response Spectrum Recorders THA system operation. l The response spectrum recorders measure both Audible and visual annunciators are provided horizontal and vertical peak acceleration for a in the main control room to indicate whether the scr!cs of frequencies pertinent to specific OBE floor accelerations have been exceeded fcr Amendment 1 3.7-15 1

ABM 23A6100AE l Standard Plant anv. A the seismic switch location. 3.7.4.5 In Senice Surveillance The peak acceleration level experienced by the Each of the seismic instruments will be l reactor building basemat is available immediately demonstrated operable by the performance of the  ; following the earthquake. This is obtained by channel check, channel calibration, and channel playing back the recorded THA data from the functional test operations at the intervals basemat location and reading the peak value from specified in Table 3.7 9. a strip chart recorder. 3.7.5 Interfaces Significant response spectra from the reactor building basemat are available immedistely 3.7.5.1 Seismic Parameters following an earthquake for comparison with the OBE and SSE response spectra. The design basis horizontal g value is 0.3g for SSE and 0.35g for OBE. These are maximum 3.7AA Comparison of Measured and Predicted free-field ground accelerations at the site as Responses measured at the existing grade level near the Nuclear Island. The response spectra are initial determination of the carthquake level presented in Subsection 3.7.1. The range of is performed immediately after the earthquake by site parameters used to establish the design comparing the rneasured response spectra from the basis seismic parameters is presented in , reactor building basemat with the OBE and SSE Appendix 3A.  : response spectra for the correspanding location. If the measured spectra exceed the OBE response 3.7.6 References spectra, the plant is shut down and a detailed ana?ysis of the earthquake motion is undertaken. 1. General Electric Company BWR/6-238 Standard Safety Analysis Report (GESSAR), Docket No. After any earthquake, the data from all STN 50-447, November 7,1975. seismic recorders and recording instruments are retrieved. When the OBE has been exceeded, the 2. E. H. Vanmarcke and C. A. Cornell, Seismic g data from these instruments are analyzed to Risk and Design Response Spectra, ASCE obtain the seismic accelerations experienced at Specialty Conference on Safety and the location of major Seismic Category I Reliability of Metal Structures, Pittsburgh, structures and equipment. The measured response Pennsylvania, November 1972. from the time history accelerographs, peak-recording accelerographs, and response spectrum 3. NUREG-0800, Standard Review Plan, Section recorders are used to determine the response 3.7.1, Draft of Resision 2. spectra at the location of each Seismic Category I structure and system. These spectra are 4. L. K. Liu, Seismic Analysis of the Boiling i compared with those used in the design to Water Rear'or, symposium on seismic l determine whether the structure or system is analysis of pressure vessel and piping I still adequate for future use. Peak recording components, First National Congress on l accelerographs mounted on equipment are used to Pressure Vessel and Piping, San Francisco, j determine whether the design limitation of that California, May 1971. ' specific equipment has been exceeded. The theoreticci structural response and mea-sured structural responses are compared to assess the degree of conservatism in the analytical pre-dictions. Seismic levels are established to de-termine whether the plant can continue to oper-ate or be shutdown. The criteria consider system design and dynamic analysis in establishing the acceptable levels for continued operation. Amendment 1 3.7-26

MM. 23A6100AE Standard Plant REV.A Table 3.71 ' DAMPING FOR DIFFERENT MATERIALS  ! Percent Critical Damping htm DE ESE Reinforced concrete structures 4 7 Welded strucural assemblies 2 4 Steel frame structures 2 3 Bolted or riveted structural assemblies 4 7 Equipment 2 3 piping systems diameter greater than 12 in. 2* 3 diameter less than or equal to 12 in. 1* 2 Reactor pressure vessel, support skirt, shroud head and separator 2 4 O Guide tubes and CRD bousings 1 2 Fuel 6 6 1 Damping values ofASME Code Case N-411 i, attemative damping Values for Response l l Spectra Analysis of Class 1, 2, and 3 Piping, Section 111, Division 1, may be . used as permitted by Regulatory Guide 1.84, Revision 24, June 1986. These damping values are applicable in analyzing piping response for Seismic and other l dynamic loads filtering through building structures in high frequencies range ' beyond 33 Hz. i l l Amendment 1 3.7 27 , i

Mkb 23A6100AE l Standard Plant RI?/. A i Table 3.7 2  ; NATURAL FREQUENCIES OF THE REACTOR BUILDING COMPLEX IN h" X DIRECTION (0 -180 AXIS) FIXED BASE CONDITION Mode No. Freauency (117) 1 3.97 2- 433 3 7.70 4 8.11 5 9.17 6 1137 7 13.6% 8 13 89 9 15.02 10 1531 11 15.79 12 16.26 13 16.82 14 18.00 15 19.73 16 20.42 17 21.0S 18 22.05 19 23.11 20 21 24.61 26.27 g 22 27.29 23 28.17 24 28.51 25 2938 26 31.10 27 32.Gt 28 32.22 29 32.58 9 Amendment 1 3.73

ABM 23461oorn 4 Standard Plant anv. A Table 3.7 3 O s NATURAL FREQUENCIES OF THE REACTOR BUILDING COMPLEX IN Y DIRECTION (90' 270 AXIS) - FIXED BASE CONDITION  : Mode No. Frequency (HZ) , i 1 321 2 4.52 3 7.03 , 4 7.65 , 5 7.73 i 6 8.65 7 1137 8 13.02 9 13.67 10 14.17 11 1532 12 15.91  ; 13 16.68 , 14 16.82 l 15 18.00 , . 16 19.25 17 19.74 l 18 21.24 , 19 22.14 , 20 23.75  ; 4 21 24.58 22 26.15 23 25.66 1 24 27.83 25 29.59 I . M M.% 2 27 31.10 28 31.63 29 32.22 1 1 'T l 1 i l;

O Amendment 1 3.7-29 I

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ABM ussioaan S_t_andard Plant any.A Table 3.7-4 NATURAL FREQUENCIES OF TIIE REACTOR BUILDING COMPLEX IN h Z DIRECTION (VERTICAL) FIXED BASE CONDITION Afode No. Freauency (11Z) 1 5.07 2 5.176 3 5.183 4 8.44 5 9.20 6 9.23 7 12.80 8 13.37 9 19.60 10 27.54 11 31.36 Table 3.7 5 NATURAL FREQUENCIES OF TIIE CONTROL O BUILDING - FIXED BASE CONDITION 31 ode No. Frecuency (11Z) Direction 1 3.71 X liORIZ 2 7.44 Y HORIZ 3 10.98 X llORIZ 4 16.21 Z VERT 5 19.87 Y llORIZ 6 22.70 X IIORIZ 7 27.33 X 110RIZ 8 34.77 X llORIZ S Amendment 1 3.7 30

ABM 2346iooxa , Standhrd Plant REV.A  ! Table 3.7 6 NUMBER OF DYNAMIC RESPONSE CYCLES EXPECTED DURING A SEISMIC EVENT FOR SYSTEMS & COMPONENTS i FREOUENCY BANDWIDTH (Hz) 0 + 10 1Q-1Q 29.E Total number of seismic cycles 168 359 643 No. of seismic cycles (0.5% of total) between 75% and 100% of peak loads 0.8 1.8 3.2 No. of seismic cycles (4.5% of total) between 50% and 75% of peak loads 7.5 16.2 28.9

                                                                                                                                                  -l l

O 1 I l O Amendment i 3.7 31 l l i

ABWR uumu Standard Plant attv.4 Table 3.7 7 DESCRIPTION OF SEISMIC INSTRUMENTATION h Setpoint Operating Comnonent Location Elevation * (s _) Eanar i Time history accelerometer Free field,160 ht from N/A - 0.01 to 1.0g sensor Reactor Building RB Time history accelerometer Reactor building founda- (-) 13.2 h1 - 0.01 to 1.0g sensor tion mat at base an RB clean zone Time history accele ometer At RB clean zone ( + ) 26.7 hi - 0.01 to 1.0g sensor Seismic trigger Free field,160 ht from N/A 0.01 025 to 0.02g Reactor Building Seismic trigger Free field, M0 hi from N/A 0.01 0.005 to 0.02g Reactor Bui ,g Peak recording accelero- Reactor Building, RWCU (-)6.7hi - 1 to 20 llz graph 'egenerative heat exchange support Peak recording accelero- Reactor Bu:! ding, RIIR - - I to 20 llz O graph line Peak recording accelero- Reactor Building. Diesel ( + ) 7.3 hi - 1 to 20 liz graph generator A support Seismic switch Reactor Building founda- (-) 13.2 h1 0.10 0.1 to 30.0 llz tion Response spectrum recorder, Reactor Building founda. (-) 13.2 hl Table Table 3.7 8 (acthe) tion mat at the base of an 3.78 Ri1 clean zone Response spectrum recorder, Control Building found- (-) 3.5 h1 Table Table 3.7-8 (acthr) tion mat 3.78 Response spectrum recorder, Reactor Building pipe - - 1.0 to 32 Itz (passhr) tunnel RilR hanger Response spectrum recorder, Reactor Building FhtCRD ( + ) 18.7 M - 1.0 to 32 llz (passive) control panel support Seismic event - recording Control equipment room - - - alarm, playback panel

• Elevations are with respect to the RPVbottom head.

Amendment 1 3.7 M L

MM DA6100AE standard Plant anv. A Table 3.7-8 ) O ser voisrs ron acrivs atseosse setcraum asconosas i Setpoint @ Operating Range Horizontal Vertical Freauency (Hz) 0.19 0.13 1.00 0.23 0.16 1.26 0.29 0.20 1.59 035 0.25 2.00 0.48 030 2.52 0.41 037 3.17 0.40 039 4.00 039 038 5.04 037 037 635 036 036 8.00 032 032 10.01 0.26 0.26 12.07 0.21 0.21 20.00 0.17 0.17 20.20 0.13 0.13 24.40 0.10 0.10 32.00 0 O Amendment 1 3.7 33

ABWR m6ian l Standard Plant RIN. A j Table 3.7 9 l SEIS511C h10NITORING INSTRUh1ENTATION SURVEILLANCE REQUIREN1ENTS 9' 1 CllANNEL I CIIANNEL CllANNEL FUNCTIONAL I INSTRUh1ENT CIIECKa CAllHRATIONu TESl*

1. TriaxialTime History Accelerographs hi R SA

2. Triaxial Peak Accelerographs NA R NA

3. Triaxial Scismic Switches hi R SA

4. Triaxial Responsc Spectrum Recorders h1 R SA ahl = hionthly l R = Refueling SA = One per 18 months NA = Not Applicable l

l O Amendment 1 3,7.M

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2) THE ROTATIONAL SPRING BETWEEN NODES 90 AND 88 IS PRESENTED i ONLY IN THE X-Z PLANE Figure 3,7-31 REACTOR BUILDING MODEL 87 592 57 Amendment 1 3716

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l ABWR ua6imau Standard Plant anv.A SECTION 3.8 CONTENTS Section Title P.ase 3.8.1 Concrete Containment 3.81 3.8.1.1 Description of the Containment 3.8-1 3.8.1.1.1 Concrete Containment 3.8-1 3.8.1.1.2 Containment Liner Plate 3.8-2 3.8.1.2 Applicable Codes, Standards, and Specifications 3.82 3.8.1.2.1 Regulations 3.8-2 3.8.1.2.2 Construction Codes of Pract'.cc 3.8-2 3.8.1.23 General Design Criteria, Regulatory Guides, 3.82 and Industry Standards 3.8.1.2.4 Containment Boundary 3.8-2 3.8.13 Loads and Load Combinations 3.8-3 3.8.13.1 Normal Loads 3.8-3 3.8.13.2 PreoperationalTesting Loads 3.8-4 3.8.133 Severe Emironmental Loads 3.8-4 3.8.13.4 Extreme Emironmental Loads 3.8-4 3.8.13.5 Abnormal Plant Loads 3.84 1 3.8.13.6 Load Combinations for the Containment 3.8-5 I Structure and Liner Plate 3.8.1.4 Design and Analysis Procedures 3.8-5 3.8.1.4.1 Containraent Cylindrical Wall, Top Slab 3.85 and Foundation hiat 3.8.1.4.1.1 Analytical hiethods 3.8-5 3.8.1.4.1.1.1 Nonaxisymmetrical Loads 3.85 0 3.8-il Amendment 1

1 l ABWR uuimau l Standar l_ Plant RIN. A ) SECTION 3.8  ! CONTENTS (Continued) O Section Title Page l 3.8.1.4.1.1.2 Axisymmetrical Loads 3.8-6 3.8.1.4.1.1.3 hiajor Penetrations 3.8-7 3.8.1.4.1.1.4 Variation of Physical biaterial Properties 3.8-7 1 3.8.1.4.1.2 Design hiethods 3.87 3.8.1.4.13 Concrete Cracking Considerations 3.87 3.8.1.4.1.4 Corrosion Prevention 3.8-8 3.8.1.4.2 Ultimate Capacity of the Containment 3.8-9 3.8.1.5 Structural Acceptance Criteria 3.8-9 3.8.1.6 hiaterial, Quality Control, and Special Construction Techniques 3.89 3.8.1.6.1 Concrete 3.89 3.8.1.6.2 Reinforcing Stcel 3.8-10 3.8.1.63 Splices of Reinforcing Steel 3.8-10 l 3.8.1.6.4 Liner Plate and Appurtenances 3.8-10 3.8.1.6.5 Quality Control 3.8 11 3.8.1.7 Testing and Insenice Inspection Requirements 3.8-11 3.8.1.7.1 StructuralIntegrity Pressure Test 3.8 11 l 3.8.1.7.2 Preoperational and Insenice Integrated Leak Rate Test 3.8-11 l 3.8.2 Steel Comoonents of the Reinforced Concrete Containment 3.8-11 3.8.2.1 Description of the Containment 3.8 11 3.8.2.1.1 Description of Penetrations 3.8 11 0 3.8-iii Amendment I

ABWR useiman Standard Plant anv. A SECTION 3.8 O CONTENTS (Continued) Section Title Page 3.8.2.1.1.1 Personnel Air Locks 3.8-11 . 3.8.2.1.1.2 Equipment Hatch 3.8-12 3.8.2.1.1.3 Other Penetrations 3E-12 3.8.2.1.1.4 Drywell Head 3.8 12

               '4R?13              Boundaries                                                   3.8-13 3.8.2.2              Applicable Codes, Standards, and Specifications              3.8 13 3.8.2.2.1           Codes and Standards                                          3.8-13 3.8.2.2.2           Code Classifications                                         3.8 13 32.2.2.3            Code Compliance                                              3.8-13 3.8.2.3             Imds and lead Combinations                                   3.8-13 3.8.2.4             Design and Analysis Procedures                               3.8-13 l

Description 3.8.2.4.1 3.8 13 32.2.4.1.1 Personnel Air Locks 3.8-13 3.8.2.4.1.2 Equipment Hatches 3.8-13 3.8.2.4.13 Penetrations 3.8-14 3.8.2.4.1.4 Drywell Heads 3.8-14 32.2.5 Structural Acceptance Criteria 3.8-14 32.2.6 Materials, Quality Control, and Special 3.8-14 Construction Techniques 3.8.2.7 Testing and Inservice Inspection Requirements 3.8 15 O 3.8-iv Amendment 1

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ABWR MA61MAE Standard PI: int nriv ^ SECTION 3.8 CONTENTS (Continued) Section lille Eagc 3.8.2.7.1 Examination Requirements 3.8-15 3.8.2.7.1.1 Impaet Testing 3.8-15 3.8.2.7.1.2 Radiography 3.8 15 3.8.2.7.13 t>'trasonic Examination of Plates 3.8 15 3.8.2.7.1.4 Magnetic Particle Examination 3.8 15 3.8.2.7.1.5 Liquid Penetrant Examination 3.8-15 3.8.2.7.2 Shop Testing Requirements 3.8 15 3.83 Concrete and Steel Internal Structures of the Qncrete Containment 3.8 16 3.83.1 Description of the Internal Structures 3.8 16 3.83.1.1 Diaphragm Floor 3.8 16 3.83.1.2 Reactor Pedestal 3.8 16 3.83.13 Reactor Shield Wall 3.8 17 3.83.1.4 Drywell Equipment and Pipe Support Structure 3.8 17 3.83.1.5 Other Internal Structu.'es 3.8-17 3.83.1.5.1 Miscellaneous Platforms 3.8 17 3.83.1.5.2 Lower Drywell Equipment Tunnel 3.8 17 3.83.1.53 Lower Drywell Personnel Tunnel 3.8-17 3.83.1.5.4 Reactor Shield Wall Stabilizer 3.8 17 3.83.2 Applicable Codes, Standards, and Specifications 3.8-17 3.833 Loads and Load Combinations 3.8 18 9 3.8 v Amendment 1

23A6100AE i Standard Plant nev. A SECTION 3.8 O CONTENTS (continued) i Section Illis Page 3.833.1 Load Definitions 3.8-18 3333.2 Load Combinations 3.8-18 3.83.4 Design and Analysis Procedures , 3.8-18 3.83.4.1 Diaphragm Floor 3.8-18 l 3.83.4.2 Reactor Pedestal 3.8-18 i l l 3.83.43 Reactor Shield Wall 3.8-19 1 3.83.4.4 Drywell Equipment and Pipe Support Structure 3.8 19 3.83.4.5 Other Internal Structures 3.8-19 3.83.5 Structural Acceptance Criteria 3.8-19 3.83.5.1 Diaphragm Floor 3.8 19 3.83.5.2 Reactor Pedestal 3.8 19 , I 3.83.53 Other Internal Structures 3.8-19 1 3.83.6 Materials, Quality Control, and Special Construction Techniques 3.8-19 I 3.83.6.1 Diaphragm Floor 3.8-19 3.83.6.2 Reactor Pedestal 3.8-19 3.83.63 Reactor Shield Wall 3.8-19 3.83.6.4 Drywell Equipment and Pipe Support Structure 3.8 19

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3.83.6.5 0ther Internal Structures 3.8-20 3.83.7 Testing and Inservice Inspection Requirements 3.8 20 3.8.4 Other Seismic Catenory I Structures 3.8-20 3.8 vi Amendment 1

ABWR uumn Standard Plant nuv. A l SECTION 3.8 1 CONTENTS (Continued) 0\ l 1 i SeclinD .TLuc Pace l 3.8.4.1 Description of the Structures 3.8-20 l 3.8.4.1.1 Reactor Building Structure 3.8 20 3.8.4.1.2 Control Building 3.8 21 3.8.4.1.4 Seismic Category 1 Cable Tray and Conduit Supports 3.8 21 3.8.4.2 Appliable Codes, Standards, and Specifications 3.8 21 3.8.4.2.1 Reactor Building 3.8 21 3.8.4.2.2 Control Building 3.8 22 3.8.4.23 Seismic Category 1 Cable Tray and Conduit Supports 3.8 23 3.8.43 Loads and Load Combinations 3.8-23 1 3.8.43.1 Reactor Building 3.8 23 g 3.8.43.1.1 Loads and Notations 3.8 23 3.8.43.1.2 Load Combinations for Concrete Members 3.8-24 3.8.43.13 Load Combinations for Steel Members 3.8-25 3.8.43.2 Control Building 3.8 25 3.8.43.4 Seismic Category 1 Cable Tray and Conduit Supports 3.8 26 3.8.4.4 Reactor Building Design and Analysis Procedures 3.8 26 3.8.4.4.1 Reactor Building 3.8 26 9 3.8-vil Amendment 1

l MM 23A6100A!! Standard Plant anv.A l SECTION 3.8 CONTENTS (Continued) Section Ii.tla Eagt 3.8.4.4.4 Seismic Category I Cable Tray and Conduit Supports 3.8-26 3.8.4.4.4.1 Cable Tray Supports 3.8-26 l l 3.8.4.4.4.2 Conduit Supports 3.8-26 3.9.4.5 Structural Acceptance Criteria 3.8 26 3.8A.5.1 Reactor Building 3.8-26 3.8.4.5.1.1 General Criteria 3.8-26 3.8.4.5.1.2 Materials Criteria 3.8-27 3.8.4.5.2 Control Bullding 3.8-27 3.8.5 Foundations 3.8-27 3.8.5.1 Description of the Foundatiors 3.8-27 3.8.5.2 Applicable Codes, Standards and Specifications 3.8-27 3.8.53 Loads and Load Combinations 3.8-27 3.8.5.4 Design and Analysis Procedures 3.8 27 3.8.5.5 Structural Acceptance Criteria 3.8-28 3.8.5.6 Materials, Quality Control, and Special Construction Techniques 3.8-28 3.8.5.7 Testing and Inservice Inspection Requirements 3.8 28 O 3.8viii Amendment 1

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l l l ABWR m iman l Standard Plant any a I 1 SECTION 3.8 TABLES O' Table Title Eage 3.81 Load Combinations, Load Factors, and Acceptance Criteria for the Reinforced Concrete Containment 3.8 29 3.8-2 Major Allowable Stresses in Concrete and Reinforcing Steel 3.8-30 3.8-3 Stress Intensity Limits 3.8 31 38-4 Codes, Standards, Specifications, and Regulations Used in the Design and Construction of Seismic Category I Internal Structures of the Containment 3.8 32 3.8-5 Load Combhunons, ' cad Factors, and Acceptance Criteria for the Reinforced Concrete Structures inside the Containment 3.8-34 3.86 lead Combinations, Load Factors, and Acceptance Criteria for Steel Structures inside the Containment 3.8-35 3.8-7 Load Combinations for Foundation Design 3.8-36 ILLUSTRATIONS Eignts Title Eage 3.8 1 RB Floor Plan El. (-) 13200 3.8-37 3.8-2 RB - Floor Plan El. (-) 6700 3.8-38 3.83 RB - Floor Plan El. (-) 0200 3.8-39 3.8-4 RB - Ground Floor Plan El 7300 3.8-40 l 3.8-5 RB Floor Plan El.13100 3.8-41 ' 3.8-6 RB Floor Plan El.18500 3.8-42 3.8-7 RB Floor Plan El. 26700 3.8-43 3.88 RB - Roof Plan El. 33200 3.8 44  ! O' 3.8ix Amendment 1 i l

23A6100AE

     . Standard Plant                                                                                           REV.A                l l

SECTION 3.8 O ' ILLUSTRATIONS (Continued) Figure Illie Page l 1 3.8-9 RB - Roof Plan El. 44700 3 3-45

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3.8-10 RB Section OO - 1800 3.8-46 i 3.8-11 RB Section 900 2700 3.8-47 3.8-12 Code Jurisdictional Containment Boundary 3.8-48 ) l 3 3-13 RB Containment Structure 1800 Finite Element Model 33-49 I l 3 3-14 Section 1x> cations in RB Containment Fmite  ! Element Model 3.8-50 l 3 3-15 RB Containment Upper Drywell Equipment Hatch 3.8-51 3 2-16 RB Containment Dr>well Head 3.8-52  ! l 3.8-17 RB RCCV Internal Structures Nomenclature 3.8 53 I 3.8 18 RB RCCV Configuration 1.8-54 3.8-19 Soil Pressure on Reactor Building 3.8-55 3.8-20 Annual Temperature Profile of Suppression Pool Water During Normal Operations of a Typical Plant in Southern States 3.8-56 1 1 i O 3.8 x Amendment 1

l l ABM ua6iconii l Standard Plant an A 3.8.1 CONCRETE CONTAINMENT interference with the containment wall , s.' reinforcement. , The containment structure is designed to house , the primary nuclear system and is part of the The foundation mat is shown in Appendix 3H. containment system whose fur.ctional requirement is to confine the potential release of The containment wall is a right, circular ' radioactive material in the event of LOCA. This cylinder,2 m (o Icet,7 inches) thick, with an subsection describes the concrete containment inside radius of 14.5 m (47 feet,7 inches) and 1 structure. Steel components of the containment has a height of 29.5 m (96 feet,9 inches) meas-that resist pressure and are not backed by ured from the top of the founddon mat to the structural concrete are discussed in Subsection bottom of the containment top slab. The main 1 3.8.2. A detailed description of the containment reinforcement in the wall consists of inside and I system is presented in Subsection 6.2. outside layers of hoop and vertical reinforce-ment and radial bars for shear reinforcement. 3.8.1.1 Description of the Containment i Details of the containment wall reinforcement I 3.8.1.1.1 Concrete Containment are shown in Appendix 3H.

   . The containment is shown in Figures 3.8-1                 Reinforcement is placed at major                    I through 3.8 8.                                    discontinuities in the wall including the intersection of the wall and foundation mat, the The structural system is a low-leakage vicinity of the wall intersection with the top reinforced concrete structure with an internal    slab, around major piping penetrations, the steel liner in the drywell and suppression        upper drywell equipment hatch and personnel chamber to serve as a leak tight membrane. The    airlock, the lower drywell equipment hatch and containment is a cylindrical shell structure      personnel airlock tunnels, and suppression                     !

which is divided by the diaphragm floor and the chamber access hatch. l O re cier nedesi > into n never dr>-cii chember. a lower drywell chamber and a suppression chamber. The arrangement of reinforcement around major The top slab of the containment is an integral openings is shown in Appendix 3H. part of the fuel pool with the pool girders rigidly connected to the containment top slab and The containment top slab is nominally 2.2 m the reactor building walls. The reactor building (7 feet,2 inches) thick. The slab thickness is floors that abut the containment are integrated increased to 2.4 m (7 feet,11 inches) beneath structurally with the concrete containment. The the fuel pool, steam dryer and steam separator containment foundation mat is continuous with the pool, and around the drywell hcad opening. reactor building foundation mat. The containment wall, top slab, reactor building floor slabs and The containment top slab main reinforcement foundation mat are constructed of cast in place, consists of a top and bottom layer of conventionally reinforced concrete, reinforcement. The top layer of reinforcement is arranged in a rectangular grid. The bottom The containment foundation mat is 5.5m (18 layer of reinforcement is arranged in a feet) thick. The foundation mat reinforcement rectangular grid and then is bent near the consists of a top layer of reinforcement, a containment wall into a radial pattern to avoid bottom layer of reinforcement, and vertical shear interference with the containment wall vertical reinforcement. The bottom layer of reinforcement reinforcement. Hoop reinforcement is provided is arranged in a rectangular grid. The top layer in the area of the drywell head opening, cf reinforcement is arranged in a rectangular grid at the center of the mat and then radiates The arrangement of reinforcement for the outward in a polar pattern in order to avoid containment top slab is shown in Appendix 3H. O Amendment 1 3 8.t

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ABM 23A6100AH Standard Plant nov ^ l 3.8.1.1.2 Contalnment IJner Plate (ASME) Boiler and Pressure Vessel Code, Division l 2, Section III, Subsection CC. 3 The internal surface of the containment is W lined with welded steel plate to form a leaktight 3.8.1.2.3 General Design Criteria, Regulatory barrier. The liner plate is fabricated from Guldes, and Industry Standards carbon steel except that stainless steel plate or clad is used on wetted surfaces of the suppres- (1) 10 CFR 50, Appendix A ' General Design sion chamber. The extent of the stainless steel Criteria for Nuclear Power Plants" Criteria liner plate in the suppression chamber is shown 1, 2, 4,16 a n d 50. Conformance is in Appendix 3H. The liner plate is stiffened by discussed in Section 3.1. use of structural sections and plates to carry the design loads and to anchor the liner plate to (2) U.S. Nuclear Regulatory Commission (NRC) the concrete. The liner plate is thickened 1o. Regulatory Guides. Regulatory Guide 1.136. cally and additional anchorage is provided at Materials, Construction and Testing of major structural attachments such as penetration Concrete Containment, sleeves, structural beam brackets, the RPV pedes-tal and the SRV quencher support connection to (3) Industry Standards the basemat, and the diaphragm floor connection Nationally recognized industry standards to the containment wall. such as those published by the American Society for testing and Materials (ASTM) and The erection of the liner is performed using the American National Standards Institute standard construction procedures. The (ANSI) as referenced by the Applicable containment wall liner and top slab liner are Codes, Standards, and Regulations are used, used as a form for concrete placement. The liner on the bottom of the suppression chamber and 3.8.1.2.4 Containment Boundary lower drywell is placed after the foundation mat concrete is in place. The jurisdictional boundary for application 3.8.1.2 Applicable Codes, Standards, and of Section III, Division 2 of the ASME code to the concrete containment is shown in Figure g Specifications 3.8-12. The boundary extends to the: The de.<ign, f abrication, construction, (1) Outside diameter of the containment wall testing, and inservice inspection of the contain- from the foundation mat to the containment ment conforms M the applicable codes, standards, top slab. specifications, and regulations listed below, i except where specifically stated otherwise. (2) The foundation mat witbir the outside diameter of the containment wall. 3.8.1.2.1 Regulations (3) The containment top slab from the drywell (1) Code of Federal Regulations, Title 10, head opening to the outside diameter of the Energy, Part 50, ' Licensing of Production containment wall, and Utilization Facilities." (4) The intersection of the RPV pedestal on top (2) Code of Federal Regulations (CFR), Title 10 of the basemat.

     - Energy, Part 100, Reactor Site Criteria, (10 CFR 100), including Appendix A thereto, (5) The intersection of the diaphragm floor with

• Seismic and Geologic Siting Criteria for the containment wall.

Nuclear Power Plants." The concrete containment pressure boundary is 3.8.1.2.2 Construction Codes of Pmetice limited to the cylindrical wall of the drywell and wetwell, and the drywell top slab. American Society of Mechanical Engineers 9 Amendment 1 382

       . .~.

Al.lM 23A6100Ati [ Standard Plant EVA l l (3) To Thermal effects and loads during f] They are included in ASME code jurisdictiori normal operating, startup or shutdown (./ boundary for design, material, fabrication, conditions, including liner plate expansion, inspection, testing, stamping, etc., requirements equipment and pipe reactions, and thermal of the code. However, the fuel pool girders and gradients based on the most critical , any other structural components which are transient or steady state thermal gradient. l integral with the containment structure are , treated the same as the containment only as far (4) Ro Pipe reactions during normal as loads and loading combinations are concerned operating or shutdown conditions based on

in the design. Similaryly, the reactor building the most critical transient or steady state floor slabs that are integrated with the conditions. ,

containment are not included in the ASME code jurisdictional boundaries but are treated the (5) Po Pressure loads resulting from the same as the containment only as far as loads and pressure difference between the interior and load combinations are concerned. exterior of the containment, considering both interior pressure changes because of The reactor pedestal and diaphragm floor slab, heating or cooling and exterior atmospheric which partitition the containment into drywell pressure variations. and suppression chamber, are not part of the containment boundary. The reactor pedestal, (6) Construction Loads - Construction loads are steel structures filled with concrete, and the loads which are applied to the containment diaphragm floor slab are designed according to from start to completion of construction. codes given in Sections 3.8.3 and 3.8.4, The definitions for D, L and To given respectively. above are applicable, but are based on actual construction methods and/or Those portions of the structure outside the conditions. indicated code jurisdictional boundary will be

 . designed, analyzed and constructed as indicated       (7) SRV - Safety / Relief Valve Loads.

in Sections 3.8.3, 3.8.4 a n d 3.8.5. The Oscillatory dynamic pressure loadings analytical models will include both the resulting from discharge of safety / relief containment and reactor building and therefore valves (SRVs) into the suppression pool. l will provide continuity in the analysis. The development of these loads is with the j methods described in Appendix 3B of this 3.8.1.3 Loads and Load Combinations SSAR. The reactor building vibration (RV3) dynamic effects shall be included in the . The containment is analyzed and designed for load combinations. The number and ) all credible conditions of loading including combinations of valves that will open during normal loads, preoperational testing loads, loads a reactor vessel pressure transient are as during severe environmental conditions, loads follows: during extreme environmental conditions and loads during abnormal plant conditions. (a) G 1 - Design pressure load on the suppression pool boundary resulting from 3.8.1.3.1 Normal toads discharge of one safety / relief valve into the suppression pool. First (1) D - Dead load of the structure and equipment actuation and subsequent actuation shall  ; plus any other permanent loads, including be considered. l vertical and lateral pressures of liquids. (b) G2 Design pressure load on the (2) L - Live loads, including any moveable suppression pool boundary resulting from equipment loads and other loads which vary discharge of two adjacent safety / relief in intensity and occurrence, such as forces yalves, first aetuation, into the exerted by the lateral pressure of soil, suppression pool. O Amendment 1 3.83

1 ABWR a346iman Standard Plant REV.A (c) GALL Design pressure load on the LOCA (LBL), based upon the calculated peak suppression pool boundary resulting from discharge of all SRVs, first actuation, pressure with an appropriate margin. into the suppression pool. (5) Pi Design accident pressure load within the containment generated by an intermediate (d) ADS Design pressure load on the break LOCA (IBL). suppression pool boundary resulting from the SRV automatic depressurization (6) P3 - Design accident pressure load within system (ADS) discharge into the the containment generated by a small break suppression pool. LOCA (SBL). 3.8.13.2 Preoperational Testing Loads (7) Y Local effects on the containment due to a LOCA. The local effects shall include the (1) P. Test loads are loads which are following: applied during the structural integrity test. (a) Y r- Load on the containment generated by the reaction of a ruptured high ener-(2) T t Thermal effects and loads during the gy pipe during the postulated event of structural integrity test. the DBA. The time dependent nature of the load and the ability of the contain-3.8.133 Severe Environmental Loads ment to deform beyond yield shall be considered in establishing the structu-(t) E - Operating Basis Earthquake (OBE) loads ral capacity necessary to resist the as defined in Section 3.7. effects of Y r-(2) W - Loads generated by the design wind Load on the containment generated specified for the plant site as defined in (b)byYjt he jet impingement from a ruptured a Section 3.3. high energy pipe during the postulated W event of the DBA. The time dependent 3.8.1.3.4 Extreme Endronmental Loads nature of the load and the ability of the containment to deform beyond yield (1) E' - Safe shutdown earthquake (SSE) loads as shall be considered in establishing the defined in Section 3.7. structural capacity necessary to resist (2) W' Loads generated by the tornado specified in Section 3.3. (c) Ym - The load on the containment re-sulting from the impact of a ruptured 3.8.1.3.5 Abnormal Plant Imds high energy pipe during the DBA. The type of impact, for example plastic or (1) FL - Ilydrostatic load due to post LOCA clastic, together with the ability of

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flooding of the containment .to a level of the containment to deform beyond yield 131 feet above the top of the foundation mat shall be considered in establishing the for fuel recovery subsequent to a design structural capacity necessary to resist basis accident. the impact. (2) Ra - Pipe reactions (including Ro) from (8) CO An oscillatory dynamic loading thermal conditions generated by a LOCA. (condensation oscillation), on the suppression pool boundary due to steam (3) Ta Thermal effects (including T o) and condensation at the vent exits during the loads generated by a LOCA. period of high steam mass flow through the vents following a LOCA. (4) Pa - Design accident pressure load within the containment generated by large break (9) CHUG An oscillatory dynamic loading g Amendment 1 384

i MkN MA6100AE Standard Plant anv. A , l (chugging) in the top vent and on the the center line of the plant that simulate the < suppression pool boundary due to steam symmetry of the whole structure. (q/ condensation inside the top vent or at the i top vent exit during the period of low steam The foundation soil is simulated by a set of mass flow in the top vent following a LOCA. horizontal and vertical springs. The soil spring constraints are calculated based on using (10) VLC Loads from component response or correction factors to account for the reactor direct fluid forces, on components located building embedment, with a shear wave velocity in the suppression pool, caused by the main of 1000 fps, a shear modulus of 3.88 x 100 vent line clearing phenomenon. Ibs/ft 2 and a Poisson ratio of 0.45. 1 (11) PS Pool swell bubble pressure The containment and reactor building 1800 ) (axisymmetrical and nonaxisymmetrical) on finite element model is shown in Figure 3.8-13. ) the suppression pool boundary due to a LOCA. 3.8.1.4.1.1.1 Nonaxisymmetrical Loads i 3.8.1.3.6 Load Combinations for the l Containment Structure and Liner Plate Nonaxisymmetrical loads imposed on the j structure include the following, and are as j

     'The containment structure is designed using defined in Subsection 3.8.1.3.                             j the loads, load combinations, and load factors listed in Table 3.8-1.                             (1) Tornadowind Loads and load combinations listed in Table    (2) Design wind 3.81 shall be used for the design of the steel liner and liner anchors, but the load factor for   (3) Safe shutdown earthquake all loads in the load combinations shall be 1.0.

(4) Operating basis earthquake ( 3.8.1.4 Design and Analysis Procedures (5) Local pipe rupture forces, including local This section describes the analytical and compartmental pressures from ruptured pipes , design procedures used in designing the in compartments inside or outside the containment, containment. 3.8.1.4.1 Containment Cylindrical Wall, (6) Pool sw:ll bubble pressure Top Slab, and Foundation Stat  ; (7) SRV actuation in the suppression pool 3.8.1.4.1.1 Analytical hfethods (8) Loadings from embedded steel brackets in the The containment stn eture is analyzed by the wall and top slab. use of the linear clastic finite element computer program STARDYNE described in Appendix SC. The The containment structure is analyzed for the containment and reactor building layout utilizes nonaxisymmetrical pool swell bubble pressure, an integrated structural system. The structure and nonaxisymmetrical pressures from discharge. is idealized as a three dimensional assemblage of beam elements, and iso. parametric membrane- An equivalent static analysis is performed bending plate elements, for the nonaxisymmetrical pool swell bubble pressure loading using the peak pressures for Since the containment and reactor building are this loading. essentially symmetrical about the center line of the plant parallel to the fuel pool girders and Input data for nonaxisymmetrical pool swell steam tunnel, only 1800 of the structure is bubble pressures is described in Subsection modeled. Boundary conditions are applied along 3.8.1.3.5, O Amendment 1 3.8-$

l MkN 23A6100AE Standard Plant REV.A l Seismic inertial forces (two orthogonal hori- structure include the following, and are as zontal and one vertical) based on the analysis defined in Section 3.8.1.3. described in Section 3.7, are applied to the fi-g nite element model as equivalent static forces. (1) Structure dead load The resulting moments and forces at various sec-tions of the containment structure are combined (2) Surcharge loads from adjacent structures by the square root of the sum of the squares (SRSS) method. (3) Ilydrostatic load from probable maximum flood The containment wall is shielded from the de-sign wind by the reactor building, which com- (4) Hydrostatic load from normal site water pletely encloses the structure. Forces from the table design wind is transmitted directly to the con-tainment wall through the reactor building (5) Ilydrostatic load from post LOCA flooding of connections. the containment The actuation of SRVs results in dynamic loads (6) Local dead and live loads from embedded on the suppression pool boundaries. These dyna- brackets, treated as axisymmetrical loads mic loads are formulated by applying a non dimen- for overall structural response sional time function to the attentuated pressures in the suppression pool which result from single (7) Dead and live loads from internal and multiple valve discharge. These attenuated structures imposed on the foundation mat pressures area calculated based upon the methodo-logy presented in Appendix 3B The non dimen- (8) Normal operating thermal gradients sional time function is oscillation of the bubble pressure within the suppression pool, which is (9) Abnormal plant thermal gradients (including shown in Appendix 3B. The magnitude of the those from LBL, IBL and SBL) pressure at any point within the pool decreases with time, with the duration of the load being (10) Preoperational test pressure lets than 1 second. This pressure time history is represented in terms of an equivalent static (11) Abnormal plant pressure loads (including load and then used as input for the structural those from LBL, IBL and SBL) analysis with a dynamic load factor. (12) Normal external pressure load The containment wall and containment basemat etc extremely stiff steel lined reinforced con- (13) Safety / relief valve actuation in suppres-crete structures which form the suppression pool sion pool b o u n d a ry. Thus, effects of fluid structure interaction upon the total containment building (14) Pool swell bubble pressure response due to dynamic suppression pool boundary loads are small. Suppression pool boundary A LOCA and SRV actuation result in dynamic loads, defined in Appendix 3B, are applied to the loads on the suppression pool boundaries. These mathematical model as rigid wall loads. The mass hydrodynamic loads are formulated by applying a of the suppression pool water has been lumped at time function to the attenuated pressures in the those node points that form the suppression pool suppression pool. The attenuated pressures are boundary. The effect of negative SRV pool boun- calculated based on the methodology presented in dary loads upon the suppression pool liner is Appendix 3B. Once the pressure time histories discussed further in Subsections 3.8.1.1.2 and are formulated, they are represented in terms of 3.8.1.4.2. an equivalent static load and then used as input for the finite element analysis with a dynamic 3.8.1.4.1.1.2 Asisymmetrical leads load factor. Axisymmetric loads imposed on the containment g Amendment 1 3.84

23A6100AE Standard Plant anv. A 3 8.1.4.1.13 Major Penetrations - (CECAP), described in Appendix 3C, is used to r- determine the extent of concrete cracking at ( The major penetrations in the containment wall these sections, and the concrete and rebar include the upper drywell equipment and personnel stresses and liner plate strains. The CECAP l hatches, the lower drywell equipment and program models a single element of unit height,

                                                                                                            )

personnel tunnels and batches, the suppression unit width, and depth equal to the thickness of , chamber access hatch, and the main steam and the wall or slab. The calculations used in  ; feedwater pipe penetrations. The state of stress CECAP assume that the concrete is isotropic and and behavior of the containment wall around these linear elastic but with zero tensile strength, openings is determined by the use of analytical CECAP also can calculate the reduced thermal numerical techniques. The analysis of the area forces and moments due to concrete cracking, around the penetrations consists of a non-linear However, the redistribution of forces and three dimensional finite element analysis with moments are not calculated. To account for the boundaries extending to a region where the :oncrete cracking effects on the redistribution discontinuity effects of the opening are of forces and moments, an interactive procedure negligible. described in Subsection 3.8.1.4.1.3 is used. Displacements compatible with the global The input data for the CECAP program consist analysis of the containment are applied at tl ese of the membrane forces, shear forces and bending boundaries. The stresses and strains in the moments calculated by the STARDYNE t nalysis. reinforcement, concrete and liner plate are The areas of the reinforcing steel in terms of obtained from the local finite element mode. The steel area to concrete cross section ratio are analysis considers concrete cracking and thermal based on a trial design shown in Appendix 3H. strains. When the calculated stresses in the steel and in the concrete meet the acceptance criteria, the 3.8.1.4.1.1.4 Variation of Pbysical Material trial design is adequate; otherwise, the trial Properties design is revised. The evaluation of contain- - ment structural adequacy is shown in Subsection In the design analysis of the containment 3.8.1.5. consideration is given to effects of possible variations in the physical properties of 3.8.1.4.13 Concrete Cracking Considerations materials on the analytical results. A high initial value of the concrete modulus of The membrane forces, shear forces and bending elasticity is used in the analysis, which will moments in the containment structure subjected conservatively determine the upper bound effects to loads are obtained by applying the STARDYNE from thermal stresses in the structure. The computer program to the finite element model variation of the soil modulus of elasticity and that was developed on the basis of an uncracked Poisson's ratio of concrete is considered in the section. This model is called the uncracked i analysis in order to constitute the most severe model. In sizing the reinforcing steel or in i structural responses. calculating the rebar stresses, the concrete is j not relied upon for resisting tension. Th. s, ) 3.8.1.4.1.2 Design Methods those portions of structures which are either in  ! membrane tension or in flexural tension are l The design of the containment structure is cracked to transfer loads from concrete to based on the membrane forces, shear forces and rebar. The CECAP program, described in Appendix bending moments for the load combinations defined 3C, is used for calculating the extent of in Subsection 3.8.1.3.6. The membrane forces, concrete cracking and the stresses in the shear forces and bending moments in selected concrete and in the steel. Because of concrete l sections are obtained by the computer program cracking that leads to stiffness changes, the ' STARDYNE as described in Section 3.8.1.4.1.1. distribution of forces and moments is different The selected sections are shown in Figure 3.814. from those calculated from the uncracked model described above. To determine the effects due The Concrete Element Cracking Analysis Program to concrete crasking a revised finite element Amendment 1 3.8-7 1

I MM 23A6100AF Standard Plant avA l model called the ' cracked model*, that takes into The suppression pool contains air-saturated, account the concrete cracking, was developed for stagnant, high purity water and is designed for applying the STARDYNE computer program. The a 60 year life. The amount of corrosion is "cracked model" is the same as the original based on the annual temperature profile of noncracked model except that the element suppression pool water for a typical plint in thicknesses for the diaphragm floor (D/F) slab southern states under normal operation (Figure and the top slab are reduced to reflect concrete 3.8-20). The following conditions can cause the cracking due to bending while the element pool temperature to rise above normal: thicknesses for the reactor building (R/B) floor slab and walls are reduced to reflect the tension (1) Reactor core isolation mode: pool cracks. In this model, the thicknesses for the temperature can rise 300F above normal RPV pedestal and reactor shield wall are not for a total of 165 days during the 60 year reduced because they are steel structures (with life; and concrete fill). The basemat and the columns are predominately subjected to compression, so no (2) Suppression pool cooling mode: pool cracking is assumed. temperature can rise 300F above normal for a total of 540 days during the 60-year The results obtained from the uncracked model lifetime. and from the cracked model are compared for SSE loads and for test pressure load cases in The corrosion allowance for type 304L stain-Appendix 3G and 3H respectively. less steel in air saturated water for any oxygen level and temperatures up to 6000F for 60 A comparison indicates that the forces and/or years is 0.005 inches. The major concern has bending moments from the uncracked model are involved the air / water interface area vchere controlling in ssme of the regions of the pitting is most likely to occur. The 0.005 inch structure and the results from the cracked model corrosion allowance is a small fraction of the are controlling in other regions, stainless steel thickness which will be a nominal 1/10 of an inch if clad carbon steel The procedures for the design and analysis of plate is used. g the liner ph te and its anchorage system are in accordance ub the provisions of the ASME Code Water und to fill the suppression pool is Section !!! Div. 2 Subarticle CC 3600. The linu either condensate or demineralized. No chem-plate analysis considers deviations in geometry icals are added to the suppression pool water. due to fabrication and erection tolerances, and variations of the assumed physical properties of Observations made on suppression pool water the liner and anchor material. The strains and quality over a period of several years indicate stresses in the liner and its anchors are within that periodic pool cleaning such as by allowable limits defined by the ASME Code, underwater vacuuming will be required as well as Section Ill, Div. 2, Subarticle CC-3720. the use of the suppression pool cleanup system to maintain water quality standards. The 3.8.1.4.1.4 Corrosion Presention suppression pool cleanup system (Subsection 9.5.9) also acts to taaintain purity levels. Type 304L stainless steel or clad carbon steel plate will be used for the containment An ultrasonic thickness measurement program liner in the wetted areas of the suppression pool will be performed to detect any general as protection against any potential pitting and corrosion at underwater positions. A visual corrosion on all wetted surfaces and at the water examination for local pitting on the underwater to-air interf ace area. The advantage of portions of the steel containment will be made cladding is greatly reduced pool maintenance / at refueling outages using underwater lighting operating costs over the life of the plant and and short focus binoculars. This covers 10% of crud removal and burial costs over the life of the surface at the first refuchng outage after the plant. the start of commercial operation, 5% 9, Amendment 1 38-8

1 l l M 23A6100AE Standard Plant REV.A l l additional surface approximately two to five 3.8.1.6.1 Concrete ,A years later and 5% at five year intervals there-() after. If pits are detected at any examination, All concrete materials are approved prior to ' representative ones are ultrasonically tested and start of construction on the basis of their i the depth of those large enough for measurement characteristics in test comparisons using ASThi will be determined. Appropriate repairs can be standard methods. Concrete aggregates and i made as required. cement, conforming to the acceptance criteria of the specifications, are obtained from approved 3.8.1.4.2 Ultimate Capacity of the Containment sources. Concrete properties are deteimined by laboratory tests. Concrete admixtures are used An analysis is performed to determine the to minimize the mixing water requirements and ultimate capacity of the containment. The increase workability. The specified compressive results of this analysis are summarized in strength of concrete at 28 days, or earlier, is: Chapter 19. Specified Strength 3.8.1.5 Structural Acceptance Criteria Structure f'c (osi) i For evaluation of the adequacy of the contain. Containment 4000 ) ment structural design, the major allowable Foundation hist 3000 1 stresses of concrete and reinforcing steel for  ! service load combinations and factored load All structural concrete is batched and placed i combinations according to AShiE III, Div. 2 are in accordance with Subarticle CC-2200 and l shown in Table 3.8-2. Article CC-4000 of the AShiE Code, Section 111, ' Division 2. The allowable shear stresses are extracted from ash 1E III, Div. 2. The allowable values are (a) Cement low when only the orthogonal reinforcement system is considered in resisting tangential shear. Cement is Type 11 conforming to the  ! However, the draft of the ash 1E Code, Section lit, Specification for Portland Cement (ASTH! C l Div. 2. revision has adopted higher allowable 150). The cement contains no more than values (500 psi). Because this draft has passed 0.60 percent by weight of alkalies the ash 1E 111 design committee, the revis;d calculated as sodium oxide plus 0.658 allowable shear stresses are used for the potassium oxide. Certified copies of containna:nt structural adequacy evaluation to material test reports showing the chemical justify that no dhgonal reinforcement is to be composition and physical properties are used in resisting lateral shear loads. obtained for each load of cement delivered.  ; 1 The structural evaluation is provided in For sites where concrete may come into con- ) Appendix 3H. tact with soils having more than 0.20% wa- l ter soluble sulfate (as SO4 ) of ground- , 3.8.1.6 hiaterial, Quality Control, and Special water with a sulfate concentration exceed- I Construction Techniques. ing 1500 ppm, only Type V cement shall be l used unless other suitable means are hiaterials used in construction of the con- employed to prevent sulfate attack and i tainment are in accordance with Regulatory Guide concrete deterioration. 1.1.36 and the ash 1E Code, Section III, Disision

2. Article CC-2000. Specifications covering all (b) Aggregates materials are in sufficient detail to assure that the structural design requirements of the work All aggregates conform to che Specification are met. for Concrete Aggregates (ASThi C 33).

O Amendment 1 3.89

MM 33A6100AE Standard Plant REV.A (c) Water meeting the design requirements specified for concrete are used. Water and ice for mixirg is clean, with a .s total solids content of not mor th '000 3.8.!.6.2 Reinforcing Steel ppm as measured by ASTM D 1888. . . .aixing water, including that contained as free Reinforcing bars for concrete are deformed water in aggregate, co :tains not more than bars meeting requirements of the Specification 250 ppm of chlorides as Cl as determined by for Deformed and Plain Billet Steel Bars for ASThi D 512. Chloride ions contained in the Concrete Reinforcement (ASThi A 615, Grade 60). aggregate are included in calculating the hiill test reports, in accordance with ASThi A total chloride ion content of the mixing 615, are obtained from the reinforcing steel water. The chloride content contributed by supplier to substantiate specification the aggregate is determined in accordance requirements, with ASThi D 1411. The test procedures are in accordance with (d) Admixtures ASTM A 370, and acceptance standards are in accordance with ASTM A 615. The concrete may also contain an air-en-training admixture and/or a water reducing 3.8.1.6.3 Splices of Reinforcing Steel admixture. The air-entraining admixture is in accordance with the Specification of Air Sleeves for reinforcing steel mechanical Entraining Admixtures for Concre:e (ASTM C splices conform to ASTM A 513, A519 or A 576 260). It is capable of entraining 3 to 6 Grades 1008 through 1030, Certified copies of percent air, is completely water soluble, material test reports indicating chemical compo-and is completely dissolved when it enters sition and physical properties are furnished by - the batch. Superplasticizers, entraining the manufacturer for each sleeve lot. from 1.5 to 4.5 percent air, may be used in concrete mixes (f' = 5000 psi, maximum) for Placing and srlicing of reinforcing bars is congested areas to improved workability and in accordance with Article CC-4300 and Subarti-g prevent the formation of voiA around rein. cle CC-3530 of the ASME Code, Section III, forcement. The water reducing admixture Division 2. conforms to the standard specification for Cbemical Admixtures for Concrete (ASTM C 3.8.1.6.4 Liner Plate and Appurtenances 494), Types A and D. Type A is used when average ambient temperature for the daylight The materials conform to all applicable period is below 700F. Type D is used when requirements of the ASME Code, Section III, average ambient air temperature for the day- Division 2. light period is 700F and above. Pozzo. lans, if used, conform to Specification for Steel plate is tested at the mill in full Fly Ash and Raw or Calcined Natural Pozzo- conformance to the applicable ASTM specifica-lans for Use in Portland Cement Concrete tions, and certified mill test reports are sup-(ASTM C 618) except that the loss on igni- pDd for review ant approval. The plate is tion shall be limited to 6 percent. Admix- visually examined for laminations and pitting. tures containing more than 1 percent by Identity of the plate is maintained throughout weight chloride ions are not used. fabrication. (e) Concrete Mix Design Dimensional tolerances for the erection of the liner plate and appurtenances are detailed Concrete mixes are designed in accordance in the Construction Specification based on the with ACI 211.1, Recommended Practice for structure geometry, liner stability, concrete Selecting Proportions for Nor nal and Heavy strength, and the construction methods to be Weight Concrete, using materials qualified used, and accepted for this work. Only mixes Amendmer t 1 3.8-10

ABM ux6too.m Standard Plant REV A 3.8.1.6.5 Quauty Control 3.8.2 STEEL COMPONENTS OF THE  ; 3 REINFORCED CONCRETE (V Quality control procedures are established in CONTAINMENT the Construction Specification and implemented during construetion and inspection. The 3.8.2.1 Description of the Containment Construction Specification covers the fabrication, furnishing, and installation of each The ABWR has a reinforced concrete structural item and specifies the inspection and containment vessel (RCCV) as described in documentation requirements to ensure that the Section 3.8.1. This section will describe the requirements of the ASME Code, Section III, following steel components of the concrete Division 2, and the applicable Regulatory Guides containment vessel: are met. (1) Personnel Air 1.ocks 3.8.1.7 Testing and inservice inspection Requirements (2) Equipment Hatches, 3.8.1.7.1 Structural Integrity Pressure Test (3) Penetrations A structural integrity test of the (4) Drywell Head containment structure is performed in accordance with Article CC-6000 of the ASME Code, Section 3.8.2.1.1. Description of Penetrations III, Division 2 and Regulatory Guide 1.1.36, af ter completion of the containment The penetrations through the RCCV include the construction. The test is conducted at 115% of following. the design pressure condition of 45 psig in both the drywell and suppression chamber, 3.8.2.1.1.1 Personnel Alr Locks simultaneously. A pressure test for the design differential pressure condition of 25 psig Two personnel air locks with an inside between the drywell and the suppression chamber diameter sufficient to provide 6 ft.,8 in., is also performed where the drywell pressure is high by 3 ft., 6 in., wide minimum clearance greater than the suppression chamber pressure, above the floor at the door way are provided. One of these air locks provides access to the During these tests the suppression chamber and upper dry well and the other provides access to spent fuel pool are filled with water to the the lower drywell via the access tunnel. normal operational water level. Deflection and concrete crack measurements are made to determine Lock and swing of the doors is by manual and that the actual structural response is within the automatic means. The locks extend radically limits predicted by the design analysis. outwa:d from the RCCV into the reacwr btdiding and ate supported by the RCCV only. The minimum i In addition to the deflection and crack clear horizontal distance not impaired by the I measurements, the first prototype containment door swing is 6 ft, structure is instrumented for the measurement of strains in accordance with the provisions of Each personnel air lock has two pressure-Subsubarticle CC-6230 of the ASME Code, Section seated doors interlocked to prevent simultanews 111, Division 2. opening of both doors and to ensure that one door is completely closed before the opposite l 3.8.1.7.2 Preoperational and Insenice door can be opened. The design is such that the  ! Integrated leak Rate Test interlocking is not defeated by postulated malfunctions of the electrical system. Signals Preoperational and inservice integrated leak and controls that indicate the operational rate testing is discussed in Section 6.2.1. status of the doors are provided. Provision is Amendment 1 3.8 11

ABWR mmu Standard Plant REV A made to permit temporary bypassing of the door 3.8.2.1.1.3 Other Penetrations interlock system during plant cold shutdown. The door operation is designed and constructed so The RCCV penetrations are categorized into g either door may be operated from inside the two basic types. These types differ with containment vessel, inside the lock, or from respect to whether the penetration is subjected outside the containment vessel. to a hot or cold operational environment. The lock is equipped with a digital readout The cold penetrations pass through the RCCV pressure transducer system to read inside and wall and are embedded directly in it. The hot outside pressures. Quick acting valves are penetrations do not come in direct contact with provided to equalize the pressure in the air lock the RCCV wall but are provided with a thermal when personnel enter or leave the containment sleeve which is attached to the RCCV wall. The vessel. Each door has two separate inflatable thermal sleeve is attached to the process pipe seals with provision for independent pressure and at distance from the RCCV wall to minimize leakage testing. The air supply to the seals is conductive heat transfer to the RCCV wall. designed so that each redundant seal on a given door is connected to an independent air supply. Besides piping penetrations, several electri-Loss of air supply does not jeopardize the cal penetrations also exist. A description of integrity of the lock. the various penetrations is given Chapter 8. 3.8.2.1.1.2 Equipment IIntch 3.8.2.1.1.4 Drpell IIcad Three equipment hatches are provided. The two A 33 feet 9 inches in diameter opening in the drywell equipment hatches are 9 feet 10 inches in RCCV or upper drywell top slab over the RPV is diameter. One of these serves the upper drywell covered with a removable steel ellipsoidal and the other serves the lower drywell via the drywell head which is part of the pressure access tunnel. The third equipment hatch is 6 boundary. The drywell head is designed for feet 6 inches in diameter and provides personnel removal during reactor refueling and for and equipment access to the suppression chamber replacement prior to reactor operation using the h airspace. reactor building cranc. One pair of mating flanges is anchored in the drywell top slab and The equipment hatch covers have a double O- the other is welded integrally with the drywell ring scaled flange with provisions to pressure head. Provisions are made for testing the test the space between the seals of the flange, flange seals without pressurizing the drywell. A means for removing .ind handling the equipment Figure 3.816 shows the drywell head. hatch cover is provided. The hoisting equipment and hoisting guides are arranged to minimize con. The drywell head is attached to the RCCV with tact between the doors and seats during opening a ' finger pin' type closure as shown in Figure and closing. The equipment hatch includes the 3.8 16. Stainless steel is used for the finger electric-motorized hoist with pushbutton control pins and mating flanges, llowever, the rest of stations, lifting . lings, hoist supports, hoist- the head shell may be carbon steel. ing guides, access platforms, and ladders for ac-cess to the dogged position of the door and The finger pin closure was selected because hoist, latches, seats, dogging devices, and tools it reduces the times required for removal and required for operation and maintenance of the installation of the head by a factor of 3 or 4. hatch. This time reduction also results in lower opera-tor doses. The structural benefit of the finger The equipment hatches and covers are entirely pin closure is the low discontinuity stress at supported by the RCCV. Figure 3.8-15 show gen- the connection as compared with a standard eral details of the equipment hatch and cover, bolted flange design. O Amendment 1 3812

MM 23A6100AE Standard Plant _ REV.A t 3.8.2.13 Boondaries Design by Analysis; and NE 3300, Vessel Design, . of ash 1E Code Section III. For the configurations kq) The boundaries of the steel components of the and loadings which are not explicitly treated in RCCV consist of those defined in Paragraph Subarticle NE 3130, the design is in accordance NE 1132, AShf E Code Section Ill, Division 1, with the applicable Subarticles designated in paragraphs (b) and (d) of Subarticles NE-3130 of 3.8.2.2 Applicable Codes, Standards, and AShiE Code Section 111. Specifications The design of nonpressure resisting parts is 3.8.2.2.1 Codes and Standards performed in accordance with the general prac-tices of the AISC hianual of Steel Construction. In addition to the codes and standards specified in Subsection 3.8.1.2.2, the following 3.8.2.4.1 Description codes and standards apply: Following are individual descriptions of the (1) American Society of Mechanical Engineers design and analysis procedures required to veri. (AShf E) Boiler and Pressure Vessel Code, fy the structural integrity of critical areas Section Ill, Division 1, Nuclear Power Plant present within the steel components of the RCCV. Components, Subsection NE, Class MC. 3.8.2.4.1.1 Personnel AirIAcks (2) AISC hianual for Steel Construction, The personnel air lock consists of four main 3.8.2.2.2 Code Classification sections: doors, bulkheads, main barrel, and reinforcing barrel with collar. The personnel The steel components of the RCCV are air locks are supported entirely by the RCCV classified as Class hic in accordance with wall. The lock barrel is welded directly to the Subarticle NA 2130, AShiE Code Section III. containment liner penetration through the RCCV wall. The personnellock and penetration through 3.8.2.23 Code Compliance the RCCV wall will be analyzed using a finite element computer program. The discon tinuity The steel components within the boundaries stresses induced by the combination of external, defined in Subsection 3.8.2.1.3, are designed, dead, and live loads, including the effects of fabricated, erected, inspected, examined, and earthquake loadings, are evaluated. The required tested in accordance with Subsection NE, Class htC analyses and limits for the result ing stress Components and Articles NA-4000 and NA 5000 of intensities are in accordance with Subarticles AShfE Code Section ill, Division 1. NE 3130 and NE 3200 of ASME Code Section III, Division 1. Structural steel attachments beyond the boundaries established for the steel components The piping system and components inside the of the RCCV are designed, fabricated, and personnel air locks are Class 2 and are designed constructed according to the AISC Manual for in accordance with ASME Code Section NC. Steel Construction. 3.8.2.4.1.2 Equipment flatches 3.8.23 Loads and Load Combinations An equipment hatch assembly consists of the The applicable loads and load combinations are equipment hatch cover and the equipment hatch described in Subsection 3.8.13. body ring which is imbedded in the RCCV wall and connects to the RCCV liner. 3.8.2.4 Design and Analysis Procedures A finite element analysis model will be used The steel components of the RCCV are designed to determine the stresses in the body ring and i in accordance with the General Design Rules of hatch cover of the equipment natch. The ' Subarticles NE-3100, General Design, NE 3200, equipment analysis and the stress intensity c Amendment 1 3813

ABWR uumu Standard Plant anv ^ limits are in accordance with Subarticle NE 3130 In addition to the structural acceptance cri-and NE 3200 of AShiE Code Section 111. The hatch cover with the bolted flange is designed in teria, the RCCV is designed to meet minimum leakage rate requirements discussed in Section g accordance with Subarticle NE 3326 of AShlE Code 6.2. Those leakage requirements also apply to Section Ill, the steel components of the RCCV. 3.8.2.4.1.3 Penetrations The combined loadings designated under

                                                 ' Normal *, ' Construction *,

• Severe Environmental',

Piping penetrations and electrical penetration "Extreme Environmental", ' Abnormal *, "Severe are subjected to various combination of piping Environmental

• and ' Abnormal /Extrerr.c Environ-reactions, mechanical, thermal and seismic loads mental" in Table 3.81 are categorized according transmitted through the RCCV wall structure. The to Level A, Lesel B, Level C and Level D service resulting forces due to various load combinations limits as defined ie NE 3113. The resulting ire combined with the effects of external and primary and local membrane, bending, and internal pressures. secondary stress intensities including compressive stresses are calculated and their The stresses in the penetrations are evaluated corresponding allowable limit is in accordance using a finite element model for stress analy- with NE 3220 of AShiE Code Section !!I.

sis. For penetrations subjected to cyclic loads, the peak stress intensities are also evaluated. In addition, the stress intensity limits for The required analysis and associated stress testing, design and post LOCA flooding conditions intensity limits are in accordance with NE 3130 are summarized in Table 3.8 3. and NE 3200 of ASME Code Section lit, Division 1. Stability against compression buckling is 3.8.2.4.1.4 Drpeli llend assured by an adequate factor of safety. The drywell head consisting of shell, finger The allowable stress limits used in the design pin closure and drywell head anchor system will be analyzed using a finite element stress and analysis of nonpressure resisting components are in accordance with 3.8.2.2.1 (2). h analysis computer program. The stresses, including discontinuity stresses induced by the 3.8.2.6 Materials, Quality Control, and combination of external pressure or internal Special Construction Techniques pressure, dead load, live load, thermal effects and seismic loads are evaluated. The required The steel components of the RCCV locks, analyses and limits for the resulting stress hatches, penetrations, and drywell head will be intensities are in accordance with NE 3130 and fabricated from the following materials: NE-3200 of the ASME Code Section ill, Division 1. ' (1) plate (SA 516 grade 70, SA 240 type 304L, The compressive stress within the knuckle SA-516 grade 60 or 70 purchased to SA 264); region caused by the internal pressure and the l compression in other regions caused by other (2) pipe (scamless SA-333 grade 1 or 6 or SA 106 loads are limited to the allowable buckling grade B or SA-312 type 304L)- stress values in accordance with NE 3222 of the i ASME Code Section III, Division 1. (3) forgings (SA 350 grade LF1 or LF2); 3.8.2.5 Structural Acceptance Criteria (t) botting (SA 320 L43 or SA 193 B7 bolts with l SA-194 7 or A325 or A490 nuts); l The structural acceptance criteria for the ' steel components of the RCCV (i.e., the basis for (5) castings (SA 216, grade WCB or SA 352, grade establishing allowable stress values, the LCB, A27, or 7036); deformation limits, and the factors of safety) are established by and in accordance with ASME (6) cold finished steel (A108 grade 1018 to Code Section lit, Subsection NE. 1050); g Amendment 1 3814

ABWR u^6too^ts Standard Plant _. nnv. A (7) bar and machine steel (A576, carbon content All radiographic inspection is made in n v not less than 0.3%); and (8) clad (SA 240 type 304L). accordance with the requirernents of Article NE 5000 of AShfE Code Section 111 except as specificd herein. All shops welds requiring radiography are radiographed in the shop after The structural steel materials located beyond intermediate or final post weld heat treatment the containment vessel boundaries are as follows: (PWHT). Radiographic examination of field weldments is performed after any required final (1) carbon steel (A36 or SA 36) and PWHT. Welds of AShfE Code, categories C and D, which cannot be examined by radiography are (2) stainless steel extruded shapes (SA-479). examine 1in accordance with paragraphs NE 5232 and NE 5242 of ASME Code Section Ill. The materials meet requirements as specified in Subarticle NE 2000 of ash 1E Code Section Ill. 3.8.2.7.13 Ultrasonic Examination of Plates The lowest senice metal temperature is 300F. The ultrasonic examination of materials in the 3.8.2.7 Testing and Inservice Inspection steel components is performed in accordance with Requirements Subarticle NB 2530 of AShfE Code Section 111. Leakage of the containment vessel, including 3.8.2.7.1.4 hiagnetic Particle Examination the steel components is described in Subsection 3.8.1.7. hiagnetic particle examination is performed on carbon steel welds in accordance with Articles 3.8.2.7.1 Examination Requirements NE-5000 and CC 5000 of AShiE Code Section III. The magnetic particle acceptance standards are in The examination and testing of all materials accordance with NE-5340 and CC-5540 0f ash 1E Code including welding materials are performed in Section llt. O V accordance with the requirements of Articles NE 2000 and NE 5000 of AShfE Code Section Ill. 3.8.2.7.1.5 Liquid Penetrant Examination Nondestructive examinations are performed only by technicians qualified in accordance with Article Liquid penetrant examination is performed on NE-5000 of the AShtE Code. all stainless steel welds in accordance with NE 5000 and CC 5000 of the ash 1E Code. All 3.8.2.7.1.1 impact Testing discontinuities exceeding those allowed are removed and the weld repaired, as necessary, in Charpy V notch impact tests are made on ma- accordance with applicable procedures. I terial, weld deposit, and the base metal weld heat affected zone employing a test temperature 3.8.2.7.2 Shop Testing Requirements not higher than 00F, The requirements of Sub-article NE-2320 of ASNIE Code Section 111 are met The shop tests of the personnel air locks in-for all materials under jurisdiction of the clude operational testing and an overpressure  ! code. Impact tests of weld deposit and base me- test. After completion of the personnel air l tal weld heat affected zone are made for each locks tests (including all latching mechanisms I welding procedure requiring ash 1E Code Section IX and interlocks) each lock is given an operational j qualifications. test consisting of repeated operating of each door and mechanism to determine whether all parts i 3.8.2.7.1.2 Radiography are operating smoothly without binding or other l defects. All defects encountered are corrt cted All radiography is performed in accordance and retested. The process of testing, corre, ting with the requirements of Article NE 5000 of ash 1E defects, and retesting are continued unhl no Code Section 111. defects are detectable. O Amendment 1 3.8 15

ABWR msmu Standard Plant nim a For the operational test, the personnel air locks are pressurized with air to the maximum Figures 3.817 and 3.818 and Figures 1.2-2 g permissible code test pressure. All welds and through 1.213 show an overview of the W seals are observed for visual signs of distress containment including the internal structures. or noticeable leakage. The lock pressure is then reduced to design pressure and a thick bubble 3.8.3.1.1 Diaphragm Floor solution is applied to all welds and seals and observed for bubbles or dry flaking as indica- The diaphragm floor serves as a barrier tions of leaks. All leaks and questionable areas between the drywell and the suppression are clearly marked for identification and chamber,. It is a reinforced concrete circular subsequent repair, slab, with an outside diameter of 14.5 m (47 ft, 7 in), and a thickness of 1.2 m (3 ft,11 in). During the overpressure testing, the inner door are blocked with holddown devices to prevent The diaphragm floor is supported by the reac-unseating of the seals. The internal pressure of tor pedestal and the containment wall. The con-the lock is reduced to atmospheric pressure and nection of the diaphragm floor to the contain-all leaks are repaired. Afterward, the lock is ment wall is a fixed support. The diaphragm again pressurized to the design pressure with air floor connection to the reactor pedestal is a and all areas suspected or known to have leaked hinged support. The diaphragm floor is pene-during the previous test are retested by the trated by 18,508 mm (20 in) diameter sleeves bubble technique. This procedure is repeated for the SRV lines. Reinforcing 1.tcel and end until no leaks are discernible. Support details are provided in Appendix 311. 3.8.3 Concrete and SteelInternal A 1/4 inch thick, carbon steel liner plate is Structures of the Concrete provided on the bottom of the diaphragm floor, Containment and is anchored to it. The liner plate serves as a form during construction and prevents the 3.83.1 De cription of the Internal Structures bypass flow of steam from the upper drywell to the suppression chamber air space during a LOCA. g The functions of the containment internal structures include: support of the reactor vessel 3.83.1.2 Reactor Pedestal radiation shielding, support of piping and equipment, and formation of the pressure A composite steel and concrete pedestal pro-suppression boundary. The containment internal vides support for the reactor pressure vessel, structures are constructed of reinforced concrete the reactor shield wall, the diaphragm floor, and structural steel. The containment internal access tunnels, horizontal vents, and the lower structures include the following: dr>well access platforms. The pedestal consists of two concentric steel shells tied together by (a) Diaphragm floor vertical steel diaphragms. The regions formed by the steel shells and the vertical diaphragms, (b) Reactor pedestal except the vents and the vent channels, are filled with concrete. There are ten drywell (c) Reactor shield wall connecting vent (DCV) channels connecting the upper drywell to the lower drywell and the (d) Drywell and equipment pipe support ttructure horizontal vents. (c) Miscellaneous platforms The wetted portion of the exterior surface of the reactor pedestal steel shell in the (f) lewer drywell equipment tunnel suppression chamber is clad with stainless steel to provide corrosion protection. The extent of (g) Lower dr>weli personnet tunnel the cladding and the reactor pedestal configuration is provided in Appendix 311. (h) Reactor shield wall stabilizer Amendment 1 3 8-16

MM 23A6100All Signdard Plant RTV A 3.8.3.13 Reactor Shield Wall 3.83.1.5.2 Lower Drpell Equipment Tunnel The reactor shield wall is supported by the A steel tunnel is provided at azimuth 180 reactor pedestal and surrounds the reactor degrees for equipment access to the lower dry. pressure vessel. Its function is to attenuate well from the reactor building. The tunnel has radiation emanating from the reactor vessel. In an inside diameter of 4.3 m (14 ft,1 in), is 20 addition, the reactor shield wall provides rum (3/4 in) in thickness, and has a flanged structural support for the reactor vessel closure at the reactor building end. The wetted stabilizer, the reactor vessel insulation and the portion of the tunnel is stainless steel or drywell equipment and pipe stipport structure. carbon steel with stainless steel cladding. The Openings are provided in the shield wall to tunnel is attached rigidly to the containment permit the routing of necessary piping to the RPV wall at one end and the reactor pedestal at the and to permit in service inspection of the RPV other end and is partially submerged in the and piping. suppression pool at normal water level. The tunnel has one or two flexible rings to The shield wall is shaped as a right accommodate differential displacement of the cylinder. The shield wall consists of two containment wall and reactor pedestal. The concentric steel cylindrical shells joined configuration of the tunnel and the connection together by horizontal and vertical steel plate details at the containment wall and reactor diaphragms. Full depth stiffeners are provided pedestal are shown in Appendix 3H. Fine motion in the reactor shield wall at the attachment control rod drive piping is routed through the locations of major pipe supports, pipewhip tunnel. The tunnel permits entry from the restraints and beam supports. The annular region reactor building into the lower drywell without between the outer and inner shells is filled with exposure to the suppression chamber atmosphere. concrete. The arrangement of the reactor shield wall is provided in Appendix 3H. 3.83.1.53 Imer Drpell Personnel Tunnel 3.83.1.4 Drpell Equipment and Pipe Support The lower dgywell personnel tunnelis located Structure at azimuth 0 and is similar to the lower drywell equipment tunnel described in Subsection The drywell equipment and pipe support 3.8.3.1.5.2. However, it has a personnel lock structure (DEPSS) consists of various structural at the reactor building end. The arrangement components such as beams and columns. Built up and details of the tunnel are shown in Appendix box shapes are used for beams and columns that 3H. must resist torsion and biaxial bending. The beams span between the reactor shield wall and 3.83.1.5.4 Reactor Shield Wall Stabilizer the vertical support columns which are anchored to the diaphragm floor. The DEPSS provides Reactor shield wall stabilizers, fabricated support for piping, pipe whip restraints, from structural steel members, span from the top mechanical equipment, electrical equipment and of the reactor shield wall to the bottom of the general access platforms and stairs. The containment top slab as shown in Appendix 3H. arrangement and details of the DEPSS are shown in The connection of the stabilizer at the reactor Appendix 3H. shield wall is designed to allow vertical and radial movement of the shield wall relative to 3.83.13 Other Internal Structures the containment top slab and to provide support for the shield wall against seismically induced 3.83.1.5.1 Miscellaneous Platforms tangential movements. Miscellaneous platforms are designed to allow 3.83.2 Applicable Codes, Standards, and access and to provide support for equipment and Specifications piping. The platforms consists of steel beams ! and grating. The design of the concrete and steel Amendment 1 3817

ABM ux6ioasu Standard Plant Rnv.A internal structures of the containment conform to (2) Construction Loads Construction loads are the applicable codes, standards, and loads which are applied to the containment a specifications and regulations listed in Table internal structures from start to completion W 3.8 4 except where specifically stated otherwise, of construction. The definitions for D, L and Toare applicable, but are based on Structure or Specific Reference actual construction methods and /or Component Number conditions. Diaphragm Floor 14 (3) RV2- Loads from component response or direct fluid forces, on components located in the Reactor Pedestal 1 13, 15 22 suppression pool, caused by safety relief valve air cleaning loads. Reactor Shield Wall 1-13, 15 22 (4) RBV Loads due to reactor building DEPSS 15-22 vibrations caused by an SRV and LOCA event. Miscellaneous platforms 15 22 (5) AP Loads and pressures directly on the reactor shield wall and loads from component L/D Equipment Tunnel 15 22 response or direct steam flow forces on components located in the reactor vessel L/D Personnel Tunnel 15-22 shield wall annulus region, caused by a rupture of a pipe within the reactor vessel Reactor Shield Wall 15 22 shield wall annulus region. Stabilizer (6) SL--Loads from component response or direct 3.833 Loads and Load Combinations fluid forces, on components located in the sloshing zone of a pool or component, caused 3.833.1 Load Definitions by the sloshing phenomenon from any dynamic event. g The loads and applicable load combinations for which the structure is designed depend on the 3.833.2 Load Combination conditions to which the particular structure is subje cted. The load combinations and associated accep-tance criteria for concrete and steel internal The containment internal structures are de- structures of the containment are listed in Ta-signed in accordance with the loads described in ble 3.8.3-5 and Table 3.8.3 6, respectively; for Appendix 3B These loads and the effects of the reactor shield wall refer to appendix 3B, these loads are considered in the design of all internal structures as applicable. The loads 3.83.4 Design and Analysis Procedures within the loading combinations are combined using the absolute sum technique. (Those loads 3.83.4.1 Diaphragm Floor which are defined as reversible in algebraic sign are combined in such a way as to produce the The design and analysis procedures used for maximum resultant stresses in the structure. All the diaphragm floor are similar to those used other loads are combined in accordance with their for the containment structure. The diaphragm direction of application to the structure.) The slab is included in the finite element model loads are defined in Subsection 3.8.1.3 except as described in Subsection 3.8.1.4.1.1. follows: 3.83.4.2 Reactor Pedestal (1) P o-Pressure loads resulting from the normal operating pressure difference between The reactor pedestal is included in the the drywell (upper and lower) and the finite element model described in Subsection suppression chamber of the containment. 3.8.1.4.1.1. I l Amendment 1 3818 l l l 1 i

MN 23A6100AE Standard Plant REV.A l The design and analysis is based on the clas- respectively. tic method. All loads are resisted by the integ-O- ral action of the inner and outer steel shells. 3.83.6 hinterials, Quality Control, and The concrete placed in the annulus between the Special Construction Techniques inner and outer shells acts to distribute loads between the steel shells, and provides stability 3.83.6.1 Diaphragm Floor to the compression elements of the pedestal. The materials, quality control, and construc-3.83.43 Reactor Shield Wall tion techniques used for the diaphragm floor and liner plate are the same as those used for the The design and analysis procedures used for containment w-" nd liner plate in Subsection the reactor shield wall are similar to those used 3.8.1.6. for the reactor pedestal described in Subsection 3.8.3.4.2. 3.83.6.2 Reactor Pedestal

• 3JJ.4.4 Drywll Equipment and Pipe Support The materials conform to all applicable Sinscture iequirements of ANSI /AISC N690 and ACI 349 and comply with the following:

The drywell equipment ao pipe support struc- lits Soecification ture is designed using the AISC working stress methods for steel safety related structures for Inner and cuter shells ASThi A441 nuclear facilities (ANSI /AISC N690). The design of beams supporting pipe whip restraints allows Internal stiffeners ASThi A441 inelastic deformations due to postulated pipe rupture loads. All safety-related items which Concrete fill f' c = 4u00 psi the inelastic beam deformations may aff:ct are evaluated to verify that no required safety 3.8.3.6.3 Reactor Shield Wall function would be compromised. O The materials conform to all applicable 3.83.4.5 Other Internal Strvetures requirements of ANSI /ASIC N690 and ACI 349 and comply with the following The design and analysis procedures used for other internal structures are similar to those lism Sne-ification used for the drywell equipment and pipe support structure as described in Subsection 3.83.4.4. Inner and outer shells ASThi A441 3.83.5 Structural Acceptance Criteria Internal stiffeners ASThi A441 l 3.83.5.1 Diaphragm Floor Concrete fill f' c = 3000 psi The calculated and allowable stresses for the Stainless Steel Clad SA.240 Type 304 L diaphragm floor are found in Appendix 311. 3.83.6.4 Drymil Equipment and Pipe Support 3.83.5.2 Reactor Pedestal Structure The calculated and allowable stresses for the The mate:ials conform to all applicable ) reactor pedestal are found in Appendix 3H. requirements of ANSI /AISC N690 and comply with I the following: ) 3.8.53 Other Internal Structures lism Specification The structural acceptance criteria for other i internal concrete or steel structures are in Structural steel and ASTH 1 A36 e accordance with ACI 349 and ANSI /AISC-N690, connections Amendment 1 3&l9

l l ABWR u461oo41 Standard Plant mmA High strength structural ASTM A572 or A441 building and control building. Figure 1.21 steci plates shows the spatial relationship of these build- g ings. The only structures in the remainder of W Bc!ts, studs, and nuts ASTM A325 or A490 plant which are in close proximity to these (dia. .>_3/4 ') structures are the turbine building and radwaste facility. They are structurally separated from Bolts, studs, and nuts ASTM A307 the Nuclear Island buildings. (dis.13/4 *) The Seismic Category I structure within the 3.83.6.5 Other laternal Structures Nuclear Island, other than the containment structures, that contains high-energy pipes is The materials conform to all applicable the reactor building. The steam tunnel walls regtdrements of ANSI /AISC N690 and comply with protect the reactor building from potential the following- impact by rupture of the high. energy pipes. This building is designed to accommodate the lif.m Specification guard pipe support forces. Miscellaneous platforms Same as Section The reactor building, steam tunnel, residual 3.8.3.6.4 heat removal (RHR) system, reactor water cleanup (RWCU) system, and reactor core isolation cool. Lower drywell equipment ASTM A516 Grade 70 ing (RCIC) system rooms are designed to handle tunnel SA 240 Type 34t L the consequences of high energy pipe breaks. The RHR, RCIC, and RWCU rooms are designed for Imwer drywell personne! ASTM A516 Grade 70 differential compartment pressures, with the tunnel SA-240 Typc 304 L associated temperature rise and jet force. Steam generated in the RHR compartment from the Reactor shield wall stabilizer postulated pipe break exits to the steam tunnel

       -tube sections           ASTM A501 through blowout panels. The steam tunnel is vented to the turbine building through the g

seismic interface restraint structure (SIRS).

      --plates                  ASTM A36          The steam tunnel, which contains several pipe-lines (e.g., main steam, feedwater, RHR), is al-3.8J.7 Testing and Inservice inspection            so designed for a compartment differential pres.

Requirements sure with the associated temperature changes and jet force. A formal program of testing and inservice in-spection is not planned for the internal struc- Seismic Category I masonry walls are not used tures except the diaphragm floor, reactor pedes- in the design. The ABWR standard Nuclear Island tal, and lower drywell access tunnels. The other does not contain seismic Category I pipelines internal structures are not directly related to buried in soil, the functioning of the containment system; therefore, no testing or inspection is performed. 3.8.4.1 Description of the Structurrs Testing and inservice inspection of the dia- 3.8.4.1.1 Reactor Building Structure phragm floor, reactor pedestal and lower drywell access tunnels are discussed in Subsection The reactor building (RB) is constructed of 3.8.1.7. reinforced concrete with a steel frame roof. The RB has four stories above the ground level 3.8.4 OTHER SEISMIC CATEGORY I and three stories below. Its shape is a i STRUCTURES rectangle of 59 meters in the E W direction,56 l meters in the N-S direction, and a height of Other Seismic Category I structures which about 57.9 meters from the top of the basemat. constitute the Nuclear Island are the reactor l Amendment 1 3 8-20 l

ABWR mama Standard Plant RL%6 The Reinforced Concrete Containment Vessel the nearest wall or ceiling to take the seismic (RCCV) in the center of the RB encloses the loads. t Reactor Pressure Vessel (RPV). The RCCV supperts the upper pool and is integrated with the RB 3.8.4.2 Applicable Codes, Standards, and structure from the basemat up through the Speclilcations elevation of the RCCV top slab. The interior floors of the RB are also integrated with the . 3.8.4.2.1 Reactor Building RCCV wall. The RB has slabs and beams which join the exterior wall. Columns tupport the floor The major portion of the reactor building, is slabs and beams. The fuel pool girders are not subjected to the abnormal and severe acci-integrated with the RCCV top slab and with RB dent conditions associated with a containment, wall columns. The RB is a hear wall structure A listing of applicable documents follows: designed to accommodate all seismic loads with its walls. Therefore, frame members such as (1) ACI 349, Code Requirements for Nuclear beams or columns are designed to accommodate Safety Related Concrete Structures; deformations of the walls in case of carthquake conditions. (2) AISC, Specification for Design, Fabrication and Erection of Structural Steel for 3.8.4.1.2 Control Building Buildings; The control building (CB) is shown in Section (3) AShfE Boiler and Pressure Vessel Code Section 1.2. III, Subsection NE, Division 1, Class hf C (for design of main steam tunnel embedment The CB houses the electrical and control and piping anchorage in the RB and CB only); instrumentation equipment, the control room for the reactor and turbine buildings as well as the (4) AWS Structural Welding Code, AWS D1.1; CB HVAC equipment. (5) AWS Structural Welding Code, AWS D12.1; J The CB is a Seismic Category I structure designed to provide missile and tornado (6) NRC publications TID 7024 and TID 25021, protection and house operating personnel. Nuclear Reactors and Earthquakes and Summary Shielding requirements are satisfied in the of Current Seismic Design Practice for determination of concrete wall and slab Nuclear Reactor Facilities; thicknesses. (7) The inservice inspection requirements for 3.8.4.1.4 Seismic Category 1 Cable Tray and the fuel pool liners in the Reactor Building Conduit Supports are in conformance with AShtE Code Section 111, Division 2. Electrical cables are carried on continuous horizontal and vertical runs of steel trays (8) NRC Regulatory Guides: supported at intervals by structural steel frames. The tray locations and elevations are (a) Regulatory Guide 1.10, hicchanical predetermined based on the requirements of the (Cadweld) Splices iri Reinforcing Bars of electrical cable network. Generally, several Category 1 Concrete Structures; trays of different sizes are grouped together and connected to a common support. (b) Regulatory Guide 1.15, Testing of Reinforcing Bars for Category 1 Concrete The support frame spacing is determined by al. Structures; lowable tray spans, which are governed by rigidi-ty and stress. The frames may be ceiling sup- (c) Regulatory Guide 1.28, Quality Assurance ported, or wall-supported, or a combination of Program Requirements (Design and both. Various type of frames form a support sys- Construction); tem with transverse and longitudinal bracing to O Amendment t 3821

ABWR m aman Standard Plant REV,A (d) Regulatory Guide 1.29, Seismic Design Personnel for the Construction Phase of Classification; Nuclear Power Plants; g (c) Regulatory Guide 1.31, Control of (c) ANSI N45.2.9, Requirements for the Stainless Steel Welding; Collection, Storage, and hiaintenance of OA Records for Nuclear Power Plants; (f) Regulatory Guide 1.44, Control of the Use of Sensitized Stainless Steel; (f) ANSI N45.4, Leakage Rate Testing of Containment Structures for Nuclear (g) Regulatory Guide 1.55, Concrete Reactors; Placement in Category I Structures; (g) ANSI N101.2, Protective Coatings (b) Regulatory Guide 1.60, Design Response (Paints) for Light Water Nuclear Spectra for Seismic Design of Nuclear Reactor Containment Facilities; and Power Plants; (h) ANSI N101.4, Quality Assurance for (i) Regulatory Guide 1.61, Quality Assurance Protectiv: Coatings Applied to Nuclear Requirements for the Design of Nuclear Facilities; Power Plants; (10) Steel Structures Painting Council Standards l (k) Regulatory Guide 1.69, Concrete l Radiation Shields for Nuclea Power (a) SSPC PA-1, Shop, Field and hiaintenance Plants. Painting; (1) Regulatory Guide 1.76, Design Basis (b) SSPC PA 2, hicasurement of Paint Film Tornado. Thickness with hiagnetic Gages; (m) Regulatory Guide 1.142, Safety Related (c) SSPC-SP-1, Solvent Cleaning; Concrete Structures for Nuclear Pow:r Plants (Other than Reactor Vessels and (d) SSPC SP 5, White Metal Blast Cleaning; Containment); and (c) SSPC SP 6, Commercial Blast Cleaning; (n) Regulatory Guide 1.94, Quality Assurance and Requirements for Installation, inspec-tion, and Testing of Structural Concrete (f) SS?C-SP-10, Near White Blast Cleaning; and Structural Steel During the Con-struction Phase of Nuclear Power Plants. (11) ACI ASCE Crmmittee 326, Shear and Diagonal Tension, ACI hianual of Concrete Practice, (9) ANSI: Part 2; (a) ANSI A58.1, Building Code Requirements (12) Applicable ASThi Specifications for for hiinimum Design leads in Building and hiatcrials and Standards; and Other Structures; (13) AASHTO Standard Specifications for liighway (b) ANSI N5.12, Protective Coatings (Paint) Bridges for truck loading area. for the Nuclear Industry; (c) ANSI N45.2, Quality Assurance Program Requirements for Nuclear Power Plants; Refer to Subsection 3.8.4.2.1. (d) A NSI N45.2.6, Q ualifica tion s of Add NRC Rules and Regulations Title 10, Chap-Inspection, Examination and Testing ter 1, Code of Federal Regulations, Part 73.2 and 73.55. g Amendment 1 3 8-22

ABM 23AsiooAs Standard Plant REV,A 3.8.4.23 Seismic Category I Cable Tmy and Ro = pipe reactions during normal f3 Conduit Supports operating or shutdown conditions V based on the most critical (1) All codes, standards, and specifications transient or steady state applicable to the building structures shall condition also apply to cable tray and conduit supports. Ra = PI Pe reactions under thermal conditions generated by the (2) AISI, Specification for the Design of

                                        ~

postulated break and including Cold formed Steel Structural Members. Ro (3) NEMA, Fittings and Supports for Conduit and Yr = equivalent static load on a Cable Assemblies, structure generated by the reaction on the broken 3.8.43 leads and Load Combinations high energy pipe during the postulated break and including a 3.8.43.1 Reactor Building calculated dynamic factor to account for the dynamic nature of The temperature and pressure loads caused by a the load LOCA do not occur on the reactor building. The reactor building ventilation system is designed Yj = jet impingement equivalent static to keep the building within operating design load on a structure generated by conditions. the postulated break and including a calculated dynamic 3.8.43.1.1 Loads and Notations factor to ac count for the dynamic nature of the load. Loads and notations are as follows: p V' Ym = missile impact equivalent static D = dead load of structure plus any load on a structure generated by other permanent load or during the postulated break, like pipe whipping, and including L = conventional floor or roof life a calculated dynamic factor to loads, movable equipment loads, and account for the dynamic nature of other variable loads such as the load. construction loads. The following live loads are used: W = wind force (Subsection 33.1.) Concrete floors and slabs (including W g = tornado load (Subsection 3.3.2) roofs) 200 psf. Stairs, stair (Tornado. generated missiles are platforms, grating floors, and des::ribed in Subsection 3.5.1.4, platforms - 100 psf. Concrete and barrier design procedures in roofs, live or snow load (not Subsection 3.5.3). concurrent) 50 psf. Construction live load on floor framing in The pressure part of Wt I5 addition to dead weight of floor - defined as either (1) 360 mph 50 psf *. velocity pressure with or without

• If the actual construction live load is greater than this value a design check of the structures will be made.

A V Amendment 1 3.8 23

ABWR uuiman Standard Plant REV.A

1. 5 p s i negative pressure Ta = thermal effects (including To) differential, of (2) 3 psi negative which may occur during a design pressure differential. accident at 1650F maximum 30 minutes after LOCA h

Pa - internal negative pressure of 3.0 psig due to tornado; accident U = for concrete structures, the section pressure at main steam tunnel piping strength required to resist design embedment loads based on the strength design method described in ACI 318. B = uplift forces created by the rise cf the ground water table 11 = loads caused by static or seismic F = internal pressures resulting from carth pressures. flooding of compartments For structural steel, S is the required sec. E' = safe shutdown earthquale (SSE) loads tion strength based on the clastic design me-as defined in Section 3.7 thods and the allowable stresses defined in Part 1 of the AISC Specification for the Design, E = operating basis carthquake (OBE) Fabrication and Erection of Structural Steel for loads as defined in Section 3.7 B uildings. To = thermal cffeets load effects 3.8.43.1.2 load Combinations for Concrete induced by normal thermal gradients Members existing through the reactor building wall and roof. Both summer For the load combinations in this subsection, and winter operating conditions are where any load reduces the effects of other considered. In all cases the loads, the corresponding coefficient for that conditions are considered of long load shall be taken as 0.9 if it can be enough duration to result in a demonstrated that the load is always present or 3 straight line temperature gradient. occurs simultaneously with the other loads. W The temperatures are as follows: Otherwise, the coefficient for that load shall be taken as rero. (1) Summer operation: (1) Normal operating conditions The strength (a) air temperature inside building - design method is used and the following 1200F load combinations are satisfied: (b) exterior temperature - 115 F U =1.4 D + 1.7 L + 13 To + 1.7 Ro+ (2) Winter operation: 1.711 + 1.4 B (a) air temperature inside building - U - 1.4 D = 1.7 L + 13 To + 1.7 Ro+ 700F 1.711 + 1.9 E i (b) exterior temperatute - (-) 400F U =1.4 D + 1.7 L + 13 To + 1.7 Ro+ 1 (3) Winter shutdown 1.711 + 1.7 W I (a) air temperature inside building - For fluid pressure F, replace 1.711 by 1.7 500F F in the last two of the three equations  : (b) cxterior temperature - (-) 400 F above. For all cases as. constructed temperature is (2) Abnormal / extreme environmental conditions - 600 F The strength design method is used and the l l 9 Amendment 1 3.8-24

ABM 23^61oo^u Standard Plant REV.A followingload combinations are satisfied: (2) Abnormal / extreme emironmental conditions - The clastic working stress design method is v U =D + L + To + Ro+ H + B . used and the following load combinations are satisfied: U = D + L + To + Ro + H + E' 1.6 S = D + L + To+Ro + E' U =D + L + To+Ro+H 1.6 S = D + L + oT + Ro+W.g U = D + L + To + Ro+H+Wt 1.6 S = D + L + To + Ro+W+H t U =D + L + Ta+ Ra + 1.5 Pa + H 1.6 S = D + L + aT + Ra+P- a U =D + L + Ta + Ra + 1.25 Pa+H+ 1.25 E + (Yr + Yj + Ym) 1.6 S = D + L + Ta+Ra + 1.0 E + Pa+ (Yj + Yr + Ym)- U = D + L + Ta + R a+P+H a

           + E' + (Yr + Yj + Ym)                             1.6 S = D + L + Ta+Ra + E' + Pa+

(Y; + Yr+Ym)- 3.8.43.13 Load Combinations for Steel Members in all these load combinations, both cases of L having its full value or being completely (1) Normal operating conditions the clastic absent are checked. working stress design method is used for the following load combinations: 3.8.43.2 Control Building S=D+L Refer to the loads, notations, and combina-tions established in Subsection 3.8.4.3.1, S=D+L+E except that fluid pressure F, accident pressure Pa , and pipe break loads Yr , Yj, Ym do S=D+L+W not exist and the live loads are as follows: Since thermal stresses due to To and R o Ali concrete floors 400 psf l are present and are secondary and self-limiting i in nature, the following combinations are also Stairs, stait platforms, grating floors, and satisfied: platforms - 100 psf 1.5 S = D + L + To+R. o Roof live or snow load (non concurrent) 50 psf 13 S = D + L + To+Ro+E Construction live load on floor framing in 1.5 S = D + L + To + W. addition to dead weight of floor 50 psf' In all these load conditions, both cases of To = thermal effects. As. constructed L having its full value or being completely temperature is 600 F. The absent are checked. temperatures inside the building are as follows: i

• If the actual construction live load is greater than this value a design check of the structures will be made.

i O Amendment 1 3 8-25 l 4

ABWR ummu Standard Plant nuv.A Operating Conditions (1) Rigid Support with Flexible Tray, in this method, trays are modeled as flexible g Control room, summer 750F clastle systems and analyzed by the response W winter 700F spectrum method. The resulting reactions llVAC room, summer 950F are used for the design of the supports, winter 600 F Other areas, summer 750F (2) Flexible Support with Flexible Tray, in winter 750F t iis method, the composite system of trays and supports is modeled and analyzed by Shutdowu condition computer as a multidegree of freedom elastic system. The support motions can be Controt room, summer 800F prescribed by the appropriate floor response winter 500F spectrum. The resulting responses are used IIVAC room, summer 1(WF to obtain design loads for the supports, winter 500F Other areas, summer 900F 3.8.4.4.4.2 Conduit Supports winter 500 F The design and analysis of conduit supports 3.8.4.3.4 Seismic Category I Cable Tray are basically the same as for cable tray and Conduit Supports supports. As conduits are more flexible and have comparatively less dead load, a rigid Loads and load. combinations for the tray and support approach is used as described in method conduit supports shall use the same criteria as (1) of cable tray support design, for the building structures where the supports are located. 3.8.4.5 Structural Acceptance Criteria 3.8.4.4 Reactor Hullding Design and Analysis 3.8.4.5.1 Reactor llullding Procedures 3.8.4.5.1.1 General Criteria g 3.8.4.4.1 Reactor flullding The first criterion is that the reactor The design procedures for the reactor building building shall provide biological shielding for are discussed in Subsection 3.8.3. The analysis plant personnel and the public outside of the procedures for the reactor building are discuud site boundary. This criterion dictates the in Subsection 3.8.1 minimum wall and roof thicknesses. 3.8.4.4.4 Seismic Category 1 Cable Tray and The second criterion is that the reactor Conduit Supports building shall protect the reinforced concrete containment from environmental hazards such as 3.8.4.4.4.1 Cable Tray Supports tornado and other site proximity generated missiles. The shielding thicknesses are Wherever possible, the supporting frames for a sufficient for this purpose. tray or group of trays are designed to have adequate rigidity to avoid causing additional The reactor building provides a means for amplification of seismic acceleration transmitted collection of fission product leakage from the by the building structures. Where rigidity reinforced concrete containment following an cannot be achieved without an excessive increase accident. in support enember size, the design of the supports is tiien based on the amplified seismic The reactor building SGTS is designed to keep load obtained from the floor response spectra. the compartments surrounding the reinforced concrete containment at a negative pressure even Thus, two methods are used in design and after a LOCA. In order to achieve a msimum analysis cf cable tray supports. in. leakage rate of 50% per day under a pressure Amendment t 3826

ABM 2346icoan Standard Plant REV A differential of 6 mm of water, the reinforcing The containment structure foundation, defined f steel is designed to remain clastic during the as within the perimeter or the exterior surface Q SSE load combinations. of the containment structure, is integral with the reactor building foundation. The 3.8.4.5.1.2 Materials Criteria containment fourdation mat details are discussed in Subsection 3.8.1.1.1. Refer to the materials criteria estt,blished in 3.8.5 for the strength and materials requirements 3.8J.2 Applicable Codes, Standards and ,

for the reinforced concrete reactor building. Specifications j 3.8.45.2 Control Building The applicable codes, standards, i specifications and regulations are discussed in I Structural acceptance criteria are defined in Subsection 3.8.1.2 for the containment the AISC Specification and ACI 318 Code. In no foundation and in Subsection 3.8.4.2 for the case does the allowable stress exceed 0.9 F other seismic Category I foundations, where Fy is the minimum specified yield stress. The design criteria preclude excessive 3.8.5.3 Loads and Load Combinations  ;

deformation of the building. The clearances ' between adjacent buildings are sufficient to The loads and load combinations for the prevent impact during a seismic event. The containment foundation mat are given in ' tornado load analysis for this building is the S ubsection 3.8.1.3. The loads and load same as the analysis for the reactor building. combinations for the other seismic category 1 I structure foundations are given in Subsection 3.8.5 FOUNDATIONS 3.8.4.3. This section describes foundations for all The loads and load combint.tions for all ' seismic Category I structures of the Nuclear seismic Category I foundations examined to check O Island. against sliding and overturning due to carth-quakes, winds and tornados, and against flo- , 3.8.5.1 Description of the Foundations tation due to floods are listed in Table 3.8 7. l The foundations of the reactor building and The lateral earth pressure loads are shown in 1 control building are reinforced concrete mat Figure 3.8-19. foundations. The configuration and relative location of these foundations are shown in 3.8.5.4 Design and Analysis Procedures Appendix 3H. The foundations of seismic category 1 These two foundation mats are separated from structures are analyzed using well established each other by a separation gap of 2 meters (6 methods where the transfer of loads from the feet,6 inches) wide to minimize the structural foundation mat to the supporting foundation interaction between the buildings. media is determined by clastic methods. The reactor building foundation is a Bearing walls and columns carry all the rectangular reinforced concrete mat 56 m (83 vertical loads from the structure to the feet,8 inches) by 59 m (193 feet,5 inches) and foundation mat. Lateral loads are transferred 5.5m (18 feet) thick. The foundation mat is to shear walls by the roof and floor constructed of cast in place conventionally diaphragms. The shear walls then transmit the reinforced concrete. It supports the reactor loads to the foundation mat, building, the containment structure, the reactor pedestal, and other internal structures. The top The design of the mat foundations for the of the foundation mat is approximately 20.2 m(66 structures of the Nuclear Island involves feet, 3 inches) below grade. The reactor primarily determining shear and moments in the building foundation is shown in Appendix 311. reinforced concrete and determining the O i Amendment 1 3 8-27

ABWR uwman Standard Plant rum a interaction of the substructure with the The calculated and allowable factors of underlying foundation medium. For a mat safety of the Nuclear Island structures for foundation supported on soil or roch, the overturning, sliding, and flotation are shown in pertinent aspects in the design are to maintain Appendix 311. tl.c bearing pressures within allow.ible limits, particularly due to overturning forces, and to 3.M.6 Materials, Quality Control, and ensure that there is adequate frictional and Special Construction Techniques passive resistance to prevent sliding of the structure when subjected to lateral loads. The foundations of seismic Category I structures are constructed of reinforced The design loads considered in analysis of the concrete using proven methods common to heavy foundations are the worst resulting forces from industrial construction. For further discussion the superstructures and loads directly applied to see Subsections 3.8.1,6 and 3.8.4.6. the foundation mat due to static and dynamic load combinations. 315.7 Testing and Intmice Inspection Requirements The applicant will evaluate the capability of the foundation to transfer shear if a A formal program of testing and inservice waterproofing system is used on the foundation, inspection is not planned and is not required for the seismic Category I structures of the The standard Nuclear Island design is Nuclear Island. developed using a range of soil conditions as detailed in Appendix 3A. The applicant will determine variations of physical properties of the site specific subgrade materials. The applicant will calculate settlement of the foundations, differential settlement between foundations for the site specific foundations medium and will design safety related systems h (i.e., piping, conduit, etc.) for the calculated settlement of the foundations. The applicant will evaluate the effect of the site specific subgrade stiffness and calculated settlement on the design of the seismic Category I structures and foundations. A detailed description of the analytical and design methods for the reactor building foundation mat including the containment foundation, is included in Section 3.8.1.4. I 3.8.5.5 Structural Acceptance Criteria The main structural criterion for the l containment portion of the foundation is adequate strength to resist loads and sufficient stiffness to protect the containment liner from excessive i strain. The acceptance criteria for the I containment portion of the foundation mat is presented in Subsection 3.8.1.5. The structural acceptance criteria for the reactor building foundations are described in Subsection 3.8.4.5. O Amendment 1 382s

ABM 2346iOOAu Standard Plant REV.A Table 3.81 'o s LOAD COMBINATIONS, LOAD FACTORS AND ACCEPTANCE CRITERIA FOR THE REINFORCED CONCRETE CONTAINMENT (1), (2), (3), (4) Accep. tance 1 cad Comb natjqi1 load Condition Cntena (6) SRVO Description No. D L Pt' Po Pa Pi Ps Tt To Ta E E' W W' Ro Ra Y FL 1Y ADS ALL LOCA SERVICE Test 1 1.0 1.0 1.0 1.0 S Construction 2 1.0 1.0 1.0 1.0 S Normal 3 1.0 1.0 1.0 1.0 1.0 1.0 1.0 S FACTORED Severe 4 1.0 1.3 1.0 1.0 1.5 1.0 1.0 1.0 U Environmental 5 1.0 1.3 1.0 1.0 13 1.0 1.0 1.0 U Extreme 6 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 U Environmental 7 1.0 1.0 1.0 1.0 1.0 1.0 1.0 U Abnormal 8 1.0 1.0 13 1.0 1.0 1.0 Note 5 U 8a 1.0 1.0 13 1.0 1.0 1.0 1.0 Note 5 U 8b 1.0 1.0 13 1.0 1.0 1.0 1.0 Note 5 U 9 1.0 1.0 1.0 1.0 1.25 1.0 Note 5 U 9a 1.0 1.0 1.0 1.0 1.25 1.0 1.0 Note 5 U 9b 1.0 1.0 1.0 1.0 1.25 1.0 1.0 Noic 5 U 10 1.0 1.0 125 1.0 1.0 1.25 Note 5 U (3 10a 1.0 1.0 1.25 1.0 1.0 125 1.25 Note 5 U O 10b 1.0 1.0 1 25 1.0 1.0 1.25 1.25 Note 5 U Abnormal / 11 1.0 1.0 1.25 1.0 1.25 10 1.0 Note 5 U Severe lla 1.0 1.0 1.25 1.0 1 25 1.0 1.0 1.0 Note 5 U Environmental lib 1.0 1.0 1.25 1.0 125 1.0 1.0 1.0 Note 5 U 12 1.0 1.0 1.25 1.0 1 25 1.0 1.0 Note 5 U 12a 1.0 1.0 1.25 1.0 1.25 1.0 1.0 1.0 Noie 5 U 12b 1.0 1.0 1.25 1.0 125 1.0 1.0 1.0 Note $ U 13 1.0 1.0 1.0 1.0 1.0 U 14 1.0 1.0 1.0 1.0 1.0 U Abnormal / 15 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 Note 5 U i Extreme 15a 1.0 1.0 1.0 1.0 1.0 1.0 1.0 10 1.0 Note 5 U l Environmental 15b 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 Note 5 U i 1 i NOTES: l

1. The loads are described in Subsection 3.&1.3 and acceptance criteria in Subsection 3.&1.5.

2, For any load combination, if the effect of any load component, other than D, reduces the combined load then l the load component is deleted from the load combination.

3. Since Pa, Pi, Ps, Ta, SRVand LOC 4 are time dependent loads; their efects will be superimpo et' accordingly.

4. Load E in load combination 13 is based on thepost-accident recovery waterflood height in the containment.

5. LOC 4 loads, CO, CHUG, l'l.C and PS are time dependant loads. ne sequence of occurrence is given in Appendh 3B. De loadfactorfar LOCA loads shall be the same as the corresponding pressure load Pa, Pi or Ps.

6. )' includes )], i'm and i'r.

7. De sequence of occurrence ofSRVloads is given in Appendh 3B.

lv (may be 1 valve 2nd pop), ADS and ALL are not concurrent, when they are indicated in the load y combination. Amendment 1 3 8-29

1 ABWR 334amac i Standard Plant nitv. A l Table 3.8 2 MAJOR ALLOWABLE STRESSES IN CONCRETE AND REINFORCING STEEL Concrete Reinforcing Steel Compression Shear Tension Service 2.4ksi (1) 80 psi (Max allowable 30ksi lead tangential shear stress carried Combination by hoop and meridional rebars) (2) Normal shear 5 greater of 45ksi (For test 3/f e' or 1.5vc pressure case) Factored 3.4ksi (1) 160 psi (Max. allowable 54ksi lead tangential shear stress carried Combination by boop and meridional reinforce-ment) (2) Nominal shear 5 8/fe' h

                                                    = 506 psi l

I l l 9 l Amendment i 3 8-M i I

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

1 l 23A6100AE Standard Plant any A i Table 3.8.3 l !O l statss isressirv 'imits i i Primary & - Primary Stresses Gen. Mem. Local Mem. Bending & Local Mem. Seconda 7 ) Stresses , Pm PL Pj + pt PL + Pj + Q Test Condition 0.75 Sy 1.15 Sy 1.15 Sy N/A

                          . Design Condition              1.0 Sm (1)                       1.5Sm                        1.5Sm                   N/A                                 !

Post LOCA Flooding The larger of The larger of The larger of 3 Smi(1) 1.2 Smc or 1.0 Sy 1.8 Smc or 1.5 Sy 1.8 Smc or 1.5 Sy 1 NOTES

1. The allowable stress intensity Sm is the Sm listed in Table 110.0 and Sy is the yield i strength listed in Tcble 12.0 of Appendix 1 of the ASME Code, Section 111.

!O i  ! i j } 1 I t !O Amendment 1 3831 1 1 I l ,. - - . - , _ . - . - .-_ .- . _ _ _ , - _ _ - . . . . . . - _ . - - - - . -. - - . - - _ . - - -

ABWR m6imt , Standard Plant um A Table 3,8-4 CODES, STANDARDS, SPECIFICATIONS, AND REGULATIONS USED IN Tile DESIGN AND CONSTRUCTION OF SEISMIC CATEGORY l h INTERNAL STRUCTURES OF Tile CONTAINMENT SPECIFICATION SPECIFICATION REFERENCE OR STANDARD NUMliER DESIGNATION TITLE 1 ACI 301 Specifications for Structural Concrete for Builders 2 ACI 307 Recommended Practice for Concrete Formwork 3 ACI 305 Recommended Practice for llot Weather Concreting 4 ACI 211.1 Recommended Practice for Selecting Proportions for Normal Weight Concrete 5 ACI 315 hianual of Standard Practice for Detailing Reinforced Normal Weight Concretc 6 ACI 306 Recommended Practice for Cold Weather Concreting 7 ACI 309 Recommeeded Practice for Consolidation of Concrete g S ACI 308 Recommended Practice for Curing Concrete 9 ACI 212 Guide for Use of Admixtures in Concrete 10 ACI 214 Recommended Practice for Evaluation of Compression Test Results of Field Concrete 11 ACI 311 Recommended Practice for Concrete inspection 12 ACI 344 Recommended Practice for Measuring, Mixing. Transporting, and Placing Concrete 13 ACI 349 Code Requirements for Nuclear Safety Related Concrete Structures 14 ACI ASME-359 ASME Iloller and Pressure Vessel Code, Section 111, Division 2, Concrete Reactor Ves.els and Containments Amendment 1 3 L32 i I I I

ABWR man Standard Plant any. A p Table 3.8-4 V CODES, STANDARDS, SPECIFICATIONS, AND REGULATIONS USED IN TIIE DESIGN AND CONSTRUCTION OF SEISMIC CATEGORY I INTERNAL STRUCTURES OF TIIE CONTAINMENT (Continued) SPECIFICATION SPECIFICATION REFERENCE OR STANDARD NUMBER DESIGNATION TIRE 15 ANSI /AISCN690 Specification for the Design, Fabrication, and Erection of Steel Safety-Related Structures for Nuclear Facilities 16 AWS D1.1 Structural Welding Code 17 NCIG-02 Visual Weld Acceptance Criteria for Structural Welding at Nuclear Power Plants 18 ANSI /ASME Quality Assurance Program Requirements for NOA 119S6 Nuclear Facilities 19 ANSI /ASME Quality Assurance Requirements for NOA 219S6 Nuclear Facilities 20 NRC Regulatory Quality Assurance Requirements for Installation, O V Guide 1.94 Inspection, and Testing of Structural Concrete and Structural Steel During the Construction Phase of Nuclear Power Plants 21 NRC Regulatory Materials for Concrete Containments Guide 1.136 (Article CC-2000 of the Code for Concrete Reactor Vessels and Containments) 22 NRC Regulatory Safety Related Concrete Structures for Nuclear Power Guide 1.142 Plants (Other than Reactor Vessels and Containments) Etolanation of Abbreviations ACI American Concrete Institute AISC American Institute of Steel Construction ] i AISI American Iron and SteelInstitute j ANSI American National Standards Institute ASME American Society for Mechanical Engineers AWS American Welding Society NCIG Nuclear Construction issues Group NRC Nuclear Regulatory Commission NOTES:

1. Unless specified, the Edition of the Specificat.~on or Standard shall be the latest issued l p for industry use.

h l Amendment 1 3 8-33 l

TABLE 3.&5 LOAD COMBINATION, LOAD FACTORS AND ACCEL'TANCE CRITERIA y FOR REINFORCED CONCRETE STRUCTURES INSIDE TIIE CONTAINMENT (!),(2) 50 > S

                                                                                                                                                              $W E

e &,$ Q . LOAD COMBINATION LOAD CONDITION ACCEFTANCE  :: CRITERIA (3) $

                                                                                                                                                              ~

00 sot 1P!YN No. D L P, P, P. Pg P. T T, T. E E' W W R, R, M SR M N I' ADS All Tees 1 1.0 1.0 1.0 11, 5 Nermet 3 1.0 1.0 1.0 1.0 1.0 1.0 1.0 U 3a 1.4 1.7 1.0 1.7 1.7 1.7 U 3b l.05 1.3 1.0 1.3 1.3 1.3 1.3 U Sevev de 1.4 1.7 3.0 1.9 5.7 1.7 1.7 U Eavseesmeenal 4b I.05 1.3 1.0 1.3 1.4 1.3 1.3 1.3 U Sa I .4 1.7 1.0 1.7 1.7 1.7 1.7 U Sb I.05 1.3 1.0 3.3 1.3 1.3 1.3 1.3 U Entrenne 6 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 U Envreesmeenal 7 1.0 1.0 1.0 1.0 1.0 1.0 1.0 10 U Abeormal S 1.0 1.0 1.5 1.0 1.0 1.25 Nees 4 U Sa 1.0 1.0 1.5 1.0 1.0 1.25 1.25 *

                                                                                                                                                   *U sb   I.s  3.0                                     3.5                   1.0                           1.0     1.25  1.25     *

• Absormalf II 1.0 1.0 3.25 1.0 1.25 1.0 1.0 1.0 * - U 5evere lla 10 1.0 1.25 1.0 1.25 3.0 1.0 1.0 1.0 -
• U Envremaaemaal lib I .0 1.0 1.25 1.0 1.25 1.0 1.0 1.0 1.0 - " U Absormal I5 1.0 10 1.0 1.0 1.0 1.0 1.0 1.0 - - U Extreme 15e 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 *
• U Eavironmental 15b 1.0 1.0 10 1.0 1.0 1.0 1.0 1.0 1.0 *
• U Notes:

1. The lands are descnhed in Subseaan 3.833.

2. (Same as Nose 2.TaNe 3.84)

3. S = nxpared sumgth so resist servia loads per ASME Code SecUm III Dev. 2 M U = requind strength so resist faaored loads per ACI 349 E$

4. W

?' 3. (Same as Note 5. TaNe 3.8-1) (Same as Note 6. TaNe 3.8-1)

                                                                                                                                                              >g   :--

U 6. (Same as Nose 7. TaNe 3.8-1) O O @

O O TAILE 3.R-6 O LOAD COMBINATION, LOAD FACTORS AND ACCEPTANCE CRITERIA fe FOR STEEL STRUCIURES INSIDE THE CONTAINMENT (1),(2)

                                                                                                                                                                                                                    =$

3 E a In 2 LOAD COMBIN ATION LOAD CONDITION ACCElTANCE $ CRITERIAII DE30tFTKN No. D L P, P, Pg P, T. T, E E W W~  % R, p5) SRV(6) IJDCA l' ADS AIL Newmat i 1.0 1.0 1.0 3 2 1.0 1.0 1.0 1.0 1.0 1.0 1.0 M Scree 3 1.0 1.0 1.0 1.0 1.0 1.0 3 Env; n: 4 1.0 1.0 1.0 1.0 1.0 1.0 3 < 5 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 SD 6 1.0 1.0 I .0 1.0 I .0 1.0 1.0 1.0 SM Exaresse 7 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.65 Enviresuunnsal S 3.0 1.0 3.0 1.0 1.0 1.0 1.0 1.0 1.65 Abnersnal 9 1.0 3.0 1.0 1.0 1.0 1.0 Note d 1.63 9a 1.0 1.0 1.0 1.0 1.0 1.0 1.0

• 1.65 96 1.0 1.0 1.0 1.0 3.0 1.0 1.0
• 1.65 Absormal/ 10 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0
• 1.65 Seves 10s 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0
• 1.05 Eav;_ 106 1.0 1.0 3.0 1.0 1.0 1.0 1.0 1.0 1.0
• I.65 Abasemal 11 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0
• I.75 Entreme 1.0 1.0 1.0
• Ita 1.0 1.0 1.0 1.0 1.0 1.0 1.75 Enviromermaal Ilb 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 "

1.75 Nees:

1. Same as &se 1. Table 3.8-5).

2. Since P., T., SRV, and IDCA are timeW loads.deir effeas wiB he superuryoned accordingly.

3. Allowable Elastic Working Stress (S)is the allowable stress lanit specified in PartIof ANSI /AISCN690 g

4. (Same as &se 5.TaNe 3.8-1) y c:

      ,                                         5.     (Same as &se 6, Table 3.81)                                                                                                                                 .<g r                                         ik     (Same as &se 7 Table 3.8-1)                                                                                                                                 > c.,

d 7. For primary plus secondary stress. ihe eBownide Innits are increased by a factor d 1.5

ABWR memm Standard Plant nix A Table 3.8 7 LOAD COMillNATIONS FOR FOUNDATION DESIGN h LOAD COMBINATION NO. LOAD CONDITION D L U E E E E E E 1 1.0 1.0 1.0 1.0 1.0 1.0 2 1.0 1.0 1.0 1.0 3 1.0 1.0 1.0 1.0 1.0 4 1.0 1.0 1.0 1.0 5 1.0 1.0 Nomenclature: D Dead Lead F Bouyant Force of Design Ground Water P Bouyant Force of Det.ign Basis flood O 11 lateral Earth Pressure L Live lead E Basic OBE Seismic load (factored from SSE analysis) E' Basic SSE Seismic Lead W Wind lead

                  %"       Tornado Wind NOTE:

load combinations I and 3 shall be esuluated [<w two cases where:

1. hw load is considered to haw infull suhte and

2. hw loan is considered comp!cte& absent l

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