ML20049H259

Forwards Gessar & General Info Attachment

ML20049H259
Person / Time
Site:	05000447
Issue date:	02/12/1982
From:	Sherwood G GENERAL ELECTRIC CO.
To:	Eisenhut D Office of Nuclear Reactor Regulation
Shared Package
ML20049H260	List: ML20049H259
References
JNF-06-82, JNF-6-82, MFN-012-82, MFN-12-82, NUDOCS 8202230003
Download: ML20049H259 (33)

Text

GENERAL h ELECTRIC NUCLEAR POWER SYSTEMS DIVISION GENERAL ELECTRIC COMPANY.17s CURTNER AVE., SAN JOSE. CALIFORNIA 9s12s MC 682, (408) 925-5040 JNF-06-82

, MFN-012-82 February 12, 1982 D d  %

6 RECEWED 4 U.S. Nuclear Regulatory Commission -

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Office of Nuclear Reactor Regulation FEB 2EN t aggyi7 }-

Washington, DC 20555 19C Attention: D. G. Eisenhut, Director

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Division of Licensing f 7

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SUBJECT:

IN THE MATTER OF 238 NUCLEAR ISLAND GENERAL ELECTRIC STANDARD SAFETY ANALYSIS REPORT (GESSAR II): DOCKET NO. STN 50-447

Reference:

Letter to Glenn G. Sherwood frcm Darrell G. Eisenhut,

" Acceptance Review of Application for Final Design Approval for 238 Nuclear Island," December 9, 1981 In accordance with the referenced acceptance letter, the General Electric Company (GE) herewith files the following documents for the docketing of its application for final design approval for the 238 nuclear island:

a) Three (3) signed originals and fifteen (15) copies of that portion of the general information attachment (Attachment 1) and b) Forty (40) copies of the General Electric Standard Safety Analysis Report, GESSAR II (under separate cover).

In addition to the above, GE has retained ten (10) copies of the general information attachment and thirty (30) copies of GESSAR II for direct distribution with instructions which might be provided later by the staff.

GESSAR II proprietary information will be provided under separate cover by February 28, 1982.

As requested by the acceptance letter, the facility-unique information in '

GESSAR II has been modified to elminate the need for colored pages.

However, GE will make a complete set of this facility-unique information (based on the TVA-Hartsville Plant) available to the staff. While this information is not required for the review, it eill aid the staff by providing " typical" facility-unique information.

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//W F202230003 820212 .

PDR ADOCK 05000 K

ELECTRIC GEN E3 AL @ lear Regulatory Commission U.S. Nuc Page 2 GE will provide all of the additional information requested by the acceptance letter (TMI-related information, Enclosure 1, and Enclosure 2) within three months of the docketing date with the exception of the FMEA and geotechnical information which will be provided by September 30, 1982.

GE considers it appropriate to delay submittal of the FMEA information since the FMEAs requested are an integral part of a total FMEA program currently in process and an earlier submittal could compromise their completeness. GE finds it necessary to delay submittal of the geo-technical information because the level of detail requested will require an appreciable effort to answer.

As agreed at the January 6, 1982 GE/NRC meeting at Bethesda, Maryland, the GESSAR II review will be conducted using a " difference" approach in which GE identifies the differences between GESSAR II and similar BWR/6's and the Staff limits its review to the differences. GE provided a preliminary review matrix at the January 28, 1982 GE/NRC meeting at San Jose, California, and agreed to provide a final review matrix at the time of docketing. This review matrix is provided in Attachment 2.

Very truly yours, Glenn G. Sherwood, Manager Nuclear Safety and Licensing Operation GGS:hmc/002081 D

UNITED STATES 0F A M E F. I C A NUCLEAR REGULATORY C0MMISSION In the Matter of )

General Electric Company ) Docket No. STN 50-447 238 Nuclear Island GESSAR II )

APPLICATION FOR REVIEW OF " GENERAL ELECTRIC STANDARD SAFETY ANALYSIS REPORT" 238 Nuclear Island GESSAR II General Electric hereby applies for Nuclear Regulatory Commission and Advisory Committee on Reactor Safeguards review of the attached General Electric Standard Safety Analysis Report, "238 Nuclear Island GESSAR II,"

pursuant to 10CFR50, Appendix 0.

Appended to this application as Attachment 1 is the information described in Paragraph 50.33(a)-(d). The 238 Nuclear Island GESSAR II document contains the technical information pursuant to Paragraph 50.34(a).

Respectfully submitted, By: , /., V[ -

A. Philip Bray /

Vice President and General Mafiager Subscribed and sworn 'o before me the 12 day of Februnrv .

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ATTACHMENT 1 GENERAL ELECTRIC CORPORATE INFORMATION (a) Name of Applicant

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General Electric Company (b) Address of Applicant General Electric Company Attn: Glenn G. Sherwood, Manager Nuclear Safety and Licensing Operation Nuclear Power Systems Division 175 Curtner Avenue, Mail Code 682 San Jose, California 95125 (c) Description of Business or Occupation of Applicant The Nuclear Energy Group of the General Electric Company has as its principal business the design and manufacture of major portions of boiling water reactors including fuel assemblies manufactured in Wilmington, North Carolina; reactor vessels; instrumentation systems; control systems; coolant systems and other major components and systems which may be identified as the Nuclear Steam Supply System. These major design and manufacturing activities are augmented and supported by a wide range of research and development programs, including operation of facilities at the Vallecitos Nuclear Center.

In addition, the General Electric Company designs a major portion of the Lalance-of plant, exclusive of the turbine generator systems, including engineered safeguard systems, containment, auxiliary building, control room fuel building, diesel generator buildings, radwaste building and related systems and structures. Other associated activities include the servicing of nuclear reactors during refueling outages, the storage of spent fuel assemblies at the Morris, Illinois fuel storage facility, and the procurement of uraniu.n dioxide. The interests of the Division include both domestic and overseas business ventures for nuclear power facilities.

(d) Corporate Data (i)

General Electric Company is incorporated in and its l principal place of business is the State of New York.

l (ii)

Directors and principal officers are listed below. All of these directors and officers are citizens of the United States of America.

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Principal Corporate Officers Chairman of the Board and Chief Executive Officer John F. Welch, Jr.

Vice Chairman of the Board and Directors John F. Burlingame Edward E. Hood, Jr.

Address of Principal Corporate Officers '

General Electric Company 3135 East Turnpike Fairfield, Connecticut 06431 Board of Directors Richard T. Baker James G. Boswell II John F. Burlingame Silas S. Cathcart Charles D. Dickey, Jr.

Lawrence E. Fouraker Henry H. Henley, Jr.

Henry L. Hillman Edward E. Hood, Jr.

Ralph Lazarus Edmund W. Littlefield George M. Low Gertrude G. Michelson Lewis T. Preston Gilbert H. Scribner, Jr.

Walter B. Wriston (iii) The General Electric Company is not owned, controlled or dominated by an alien, a foreign corporation or a foreign government.

RV:hmc/D02083-2

ATTACHMENT 2 GESSAR II REVIEW MATRIX LEGEND j

P PERRY C -

CLINTON GG -

GRAND GULF RB -

RIVER BEND PDA - PRELIMINARY DESIGN APPROVAL U UNIQUE (FULL REVIEW)

A -

APPLICANT f

N/A -

NOT APPLICABLE TO GESSAR II i

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167J1 7 - -- + - -~-,-__m . _ _ _ _ _ _ _ _ _ _ _ , . _ _ _ _ , , _ _ _ _ _ , _ _

CHAPTER 1 Introduction and General Description of Plant Review Base

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1.1 Introduction PDA 1.1.1 Type of License Required A 1.1. 2 Identification of Applicant U 1.1.3 Numter of Plant Units A 1.1.4 Description of Location U 1.1. 5 Type of Nuclear Steam Supply System U 1.1.6 Type of Containment U 1.1. 7 Core Thermal Power Levels PDA 1.1.8 Scheduled Completion and Operation Dates 1.2 General Plant Description PDA 1.2.1 Principal Design Criteria PDA 1.2.2 Plant Description 1.3 Comparison Tables U 1. 3.1 Comparisons with Similar Facility Designs U 1.3.2 Comparisons of Final and Preliminary Information 1.4 Identification of Agents and Contractors A 1.4.1 Applicant U 1.4.2 Architect Engineer - Nuclear Island Design PDA 1.4.3 Nuclear Steam Supply System Supplier A 1.4.4 Turbine-Generator Vendor A 1.4.5 Consultants 1.5 Requirements for Further Technical Information P 1.5.1 Current Development Programs PDA 1.5.2 PSAR Commitment Items U 1.6 Material Incorporated by Reference 1.7 Drawings and Other Detailed Information U 1.7.1 Electrical, Instrumentation, and Control Drawings U 1.7.2 Piping and Instrumentation Diagrams U 1.7.3 Abbreviations and Symbols i

! U 1.8 Conformance to NRC Regulatory Guicas l

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Review

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1.9 Standard Designs U 1.9.1 Interfaces U 1.9.2 Exceptions U App 1A TMI P App 1B Unresolved Safety Issues U App IC NRC Additional Guidance for GESSAR II i

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CHAPTER 2 Site Characteristics Review Base

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U 2.0 Summary 2.1 Geography and Demography A 2.1.1 Site Location and Description A 2.1.2 Exclusion Area Authority and Control A 2.1.3 Population Distribution 2.2 Nearby Industrial, Transportation and Military Facilities A 2.2.1 Location and Routes A 2.2.2 Descriptions A 2.2.3 Evaluation of Potential Accidents 2.3 Meteorology A 2.3.1 Regional Climatology A 2.3.2 Local Meteorology A 2.3.3 Onsite Meteorological tieasurements Program A 2.3.4 Short-Term Atmospheric Diffusion Estimates A 2.3.5 Long-Term Atmospheric Diffusion Estimates U 2.4 Hydrologic Engineering A 2.4.1 Hydrologic Description U 2.4.2 Floods A 2.4.3 Probable Maximum Flood (PMF) on Streams and Rive-s A 2.4.4 Potential Dam Failures, Seismically Induced A 2.4.5 Probaole Maximum Surge and Seiche Flooding A 2.4.6 Probable Maximum Tsunami Flooding A 2.4.7 Ice Effects A 2.4.8 Cooling Water Canals and Reservoirs A 2.4.9 Channel Diversions A 2.4.10 Flooding Protection Requirements A 2.4.11 Low Water Considerations A 2.4.12 Dispersion, Dilution, and Travel Times of Accidental Releases of Liquid Effluents in Surface Waters A 2.4.13 Ground Water A 2.4.14 Technical Specifications and Emergency Operation c

Requirements 2.5 Geology, Seismology, and Geotechnical Engineering A 2.5.1 Basic Geologic and Seismic Information A 2.5.2 Vibratory Ground Motion A 2.5.3 Surface Faulting A 2.5.4 Stability of Subsurface Materials and Foundations A 2.5.5 Stability of Slopes A 2.5.6 Embankments and Dams rf:csc/167M3 2/11/82

CHAPTER 3 Design of Structures, Components, Equipment, and Systems Review Base 3.1 Conformance with NRC General Design Criteria GG 3.1.1 Summary Description GG 3.1.2 Criterion Conformance GG 3.2 Classification of Structures, Components, and Systems GG 3.2.1 Seismic Classification GG 3.2.2 System Quality Group Classifications GG 3.2.3 System Safety Classifications GG 3.2.4 Quality Assurance GG 3.2.5 Correlation of Safety Classes with Industry Codes U 3.3 Wind and Tornado Loadings PDA 3.3.1 Wind Loadings PDA 3.3.2 Tornado Loadings U 3.3.3 BOP Interface U 3.4 Water Level (Flood) Design P 3.4.1 Flood Protection P 3.4.2 Analytical Test Procedure U 3.4.3 B0P Interface PDA 3.5 Missile Protection PDA 3.5.1 Missile Selection and Description PDA 3.5.2 Structures, Systems, and Components to be l Protected from Externally Generated Missiles PDA 3.5.3 Barrier Design Procedures l PDA 3.5.4 B0P Interface PDA 3.6 Protection Against Dynamic Effects Associated With The Postulated Rupture of Piping PDA 3.6.1 Postulated Piping Failures in Fluid Systems Inside and Outside of Containment PDA 3.6.2 Determination of Break Locations and Dynamic Effects s Associated with the Postulated Rupture of Piping l

U 3.7 Seismic Design PDA 3.7.1 Seismic Input PDA 3.7.2 Seismic System Analysis PDA 3.7.3 Seismic Subsystem Analysis PDA 3.7.4 Seismic Instrumentation U 3.7.5 BOP Interface rf:csc/167M4 2/11/82 l

Review Base PDA 3.8 Design of Seismic Category I Structures PDA 3.8.1 Concrete Containment PDA 3.8.2 Steel Containment PDA 3.8.3 Concrete and Steel Internal Structures of Steel Containment

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PDA 3.8.4 Other Seismic Category I Structures PDA 3.8.5 Foundations PDA 3.8.6 BOP Interface 3.9 Mechanical Systems and Components P 3.9.' Special Topics for Mechanical Components P 3.9.F Dynamic Testing and Analysis P 3.9.3 ASME Code Class 1, 2 and 3 Components, Component Supports, and Core Support Structures P 3.9.4 Control Rod Drive Sytem P 3.9.5 Reactor Pressure Vessels Internals U 3.9.6 Inservice Testing of Pumps and Valves U 3.9.7 BOP Interface U 3.10 Seismic Qualifications of Seismic Category I Instrumen-tation and Electrical Equipment (Including Hydrodynamic Efforts)

P 3.10.1 Seismic Qualification Criteria (Including Hydrodynamic Loads)

P 3.10.2 Methods and Prc:edures for Qualifying Electrical Equipment and Instrurentation P 3.10.3 Methods and Prccedure of Seismic Analysis or Testing of Supports of Electrical Equipment and Instrumenta-tion (Including Hydrodynamic Loads)

U 3.10.4 Seismic Qualification Tests and Analyses (Including Hydrodynamic Loads)

U 3.11 Environmental Design of Safety-Related Mechanical and Electrical Equipment P 3.11.1 Equipment Identification and Environmental Conditions P 3.11.2 Qualification Tests and Analyses U 3.11.3 Qualification Results U 3.11.4 Loss of Ventilation P 3.11.5 Estimated Chemical and Radiation Environment PDA App 3A Seismic Soil-Structure Interaction Analysis of the Nuclear Island rf:csc/167M5 2/12/82

Review Base

• App 3B Containment Loads 4

U App 3C Computer Programs Used in the Design of Seismic Category I Structures U App 3D Analysis of Recirculation Motor and Pump Under Accident Conditions U App 3E Description of Safety Related Mechanical and Electrical Equipment U App 3F Dynamic Buckling Criteria fnr Containment Vessel U App 3G Pipe Failure Analysis l

U App 3H Effect of Concrete Annulus Below Elevation (-)

5 Feet, 3 Inch on Seismic Design Loads and Building Responses l

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• Review completed on GESSAR 11 pre-docket version.

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CHAPTER 4 Reactor Review Base

_C 4.1 Summary Description C 4.1.1 Reactor Vessel C 4.1.2 Reactor Internal Components C 4.1.3 Reactivity Control Systems C 4.1.4 Analysis Techniques 4.2 Fuel System Design C 4.2.1 General and Detailed Design Base C 4.2.2 General Design Description C 4.2.3 Design Evaluations C 4.2.4 Testing and Inspection C 4.2.5 Operating and Developmental Experience C 4.3 Nuclear Design C 4.3.1 Design Bases C 4.3.2 Description C 4.3.3 Analytical Methods C 4.3.4 Changes 4.4 Thermal - Hydraulic Design C 4.4.1 Design Basis C 4.4.2 Description of Thermal Hydraulic Design of the Reactor Core C 4.4.3 Description of the Thermal and Hydraulic Design of the Reactor Coolant System C 4.4.4 Evaluation C 4.4.5 Testing and Verification C 4.4.6 Instrumentation Requirements l

4.5 Reactor inaterials C 4.5.1 Control Rod System Structural Materials C 4.5.2 Reactor Internal Materials C 4.5.3 Control Rod Drive Housing Supports

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Review Base C 4.6 Functional Design of Reactivity Control Systems C 4.6.1 Information for Control Rod Drive System (CRDs)

C 4.6.2 Evaluations of the CRDs

_ C 4.6.3 Testing and Verification of the CRDs C 4.6.4 Information for Combined Performance of Reactivity Systems C 4.6.5 Evaluation of Combined Performance i

C App 4A Control Rod Patterns and Associated Power Distribution for Typical BWR 4

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CHAPTER 5 Reactor Coolant System and Connected Systems Review Base

~C 5.1 Summary Description C 5.1.1 Schematic Flow Diagram C 5.1.2 Piping and Instrumentation Diagram C 5.1. 3 Elevation Drawing C 5.2 Integrity of Reactor Coolant Pressure Boundary V 5.2.1 Compliance with Codes and Code Cases C 5.2.2 Overpressure Protection C 5.2.3 Reactor Coolant Pressure Boundary Materials C 5.2.4 Inservice Inspection and Testing of Reactor Coolant Pressure Boundary C 5.2.5 Reactor Coolant Pressure Boundary and ECCS System Leakage Detection System 5.3 Reacter Vessel C 5.3.1 Reactor Vessel Materials ,

C 5.3.2 Pressure / Temperature Limits C 5.3.3 Reactor Vessel Integrity 5.4 Component and Subsystem Design C 5.4.1 Reactor Recirculation Pumps N/A 5.4.2 Steam Generators C 5.4.3 Reactor Coolant Piping C 5.4.4 Main Steam Line Flow Restrictors C 5.4.5 Main Steam Line Isolation System C 5.4.6 Reactor Core Isolation Cooling System (RCIC)

C 5.4.7 Residual Heat Removal System C 5.4.8 Reactor Water Cleanup System C 5.4.9 Main Steam Lines and Feedwater Piping

, N/A 5.4.10 Pressurizer i N/A 5.4.11 Pressurizer Relief Valve Discharge System l C 5.4.12 Valves

! C 5.4.13 Safety and Relief Valves C 5.4.14 Component Supports l

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CHAPTER 6 Engineered Safety Features Review Base

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P 6.0 General P 6.1 Engineered Safety Feature Materials P 6.1.1 Metallic Materials U 6.1.2 Organic Materials 6.2 Containment Systems PDA 6.2.1 Containment. Functional Des'gn PDA 6.2.2 Containment Heat Removal S3 item U 6.2.3 Secondary Containment Funct.onal Design PDA 6.2.4 Containment Isolation System PDA 6.2.5 Combustible Gas Control in Containment PDA 6.2.6 Containment Leakage Testing PDA 6.2.7 Suppression Pool Makeup System 6.3 Emergency Core Cooling Systems P 6.3.1 Design Bases and Summary Description P 6.3.2 System Design P 6.3.3 ECCS Performance Evaluation P 6.3.4 Tests and Inspections P 6.3.5 Instrumentation Requirements U 6.4 Habitability Systems U 6.4.1 Design Basis U 6.4.2 System Design U 6.4.3 Systems Operational Procedures U 6.4.4 Design Evaluations U 6.4.5 Testing and Inspection U 6.4.6 Instrumentation Requirements U 6.4.7 Nuclear Island /B0P Interface 6.5 fission Product Removal and Control Systems PDA 6.5.1 Standby Gas Treatment System (SGTS)

N/A 6.5.2 Containment Spray Systems U 6.5.3 Fission Product Control Systems N/A 6.5.4 Ice Condensers as a Fission Product Control System 9

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Review Base 6.6 Inservice Inspection of Class 2 and 3 Components U 6.6.1 Components Subject to Examination U 6.6.2 Accessibility A 6.6.3 Exe.mination Techniques and Procedures

~ A 6.6.4 Inspection Intervals A 6.6.5 Examination Categories and Requirements A 6.6.6 Evaluation of Examination Results i A 6.6.7 System Pressure Tests A 6.6.8 Augmented Inservice Inspection to Protect Ageinst Postulated Piping Failures RB 6.7 Main Steam Positive Leakage Control System (MSPLCS)

RB 6.7.1 Design Bases RB 6.7.2 System Description RB 6.7.3 System Evaluation RB 6.7.4 Inspection and Testing RB 6.7.5 Instrumentation Requirements

6. 8 Pneumatic Supply System U 6.8.1 Design Bases U 6.8.2 System Description U 6.8.3 System Evaluation U 6.8.4 Inspection and Testing Requirements U 6.8.5 Instrumentation Requirements U App 6A Improved Decay Heat Correlation for LOCA Analysis i

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CHAPTER 7 Instrumentation and Control Systems Review Base PDA 7.1 Introduction (All Systems)

PDA 7.1.1 Identification of Safety-Related Systems PDA 7.1. 2 Identification of Safety and Power Generation Criteria 7.2 Reactor Protection (Trip) System (RPS)

C 7.2.1 Description C 7.2.2 Conformance Analysis 7.3 Engineered Safety Features System, Instrumentation and Control '

7.3.1 Description 7.3.2 Analysis C -HPCS C -ADS C -LPCS C -RHR/LPCI C -CRVICS RB -MSPLCS C -RHR/Contm't Spray C -RHR/Suppr. Pool Cooling U -Suppr. Pool Makeup U -Combustible Gas Control U -SGTS U -Sh. Bldg. Annulus Mixing U -Secondary Containment Isol.

U -Prim. Containment Isol. LCS C -Standby Power U -D-G Support Systems U -Essential Service Water U -ESF Area Cooling U -Pneumatic Supply U -C.B. Atmos. Control U -C.B. Chilled Water 7.4 Systems Required for Safe Shutdown i 7.4.1 Description 7.4.2 Analysis C -RCIC C -RHR/ Shutdown Cooling C -SLC C -Remote Shutdown rf:csc/167M12 2/11/82

Review Base

7. 5 Safety-Related Display Ins rumentation 7.5.1 Description 7.5.2 Analysis

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C -Nuclenet Control Console C -BOP Benchboard C -Standby Info. Panel C -Supervisory Mon. Console C -Rx Core Cooling BB C -Display Control System 7.6 All Other Instrumentation Systems Required for Safety 7.6.1 Description 7.6.2 Analysis 7.6.3 Additional Design Considerations Analyses 7.6.4 References C -Neutron Monitoring P -Process Rad. Mon.

C -Refueling Inte locks C -Leak Detection C -Rod Pattern Control C -HP/LP System Interlock C -Recirc Pump Trip GG -FPCCS U -DW/Contm't Vacuum Relief U -Vent & Pressure Control C -CAMS U -Suppression Pool Temp. Mon.

7.7 Control Systems Not Required for Safety 7.7.1 Description 7.7.2 Analysis 7.7.3 References U -RPV Instrumentation C -Rod Control & Info.

C -Recirc Flow Control C -Feedwater Control C -Performance Monitoring Sys.

U -Radwaste l C -Leak Detection C -Rod Block Trp U -Fire Protection U -Drywell Chiller & Cooling U -Plant Instrument Air C -Neutron Monitoring rf:csc/167M13 2/11/82

Review Base U 7.8 NI/80P Interfaces U 7.8.1 Essential Service Water (Supply) System Instrumen-tation and Controls

_U 7.8.2 Diesel Generator Fuel Oil Transfer System

• App 7A I&C Elementary Diagrams

• Mixed review base (see 7.2 - 7.7)

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CHAPTER 8 Electric Power I

Review Base 8.1 Introduction A 8.1.1 Utility Grid Description U 8.1.2 Onsite Electric Power System U 8.1.3 Design Bases 1

8.2 Offsite Power System A 8.2.1 Description A 8.2.2 Analysis U 8.2.3 Nuclear Island - BOP Interface 8.3 Onsite Power Systems PDA 8.3.1 AC Power Systems PDA 8.3.2 DC Fower Systems U 8.3.3 Fire Protection of Cable Systems i

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CHAPTER 9 AUXILIARY SYSTEMS Review Base

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9.1 Fuel Storage and Handling P 9.1.1 New Fuel Storage (High Density)

P 9.1.2 Spent Fuel Storage (High Density)

P 9.1.3 Fuel Pool Cooling and Cleanup System P 9.1.4 Fuel-Handling System 9.2 Water Systems PDA 9.2.1 Essential Service Water (ESW) System PDA 9.2.2 Closed Cooling Water System PDA 9.2.3 Demineralized Water Makeup System PDA 9.2.4 Potable and Sanitary Water Systems PDA 9.2.5 Ultimate Heat Sink PDA 9.2.6 Condensate Storage Facilities and Distribution System PDA 9.2.7 Plant Chilled Water Systems PDA 9.2.8 Heated Water Systems U 9.2.9 Nuclear Island / BOP Interfaces 9.3 Process Auxiliaries P 9.3.1 Compressed Air Systems P 9.3.2 Process Sampling System P 9.3.3 Floor and Equipmerit Drainage Systems N/A 9.3.4 Chemical and Volume Control System (PWR)

P 9.3.5 Standby Liquid Control System 9.4 Air Conditioning, Heating, Cooling and Ventilation System V 9.4.1 Control Room HVAC System U 9.4.2 Fuel Building HVAC System U 9.4.3 Auxiliary Building HVAC Systems U 9.4.4 Turbine Building Area Ventilation System PDA 9.4.5 Reactor Building HVAC System V 9.4.6 Radwaste Building HVAC System U 9.4.7 Diesel-Generator Buildings HVAC Systems

9. 5 Other Auxiliary Systems U 9.5.1 Fire Protection System U 9.5.2 Communications Systems U 9.5.3 Lighting Systems U 9.5.4 Diesel-Generator Fuel Oil Storage and Transfer System U- 9.5.5 Diesel-Generator Cooling Water System rf:csc/167M16 2/12/82

Review

. Base U 9.5.6 Diesel-Generator Starting Air System U 9.5.7 Diesel Engine Lubrication System U 9.5.8 Diesel Generator Combustion Air Intake and Exhaust System

- U 9.5.9 Suppression Pool Cleanup System U 9.5.10 Nuclear Island - GOP Interface U App 9A Fire Hazard Analysis l

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CHAPTER '.0 Steam and Power Conversion Steam Review Base A 10.1 Summary Description 10.2 Turbine Generator C 10.2.1 Design Bases A 10.2.2 System Description A 10.2.3 Turbine Disk Integrity A 10.2.4 Evaluation 10.3 Main Steam Supply A 10.3.1 Design Bases A 10.3.2 Description A 10.3.3 Evaluation A 10.3.4 Inspection & Testing Requirements A 10.3.5 Water Chemistry A 10.3.6 Steam and Feedwater System Materials 10.4 Other Features of Steam and Power Conversion System A 10.4.1 Main Condensers A 10.4.2 Condenser Air Removal System A 10./ 3 Main Condenser Evacuation System P 10.4.4 Turbine Bypass System A 10.4.5 Circulating Water System A 10.4.6 Condensate Cleanup System P 10.4.7 Condensate and Feedwater System N/A 10.4.8 Steam Generator Blowdown System (PWR)

N/A 10.4.9 Auxiliary Feedwater System (PWR) l l

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CHAPTER 11 Radioactive Waste Management Review Base P 11.1 Source Terms P 11.1.1 Fission Products P 11.1.2 Activation Products P 11.1.3 Tritium P 11.1.4 Fuel Fission Production Inventory and Fuel Experience P 11.1.5 Process Leakage Sources P 11.1.6 Radwaste System P 11.1.7 Radioactive Sources in the Gas Treatment System GG 11.1.8 Source Terms for Component Failures C 11.1.9 Other Releases 11.2 Liquid Waste Management System PDA 11.2.1 Design Basis PDA 11.2.2 System Descriptions PDA 11.2.3 Estimated Releases 11.3 Gaseous Waste Management Systems P 11.3.1 Design Bases P 11.3.2 Main Condenser Steam Jet Air Ejector Low-Temp RECHAR System Description P 11.3.3 RECHAR System Operating Procedure C. 11.3.4 Radioactive Releases 11.4 Solid Radwaste System PDA 11.4.1 Design Bases PDA 11.4.2 System Description 11.5 Process and Effluent Radiological Monitoring and Sampling Systems GG 11.5.1 Design Bases C 11.5.2 System Description P 11.5.3 Effluent Monitoring and Sampling P 11.5.4 Process Monitoring and Sampling P 11.5.5 Calibration and Maintenance A 11.6 Offsite Radiological Monitoring Program l

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CHAPTER 12 Radiation Protection Review Base 12.1 Ensuring that Occupational Radiation Exposures Are as Low as Reasonably Achievable (ALARA)

PDA 12.1.1 Policy Considerations PDA 12.1.2 Design Considerations PDA 12.1.3 Operational Considerations 12.2 Radiation Sources P 12.2.1 Contained Sources P 12.2.2 Airborne Radioactive Material Sources 12.3 Radiation Protection Design Features PDA 12.3.1 Facility Design Features PDA 12.3.2 Shielding PDA 12.3.3 Ventilation PDA 12.3.4 Area Radiation and Airborne Radioactivity Monitors A 12.4 Dose Assessment A 12.5 Health Physics Program rf:csc/167M20 2/11/82

CHAPTER 13 Conduct of Operations Review Base 13.1 Organizational Structure of Applicant A 13.1.1 Management and Technical Support Organization A 13.1.2 Operating Organization A 13.1.3 Qualifications of Nuclear Plant Personnel 13.2 Training A 13.2.1 Plant Staff Training Program A 13.2.2 Replacement and Retraining A 13.2.3 Applicable NRC Documents 13.3 Emergency Planning A 13.3.1 Preliminary Planning A 13.3.2 Emergency Plan A 13.3.3 BOP Interface 13.4 Review and Audit A 13.4.1 Onsite Review A 13.4.2 Independent Review A 13.4.3 Audit Program 13.5 Plant Procedures A 13.5.1 Administrative Procedures A 13.5.2 Operating and Maintenance Procedures 13.5 Industrial Security A 13.6.1 Preliminary Planning A 13.6.2 Security Plan U 13.6.3 B0P Interface l

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CHAPTER 14 Initial Test Program Review

- Base U 14.1 Test Program A 14.1.1 Administrative Procedures (Testing)

A 14.1.2 Administrative Procedures (Modifications)

U 14.1.3 Test Objectives and Procedures U 14.1.4 Fuel Loading and Initial Operation A 14.1.5 Administrative Prceedures (System Operation)

U 14.2 Specific Information to be Included in Safety Analysis Reports U 14.2.1 Summary of Test Program and Objectives-GG 14.2.2 Organization and Staffing A 14.2.3 Test Procedures A 14.2.4 Conduct of Test Program P 14.2.5 Review, Evaluation and Approval of Test Results P 14.2.6 Test Records U 14.2.7 Conformance of Test Programs with Regulatory Guides P 14.2.8 Utilization of Reactor Operating and Testing Experiences in the Development of Test Program P 14.2.9 Trial Use of Plant Operating and Emergency Procedures GG 14.2.10 Initial Fuel Loading and Initial Criticality A 14.2.11 Test Program Schedule P 14.2.12 Individual Test Descriptions l

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CHAPTER 15 Accident Analyses Review Base _

P 15.0 General P 15.0.1 Analytical Objective P 15.0.2 Analytical Categories P 15.0.3 Event Evaluation P 15.0.4 Nuclear Safety Operational Analysis (NOSA)

Relationship 15.1 Decrease in Reactor Coolant Temperature P 15.1.1 Loss of Feedwater Heating P 15.1.2 Feedwater Controller Failure - Maximum Demand '

P 15.1.3 Pressure Regulator Failure - Open P 15.1.4 Inadvertent Safety / Relief Valve Opening P 15.1.5 Spectrum of. Steam System Piping Failures Inside and Outside of Containment in a PWR P 15.1.6 Inadvertent RHR Shutdown Cooling Operation 15.2 Increase in Reactor Presiiire P 15.2.1 Pressure Regulator Failure - Closed P 15.2.2 Generator Load Rejection P 15.2.3 Turbine Trip ~ ^

P 15.2.4 MSLIV Closures - --

P 15.2.5 Loss of Condenser Vacuum P 15.2.6 Loss of OffSite AC Power P 15.2.7 Loss of Feedwater Flow '

P 15.2.8 Feedsater Line Break P 15.2.9 Failure of RHR Shutdiwn Cooling 15.3 Decrease in Reactor Coolant System Flow Rate P 15.3.1 Recirculation Pump Trip P 15.3.2 Recirculation Flow Control Failure -

Decreasing Flow P 15.3.3 Recirculation Pump Seizure P 15.3.4 Recirculation Pump Shaft Break'

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15.4 Reactivity and Power Distributfon Anomalies P 15.4.1 Rod Withdrawal Error - Low. Powe'c '

P 15.4.2 Rod Withdrawal Error at Power P 15.4.3 Control Rod Maloperation (System halfunction or Operator Error) ..

P 15.4.4 Abnormal Startup of Idle Recirculation Pump P 15.4.5 Recirculation Flow Control with Increasing Flow

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Review Base N/A 15.4.6 Chemical and Volume Control System Malfunctions P 15.4.7 Misplaced Bundles Accident P 15.4.8 Spectrum of Rod Ejection Assemblies

. P 15.4.9 Control Rod Drop Accident 15.5 Increase in Reactor Coolant Inventory P 15.5.1 Inadvertent HPCS Startup N/A 15.5.2 Chemical Volume Control System Malfunction (for Operator Error)

P 15.5.3 BWR Transients Which Increase Reactor Coolant Inventory 15.6 Decrease in Reactor Coolant Inventory P 15.6.1 Inadvertent Safety / Relief Valve Opening N/A 15.6.2 Failure of Small Lines Carrying Primary Coolant Outside Containment N/A 15.6.3 Steam Generator Tube Failure P 15.6.4 Steam System Piping Break Outside Containment P 15.6.5 Loss-of-Coolant Accidents (Resulting from Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary) - Inside Containment 15.6.6 Feedwater Line Break-Outside Coittainment 15.7 Radioactive Release from Subsystems and Components P 15.7.1 Radioactive Waste System Leak or Failure P 15.7.2 Liquid Radioactive System Failure .

P 15.7.3 Postulated Radioactive Release, Due to Liquid Radwaste Tank Failure P 15.7.4 Fuel-Handling Accident l P 15.7.5 Spent Fuel Cask Drop Accidents U 15.8 ATWS U 15.9 Severe Accidents P App 15A Plant Nuclear Safety Operational Analysis P App'15B BWR/6 Generic Rod Withdrawal Error Analysis U App 15C Failure Modes and Effects Analysis

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CHAPTER 16 Standard Technical Specifications for General Electric Boiling Water Reactors Review Base GG 16.1 Definitions GG 16.1.1 Action GG 16.1.2 Average Planar Exposure GG 16.1.3 Average Planar Linear Heat Generation Rate GG 16.1.4 Channel Calibration GG 16.1.5 Channel Check GG 16.1.6 Channel Functional Test GG 16.1.7 Core Alteration GG 16.1.8 Critical Power Ratio GG 16.1.9 Dose Equivalent I-131 GG 16.1.10 E-Average Disintegration Energy GG 16.1.11 Emergency Core Cooling System (ECCS) Response Time GG 16.1.12 Frequency Notation GG 16.1.13 Identified Leakage GG 16.1.14 Isolation System Response Time GG 16.1.15 Limiting Control Rod Pattern GG 16.1.16 Linear Heat Generation Rate GG 16.1.17 Logic System Functional Test GG 16.1.18 Maximum Total Peaking Factor <

GG 16.1.19 Minimum Critical Power Ratio GG 16.1.20 Operable - Operability GG 16.1.21 Operationai Condition (Condition)

GG 16.1.22 Physics Test GG 16.1.23 Pressure Boundary Leakage GG 16.1.24 Primary Containment Integrity GG 16.1.25 Rated Thermal Power GG 16.1.26 Reactor Protection System Response Time GG 16.1.27 Recirculation Pump Trip System Response Time GG 16.1.28 Reportable Occurrence GG 16.1.29 Rod Density GG 16.1.30 Se.:ondary Containment Integrity GG 16.1.31 Shutdown Margin GG 16.1.32 Staggered Test Basis GG 16.1.33 Thermal Power GG 16.1.34 Total Peaking Factor GG 16.1.35 Unidentified Leakage 16.2 Safety Limits and Limiting Safety System Settings GG 16.2.1 Safety Limits GG 16.2.2 Limiting Safety System Settings rf:csc/167M25 2/12/82

Review l Base 16.82 Safety limits GG 16.B2.1 Bases GG 16.B2.2 Limiting Safety System Settings

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16.3/4 Limiting Conditions for Operation and Surveillance Requirements GG 16.3/4.0 Applicability

, GG 16.3/4.1 Surveillance Requirements GG 16.3/4.2 Power Distribution Limits GG 16.3/4.3 Instrumentation GG 16.3/4.4 Reactor Coolant System GG 16.3/4.5 Emergency Core Cooling Systems GG 16.3/4.6 Containment Systems GG 16.3/4.7 Plant Systems GG 16.3/4.8 Electrical Power Systems GG 16.3/4.9 Refueling Operations GG 16.3/4.10 Special Test Exceptions 16.B3/4.0 Applicability GG 16.83/4.1 Reactivity Control Systems GG 16.B3/4.2 Power Distribution Limits GG 16.B3/4.3 Instrumentation GG 16.B3/4.4 Reactor Coolant System GG 16.B3/4.5 Emergency Core Cooling System GG 16.83/4.6 Containment Systems GG 16.83/4.7 Plant Systems GG 16.83/4.8 Electrical Power Systems GG 16.B3/4.9 Refueling Operations GG 16.B3/4.10 Special Test Exceptions 16.5 Design Features A 16.5.1 Site GG 16.5.2 Containment j GG 16.5.3 Reactor Core GG 16.5.4 Reactor Coolant System A 16.5.5 Meteorological Tower Location GG 16.5.6 Fuel Storage GG 16.5.7 Component Cyclic or Transient Limit l

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j CHAPTER 17 Quality Assurance a ,

! Review Base U 17.1 Quality Assurance During Design and Construction  !

A 17.2 Quality Assurance During the Operating Phase i

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