ML20128B705

Amend 25 to ABWR Ssasr

ML20128B705
Person / Time
Site:	05200001
Issue date:	01/29/1993
From:	GENERAL ELECTRIC CO.
To:
Shared Package
ML20128B393	List: ML20128B392 ML20128B705
References
NUDOCS 9302030153
Download: ML20128B705 (116)

Text

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.1 ABWR- -

n^am^c Standard Plant nrw. c CHAPTER 11 f1'd

. TdBLE OF CONTENTS Scrilan Title.-

Bige 1-INTRODUCTION AND GENERAL DESCRIPTION OF PLANT-'

- 1.1 JNTRODUCT13N 1.1 1 1.1.1 Format & Contents 1.11-1.1.2.

ABWR Standard Plant Scope L1-1:

1.13 Engineering Documentation 1.1.1 L1.4 Type of License Required 1.11 1.1.5 Number of Plant Units 1.11 1.L6 Description of Location 1.1 1 1.1.7 Type of Nuclear S; cam Supply 1.1 1 1.1.8 Type of Containment

- 1.1 1.

1.1.9 Core Thermal Power Levels -

1.1 1 =

~

1.2 GENERAL PLANT DESCRIPTION 1.2 1 1.2.1 Principal Design Criteria L2 L2.2 Plant Description 1.2 13 COMPARISON TABLES 13 13.1 -.

Nuclear Steam Supply System Design Characteristics 13 13.2 Engineered Safety Features Design Characteristics '

13-1 133 Containment Design Characteristics 13-1 13.4 Structural Design Characteristics 13 a 13.5

' Instrumentation and Electrical Systems Design Characteristics 13-1 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS 1.4 1 1-il Amendment 8 -

'9302030153-930129 PDR-.ADOCK 05200001 A

PDR

ABWR

. 2munc Standard Plant nov.c CIIAPTER 1 TAllLE OF CONTENTS (Continued)

Secilon Title Eage 1.5 REOUIREh1ENTS FOR FURTilEILTECIINICAL INFORNIATIOS 1.5-1 1.6 M ATERI AL INCORPOR ATED IW REFERENCE 1.6 1 1.7 DRAWINGS 1.71 1.7.1 Piping and Instrumentation and Process -

Flow Drawings 1.7.2 Instrument, Control and Electrical Drawings 1,7-1 1.73 ASME Standard Units Metric Conversion 1,71-Factors 1.7.4 Metric Corwersion to ASME Standard Units 1,71 1.7.5 Drawing Standards 1.7-1 l

1.7.6 COL License Information 1.7-1 1.8 CONFORA1ANCE WITII STANDARD REV1EW PLAN AND APPLICAlllLITY OF CODES AND STANDARDS 1.8-1 1.8.1 Conformance With Standard Resiew Plan 1.8-1 1.8.2 Applicability of Codes and Standards 1.8-1 1.83 COL License Information 1.8-1 1.9 COL L.lCENSE INFORhl ATION 1.9-1 APPENDIX 1 A RESPONSES TO TMI RELATED MATTERS 1 iii Amendment 25

ABWR 23MMAC Standard Plant nev e 1.2.2,13 3 Isolated Phase Ilus Fiber optic dataways are not restricted to raceway

/ T classifications, but would generally be run with V

The isolated phase buses duct system provides control cables due to their common destinations.

electrical interconnection frorr. the main generator output terminals to the low voltage generator 1.2.2.13.9 Grounding Wire breaker and from the low voltage generator breaker to the low voltage terminals of the main transformer, Grounding wire is summarized in Subsection and the unit auxiliary transformers. During the time 83.1.1.6.2.

the main generator is off line, the low voltage generator breaker is open and power is fed to the 1.2.2.13.10 Electrical Wiring Penetration unit auxiliary transformers by back feeding from the main transformer. During startup the generator Electrical wiring penetrations are summarized breaker is closed at about 7% power to provide in subsection 83.14.1.2(7).

power to the main and the unit auxiliary transformers for normal operation of the plant.

1.2.2.13.11 Combustion Turbine Generator A package cooling unit is supplied with the The primary function of the combustion turbine isolated bus duct system.

generator (CTG) is to act as a standby on-site l non-safety power source to feed permanent 1.2.2.13.4 Non. Segregated Phase Bus non safety loads during loss of offsite power (LOOP) events.

The non. segregated phase bus provides the electrical interconnection between the unit auxiliary The unit also provides an alternate AC power transformers and their associated 6.9kv metal-clad source in case of a station blackout event, as switchgear, defined by Appendix B of Regulatory Guide 1.155.

1.2.2.13.5 Metalclad Switchgear 1.2.2.13.12 Direct Current Power Supply

!g!

U The metal-clad switchgear distributes the 6.9kv The plant has four independent Class 1E i

power. Circuit breakers are drawout type, stored 125-volt de power systems.

energy vacuum breakers. The switchgear interrupting rating shall be determined in accordance 1.2.2.13.12.1 Unit Auxillary DC Power System with requirements of ANSI C37.10.

The unit auxiliary DC power system supplies 1.2.2.13.6 Power Center power to unit DC loads that are nonsafety-related.

The system con ists of two battery chargers, two The power center is summarized in Subsection batteries, two motor control centers, and two 83.1.1.2.1.

distribution panels.

1.2.2.13.7 Motor Control Center 1.2.2.13.12.2 Unit Class IE DC Power System The motor control center is summarized in The unit Class 1E DC power system supplies Subsection 83.1.1.2.2.

125 VDC power to the unit Class IE loads.

Battery chargers are the primary power sources.

1.2.2.13.8 Raceway System The system, which includes storage batteries that serve as standby power sources, is divided into four The raceway system is a plant wide network divisions, each with its own independent comprised of metallic cable trays, metallic conduits distribution network, battery, charger, and and supports. Raceways are classified for carrying redundant load group.

medium voltage power cables, low voltage power cabies, controI cabies and Iow Ieve1 1.2.2.13.13 Emergency Diesel Generator System signal / instrumentation cables. Didsional cables are O

routed in separate cable raceways for each division.

The emergency diesel generator system is l C supplied by three diesel generators. Each Class 1E l

l.

l Amendment 25 1.2-16.3 t

ABWR m6mc Standard Plant Rm C division is supplied by a separate diesel generator.

1.2.2.13.17 IJghting and Servicing Power Supply There are no provisions for transfer *ing Class 1E buses beiween standby ac power supplies or The design basis for the lighting facilities is the supplying more than one engineered safety feature standard for the Illuminating Engineering Society.

(ESP) from one diesel generator. This one-to-one Special attention is given to areas where proper relationship ensures that a failure of one diesel lighting is imperative during normal and generator can effect only one ESF division. The emergency operations. The system design diesel generators are housed in the reactor building precludes the use of mercury vapor fixtures in the which is a Seismic Category I structure, to comply containment and the fuel handling areas. The with applicable NRC and IEEE design guides and normallighting systems are fed from the unit criteria.

auxiliary transformers. Emergency power is supplied by engineered safety buses backed up by 1.2.2.13.14 Vital AC Power Supply diesel generators. Normal operation and regular simulated offsite power loss tests verify system 1.2.2.13.14.1 Safety System Protection System integrity.

Power System 1.2.2.14 Power Transmission Systems Four divisions of the safety system logic and control (SSLC) power system provide an 1.2.2.14.1 Reserve Transformer uninterruptible Class IE source of 120-VAC single phase control power. The primary power source for The reserve auxiliary transformer provides the the SSLC power system is the Class 1E AC power alternate preferred feed for the safety-related system. On loss of AC power, the appropriate buses M/C, C, D, & E. It also provides an divisional battery immediately assumes load without alternate feed to 6.9kv bus M/C B1 which supplies interruption. When AC power is restored. It the "B' train for plant imestment protection loads.

resumes the load without interruption.

The "A" train plant investment protection load alternate feed is from the combustion turbine via 8

1.2.2.13.14.2 Uninterruptible Pour System 6.9ky bus M/C A1.

W The uninterruptible power system (UPS) 1.2.2.15 Containment and Emironmental supplies regulated 120 VAC single phase power to Control Systems non Class 1E instrument and control loads which require an uninterruptible source of power. The 1.2.2.15.1 Primary Containment System power sources for the UPS are similar to those for the SSLC, but are non-Class 1E.

The primary containment system design for this plant incorporates the drywell/ pressure 1.2.2.13.14.3 Reactor Protection System suppression feature of previous BWR containment Alternate Current Power Supply designs into a dry containment type structure. In fulfilling its design basis as a fission product The reactor protection system alternate current barrier, the primary containment is a low leakage power supply is summarized in Subsection structure even at the increased pressures that could 8.3.1.1.4.2.2.

follow a main steamline rupture or a fluid system line break.

1.2.2.13.15 Instrument and Control Power Supply The main features of the containment design The instrument and control power supply include:

provides 120 VAC single phase power to instrument and control loads which do not require an (1) the drywell, a cylindrical steel lined uninterruptible power source.

reinforced concrete structure surrounding the reactor pressure vessel (RPV);

,1.2.2.13.16 Communication System (2) a suppression pool filled with water which The communication system is summarized in serves as a heat sink during normal Subsection 9.5.2.

operation and accident conditions; Amendment 20 t.216A

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n LABWRL tum.c iStandard Plant --

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- SECTION 1.7..

_m a y i/

-CONTENTS Secilon Thic Page 1.7.1 Pininn and Instrumentation and Process riow Drawines -

1.71) 1.7.2 Instrument. Control and Electrical Drawinns.

1.71-

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~

1.7.3 -

ASME Standard Units Metric Conversion Factors

-1.71'-

1.7.4

-Metric Conversion to ASME Stnnard Units 3 1.71 1.7.5 Drawinn Standards

- 1.7 1 --

l 1.7 6 COL License Information 1.71-SECTION 1.7 TABLES Table Htl.c Page p

V 1.71 Piping and Instrumentation and Process Flow -

Diagrams

, 1.72-1.7-2 Instrument Engineering, Interlock Block and Single Line Diagrams - 1.7-5 1.7-3 ASME Standard Units Metric Conversion Factors 1.75.2 1.7-4 Conversion Tables Metric to ASME Stanard Units

' 1.75.5 1.7-5 Drawing Standards 1.7-5.7 ILLUSTRATIONS Figure Titis Page

-1.7-1 Piping and Instrumentation Diagram Symbols

- 1.7-6 1.72 GraphicalSymbols for Use in IBDs.

1.7-8 A

1.7 u 1

QJ Amendment 25

- L T

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o ABWR m6mc Standard Plant nev e 1.7 DRAWINGS I

e 1.7.1 Piping and Instrumentation and Process Flow Drawings Table 1.7.1 contains a list of system Piping and Instrumentation diagrams (P&lD) and process flow diagrams (PFD) provided in the ABWR SSAR.

Figure 1.71 defines the sysmbols used on these drawings.

1.7.2 Instrument, Control and Electrical Drawings i

Interlocking block diagrams (IBD), instrument engineering diagrams (IED) and single line diagrams (SLD) are listed in Table 1.7-2. Figure 1.7-2 defines the graphic symbols used in the IBDs.

1.7.3 ASME Standard Units Metric Conversion Factors The ASME Standard units are applied with the numerical values converted to the metric system as listed in Table 1.7-3.

,o 1.7.4 Metric Conversion to ASME (J

Standard Units Selected flow, pressure, temperature and length metric units are converted to ASME standard units as tabulated in Table 1.7-4.

1.7.5 Drawing Standards Guidelines for identifying systems, facilities, equipment types and numbers and for drawing P&lD's and FPD's are treated in Table 1.7-5.

1.7.6 COL License Information COL applicants shall complete P&lD pipe schedules indicated as: COL applicant.

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Amcodment 25 1,7-t

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. 23A6100AC '

p Standard Plant -' ~

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My SECTION 1.8L u X-Ul:

CONTENTS--

Sectlon -

Iille Eage

e 1.8.1

- Conformance With Standard Review Plan <

1&1)

[

.i 1.8.2 Applicability of Codes and Standards ~

'1.8-l'-

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1.8.3 Applicability of Experience Information 1.8-1 i l'

1.8.4 COL License Information -

--1.8 14 TABLES:

Table 11112 Eage 1.8-1 Summary of Differences From SRP Section 1 1.8 1&2 -

Summary of Differences From SRP Section 2' 1 & 31 1.8-3

_ Summary of Differences From SRP Se'etion 3.

1.8-4 1.8-4 Summary of Differences From SRP Section 4 -

~ 1.8$ H-6 1.8-5 Summary of Differences From SRP Section 5 ?

1.8-6 U

1.8-6 Summary of Differences From SRP Section 6 '

1.8-7, 1&7 Summary of Differences From SRP Section 7 1.8-8 c 1.88 Summary of Differences From SRP Section 8

- 1&9 '

1.8-9 Summary of Differences From SRP Section 9 -

1.8-10 =

4,

1.8 10 Summary of Differences From SRP Section 10 '

1.8-1 11 1

1.8-11 Summary of Differences From SRP Section 11:

1.8 '

1.8 12-Summary of Differences From SRP Section 12 -

1.8-13 1.8-13 Summary of Differences From SRP Section 13 1 & 14-r 1.8 - Summary of Differences From SRP Section 14 --

1.8-15i 1

1.8-15 Summary of Differences From SRP Section 15

- 1&l6?-

' ISII

~ Amendment 25 -

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ABWR usamac Standard Plant nrv. c SECTION L8 9

TAllLES Table 11tk 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&l8 1.8-18 Summary of Differences From SRP Section 18 1.8-19 1.8-19 Standard Review Plans and Branch Tecimical Positions Applicable to ABWR 1.8-20 1.8-20 NRC Regulatory Guides Applicable to ABWR 1.8 1.8-21 Industrial Codes and Standards Applicable to ABWR 1.8-51 1 & 22 Experience Information Applicable to ABWR 1.8/;7 9

1&ili Amendment 12

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Standard Plant ium e SECTION 1.9

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CONTENTS Ecetton Ilile Eage 1.9 COLLiginse informatum -

1.91 TABLES Table Ill!c Page 1.9-1 Summary of ABWR Standard Plant COL License Information 1.9-2

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1.9 -11 Amendment 25

^

ABWR 23A6100AC Standard Plant va c 1.9 COL LICENSE INFORMATION

~

U The ABWR SSAR presents the AEWR Stan-aed !" a' doe 8.i incorporating the nuclear islaad, -

turbine island and radwaste f acility. Althoughuis scope is essentially a total plant, there is a modest amount of information that must be addressed by the COL applicant. The purpose _ of this section is to identify the SSAR sections where descriptions of the COL license information are presented.

The COL licensi information is summarized in Table 1.9-1 in :he order they are presented in the SSAR. An item number has been essigned to each r3 entry to facilitate future identification.

N]

,o Amendment 25 g.

o ABWR m uwxc-Standard Plant' am e -

Table I.91" N )/J y

SUMMARY

OF ABWR STANDARD PLANT INTERFACES

COL LICENSE INFORMATION?

ITEM -

1NTERFACE NO.

SUBJECT.

~

TYPE SURSECTION ~

1.1 Standard review plan r,ections for remainder of Confirmato_ry-1.8.4 plant identified as " Interface" in Table 1.F.19 1.2 Applicability of regulatory guldes for remainder Confirmatory '1.8.4 of plant indentified as " Interface" in Table 1.8-20 13 Applicability of Experience Information Confirmatory /1.8.4 for remainder of plant identified as Procedural

' Interface

• in Table 1.8 1.4 Emergency procedures and emergency procedures Procedural

- 1A3.1 training program 1.5 Procedures for removing safety-related systems -

Procedural

.1A3.2 -

from service 1.6 Inplant radiatiom momtoring.

Procedural 1A33 1.7 Reporting of Failures of Reactor System Procedural 1 A3.5 -

Rclief Valves 1.8 Report on ECCS Outage Procedural

- IA/>.6 2.1 Envelope of ABWR Standard Plant Site Design - Design &

2.2.1 -

Parameters Confirmatory 2.2 Standard Review Plan Site Character'.stics Confirmatory 2.2.2 23 CRAC 2 Computer Code Calculations Confirmatory 2.23 3.1 Site-Specific Design Basis Wind Confirmatory 333.1.

3.2 Site-Specific Design Basis Tornado Confirmatory 333.2 e

33 Effect of remainder of plant structures, Confirmatory 3333 systems and components not designed to tornado loads 3.4 Flood Elevation Design 3.43.1 3.5 Ground Water Elevation Design 3.43.2 3.6 Protection of ultimate heat sink Confirmatory 3.5.4.1 Amendment 25.

1.9 -

w.

ABWR 23umc Standard Plant wc Table 1.91

SUMMARY

OF ABWR STANDARD PLANT INTERFACES l

COL LICENSE INFORMATION (Continued).

ITEM INTERFACE NO.

SUlijECT TYPE SUllSECTION 3.7 Missies generated by natural phenomena from Confirmatory 3.5.4.2 remainder of plant 3.8 Site proximity missiles and aircraft hazards Confirmatory 3.5.43 3.9 Protection against secondary missiles inside Confirmatory 3.5.4.4 containment 3.9a Impact of Failure of Non Safety-Related Confirmatory 3.5.4.5 Items Due to Design Basis Tornado 3.10 Details of pipe break analysis results Confirmatory 3.6.4.1 and protection methods 3.11 Leak-before-break analysis results Confirmatory 3.6.4.2 3.11a Seismic Parameters Confirmatory 3.7.5.1 3.12 Foundation Waterproofing Confirmatory 3.8.6.1 3.13 Site Specific Physical Properties and Confirmatory 3.8.6.2 Foundation Settlement 3.14 Reactor Internals Vibration Analysis, Confirmatory 3.9.7.1 Measurement and Inspection Programs 3.15 ASME Class 2 or 3 Quality Group Confirmatory 3.9.7.2 Components with 60 Year Design Life 3.15a Pump and Valve Inservice Testing Program Confirmatory 3.9.7.3 3.15b Audits of Design Specifications and Design Reports Confirmatory 3.9.7.4 3.16 Equipment qualification report Confirmatory 3.10.5.1 3.17 Dynamie qualification report Confirmatory 3.10.5.2 3.18 Erwironmental Qualification Document Confirmatory 3.11.6.1 3.19 Erwiromental Qualification Records Confirmatory 3.11.6.2 4.1 CRD Inspection Program Procedural 4.53.1 5.1 Water Chemistry Design 5.2.6.1 Amendmer.t 25 1.9-3

r ABWR uume,

Standard Plant

- am e ;

Table 1.91

SUMMARY

OF AllWR STANDARD PLANT INTERFACES

. I

' COL LICENSE INFORMATION (Continued):

ITEM -

INTERFACE NO.

SUHJECT

> Ti'PE SUllSECTION 5.2 Conversion ofIndicators

' Procedural - 5.2,6.2 -

53-Practure Toughness Data Confirmatory 53.4.13 5.4 Materials and Surveillance Capsule Confirmatory 53.4.2 6.1 -

Protection Coatings and Organic Materials Confirmatory 6.13.1' I

6.2 External Temperature Confirmatory 6.4.7.1 63 Meterology (X/Os)

Confirmatory 6.4.7.2 :

6.4 Toxic Gases Confirmatory 6.4.73 -

7.1 Effects of Sation Blackout on 1IVAC Confirmatory 7.8.1 7.2 Deleted 73 Localized liigh licat Spots in Semiconductor Confirmatory 7.8.2 Material for Computing Devices 8.1 Stability of offsite power system Confirmatory 8.1.4.1 8.2 Diesel Generator Reliability Procedural 8.1.4.2 83 Class IE Feeder Circuits Design 8.23.1 8.4 Non class IE Feeders Design 8.23.2-8.5 Specific ABWR Standard Plant / remainder of plant Design 8.233 power sysytem interfaces 8.6 Interupting Capability of Electrical Confirmatory 83.4.1 Distribution Equipment 8.7.

Diesel Generator Design Details Confirmatory 83.4.2 --

8.8 Certified Proof Tests on Cable Samples Confirmatory 83.43 8.9 Electrical Penetration Assemblies Confirmatory 83.4.4 8.10 - Analysis Testing for Spatial Seperation Confirmatory 83.4.5 per IEEE 304 Q)

Amendment 25

' 19-4. -

i,7-

ABMTt a3463ooac Standard Plant nu c Table 1.91

SUMMARY

OF ABWR STANDARD PLANT INTERFACES COL LICENSE INFORMATION (Continued)

ITEM INTERFACE NO, SUBJECT

'IYPE SUBSECTION 8.11 DC Voltage Analysis Confirmatory 83.4.6 8.12 Seismic Qualification of Eyewash Equipment Confirmatory 83.4.7 8.13 Diesel Generator Load Table Changes Confirmatory 83.4.8 8.14 Offsite Power Supply Arrangements Procedural 83.4.9 8.15 Diesel Generator Qualification Tests Confirmatory 83.4.10 8.16 Defective Refurbished Circuit Breakers Confirmatory 83.4.11 8.17 Minimum Starting Voltages for Class Confirmatory 8 3.4.12 1E Motors 9.1 New Fuel Storage Racks Criticality Analysis Confirmatory 9.1.6.1 9.2 New Fuct Storage Racks Dynamic and Impact Confirmatory 9.1.6.2 Analysis 93 Spent Fuel Storage Racks Criticality Analysis Confirmatory 9.1.63 9.4 Spent Fuel Storage Rack Load Drop Analysis Confirmatory 9.1.6.4 9.5 Ultimate heat sink capability Design 9.2.17.1 9.6 Makeup water system capability Design 9.2.17.2 9.7 Potab!c and Sanitary Water System Design 9.2.173 9.8 Radioactive Drain Transfer System Collection Design 9 3.12.1 Piping 9.9 Contamination of DG Combustion Air intake Confirmatory 9.5.13.1 9,10 Use of Communication System in Emergencies Procedural 9.5.13.2 9.11 Maintenance and Testing Procedures for Proc.: dural 9.5.13 3 Communication Equipment 9.12 Fire Hazard Analysis Database Confirmatory 9A.63 10.1 Low Pressure Turbine Disk Confirmatory 10.2.5.1 Fracture Toughness Amcodment 2$

1.9-5

7 -

ABWR 2mie Standard Plant-

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Table 1.91 A

Ld

SUhth1ARY OF ABWR STANDARD PIANT INTERFACES COL LICENSE INFORh1ATION (Continued) i ITEM-INTERFACE NO.

SUlUECT TYPE SUBSECTION -

10.2 Turbine Design Overspeed Design 10.2.5.2~

12.1 Regulatory Guide 8.10 Confirmatory - 12.1.4.1 12.2 Regulatory Guide 1.8 Confirmatory 12,1.4.2 123 Occupational radiation exposure Procedural 12.1.4 3 13.1 Physical SecurityInterfaces Design &

13.6 3 '

Procedural 18.1 Main Control Room Design &

18.5 Confirmatory 19.1 Long term training upgrade Procedural

- 19A3.1 19.2 Long. term program of upgrading of procedures Procedural -

- 19A3.2 193 Purge system reliability Procedural 19A33 -

19.4 Licensing emergency support facility Procedural 19A3,4 19.5 In. plant radiation monitoring Procedural 19A3.5 19.6 Feedback of operating, design and construction Proced'ral 19A3.6 experience 19.7

_ Organi2.ation and staffing to oversee design and Procedural 19A3.7 construction 19.8 Quality Assurance Program Design 19B3.1 19.9 Prevention of Core Damage Procedural 19B3.2 -

19.10 Protection from ExternalThreats Design 19B33 '

.19.11-Ultimate Heat Sink Models

. Design 19B33 d

19.12 -

Ultimate Heat Sink Reliability Design 19B3.5 19.12a Main Transformer Design Design 19B3.6 19.13 Plant Siting Procedural 19B3.7 19.14 Interdesciplinary Design Reviews Procedural 19B3.8 Amendment 25 194-

p AmW 2346iooxc SJJlndimi Plant uvc 1A.2 NRC POSITIONS / RESPONSES GE has participated, and continues to par.

ticipate,in the BWR Owners' Group program to de-IA.2.1 Short Term Accident Analysis velop emergency procedure guidelines for General Procedure Revision (l.C,1(3)]

Electric BWRs. The resulting emergency procedure guidelines are generally applicable to the ABWR as NRC Position are the transient and accident analyses. Folloling is a brief description of the submittals to ds*c, and a in letters of September 13 and 27, October 10 justification of their adequacy to support guideline and 30, and November 9,1979 (References 4 through development.

8), the Office of N. : lear Reactor Regulation re-quired licensees of operating plants, applicants for (1) Description of Submittals operating licenses and licensees of plants under con-struction to perform analyses of transients and acci-(a)

NEDO 24708, AdditionalInformation Re.

dents, prepare emergency procedure guidelines, up-quiredfor NRC Staff Generic Report on Boll-grade emergency procedures, including procedures ing Water Reactors, August,1979.

for operating with natural circulation conditions, and to conduct operator retraining (see also item (b)

NEDO 24708A, Revision 1, Additionallnfor.

I.A.2.1). Emergency procedures are required to be marion Requiredfor NRC Staff Generic Report consistent with the actions necessary to cope with the on Boiling Water Reactors, December,1980.-

transients and accidents analyzed. Analysis of tran-This report was issued via the letter from D.

sients and accidents were to be completed in early B. Waters (BWR Owners' Group) to D. G.

1980 and implementation of procedures and retrain-Eisenhut (NRC) dated March 20,1981.

ing were to be completed 3 months after emergency procedure guidelines were established; however, (c)

BWR Emergency Procedure Guidelines (Re-some difficulty in completing these requirements has vision 0) - submitted in prepublication form been experienced. Clarification of the scope of the June 30,1980, task and appropriate schedule revisions are being developed. In (be course of review of these matters (d)

BWR Emergency Procedure Guidelines (Re-on Babcock and Wilcox (B&W)- designed plants, vision 1) Issued via the letter from D. B, the staff will follow up on the bulletin and order mat.

Waters (BWR Owners' Group) to D. G.

ters relating to analysis methods and results, as listed Eisenhut (NRC) dated January 31,1981, in NUREG-0660, Appendix C (see Table C.1, items 3,4,16.18,24,25,26,27; Table C.2, items 4,12,17, (e)

BWR Emergency Procedure Guidelines (Re-18, 19, 20; and Table C.3, Items 6,35,37,38,39,41, vision 2) submitted in prepublication form 47,55,57).

June 1,1982, Letter BWROG-8219 from T. J.

Dente (BWR Owners' Group) to D. G.

Response

Eisenhut (NRC).

In tbc clarification of the NUREG-0737 require-(1)

BWR Emergency Procedure Guidelines (Re-ment for reanalysis of transients and accidents and vision 3), submitted in prepublication form inadequate core cooling and preparation of gu;de-December 22,1982, Letter BWROG-8262 lines for development of emergency procedures, from T. J. Dente (BWR Owners' Group) to NUREG 0737 states:

D. G. Eisenhut (NRC).

Owners' group or vendor submittals may be refer-(g)

NEDO 31331, BWR Emergency Procec*u.

enced as appropriate to support this reanalysis. If Guidelines (Revision 4), submitted April 23, owners' group or vendor submittals have already 1987, Letter BWROG 8717, irom T. A.

been forwarded to the stafffor review, a briefde-Pickens (BWR Owners' Group) to T. Murley scription of the submittals and justification of (NRC).

their adequacy to support guideline development is all that is required.

(2) Adequacy of Submittals Amendment 6 1A.21

ABWR mmc Standard Plant unv c -

The submittals described in (1) above have been 0700. A DCRDR specified in NUREG.0737 is not _

discussed and reviewed extensively among the BWR required by SRP Section 18.1.

Owners' Group, the General Electric Company, and the NRC Staff.

1A.2.3 Control Room Design - Plant Safety Parameter Display Console The NRC has extensively reviewed the latest re-

[1.D.2]

} vision (Revision 4) of the Emergenc) Procedures Guidelines and issued a SER, Safety Eratuation of -

NRC Position BWR Owners' Group Emergency Procedure Guide-lines. Revision 4, NEDO-31331, March 1987, letter in accordance with Task Action Plan I.D.2, each from A. C. Thadani, NRC Office of Nuclear Reactor applicant and licensee shcIl install a safety parameter Regulation, to D. Giace, Chairman of BWR Owners' display system (SPDS) that will display to operating Group, dated September 12,1988. The SER con-personnel a minimum set of parameters which define cludes that this document is acceptable for imple-the safety status of the plant. This can be attained mentation. It further states that the SER closes all through continuous indication of direct and derived the open items carried from the previous revisions of variables as necessary to assess plant safety status, the EPG.

Response

GE believes that in view of these findings, no further detailed justification of the analyses or guide-The functions of the SPDS will be integrated into lines is necessary at this time. Interface require-the overall control room design, as permitted by SRP ments pertaining to emergency procedures are dis-Section 18.2.

cussed in Subsection 1A.3.1.

IA.2.4 Scope of Test Pmgram - Preoper.

I A.2.2 Control Room Design Reviews -

ational and Lower Power Testing [I.G.1]

Guidelines and Requirements [I.D.1(1)]

NRC Position NRC Position Supplement operator training by completing the In accordance with task Action Plan 1.D.1.(1),

special low-power test program. Tests may be ob-all licensees and applicants for operating licenses will served by other shifts or repeated on other shifts to be required to conduct a detailed control-room de-provide training to the operators, sign review to identify and correct design deficien-cies. This detailed control-room design review is

Response

expected to take about a year. Therefore, the Office of Nuclear Reactor Regulation (NRR) requires that The initial test program presents an excellent op-those applicants for operating licenses who are portunity for licensed operators and other plant staff unable to complete this review prior to issuance of a members to gain valuable experience and training license make preliminary assessments of their con-and in fact theses benefits are objectives of the pro-trol rooms to identify significant human factors and gram (see Subsection 14.2.1). The degree to which instrumentation problems and establish a schedule the potential benefit is realized will depend on such approved by NRC for correcting deficiencies. These plant specific factors as the organizational makeup of i

applicants will be required to complete the more the startup group and overall plant staff (see Subsec.

detailed control room reviews on the same schedule tions 14.2.2 and 13.1), as well as how the test pro-as licensees with operating plants, gram is conducted (see Subsection 14.2.4).

Response

The test program described in Chaper 14 is con-sistent with the BWR Owners' Group response to The design of the main control room will utilize item I.G 1 of NUREG-0737 as documented in a i

accepted human factors cr.gineering principles, in-letter of February 4,1981 from D. B. Waters to D.

corporating the results of a full systems analysis G. Eisenhut.

similar to that described in Appendix B of NUREG-O Amendment 25 1A.2-2 l

ABWR 2mesc Standard Plant uvc automatic reopening of containment isolation (5) The ABWR Standard Plant design is consistant

[V valves. Reopening of containment isolation w.:h this position.

)

valves shall require deliberate operator action.

(6) All ABWR containment purge valves meet the (5) The containment setpoint pressure that initiates criteria provided in BTP CSB 6-4. The main 22' containment isolation for non essential penetra-purge valves are fail-closed and are maintained.

tions must be reduced to the minimum compat-closed through power operation as defined in the ible with normal operating conditions, plant technical specifications. All purge and vent valves are remote pneumatically operated, fail (6) Containment purge valves that do not satisfy the closed and receive containment isolation signals, operability criteria set forth in Branch Technical Certain vent valves can be opened manually in Position CSD 6-1 or the StaJ Interim Position of the presents of an isolation signal, to permit October 23,1979 must be scaled closed as de-venting through the SGTS, fined in SRP 6.2.4, item II.6.f during operational cor.Jitions 1, 2,3, and 4. Furthermore, these (7)In the ABWR design, the containment purge and valves must be verified to be closed at least vent isolation valves will be automatically isolated every 31 days.

on high radiation levels detected in the reactor building HVAC air exhaust or in the fuel (1) Containment purge and vent isolation valves handling area air exhaust.

must close on a high radiation signal.

Response

(1) The bolation provisiont described in the Stan-dard leicw Plan, Subsection 6.2.4 (i.e., that there be diversity in the parameters sensed for the initiation of containment isolation) were re-g viewed in conjunction with the ABWR Standard t

Plant design. It was determined that the ABWR Standard Plan is designed in accordance with these recommendations of the SRP.

(2) This request appears to be directed primarily toward operating plants. Ilowever, the classifi-cation of structures, systems and components for the ABWR Standard Plant design is addressed in Section 3.2 of this SSAR. The basis for classi-fication is also presented in Section 3.2. The ESF system, with remote manual valves with leakage detectica outside containment are delinated in Tables 6.2-7. The ABWR Standard Plant fully conforms with the NRC position so far as it relates to the new equipment supplier.

(3) All non-essential systems comply with the NRC position to automatically isolate by the contain-ment isolation signals, and by redundant safety grade isolation valves.

(4) Control systems for automatic containment iso-lation valves are designed in accordance with this position for the ABWR Standard Plant

(]

Design.

U Amendment 25 tA.2-9

ABWR 23A61@AC b 1Illid Q Id Elillit RFV C 1 A.2.15 Additional Accident Monitoring the top of the containment sump. A wide range in-lustrumentation [II.F.1(1)]

strument shall also be provided for BWRs and shall cover the range from the bottom of the containment NRC Position to the elevation equivalent to a 600,000 gallon capacity. For BWRs, a wide range instrument shall Noble gas effluent monitors shall be installed be provided and cover the range from the bottom to w?th an extended range designed to function during 5 feet above the normal water level of the suppres-accident conditions as well as during normal operat-sion pool.

ing conditions, hiultiple monitors are considered necessary to cover the ranges of interest.

A continuous indication of hydrogen concentra-tion in the containment atmosphere shall be pro-(1) Noble gas efflynt monitors with an upper range vided in the control room. hicasurement capability _

capacity of 10 Ci/cc (Xc-133) are considered to shall be provided over the range of 0 to 10% hydro-be practical and should be installed in all gen concentration under both positive and negative I

operating plants.

ambient pressure.

(2) Noble gas effluent monitoring shall be provided

Response

for the total range of concentration extending from normal condition (as low as reasonably GE believes the requirements of Regulatory achievable (ALARA)) concentrations to a maxi-Guide 1.97, Revision 3, incorporate the above re-mum of 10~Ci/cc (Xe-133) h1ultiple monitors quirements. Section 7.5 compares the ABWR design are considered to be necessary to cover the against this Regulatory Guide, ranges of interest. The range capacity of individual monitors should overlap by a factor of IA.2.16 Identification of and Recovery ten.

From Conditions Leading to inadequate -

Core Cooling [II.F.2}

Because iodine gaseous effluent monitors for the accident condition are not considered to be prac-NRC Position tical at this time, capability for efflu nt monitoring of radiciodines for the accident condition shall be Licensees shall provide a description of any addi-provided with sampling conducted by absorption on tional instrumentation controls (primary or backup) charcoal or other media, followed by onsite labora-proposed for the plant to supplement existing instru-tory analysis.

mentation (including primary coolant saturation monitors) in order to provide an unambiguous, easy-In-contrinment tadiation level monitors with a to-interpret indication of inadequate core cooling h

maximum range of 10 rad /hr shall be installed. A (ICC). A description of the functional design re-minimum of two such monitors that are physically quirements for the system shall also be included. A separated shall be provided. hionitors shall be description of the procedures to be used with the developed and qualified to function in an accident proposed equipment, the analysis used in developing environment, these procedures, and a schedule for installing the equipment shall be provided.

A continuous indication of containment pressure shall be provided in the control room of each operat.

Response

ing reactor, hicasurement and indication capability shall include three times the design pressure of the The direct water level instrumentation provided containment for concrete, four times the design pres-in the ABWR design is capable of detecting sure for steel, and -5 psig for all containments, conditions indicative of inadequate core cooling A continuous indication of containment water The ABWR has two sets of four wide range level shall be provided in the control room for all reactor water level sensing units (eight total) which plants. A narrow range instrument shall be provided are used in two separate two out of four logics which for BWRs and cover the range from the bottom to initiate ECCS and other safety functions. Each set of Amendment 6 1 A.2-10

ABWR macue Sli1Hditrd Phtnt utiv c differential pressure signals which isolate the RCIC (5) Earlier initiation of ECC systems,

[,]

turbine are processed through the leak detection and

'V isolation system (LDS). Spurious trips are avoided (6) lleat removal through emergency condensers, because the RCIC has a bypass start system con-trolled by valves F037 and F(MS (see Figure 5.4-8, (7) Offset valve setpoints to open fewer valves per RCIC P&lD).

challenge, On receipt of RCIC start signals, bypass valve (8) Installation of additional relief valves with a F045 opens to pressurize the line downstream and block or isolation valve feature to eliminate accelerate the turbine. The bypass line via F045 is opening of the safety / relief valves (SRV's),

small (1 inch) and naturally limits the initial flow consistent with the ASME Code,.

surge such that a differential pressure spike in the upstream pipe will not occur.

(9) Increasing the high steam line flow setpoint for -

main steam line isolation valve (MSIV) closure, After a predetermined delay (approximately i

5-10 seconds), steam supply valve F037 opens to ad-(10) Lowering the pressure setpoint for MSIV Clo-mit full steam flow to the turbine. At this stage, the sur.

line downstream is already pressurized Thus, it is highly unlikely that a differential pressure spike (11) Reducing the testing frequency of the MSIV's, could occur during any phase of the normal start-up process.

(12) Mme stringent valve leakage criteri, aad 1A.2.24 Reduction of Challenges and (13) Early removal of leaking valves.

Failures of Relief Valves Feasibility Study and System Modification An investigation of the feasibility and constraints

[II.K.3(16))

of reducing challenges te the relief valves by use of

.n the aforemen'ioned methods should be conducted.

(

NRC Position Other methodi should also be included in the feasi-bility study. Those changes which are shown to The record of relief valve failures to close for all reduce relief valve challenges without compromising boiling water reactors (BWRs)in the past 3 years of the performance of the relief valves or other systems plant operation is appproximately 30 in 73 reactor-should be implemented. Challenges to the relief years (0.41 failures per reactor-year). This has dem-valves should be reduced substantially(by an order onstrated that the failure of a relief valve to close of magnitude),

would be the most likely cause of a small-break loss-of-coolant accident (LOCA). The high failure rate is

Response

the result of a high relief-valve challenge rate and a relatively high failure rate pei challenge (0.16 fail.

General Electric and the BWR Owners' Group _ l urcs per challenge). Typically, five valves are chal-reponded to this requirement in Reference 6. This lenged in each event. Tnis results in an equivalent response, which was based on a review of existing failure rate per challenge of 0.03. The challenge and operating information on the challange rate of relief failure rates can be reduced in the following ways:

valves, concluded that the BWR/6 product line had already achieved the " order of magnatude' level of (1) Additional anticipatory scram on loss of feedwa-reduction in SRV challange rate. The ABWR relief

ter, valve system also has similar design features which also reduce the SRV chaPenge rate. With regard to (2) Revised relief-valve actuation setpoints, inadvertently opened relief valves (IORV), the BWR/6 pint design evaluated for the Owners' (3) increased emergency core cooling (ECC) flow, Group report reflected a reduced level if IORC wapared with previous design because of (4) Lower operating pressures, OO Amendment 15 1A.214

ABWR 2 mime Standard Plant nev c climination of the pilot operated relief valve type of design, The ABWR design has also eliminated the pilot operated relief valve type of design.

For the ABWR which has solid state logie design with redundancy, the likelihood of an IORV is the same or less than the BWR/6 design evaluated in conection with the Owners' Group report. The

. redundant solid state design has been selected in order that the frequency of IORV with solid state logic becomes low enough so as to achieve the order of magnitude reduction in total SRV challange rate required by NUREG 0737.

The redundant solid state design for SRV operation in the pressure relief mode consists of two -

duplicated microproccessor channels. Each microproccessor channel activates a separate load driver and both load drivers must be activated to cause operation of the SRV's in the_ relief mode.

Operation of the SRV's in the ADS mode also requires activation of two microproccessors channels with separate load drivers to prevent unwanted SRV operation; however, two separate dual channel systems are used to assure reliable operation in the ADS mode Reliable operation in the pressure relief mode is assured by direct opening of the SRV against spring force, O

Amendment 6 1A.2-14a

e 4

ABWR- -

23A61c0AC :

Standard Plant-

' any c1 11A.2.25 _ Report on Outages of Emergency?

See Sabsection 1 Ai3.5 for7nterface (l

Core Cooling Systems Licensee Report Requirements;

(./

and Proposed Technical Specillcat ion i Changes [II.K.3(17))-

1A.2.26; Modification of Automatic De-L Pressurization System Logic Feasl<

NRC Position.

bility for increased Diversity for Somei Event Sequences [II.K.3(18)]:

Several components of the emergency core cooling (ECC) systems are permitted by technical NRC PositionJ

^

r specifications to have substantial outage times (e.g.,

72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for one diesel generator; 14 days for the The automatic ddpressurization system (AD'S) -

HPCI system). In addition, there are no cumulative actuation logic should be modified to eliminate the' outage time limitations for ECC systems.- thensees need for manual actuation to assure adequate core -

should submit a report detailing outage dates and cooling; A feasibility and risk assessment study is ~

lengths of outages for all ECC systems for the last 5 required to determine the optimum approach. One; years of operation. The report should also include possible scheme that should be considered is ADSi the causes of the outages (i.e., controller failure, actuation on low reactor vessel water level provided spurious isolation).

no high-pressure coolant injection (HPCI) or high.

pressure coolant system (IIPCS) flow enists and al Clarillcation low-pressure emergency core cooling (ECC) system.

is running. This logic would complement, not -

The present technical specifications contain replace, the existing ADS actuation logic, limits on allowable outage times for ECC systems and components. However, there are no cumulative

Response

outage time limitations on these same systems; it is possible that ECC equipment could meet present An 8 minute high drpell pressure bypass timer -

technical specification requireme.nts but have a high has been added to the ADS initiation logic to address ;

unavailability because of frequent outages within the -

TMI action item II.K.3.18; This timer will initiate on N

allowable technical specifications.

a Low Water Level I signal. When it times out,it bypasses the need for a high drywell signal to initiate -

The licensees should submit a report detailing the standard ADS initiation logic.-

outage dates and length of outages for all ECC systems for the last 5 years of operation, including '

For all LOCAs inside the containment, a high -

causes of the outages. This report will provide the drywell signal will be present and ADS will actuate staff with a quantification of historical unreliability 29 seconds after a Low Water Level.1 signalis due to test and maintenance outages, which will be reached. Tall LOCAs outside the containment.

requirements in the technical specifications.

. become rapid.y isolated and any one of th'e three high pressure.iCCS can control the water level. The?.

Based on the above guidance and clarification, a -

high drywell pressure bypass timer in the ADS detailed report should be submitted. The report initiation logic will only affect the LOCA response if should contain (1) outage dates and duration of all high pr;ssure ECCS fail following a break outside outages;(2) causes of the outage;(3) ECC systems the con'ainment. For this case the ADS will..

or components involved in the outage; and (4) autom.tically initiate within 509 seconds (8 minute corrective action taken. Tests and maintenance timet plu: 29 second standard ADS logie delay),

outages should be included in the above listings.

following a Low Water Levell signal, which are to cover the last 5 years of operation. - The licensee should propose changes to improve the availability of ECC equipment,if needed.

Applicants for an operating license sha.1 establish a plan to meet these requirements.

Response

Amendment 25

- tA.2.t3

ABWR 23yi c

SinfidRulflRDI uvc O

4 i

i i

j l

I i

l l

O.

i l

l l

O.

Amendment 21 1A,215a l

l I-

~.. ~ - - -.. - -

-. ~

AllWR m im4c Sillnd11rd Plllill

%r i

1AA.2 SUMMAlW OF SillELDING (1) The period of interest begins with the planiin a C

DESIGN REV1EW safe shutdown condition. Thus, the varioui safety A

related systems needed to achieve safe shutdown Sncrat alternatives are potentially available to conditions have pcrformed, and only the the designer to assure continued equipment engineered safety features systems (Chapter 6) availability and performance under post accident and auxiliaries, as described later, are required to conditions. One is to provide redundant systems maintoin this condition.

and/or components which are quatined to operate in the expected environment. Another is to provide (2) 11ased upon the accident source terms ofl Regulatory Guides 1.3 and 1.7 and Standard operator access to conduct the operations and to maintain the equipment. This latter alternative Review Plan 15.6.5, and normal operations the would generally be accompanied by appropriate vital equipment exposures will be within g shielding and in many cases would be dif0 cult if not maximum required envelope exposure of 3.6 x 10 impossible to carry out.

Rgds for equipment in primary containment,p x 10 Rads for equipment in ECCS rooms,9x10 in General Electric has taken the nrst approach SOTS rooms and pumps and valves per Tables and furthermore has designed the plant t.o that most 31.3 6,31.3 7. 31.315, and 31.316 where the responses to transient conditions are automatic, integrated exposure is for six months. All vital including achieving and maintaining safe shutdown equipment will be environmentally qualined. This conditions. The design basis ist the ABWR exposure envelope is no' time dependent after 100 Standard Plant is to require safety related equipment days to be appropriately environmentally qualified and operable Itom the control room. As a result of this (3) It is not neceuary for operating personnel to have design philosophy and as shown by this review, no access to any place other than the control room, changes are necessary to assure that personnel the technical support center, the post. accident acccu is adequate or that safety equipment is not sampling station, the sample analysis area, and the degraded because of post accident operation.

Safety related nitrogen supply bottles to operate g

the equipment of interest during the 100 day As part of the design of the ABWR Standard period. The control room, technical center and Plant it was necessary to establish the environmental sample analysis area are designed to be accessible conditions for qualification of safety related post accident. The latter areas are considered equipment. A result of this design work was an accesible on a controlled exposure basis. Direct environmental requirement establishing ihe shine from the containmrnt is less than 0.5 R/hr integratul dose that the equipment must be able to within four hours post accident.

withstand. These values are listed in Appendix 31.

(4) Access to radwaste is not required, but the Another aspect of the review was the manner in radwaste building is accessible since primary which the safety related equipment is arranged and containment sump discharges are isolated and operated during normal and abnormal operation and secondary containment sump pump power is shed postulated accidents. The essence of the ABWR at the onset of the accident. Thus, fission Standard Plant is to achieve and maintain a safe products are not transported to radwaste. The shutdown condition for all postulated accident combustible gas control system is operated from conditions with all operator actions being conducted the control roomt the ABWR does not have a from outside the primary and secondary containment containment isolation reset control arca or a rones, principally frorr the control room.

manual ECCS alignment area. These functions are provided in the control room.

The purposes of this reviewis first to verify that where equipment access is required, it is reasonably (5) Following an accident, access is available to accessible outside the primary and secondary electrical equipment rooms containing motor containment rones. Secondly, the review should control centers and corridors in the upper reactor verify that inaccessible equipment is erwironmentally building Section 12.3.6. This is based on qualiGed and is operable from the control room, radiation shine from the ECCS rooms and primary c ntainmenu th is no ahne The results of the review are:

Amendment 15 tAA2-1

AllWR mnme 811111dilfdlllli 1tLS radiation source in the electrical equipment rooms and ECCS corridor areas. While not necessary to maintain safe shutdown, such access can be uref.Iin extending system functionality and in plant recovery.

(6) The emergency power supplies (diesel generators) are accessible, llowever, access is not necessary since the equipment is environmentally qualified.

9 O

Amendment 10 1AA.2 2

ABWR m-c Standani Phtnt mc l AA.5 RESULTS OF Tile REVIEW Tables 1AA 2 thru s are generated:

l qb 1 AA.5.1 Splems Requirect l'ost Accicient (i) to show what major equipment and systems are required to function and thereby define the This section establishes the various systems systems for review, and l rquipment which are required to function following an accident along with their locations. The expected (ii) to sho y the redundnat equipment locations by habitability conditions and access and control needs divisional isolated rcom or atea and are identified for the required [wt accident period.

containment or building.

I AA.5.1.1 Necessary Post. Accident functions and 1 AA.$.1.2 I mergency Core Cooling Systems and Systems Ausillaries Following an accident and assuming that Table 1AA 2 lists various systems related to immediate plant recovery is not possible, the cooling the fuel under post accident conditions as following functions

• are necessary:

described in Section 6.3 and Subsection 9.4.5.2 IIVAC. This table shows ECCS equipment and (1) Reacthity control equipment coolers in an ECCS toom, justrumentation transmitters are in adjoining areas.

(2) Reactor core cooling The required power and cooling water in the same division are described in Section 1AA.5.1.5. All (3) Reactor coolant system integrity perform together to provide an ECCS function.

(4) Primary reactor containment Integrity, and The automatic depressurization system (ADS) function is described in Subsection 1.2.2.4.8.2. A (5) Radioactive effluent control postulated non break or small break accident could

(

require continued need for the depressurization

(

Reacthity controlis a r.hort term function and function until the RilR system is placed in the is achieved when the reactor is shutdown. The shutdown reactor cooling mode, in the case of a remaining functions are achieved in the longer term non-break or a small break accident, the majority of post accident period by use of:

the fission products would be released sia the safety relief valves to the suppression pool and hence to the (a) The cmerge cy core cooling system (ECCS) containment rather than direct mixing through the and their auxiliaries (for reactor core cooling).

supersession pool vents as v/ould occur following a DilA LOCA. In either case the diatribution of fisJon (b) The combustible gas control system (CGCS) products is assumed to be the same as for the and auxiliaries (for primary containment and DilA LOCA even though realistically a significant reactor coolant system integrity),

portion of halogens and solid fission products would be retained in the reactor pressure vessel. Thus, the (c) The Gssion product removal and control system results as they apply to the ADS are very conservative.

and auxiliaries (for radioactive effluent The pneumatic nitrogen supply for the ADS and control), and other containment valves is included in Table IAA 3 as a portion of the combustible gas control. The hand (d) Instrumentation and controls and power for operated nitrogen reserve supply valves P54 F017C accident monitoring and functioning of the and D are accessible outside the secondary necessary systems and associated habitability containment, if needed, to mitigate a large leak.

systems.

The high pressure core flooder (IIPCF) and the low pressure flooder (LPFL) functions are described

• ANSI /ANS U Cn ena for Accident Monitoring Functions in in Subsection 1.2.2.4.8.1.1 and 1.2.2.4.8.3 respectively, light Water Reacton The cooling function can also satisfy the containment q

cooling function in that by cooling suppression poo!

Q water, which is the source of water flowing to the reactor, the containment source of heat is also Amendment 11 1AA.5-1

AlnVR mmc 81.Il_Ildfifil Dillli ket C remmed. The wetwell/drywell sprays are described Enginected safety feature filter systems are the in Subsection 1.2.2 31.9.4.

Standby gas treatment sy. tem (SGTS) and the control building outdoor air cleanup system. Both consht of The fuel pool cooling function (Subsection redundant systems designed for accident conditions 1.2.2.8.2) is also included on the basis that a recently and are controlled from the contr01 room. The SGTS unloaded fuel batch could require continued cooling filters the gaseous effluent from the primary and during the post. accident period. The equipment is secondary containment when required to limit the environmentally qualified so access is not required discharge of radioactivity to the environment. The and redundancy is included in system components.

system function is described in Subsection 1.2.2.4.1.6.

The location of selected associated valves and A portion of the control building heating instrument transmittrra are included. These do not ventilating and air conditioning (llVAC) provides represent all of this type of equipment which is detection and limits the introduction of radioactive environmentally qualified, safety.related, or included material and smoke into the control room. This in the systems of Table 3.21. It does however, portion is described in Subsection 9.4.1.1.3.

represent principal components which are needed to opcrate, generally during post accident operations.

The CAhtS described in the previous section For example, most ECCS system valves are normally also measures and records containment area radiation open, and only a pump ditch *rge valve needs to open under post accident conditions. A post accident to direct water to the reactor. Similarly, the sampling system (PASS) obtains containment instrument transmitters shown are those which atmosphere and reactor water samples for chemical would provide information on long term system and radiochemical analysis in the laboratory. Delayed performance post a:cident. Contrcl room sampling, shielding, remote operated valves and instrumentation is not listed since it is all in an sample transporting casts are utilized to reduce accessible area where no irradiation degradation radiation exposure. The samples are manually would be expected. Passive elements such as transported between the PASS toom in the reactor thermocouples and flow sensors are not listed building and the analysis laboratory in the service although they are environmentally qualified. The building. The system is described in Subsection l components listed under main steam (U21) are those 9.3.2.3.1. Table 1A A 4 lists the fission product for ECCS function or monitoring reactor vessel level.

removal control components and locations.

Suppression poollevelis included with the llPCF instrumentation, l AA.S.I.5 Instrumentation und Control, Power, and 1 AA.5.1.3 Combustible Gas Control Splems and Ausillatles hiost of the post accident instrumentation and control system equipment is listed with the applicable Flammability control in the primary equipment in Tables 1AA.2,1AA 3 and 1AA 4. The l containment is achieved by an inert atmosphere remaining instrumentation and control equipment is during all plant operating modes except operator included with the power and habitability systems access for rcfueling anti maintenance and a equipment listed in Table 1AA 5. Instrumentation is recombiner system to control oxygen produced by consistent with the post accident phase variables radiolysis. The high pressure nitrogen (llplN) gas monitored by the post accident monitoring (PAht) supply is described in Subsection 1.2.2.8.8. The system listed in Table 7.5.2.

containment atmospheric monitoring s) stem (CAhtS) measures and records containment Standby AC power is supplied by three diesel oxygen / hydrogen concentrations under post accident generators in separate electrical divisions as described conditions. It is automatically initiated by detection in Subsection 1.2.2.4.1.8. The diesel generators, of loss of coolant accident (LOCA) and is described switchgear and motor control centers are included in in Subscetion 7.6.1.6. Table 1 AA 3 lists the the unit Class 1E AC power system described in combustible gas control principal comporents and Subsection 1.2.2.5.1.2. Storage batteries are the their locations, standby power rource for the unit Class 1E DC power system described in Subsection 1.2.2.5.1.7. The safety l AA.5.1.4 Fission Product Remmal and Control system logic and control power system is described in Sptems and Ausillaries Subsection 1.2.2.5.1.3.

Amendment 23 1AA.5 2

ABWR mime SJan11ar11 Plant ivu:

liabitability systems ensure that the operator

,m

('v) can semain in the control oom and tale appropriate action for post accident operations. The control building includes all the invrumentation and controls necessary for operating the systems required under post accident conditions.

The control room, controf and reactor building flVAC essential equipment are a portion of the plant environmental control of temperature, pressure, humidity and airborne contamination described in Subsection 1.2.2.8.10(1),(4), (5), (7) and l (8). IIVAC units controlling the local rooin environments are included with respective equiprnent in Tables 1AA4,1AA-3 and 1 AA-4. The major HVAC equip:nent at.d locations are listed in Table IAA 5.

The reacte b %n sooling water (RIICW) system provide. 460 unter to :lesignated equipment in the resetor building including containment as described in Subsection 1.2.2.8.1.

The IIVAC emergency cooling water (llECW) system provides chilled water to designated equipment in the control bu;1 ding as described in

, subsection 1.2.2.8.4.

9 m

(v Arnendment 25 1AA.5-3

[.

-AllWR:

zwun R

Satidatd. Plant niv ti CilAPTER 2 i

TAllLE OF CONTENTS l

Secdun This Ease 2

SITE CIIARACTERISTICS 2.0 SUM 51ARY 2.0-1

?

2.1 LIMl'fS IMPOSED ON SRP SECTION 11 ACCEPTANCE CRITERIA fly AHWR DESIGN 2.11-2.2 RI:OUIREMl?NTS FOR DE11:R$11 NATION DI 2.21 i

Aligg3nI.ACCElrrAlllLl1Y

-i 2.2,1 Design 11ases Events 2.21-2.2.2 Severe Accident 2.21 l

23 COL LICENSE INI'ORM ATION -

23 1 23.1 Erwelope of AllWR Standard Plant 23-1 Site Design Parameters 23.2 Standard Review Plan Site Characteristics 23 1 e

233 CRAC 2 Computer Code Calculations 23 1 APPENDlX 2A INPUT TO CR AC 2 COMPUTER COI)E l OR

~

-I DI~rERh11 NATION OF AllWR SITE ACCEL'TAltiLITY 3

a M

Amendment 2$

--,.6.

ABWR mmen Etandard Plant n~ n -

2.2 REOUIREMENTS FOR UTluTY are to be supplied by the licensing utility O

DETEl0611 NATION OF AllWR SITE as specified in the CRac 2 manual ACCEPTAlllLITY (NUREG/CR 2326) and are site specific.

This section provides the requirements for the The basic reference case assumes no evacuation determination of ABWR site acceptability, or radiation shiciding (Subgroup Evacuation) for risk Acceptability is required from the standpoint of both and dose calculations. However,if the results of design bases events and severe accident, such an evaluation for a speelfic site are unacceptable, site specific evacuation and shiciding 2.2.1 Design Hases Events parameters may be substituted in lieu of the -

reference values in Subgroup Evacuation.

For design bases events, the site is acceptable if all of the site characteristics fall whhin the envelope Analysis: The analysis for evaluation of a specific site:

of ABWR Standard Plant site design parameters will be accomplished with the CRAC 2 computer given in Table 2.0.L For cases where a characteristic code as modified through Sandia National Labo.

i exceeds its envelope it will be necessary for the ratory mod 46. Basic input and code characteristics applicant referencing the ABWR design to submit ate described in NUREG/CR.2326 and analyses to demonstrate that the overall set of site NUREG/CR 2552.

characteristics do not exceed the capability of the design.

2.2.2 Severe Accidents The ABWR PRA results were calculated for an average or typical site, as outlined in Subsection 19E.3. Although these results form a good basis for assessing the general ABWR capability to satisfy offsite dose related goals, they do not form a basis for concluding that the ABWR would meet dose reWcd goals at a specific site whose characteristics ensaot be defined at the point of ABWR certification. Consistent with the certification concept that all key technicalissues be resolved before certification,it is appropriate to define the piocess for determining future site acceptability.

This process is defined below in terms of (1) acceptance criteria, (2) data input, and (3) analysis, Acceptance Criteria: Site acceptability for severe accidents will be based upon a calculation using the CRAC 2 computer code. The results of such a calculation will be compared to the goal. of Table 19E.3 7 as shown in Table 2.21. The site will be deemed acceptable if the results fall within the given

goals, Data Inout: The input to the CRAC 2 computer code will be a combination of ABWR and site parameters.

The CRAC 2 code input is divided into specific areas. The areas defined in Table 2.2 2 as ABWR will be used as input with their specific data listed in Appendix 2A The areas defined as GENERAL are also provided in Appendix 2A. The areas defined as Amendment 25 2.21

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

ABWR 2m m StandanLPlant neo 2.3 COL LICENSE INFORMATION (1)

Design adequacy is established if floor (V')

resp (mse spectra are bounded by Section 30.4 2J.! rnselope of Standard Plant Design spectra (or the actual spectra considered in Parameters design if applicable) at key locations. The site unique response spectra used for comparison 2.3.1.1 Non.Scismic Design Parameters need not be broadened since uncertainties in the structural frequencies have been Compliance with the envelope of AllWR Standard accounted for in the smooth broadened site Plant site non scismi: design parameters of Table envelope spectra.

2.01 shall be demont,trated for design bases events.

(See Subsection 2.2.1)

(2)

If not, examine whether the deviations are at major resonant frequencies of the component 2.3.1.2 Seismic Design Parameters under consideration, if not, design adequacy is confirmed. Otherwise, perform analysis To confirm the seismic design adequacy of the and/or testing to demonstrate that the standard plant, the COI applicants shall demonstrate acceptasice criteria given in design that the eight (8) site dependent conditions specined specifications are met.

in Section 3A.1 are satisfied. In meeting these eight conditions, the compliance with the site envelope If the soil properties of the site vary very abruptly parameters shown in Table 2.01 for soil properties with depth (site dependent condition 7), a site and seismology is also catablished, specific SSE SSI analysis is required. The evaluation procedures and acceptance criteria specified above if there is any deviation of the eight are applicable, site dependent conditions, a site specine evaluation is required. The type of evaluation will vary depending If the soll bearing capacity at the site is not on the deviation. If the deviation is for condition 1 adequate to accommodate the standard plant design p

(peak ground acceleration),2 (ground response loads (site dependent condition 8), the foundation Q

spectra), or 6 (shear wave vehicity), a site specific SSE material shall be removed and replaced with better soil structure interaction analysis (SSI) is required.

material to achieve the required bearing capacity.

The calculated site unique responses are compared to Alternatively, the applicant referencing the ABWR the site-envelope responses defined in Section 30.4 to design may perform a site specific analysis to confirm the seismic design adequacy of the standard demonstrate that the site has an adequate bearing plant according to the following prccedures and capacity against the site unique loads, acceptance criteria.

The site dependent conditions 3 (liquefaction The Seismic Category I structures including the potential) and 4 (fault displacement potential)

RPV and its internal components that are included in require site speciGc investigation.

the SSI analysis model:

A site specific evaluation is required if the (1) Design adequacy is established if maximum embedment depths of Seismic Category I buildings structural responses in terms of force, moment, or deviate from those from the standard plant design acceleration are bounded by the Section 30.4 (site dependent condition 5). The evaluation responses (or the actual scismic loads considered procedure and acceptance criteria are the same as in design if applicable) at key locations.

those defined above for the site specific SSE SSI analysis.

(2)If not, calculate resulting SSE stresses. Design adequacy is confirmed if combined stresses due to 2.3.2 Standard Heslew Plant Characteristics l

SSE and other appropriate loads are within design code allowable limits.

Identification and description of all differences from SRP Section 11 Acceptance Criteria for site For Seismic Category I equipment and piping characteristics (as augmented by Table 2.11) shall whose seismic input is in the form of floor response be provided. Where such differences exist, the (3

spectra:

evaluation shall discuss how the alternate site

'V characteristic is acceptable. In addition, the COi, Amendment 25 23-t

ABWR m-n Shilli!Md.Elillit lin Il applicant will provide / address the following:

2J.2.7 Geology and Selsmology 23.2.1 Ilydrologic Features Description COL applicants will submit the geology and scismology investigations required by 10CFR50, COL applicants will provide a detailed Appendix A, and 10CFR100, Appendix A. Regu.

description of all major hydrologic features on or in 8atory Guide 1.70 and the SRP. All information the sicinity of the site. They will also provide a relevant to local, regional, and site geology and specific description of the site and all safety related scismologicalinvestigation will be obtained and elevations, structures, exterior accesses, equipment, presented to the NRC in accordance with Regulatory and systems from the standpoint of hydrology Guide 1.70. The NRC evaluates this information to considerations, establish that a safe shutdown carthquake of 03g or less is adequate for the site and the potential for 23.2.2 fee iloating or lilockage surface faulting.

COL applicants will demonstrate that 23.2.8 Vibratory Ground hiotion safety.related facilities and water supply are not affected by ice fheding or bh>ckage.

COL applicants will develop site specific geological, seismological, and geotechnical data and 23.23 II)draulle Design of Channels and will submit these data to the NRC for review. These Resenoirs data should be comparable to lhe design basis assumptions regarding the SSE, including the COL applicants will provide Ihe basis for the verification of the ground motion response spectra.

hydraulic design of channels and reservoirs used to transport and impound plant cooling and for 23.2.9 Surface Faulting protection of related structures.

COL applicants will develop site specific 23.2.4 Cooling Water Supply geological data to ensure that no potential exists for surface faulting at the site.

COL applicants will identify natural events that may reduce or limit the available cooling water 23.2.10 Stability and Subsurface Material and supply and ensure that an adequate water supply will Fmmdation exist to operate or shutdown the plant as required.

COL applicants will develop and submit to the 2J.2.5 Surface Water Dispersion of Fmergency NRC site-specific geotechnical data to demonstrate Operation and Water Supp!y that they are comparable to the design assumptions concerning the soil. deposit depths, the soil profile COL applicants will provide information on the and properties, and the ground water level, ab lity of the surface water environtnent to disperse, Particular attention should be paid to the dilute, or concentrate accidental releases of assumptions for the depth of embedment in the case radioactive effluents, specifically information about of rock and the three cases of soil-deposit depths for the effects of these releases on existing and known which fixed values of depths are assumed. The COL future use of surface water resources.

applicant will demonstrate that the envelope of structural response with fixed soil depth will cover 23.2.6 Technical Specifications and Lmerycucy completely the cases for which the soil deposit Operation and Shutdown Water Supply depths and properties are different from those assumed in the SSAR.

COL applicants willidentify the technical specifications and emergency procedures required to 2J.2.11 Site and Facilities -

implement flood protection for safety related facilities and provide assurance of an adequate water COL applicants will provide a detailed description supply to shutdown and cool the reactor.

of the site conditions and geologic features and demonstrate the site characteristics are enveloped by the 03g peak horizontal ground acceleration for the SSE. The description willinclude site topographical Amendment 23 2.3-2

> /Mb 23Ar,100At!

M(IRilitgl_[311til RtVJ operation, calibration operation, or test to observed on control room instruruntation. More

(]

verify operational availability impairs the importantly, the hydraulic control unit scram G/

ability of the systein to perform its intended accumulator level is continuously monitored.

safety function. Additionally, the system design assures that when a scram trip point is exceeded, The main steamline isolation valves may be there is a high scram probability. Ilowever, 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 thel 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, electrical transients, and can be deterruined.

physical events from impairing the ability of the system to respond correctly.

The RiiR system testing can be performed during normal operation. Main system pumps can The reactor 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 reactor variable exceed the corrective operating procedures also permit testing the action setpoint.

supply valves of the three RilR lines The lower pressure flooder mode can be tested after reac.

The reactor protection system initiates an tor shutdown. Each active component of the ECCS automatic reactor shutdown if the monitored plant provided to operate in a design basis accident variables exceed preestablished limits. This is designed to be operable for test purposes system in arranged as four separately powered during normal operation of the nuclear system.

I divisions.

The high functional reliability, redundancy, Each division has a logic which can produce an and intervice testability of the protection automatic trip signal. The logic scheme is a system satisfy the requirements specified in two out of four arrangement. The reaetor 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 ver!fies the ability to de energize the scram pilot valve Chapter /

solenoids. Indicating lights verify that the Section Illk actuator contacts have opened. This capability for a thorough testing program significantly (1) 1.2.1 Principal Design Criteria increases reliability.

(2) 4.6 Functional Design of Reactivity Control rod drive operability can be tested Control Systems during normal reactor operation. Rod position indicators and in core neutron detectors are used (3) 5.4.5 Main Steamline Isolt' ion System to verify control rod movement. Each control rod can be withdrawn one step and then reinserted to (4) 5.4.7 Residuallleat Removal System 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 t;me. Control rod mechanism overdrive (6) 63 Emergency Core Cooling Systems demonstrates rod to dri'c coupling integrity.

Ilydraulic supply subsysti m pressures can be (7) 7.2 Reactor Protection System Ov Amendment 1 11 13

ABWR 2-n S.tlltidat(U.4RRt RIV H (8) 7.3.1.1 Emergency Core Cooling Syster is -

the protection system allows operational system and instrumentation and Control testing by the use of an independent input for 73.2.1 each actuator logic. When an essential moni-tored variable exceeds its scram trip point, it (9 73.1.2 Leak Detection and Isolation is sensed by four independent sensors each 10 l

) and Systern Irntrumentation cated in a separate instrumentation channel. A 73.2.2 and Controls bypass of any single channel is permitted for maintenance operation, calibration operation, (10) 7.6.1.2 Process Radiation Monitoring-test, etc. This leaves three channels per moni-and Instrumentation and Controls tored variable, each of which is capable of ini-7.6.2.2 tlating a scram. Only two actuator logics must trip to initiate a scram. Thus, the two out of.

(11) 15 Accident Analyses four arrangement assures that a scram occurs as a monitored variable exceeds its scram setting 3.1.2.3.3 Criterion 22. Protection System Independence The protection system meets the design requirernents for functional and physical 3.1.2.33.1 Criterion 22 Statement independence as specified in Criterion 22.

The protection system thall be designed to For further discussion, see the following assure that the effects of natural phenomena and sections:

of normal operating, maintenance, testing, and postulated accident conditions on redundant Chapter /

channels do not result in loss of the protection SIIllpJ1 Illk function, or shall be demonstrated to be 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) 4.6 Functional Design of Reactivity operation, shall be used to the extent practical Control Systems to prevent lo s of the protection function.

(3) 5.4.5 Main Steamline Isolation System 3.1.23.3.2 Evaluation Against Criterion 22 (4) 5.4.7 Residuallleat Removal System Components of protection systems are designed so that the mechanical, thermal and radiological (5) 63 Emergency Core Cooling Systems 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 funetlon.

(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. Multi-(8) 73.1.2 Leak Detection and isolation plexed signals are carried by fiber optic medium and System to assure control signal isolation.

73.2.2 The reactor protection system is designed to (9) 7.6.1.2 Process Radiation Monitoring -

permit maintenance and diagnostic work while the and Instrumentation and Controls reactor is operating without restricting the 7.62.2 plant operation or hindering the output of safety functions. The flexibility in design afforded (10) 15 Accident Analyses O

Amendment 25 3.1 16

-~...w.

s ABWR-mmorm Stan_dard Plant -

nw A TAHLE 3.21 l

O t

CI ASSIFICATION SOMMARY (Continued)

[

i Table Table 3.2 1 MPL 3.21 MPL ltem No.

N_umlaI**

Illlt item No.

Number **

Ihig J

Nuclear ruel N12 N36 -

Extraction System J1 J11 Nucicar Fuel N13 N37 Turbine Bypass System J2 J12 Fuel Channel N14

. N38 Reactor Feedwater Pump Driver '

K Radioactive Waste Sntem

-- N15 N39~

Turbine Auxiliary Steam System K1 K17 Radwaste System N16 N41 Generator Ilydrogen Gas Cooling System N

Pour Cycle Sntems N17 -

N42

-NI N11 Turbine Main Steam System N18 -

N43 Generator Cooling System.

N2 N21 Condensate, recdwater and N19 N44

. Generator Scaling Oil System Condensate Air Extraction System N20 N51 Exciter N3 N22 Ileater, Drain and Vent System N21 N61 Main Condenser N4 N25 Condensate Purification System N22 N62

' Offgas System ;

N5 N26

. Condensate Filter Facility N23 N71' Circulating Water System -

N6 N27

. Condensate Demineralizer N24 N72

' Condenser Cleanup System -

N7 N31 Main Turbine P-Statlon Auxilatry Sntems.

N8 N32 Turbine Control Syste P1

.P11 Makeup Water System -

tl

-t (Purified).

N9 N33 Turbine Gland Steam System

'P2 P13 Makeup Water System

.N10 N34 Turbine Lubricating Oil (Condensate)

System N11 N35

' Moisture Separator lleater

~

These systems or subsystems thereof, have a primary function that is ' safety related, As shown ln -

the balance of this Table, some of these systems contain non safety related components and,.

n conversely, s.me systems whose primaryfunctions are non safety related contain components that.

R

- have been designated safety related.

Master Pans List Number designatedfor the system Amendment 25 3.2-7.1

~

.~.

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

ABWR mwen Sandard Plan!

nw n g

TA1112 3.2 1 CLASSIFICATION SUMMAltY (Continued)

Table Table 3.21 MPL 3.21 MPL ltem No.

Nu mber*

• Eth item N6-b'ninlarl'

'lllh P

Station Auxillary Syiltnu (Continued)

Pl7 P73 Ilydrogen Water Chemistry System P3 P21 Reactor 13uilding Cooling Water System

• P18 P74 Zine injection System P4 P22 Turbine !!uilding Cooling P19 P81 Dreathing Air System Water System P20 P91 Sampling System (includes PASS) l PS P24 IIVAC Normal Cooling Water System P21 P92 Frecre Protection System P6 P25 llVAC Emergency Cooling Water P22 P95 f ron injection System System R

Station llectrical Systems P7 P32 Oxygen Injection System R1 R10 Electrical Power Distribution P8 P40 Ultimate llent Sink System P9 P41 Reae:oi Service Wate System R2 R11 Unit AuxiliaryTransformer P10 P42 Turbine Senice Water System R3 R13 Isolated Phase Bus Pil P51 Station Instrument Air System R4 R21 Non-Segreated Phase ilus P12 P52 Instrument Air System R$

R22 Metalelad Switehgear l

P13 P54 liigh Pressure Nitrogen Gas R6 R23 Power Center Supply System R7 R24 Motor Control Center P14 P61 lleating Steam and Condensate Water Return System R8 R31 Raceway System P15 P62 Ilouse Boiler R9 R34 Grounding Wire P16 P63 llot Water lleating System R10 R35 Elcettical Wiring Penetration These systems or subsystems thereof, have a primary function that is safety-related. As shown in the balance of this Table, some of these systems contain non safety related components and, conversely, some systems whose primary functions are non safety related contain components that have been designated safety-related, Master Parts List Number designatedfor the system O

Amendment 21 32-7.2

r ABWR mamAn ShindanW}llRt PTV. A

^(3 TA11IE 3.21 V

CLASSIFICATION SUMMAl(Y (Continued)

Quallly Group Quality Safet4 tern.

Classi-Assurance Selsmic l'titidpal Component" Gait Llau#

lltallEn Etntdtutitill' QttracI[

b'k.

C2 Citi) Sptem (Continued) 6.

CRD Drive wates cumps N

SC D

E 7.

Control Rod Drive 1/3 C

A/--

Il I

II I

8.

Illectrical modules with 3

C,SC safety function 9.

Cable with safety-rclated 3

C,SC,X D

1

10. ATWS Equipment associated N

C E

(cc) with the Alternate Rod insert (ARI) functions C3 l'cedwater Control Splem N

C.T,5C, E

X V[

s Amendment 25 3.2 11.1

AllWR mamn S11u11hin!I'lant nuv n h

TA11LE 3.21 CIASSIFICATION SUMMAin' (Continued)

Quality Group Quality Safety laica.

Classi.

Assurance Scismic l' dad 191 Cn!14Tutul" Clalib

{lggd Mu h M' CdN DM I

gigge l C4 Standby Liquid ControlSystem 1.

Standby liquid control 2

SC 11 B

I (u) h tank including supgotts n

2.

Pump including supports 2

SC B

B I

(u) 3.

Pump motor 2

SC Il 1

(u) 4.

Valves - injection 1

SC A

B I

(u) 5.

Valves within injection 1

C.SC A

B I

(u) valves 6.

Valves beyond injection 2

SC Il B

I (g,u) valves 7.

Piping including supports 1

C,SC A

B i

(g.u)

Q within injection valves n

8.

Piping including supports 2

SC B

B 1

(g u) beyond injection valves 9.

Electrical equipment 3/N SC,X B/E I/---

(cc)

?

and desices 5

10. Cable 3/N SC,X B/E 1/

(cc) l CS Neutron Afonitoring System 1.

Electrical modules -

3 SC,X B

i SRNM, LPRM and LPRM 2.

Cable - SRNM and LPRM 3

C,SC,X B

1 RZ 3.

Detector and tube 2/3 C

B/C B

1 assembly i

l O

Amendment 20 3 2-12 1

1

ABWR zw.ioaru Standard Plant awn

,\\

(O TA11I.E 3.2 1

)

CIASSIFICATION SUMMAltY (Continued)

Quality Group Quality Safety Loca.

Classi.

Assurance Seismic b

ggd Requirement' CMrguni

$10 l'rincipal Componenta gggg gge El RilR Splem (Continued)

10. Jockey pumps and motors 2

SC Il 11 I

including supports n

O

\\

Ch V

Amendment 20 3.2-14.1

ABWR 2mmu Standantflant IUV A g

TAlllE 3.21 CIASSIFICATION SUMMAltY (Continued)

Quality Group Quality Saftig thea.

Classi.

Anurance Scismie l'dadbal Conmentfd" Gail Ll191 Ikallan lituuktintul' Catecon b'eln I

E2 I"

"'tst ure Core Flooder S

1.

Reactor pressure ver,sel 1/2 C.SC A/II 11 1

(g) injection line and connected piping including supports with-in outcrmost isolation valve

• 2.

All other piping inclading 2/3 SC,0 B/C B

1 (g) supports" 3.

Main Pump 2

SC D

11 1

4.

Main Pump Motor 3

SC B

I 5.

Valves other isolation 1

C.SC A

B I

(g) and within the reactor pressure vesselinjection line and connected lines l

6.

All other valves 2/3 SC B/C 11 1

(g) 7.

Electrical moduler, with safety-3 C SC,X B

1 related functions 8.

Cable with safety-related 3

C,SC,X B

1 function E3 leak Detection and Isolation Sptem

1. Temperature sensors 3/N C.SC,T B/E 1/- - (7) 2.

Pressure transmitters 3

C SC B

1/ -

(r) 3.

Differential pressure 3

C,5C B

1/ -

(z) trannuitters (flow)

The ECCS high pressure core flooder spargers are part of the Reactor Pressure l'essel System, see item Bl.S.

Pool suction piping, suction piping from condensate storage tank, test line to pool, pump discharge piping and return line lo pool.

Amendment 25 3.2 13

ABWR muu i

SlilTHlMilltilli ITV.11 3.3 WIND AND ToltNADO LOADINGS Reference 1. Reference 2 is used to obtain the O

rflective wind pressures for cases which Refer.

AllWR Standard Plant structures which are ence 1 does not cover. Since the Seismic Cat.

Scismic Category I arc designed for tornado and egory I structures are not slender or flexible, extreme wind phenomena.

vortex. shedding analysis is not required.ind the 3.3.1 Wind leadings 3.3.2 Tornado Imadings 33.1.1 Design Wind Velocity 33.2.1 Appikable Design Parameters Seismic Category I structures are designed to withstand a design wind velocity of 130 mph at an The design basis tornado is described by the elevation of 33 feet above grade with a recur.

following parameters:

rence interval of 100 years. See Subsection 3.3.3.1 for interfacc: requirement.

(1) A maximum tornado wind speed of 300 mph at a radius of 150 feet from the center of the 33.1.2 Determination of Applied l'arces tornadot The design wind velocity is converted to (2) A maximum translational velocity of 60:nph; velocity pressure in accordance with Reference 1 using the fortnula:

(3) A maximum tangential velocity of 240 mph, based on the translational velocity of 60 q,

= 0.00236 K, (lVT mph; where K

= the velocity pressure exposurc (4) A maximum atmospheric pressure drop of 2.00 coefficient which depends upon the psi with a rate of the pressure change of type of exposure and height (r) 1.2 psi per secondt and O

above ground per Table 6 of Reference 1.

($) The spectrum of tornado generated missiles and their pertinent characteristics as given 1

= the importance factor which depends in Subsection 3.5.1.4.

on the type of exposure; appropriate values of I are listed in Table See Subsection 3.3.3.2 for COL lleense 3.3-1, information.

V

= design wind velocity of 130 mph, and 3.3.2.2 Determination of Forces on Structures q'

- velocity pressure in psf The procedures of transforming the tornado loading into effective loads and the distribu-The velocity pressure (q ) distribution with tion across the structures are in accordance height for exposure types C and D of Reference 1 with Reference 4. The procedure for transform-are given in Table 3.3-2.

ing the tornado-generated missile impact into an effective or equivalent static load on struc-The design wind pressures and forces for tures is given in Subsection 3.5.3.1. The load-buildings, components and cladding, and other ing combinations of the individual tornado load-structures at various heights above the ground ing components and the load factors are in accor-are obtained, in accordance with Table 4 of dance with Reference 4.

Reference 1 by multiplying the velocity pressure by the appropriate pressure coefficients and gust The reactor building and control building are factors. Gust factors are in accordance with not vented structures. The exposed exterior Table 8 of Reference 1. Appropriate pressure roofs and walls of these structures are designed coefficients are in accordance with Figures 2, for the 2.00 psi pressure drop. Tornado dampers 3a,3b,4, and Tables 9 and 11 through 16 of Amendmeat 23 3.31

ABWR zwun Silttid1111PhtEl RIV 11 are provided on all air intake and exhaust 3.

Deleted openings. These dampers are designed to withstand a negative 1.46 psi pressure.

33.23 lifl~rct ofl' allure of Structures or Components Not Designed for Tornado loads 4 tiechtel Topical fleport 13C TOP 3-A, Revision 3, Tornado and Extreme Hind Design Critcria All safety related system and cornponents are for Nur/ car rower r/ ants.

protected within tornado. resistant structures.

See Subsection 3333 for interface requirement.

3.3.3 Inierfaces 3J3.1 Sitr Specine Design llasts Wind The site specific design basis wind shall not exceed the design basis wind given in Table 2.01 (See Subsection 2.2.1).

3.3.3.2 Site Specine Design llasts Torn do The site specific design basis torr, ado shall not exceed the design basis tornado given in Table 2.01 (See Subsection 2.2.1).

3.3.3.3 1;ffect of Remalnder of Plant Strue.

tures, Systems, and Components not Designed for Tornado Loads All remainder of plant structures, systems, and components not designed for tornado loads shall be analyred for the site specific loadings to ensure that their mode of failure will not effect the ability of the Seismic Category i AllWR Standard Plant structures, systems, and compo-nents to perform their intended safety functions.

(See Subsection 33.2.3) 3.3.4 References 1.

ANSI Standard A58.1, Minimum Desip: Loads for Buildings and Other Structures, Committee A. 58.1, American National Standards institute.

2.

ASCE Paper No. 3269, li'ind Forces on Structurcs, Transactions of the American Society of Civil Engineers, Vol.126, Part II, O

Amendment 25 3}2

ABWR man Sinndarditint nty n Flooding on this level may also wur from h

room cooling systems or from firefighting V

cfforts. Cooling system failures in air supply, exhaust or filter rooms may allow flooding at the rate of.3 cubic meter / minute (80 gpm) which will flow out into adjacent corridor areas if undetected for 10 minutes, the approximate 3 cubic meter (800 gallons) telcased may create a 3.4.1.1.2.1.6 Daluation of iloor 600 0F) depth of a few millimeters over the available floor area; a very limited amount of water will Flooding events at this floor level may cascade down the stairwells. Divisional areas involve f ucl oil as well as water. Those encompassing the thice emergency electric supply divisional rooms associated with the emergency fans and the RIP A exhaust willinclude raised diesel generator fuel tank and cooling system, sills to preclude water intrusion although water have the potential of leakage from the fuel depth will be slight. Equipment pedestals will storage tanks. These rooms must accommodate also minimire flooding impact on all equipment.

leakage of 11.4 cubic meter (3000 gallons) for each division. Twenty cm (8 inches) sills on Firefighting activities in this area would entry to these areas successfully contain all the cause water inflow of.57 cuble meter / minute volume in the tanks. Leakaga from these tanks (150 gpm) under controlled conditions and will also be monitored through safety grade level expected water intrusion is no more than that indication and alarm equipment so that protracted

above, leakage as well as gross leakage can be identified. The rooms are protected by CO, 3.4.1.1.2.1.8 Evaluation of Floor 800 (4F) firefighting system. Water flooding may occur from the cooling system at about.15 cubic Flooding on this floor can be caused by p

meter / minutes (41 gpm). If undetected for rupture of the RCW surge tanks A,11 & C piping.

Q several hours water may begin cascading down the llowever, each tank and its associated piping is nearest stairwell but is prevented from entering located in a separate compartment which can be other division areas by raised sills.

sealed off in the event of accidental flooding.

The use of raised sills on entry ways will in the SGTS areas, the room cooling equipment contain the scepage to the flooded area. Also, may cause flooding at a rate.15 cubic meter /

the use of pedestals for equipment installation minute (41 gpm). Raised sills prevent intrusion of the RIP supply and exhaused fans and for the of water into rooms of another division.

DG.C exhaust fans will guard against flooding Flooding may also occur from manual firefighting this equipment.

in equipment maintenance areas or from leakage from the standby liquid control tanks. Maximum Flooding in the main reactor hall may occur tank leak rate will bc.1 cubic meter / minute (25 from reactor service operations, but will be gpm) so that a sesponse to tank level alarms drained into service pools. Firefighting water within 10 minutes will limit loss to one cubic expended into this area would occur at a maximum will spread over the large service area {

rate of.57 cubic meter / minute (150 gpm) but meter (or 250 gallons). Large floor areas permit spr:ad of water at limited depth.

available. Minor amounts of water may find the 3.4.1.1.2.1.7 Evaluation of i loor 700 (M4F) way to stairwells, but would not impede operations.

Flooding in the FMCRD panel rooms may occur from firefighting activities at an input rate of 3.4.1.1.2.1.9 Flooding Summary Evaluation

.57 cubic rneters/ minute (150 gpm). Since these activities are manually controlled, any excessive Floor by floor analysis of potential pipe depth of water will be noted and action taken to failure generated flooding events in the reactor mitigate water intrusion to other areas, building shows the following:

Ov Amendment 25 345

AIlWR umu Sl!1fidllIdhlt[1t fut n (1) Where estensive flooding may occur in a blowdown will cause most of the steam to vent division rated compartment, propagation to out of the tunnel into the turbine building, other divisions is prevealed by watertight Water or steam cannot enter the control T

doors or scaled hatches. Flooding in one building. See Section 3.6.1.3.2.3 for a division is lirnited to that division and description of the subcompartment pressurization flood water cannot propagate to other analysis performed for the steam tunnel.

divisions.

Moderate energy water services in the control (2) Leakage of water from large circulating building comprise 28. inch service water lines, water lines, such as reactor building 18 inch cooling water lines,6. inch cooling cooling water lines rnay flood rooms and water lines to the chiller condenser,6. inch corridors, but through sump alarms and fire protection lines, and 6 inch chilled water leakage detection systems the control room heater lines. Smaller lines supply drinking is alerted and can control flooding by water, sanitary water and makeup for the chilled system isolation. Divisional areas are water system. Areas with water pipe routed protected by watertight doors, or where only through are supplied with floor drains and curbs limited water depth can occur, by raised to route leakage to the basement floor so that sills with pedestal mounted equipment within control or computer equipment is not subjected the protected rooms.

to water. In those areas where water infusion cannot be tolerated, the access sills are (3) Limited flooding that may occur from manual raised.

firefighting or from lines and tanks having limited inventory is restrained from Maximum flooding may occur from leakage in a entering division areas by raised sills and 28 inch service water line at a maximum rate of elevation differences.

12.0 cubic meters / minute (3150 gpm). Early detection by alarm to control room personnel Therefore, within the reactor building, will limit the extent of flooding which will internal flooding events as postulated will not also be mitigated by drainage to exterior of the prevent the safe shutdown of the reactor.

building. The expected release of a service water leak.is limited to line volume plus 3A.1.1.2.2 1 valuation of Control Ilullding operator response time times leakage rate. The Flooding thents assumed operator response time is 30 minutes to close isolation valves and turn off the pump in The control building is' a seven story the affected service water division. Water will building. It houses in separate areas, the be contained inside a division of closed cooling control room proper, control and instrument water equipment rooms in the bottom level of the cabinets with power supplies, closed cooling control building. A maximum of 2.15 meters of water pumps and heat exchangers, mechanical water in a divisional room is expected. Water equipment (IIVAC and chillers) necessary for tight doors will confine the water to a building occupation and environmental control for division, computer and control equipment, and the steam tunnel.

The failure of a cooling water line in the rucchanical rooms of the turbine building may l

The only high energy lines in the control result in a leak of 0.6 cubic meter / minute (160 building are the mainsteam lines and feedwater gpm). Early detection by control room personnel lines which pass through the steam tunnel will limit the extent of flooding. Total connecting the reactor building to the turbine release from the chilled water system will be building. There are no openings into the control limited to line inventory and surge tank volume, building from the steam tunnel. The tunnel is spillage of more than 6 cubic meters (1500 scaled at the reactor building end and open at gallons) is unlikely. Elevation differences and the turbine building end, it consists of separation of the mechanical functions from the reinforced concrete with 2 meter thick walls, remainder of the control building prevent Any break in a mainsteam or a feedwater line will propagation of the water to the control area.

flood the steam tunnel with steam. The rate of Amendment 20 344

ABWR m6-Shinillttif111111 nf!v. n 3J 1.1 Internally Generated blissiles (Outside system maintenance program including

[_}

Containment) probability calculations of turbine missile U

generation based on the NGC approved These missiles are considered to be those methodology (such as Reference 10), or missiles resulting internally from plant equipment failures within the ADWR Standard Plant (2) Volumetrically inspect all low pressure but outside containment.

turbine rotors at the second refueling outage and every other (alternate) refueling 3.5.1.1.1 Hotating Equipment outage thereafter until a maintenance l 3.5.1.1.1.1 bilsslie Characterir.ation (3) Meet the minimem requirement for the Equipment within the general categories of probability of turbine missile generation pumps, fans, blowers, dicscl generators, compres-given in Table 3.51.

sors, and turbines and, in particular, componcnts in systems normally functioning during power re-3.5.1.1.I.4 Other hilssile Analpls actor operation, has been examined for any possi-ble source of credible a-d significant missiles.

No remaining credible missiles meet the significance criteria,o[ having a probability 3.5.1.1.lJ RCIC Steam Turbine (P ) greater than 10 per year for rotating 4

or pressurired equipment, because either:

The RCIC steam turbine driving the pump is not a credible source of missiles. It is provided (1) The equipment design and manufacturing with mechanical overspeed protection as well as criteria mentioned pregously result in automatic governing: very extensive industrial (P ) being less than 10' per year; or 3

and nuclear experience with this model of turbine has never resulted in a missile which penetrated (2) Sufficient physical separation (barriers (q

the turbine casing, and/or distance) of safety related and redundant equipment exists so that the 3.5.1.1.1.3 hiain Strum Turbine combinep probability (P3 2

x P ) I' than 10 per year.

Acceptance criteria 1 of SRP Section 3.5.1.3 considers a plant with a favorable turbine gen.

These conclusions are arrived at by noting crator placement and orientation and adhering to that pumps, fans, and the like are AC powered, the guidelines of Regulatory Guide : 115 ad.

Their speed is governed by the frequency of the equately protected against turbine missile haz. AC power supply. Since the AC power supply ards. Further, this criterion specifies that frequency variation is limited to a narrow exclusions of safety related structures, systems range, it la not likely they will attain an or components from icw trajectory turbine missile overspeed condition. At rated speed, if a piece strike zones constiletes adequate protection such as a fan biade breaks off, i: will not against low trajectory turbine missiles. The penetrate the casing. The issue 'of missile turbine generator placement and orientation of generation in rotating machinery is a general the ADWR Standard Plant meets the guidelines of safety problem which is not limited to nuclear Regulatory Guide 1.115 as illustrated in Figure applications. The designers and manufacturers 3.3 2.

of these equipment consider this factor as a requirement in their design. Industrial

'dition, the applicant referencing the experience and studies conducted on system AI3WR aesign shali; components indicate that the probability of a missile escaping the casing is very low, GE has (1) Submit for NRC approval, within three years also conducted a study on potential missile of obtaining an operating license, a turbine generation from electrical machines (motors, exciters, generators), flexible couplings and

/]

fluid drives, One example where missile V

generation is significant is in fluid drives Amendment 25 3.5-3

ABWR mimn SlRIldard Plant RIV H 3.5.1.1.2 Pressurized Components 3.5.1.1.41 hilssile Characterization Potential missiles which could result from the fallore of pressurized components are analyzed in this subsection. These potential missiles may be categorized as contained fluid energy missiles or stored strain energy (clastic) missiles. These potential missiles have been conservatively evaluated against the design criteria in Subsection 3.5.1.

Examples of potential contained fluid energy missiles are valve bonnets, valve stems, and retaining bolts. Valve bonnets are considered jet propelled missiles and have been analyzed as such. Yalve stems have been analyzed as piston type missiles, while retaining bolts are examples of stored strain energy missiles.

3.5.1.1.2.2 bilsslie Analyses Pressutired components outside the contain-ment capable of producing missiles have been reviewed. Although piping failures could result O

O Amendment 16 33-3.1

ABWR mwm SIRudREdERI11 WV. H in significant dynamic effects if perrnitted to (2) Valve Stems. All the isolation valves

"]

whip, they do not form missiles as such because installed in the reactor coolant systems (J

the whipping section remains attached to the have stems with a back seat which climinates remainder of the whip. Since Secilon 3.6 the possibility of ejecting valve stems even addresses the dynamic effects associated with if the stem threads fall. Since a double pipe breaks, pipes are not included here as failure of highly reliable components would potential internal missiles, be required to produce a valve stem mistile, the overall p less than 10,r.pbability of occurrence is All pressurized equipment and sections of pip.

per year. llence valve ing that may periodienlly become isolated under stems can be dismissed as a source of pressure are provided with pressure relici valves missiles.

acceptable under the AShfE Code, Section Ill.

(3) Pressure Vessels Moderate energy vessels The only remaining pressurized components less than 275 psig are not credible missile considered to be potentially capable of producing sources. The pneumatic system air bottles missiles are:

are designed for 2500 psig to the ASME Code, Section til requirements. These bottles are -

(1) valve bonnets (large and small);

not considered a credible source of missiles for the following qualitative analysis:

(2) valve stems; (a) The bottles are fabricated from heavy.

(3) pressure ves els; wall rolled stect; (4) thermowells; (b) The operating orientation is vertical with the ends facing concrete slabs.

(5) retaining boks; and The bottles are topped with steel covers thick enough to preclude penetration by (6) blowout pancis.

a missile.

These arc analyzed as follows:

(c) The fill connection is protected by a permanent steci collar.

Valves of ANSI 900 (1) Valve llunnets psig and above and constructed in accordance (d) The bottles are strapped in a rack to with the ASME Code, Section til are prevent them from toppling over. The presrure. seal bonnet type valves, Valve rack is seismically designed to the ASME bonnets are prevented from becoming missiles Code, Section lil, Subsection NF by limiting stresses in the bolting to those require ments, defined by the ASME Code and by designing flanges in accordance with applicable code (4) Thermowells - Thermowells are welded to requirements. Safety factors involved socket connections which in turn are l against failure of these type bonnets are sufficiently high that _these pressure seal type valves are not considered a potential missile source (Ref. 9).

Most valves of ANSI 600 psig rating and below are valves with bolted bonnets. These type valves were analyzed for the safety factors against failure, and, coupled with the low historical incidents of complete severance failure, were determined to not be a potential missile source (Ref. 9).

O Amendment 15 3h

ABWR

==

Sinndard I!!anL niv n valent static load concentrated at the impact impact the safety function of a safety related

/ T area is determined. The structural response to systems and components will be provided to the V

this loud, in conjunction with other appropriate NRC by tbc applicant referencing the ADWR design loads, is evaluated using an analysis design. (See Subsection 3.5.1.4).

procedure similar to that in Reference 6 for rigid missiles, and the procedure in Reference 7 3.5.4.6 Turbine System Maintenance Program for deformable missiles.

A turbine system maintenance program 3.5.4 Interfaces including probability calculations of turbine missile generation meeting the minimum 3.$A.1 l'rotection of Ultimate llent Sink requirement for the probability of missile generation shall be provided to the NRC (See Comp!!ance with Regulatory Guide 1.27 as Subsection 3.5.1.1.3),

related to the ultimate heat sink and connecting conduits being capable of withstanding the 3.5.5 References effects of externally genetated missiles shall be demonstrated (See Subsection 3.5.2).

1.

C. V. Moore, The Design of Barricades for Hasardous Prc.ssure Systerss, Nuclear 3.5A.2 Missiles Generated by Natural Phenomena Engineering and Design, Vol. $,1967.

from Remainder of Hant Structures, Systems and Components 2.

F. J. Moody, Prediction of Blowdown Thrust and let Forces, ASME Publication 691 T 31 The remainder of plant structures, systems, August 1969.

and components shall be analytically checked to ensure that during a site specific tornado they 3.

A. Amirikan, Design of Protectirc Struc.

will not generate missiles exceeding the missiles tures, Bureau of Yards and Docks, Publica.

considered under Subsection 3.5.1.4.

Lion No. NAVDOCKS P 51, Department of the

[ )/

Navy, Washington, D.C., August 1960.

3.5AJ Site Proximity Miniles and Aircraft liarards.

4.

A. E. Stephenson, Full Scale Tornado Mis-site Impact Tests, EPRI NP 440, July 1977, Analyses shall be provided that demonstrate prepared ic t Electric Power Research that the probability of site proximity missiles Institute by bandia Laboratories.

(including aircraft) impacting the ABWR Standard Plant and causing consequences greater t,hpn 10CFR 5.

W. B. Cottrell and A. W. Savolainen, U. S.

Part 100 exposure guidelines is.<_10 per year Reactor Containment Technology, ORNL-(See Subsection 3.5.1.6).

NSIC 5, Vol.1, chapter 6, Ook Ridge No.

tional Laboratory.

3.5AA Secondary Missiles inside Containment 6.

R. A. Williamson and R. R. Alvy, Impaci Protection against the secondary missiles Effect ol Fragments Striking Structural inside containment described in Subsection Elcments, llotmes and Narver, Inc., Revised 3.5.1.2.3 shall be demonstrated.

November 1973.

7.

J. D. Riera, On the Stress Analysis of 3.5A.5 Impact ot Vallurc otNon Safety Related Structures Subjected to Aircraft impact Structures Systems,and Components Due to n Forces, Nuclear Engineering and Design, Design llasts Tornado North Holland Publishing Co., Vol. 8,1968.

An evaluation of all non safety related 8.

Deleted structures, systems, and components (not housed in a tornado structure) whose failure due to a p

design basis tornado missile that could adversely LJ Amendment 15 3$8

ABWR maan Sinndard Pinnt nwa 9.

River Dend Station Updated Safety Analysis Report, Docket No. 50 458, Volurne 6, pgs.

3.5 4 and 3.5 5, August 1987.

10 NUREG 1048, Safety Evaluation Report Related to tbc Operation of flope Creek Generating Station, Supplemcot No. 6, July 1986.

O O

Amendment 16 35-81

ABWR 2Wiman Mitadard Plant RW. H Subsection 3.9.2.5.

Dynamic analysis is per.

3.9.5.33 Design leading Categories g

i fortned by coupling the lumped mass model of the 4

V reactor vessel end internals with the building The basis for determining faulted dynamic model to detenmine the system natural frequencies event loads on the reactor internals la shown in j

and node shapes. The relative displacement, Sections 3.7, 3.8 and Subsections 3.9.2.5, acceleration, an:1 load response is then deter.

3.9.5.2.3 and 3.9.5.2.4. Table 3.9 2 shows the mined by either the time history method or the load combinations used in the analysis.

response spectrum method.

Core support structures and safety class 3.9.!J Design liases internals stress litnits are consistent with ASME Code Section 111. Subsection NO. For 3.9.53.1 Safety Design liases these components, Level A, D, C, and D service limits are applied to the normal, upset, The reactor internals including core support emergency, and faulted loading conditions, structures shall meet the following safety design respectively, a defined in the design bases:

specification. Stress int 6,nsity and other design limits are discussed in Subsections (1) The reactor vessel nonles and internals 3.9.5.3.5 and 3.9.5.3.6 shall be so arranged as to provide a floodable volume in which the core can be 3.9.5J.4 Response ofInternals Due to Steam adequately cooled in the event of a breach IJne Ilreak Accident in the nuclear system process barrier external to the reactor vessel; As described in Subsection 3.9.5.2.3.2, the maximum pressure loads acting on the reactor (2) Deformation of internals shall be limited to internal components result from steam line break assure that the control rods and core upstream of the main steam isolation valve and,

/'

standby cooling systems can perform their on some components, the loads are greatest with

(

safety.related functiont,; and operation at the minimum power ast 'sted with the maximum core flow (Table 3.94.,

se2).

(3) Mechanical design of applicable structures This has been substantiated by the analytical shall assure that safety design bases (1) cmnparison of liquid versus steam line breaks and (2) are satisfied so that the safe and by :he investigation of the effects of core shutdown of the plant and removal of decay power and core flow, heat are not impaired.

it has also been pointed out that, although 3 9.5J.2 Power Generation Desigu llases possible, it is not probable that the reactor would be operating at the rather abnormal The reactor internals including core support ec ndition of minimum power and maximum ccre structures shall be designed to the following flow. More calistically, the reactor would be power generation design bases:

at or near a full power condition and thus the maximum pressure loads acting on the internal (1) The internals shall provide the proper components would be as listed under Case 1 in toolant distribution during all anticipated Table 3.9 3.

normal operating conditiors to full power meration of the core without fuel damage; 3.933.5 Stress and Fatigue Limits for Core Support Structures (2) The internals shall be arranged to f acilitate refueling operations; and The design and construction of the core support structurce are in accordance with ASME (3) The internals shall be designed to Code Section 111, Subsection NG.

f acilitate inspection.

OV Amendment 7 3.9-43

ABWR 2umn Standard Plant nrv n 3.9.5.3.6 Stress, Deformation, nod Fatigue 3.9.6 In service Testing of Pumps rad Valves Limits for Safety Class and Other Reactor Internals (Except Core Support Structures)

In-service testing of safety related pumps and valves will be performed in accordance with For safety class reactor internals, the stress therequirementsofASME/AN". loma 198S Addenda deformation and fatigue criteria listed in Tablea to ASME/ ANSI OM 1987, Parts 1,6 and 10. Table 3.9 4 through 3.9 7 are based on the criteria 3.9 8 lists the in service testing parameters established in applicable codes and stardards for and frequencies for the safety related putnps and similar equipment, by manufacturers standards, or valves. The reason for each code defined by empirical methods based on field experienee testing exception or justification for each code and testing. For the quantity SF (minimum exemption request is noted in the description of safety factor) appearing in those;, tables, the the affected pump or valve. Valves having a following values are used:

containment isolation function are also noted in the listing. In service inspution is discussed Senice Senice in Subsection 5.2.4 and 6.6.

gp Lntl Condition

_mla Details of the in-service testing program, A

Normal 2.25 including test schedules and frequencies will be B

Upset 2.25 reported in the in service inspection and C

Emergency 1.5 testing plan which will be provided by the D

Faulted 1.125 applicant referencing the ABWR design. The plan will integrate the applicable test requirements Components inside the.cactor pressure vessel for safety related pumps and valves including such as control rods which must move during those listed in the technical specifications accident condition have been examined to (Chapter 16) and the containment isolation determine if adequate clearances exist during system, (Subsection 6.2.4). For example, the emergency and faulted conditions. No mechanical periodic leak testing of the reactor coolant clearance problems have been identified. The pressure isolation valves in Table 3.9-9 will be forcing functions applicable to the reactor performed in accordance with Chapter 16 internals are discussed in Subsection 3.9.2.5.

Surveillance Requirement SR 3.6.1.5.10. This plan will include baseline pre service testing The design criteria, loading conditions, and to support the periodic in service testing of analyses that provide the basis for the design of the components. Depending on the test results, the safety class reactor internals other than the the plan will provide a commitment to core support structures meet the guidelines of disassemble and inspect the safety related pumps NG 3000 and are constructed so as not to and valves when limits of the OM Code are adversely affect the integrity of the core exceeded, as described in the following support structures (NG 1122).

paragraphs. The primary elements of this plan, including the requirements of Generic Letter The design requirements for equipment 89-10 for motor operated valves, are delineated classified er non-safety (othe.) class internals in the subsections to follow. (See Subsection (e.g., steat dryers and shroud heads) are 3.9.7.3 f or COL lice nse in f orm a tion specified with appropriate consideration of the re q uir e m e n t s),

intended service of the equipment and expected plant and environmental conditions under which it 3.9.6.1 In. service Testing of Safety Related will operate. Where Code design requirements are Puraps not applicable, accepted industry or engineering practices are used.

The ABWR safety-related pumps and piping configurations accommodate in service testing at a flow rate at least as large as the maximum design llow for the pump, in addition, the O

Amendment 25 3.9-44 l

ABWR mama Standard Plant myn

- siring of each minNum recirculation flow path is experience. (See Suloction 3.9.7.3(1) for COL evaluated to assure that its use under all license information requirements.)

analy2Jd conditions will not result in degradation of the pump. The flow rate through 3.9.6.2.2 hiotor Operated Valves minimum recirculation flow paths can also be periodically measured to verify that flow is in The motor operated valve (h10V) equipment accordance with the design specification, specifications require the incorporation of the results of ai'her in situ or prototype testing The safety related pumps are provided with with ful e and pressure or full differential instrumentation to verify that the net positive pressurt e ify the proper sizing and correct suction head (NPSil) is greater than or equal to switch se.ags of the valves. Guidelines to the NPSil required during all modes of pump justify prototype testing are contained in operation. These pumps can be disassembled for Generic Letter 89-10, Supplement 1, Questions 22 i evaluation when Part 6 testing results in a and 24 through 28. The COL applicant will deviation which falls within the " required action provide a study to determine the optimal range." The Code provides criteria limits for frequency for valve stroking during in service the test parameters identified in Table 3.9 8. A testing such that unnecessary testing and damage program will be developed by the COL applicant to is not done to the valve as a result of the establish the frequency and the extent of testing. (See Subsection 3.9.7.3(2) for COL disassembly and inspection based on suspected license inforrr ation requirements),

degradation of all safetym lated pumps, including the basis for the lo quency and the The concerns and issues identified in extent of each disassembly. Tu program may be Generic Letter 89-10 for h10Vs will be addressed revised throughout the plant life to minimize prior to plant startup. The method of assessing disassembly based on past disassembly the loads, the method of sizing the actuators, experience. (See Subsection 3.9.7.3(1) for COL and the setting of the torque and limit switches license information requirements.)

will be specifically addressed. (See Subsection g

3.o.7.3(3) for COL license information 3.9.6.2 Inser Ice Testing of Safety Related requirements).

Vahes The in-service testing of MOVs will rely on 3.9.6.2.1 Check Vahes diagnostic techniques that are consistent with the state of the art and which will permit an of the performance of the valve under All ABWR safety related piping systems assessmee incorporate provisions for testing to demonstrate actual loaoing. Periodic testing per Part 10 l the operability of the check valves under design will be conducted under adequate differential conditions. In-service testmg will incorporate pressure and flow conditions that allow a the use of advance non-intrusive techniques to justifiable demonstration of continuing htOV periodically assess degradation and the capability for design basis conditions, perbrmance characteristics of the check valves.

including' recovery from inadvertent valve l The Part 10 tests will be pe: formed, and check positioning. MOVs that fail the acceptance valves that f ail to exhibit the required criteria, and are

• declared inoperable." for performance can be disassembled for evaluation.

stroke tests and leakage rate can be The Code provides criteria limits for the test disassembled for evaluation. The Code provides parameters identified in Table 3.9 8. A program criteria limits for the test parameters will be developed by the COL applicant to identified in Table 3.9-8. A program will be establish the frequency and the extent of developed by the COL applicant to establish the disassembly and inspection based on suspected frequency and the extent of disassembly and degradation of all safety-related pumps, inspection based on suspected degradation of all including the basis for the freque ccy and the safety-related "MOV's", meluding the basis for extent of each disassembly. The pagram may be the frequency and the extent of each revised throughout the plant life to minimize disassembly. The program may be revised disassembly based on past disassembly thioughout the plant Iife to minimize k

Amendmem ?.5 39411 l

ABWR 2-mn Standard Plant nw n disassembly based on past disassembly exper-lence. (See Subsection 3.9.7.3(1) for COL license information requirements.)

3.9.6.2.3 Isolation Yalve leak Tests The leak tight integrity will be verified.

for each valve relied upon to provide a leak. tight function. These valves include:

(1) pressure isolation valves valves that provide isolation of pressure differential from one part of a system from another or between systems; (2) temperature isolation valves - valves whose leakage may cause unacceptable thermal loading on supports or stratification in the piping and thermal loading on supports or whose leakage may cause steam binding of pumps; and (3) containment isolation valves valves that perform a containment isolation function in accordance with the Evaluation Against Crit e rion 54, S u bsection 3.1.2,$.5.2, including valves that may be exempted from Appendix J, Type C testing but whose leakage may cause loss of suppression pool water inventory.

Leakage rate testing of valves will be in accordance with Part 10, Paragraphs 4.2.2.2 and 4.2.2.3. An example is the fusible plug valves that provide a lower drywell flood for severe accidents described in Subsection 9.5.12, The valves are safety related due to the function of retaining suppression pool water as shown in Figure 9.5-3. These special valves are noted here and not in Table 3.9 8. The fusible plug valve is a nonreclosing pressure relief device and the Code requires replacement of each at a maximum of 5 year intervals.

O Amendment 25 3.9442

ABWR.

mamn Standard Plant nn n -

h Table 3,9 8 (Continued)

V "N. SERVICE TESTING SAFETLRELATED PUMPS AND VALVESL P25 IIVAC Emergency Cooling Water System Valves (Continued)

Safety Code Valve Test-Test.- SSAR' Class Cat. Func. Para Freq. Fig.

No. Qty Description (h)(1)

- (a)

(c)

(d)

(e)

(f)

-(g)

F022 3 IIECW supplyIo DG rone cooler Temp 3

B A-S E2-9.2 3(1,2,3)-

Cont Valve F023 3 Maint viv at IIECW supply to DG zone 3

B P

E1 9.2-3(1,2,3).

cooler TCV F024 6 Maint viv at ilECW supply to DG zone 3

B-P El -

9.2-3(1,2,3)'-

cooler FD25 6 Maint viv at ilECW return from DO zone 3

B P

El:

9.2-3(1,2,3)-

cooler F026 3 TCV byp viv at IIECW supply to DG zone 3

B P

El 9.2-3(1,2,3) cooler FU30 3 Chemical addition tank return viv 3

B P

E1.

9.23(1,2,3) from 11ECW F031 3 Chemical addition tank feed valve to llECW 3 B

P El 9.23(1,2,3)

F050 2 Makc up Water Purificd (MUWP)line to 3

C_

A S

E2 9.23(1,2,3) pump suction check valve FD70 5 Pump disch line drain valve 3

B P

_ El -- l 9.2-3(1,2,3)--

.O_

F400 $

Pump drain line valve 3

B P

E1-9.2-3(1,2,3),

F401 5 Pump bearing cooling wtr needle viv.

3 B-P E1

- 9.2-3(1,2,3)-

F402 3 Refrig outlet line sample valve 3

B P

El 9.23(1,2,3)

F700 5 Pump disch line pressure instr line root valve 3 B

P E1 9.23(1,2,3)

F701 5 FE P25-FE003 upstrm instr line root valve 3

B P

El 9.2-3(1,2,3)

F702 5 FE P25-FE003 dwnstrm instr line root v-tve 3

B P

El

9.2-3(1,2,3)

F703 5 Pump suction pressure instr line root valve 3

B P

E1 9.2-3(1,2,3)

F704 6 Pump suct/disch line dpt instr line root viv 3

B P

El 9.2-3(1,2,3)

P41 Reactor Service Water System Valves F001 6 Pump discharge line check valve 3

C A

S E2 9.2-7(1,2,3)

F002 6 Pump discharge line maintenance valve 3

B P

El 9.2-7(1,2,3)

F003 9 Senice water inlet valve to RCW System 3

B P

P 2 yrs 9.27(1,2,3)

- heat exchanger F004 6 Senice water inlet valve to senice 3

B P

P-2 yrs 9.2-7(1,2,3) water strainer F005 _9 Service water outlet valve from RCW 3-B-

F P.

2 yrs 9.2-7(1,2,3) -

heat exchanger,

F006 6 Senice water strainer blowout valve 3

B P-P 2 yrs 92-7(1,2,3)

F007-9 Supply line from Domestic water check valve 3 C

P El 92 7(1,2,3)

F008 9 Supply line from Domestic water check valve 3 C

P El 9.2-7(1,2,3)

F009 9 Supply valve from Domestic Water (DW) Sys 3 B

A P

2 yrs 9.2-7(1,2,3)

S E2 Amendment 23 3.9-58.24

,w

~

9

,,,y N-

ABWR DAMAU Stafidard Plant nu n Table 3.9 8 (Continued) g IN SERVICE TESTING SAFETY RELATED PUMPS AND VALVES P41 Reactor Service Water System Valves (Continued)

Safet) Code Valve Test Test SSAR Class Cat. Func. Para Freq. Fig.

No. Qty Description (h)(1)

(a)

(c)

(d)

(e)

(f)

(g)

F010 9 RCW llX tube side (senice water side) 3 C

P R

Syrs 9.27(1,2,3) relief valve F011 9 Bypass line around RCW IIX outlet line 3

C P-El 9.27(1,2,3) outlet valve MOV P41-F005 F012 9 Senice water sampling salve 3

B P

E1 9.2-7(1,2,3)

F013 6 Senice water strainer outlet valve 3

B A

P 2 yrs 9.2-7(1,2,3)

S E2 F014 3 Common senice water strainer outlet valve 3

B P

P 2 yrs 9.2-7(1,2,3)

F015 3 Discharge line to discharge canal MOV 3

B P

El 9.27(1,2,3)

F501 9 RCW llX shell side drain vahe to SWSD 3

B P

El 9.2-7(1,23) -

F502 9 RCW IIX shell side vent valve to SWSD 3

B P

El 9.2-7(1,2,3)

F503 9 RCW HX shell side drain valve to SWSD 3

B P

El 9.2-7(1,2,3)

F504 9 RCW 11X shell side vent salve to SWSD 3

B P

El 9.2-7(1,2,3) -

F701 6 Pump discharge pressure instr root valve 3

B P

El 9.2 7(1,2,3) ---

F702 3 Senice water supply pressure instr root valve 3 B

P El 9.2-7(1,2,3)

F703 6 Diff P across senice water strainer 3

B P

El 9.2-7(1,2,3) upstream instrument root valve F704 6 Diff P across senice water strainer 3

B P

El 9.2-7(1.2,3) downstream instrument root valve F705 9 Senice water diff P across RCW liX 3

B P

El 9.2 7(1,2,3)-

upstream instr root valve F706 9 Scnice water diff P across RCW 11X 3

B P

El 9.2-7(1,2,3) downstream instr root vale.

P51 Service Air System Valves F131 1 Outboard isolation manual valve 2

A-1,P L

RO 93-7 F132 1 Inboard isolation manual valve 2

A 1,P L

RO 93-7 PS2 Instrument Air System Valves F276 1 Outboard isolation valve 2

A 1,A 1.,P -

RO 93-6 F277 1 Inboard isolation check valve 2

A,C I.A 1,P RO 93-6 l

O Amendment 25 3.9-58.25

r m

LABWRT m uu Standard Plant

- Rev. h -

?

Table 3.9 8 (Continucd)!

IN. SERVICE TESTING SAFETY.RELATED PUMPS AND VALVES i

T19 Flammability Control System Valves.

Safety Code Valve Testi Test SSAR' Class Cat. ! Func. Para. Freq.- Fig.

No, Qty Description (h)(1)

(a)

(c)

(d)

'(e)

< (f)-

(g)

F003 2 Flow con:rol valve for the FCS inlet line 3-B

-A P-

2 yrs; 6.2-40 fron drywell S

3 so FON 2 Blower bypass line flow control valve 3

. B A

P

-2 yrs 6.2-40

'S 3mo-F005 2 Blower discharge line to wetwell check 3

- C A

S-RO-6.2-40 valve (h9)

F006 2 Discharge line to wetwelloutboatd

'2 A

I,A 1,P - 2 yrs - 6.2-40 '

isolation valve S

3 mo F007 2 Discharge line to wetwellinboard.

2-A 1,A -

1,P - - 2 yrs - 6.2-40 --

isolation valve S

3mo F008 2 Cooling water supply line from the RHR 3

B.

A-P 2 yrs - 6.2-40 System MOV S

-3 mo F009 2 Cooling water supply hne maintenance valve 3

D P

E1.

6.2-40 F010 2 Cooling water supply line admission MOV 3

B.

A.

P 2 yrs G.2-40 S

. 3 mo -

F013 2 Inlet line from drywc!! drain line valve 3

B P

E1 6.2-40 FD14 2 Blower drain line valve 3

B P

El J 6.2-40 F015 2 Blower discharge line to wetwell pressure 2.

A,C 1,A _

R 5 yrs 6.2-40 relief valve L

RO-F016 2 Blower discharge line to wetwell pressure 2

A,C 1,A 1,S RO 6.2-40 ~

reliefline check valve (h3)

F501 2 Inlet line from drywell test line valve 2

B P

El 6.2 F502 2 Discharge line to wetwell test line valve 2

B P

El 6.2-40 F504 2 Blower suction line test line valve 3

- B P'

El i 6.2-40 F505 2 Blower discharge line test line valve 3

B P

El 6.2-40:

F506 2 Drain line to Low Conductivity Waste

.3 B

P E1

-6.2-40

.(LCW) valve F507 2 Cooling water supply line test line valve 3

B P

E1

. 6.2-40 F701 2 FE T49-FE002 upstream instrument line 3

B P

El 6.2-40 root valve F702 2 FE T49-FE002 downstream instrument line 3

B P

El 6.2-40 root valve F703 2 Blower suction line pressure instrument line 3

- B P

E1

. 6.2-40.

root valve F704 2 FE T49-FEON upstream instrument line 3

B P

E1-6.2-40 root valve F705 2 FE T49-FE004 downstrcam instrument line 3

B P

El 6.2-40.

root valve Amendment 23 3.9 58.30 M

.a t=

b ABWR-

. amu -

Standard Plant nn. n -

Table 3.9 8 (Continued)

IN. SERVICE TESTING SAFETY.RELATED PUMPS ANI) VALVES -

U41 Ileating, Ventilating and Air Conditioning System Valves Safety Code-Valve - Test : Test - SSAR L Class-Cat. Fune. _ Parn : Freq. - Fig._

- No. Qty Description (h)(l)

-(a)

(c)

(d)

(e)

(f)

(g)

F001 2_

Reactor area supply isolation valve

- 2_ -

B A

P' 2 yrs

. 9.4 3(1)

S 3 mo-F002 2 Reactor area exhaust isolation valve 2

-B

_A P:

_ 2 yrs :

9.4-3(1) ~

S

'3 mo -

TV03 3 Reactor bldg area divisional llVAC supply 2-B A

_P 2 yrs 9.4-3(1) isolation valve S -

3 :no F004 3 Reactor bldg area divisional liVAC exhaust 2 B

A P

2 yrs 9.4 3(1)-

isolation valve :

S 3 mo F007 4 MCR area liVAC bypass line isolation valve 2 B

-A P-2 yrs - 9.4-1(1,2).

S 3 mo T008 4 - MCR area IWAC supplyisolation valve 2

B A

_P 2 yrs 9.41(1,2)

S 3 mo FD09 4 MCR area IIVAC emergency liVAC supply 2 B?

!A P'

2 yrs 9.41(1,2) -

S 3 mo IV10 4 MCR area llVAC exhaust isolation valve-2 B

A P

2 yrs-; 9.41(1,2)

S_

- 3 mo_

-~

~

YS2 Oil Storage Transfer System Valves --

F001 6 D/G transfer pump discharge lint. check viv_ 3 C

A-S 3mo 9.5-6 --

F002 3 D/G transfer pump discharge line relief viv - 3 C

A-R

-5 yrs 9.5-6 F003 3 D/G transfer pump discharge line ball (plug) B-P.

El -

9.5-6 valve F004 3 D/G fuel oil day tank return to storage 3

B P

E1-9.5-6 --

tank valve F501 3 D/G transfer pump discharge lin: drain viv 3

B P

El' 9.5 F502 3 D/G transfer pump discharge line sent viv

-3 B1 P

E1

-9.5-6c

_0 Amendment 23 3.9-58.31 -

r m.

..-.-w

?

k' LABWRL mamu

Standard Plant ne n 7

Table 3.9 8 (Continued)'-

1 IN SERVICE TESTING SAFETY RELATED PUMPS AND VALVES -

l NOTES:

(a) 1,2, or 3 Safety Classification, SSAR Subsection 3.23.

-(b)

Pump test parameters per ASME/ ANSI OMa 1988 Addenda to ASME/ ANSI OM 1987, Part 6:

b N. -

Speed-Pd-Discharge Pressure i

Pi.

Inlet Pressure 0-Flow Rate Vd-Peak to-peak vibration displacement Vv.

Peak vibration velocity

.(c)

A, B, C or D - Valve category per ASME/ ANSI OMa.1988 Addenda to'ASME/ ANSI OM 1987, Part l'.-[

and 10.

1 (d)

Valve function:

1-Primary containment isolation, SSAR Subsectiori 6.2.4 -

A or P - Active or passive per ASME Code in (c) above (Part 10, Paragraph 13).

(c)

Valve test parameters per ASME Code in (c) above:

L-Leakage rate (Part 10, Paragraph 4.2.2, SSAR Table 6,2-7 for valves with function I in (d) above)).

P-Local position verification (Part 10, Paragraph 4.1)

R-Relief vahe test including visual examination set pressure and seat tightness testing (Part 10, Paragraph 43.1 and Part 1, Paragraph 133 and 13.4).

S-Stroke exercise Category A or B (Part 10, Paragraphs 4.2.1.1,4.2.1.2) -

Category C (Part 10, Paragraphs 43.2.1,43.2.2,43.2.4)

X-Explosive charge test (Part 10, Paragraph 4.4.1)

(f)

Pump or valve test exclusions, alternatives and frequency per ASME Code in (b) or (c) above or Appendix 1:

-- CS-Cold shutdown RO-Refu; ling outage and/or no case greater than two years.

El-Used for operating convenience,i.e., passive vent; drain, instrument, test, maintenance <

valves, or a system control valve. Test are not required (Part 10, Paragraph 1.2),

_ _ _fl _

E2-In regular use. Test frequencyis not required provided the test parameters are analyzed p

and recorded at an operation interval not exceeding three months.

Category A or B, Stroke (Part 10, Paragraph 4.2.1.5).

Category C, Stroke (Part 10, Paragraph 43.23).

E3-Operability rest every six months. Set pressure and leak test every refueling outage.

(Part 1, Paragraph 13.43).

ll E10.-

In Regular use. Test frequency is'not required prcvided the test parameters are _

recorded at least once every three months of operation (Part 6, Paragraph 53).

l~

Ell-Lacking required fluid inventory. Test shall be performed at least once every two years with required fluid inventory provided (Part 6, Paragraph 5.5).'

l Amendment 25 3.9-58.31.1--

n

ABWR 23A61MAE Standard Plan l

' Rev. H Tsble 3.9 8 (Continued) g-IN SERVICE TESTING.iAFETY.RELATED PUMPS AND VALVES -

NOTES (Continued):

(g)

Piping and instrument symbols ano abbreviations are defined in Figure 1.71. Figure page numbers are shown in parenthesis ().

(h)

Reasons for code defined testing exceptions (Part 10, Paragraphs 4.2.1.2 and 4.3.2.2).

l (hl) Inaccessible inerted containment and/or steam tunnel radiation during power operations.

(h2) Avoids valve damage and impacts on power operations.

(h3) Avoids impacts on power operations.

(h4) A temporary crosstic is necessary to carry the ongoing cooling loads. A permanent crosstic would -

violate divisional separation.

(h5) Avoids cold / hot water injection to RPV during power operations, j

(h6) Maintain pressure isolation during normal operation.

(h7) Inventory available only during refueling outage.

(h8) Testing at power will impact operation because the valves do not automatically isolate with a '

LOCA signal.

(h9) Test connection size is insufficient for full now test during operation. Therefore, test part stroke during plant operation and full stroke during refueling outage. A test connection size which would be sufficient for full flow tests would pressurize the secondary containment beyond specified limits, thus affecting power operation.

(i)

Summaryjustification for code exemption request (Part 6, Paragraph 5.2, or Part 10, Paragraph 6.2).

l (il) The piping is maintained full by a small fraction of the pump's flow capacity. These pumps may.

I be a constant speed centrifugal type with a cooling by-pass loop. Normal operation will be near minimum flow in the flat or constant region of the pressure / flow performance curve. Therefore, a flow measurement would not be useful. The pumps will be designed and analyzed to withstand low flow operation without significant degradation.

O Amendment 25 3.9-583 t.2

_ = - _

h ABWR-mams -:

Stamf ard Plant nev. 4 -

- n/f

. SECTION 4C-

~ ~\\_

- CONTENTS Section Iltle j'agt 4C CONTROL ROD LICENSING ACCEPTANCE CRITERIA-4 4C.1 Introduction 4C-1 4C.2 General Criteria 4C 1 4CJ-. Basis for Acceptance -

' 4C 1 4C3.1' Stess, Strain and Fatigue 4C-1 4C,3.2 Control Rod Insertion

~ 4C 1 4C33 Control Rod Material 4C-1 4C3.4 Reactivity

4C 1.

l 4 C3.5 Surveillance Criteria 4C 1.1 4C.4 Refrences 4C-1,1 -

OV

t. j i:

L I'

~

l 4C*ii Amendment 23 i

ABWR 2mmn -

Standard Plant iuw. c SECTION 5A

.]

CONTENTS (Continued)

Section Title Eage 5.4.9.2 Power Generation Design Bases 5.4 27 5.4.93 Description

.5.4-28 5.4.9.4 Safety Evaluation 5.4-28 5.4.9.5 Inspection and Testing 5.4-28 5.4.10 Pressurircr 5.4-28 5.4.11 Pressurizer Rt]lef Discharee System 5.4-28 5.4.12 Valves 5.4-28.1 5.4.12.1 Safety Design Bases 5.4 28.1 5.4.12.2 Description 5.4-29 5.4.123 Safety Evaluation 5.4 29 5.4.12.4 Inspeetion and Testing 5.4-29 5.4.13 Safetv/ Relief Valves 5.4-29 =

5.4.13.1 Safety Design Bases 5.4-29 5.4.13.2 Description 5.4-30 5.4.13 3 Safety Evaluation 5.4-30 5.4.13.4 (Deleted) 5.4-30 5.4.14 Comnonent Sunnorts 5.4-30 5.4.14.1 Safety Design Bases 5.4 30 -

5.4.14.2 Description 5.4-30 5.4.14 3 Safety Evaluation 5.4 5.4.14.4 Inspection and Testing 5.430 5.4.15 References 5.4-30 5.4-vi Amendment 25

ABWR ammn Standard Plant nrv. c -

SECTION 5.4 O

TABLES Table Illle Page 5.4-1 Reactor Recirculation System Design 5,4 32 Characteristics 5.4-2 Design Parameters for RCIC System Components 5.4-33 5.4-3 RHR Pump / Valve Logic 5.4 38 5.4-4 RHR Ileat Exchanger Design and 5.4-40 Performance Data -

5.45 Component and Subsystem Relief Valves 5.4-41 5.4-6 Reactor Water Cleanup System Equipment Design Data 5.4-43 ILLUSTRATIONS Figure Iule Page 5.4 1 Reactor Internal Pump Cross Section 5.4-44 5.4-2 ABWR Reeirculation Flow Path 5.4-45 5.4-3 Reactor laternal Pump Reference Characteristics 5.4-46 5.4-4 Reactor Recirculation System P&ID 5.4-47 5.4-5 Reactor Recirculation System PFD 5.4-48 5.4-6 Main Steamline Flow Restrictor 5.4-49 5.4-7 Main Steam Isolation Valve 5.4-50 5,4-8 Reactor Core Isolation Cooling System P&ID 5.4-51 5.4-9 RCIC System PFD 5.4-53 5.4-10 Residual Heat Removal P&ID 5.4-55 5.4-11 RHR PFD 5.4-59 5.4-12 Reactor Water Cleanup System P&ID 5.4-61 5.4-13 Reactor Water Cleanup System PFD 5.4-63 5.4-vil Arnendment 15

1 ABWR maman Standard Plant RI?V A motor cavities are purged by water from the the high pressure block valves is designed to

/3 control rod drive system. CUW system return flow Quality Group D.

)

is directed to either the nuclear boiler r.ystem x

(feedwater lines), directly to the RPV through A tabulation of CUW system equipment data, the RPV head spray, suppression pool or radwaste including temperature pressure and flow capacity through the CUW dump line. CUW filter-is provided in Table 5.4-6, demineralizer backwash is to the backwash receiving tank (BWRT) located in the FPC (BWRT SA.9 Main Steamlines and Feedwater Piping accommodates backwash from the CUW the FPC, and the suppression pool cleanup system). The 5.4.9.1 Safety Design Bases non-regenerative heat exchanger is cooled by the reactor building cooling water system. Other In order to satisfy the safety design bases, utility or support interfaces exist with the the main steam and feedwater lines are designed instrument air system and the condensate and as follows:

plant air systems for the filter-demineralizer backwash.

(1) The main steam, feedwater, and associated drain lines are protected from potential The type of pressure precoat cleanup system damage due to fluid jets, missiles, reaction used in this system was first put into operation forces, pressures, and temperatures in 1971 and has been in use in all BWR plants resulting from pipe breaks, brought on line since then. Operating plant experience has shown that the CUW system, de-(2) The main steam, feedwater, and drain lines signed in accordance with these criteria, are designed to accommodate stresses from provides the required BWR water quality. The internal pressures and carthquake loads ABWR CUW system capacity has been increased to a without a failure that could lead to the nominal of 2% of rated feedwater from the release of radioactivity in excess of the original 1% sire. This added capacity provides -

guideline values in published regulations.

()

additional margin against primary system

'V intrusions and component availability. The (3) The main steam and feedwater lines are nonregenerative heat exchanger is sized to acces-sible for inservice testing and maintain the required process temperature for inspection.

100% system flow. During periods of water rejection to the suppression pool or radwaste, (4) The main steamlines are analyzed for dynamic CUW system flow may be reduced slightly to loadings due to fast closure of the turbine compensate for the loss of cooling flow through stop valves.

the RPV return side of the regenerative heat exchanger.

(5) The main steam and feedwater piping from the reactor through the scismic interface The CUW system is classified as a nonsafety restraint is designed as Seismic Category I.

system. The reactor isolation valves are l classified as safety related. System piping and (6) The main steam and feedwater piping and components within the drywell, including the smaller connected lines are designed in suction piping up to and including the outboard accordance with the requirements of Table suction isolation valve, and all containment 3.2-1.

isolation valve including interconnecting piping assembly, are Scismic Category I, Ouality Group 5.4.9.2 Power Generation Design Bases l A. All other non-safety equipment is designed as Nonseismic, Quality Group C. Low pressure piping (1) The main steamlines are designed to conduct in the backwash and precoat area downstream of steam from the reactor vessel over the full range of reactor power operation.

f Amendment 23 SM7

3 cn LABWR1

%=a

' Standard Plant RFV C '

(2)- The feedwater_ lines are designed to conduct 1 ~ Group A from the reactor pressure vessel out to-

~

water to the reactor vessel'over the_ full- 'and including thel outboard isolation _ valve,.

T range of reactor' power operation.

_ Ouality Group B from the.. outboard isolation' valve to and including the _ seismic interface, 5.4.9.3 = Description

restraint, and Ouallty Group.D, bey'ond the;

_ shutoff valve. The feedwater ' piping and allL The main steam piping is described in Section ' connected piping'of 21/2; inch larger nominalj 10.3. The main steam and feedwater piping from size is Seismic Category I only from the reactor the reactor through the containment isolation ' pressure vessel out to, and including, the interfaces is diagrammed in Figure 5.13.

seismic' interface' restraint.

~

As discussed in Table 3.21 and shown in The materials-used in the piping are.in

~

Figure 5.1-3, the main steamlines are Quality accordance with the applicable design code and

~

Group A from the reactor vessel out to and includ _ supplementary requirements described in Section~

ing the outboard MSIV_ and Quality Group B from - 3.2. The valve between the outboard isolation.

the outboard MSIVs to the turbin'c stop valve, valve and the shutoff valve upstream of the RilRT They are also Seismic Category I only from the-entry to the feedwater line is to effect a -

t reactor pressure vessel out to the seismic inter-closed loop outside containment _(CLOC) for, face restraint.

containment bypass leakage control (Subsections-6.2.6 and 6.5.3).

The feedwater piping consists'of two 22 inch diameter lines from the feedwater supply header The general requirements of the feedwater to the reactor. On each of the feedwater lines system are described in Subsections 7.L1.7, ~

from the common feedwater supply header, there 7.7.1.4, 7.7.2.4i and 10.4.7.

shall be a seismic interface restraint. The seismic interface restraint serves as the. 5.4.9.4 Safety Evaluation boundary between the Scismic Category I piping c

1 and the nonoeismic piping. Downstream of the Differential pressure on reactor ' internals:

seismic interface restraint, there is a remote under the assumed accident condition o r a rup- -

manual, motor operated valve powered by a-tured steamline is limited by the use'of flow.

non safety grade bus. These motor-operated restrictors and by the use'of four main : team -

valves serve as the shutoff valves for the lines. All main steam and feedwater piping _wili ' >

feedwater lines. Isolation of each line is be designed in accordance with the requirements ~

accomplished by two containment isolation valves defined in Section 3.2. t Design of the piping in-consisting of one check valve inside the drywell accordance with these requirements ensures and one positive closing check valve outside the meeting the safety design bases, containment (Figure 5.1-3). The closing check valve outside the containment is a spring closing; 5.4 9.5 Inspection and Testing.

check valve that is held open by air. These"

]

check valves will be qualified to withstand the

. Testing is carried out-in accordance with:

1 dynamic effects of a feedwater line break outside Subsection 3.9.6 and Chapter 14. I:ncevice 9

containment. Inside the containment, downstream inspection is considered in the <'.esign of the' d

of the inboard FW line check valve, there is a main _ steam and feedwater piping. This consider :-

, manual maintenance valve (B21 F005) ation assures adequate working spsce and access for the inspection of selected compoaents.

- The design temperature and pressure of the feedwater line is the same as that of the reactor 5.4.10 Pressurizer inlet nozzle (i.e.,1250 psig and 575 F)for turbine driven feedwater pumps.

.Not Applicable to BWR -

5.4.11 Pressurizer Relief Discharge System i

As discussed in Table 3.2-1 and shown in Not Applicable to BWR 1"

Figure 5.13 the feedwater piping is Quality L,

Amendment 25 5.4-28

_L.

ABWR-numn Standanwlant nw c C]'

/

5.4.12 Valves 5.4.12.1 Safety Design llases Line valves, such as gate, globe, and check O

l Amendment 25 5.4-28.1

ABWR

--r Standard Plant myn Q.

tem (see Section 7.6). The objective of steamlines will result in trip of the Q

the main steamline radistion monitoring MSIV isolation logic to close the MSIVs subsystem is to monitor. for 'the gross' and main steam drain valves. Valve iso-release of fission products from the lation is annunciated in the control fuel, and upon indication of such re-

room, lease, initiate appropriate action to limit fuel damage and further release of (e) Main Steam Line 14w Pressure Monitoring fission products.

Four pressure transmitters are provided The process radiation monitor detectors to sense the pressure downstream of the are physically located near the main outboard MSIVs and to initiate MSIV iso-steamlines just downstream of the out-lation on low pressure indications, board main steamline isuiation valves.

These transmitters are located as close The detectors are geometrically arranged as possible to the turbine stop valves, to detect significant increases in ra-diation level with any number of main Steam pressure at the turbine inlet is steamlines in operation.

monitored to provide protection against a rapid depressurization of the reactor When a significant increase in the main vessel, which could be caused by the steamline radiation level is detected, turbine bypass valves failing to the trip signals are transmitted to the reac-fully open position. The low pressure for protection system (RPS). The RPS indication is annunciated in the initiates reactor trip and signals the control room.

LDS to initiate closure of all main steamline isolation vaIves and the (f) Main Condenser Low Vacuum Monitoring steamline drain valves.

(9 Low main condenser vacuum could indi.

V (c) Main Steam Line Area Temperature cate that primary reactor coolant is Monitors being lost through the main condenser.

Four divisional channels of the main-Single element thertnocouples are located condenser pressure monitoring are pro-in the MSL tunnel area to monitor for vided by the nuclear boiler system.

high ambient temperature. The detectors The LDS utilizes the low vacuum signal are shielded so that they are sensitive from the rnain steam monitoring channels to MSL area ambient temperature and not to trip the MSIV isolation logic on low to radiated heat from hot equipment.

condenser vacuum, thereby closing the The element provides input to the LDS MISVs and steamline drain valves. The for MSIV isolation when a preset high condenser vacuum signal can be bypassed temperature condition (potentially by a manual keylocked bypass switch in indicative of a main steamline steam the control room during startup and leak) is detected, shutdown operation.

(d) Main Steam Line Flow Monitoring (g) CUW Differential Flow Monitoring Differential pressure transmitters sense The suction and discharge flows of the the pressure difference across a single reactor water cleanup system are flow element in each MSL. The setting monitored for flow differences. Flow is selected high enough to permit clo-differences greater than preset values sure of one MSIV for testing at rated cause alarm and isolation. Delay tim-power without causing isolation of the ers provide for delaying the isolation other MSLs, yet low enough to permit signal to accommodate normal system p

early detection of a steamline break.

surge conditions. Two divisional chan-

%.j H1gh flow in any two of the four main nels of differential flow are provided Amendment 4 7.3-19b

ABWR.

m imar Standard Plant nrv.n by the LDS for this function as fol.

(j) RCIC Steam line Flow hionhoring lows: flow in the CUW suction line from the reactor and flow in the CUW return The steam supply line fro motive power lines to the reactor and to the main con-to drive the RCIC turbine is monitored denser are monitored by six differential for abnormal flow. Two channels of 1

pressure (flow) transmitters (two for flow measurements are provided by LDS cach line). The outputs of the flow for detection of steamline breaks by transmitters in the suction line are com-flow transmitters which sense differen-pared with the flow outputs from the dis-tial pressure across elbow taps in the charge lines by electronic equipment steamline. High steam flow or an in-which trips on high differential flow, strument line break will result in the The six transmitters and electronic closure of the RCIC steamline isolation equipment are arranged to provide one Di-valves, and warm up bypass valve, and vision I and one Division 11 measure-turbine trip if running. A trip signal ment and trip function. The Division from Division 11 isolation logic will 11 channel trip will close the inboard close the outboard isolation valve CUW isolation valves and Division I while a Division I trip will _close the channel trip will close the RWCU out-inboard RCIC steamline isolation valve board isolation valves, and warm up bypass valve. Any isola-tion signal to the RCIC logic will also (h) Drywell Pressure hionitoring trip the RCIC turbine. The elbows and taps are shown on the RCIC P&lD (Figure Drywell pressure is monitored by four di-5.4-8). The transmitters and assocl-l visional pressure transmitters relative ated trip channels are shown on the LDS to containment pressure. These transmit-IED (Figure 5.2-8),

ters are provided by the nuclear boiler system and are shared with other (k) Drywed Temperature hionitoring systems. The transmitters are mounted in local panels within the reactor build-The ambient temperature within the ing. Instrument sensing lines that drywell is monitored by four single ele.

connect the transmitters with the ment thermocouples located in the verti-drywell interior physically interface cal direction within.the drywell with the containment system, equally spaced. An abnormal increase in drywell temperature could indicate a Four channels (one in each of the four leak within the drywell, Ambient tem-divisions) provide signals to LDS isola-peratures within the drywell are re-tion logic.

corded and alarmed in the control room.

(i) Drywell Cooler Condensate Flow hionitor-(1) Valve Leakage hionitoring ing Large remote power-operated valves lo-The condensate flow rates from the cated in the drywell for the nuclear dr>well atmosphere coolers are monitored boiler, reactor water cleanup, reactor for high drain flow, which indicate core isolation cooling, and residual leaks from piping or equipment within heat removal systems are fitted with the drywell This flow is monitored by drain lines from the valve stems. Each one channel of flow instrumentation lo-drain line is located between two sets cated to measure flow in the common con-of valve stem packing. Leakage through densate cooler drain line which drains the inner packing is carried to the the condensate from all of the drywell drywell equipment drain sump. Leakage coolers to the drywell floor drain during hydrotesting may be observed in sump. The high flow indication is drain line sight glasse4 installed in alarmed in the control room.

each drain line. A remote operated so-Amendment 25 73-19e

ABWR maw Standard Plant nrv. n lation requirements during a loss of coolant Power supply failures p) accident inside the drywell. Components of the LDS that are located inside the drywell Individual valve position indication ad-and that must operate during a jacent to valve control switches loss of coolant accident are the cables, control mechanisms and valve operators of All nonessential indications and alarms isolation valves inside the drywell. These (i.e., annunciator, computer inputs) are isolation components are required to be func-electrically and physically isolated from tional in a loss of-coolant accident environ-the isolation logics to preserve the integ-ment (see Section 3.11). Electrical cables rity of the isolation function in the event are selected with insulation designed for of a failure in nonsafety-related equip-this service. Closing mechanisms and valve ment. See Chapter 16 for setpoints and mar-operators are considered satisfactory for gin.

use in the isolation control system only after completion of environmental testing The CUW isolation logic receives inputs under loss of coolant accident conditions or originating from starting the standby submittal of evidence from the manufacturer liquid control system (SLC). These input describing the results of suitable prior signals are required to isolate CUW when tests.

SLC is started. The RiiR system isolation logic is provided with input signals from (12) Operational Considerations pressure transmitters monitoring reactor pressure. These pressure transmitters The LDS is on continuously to monitor con-prevent opening the RHR shutdown cooling tainment leakage during normal plant op-valves and CUW head spray valve whenever cration. The system will automatically func-the reactor pressure is above a preset tion to isolate a reactor coolant leak exter-value. This signal is pro vided as an in-nal to the containment and prevent unaccept-terlock and is not provided for containment L

able radiological releases from the contain-or reactor vessel isolation.

ment following detection of a leakage within the containment. No operator action is re-(13) Parts of System Not R quired for Safety quired for at least 10 minutes following system initiation.

The nonsafety related portions of the LDS include the circuits that' drive annun-The following information is alarmed and/or ciators and the computer. Other instrumen-indicated in the control room. Indication nation considered nonsafety related are is provided by instruments, displays, record-those indicators which are provided for op-ers, status lights, computer readout or an-erator information, but are not essential nunciator alarms.

to correct operator action.

Manual system level isolation armed and initiated Instrument channel trips isolation logic trips Logic failures or out of service All bypasses Valve overrides

!Q Test status Amendment 4 7.319j

ABWR 2mme Standard Plant nev.n 7.3.1.1.3 RilR/Wetwell and Drywell Spray Cooling pump loops, each loop with its own sepa-Mode Instrumentation and Controls rate discharge valve. All components per-tinent to wetwell and drywell spray opera-(1) System Identification tion are located outside of the drywell.

Wetwell and drywell spray cooling (WDSC) is Motive and control power for the two loops an operating mode of the residual heat remo-of wetwell and drywell spray instrumenta-val system, it is designed to provide the tion and con. trol equipment are the same as capability of condensing steam in the wet-those used for RilR B and RilR C.

well air volume and the containment atmos-phere and removing heat from the suppression The drywell spray cooling system can be pool water volume. The mode is manually in.

manually initiated from the control room if itiated when necessary, drywell pressure is above the high setpoint, allowing the operator to act in the event of The RiiR System is shown in (P&lD) Figure a loss-of coolant accident, in the absence 5.4-10.

of high drywell pressure conditions, the drywell spray valves cannot be opened.

(2) Supporting Systems (Power Sources)]

The wetwell spray cooling can be manually Power for the RilR system pumps B and C is initiated in the control room. The operator supplied from two independent AC buses that relies on the instrumentation that provides can receive standby AC power. Motive and indication of the wetwell air space control power for the two divisions of wet-temperature condition when initiating this well and drywell spray cooling and instru.

mode. No interlock is provided, mentation and cor. trol equipment are the same as those used for LPFL B and C, respectively (a) Initiating Circuits (see Subsection 73.1.1.4).

Drvwell Snrav B:

(3) Equipment Design Drywell pressure is monitorcU by four Control and instrumentation for the fol-shared pressure transmitters mounted in lowing equipment is required for this mode instrument racks in the containment, of operation:

Signals from these transmitters are Two RHR main system pumps, routed to the local multiplexer units which convert analog to digital signals Pump suction valves, and send them through fiber optic links for logic processing in the control Dr>well spray discharge valves.

room. Any two out of four signals pro vide the permissive to initiate the Wetwell spray discharge valves.

WDSC Variables needed for the operation of the initiation logic for drywell spray "B" drywell spray equipment are high pressure is identical to drywell spray "C".

conditions in the drywell air space. The instrumentation for wetwell and drywell Wetwell Snrav 8:

spray operation ensures that water will be routed from the suppression pool to the The initiation of wetwell spray is man-wetwell and drywell air volumes, ual and does not have an interlock. The operator bases judgment on the instru.

Wetwell and drywell spray operation uses two mentation indication of the condition of 9

Amendment 25 7.3 20

ABWR 2.sacioore Standard Plant un 7A.7 HESPONSES TO SUllSECTIONS 7A.5 &

generally found to be not applicable to the

[]

7A.6; COMPUTER ll ARDWARE AND SOFTWARE BWR/ABWR reactor design philosophy.

V Items 7A.5(1) and 7A.5(2):

The NUREG discusses a " core protection calculator system (CPCS)* which is designed to Criteria and guidelines stated in provide reactor protection for two conditions: (1)

ANSI /IEEE-ANS 7.4.3.2, as endorsed by Regulatory low local departure from nucleate bolling ratio Guide 1.152, have been used as a basis for design (DNBR), and (2) high local linear power density.

procedures established for programmable digital equipment.

For condition (1),"DNBR" is associated with PWRs and is not applicable to BWRs. For condition All programmable digital equipment utilized for (2), power density is determined via the neutron safety-related functions are qualift d in accordance monitoring system (NMS), similar to methods used with safety criteria and with the safety system design in operating BWRs. (See Subsection 7.6.1.1 for basis with which they interface.

discussion of the NMS.)

A structured, engineered approach to the The ABWR design of the reactor protection development of both hardware and software is system utilizes microprocessor technology for logic implemented to assure that the design proceeds decisions based on analog input from various along the lines of the requirement specifications ant' sensors. This philosophy is much the same as that of has traceable documentation.

GESSAR 11 and the Clinton BW 1, except in those designs, solid state CMOS accepted digital signals Verification and validation (V&V)includer the from analog trip modules (ATM). In the ABWR establishment of test and evaluation criteria, t..

design, ahe microprocessors perform the functions of development of test and evaluation procedures, the both the CMOS and the ATM.

testing of the integrated hardware and software, and the installation of the hardware and software in the The important distinction is that the ABWR uses a tO field.

modern form of digital compu!cr device (i.e.,

V microprocessors) for the same reasons relays anj in accordance with the step-by-step verification solid-state devices were used in earlier designs (i.e.,

process, design resiews are performed at the system mak;ng simple logic decisions); not for making functional and performance requirements complex calculations for which protective action is specification / task analysis and allocation of functions dependent.

level, the hardware design and the software design level, the test and evaluation criteria and procedures items 7A.5(4) and 7A.6(4):

level, and the personnel requirements and operating / maintenance plan level. Such reviews are The guidelines of NUREG-0493 have been used conducted by knowledgeable and experienced system to perform analysis of several possible different engineers, software engineers, hardware engineers, configurations of the safety system logic and control etc., who are not directly responsible for the design, (SSLC) network. Analyses have been performed at but who may be from the same organization.

Ihe system design level to assure adequate defense-in-depth and/or diversity principles were Figure 7A.7-1 illustrates the structure utilized for incorporated at acceptable cost, it is recognized that ABWR control and instrumentation system design such requirements are in addition to positions on which incorporates subject guidelines.

safety-related protection systems (such as the single failure criterion) taken previously in other Regulatory Guides.

Items 7A.5(3) and 7A.6(2):

In order to reduce plant construction costs and NUREG-0308, " Safety Evaluation Report -

simplify maintenance operation, the ABWR Arkansas Nuclear 1, Unit 2* was reviewed and protection sytems are designed with a " shared O

V Amendment 22 7A.7-1

r ABWR umw Standani Plant un sensors" concept. The SSLC is the central processing hydra u1ic ser a m.

11 o w e v e r, Ihf.

mechanism and produces logic decisions for both Class 1E to-non Class 1E isolated interface is a RPS and ESF safety system functions. Redundancy special case for mitigation of ATWS and is not a and

• single failure" requirements are enhanced by a control system interface, full four division modular design using 2-out of 4 voting logic on inputs derived from LOCA signals The ABWR demonstrates strong multi system which consist of diverse parameters (i.e., reactor low diversityin its capability to shut down and cool the level and high drywell pressure), hiany additional reactor core There are four distinct systems for signals are provided,in groups of four or more, to controlling reactivity and four distinct systems for initiate RPS scram (see Table 7.2-2).

cooling the core.

With its inherent advantages, it is also recognized REACTOR SHUT-DOWN SYSTEMS that such design integration (i.e., shared sensors) theoretically escalates the effects of potential (1) The RPS " failsafe" (i.e., scram on loss of l

common. mode failures (ChiF). Therefore, SSLC power or data communcations) hydraulic system architecture is designed to provide maximum scram (Section 7.2.1.1.4).

separation of system functions by using separate digital trip modules (DThis) and trip logie units (2) The ATWS-mitigating de-power-actuated air (TLUs) for RPS/htSIV logic processing and for header dump valves (alternate rod insertion LDS/ECCS logic processing within each of the four

[ARI]) scram (7.2.1.1.4.5),

essential power divisions. Thus, setpoint comparisons within individual DThis are associated (3) The ATWS-mitigating rod run in function with logically separate initiation tasks.

utilizing fine-motion control rod drive (7.7.1.2.2).

Sensor signals are sent to each DThi on separate or redundant data links such that distribution of (4) The standby liquid control system (7.4.1.2).

DTM functions results in minimum interdependence between echelons of defense. For reactor level REACTOR CORE COOI ING SYSTEMS sensing, the RPS scram function utilizes narrow-range transmitters while the ECCS functions (1) The feedwater control system (7.7.1.4).

utilize the wide-range transmitters. The diverse high drywell signals are shared within the 2-out of-4 (2) The high pressure core flooder system voting logic. In addition, all automatic protective (73.1.1.1.1).

functions are backed up by manual controls. These concepts are illustrated in Figure 7A-2-1.

(3) The turbine-driven reactor core isolation cooling system (73.1.1.13).

As a general rule, shared sensors for protection systems are not used for control systems (i.e.,

(4) The low pressure flooder mode of RHR feedwater, recire, etc.). However, the end of-cycle (73.1.1.4).

(EOC) recire pump trip signals originate from the same turbine stop valve closure or turbine control The remote shutdown system (RSS) also provides valve fast closure sensors which contribute to scram.

an independent means of actuating core cooling These are Class IE sensors, but they are not shared functions diverse from the plant main control room.

with other protection systems and the interface with the recirc system is naturally isolated via fiber-optic In summary, the ABWR design has incorporated

cable, defense-in-depth principles through maintaining separation of control and protection functions even Another use for some of the protection shared though sensors are shared within protection systems, signals involves the ATWS trip which activates the In addition, the shared sensors are designed within a fine motion control rod drive (FMCRD) run-in and full four division architecture with 2-out-of-4 voting alternate rod insertion (ARI) as diverse back up to logic.

O Amendment 22 7A.7 2

a ABWR usamar-Standard Plant n~ n

[

')

Diversity principles are incorporated at both the associated with the protection systems are Class IE v

signal and system levels: (1) Diverse parameters are -

and are qualified to the same standards as the.

monitored to automatically initiate protective actions protection systems, which are also manually controllablet (2) hiultiple diverse systems are available to both shut down the Independence of the four SSLC electrical divisions reactor and to cool its core, is retained by using fiber optic cable for cross divisional communication such as the Therefore, the ABWR fully meets the intent of 2-out of 4 voting :ogic. Capability for test and NUREG-0493.

calibration is greatly enhanced by the SSLC's self test subsystem (STS) as described in Subsection 7.1.2.1.6.

Items 7A.5(5) and 7A.6(5):

In summary, the hardware and software functions NUREG-0i91 has been reviewed and determined of the microprocessors used in the SSLC comply to be a precursor to NUREG-0493 for which GE has with applicable portions ofIEEE 603 and Regulatory z

stated full compliance as detailed above. Therefore, Guide 1.153 (i.e., quality, qualification, testability, the ABWR design is also consistent with the intent iidependence). The remaining portions, which apply of NUREG-N91.

to the nuclear safety systems, are not compromised (yet enhanced by self test) by the SSLC design.

Items 6(1) and 6(3):

IEEE 603 has been reviewed, as has Regulatory Guide 1.153 which endorses IEEE 603.

The microprocessor hardware and software which n) make up the safety system logic and control (SSLC)

_('

is designed to make logie decisions which automatically initiate safety actions based on input from instrument monitored parameters for several nuclear safety systems. As shown in Figure 7A.7-1, the SSLC is not a nuclear safety system ofitself, but is a means by which the nuclear safety systems accomplish their functions. In that sense, the SSLC is a component that integrates the nuclear safety

systems, hiost positions stated in IEEE 603 (as endorsed by RG 1.153) pertain to the nuclear safety systems, and are similar to those of IEEE 279, which are addressed for each system in the analysis sections of Chapter 7. Safety system design bases are described for all C&l systems in Section 7.1, beginning at Subsection 7.1.2.2. Setpoints and margin may be found in Chapter 16.

The safety system criteria in Section 5 of IEEE 603 is not compromised by the introduction of the SSLC.

All positions regarding single-failure, completion of protective actions, etc., are designed into the protection systems. All SSLC components x,-

Amendment 22 7A.7-3

ABWR 2aciman Standard Plant nev n Notes for Tables 9A,61,2,3,4,5 & 6

/^$

FIRE IIAZARD ANALYSIS

()

Equipment Database m.

Column IIcadings:

Item No.

Where listed, this is a serialized number which was added to the database to expedite tracking of the individual records.

MPL Number This is the master parts list number for the device. if the MPL number for a device was not known, a number was made up and an asterisk placed at the right hand end of the number. This facilitates the tracking of desices.

Elect. Division The nuinber or alpha character shown in this column is the electrical divisional assignment utilized for the analysis, 1

Division i 2

Division 11 3

DiW.on 111 4

Dhision IV N

Non-divisional (Not Class 1E)

O

?

Insufficient information to determine dhisional assignment v

Multiple numbers per item indicate that there are multiple divisions associated with the item.

Elev. Location Indicates the assigned elevation in millimeters for the location of the device. A "99999" in the column indicates that the elevation is unknown.

Location Number Coord.

The building column coordinate for the location of the device in the reactor building.

Location Alpha Coord.

The building row coordinate for the location of the device in the reactor building.

Description Short description of the device.

System Drawing The drawing number which identifies the device as being part of the plant design. Usually the drawing number is for the system P&lD A question mark in this column indicates a drawing was not available,

,A YL Amendment 22 9A.6 2

ABWR

. main Slamlard Plant ng Notes for Tables 9A 6-1 and 2 (Continued)

FIRE IIA 7ARD ANAINSIS' 4

Equipment Database.

.W

- Div. Assign. Verification Drawing No.

This is the number for the drawing used to verify the correct disisional assignment for the device. 'A

~

question mark indicates that a drawing was not available.

Status of Device Location The information in this column indicates the validity of the device location information. The entries have the following meanings.

K The location of the drawing was given on the drawing listed in the ' Device Locat. ion Drawing" column.

V Device location could not be found on any drawing.

?V Device may be shown on a drawing but it is not labeled so it can be identified.

F Location of device not known Location will be determined as the detailed design progresses.

Device Location Drawing Drawing number from which indicated location of the device was determined. O ;stion mark indicates no drawing was available.

Mech. or Ins / Elect. Penet.

O Some of the devices on the list are electrical, instrumentation or mechanical penetrations. The entries have the following meanings:

N/A Device is not a penetration.

M Mechanical penetration.

I/E Instrumentation or Electrical Penetration.

E electricalPenetration.

Room Number This is a room number that was assigned strictly for use in th-FHA to v..iquely identify areas in the plant.

l The numbers are also shown on the FHA Equipment Drawings, I k.e 9A.4-1 through 9A.4-8.

Cables The power and instrumentation cables for each piece of equipment are considered to be part of the equipment listed.

i 1

O l.

Amendment 25 9A.G3

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Standard Plant 4 Rev A--

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SECTION 131

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--TABLE OF CONTENTS a

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Title

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' 13.3 EMERGENCY PLANNING '-

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13 3.1' COL License Information 133 1,

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13 3.1.1 Emergency Plans

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Amendment 25 1-

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o Standard Plant am:A 113.3 EMERGENCY PLANNING;

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~ :) Emergency planning is_ hot within the scope of:-

the ABWR design, Ilowever, there.are design:

1 features, facilities; functions, and equipment:

l necessary for emergency planning that must be--

iconsidered in the design bases of a standard plant.- -

- Table 13.31 is a' summary of the ABWR' design 4

considerations pertaining to emergency planning.--

The COL applicant will provide emergency plans in ;

accoradnce with 10CFR5033(g) and 52.79(d). See Subsection 133.1.1 for COL license information.'

s 13.3.1 COL License Information ?

133.1.1 Emeryney Plans -

{

The COL applicant will provide emergency p'mc in accordance with 10 CFR 5033(g) and'-

52.79hl). (See Section 133) -

i s

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Amendment 25 -

~ 1331-l J

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LABWR avimm Standard Plant Rd A --

Table 13.31 -

p/

(

ABWR DESIGN CONSIDERATIONS FOR EMERGENCY PLANNING REQUIREMENTS primary Document /

Emergency Planning -

~AHWR facility Sts.tj,en Renulrrments

- Deslan Consideration TechnicalSupport NUREG 06%/ The TSC is an onsite The ABWR Standard -

Center (TSC) 1.3.1 facility located close Plant will comply to the control room with all the TSC that shall proside

- design requirements.

plant management and Specifically, a TSC technical support to -

- of sufficient size to the reactor operating support 25 people is '

h personnellocated in -

located in the-the control ruom dur-senice building L ing emergency cond-adjacent to the ions,it shall have

control building.

technical dcta displays The necessary fac-and plant records ilities and equip-available to assist ment called for in in the detailed ana-Section 2 of lysis and diagnosis of NUREG-06%.

abnormal plant condit-

?

ions and any signincant release of radioactivity-to the environment. The TSC shall be the primary communications center for the plant during an emergencyJ A senior ofScial, designated by'-

the licensee, shall use the resources of the TSC -

- to assist the control room operators by handl-~

ing the administrative

' items, technical evaluat-ions, and contact with offsite activities, reliev- -

ing them of these functions.

The TSC facilities may also be used for perforrning normal functions, such as shift technicalsupenisor and plant oiperatio'ns/

ij maintenance analysis.

functions, as well as for emergencies.

Amendment 25 13.3-2~

t q

ABWR uuiei Standard Plant RmA Table 13.31 AllWR DESIGN CONSIDERATIONS FOR EMERGENCY PIANNING REQUIREMENTS (Continued)

Primary Document /

Emergency Planning AllWR Diciliti SEl!QD ReauIrements Desien Consideration OperationalSupport NUREG 06%/

The OSC is an onsite The ABWR Standard Plant Center 13.2 assembly area separate will comply with all the (OSC) from the control room OSC design requirements.

and.he TSC where Specifically, the lunch licensee operations room adjacent to the TSC support personnel in the service building report in an emergency.

which is adjacent to the There is direct commun, control building will be ications between the indentified as the OSC.

OSC and the control The COL applicant is respon-room and between the sible for identifying the OSC and the TSC so communication interfaces for that the personnel the inclusion in the detailed reporting to the OSC design of the control room and can be assigned to TSC. The detailed require-duties in support of ments are prosided in section emergency operations.

3 of NUREG-06%.

Emergency Oper.

NUREG %%/

The EOF is a offsite The EOF is not within ations Facility 133 support facility for the scope of the ABWR (EOF) the management of over Standard Plant. It alllicensee emergency is the responsibility response, coordination of the applicant of radiological and referencing the ABWR embonmental assess-to identify his EOF ments and determination and the communication of re.ommended public interfaces for inclu-protective actions, sion in the detailed The EOF has appropriate design of the TSC and technical data displays control room. The and plant records to detailed requirements assist in the dagnosis are provided in of plant conditions to Section 4 of evaluate the potential NUREG 06%.

or actual release of radioactive materials to the emirnomnet. A senior licensee officialin the EOF organizes and mamages licensee offsite resources to support the TSC and the control room operators.

l assembly area seperate from the Amendment 25 133-3 l

l

l,,

ABWR nu.u a Standard Plant am n.

Table 13.31

(

s AllWR DESIGN CONSIDERATIONS FOR EMERGENCY l'l>NNING HEQUIREMENTS (Continued)-

Primary Document /

1;mensency Planning AllWR fulllu Sttilmi Rtuultrintatanian ConhlderAllita controt room and the TSC, shall be prmided for operations support personnel to report in an emergency. There shall be direct comm. ucations between the OSC and control room and between the OSC and the 'lJC so that the perronnel report. Inh Io the OSC can be assigned to duties in - emergency operations. Emergency Oper. NUREO.0654/ Bach licensee shall Not within the scope. ations Center !!JI.6 makn prevision to of the ABWR Standard (EOC) acquire data from or Plant. Ilowever, no. O for emergency access - inpact on ABWR-to offsite m_oritoring design. equipment including geophysleal phenomena and radiological monitors. Laboratory Facil-NUREG4M54/ Provisions to acquire Resposibility of COL itles, Fixed or -II.ll.6.c data from or for emer. - applicant. ABWR-hiobile gency access to off-design allows site monitoring and applicant to meet . analysis equipment for this requirement. laboratory facilitics, fixed or mobile. Post Accident NUREG-0737/ Post accident sampl. Post accident sampl-Sampling System 11.B3 ing capability ing system of Sub-r.ection 93.2 meets requirements except as descirbed in -- Section 1A.2.7 for - the upper limit of activity in the' samples at the time they are taken. Amendment 23 13.bt ~

-j q

ABWR mamm Silitidttt1Ll'lttill ><~ n Table 13.31 AllWR 1)ESIGN CONF'l)EllATIONS Folt Eh1ERGENCY l'IANNING REQlllREA1ENTS (Continued) Primary Ihwument/ 1:mergency Planning AllWit DKUlli Snllun Rtuuktaltnis Iblen Consitltrat193 Ons'.tc Decontam-10CFI(50 Provislens shall bc D(contamination of ination Facility Appendis 11/ made and described of onsite individuals IV.fL3 for facilitics and will be provided by supplies at the site the COL applicant for decontamination in the senice of onsite indiGE als. building adjacent to the main thange rooms (See Figure 1.2 20). O O Amendment 23 13.3.$

AllWR mwmu Stafnigidl81gnt Riv 11 SECTION 13.6 gO CONTENTS i Sedlun Title l' age 13.6.1 Erelimin.nn.fhuuth!g 13.6 1 13.6.2 &culity.flatl 13.6 1 13.6 3 col,I,lcense Infortnation 13.G2 13.6 3.1 Introduction 13.42 13.6.3.2 Design Ilases 13.42 13.633 Vital A:eas 13.6-2 13.6 3.4 Methods of Access Control 13.6 2 13.6 3.5 Access Control and Security Measures Through Exterior Doors to the Nuclear Island 13.G2 13.6 3.6 Ilullet Resisting Walls and Doors, Security Grills and Screens 13.6 3 1 13.6 3.7 Compatibility with the Remainder of the Plant 13.G3 13.6 3.8 Security Contingency, and Gaurd Trainint; Plans 13.6 3 13.6 3.9 Achievemnet of Operational Status 13.6 3 13.6 3.10 Vital Area Clusification of Central and Secondary Alarm Stations 13.G3 13.6 3.11 Verification that Vital Systems Can Perform Their Safety Function 13.6-3 13.6 3.12 Evaluation of Security Response Forec 13.6-3 13.6 3.13 Compatibility with R.G. 5.12, Positive Control Requirements, and Record Keeping 13.6 3.1 13.Gli Amendment 23 f

ABWR meu Sinndard.Phnt nts.o SECTION 13.6 O, TAllLES Inhle Hile tage 13.6 1 Reactor Building Control Measures 13.G4 j 13.G2 Control Building Control Hensures 13.G5 j ILLUSTRATIONS FIGUHE Elle Eagt 13.6 1 Reactor fluilding. Vital Areas at Moor 13.G6 Elevation (-) 8200mm 13.6 2 Reactor Building, Vital Areas at Moor 13.G7 Elevation (-) 1700mm 13.6 3 Reactor Building, Vital Areas at Moor 13.G8 Elevation 4800mm 13.6 4 Reactor Building, Vital Areas et Moor 13.6 9 Elevation 12,300mm 13.G5 Reactor Building, Vital Areas at Moor 13.6-10 Elevatior,18,100mm E E 13.6-6 Reactor Building, Vital Areas at Moor 13.6-11 Elevation 23,500mm 13.6-7 Reactor Building, Vital Areas at Elevation 13.G12 27,200mm 13.G8 Reactor Building, Vital Araes at Elevation 13.6-13 31,700mm 13.G9 Contr01 Building, Vital Areas at Elevation 13.6 14 l (-) 13,150mm l 13.610 Control Duilding, Vital Areas at Elevation 13.6-15 (-) 7100mm l 13.6 11 Control Building, Vital Areas at Elevation 13.6-16 (-) 1450mm 13.6 12 Control Building, Vital Areas at Elevation l.6-17 2950mm O 13.Glii Amendment 8

AllWR zw.mu S11:ndunlJ'J.unt rox. n i l 13.6.3 COL 1,1 cense information \\J SAFEGUARDS INFORMATION Provided under separate cover. (Ineludes pages 13.6-2 through 13.6-2.2) Amendment 25 1162

ABWR 1 2w=w Sinndnrdi'lani Riv 11 \\ O l SAFEGUAHDS INFORMATION.Provided under separate cover. (Continuation of Sul>section 13.6.3,

• COL License Information. Physical Securit()

O O Amendment 23 1143

i l AllWR me SandaulEhnt FIN H pool to the core, with each asrembly being test procedure preparation will be scheduled O identified by number before being placed in the such that approv 2 ?rocedures are available correct coordinate position. The procedure approsimately 60 days prior to their intended controlling this movement will specify that use or 60 days prior le fuel load for power shutdown margin and suberitical checks be made at ascension test procedures. Although there is predetermined intervals throughout the loading, considerable flexibility available in the thus ensuring safe loading increments. Innessel sequencing of testing within a given phase there neutron monitors provide continuous indication of is also a basic order that will result in the the core flux level as cach assembly is added. A most cificient schedule. During the preopera-complete check is made of the fully loaded core tional phase, testing should be performed as to ascertain that all assemblies are propersy systein turnover from construction allows. Ilow. installed, correctly oriented, and occupying ever, the interdependence of systems should also their designated positions, be considered so that comruon support systems, such as electrical power distribution, service 14.2.10.3 Pre Criticality Testing and instrument air, and the various makeup water and cooling water systems, are tested as early Prior to initial criticality the shutdown as possible. Sequencing of testing during the margin shall be verified for the fully loaded startup phase will depend primarily on specified core. The control rods shall be functional and power and flow conditions and intersystem pre-scram tested with the fuel in place. The post requisites. To the extent practicable, the fuel load flow test of the reactor internals vi. schedule should establish that, prior to ex. Station assessment program, if applicable, shall cceding 25% power, the test requirements will be be conducted at this time as well. Additionally, met for those, plant structures, r,ystems, and a final verification that the require:d technical components that are relied on to prevent, limit, specifica' , surveillances have been performed or mitigate the consequences of postulated acci. shall be made, dents. Additionally, testing shall be sequenc. ed so that the riafety of the plant is never O 14.2.10.4 luitial Criticatify totally dependent on untested systems, compo-nents, or features. Power assension testing Initial criticality shall be achieved in an will be conducted in essentially three phases: orderly, controlled fashior, following specific (1) initial fuel tonding and open vessel detailed procedures in an approved rod withdrawal testing: (2) Testing during nuclear heatup to sequence. Core neutron flux shall be continuous-rated temperature and pressure: and (3) power ly monitored during the approach to criticality operation testing from 5% to 100% rated power. and periodically compared to predictions to allow Further, power operation testing will be divided early detection and evalaation of potential ano-into three sequential testing plateaus as shown malies. on Figure 14 21. The testing plateaus consist l of low power testing at less than 25% power, mid 14.2,11 Test Program Schedule power testing up to about 75% power between approximately the 50% and 75% rad lines, and The schedule, relative to the initial fuel high power testing along the 100% rad line up to load date, for conducting each major phase of the rated power. Thus, there will be a total of initial test program will be provided by the ap-five different testing plateaus designated as plicant referencing the ABWR Standard Plant described on Figure 14.21. Table 14.21 design. This includas the timetabic for ge. indicates in which testing plateaus the various neration, review, and approval of procedures as power ascension tests will be performed, well as the actual testing and analysis of re. Although the cider of testing within a given sults. As a minimum, at least 9 months should be plateau is somewhat flexible, the normal allowed for conducting the preoperational phase recommended sequence of tests would be: (1) core prior to the fuel loading date and at least 3 performance analysis: (2) steady state tests: months should be allowed for conducting the (3) control system tening: (4) system transient sinttup and power ascension testing that commen-tests: and (5) major plant transients (including ces with fuel loading. To allow for NRC review, trips). Also, for a given testing plateau, Amendmerit 23 IL24

ABWR mim Sinndard l'Initi RI'V. Il testing at lower power levels should generally be performed prior to that at higher power levels. The detailed testing schedule will be generated by the applicant referencing the AllWR Standard Plant design and will be made available to the NRC prior to actual implementation. The r,chedule will then be maintained at the job site so that it may be updated and continually optimited to reflect actual progress and subsequent revised projections. 14.2.12 IridividualTest Descriptions 14.2.12.1 Preoperational Test Procedures The following general descriptions relate the objectives of each preoperational test. During the final construction phase, it may be necessary to modify the preoperational test methods as operating and preoperational test procedures are developed. Consequently, methods in the following descriptions are general, not specific. O O Amendment 21 14.241 .1

ABWR 23uiman Standard Plant REV. A'. '( . APPENDIX 18F-TAllLE OF CONTENTS Secilon Title Page 18F.1 - INTRODUCTION 18F-1 l'F-1 18 F.1.1 Important Operator Actions from the PMA ' 8 18F.1.2 Inventory of Controls. Displays,and Alarms ISP 18F.1.2.1 Inventory of Minimum Controls. 18F-1_ 18F.1.2.2 Inventory of Minimum Displays 18F-2 18F.1.2.3 Inventoryof Minimum Alarms 18F-2

i O

TAIlLES Table Title Page-18F-1 inventory Of Controls 11ased Upon the AllWR EIUs And PRA 18F-3 1 81 -2 inventory Of Displays liased Upon the AllWR Elms And PRA- .181L10. I l'3 Inventary of Alarms liased Upon the AllWR EPGs And PRA 18F-13 181L4 Deleted 18F-5 Deleted 181L6 - Deleted O Amendment 25 IBF-11 -l

AllWR moiman Standani Pitmt intv. A O APPENDIX 18F TAllLES (cont'd) TnMe Title Page 18167 Deleted 18F-8 Deleted 18F-9 Deleted 181'10 Deleted 18F-ll Deleted 181L12 Deleted 18IL13.1 Deicted 181' 13.2 Deleted 181Ll3.3 Deleted O Amendment 25 ISI'iii

1 ABWR 244sion4x Standard Plant uy4 l !D) 18F,1 INTRODUCTION (1) 11ackup manualinitiation ofIIPCF, (2) Recovery of feedwater following a scram,

1. "" "

• I"l##I " b""

" E "'*" 'lhis appendit contams the results of an analysis of

• """" #E""#

infonnation and control needs of the main control rootn U """" ##" ' #'.# "" "" ^ operators to identif y a setof tnimmum wntrols, du plays, and alarms. 'the analysis is based upon the operation Ab.'"E#"#7

1. E*""

W gnment anOnWahn oWewatu for m, strategies given in the AllWR Emergency Pmcedure "I# Guidelines (lil4h) as presented hi Appendix 18A and AHgnmmta n d no awa u mdryweH upon the significant operator actions detennined by the Pmbabilistic Risk Assessment (PRA) and given in [g'"[dopf wdwell spray using RilR, g subsecuon 1 SF.1.1. 'I he minimum inventoryof controls' (9) Isoladon of water sources in an internal flooding, displays, and alarms from this analysis is g iven in TaNes (10) initia00n of standby RilR in event of failure of 18F 1,18F-2, and 1XF 3. The infonnation and controls operating RilR during shutdown operadons. Idenu fied m these tables do not necessarily include timse frotn other design requirernents (such as those frorn These actions are already included in the ElYh; thus no Section 18 4.2.11, SPDS). further analyses are required. 'the following g uidelmes, developed from a tescarch 18F.1.2 -Inventory of Controls, Displays, program of advanced centrol panei designs.were used to specify the type of implementation device for controls, and Alarms displays, and alarms: Tables 18F 1,18F-2, and 18F-3 are inventories of l (a) Fixed Position Controls ininimum mntrols, fixed displays, and fixed alanns + hianual starting and resctting of safety systems, necessary based upon the operadonal analyses of each

• Mamtal starling of emergency backups, step of the AllWR EPGs and certain PRA important
• Mode switches for initiation of automation operator actions.

sequences. (b) DivisionalVDUs 18F.1.2.1 Inventory of blinimum C<mtrois

• Individual controls of safety system components,

+ 1.ineup displays of safety systems. Tnble 18F-1 is the inventory of minimum controls I (c) Non-Divisionet VDUs necessary for execution of emergency operating

• Monitmingofnon-safetysystemsandcontrolof pmcedures. In Table 18F 1, fired position mntrols that l

individual controls of non-safety systems, are Class IE control devices are indicated by bold fact: . Individualalarms. t)pe and :apitdized letters. Non-safety related fixed (d) Fited Position Alanns tmsinon ontrol devices are indicated by non-capitalized + Important plant level and system level alarms. .ype letters followed by a star character (*). All wntrol (e) Fixed Position Display devices Iceated on divisional VDUs are pmvided by . Monitormg ofimportant plant parameters, safety-related system controllers. Controls on the VDUs (driven by pmcess computer system) are provided by 18F 1.1 ImportantOperatorActionsfrom Hon-safetyrelatedsystems. Inaddition non-safetyrelated gpg controls and display capabibty as described in subsecnon 18.4.3.2, Non-Safety System, are provided by VDUs that are independent of the pmcess computer system. These The following operator actions are considered to be process-computcr-independent VDUs pror.de a backup imponant operator aedons in the AllWR PRA(refer to convol and display capability to the process-computer-subsections 19D.7): dnven VDU5 not specifically indicated on Tables 18F-1. l h d Anndant M ggp.)

AllWR meim Standard Plant at u O 1N1'.1.2.2 Inventory of hilnimum Displays Displays that are required are surnmarized in Table l 18F 2. In Table 18F.2, fixed displays provided from Class 111inntrutnents are similarly indicated by bold face type and capitalized letters. Those that are provided by non-safety related instruments are indicated by non-capitaliied letters by a star character (*). A display paratucter that is a Regulatory Guide 1.97 parameter is indicated by a double star character (* * ) following an entry in this table. I EF.1.2.3 Inunfory of hiinimum Alarms l The fixed position alanus are summarized in Table 18F-3. Fixed posiuou alama provided from Class 113 instruments are similarly indicated by bold face type and capitalized letiers. 'those that are provided by non-stfety related instruments are indicated by non-capitalitalletters by a star character (*). O O Amendment 25 Igjt2

ABWR mmaa Standard Plant nix. x TAIILE llW-1 INVENTORY OF CONTROLS IIASED UPON Tile AllWR EPGs AND PRA 3 i NO. TIXED l'OSITION No. FIXED l'OSITION 1-hhNt'At hCR AM INITI ATION 5W(4), 20 RCic sykTlal sTAsisht MODr. INI11ATION EW, 2 M4NUA',50 RAM INITIA110N sWilo. 21 Condensate pump Standby Mode I Initiation Switches (3) e, 3 REAt.1uR MOf>E $W 22 Reactor feedpump Standby Mode Initiation Switches (3) 6 4 IsrY. I M AIN x1 LAM LtNE hasrAL 23 Condensate pump Startup Mode I3OIATION sw initiation Switches (3) *. 5 DIV.ll MAIN STEAM LINE MANUAL 24 Reactor feedpump Martup Mode 1%OIATION NW Initiation Switches (3) 6 ti DIV. Ill MAIN 51 TAM IJNr MANt'AL 2$ bif( A) PCMP Cs. ISOIA110N bW 7 DIV. IV MAIN hTEAM IJNE M ANUAL 26 sLc (hi rt'MP ts. isulA110N EW 8 PRIM ARY (UNTAIN'MI:NT 27 Alts (A)INit! BIT AW. DIV.1 MANOA1, ISOLA 110N EW, 9 PRI%l AR Y (DNT AINMEWT 28 ADS (B) INillBIT SW, DIV. ll M ANUAL ! SOLATION AW, 10 PRIMARY CONTAINMENT 29 RilR(A) 6TANI>BY MODE 5W. DIV.Ill MANUAL L%OtATION SW. 1I RCIC INITIA110N sW 30 klikiB STANDBY Mol:E sW, .12 IIPCF (in IN111A110N SW ll RilR(O STANDBY MODE SW. 13 IIPor to lNTT1ATION sw 32 M AIN STEAM isotATION VALVE Ou N COOL 2No 60 milutni st traiAslON POOL MODF. INlT14110N SW, (XNil. LNG MODE INiTIA110N $W, 4] RilR EiMit"l1Hm N (Y3OIJNG 61 kIIRM kl'PPRIANION POOL MODE INITIA110N NH, (XxlLLNG MODE INIT14110N EW, 42 Rilke stit'llwm N Cf M)lJNG 62 Rilkipp DR YWI LL EPRAY Motit MODE INITIA110N SW, IN!11ATION SW, 43 ARl( A) Manual Initiation Sw *. ') 3 RIIR(O liuYwtLL krRAY MODE INii14 TION 8%, 44 ARI(II) ManualInitiation Sw. 64 scis(A)IN111ATION sw. 45 Recirculation Runback 65 M;1rus> tNIT14110N sW. Initiation SW( A) e, 16 Recirculation Runback 66 kilR< s> Wi~ty tti, $PR AY Initiation SW(II) e, INrn4110N MODE sw. 47 RIP Starl/Stop CS (10) *, 67 plin'o WE1M Elt, $ PRAY INITIATION MODE SW, 48 ARl(A) logic Reset SW *, 68 tilV 1 M ant!AL Alls CllANNEL i INII1ATION EW, 49 ARl(II) logic Reset SW *, 69 DIY I M ant!AL A!W CitANNEL 3 tNrn4T10N sW, 50 OkD CalraclNo W A1TR PRtAst'at 70 DIV 11 M4NOAL ADN Cll4NNEL 1 IDW SCRAM BYPASS bM( Al, INIT14 TION SW, 5l CRD Cll4RGING % A1TR PRI $51'R E 71 DIV II AD% M4NOAL ADS CllANNEL 2 tow bCHAM BYPASS SWtpl. ENITI ATION EW, 52 Cup CliARGING WATER PRES 5t'kE 72 RCIC DIV. I ISOLATION IIM3tC LA)W hCR AM BirAh5 SW(0. ElSET 5%, 53 Cup ClltantNo w ATEu ralast'kE 7; Rc1C Div. II ISOLA 110N !.OGIC LI'W hCH AM BYPASS h%(D), RDET SW, 54 MANUAL SCRAM RDET SW, 55 RPs Div.I Tulr nEstr sw, 56 RPs Div,11 TRIP RLsET SW, 57 RPS IIIY, til TRIP Rt3ET SW, 58 ars l>IV. IV Trit-

• T SW, O

-5

ABWR 2WlWAk Standard Plant aiy., .a O TAIILE INF-1 (cosit'd) INVENTORY OF CONTROL,S IIASED UPON TIIE AllWR EPGs AND PRA l i NO, I'lXED 14)SITION CONTROL S OUTSIDE 01' Tif E M AIN CONTROI ROOM l 1 kilR(C) rnanual valves for firewster injection (F101,102,103), 2 RWCLI regenerative heat exchangermanual bypass valve 3 Turbine building !!VAC system controls O O - Amendment 25 isF-5 --

AllWR 2 mimin stanciant Plant unu TAILLE 18F-1 (cont'd) INVENTORY OF CONTROLS IIASED UPON Tile AllWR EPGs AND Pita NO. I'lXED CONTitol,S 1.OCATED AT M AIN CONTi(Oi, ItOOM Al(EA l'ANEl,S 1 17 re protection system motor pump control SW *, 2 1 ire protection system diesel pump control SW *, 3 "A" bCR4M h0LI NO!!) MAIN POM l.k hki'Akt R (3 4

• B" %CR AM h0LENOID MAIN POM l~R bkl.Akt R (3 3

RPhl>IV.I 1RIPINillbli$% RP5 IV.11 TRIP INIlflitT SW 6 y RPfl It!V. til T*IP !NIllBli EW g RPs t>IV. IV 1 RIP LNit! BIT SW ROD M ORTil hflN! Mill:R BYP A%5 9 yg CAkt* A) OPI RATING MOlit: $W, gg CAM %B) Off RATING MODE SW, gg CAMM A) 5AMPLI: bl:LI:t7 5W, 33 CAMkBP $4MPl.E El:1.E.cT $W. g4 tCSA)COvitOt6W. 33 FCh t) n>NTROL E%, O Amendnent 25 18F--6 l. 1

H ABWR - 234t.ilxlia Standard Plant gn. 3 l l 1 TAllLE 18F-1 (Cont'd) 1 INVENTORY OF CONTROLS IIASED UPON TIIE AllWR EPGs AND PRA I NO. DIV. VDU N O. DIV.VDU I flP(T(B) $Y$18.M CON 1Hul A, 19. EllR BI Pl'MP RIK48 0 Kil I R I AN CON 1ROL, 2 IIPCl(C) EY$11AI MkYlkul A 20 kilRO PI'Mr k(K'M C(M8tJ R I AN M)NT HOL, 3 R(1C $YFilAl CON 1*OlA, gj IIP (T(Be PtfMP ROOM C(xtij R $ AN CON 1ROL, i 4 kllR( A) NY811 Al MIN 1ROI A, 22 IIP (TIC) PCMP kOOM COOLI N I AN

CONTROL, 5

kIIR(s) synn.M M W1*OLA, 23 RCIC Pl'MP kOOM "M'll k i AN M W"IR00, 6 kilR(C) EYST1.M LTINTuota. 24 F(N Al kOOM COOI.l:k l AN (Y)N1*01.. 7 M41N sTI AM I.lNr. lik4tN INauskin 23 lom e> R(xal M)Ot.ER FAN CONI rot. AND OlJillO4RD lbOIATION VAIM

CONTROlA, il NRY CON 1*OL $WITritra (141 26 I?OAI ROOM (YX)lJ R I AN (UN11 tut.

9 $G15tA)$YSTEM CONTROLA, 27 ITotp> ROOM Mxil.l u FAN CON 1ROI, 10 M11NelsystTM CONikots* 2g ITEI. POOL COOtJNa sYsTIAf k M)NTROIA, 11 RhllVAC isotA110N val.YE CONTROLS, 29 Rt1C RTEAM lJNE isOtAllt;iN IAx;tC NYPAssl3 ( AkL A li'MPl.kAll'Rl: tilull, RPV PREMt'RI' I.OW,51 TAM lJNE PR E%CR E l OW, R UIC Tt'RhlNI. l'KilAl'51 PRIMt'RE RAl%I, )2 Al%IO% Pill kl0 CONTROL. NV81138 30 mw CtiIsOIA110N LOGIC BYrAss IM11ATION % Al.VF CON 1 RolA, (SL A' INI114 TION, REGENI R ATIV E IIEAT l.XCilANGl N ARIA 17MPl'R ATT'RE lilGil RPV WATElt 1J:V112), {3 REACTUR IWILDINO IIV Ac th01A110N 3) MSIV AND MAIN NTE AM IJNE l#R AIN VAL) E CONTROLA, 1401AT10N LOGIC BYPA$$ (l l:VI.L 1.11, ht%L 111011 R 41 14 TION, MM. IIIGil FI A)W, M%I, TUNNLt. AREA TEMPIRAll'RE lilull,Mst,it'RRINI: ARIA. ITMPER ATt'RE llIGil, 14Mi30 BYPAM IRPV l 8:VFlJ) OF j4 M3TR A) kt Knl COOLER FAN (Ylvl*01., - 32 kBilVAC 6Y$1LM 1501.A110N val.VIA ' "3'CMA$,'g"E'$3dd$'.h Mi( 15 MITS BP RO(^t MSOLER FAN CON 1ROL. 33 R 1%OLA110N val.V).N. 16 CAM % A) kOOM COOLER FAN CONTROL, 34' LIMilC BYPAM Of filOll DRYMILl'ON IOR PRIMI kt AND RPV WATi k 11VIL 31sOLATI klACTUR bt'lLDING llVAC, 17 CAht% B) ROOM COO 1XR I AN CONTROL, 33 tilGil RPV WA tTR I.EVEL (LEVEL S) llPCI' (N,llT110N VALVE CLAksORE lax;ICBYrAA% ( RllRi A) Pl'MP ROOM MiOtJ R 5 1g I AN CONTROL, Amendment 25 18F-7 e .w .U- .m-. ,-,w'- 4 .4 m.. ~ .,m,--..,.- -#--,w.-, m r-n ,=. r

• w t

---r t v

AllWR 2wma Standard Plant w,a TA tile ltiF-1 (cont'd) INVENTOltY OF CONTROLS IIASED UPON TIIE AllWit EPGs AND Pita NO. Yl)U 1 Condensate and feedwater system e coritrols (Dedicattsi system VDU), 2 CRD system tuntrols, 3 Condenst'e makeup water system

controls, 4

$PClf system controls, Feedwater control system controls (I)edicated system VDU) 5 Pressure control system controls (Dedicated system VDU), 6 Main steam system controls, 7 RWCU system controls. 8 Rod control and information system

controls, 9

Drywell Cooling syswm controls, 10 Nitrogen vent and purge snode of ACS, 11 Containtnent purge mode of containment supply and purge subsystem of RlillVAC, I2 RillIVAC system controls, 13 Atmospheric Contml system controls, 14 Main steam /feedwater tunnel llVAC system controls, 15 RPV head vent valve controls, 16 logic bypasses for Alternate Rod insertion (AR!l(high RPV pressure. RPV Water level 7). !7 logic bypass of liigh RPV Water level (level 8) trip of reactor feedpumps, 18 I*gic bypasses of drywell cooling fans 39d associated cooling water (RLW)(Drywell Pressure ilip.h, RPV Water level 1] O Amendarat 25 18F-8 f

ABWIt 234< icurit Standard Plant EV. A (. i TAlli,E 18F-2 j INVENTORY OF DISI't,AYS IIASED UI'ON Tile AllWR EPGs AND l'RA j NO, I'lXICD l'OSII~lON N O, I'lXI:D l'OSITION 1 kP) WAllR Ittil e s, 24 IWCT(IO INJI CT10N V AL% 1. %1 A11$. 2 kClO TI' kit!Ni' &Pfl D, 23 IIPCl101NJI C110N VALYL: 81 ATlW. 3 WI THil L Pkl ML'kL * *. 26 Rif f4AlllA3W * *, 4 XITPkiMION POOL Bt'LK AVI R AGE 27 kill 44) INJL CI10N V ALVI. $TATT8 l TEMPLk A1Tkt e s, 1 i llP(T(lollOW te, 28 kilRIBillk% 4 4, 6 IIP (Tir+,linW ee, 29 NIlRiB) INJEf.710N V Al.YE 51 A105, 7 RPV PNI Mt'Rt *e, 30 kalk(Cll1AlW * *, 8 lin)MI:I1. PEtsM'RL

• e, 3l kilHIO tNJLCi1ON VALVE lif A108 9

REA(TOR POWI N LLVI:1., 32 T.MI:RGlWCY DItal L 6ENERATOR ( A) (Ni(TkON ILL"E. APkM) 44 OPl.R ATING ST ATI'8 * *, 10 ktAt ION roMI.R I.l:Vtt isRNw

• e, 33 1:Mrkal3Cy DIlsti of;NIR Afon (B, OPER411NU 6~tA115 *e, 11 kl.4CION 111tRM AL POWI R
• e, 34 l AILNUl;NCY DIFELL GLNI M A1DR (O OPERATING 5 Tait $ e e, 12 MsIV POST 110N STA11'N (INB(ARD AND 33 pal %I ARY CONT AINMI:NT OLTilO4kp V AIA l.5) t e W ATER ll:Vl:L t e, 13 k LAUTOR MOpt hw r1CII 36 RPV Water level 3 indwation e, ki'N Mol#E INIHC4110N, 14 hiain turtine stop valves ttatus -

37 RPV water level R indication *, (inain turbine tnp statual e, 15 Main tuttiine control valves status 38 CONDLN541t kTORAct: TANK (main generator Ulp status) *, WATlHIlMt ee, 16 M AIN ETI.AM IJNE RADl4110N * *, 39 liLC Pt'MiiAI DISCII ANGE PktEM'kE e t j7 SCR AM sol F.NOID IJGillwsi ETATts. 4o gLC rttMP(lo ptsCllARot rkist itt,e 1R MAN!!AL hCR AM $WtA) INDICATING 43 Al6 ( A)INit1Bli NW IN AtTO I OGIC OtT IJGilT NTAll'h, OF ELRVICS/INillBIT POSITION, 19 M ant'Al. hCRAM EWtul 42 Alls (BilNil!Bli SW IN AITO I.OGlC utT - INDICA 11NU llGiff SI'A1TM, OF $tRVICS/1NillBIT PostTION, 20 klACION MODE 5WirrilIN ElitTioWN 43 TOP OF ACi1VE Ft'tL WATER LFVFL POSl110N. INDICATION. 21 kPV isOI.ATION sTATt% DISPLAY e

• 44 Main condenwr pressure *.

22 kaC FIUM * *. 45 SRV POs!110Ns **, 73 kCIC INJi Cf10N VALVE STATTS. 46 St'PPR13% ION POOL LEVIL

• e,

\\ b Amendment 25 338 9-

AllWR n^cim^n Standard Plant kiY. A TAllLE 18F-2 (cont'd) INVENTOltY OF DISPLAYS IIASED UPON Tile AllWit EPGs AND Pita NO. I'lXI D l'OSITION 47 hl4tN $11.AAf 1.INI II.OW * *, 4g $RNhl RLAl* ION Pl klOD, 49 SEC ftORO*i T ANK M AU.R (IVI L * *. 50 Main generator breaker status e, $1 Recirculation pump speeds *. 52 AVI It AGl: l*1M Ell-11'MI'ERAM'RI: * *. 33 WE1MTil. IIVDR(01.N (UNCLNTRA110N l.EV1 L

• 6 34 imYM LLL II) DROGEN (UNCINT R A110N 11 VEL * *.

33 l*YMilJ. OXYGEN tuscLNut Anon * *

• 96 WF1M I:1.L OXYGEN (UNCLN11tA110N *
• 57 I m A) DrERAttNa sTAits, 58 Whl0 0rEkAUNU 81 AH5 59 Main Stack radiation level **,

60 Ttme 6 61 lA3Wl:R DRYwtLL WAftR LEVEL * *. 62 DRYM F.LL RADIATION IJVEL *

• 63 WETM Ell, it ADIA110N LEVI;L a t O

Arnendment 25 18F-10

AHWR-mm^n Standard Plant RTV.A t j TAHLE INF-2 (cont'd) INVENTORY OF DISPLAYS IIASED UPON Tile AHWR EPGs AND PRA No. DIV. VDU l R F A CTOR ITIllitNG 181t Ff;kiW11 Al, PRIA%t'kE, 2 Rt:AtTok alttt>tNO IIVAc fattat'ri rat:1A110N LEVI'L *

• 3 HTLllAN!)ljNG ANF.A VlW111A110.%

13tL4t'51 RAltl A110N t.EVFL *

• 1 4

kilR(A) Pt'MP NOOM OMMAR Orf,RA110N STAit% klIN(M) PCMF ROOM MK) TAR OPLRA110N STATi% - l 6 RiiR(c)ITMP R(x)M nlotAR l OPl ILA110N STATT8, ) 7 IIP (T(B) Pt'Mr ROOM (4KMJ.R OPFRA110N STATt% 8 IIPCT(C) Pt'MP R(MIM GM)Lt.R Orl R A110N STATL'a, 1 9 kcle PUMP N00M nK)LLR Ort RA110N ETATOS, i 10 s t3(A) Room nolyR oPrRAilos

ATATTS, II FCb(B) NOOM (XX)t1R OPER4110N r

STAit's, 12 iPC(A) ROOM MX)lJ R OPERATION STATt's, I3 FPC(B) k(M tM (YK)lJ.R OPLRA110N $TATtJS, I4 MIDI (A) kuoM MiOIAR OFFRA110N - $ TAT 05. - 15 suis (n) kOOM uMilAR OPERA 110N 81 A1TS, . l I0 CAMS (A) 400M OxtliR OPS Rfl10N &TATt!S, l7 CAM 1(B) R00M QX) TAR OPLR4'110N 5TATUS. I ) -i A.urodneat 23 18F-ll r, h l ' .I -1 . ~, u,,, __.a_,..--_z.._... a.--~- l

AllWR 2w=a Standard Plant ni u TAlllE INF-3 INVENTOltY OF Al,AltMS IIASED UI'ON Tile AllWit El*Gs AND l'ItA NO. I'lX1:D l'OSITION N O. I'lX1 D l'OSITION 1 ludicated 1(l'Y Waler level 23 Al'* Al lDUIC INIllAl d l'. Abnortnal e. 2 NPV W AT1 N t.h I L 3, 24 A t>% bi IDGIC INiil A16.II, 3 H P) PRIMt NI; ttlGil. 2$ RPV % A1FN I M il. 510P OF AUl1VL 11t 1. 4 liNYWI LL PklMt'ki: filGil, 26 $kV OPLN, Ni l'1kON ll.t'X lilGil.IllGil 2 *t M AIN BTIAM LINE B1DW 111G11, 6 APRM i>OWN% cal.E. 28 IIPIN( A) EYSTI:M 11kot'Bl.L AIARMi, 7 M ITRON MON 110ktNG EVkitM 29 11 PIN # Bi kVNTLM (1 NOl'btl. AL ARM A (1 kot'Bl.E ALARM), 8 MslY CIAmt'kt:, 30 LEAK lirl e cl10N tsOtA110N, 9 Culi Cit AmGING % ATl R Paint Rt 3I RWCU System (Trouble Alarm) *, t 0%, 10 RAPID CORE 61DW t)LCklACL, 32 kt:ACIOR Fl RIOli kilORT, 11 h{ain Turbine Trip e, 33 AlH lilV 1 l%IIIBIlllWAt'TU DIT Uf bl kVICE 12 Main Generator Trip *, 34 Al'8 l'lV. Il lNiilBlitiv4UTo OLT or hiRVICE 13 M AIN STE AM (JNL R Al>lA110N illGil. 3$ M'PrRIwON POOL Bl't.K AVI RAGE l tMPI KAll'kL 1I1011, I4 HPV t i VLL 3 INOIAT10N 36 lik t % I:LL AVik AGE TEMPI RATt'RF. IN COM PL's:1 L,

IIIGil, 1$

kPV l.I% iI,2 IsotA110N 37 M'PPkl.AslON M)OL n AltR IM1)M P1 I:TI., El VI L 11101110%, !6 NPV t FVI1. l.MikVMiLt. PRI wt'ar 38 CAhts liv 02 lk:VI:t tilGil, tilGilisotA110% LNCOMPLL1 E, 17 RPV M ATFk Il.VI:L 2, 39 CAht% A) SYST).M ABNORM AL, 18 k PV M ATl R t,I VI:1.1.3, 40 CAMSBiSYhTEM ABNORMAL, 19 kPV WATTRt.n ILI, 41 Reactor llullding AP I,ow, 20 Control Rod Not inserted 42 Area Temperature liigh *, To/lleyond blSilWP *, 21 RPV Water f.evel 8 *, 43 Area IIVAC AT liigh *, 22 1: ire Pmtection System 44 k BIIV AC EXil AttST R Attl AT10N iilGli (tmuble alarm) e. O l l Amendment 25 181L12 ~

AlnVR 2mimia Standard Plant ui:v,4 O TAllLE INF-3 (cont'd) INVENTORY OF AI, ARMS IIASED UI'ON Tile AllWR El'Gs AND l'RA NO. I'lXED l'OSITION N O. DIV.VDU d$ Reactor fluilding Area R3diation 1 k P8 l'IV. I 1 KIP l%IllBlitil. liigh e. 46 Reactor llullding 1loor thain 2 ers t>IV. tl 1*Ir istilart e.fi. Surnp Water level liigh-liigh *. j 47 killlVAC(System'Irouble Alarm) *. 3 kP8 l'IV til 1 KIP INIIIB311:t'. 48 Stack Radioactivity liigh *, 4 RPs t>ty,IV 1*tr INiusritts, 49 RCW radioactivity high

• 9 Is

((QAT,1 WATER LEVr.L. 50 Radwnste effluent radioactivity high *. 6 DRyws:1.L RAlitaticN t'PNCAlt. $1 Turbine lluilding Ventilation 7 Wl:TMI:LL RAl'IA110N t'rscAtt', System OllVS) (system bouble alarm) *. 52 Radiation Monitor liigh *, 8 [jm[a inurws LL wats:n t. vs.L $3 kCIC Sy5Tl;M (TRO(1BLF. Al.Akhfl. 54 firCrtcisysitM tlkOl'Bil ALAkhlh tircr <n sysitu (1 RtWBli At. ARM), O

Amendarut 25 18F-13

ABWR man bl.11Rd11EdMllli Etyd provided at all vehicle and personnel access This requirement is necessary to meet the intent /* portals. This shallinclude but not be limited to of 10CFR73.55(d). \\ the following: (a) Security force siewing of numbered picture (a) Means for positive identincation ofindhid-badges is an acceptable method of verifying uals required and allowed access into the access authorization of individuals prior to protected area shall be provided. permitting their entry to the protected area. This method may be supplemented, at an (b) Means to search individuals requiring indhidual utility's discretion, with electronic access into protected areas shall be pro-access control technology. vided. This shallinclude detection and alattns for Grearms, explosives, and incen-(b) To prevent the unauthorlied entry of these diary devices that may be concealed on the devices into the protected area. It is antici. individual or be included in a package or pated that advances in metal and explosive material that is to be inside the protected detectors will enable the designer to detect area. This shall be accomplished by the the presence of all explosive devices includ-use of such detection devices as metal ing non nitrate based explosives, detectors, explosive detectors, and x ray machines. (c) This requirement is designed to assure that the last point of access controlinto the pro. (c) Means shall be provided to controllast tected area will not be compromised by an access into the protected aren from within armed intruder. a bullet resistant structure that requires a permit / concur actuation befort; access into (d) This can be accomplished by using CCTV, the protected area is allowed. card readers, and/or hard key and lock con-trols as permit / concur devices. (d) Means shall be provided for positive iden-tification of individuals requiring and (c) Alarms to indicate unauthorized access at. \\ allowed access into the vital areas (i.e., tempts will permit expeditious security control room, reauor building, battery force response,if appropriate. Logging of rooms, etc.), ingress and egress will allow surveillance and review of access patterns ofindividuals, (e) In all cases, the access portal shall be and evaluation of appropriateness of access alarmed to CnS and SAS if improper authorizatian. acccss is attempted. Access shall be com-puter logged (i.e., time of day, employee, (f) This requirement will minimize the number date, area, etc.) for each egress (exit) and of locl s required to control security, health ingress (entry) and controlled to prevent physics and fire protection common doors. or allow access of specified individuals llowever, caution must be used to assure all

only, safeguards information is properly controlled in accordance with 10CFR73.21.

(f) Vital area ingress and egress shall be inter. f aced with health physics and fire (19) Communications; Chapter 9, Section 5.2.11.1, protection access control requirements to Rev.0 assure that only one door control mechar, ism (i.e., lock, readers, status) is The security communications subsystem shall required. Security shall maintain positive meet the following requirements: access control over all multi discipline vital access areas. (a) Each onsite security officer, watchman, or armed response indhidual shall be provided Engineering Rationale b'~~N Amendment 22 191L215

ABWR mwm Standard Plant wa with continuous communications with an by NRC and the overall system power demands, individualin each continuously manned alarm station (i.e., CAS, SAS, PAP). This may be accomplished by using multi-frequency radio or microwave transmitted two-way voice communications. (b) Communications shall be provided between the locallaw enforcement The poveer subsystem shall be alarmed to the authorities and the nuclear plant CAS and CAS and SAS to assure the continuously SAS. manned security station operators are aware or any power system failure or attempt at tamper. (c) Communications shall be provided ing. between the main plant control room and nuclear plant CAS and SAS (i.e., dedicated Engineering Rationale telephone service that does not have an> terminations outside the protected area To assure that the security system will have a re-boundary, radio, et c.). liable source of power io maintain alarm, con-trol, and response functions necessary to prevent (d) Communication system failure or tamper undetected intrusion into vital or protected attempts shall alarm to the CAS and SAS, areas, it is recognized that PA boundary light. ing is such a large load that it would require a Engineering Rationale dedicated independent power source to main-tain it. Therefore, this lighting load is exempted The communications requirements are from UPS power requirements, provided provided to meet the intent of 10CFR73.555(c) adequate compensation measures are taken. and (f). (21) Data Management; Chapter 9, Sections 5.2.13.1 (a) To allow the CAS, SAS, or PAP operator and 5.2.13.2, Rev. 0 to have knowledge of all security personnel locations and capabilities. The overall station security system shall utilize " host" online redundant central processing units (b) Provides CAS or SAS operator with means (CPUs). These CPUs should interface with to direct the response team and call for remote processmg units (i.e., CPUs, micropro-offsite assistance if required. cessors, minicomputers) that have on board memory to allow them to stand alone, for a de-(c) Informs plant control room operator with fined period, should communications with the assessment capability to mitigate sabotage host CPUs be interrupted. initiated 1 OCA attempt. These ' host

• CPUs shall be required to have all (d) Alarms indicating security communications data transmissions supervised and alarmed if a failure or tampering will expedite restora-failure or tamper attempt occurs, tion of senice and provide early indication of a potential sabotage threat to other All Security CPUs shall be dedicated to security plant systems.

functions only and shall not perform any other processing functions (e.g., used to incorporate (20) Power Source; Chapter 9, Section 5.2.12.1, the fire detection or protection system monitor. Rov. O ing or control). The security power subsystem shall be a Engineering Rationale non-interruptible power source capable of rnecting the minimum requirements imposed This requirement is intended to assure that data is processed quickly, accurately, and on a Amendment 25 19D.2-16

ABWR mmas ElRildnr.dfhml Rm A 1911.3 COL LICENSE INFORMATION 1911.3.8 Interdisciplinary Design Revleus (U) 1911.3.1 Quality Assurance Program COL applicants referencing the ABWR design shall establish an interdisciplinary design review COL applicants referencing the A13WR design group and direct reviews for site specific design and shall have a Quality Assurance Program satisfying construction work as required by Subsection the requ!rements of Subsection 1911.2.1(2) including 19B.2.2$(4). the right to impose additional quality assurance requirements. 1911.3.9 Sabotage Vulnerability During Plant Shutdown 1911.3.2 Presention of Core Damage The sabotage vulnerability analysis required by COL applicants referencing the ABWR design Subsection 19B.2.4(10) has been performed for the shall approve applicable design deviations in ABWR and is contained in Appendix 19C.110 wever, divisional totalindependence and separation both applicants referencing the ABWR design shallinclude mechanically and electrically as required by prmision in the plant start.up procedures to inspect Subsections 1911.2.3(3),19B.2.4(2).19B.2.5(3), and critical safety equipment within the containment for 1911.2.11(2). possible tampering just prior to sealing the containment in preparation for start up. Such 1911.3.3 Protection from External Threats equipment includes the ADS /SRV valves and associated accumulators and their charging lines and COL applicants :cierencing the AllWR design the inboard valves associated with the emergency core shall evaluate listed man.made hazards except cooling systems (l.c.. IIPCF, RilR and RCIC). sabotage on a site unique basis as required by Subsection 19B.2.4(6). 1911.3.10 Impact of Security System on Plant Operation, Testing nnd Maintenance 1911.3.4 Ultimate llent Sink Models in the design of the security system, applicants COL applicants referencing the ABWR design referencing the ABWR design shall include an shallimplement the development of predictive evaluation of its impact on plant operation, testing analytical models as required by Subsection and maintenance. This evaluation shall be conducted 19B.2.9(1) through 19B.2.9(4), as required by Subsections 19Il.2.4(12) and 9.5.13.11. This analysis should include consideration of an 1911.3.5 Ultirnate llent Sink Reliability emergency requiring evacuation of the control room in the control building to the remote shutdown panel COL applicants referencing the ABWR design in the reactor building. shall have an ultimate heat sink design goal for the service water flow as required by Subsection 1911.3.11 Security Plan Compatibility with 19B.2.10(1) tbrough 19B.2.10(13). ALWR Requirements 1911.3.6 Main Trensformer Design The AllWR security plan will comply with the ALWR requirements as defined in 19B.2.4. Future COL applicants referencing the ABWR design amendments of the ALWR Requirements Document shall provide main transformet fi.e protection as must be reviewed for ABWR compliance by the required by Subsection 1911.2.18. applicants referencing the ABWR Standard Plant. 1911.3.7 Plant Siting 1911.3.12 Plant Security Systems Electrical Requirements COL applicants referencing the ABWR design shall approve in writing the listed final design COL applicants will provide non Class IE vital parameters to be used at the plant site as required by (uninterruptible) ac power for the site security Subsection 19B.2.19(3). system. [See Subsection 19B.2.4(20)). Amendment 25 19D.11

ABWR 334amas Slandard Plant um ^ 1911.3.13 Ilolting Degradation or Failure COL applicants shall provide the bolting inforniation detailed in Subsection 1911.2.12(6). 1911.3.14 Outside Sabotage COL applicants shall provide sufficient analyses to ensure that the plant is adequately protected frorn acts of outsider satiotage. O ._a _.

ABWR zwim Etandfinifjnnt lov n 435.37 \\d in the description of the DC power system in section 8.3.2.1 it is stated that the operating voltage range of Class 1E DC loads is 105 to 140 V. It is also stated that maximum equalizing charge voltage for Class 1E batteries is 140 VDC, and the DC system minimum discharge voltage at the end of tne discharge period is 1.75 VDC per cell. For a 125 VDC lead acid battery with 60 cells.1.75 VDC per cell equates to a final discharge voltage of 105 "DC at the battery terminals. This is the same as the stated minimum operating voltage of the Class 1E de loads. There is therefore no allowance for voltage drop from the battery terminals to the terminals of the Class 1E loads at the final voltage value of 1.75 VDC per cell. Please address this discrepancy. Also, provide the results of your DC voltage analysis showing battery terminal voltage and worst case DC load terminal voltage at each step of the Class 1E battery loading profile. See the following question with regard to the battery loading profile. 435.38 Section 8.3.2.1 addresses the DC power systems in general and section 8.3.2.1.3.2 specifically addresses battery capacity. With regard to battery capacity, section 8.3.2.1.3.2 states that battery capacity is sufficient to satisfy a safety load demand profile under the conditions of a LOCA and loss of preferred power, the batteries have sufficient stored energy to operate connected essential loads continuously for at least two hours without recharging. (a) Provide the stated load demand profiles and a breakdown of the loading during this demand. (b) Provide the manufacturer's ampere hour rating of the batteries at the two hour rate and at the eight hour rete, and provide the one minute ampere rating of the barriert,. (c) Address station blackout with regard to battery capacity. If a station blackout coping analysis is being prepared for the ABWR, provide a battery load demand profile for the coping duration. Provide a breakdown of the loading during this demand. 435.39 in $cction 6.3.2.1 it is stated that each 125 VDC battery is provided with a charger and a standby charger shared by two divisions, each of which is capable of recharging its battery from a discharged state to a fully charged state while handling the normal, steady state DC load. (a) Provide the continuous and current limited output ratings of the battery chargers. (b) in accordance with position C.1.b of R.G.1.32, Rev 2 verify that the capacity of the battery charger supply is based on the hted combined demands of the various steady state loads and the charging capacity to restore the battery from the design minimum charge state to the fully charged state, irrerpective of the charge status of the plant during which these demands occur. (c) Verify that the battery charger can operate stably as a battery eliminator (i.e., with the charger remaining connected to supply the loads while the battery is disconnected from the loads). (d) Verify that no reverse DC current can flow into the battery charger output from the battery, during periods of low AC input battery charger voltage or during total loss of low AC input voltage to the charger, tO Arnendment 10 20.2-14.10 l

L ABWR umu StandanLPlant nrva 435.40 l Section 8.3.2.1 and Figure 8.3 8 identify the connection of the non Class IE 250 VDC battery chargers to divisions 1 and 3 of the Class 1E system. Identify the isolation devices used at this interface. Are the Class 1E bret.kcrs shown at the interface, tripped on an accident signal? If not, they should be, or else redur.Jant qualified breakers should be provided. 435.41 Section 8.3.2.1.2 very generally identifies the type of loads fed from the 125 VDC Class 1E power system. Please provide a more specific breakdown of the loads fed from each division of the 125 VDC Class 1E power system. 435.42 l In Section 8.3.2.1.3 it is stated that an emergency eyewash is installed in each battery room. In order to ensure that water cannot be int.dvertently splashed on the batteries the eyewash stations should be located away from the batteries and the eyewash installation and its piping should be seismically qualified. Please verify that this is the case. 435.43 Section 8.3.2.1.3.3 states that battery rooms are ventilated to remove the minor amounts of gas produced during charging of battenes. Verily that, in accordance with position C.1 of R.G.1.128 the ventilation system will limit hydrogen concent. ration to less than two percent by volume at any location within the battery area. Also, in accordance, with position C.6.e of R.G.1.1.28, verify that ventilation air flow sensors are installed in the battery rooms with their associated alarms installed in the control room. 435.44 With regard to the DC power systems, section 8.3.2.2.1 states that all abnormal conditions of important system parameters such as charger failure or low bus voltage are annunciated in the main control roorn and/or locally. Please identify the specific meters and alarms used for monitoring the status of the Ct. ass 1E DC power systems and indicate weather they are located in the main control room and/or locally. As a minimum the following indicstions and alarms should be provided in the control room: Battery current (ammeter-charge / discharge) Battery charger output current (ammeter) DC bus voltage (voltmeter) Battery charger output voltage (vohmeter) Batteg discharge alarm DC bus undervoltage and overvoltage alarm DC bus ground alarm (for ungrouded system) Battery breaker open alarm Battery charger trouble alarm (one alarm for a numixr of abnormal conditions which are usually indicated locally) Because the ABWR is an advanced reactor design, you should consider the use of state-of-the-art art battery and electt: cal system monitoring system to assure immediate notification of battery and electrical system problems and to provide for the monitoring of at least the individual cell parameters c f the batteries and the status of the various electrical system circuits, and ideally Amendment 25 20.2 14.11 i ___________________________________._______________._.._______________.________.___________________m

ABWR ummar Standard Plant Rn n 20.2.15 Chapter 15 Questions g. 420.28 L Section 15.A.2.2 defines

• Safety
• and " Power Generation.* The staff did not locate definitions for "important to safety" and " safety related" which are used in Chapter 7. (15A) 420.96 The safety system auxiliaries (Figure 15A.6-1) should be modified to include any IIVAC required to assure continued operation of the electronics. (15A.6) 420.!!8 Describe when appropriate operator action in seconds is required to prever.t significant radiological impact. (15.2.4.5.1) 420.122 is the instrumentation required for the operator to verify bypass valve performance and rel cf valve operation IE or N.1E? (15.2.2.2.1.4) 420.123 SSAR 15B.4 describes the essential multiplexing system (EhtS) in some detail. SSAR FMure 7A.21 states that the design is not limited to this configuretion. It is understanding that the EhtS design -

is still in a preliminary design stage. Is SSAR 15B.4 still accurate and is the design limited to (] that configuration? (15B4) l V 420.124 The Ph1EA submitted in SSAR 15B.4 is inadequate for a safety evaluation supporting the design certification. The Ph1EA appears to the str.M ', be oversimplified with one line item each for component failures and does not address potential software complications. The staff requests clarification of how this Fh1EA was developed given that the syst,m design has not been finalized. The staff also believes that software failures need to be evaluated. The failure modes investigated should include, as a minimum, stall, runaway, lockup, interruption / restoration, clock and timing faults, counter overflow, missing / corrupt data, and effects of hardware faults on software.(15D4) 430.58 The accident analyzed under this section considers only the airborne radioactivity that may be released due to potential failure of a concentrated waste tank in the radwaste enclosure. The SRP acceptance criteria, however, requires demonstration that the liquid radwaste concentration at the nearest potable water supply in an unrestricted area resulting from transport of the liquid radwaste to the unrestricted area does not exceed the radionuclide concentration limits specified in 10 CFR Part 20, Appendix B Table II, Column 2. Such a demonstration will require information on possible dilution and/or decay during transit which, in turn, will depend upon site specific data such as surfere and ground water hydrology and the parameters governing liquid waste movement through the soil. Additionally, special design features (e.g., steel liners or walls in the radwaste enclosure) may be provided as part of the liquid radwaste treatment systems at certain sites. The staff will, therefore, review the site specific characteristics rnentioned above individually for each plant referencing the ABWR and confine its review of ABWR, only to the choiec of the liquid radwaste tank. ( Therefore, provide informadan on the following: (15.73) w Amendocnt 2$ 20.2-21

ABWR ammxt SLpndard Plant ntw n (a) Basis for determining the concentrated waste tank as the worst tank (this may very well be the case, but in the absence of information on the capacities of major tanks, particularly the waste holdup tanks, it is hard to conclude that the above tank both in terms of radionuclide concentrations and inventories will turn out to be the worst tank). (b) Radionuclide source terms, particularly for the long-lived radionuclides such as Cs-137 and Sr.90 (these may be the critical isotopes for sites that can claim only decay credit during transit) in the major liquid radwaste tanks. 440.108 Provide further justification for the fact that the input parameters and initial conditions for analyzed events are conservative. Provide a list of what parameters will be checked at startup and I which will be in the Technical Specifications. You should define the range of operating conditions and fuel types for which your input parameters will remain valid. For example, would these parameters valid for 9x9 or ;x7 fuel or similar large change in the fact lattice. (15) 470.1 Subsection 15.6.2 of the ABWR FSAR provides your analysis for the radiological consequences of a failure of small lines carrying primary coolant outside of containment. This analysis only considers the failure of an instrument line with a 1/4 inch flow restricting orifice. Show that this failure scenrio prov! des the me;t sacre radioactive releases of any postulated failure of a small line. Your evalution dould iralude lines that meet GDC 55 as well as small lines exempt from GDC 55, 470.2 O Provide a justification for your assumption that the plant continues to operate (and therefore no iodine peaking is experien:ed) during a small line break outside containment (Subsection 15.6.2) accident scenario. Also provide the basis for the assumption that the release duration is only two hours. 4703 Subsection 15.6.4.5.1.1 of the FSAR gives the iodine source term (concentration and isotopic mix) used to analyze the steam line break outside of containment accident. The noble gas source term, however, is not addressed, Provide the noble gas source term used. Also, the table in Subsection 15.6.4.5.1.1 seems heavily weighted to the shorter lived activities (i.e., (1134). Provide the bases for the isotopic mix used in your analysis (iodine and noble gas). 470.4 Subsection 15.6.5.5 states that the analysis is based on assumptions provided in Regulatory Guide 13 except where noted. For all assumptions (e.g., release assumed to occur one hour after accident initiation, the chemical species fractions for iodine, the temporal decrease in primary containment leakage rates, credit for condenser leakage rates, and dose conversion factors) which deviate from NRC guidance such as regulatory guides and ICRP2, provide a detailed description of the justification for the deviation or a reference to another section of the SSAR where the deviations are discussed in detail. Provide a comparison of the dose estimates using these assumptions versus those which would resuit from using the NRC guidance. O Amendment 25 20.2-21.1 l

ABWR 2mor Etandard Plant nn n 470.5 (g) Provide a discussion of, or reference to, the analysis of the radiological consequences of leakage from engineered safety feature components after a design basis LOCA. 470.6 For the spent fuel cask drop accident, what is the assumed period for decay from the stated power condition? What is the justification for that assumption? 470.7 The tables in Chapter 15 should be checked and revised as appropriate. In several cases the footnotes contain typographical errors related to defining the scientific notation. Table 15. 12 also appears to contain inappropriate references to Table 15.716, rather that Table 15.713. 470.8 it is stated that Regulatory Guides 1.3 and 1.45 were used in the calculations of X/O values, Based on the values presented, it appears as though a Pasquill stability Class F and one neter per second wind speed were assumed, with adjustment for meander per Figure 3 of Regulatory Guide 1.145. If th/s is not the case, describe the assumptions and justification used in calculating the X/O values which are used in Chapter 15 dose assessments. 470.9 The SGTS filter efficiencies of 99% for inorganic and organic iodine are higher than the 90% and 70% values, respectively, assumed in Regulatory Guide 1.25 if it can be shown that the building (~ atmosphere is exhausted through adsorbers designed to remove iodine. Provide a justification for the ( use of the higher values. 470.10 Dose related factors such as breathing rates, iodine conversion factors and finite versus infinite cloud assumptions for calculating the whole body dose are not stated explicitly, although reference is made to Regulatory Guide 1.25 and another document. State these assumptions explicitly and justify use of any values which deviate from Regulatory Guide 1.25. 440.109 Provide an analysis of the loss ofinstrument air (nitrogen). (15) 440.110 In SSAR Table 15.0 2, the following transients are not categorized as moderate frequency event [ Category (a)] (a) Runout of two feedwater pumps (Cat.c) (b) Opening of all Control and Bypass Valves (Cat.c) (c) Pressure Regulator Downscale failure (Cat.c) (d) Generator load Rejection, Failure of One Bypass Valve (Cat.b) (e) Generator Load Rejection with Bypass Off (Cat.c) (f) Turbine Trip with Failure of One Bypass Valve (Cat.b) f (g) Turbine trip, Bypass Off (Cat.c) (3) (h) Loss of Aux. Power Transformer and one S/up transformer (Cat c) Amendment 25 20.2-21.2 l

ABWR muur Standani Plant um n (i) Tr;p of all Reactor Internal Pumps (Cat.c) (j) Fast Runback of all Reactor Internal Pumps (Cat.c) (k) Inadvertent HPCF pump start-up (Cat.b) Category b refers to Infrequent event and Category c refers to limiting faults. -The above categorization of transients is a significant deviation from the SRP and hence sufficient justification must be submitted to support the change in the categorization. (15) 440.111 Provide a table similar to 15.0 2 showing your evaluation of anticipated transients with single failure. List the single failure chosen for each event and provide a justification for why the chosen failure is the most limiting. (15) 440.112 Provide the following: (1) A listing of all equipment which is not classified as safety-related but is assumed in FSAR analyses to mitigate the consequences of transients nr accidents. (2) Justification for the assumption of operability of this equipment based upon equipment quality, reliability, and proposed surveillance requirements. -(3) Discuss the consequences of those events concerning (i) 1. umber of fuel failures, (ii) delta CPR and (iii) delta perk pressure that would result if only safety grade systems or components were considered in the specific transients analyses taking credit of non safety grade systems or components. (15) 440.113 You have classified the trip of t l reactor internal pumps as a limiting fault. This i i your assumption that the loss o. greater than three reactor internal pumps is 10'g based o per year. Provide operating experience data to justify this failure rate. (15) 440.114 The ABWR feedwater control system and the steam bypass and pressure control system use a triplicated digital system. You claim that no single failure in these systems will cause a minimum demand to all turbine control valves and bypass valves or the runout of two feedwater pumps. (15) (a) What is the reliability of the system? (b) Wriat design feature of these systems prevent common mode failure to more than one channel? (c) What protection is provided in these systems against a technician disabling a second th:.nnel while performing maintenance on the first. (d) What are the most limiting events for the case where two channels are lost in these systems? O Amendment 25 20121.3 l

ABWR msiwur Standard Plant imv u 440.115 , ~ - ( ) Provide further analysis and numerical justification for your assertion that FMCRD design is '~' equivalent to an ARI system and that the SLCS is not required to vespor.d to an ATWS. (15)' 440.116 For each transient and accident, identify the computer code used in the analysis in the respective section of Chapter 15. (15) /~N \\_, o) L An.cadment 15 20,2-21A l

ABWR u m mar Standardflant wn 20.311 Response to Eleventh RAI. Reference 11 QUESTION 281.15 in a letter from Thomas E. Murley, NRR, to Ricardo Artigas, G.E. dated August 7,1987, the staff presided the ABWR licensing review bases as well as the scope and content of the ABWR Standard Safety Analysis Report (SSAR). In Section 8.7, Water Chemistry Guidelines, of the referenced letter, it states that G.E. has commited to using BWR Owners Group water chemistry guidelines. These guidelines are necessary to maintain proper water chemistry in BWR cooling systems to prevent intergranular stress corrision cracking of auustenitic stainless steel piping and components and to minimize corrision and erosion /corrision induced piping wall thinning in single-phase and two-phase high energy carbon steel piping. Water chemistry is also important for the minimization of plant radiation levels due to activated corrision products. Section 10.4.63 of the ABWR indicates that the condensate cleanup system complies with Regulatory Guide 1.56. Section 10.4 should indicate that the system meets the guidelines published in: EPRI NP 4947 SR, BWR 11ydrogen Water Chemistry Guidelines 1987 Revision, dated October 1988. EPRI NP 5283-SR-A, Guidelines for Permanent BWR llydrogen Water Chemistry-1987 Revision, dated September 1987. The use of zine injection as a means of controlling UWR radiation field build-up should be dicussed. RESPONSE 281.15 A new Subsection 93.9 has been added to describe the hydrogen addition system. Revised Subsection 5.23 indicates that the guidelines in EPRI NP-4947 SR, BWR Ilydrogen Water Chemistry Guidelines 1987 p Revision, October 1988 and EPRI NP-5283-SR-A, Guidelines for Permanent BWR liydrogen Water Chemistry-1987 Revision, September 1987 will be met. This is also indicated in new Subsection 93.9, s Subsection 93.11 has been added to describe the zine addition system. QUESTION 281.16 In Section 10.4.63, the ABWR SSAR indicates that the. condensate cleanup system removes some radioactive material, activated corrision products and fission products that are carried over from the reactor. More important functions involve removal of coridensate system corrision products, and possible impurities from condenser leakage to assure meeting BWR liydrogen Water Chemistry Guidelines. This should be discussed. RESPONSE 210.16 Subsection 5.23.2.23 has been modified to discuss the removal of condensate system corrision and possible impurities from condenser icakage. QUESTION 210,17 The condensate (Figure 10.4-4) and feedwater (Figure 10.4-7) system diagrams do not indicate the location of the oxygen injection into the condensate system and hydrogen and zinc oxide into the feedwater system. This information should be provided. O Amendment 11 20.3-311

ABWR mar Standard Plant %.n RESPONSE 210.17 The location of oxygen addition for the condensate system is in Subsection 93.10. The location of hyd ogen adJition to the feedwater system will be shown in Subsection 93.9. The location of zine addition to the feedwater system is in Subsection 93.11. QUESTION 210.18 Section 10.4 does not discuss design improvements invohing material selection, water chemistry, system temperatures, piping design and hydrodynamic conditions that are necessary to control erosion / corrosion. The - EPRI CHEChtATE or other crosion/ corrosion computer codes may be useful design tools to minimize wall thinning due to crosion/ corrosion. The ABWR SSAR should discuss design considerations to minimize crosion/ corrosion and procedures and administrative controls to assure that the structuralintegrity of sirigle-phase and two-phase high energy carbon steel piping system is maintained. RESPONSE 210.18 l A discussion on the control of erosion corrosion of carbon steel has been added to Subsectic<n 5.23.2.23. QUESTION 420.123 (15B4) S3AR 15B.4 describes the essential multiplexing system (EhQ in some detail. SSAR Figu e 7A.2-1 states that the design is not limited to this configuration. It is our understanding that the EMS design 'l is stillin a preliminary design stage. Is SSAR 15B.4 still' accurate and is the design limited sto that configuration? RESPONSE 420.123 SSAR 15B.4 is an accurate system level description of EMS and reflects the components described in the ~ EMS design specification and SSLC design specification, and is the chosen system conGguration. The exact hardware implementation is not specified for design certification, since potential vendors could accomplish the multiplexing functions in several ways, given the restriction that qualification requirements must be met, liowever, certain design details to be imposed on such vendors are discussed in the various responses given in SSAR 7A.2. liardware and software ("firmware") requirements down to the module level of the equipment are described. I The EMS design is presently defined to the level of the type of processing components needed to perform the data transmission task. The design requires remote multiplexing units, control room multiplexing units and fiber optic interconnecting links. The bi-directional, dual redundant token ring topology is the chosen configuration for these components, and is the configuration shown in SSAR Figure 7A.21. However, the multiplexing tasks shown in the figure could also be accomplished by the same components arranged in a star, l bus, or point to-point architecture (all still using a dual redundant configuration). This part of the design will be determined during the detailed design phase, depending upon the required system speed (data throughput), response time, the vendor's communications protocols, error detection / correction methods, and available hardware / software designs. GE believes that specifying the exact EMS configuration at the design certification stage could skew competitive bidding for potential vendors of the equipment. The system requirements imposed on the multiplexed safety system design (e.g., single failure proof, signal isolation,2-out of-4 system logic, bypassing of failcd components, self test, easy repair, periodic surveillance, highly reliable materials, verification and validation of software, and integration testing) are sufficient to provide a qualified design. Amendment 23 20.3-312

ABWR 2mmr Standard Plant un electromagnetic or radio frequency interference (EMI/RFI). However, near the electrical to opticalinterfaces (')\\ at the transmitting and receiving cabinets, the equipment imernals must be protected from EMl/RFI effects, C such as that produced by keying portable radio transceivers near control cabinets. Protection can take the form of RF shielding of instruments and cabinets, proper instrument grounding and power distribution grounding, and filtering of input / output signal and power line conductors, as recommended in ABWR system design documents. Tests to demonstrate immunity of safety protection equipment to radio transceiver broadcasts will be developed using guidelines described in the following documents: (1) ANSI /IEEE C37.90.2 - 1987,IEEE Trial-Use Standard, Withstand Capability of Relay Systems to Radiated Electromagnetic Interference from Transceivers. Section 5.5.3 of this standard describes tests for digital equipment using clocked logic circuits. Thus, appropriate pre-operational or periodic surveillance tests will be developed for protective relays,if used, and for microivocessor-be ed EMS /SSLC equi en: rpec:hed n; +.BWR design docements. (2) ANSI C63.12-1987, American National Staadard for Electromagnetic Compatibility Limits - Recommended Practice. Plant areas that cannot be shown to be EMI/RFI compatible will be posted as'NO RADIO USE" zones. Tbc following areas are out of t e ABWR Standaid Plant design scope, and shall be included in the h (Lj appheat fire p otecuca program: (1) Main transformer ( (2) Equipment entry lock C (3) Fire protection pumphouse (4) Ultimate heat sink The applicant's fire protection program shall comply with the SRP Section 9.5.1, with ability to bring the plant to safe shutdown condition following a complete Ere area burnout without a need for recovery, QUESTION 430315 The fire hazard analysis provided as Appendix 9A listed several components within the rooms of each fire area in the reactor building. However, specific cables (power and instrumentation) were not identified in the equipment listings (Tables 9A.b-1 and 9A.b.2). The failure of these cables will have to be included in a safe shutdown analysis. Additionally, the equipment listed in these two tables showed that equipment powered by separate divisions of AC power (Division 1 and 2 for example) are located in the same reactor building Gre zones. From the information in Appendix 9A,it is nbt possible to determine if the failure of this equipment could affect the operability of required safe shutdown equipment in other fire areas. This equipment should be addressed in the safe shutdown analysis, including an associated circuit analysis. (9.5.1) RESPONSE 430316 (A) The power and instrumentation cables for each piece of equipment are considered to be part of the equipment listed in Tables 9A,6-1 & 9A.6-2, and therefore, are not listed separately (see Table 9A.6 and 2 notes). The Gre hazard analysis included these cables as part of the equipment located within the Erc area, and their failure was considered as part of the equipment failure in the safe shutdown analysis. (%v) t Amendment 22 20.3-375

ABWR nunwr Shwiilrd Plant mp l A new Subsect on,9.51.2.11 has been cdded. Information for the remainder of the response to this i question is 14en from there. The systems whose primary functions are to provide core cooling, and to bring the plant to a safe shutdown condition, have three independent mechanical and electrical safety related divisions (mechanical Divisions A, B, C and electrical Divisions I,II, III). Each division is capable of bringing the plant to a safe shtadown condition whether the system is initiated manually or automatically. The plant layout is such that redundant portions of safety related systems are located in different fire areas, therefore if one division becomes disabled due to a fire (complete burnout is acceptable) there are still two independent redundant divisions available to provide core cooling. The system initiation logic is two-out of.four logic, but if one division becomes disabled (e.g., due to a fire) the system initiation logic reverts to two-out of three logic. The initiation circuitry is located in the control room. There are some instances where, for overriding technical reasons, equipment from more than one safety division is purposely mounted in the same nie area. For example, the equipment mounted in Rocm 110 of the reactor building is for the Division I portion of the RHR system. There are also two leak detection thermocouples of Divisions I and 11 mounted in the same room. The Division 11 leak detection tiiermocouple provides redundancy for leak detection initiation to control the singic division of equipment contained in the room. if the room is burned out completely, the loss of Division Iis acceptable, because there are two independent redundant divisions (II, & 111) available in different fire areas of the building to bring the plant to a safe thutdown condition. The loss of the Division 11 leak detection thermocouples is acceptable because their loss will not impair the operation of the remaining portion of the Division 11 leak detectioc system (because of the nature of the thermocouple itself and its operation, the loss would be isolated locally. The acceptability of each case of intruding divisional devices, su ch as this, is analyzed on a case-by case basis and reported in Section 9A.5. In summary, each safety-related device (cables included) is assigncd to a safety division. External to the control room and primary containment, devices of different safety divisions are located in different fire areas unless over-riding technical considerations dictate exceptions. All exceptions are analyzed and the analysis reported in Section 9A.5. l With the above facts known and a set of fire area separation drawings,(Figure 9A.41 through 9A.4-32), it is possible to confirm the acceptability of the chosen locations for the devices listed on Tables 9A.6-1 and 9A.6-2. (B) Due to the attention paid to separation and the isolation capabilities of the fiber optic data transmission systems, there are no associated circuits in the ABWR Standard Plant Design. QUESTION 430317 Section 9.5.1.2.1 should be expanded to include the fire protection water supply system. RESPONSE 430317 The fire protection water supply system is discussed in Subsection 9.5.1.2.5 as part of Amendment 14. QUESTION 430.318 Section 9.5.1.2.2 states that a manually operated carbon dioxide (CO2) fire suppression system will be provided for the diesel generated rooms, including the day tank rooms. This does not correspond to the guidance provided in NUREG-0800, CMEB BTP 9.51, Section C.7.i which specifies automatic fire suppression for the emergency diesel generators. This section should be changed to show automatic fire suppression or expanded tojustify how manual suppression provides either equivalent or superior protection. Amendment 25 201 376}}