Text

Interim SER Re Design & Const of Facilities
	ML20148C302
Person / Time
Site:	Sundesert
Issue date:	10/31/1978
From:	; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20148C271	List: ML20148C266; ML20148C302;
References
	NUREG-0469, NUREG-469, NUDOCS 7811020010
	Download: ML20148C302 (299)
	v • d • e

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1 NUREG 0469 I

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f k INTERIM l SAFE TY EVALUATION REPORT '

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1 related to the design and construction l

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SUNDESERT NUCLEAR PLANT, UNITS 1 & 2 l

SAN DIEGO GAS AND ELECTRIC COMPANY l

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Docket Nos. 50-582 & 50 583 i

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d Published: October 1978 ,

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h U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION f

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1 Available from National Technical Information Service

Springfield, Virginia 22161 Price: Printed Copy $11.75 Microfiche $3.00 l

l The price of this document for requesters outside of the North American Continent can be obtained from the National Technical Information Service.

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i NUREG-0469 October 16, 1978 4

lk'ILRIM SAFETY EVALUATION REPORT BY THE OFFICE OF NUCLEAR i:EACTOR REGULAi!ON U.S. NUCLEAR RlGUIATORY COMMISSION IN THE MATTER OF SAN DIEGO GAS AND ELECTRIC COMPANY SUNDESERT NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-582 AND 50-583 i

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TABLE Of CONTENTS PAGE 1.0 INFRODUCTION AND GENERAL DISCUSSION. . 1-1 1.1 Introduction. , 1-1 1.2 General Plant Description. . , 1-3

1. 3 Comparison With Similar f acility Designs. . , 1-5 1.4 Identification of Agents and Contractors. 1-5

1. 5 Summary of Principal Review Matters. . 1-6

1. 6 Facility Modifications-as a Result of Staff Review. 1-7

1. 7 Requirements for future Technical Information. 1-8 1.8 Outstanding items. 1-8 f.9 Generic Issues. . , . 1-12 1.10 Standard Review Plan. , , 1-12

2. 0 SITE CHARACTERISTICS. . . 2-1 2.1 Geography and Demography. 2-1 2.2 Nearby Industrial, Transportation, and Military facilities. . 2-6 2.3 Meteorology. . . 2-7 ,

l 2.3.1 Regional Climatology. ., . . 2-8 l 2.3.2 Local Meteorology. .. . 2-10 2.3.3 Onsite Meteorological Measurements Program. . , 2-11 2.3.4 Short-Term (Accident) Diffusion Conditions. 2-14 2.3.5 tong-Term (Routine) Diffusion Estimates. . 2-16 2.3.6 Conclusions. . 2-17 2.4 Hydrologic Engineering. .. . . . 2-17 2

2.4.1 Hydrologic Description. . . . 2-17 2.4.2 Flood Potential. . . , 2-19 2.4.3 Water Supply. , 2-19

) 2.4.4 Ground Water. . 2-20 2.4.5 Conclusions. . 2-21 2.5 Geology and Seismology. . . 2 21 2.5.1 Regional Geology. 2-21 2.5.2 Tectonic Province and Regional Tectonics. . , 2-25 1

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TABLE OF CONTENTS (Continued)

PAGE 2.5.3 Site Geology. 2-27 2.5.4 Surface faulting. 2-28 2.5.5 Regional Seismicity. . 2-28 2.5.6 Design Basis Earthquakes. 2-30 2.5.7 Conclusions. , 2-34

2. 6 Geotechnical Engineering. . 2-34 2.6.1 Stability of Subsurface M.nerials and Foundations. 2-34 2.6.2 Stability of Slopes. . 2-37 2.6.3 Embankments. . 2-38 1 2.6.4 Conclusions. . 2-38

3. 0 DESIGN CRITERIA FOR STRUCTURES, SYSTEM 9 AND COMP 0NENTS. 3-1 3.1 Conformance With General Design Criteria. 3-1 3.2 Classification of Structures, Systems and Components. . 3-1 3.2.1 Seismic Classification. . . 3-1 3.2.2 System Quality Group Classification. . 3-2 3.3 Wind and Tornado Design. . 33 3.4 Water Level (Flood) Design.. 3-4 3.5 Missile Protection. . . . 3-4 l

3.5.1 Missile Selection and Protection Criteria. .. 3-4 I l'

3.5.2 Barrier Design Procedures. . 3-7 3.6 Protection Against Dynamic Effects Associated With the Postulated Rupture of Piping. . , 3-7

3. 7 Seismic Design. . 3-9 3.8 Design of Seismic Category I Structures. . 3-9 4 3.9 Mechanical Systems and Components. . 3-10 3.9.1 Design Transients and Analytical Methods. 3-10 3.9.2 Dynamic Testing and Analysis. , 3-12 3.9.3 Loading Cnmbinations, Design Transients and Stress Limits. 4 3-14 !

3.9.4 Control Rod Drive Systems. . 3-17 !

3.9.5 Pump and Valve Operability Assurance Program. 3-18 l 3.9.6 Inservice Testing of Pumps and Valves. . 3-20 m

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LABLEOFCONTENTS(Continued)

PACE 3.10 Seismic Qualification of Seismic Category I Instrumentation and Electrical Equipment. .. . . .. .. . . . . .... . 3-20 3.11 Environmental Design and Qualification of Mechanical and Electrical Fquipment. . .. . 3-20 4.0 REACTOR.. . , , 4-1 4.1 Summary Description. . , . .

4-1 4.2 Mechanical Design. . , .. . ,, , . 4-I 4.2.1 Fuel., 4-1 4.2.2 Reactor Vessel Internals. . . . 4-8 4.2.3 Reactivity Control Systems. . . 4-10 4.3 Nuclear Design. . . . . . 4-12 4.3.1 Design Bases. , , . .. 4-13 4.3.2 Design Description. 4-13 4.3.3 Analytical Methods. . . . . .. . .. 4-17 4.3.4 Summary of Evaluation and conclusions. . . 4-17 4.4 Thermal and Hydraulic Design. . . . 4-18 i

4.4.1 Thermal-Hydraulic Design Criteria and Design Bases. . 4-18 j 4.4.2 Thermal-Hydraulic Analytical Methods. , 4-21 4.4.3 Thermal-Hydraulic Design Comparison. . .. . . . . 4-22 4.4.4 Conclusions. . . . ,, . . 4-24 5.0 REACTOR COOLANT SYSTEM. . . . . . . . 5-1 5.1 Summary Description. . . . . . . . . 5-1

) 5.2 Integrity of the Reactor Coolant Pressure Boundary.. . .... 5-1 ;

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5. 2.1 Coopliance With Codes and Code Cases. 5-1 l 5.2.2 Overpressurization Protection. .

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5. 2. 3 Reactor Coolant Pressure Boundary Materials. . 5-4 Inservice Inspection Program.. 5- 7 5.2.4 .. . . .. . .

I 5.2.5 Reactor Coolarit Pressure Boundary leakage Detection. . 5-7 l

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I TABLE OF CONTENTS (Continued)

PAGE 5.3 React'or Vessel. . . . . 5-10 5.3.1 Reactor Vessel Materials. . . 5-10 5.3.2 Pressure-Temperature Limit >. . , 5-11 5.3.3 Reactor Vessel Integrity. . . 5-12 5.4 Component and Subsystem Design. . . 5-13 5.4.1 Reactor Coolant Pumps. 5-13 5.4.2 Steam Generators. . . 5-15 l 5.4.3 Residual Heat Removal System. . , 5-18 I

5. 5 Loose Parts Monitoring System, 5-21 1

6.0 ENGINEERED SAFETY FEATURES. 6-1 I 6.1 Design Considerations. 6-1 6.2 Containment Systems. 6-1 6.2.1 Containment functional Design. 6-1 6.2.2 Containment Heat Removal Systems. 6-7 6.2.3 Containment Isolation System. . 6-9 i 6.2.4 Combustible Gas Control System. 6-12 6.2.5 Containment Leakage Testing System. 6-13

6. 3 Emergency Core Cooling System. ... .

. 6-13 6.3.1 Design Bases.. .. . 6-13 6.3.2 Design Description. . . . . 6 14

, 6.3.3 Design Evaluation. . . 6-16 6.3.4 Performance Evaluation. . , 6-19 f 6.3.5 Tests and Inspection. 6-21 6.3.6 Conclusions. . . . 6-22 6.4 Control Room Habitability. . 6-22 L

l 6.4.1 General Description. . . . 6-22 6.4.2 Radiation Protection Provisions. . . 6-23 6.4.3 Toxic Vapor Protection Provisions. .. 6-23 iv i -. . -_ _ _ . _ _ _ _ _ _ _ _ _ _ l

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-TABLE OF-CONTENTS (Continued)

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6.5 Engineered Safety Features Atmosphere Cleanup Systems. 6-25 6.5.1 Summary Description. 6-25 6.5.2 Annulus Building filtratien System. 6-25 6.5.3 Fuel Building Filtration System. . 6-26 6.5.4 Control Room Standby Filtration System. 6-26 6.5.5 Containment Spray System. . 6-27 6.6 Inservice inspection of Engineered Safety Features. 6-28 6.7 Engineered Safety Features Metallic Materials. 6-29 5.8 Organic Materials. .

6-29 7.0 INSTRUMENTATION AND CONTROLS. . . 7-1 8.0 ELECTRIC POWER SYSTEMS. 8-1 8.1 General Discussion. 8-1

8. 2 Offsite Power System. 8-l 8.3 (nsite Power Systems. , , 8-3 8.3.' Alternating Current Power System. . 8-3 8.3.2 Direct Current Power System. . 8-4 8-6 !

8.3.3 Physical independence of Electrical Power Systems..

8.3.4 Electrical Penetrations. , . . 8-6 8.3.5 Environmental Qualifications of Class lE Equipment. . 8-7 l

9.0 AUXILIARY SYSTEMS. . . . . . . 9-1 1

l 9.1 Fuel Storage and Handling. . 9-2 ,

9.1.1 New and Spent fuel $torage. . . . 9-2 ;

y 9.1.2 Fuel Pool Cooling and Purification Systems. 9-2 9.1.3 Fuel Handling System. . 9-3

9. 2 Water Systems. . . .. . . 9-4 9.2.1 Service Water System. 9-4 9.2.2 Reactor Plant Component Cooling Water System. 9-5 9.2.3 Ultimate Heat Sink. , 9-7 9.2.4 Condensate Storage Facilities. .

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j 9.3 Process Auxiliaries. 9-8 4 9.3.1 Compressed Air Systems. . . 9-8 ,

9.3.2 Process Sampling System. . 9-9 I

9.3.3 Vent and Drain Systems. 9-9 l 9.3.4 Chemical and Volume Control System. 9-10 i

t j 9.4 Air Conditioning, Heating, Cooling and Ventilation Systems. 9-11 1

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9.4.1 Control Building Heating, Ventilation and Air Conditioning j Systems. . . . . 9-11 l 9.4.2 Fuel Building Ventilation System. 9-12

! 9.4.3 Annulus Building ventilation System. .. . . 9-13 1 i i j 9.4.4 Standby Diesel Generator Butiding Ventilation System. 9-13 I 9.4.5 Service Water Pump House tentilation System. . 9-14

} 9.5 Other Auxiliary Systems. 9-14

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9.5.1 Fire Protection System. . 9-14 9.5.2 Communication System. 9-14 9.5.3 Lighting System. 9-15 f

9.5.4 Diesel Generator Auxiliary Systems. 9-15 ,

i 10.0 STEAM AND POWER CONVERSION SYSTEM. 10-1 i

k i 10.1 Summary Description. . .. . 10-1 1

j 10.2 Turbine-Generator. . 10-1

{ 10.3 Main Steam Supply System. , . 10-2 j 10.4 Other Features of Steam and Power Conversion System. 10-3 i 10.4.1 Main Condenser System. . 10-3 10.4.2 Main Condenser Evacuation System.. . 10-3 10.4.3 Turbine Gland Sealing System. .. .< . 10-4 j 10.4.4 Turbine Bypass System. 10-4 j 10.4.5 Circulating Water System. . 10-5 j 10.4.6 Condensate and Feedwater Systems. . . 10-5

10.4.7 Auxiliary Feedwater System. . . 10-6 4

10.5 Steam and Feedwater System Materials. . . 10-7 4

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TABLE OF CONTENTS (Continued)

PAGE 11.0 RADI0 ACTIVE WASTE MANAGEMENT. .. , . 11-1 11.1 Summary Description. ,

11-1 11.2 Radioactive Waste Treatment System Description and Evaluation. . 11-7 11.2.1 Liquid Radioactive Was',e Treatment Systems. . .. 11-7 11.2.2 Gaseous Radioactive Waste Treatment Systems. 11-11 11.2.3 Solid Radioactive Waste Treatment Systems. 11-16 11.3 Process and Effluent Radiological Monitoring Systems. , 11-18 ,

11.4 Conclusions. .

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12-1 12.0 RADIATION PROTECTION. . .

12.1 Assuring That Occupational Radiation Exposures Are As Low As Is Reasonably Achievable. .

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12.2 Radiation Sources. .

12.3 Radiation Protection Design Features.. . 12-3 !

12.4 Dose Assessment. . . . . 12-5 l 12-6 I 12.5 Health Physics Program.

13.0 CONDUCT OF OPERATIONS.. ,

13-1 13.1 Organizational Structure of Applicant. . .. 13-1 13.2 Training Program. . .

. 13-2 13.3 Emergency Planning. . .. . 13-3 13.4 Review and Audit. . . . . . . 13-8 13.5 Plant Procedures. . . . . . 13-8 13.6 Industrial Security. . . .. . 13-8 14-1 14.0 INITIAL TEST PROGRAMS. . . . . .

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15-1 15.0 ACCIDENT ANALYSES. . . .. . .

15.1 General. . . . . . 15-1

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Classification of Events... 15-1 1$.1.1 ... . . . .

Input Parameters for Transient and Accident Analyses.. 15-1 15.1.2 15.1.3 Analytical Techniques. .. .. .. 15-5 vii

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PAGE 15.1 Moderate frequency Transients. .

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15.2.1 Increase in Heat Removal by the Secondary System. 15-8 15.2.2 Decrease in Heat Removal tr the Secondary System. 15-8 15.2.3 Decrease in Reactor Cocle System flow Rate. . 15-9 15.2.4 Core Reactivity Insertion. . 15-9 15.2.5 Decrease in Reactor Coolant Inventory. . 15-11 15.2.6 Increase in Reactor Coolant Inventory. ., 15-11 15.2.7 Rod Cluster Control Assembly Misalignment. . 15-11 15.2.8 Summary and Conclur:ons. 15-12 l

15.3 Infrequent Incidents and Postulated Accidents. . , 15-12 15.3.1 Feedwater system Piping Breaks. 15-13 15.3.2 Spectrum of Steam Piping failures Outside of Containment. 15-14 15.3.3 Reactor Coolant Pump Rotor Seizure. 15-15 15.3.4 Spectrum of Piping Breaks Within the Reactor Coolant Pressure Boundary., .

. 15-15 15.3.5 Inadvertent Loading of a fuel Assembly Into an Improper Position. .

15-15 l 15.3.6 Rupture of a Control Rod Drive Mechanism Housing (Rod Cluster Control Assembly Ejection). 15-16 15.3.7 Summary and Conclusions. )

15-17 15.4 Anticipated Transients Without Scram. . 15-17 15.5 Radiological Consequences of Accidents. 15-18 '

15.5.1 Loss-of-Coolant Accident. ..

15-20 15.5.2 Fuel Handling Accident. . . 15-20 15.5.3 Cask Drop Accident. 15-24 15.5.4 Rod Ejection Accident. . .. . .

15-24 15.5.5 Postulated Radioactive Releases Due to Liquid Tank Failures. 15-27 1 16.0 TECHNICAL SPECIFICATIONS. . , .. 16-1 17.0 QUALITY ASSURANCE. .

17-1 17.1 General. . . ,. ... .. 17-1, 17.2 San Diego Gas and Ele-tric Company. . . . . 17-1 I 17.3 Westinghouse Electr'c Corporation. . . .. 17-6 viii e

TABLE OF CONTENTS (Continued)

PAGE 17.4 Stone i Webster Engineering Corporation. .

17-9 17.5 Implementation of the Quality Assurance Program. . . 17-12 17.6 Conclusions... , . . .

17-12 18.0 REVIEW BY THE ADVISORY COMMITTEE ON REACTOR SA;EGUARDS. . . 18-1 19-1 19.0 COMMON DEFENSE AND SECURITY.. . .

.. 20-1 20.0 FINANCIAL QUALIFICATIONS. .

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21.0 CONCLUSION

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APPEN01 A - CHRONOLOGY OF RADIOLOGICAL SAFCTY REVIEW OF SUNDEstns dVCLEAR PLANT, A-1 UNil5 1 AND 2. . .

B-1 8 - BIBl.10 GRAPHY FOR THE SUNDESERT PLANT SAFETY EVALUATION REPORT.

. C-1 0 - ADVISORY COMMITTEE ON REACTOR SAFEGUARDS - GENERIC MATTERS.

D 'NTERIM TECHNICAL POSITION, FUNCTIONAL CAPABILITY OF PASSIVE PIPING

, D-1 COMPONENTS.

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LIST Of TABLES j

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2.1 ESTIMATED POPULATION DISTRIBUTION WITHIN A 50 MILE RADIUS OF THE SUNDE5ERT SITE (PERMANENT RESIDENTS) 1985. 2-5 1

2.2 SUNDESERT METEOROLOGICAL INSTRUMENTATION. 2-13

, 2. 3 5HORT-TERM RELATIVE CONCENTRATION VALUES BY DOWNWIND l j DIRECTION - SUNDE5ERT SITE. 2-15 l

l 2. 4 SHORT-TERM RELATIVE CONCENTRATION VALUES USED FOR !

j ACCIDENT ANALYSIS - SUNDESERT SITE. 2-15 l l 2. 5

SUMMARY

OF LONG-TERM RELATIVE CONCENTRATION AND RELATIVE DISPOSITION VALUES FOR SELECTED LOCATIONS NEAR THE SUNDESERT SITE. . 2-18

! 4.1 FUEL MECHANICAL DESIGN COMPARISON. . 4-2 i

j 4.2 RANGE OF DESIGN PARAMETER EXPERIENCE. . . 4-5

! l 4.3 THERMAL-HYDRAULIC DESIGN COMPARISDN. 4-23 a

j 4.4 FUEL DESIGN COMPARISON. . 4-23 J

6.1 SUBCOMPARTMENT PRES 5URE DIFFERENTIALS. 6-6 i

9.1 PROCESS SYSTEM SAMPLE LOCATIONS. 9-10 11.1 PRINCIPAL PARtMETERS USED IN ESTIMATING RELEASES OF RADIOLOGICAL MATERIAL IN EFFLUENTS FROM SUNDESERT NUCLEAR PLANT, l

) UN'T n % 1 *:ib 2. , . 11-3 J

! 11.2 CALCULATED RELEASES OF RADI0 ACTIVE MATERIALS IN GASEOUS l EFFLVENTS FROM SUNDE5ERT NUCLEAR PLANT, UNIT N05, 1 AND 2. 11-4 11.3 U CULATED DOSES TO A MAXIMUM INDIVIOUAL AND THE 50-MILE 400LATION FROM SUNDESERT NUCLEAR PLANT UNIT N05. 1 AND 2. 11-5 11.4 PRINCIPAL PARAMETERS USED IN THE COST-BENEFIT ANALYSIS FOR SUNDE5ERT NUCLEAR PLANT, UNIT N05. 1 AND 2. . . 11-6 X

LISTOFTABLES(Continuey PAGE a

11.5 DESIGN PARAMETERS OF PRINCIPAL COMPONENTS CONSIDERED IN THE EVALUATION OF LIQUID AND GASEOUS RADIDACTIVE WASTE TREATMENT SYSTEMS. . . .

11-10 11.6 PROCESS AND EFFLUENT MONITORING LOCATIONS.

11-19' 13.1 ORGANIZATIONS AND AGENCIES - EMERGENCY PLANNING FOR THE SUNDESERT NUCLEAR FACILITY. 13-5

'5.1 CATEGCRIES OF TYPICAL TRANSIENTS AND FAUL15. 15-2 15,2 TRIP POINTS AND TIME DELAYS TO TRIP ASSUMED IN ACCIDENT ANALYSES. 15-3 1

15.3 CALCULATED RADIOLOGICAL CONSEQUENCES OF DESIGN BASIS ACCIDENTS. 15-19 15.4 LOSS-OF-COOLANT ACCIDENT ASSUMPTIONS AND INPUT PARAMETERS. 15-21 15.5 FUEL HANDLING ACCIDENT ASSUMPTIONS AND INPUT PARAMETERS. 15-23 15.6 CASK DROP ACCIDENT ASSUMPTIONS AND ]NPUT PARAMETERS. . . 15-25 15.7 ROD EJECTION ACCICENT ASSUMPTIONS AND ]NPUT PARAMETERS. 15-26 l

17.I REGULATORY GUIDANCE APPLICABLE TO QUALITY ASSURANCE PROGRAMS. 17-2 l

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I PAGE 1.1 GENERAL SITE LOCATION, SUNDESERT SITE. 1-2 2.1 AREA WITHIN 10 MILES OF THE SITE. . . 2-2 I

2.2 LAND OWNERSHIP - SUNDESERT SITE. . 2-3 2.3 DIRECTIONAL FREQUENCY OF WIND - SUNDESERT SITE. 2-12 11.1 LIQUID WASTE TREATMENT SYSTEMS, SUt: DESERT NUCLEAR PLANT. 11-8 11.2 GASEOUS WASTE TREATMENT SYSTEMS, SUf!DESERI NUCLEAR PLANT. 11-13 17.1 SAN DIEGO GAS AND ELECTRIC COMPANY ORGANIZATION. 17-3 17.2 SAN DIEGO GAS AND ELECTRIC COMPANY NUCLEAR POWER PLANT PROJECT ORGANIZAlION. 17-5 17.3 WESTINGHOUSE NUCLEAR ENERGY SYSTEMS (NES)

ORGANIZATION (WATER REACTORS). 17-7 i

17.4 STONE & WEBSTER ENGINEERING CORPORAlION ORGANIZATION FOR QUALITY ASSURANCE. 17-10 i

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1.0 IN1R00VCT10N AND GENERAL OlSCUSSION

1. I Introduction The San Diego Gas and Electric Company (applicant) filed with the Nuclear Regulatory Commission (Commission) an application docketed on April 12, 1977, for licenses to construct and operate its proposeo Sunde3ert Nuclear Plant, Units 1 and 2 (plant or f acility) on a site located in Riverside County, California, as shown in Figure I. I, approximately 16 miles southwest cf Elythe, California.

Prior to the submittal of the application, the applicant had requested that the Commission's staf f, the U.S. Geological Survey and the Advisory Committee on Reactor Safeguards review several matters (as discussed in Section 2.0 of this report),

relating to the suitability of the proposed site. Information on these matters was presented in a report, "Early Site Review Report," submitted by the applicant on April 16, 1975, for our review. The results of our evaluation of these matters and the evaluation by the U.S. Geological Survey are presented in NUREG-0171, "Early {

Site Review Report for the Sundesert Site," dated February 10, 1977. The results of the review by the Advisory Committee on Reactor Safeguards are provided in letters, dated March 16, 1977 and May 6, 1977, which are included in Supplement No. 1.

dated June 16, 1977, to NUREG-0171.

A Preliminary Safety Analysis Report was submitted with the application and included, as Appendix A, the above cited "Early Site Review Report," which we had previously reviewed. The information in the Preliminary Safety Analysis Report has been supplemented by Amendment Nos. I through 14 as a result of changes initiated by the applicant and also due to requests for information which we made during our review of the document. The Preliminary Safety Analysis Report and its amendments are available for public inspection at the U.S. Nuclear Regulatory Commission Public Document Room, 1717 H Street, Washington, D.C. 20555; at the Palo Verde Valley District Library,125 West Chanslorway, Blythe, California 92555; and at the San l l

Oiego County Law Library,1105 Front Street, San Diego, California 92101.

During tha ccoit' cf our review, the applicant advised us by letter, dated May 2, 3

1978, that due to actions taken by the State of California, work on the proposed Sundesert facility was beinc cuspended except for those steps necessary to reserve the site and water supply fc 'uture use in meeting the electric energy needs of Southern California. In the letter, the applicant requested that we complete our evaluation of the Preliminary Safety Analysis Report to the maximum extent possible, 1 recognizing that, in some areas of the review, it will not be possible for the 1

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Staff to reach conclusions. Following receipt of this information we terminated our review efforts, except for that required to document the status of our review to date.

The purpose of this interim Safety Evaluation Report (report) is to summarize the results of the technical evaluation of the proposed Sundesert plant performed by the Commission's staff to date and to delineate the scope of the technical matters considered in evaluating the radiological safety aspects of the Sundesert f acility.

Whefe applicable, this evaluation incorporates the evaluation of the site-related matters presented in NUREG-0171 and its supplement as discussed above. Aspects of the environmental impact considere<f in the review of the Sundesert fac-ility, in accordance with 10 CFR Part 51 of the Commission's regulations, will be discussed

in the Commission's Final Environmental statement which is expected to be issued during October 1978.

Section 1.8 of this report summarizes the outstanding issues which require resolu-tion before we can compit;e our review of the radiological safety aspects of the Sundesert facility. Should the applicant resume work on the Sundesert facility and requests us to complete our review of the application, we will do so Consistent with the licensing requirements in effect at that time, which will include resolu-tion of these outstanding matters as well as any new safety significant considera-tions that develop in the interim. At that time our complete review will be provided in a future Safety Evaluation Report.

I Due to the applicant's suspension of work on Sundesert, the Advisory Committee on Reactor Safeguards is not currently reviewing the application. If the applicant subsequently requests us to complete our review, the Committee will be requested to review the application.

Appendix A to this report provides a chronology of our principal actions related to the processing of he radiological safety aspects of the application to datd, Appendix B is a listing of the bibliography used in our review.

1.2 General Plant Description Each of the two units of the Sundesert plant will incorporate a closed-cycle,

) pressurized water nuclear steam supply system, a tandem compound turbine generator, er.gineered safety features, radioactive waste systems, a fuel handling system, structures, other onsite f acilities, and the necessary auxiliaries required for a 7

complete and operable nuclear power plant.

Each nuclear steam supply system (to be supplied by the Wes'inghouse Electric Corporation) will consist of a three-loop reactor coolant syJtem designed for a core power output of 2775 thermal megmtts. The reactor cort will be composed of uranium dioxide pellets enclosed in Zircaloy tubes with welded e N plugs. The fuel 1-3

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] tubes will be grouped and supported in assemblies. The reactor core will in hially ;

l consistofthreeregionseachcontairtingfuelofadifferentenrichmentofuranium-[35. ,

) Water will serve as both the moderator and the coolant and will be circulated l I

i through the reactor vessel and ctre Dv wee coolant pumns. l l

{ The water, heated by the reactor., will flow through three steam generators where j heat will be transferred to the stcondary (steam) system. The water will then flow j back to the pumps to repeat the cycle. An electrically-heated pressurizer, to be j connected to the hot leg of one of the reactor coolant loops, will establish and

] maintain tha reactor coolant pressure, and will provide a surge chamber and a water j reserve to accommodate reactor coolant volume changes during operation, i The reactor power level will be regulateo oy control rod movement. Requirements I

for slower changes in the core reactivity will be achieved by changing the boric j acid concentration in the reactor coolent. A reactor protection system will be provided to automatically initiate appropriate corrective action whenever a plant condition monitored by the system approe-hes preestablished limits. The reactor protection system and an engineered safety features actuation system will act to shut down the reactor, close isolation valves, and initiate operation of the engi-neered safety features should any or all of these actions be required.

i i

The facility cesign for each unit wili include systems and design features which

). are previded to function during or fcilowing a postulated accident to prevent or '

j reduce the release of fission products from the facility. These engineered safety features will be designed to retain, within acceptable leakage limits, the fission f products which might be released from the reactor fuel, to mitigate e damage to l

the fuel cladding and other barriers, to provide for the protection of the public j and station personnel, and to provije for the removal and cleanup of fission l products within the plant struttures.

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i The containment building will consist of a re;nforced con: rete, steel lined struc-ture which will house the nuclear steam supply system and portions of the engi-neered safety features systems. The containment building will be provided with spray and filtration systems to limit the containment pressure, and to remove and

retain, within ccntainment leakage limits, radioactive material that might l

} otherwise be released in the unlikely event of an accident.

i

{ The annulus building, to be constructed on the same continuous mat as the contain-ment building, will surround the containment building along its entire periphery, The annulus building will housa portions of the engineered safety features systems,

{ auxiliary systems, the new and spent fuel storage facilities, radioactive liquid

and gaseous waste systems, and other equipment. Each of the trains of the engi-neered safety features systems will be physically separated from each other by distance and physical barriers.

1-4

The steam and power conversion system will De designed to utilize the steam produced in the three steam generators, converting it into electrical energy by means of the steam turbine generator to be locatec in the turbine build ing. The service water system will remove waste neat f rom tne main steam condenser to the natural draf t cooling tower where it will be dissirated to the atmosphere.

The service water system also will provide Cooling water for ali Componcnts neces-sary for the engineered safety features in the unlikely event of an acciaent. The system for each unit will consist of two independent trains, each dissipating the neat to the atmosphere by means of a mechanical draf t cooling tower. The basin in each cooling tower will have an adequate water supply to maintain both units of the plant in a safe shutdown condition fcr 30 days witnout replenishment.

The emergency core co: ling system will consist of accumulator tarAs and high pres-sure injection and lo+ pressure injection systems, with provisions for recircu-lation of the borated ,colant af ter the end of the injection phase of a postulated loss-of-coolant accider Various combinations of these systems will assure core cooling for the complete range of postulated coolant pipe break sizes.

The facility will be provioed with electrical power from two independent offsite power sources and independent and recundant onsite emergency diesel power supplies capable of supplying power to the engineered safety features system buses.

i 1.3 Comparison With Similar Facility Desi g l

l fhe principal design features of the Sundesert plant are similar to those we have previously evaluated and approved on other nuclear power plants. To the extent feasible and appropriate, we have made use of our previous evaluations of these similar plants in conducting our review of this facility. Where this was done, the appropriate sections of this report identify the other facilities involved.

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! The nuclear steam supply system for this f acility is similar to other nuclear steam supply systems furnished by the Westinghouse Electric Corporation. These include the Koshkonong Nuclear Plant, Unit Nos. I and 2 (Docket Nos. STN 50-502 and STN 50-503) 1 l

and the Beaver Valley Power Station, Unit No. 1 (Docket No. 50-334).

I I

k The balance-of plant design by the Stone & Webster Engineering Corporation includes structures and systems whose layout and design are similar to those furnished for the Stone & Webster Standard Reference Nuclear Power Plant (Docket No. STN 50-495)

)

for the Westinghouse nuclear steam supply system.

1.4 Identification of Agents and Contractor.s The San Diego Gas and Electric Company is the applicant for the construction permit application for the Sundesert plant, and is responsible for the design, construction and operation of the facility.

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j lhe applicant had contracted with several crganizations to provide design, manuf ac-turing, construction and consulting services in support of the proposed plant. As i

i, stated in Section 1.1 of this report, the applicant has suspended all work on the j plant, except for thnse steps neM asary to reserve the site and water supply for f future use. The status of the contr& cts with the organizations invohed as a j result of this suspension has not been provided by the applicant. Prior to the f

f suspension, the contracted organizations had the following responsibilities, as stated in the Preliminary Saf ety Analysis Report.

The applicant engaged the Westinghouse Electric Corporation to design, manufacture 1 and deliver to the site the nuclear Steam supply system and the turbine generator.

l.

j The applicant selected the Stone & Webster Engineering Corporation to design the ;

f acility outside the sccpe of the nuclear steam supply system, and to manage the i 3

i constraction of the f3cility, i

4

] The applicant also employed consultarts in certain specializec areas; for example, j

! fugro, Incorporated, for the geotechnical aspects of nuclear power plant siting, j EDS Nuclear, Incorporated, for seismic analysis, quality assurance programs and ,

j testing programs, Pickard, Lowe and Garrick, jncorporated, for fuel performance and l l other technical disciplines, and Climet Instrument Company for meteorological

{

f monitoring systems. I i

4

. l.5 Summary of Principal Review Matters 3 Our technical review and evaluation of the information submitted by the applicant j considered the principal matiers summarized below, although, because our current j review efforts have been suspended, not all aspects of these matters have been i completed:

? l 1 (1) We eval o ted the population censity and land use characteristics of the site f -environs and the physical characteristics of the site (including seismology,

?

meteorology, geology, and hydrology) to establish that these characteristics I have been determined adequately and have been given appropriate consideration

{ in the plant design, and that the site characteristics are in accordance with j the Commission's siting criteria (10 CFR Part 100), taking into consideration j the design of the f acility, including the engineered safety features provided.

t (2) We have evaluated the design, fabrication, construction and testing criteria, and expected performance characteristics of the plant structures, systems, and components important to safety to determine that they are in accord with the i

Commission's General Des'ign Criteria, Quality Assurance Criteria, Regulatory l

Guides, and other appropriate rules, codes and standards, and that any departures I

from these criteria, codes and standar:t have been identified and justified.

4 1

16 i

S

l (3) We evaluated the expected response of the facility to various anticipated operating transients and to a broad spectrum of postulated accidents. Based on this evaluation, we determined that the potential consequences of a few highly unlikely postulated accidents (design basis accidents) would exceed those of all other accidents confidered. We perfo.med conservative analyses of these design basis accidents to determine that the calculated potential offsite radiation doses that might result in the very unlikely event of their occurrence would not exceed the Coa. mission's guidelines for site acceptability given in 10 CFR Part 100.

(4) We evaluated the applicant's engineerirg and construction organization, plans for the conduct of plant operations (including the organizational structare and the general qualifications et operating and technical support personnel),

the plans for industrial security, and the planning for emergency actions to be taken in the unlikely evert of an accident that might af fect the general public, to determine that the applicant will be technically qualified to safely operate the facility.

(5) We evaluated the design of the systems provided for control of the radio-logical effluents from the facility to determine that these systems will be capable of controlling the release of radioactive wastes from the facility within the limits of the Commission's regulations (10 CFR Part 20), and that equipment to be provided will be capable of being operated by the applicant in such a manner as to reduce radioactive releases to levels that are as low as reasonably achievable within the context of the Commission's regulations (10 CFR Part 50), and to meet the dose design objectives of Appendix 1 to 10 CFR Part 50.

(6) We evaluated the applicant's quality assurance program for the design and I I

construction of the facility to assure that the program complies with the Commission's regulations (10 CFR Part 50) and that the applicant will have proper controls over the facility design and construction such that there will be a high degree of assurance that, when completed, the facility can be operated safely and reliably.

The applicant also p'ovidad financial data ard information as required by the i Commission's retul,tions (Section 50.33(f) of 10 CFR Part 50 and Appendix C to 10 CFR Part 50). To assure that we have the latest information to make a determi-nation of the financial qualifications of an applicant, it is our current practice to review this information during the later stages of our review of an application.

1.6 Facility Modifications as a Result of Staff Review During the review of the application for the Sundesert plant, numerous meetings were held with the applicant's representatives, its contractors and consultants to 1-7

I-i i

! discuss the proposed facility and the technical material submitted. A chronological I- listing of the meetings and other sicnificant events is given in Appendix A to this.

report. During the course of the review, tne applicant proposed or we requested a number of technical and administrativt change- These ate described in various f amendments to the original application, and are discussed in appropriate sections

. of this report. '

l.7 Requirements for Future Technical Infortration I

j The applicant identified, in Section 1.5 of the Preliminary Safety Analysis Report, ongoing research and development programs by Westinghouse which are directed toward g confirming the design adequacy of certain components in the nuclear steam supply j system. The objectives and description of these programs are provided by reference

} to Westinghouse Topical Reports submitted for our review. These test proorams, as l delineated in the topical reports, are discussed further in Sections 4.2 and 4.4 of j th_is report.

4 Based on our review of the verification programs, we have determined that the j inferrr.ation to be obtained from these test programs is of the type that, in accord-a ance with the provisions of Section 50.35 of 10 CFR Part 50, can be left for later

consideration and may be supplied in a Final Safety Analysis Report.

l 1.8 Outstanding Items 2

l As a result of our review to date, we have identified certain outstanding issues where we (1) have not completed our review, (2) need additional information from the applicant to complete our review, or (3) have established positions to which 4

the applicant has not yet committed. These items are summarized below and are

discussed further in the indicated sections of this report. As stated in Section 1.1

, of this report, should the applicant resume work on the Sundesert facility and j requests us to complete our review of the application, we will do so consistent j with the licensing requirements in effect at that time, which will include resolution j of these outstanding matters, as well as any new safety significant considerations i that develop in the interim.

I (1) We cannot conclude that the applicant has the authority to determine all activities within the exclusion area until it has acquired, from the Bureau of ;

I land Management, the land within the exclusion area (Section 2.1).

i (2) We have not completed our review of the minimum winter temperature proposed by l the applicant for general plant design (Section 2.3.1).

3 3

e j (3) We require that the applicant commit to geologically map site excavations for

] seismic Category I structures and for seismic Category I buried pipelines (Section 2.5.3),

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

(4) We require that the applicant revise its compaction criteria for structural ,

backfill beneath seismic Category I structures to also include the criteria from ASTM Standard 0-2049 (Section 2.6.1).

(5) We require that the fuel hvilding and the building filtration system and isolation dampers t4e designed as seismic Category 1 (Sections 3.2.1, 6.5.3, 9.4.2 and 15.5. 2).

P (6) We have not completed our review of the design procedures for combining wind and tornado loads with other applicable loads (Section 3.3).

(7) We require that the applicant postulate and evaluate potential missiles, within areas of the facility locateo outrice containment, resulting from overspeed of rotating machinery and from failure of nressurized components and systems to assure the adequacy of plant separation feat.cas and barrier design (Section 3.5.1), l i

(8) We have not completed our review of the barrier design procedures for protec- ]

tion against postulated missiles (Section 3.5.2). i I

' (9) We require that the piping system downstream of the turbine-driven auxillary feedwater pump discharge be considered as a high energy system and that protec-tion against postulated pipe breaks in this system be provided accoroingly ,

(section 3.6).

(10) We have not completed our review of the seismic design analysis procedures (Section 3.7).

(11) We nave not completed our review of the design criteria for seismic Category I structuret (Section 3.8).

(12) We require additional information from the applicant to justify the exceptions taken to Regulatory Guide 1.124 regarding the critical buckling strength for l

plate type supports (section 3.9.1).

(13) We require that the load effects from applicable dynamic loads be combined by

.f the absolute sum method for the design of mechanical systems and components (Section 3.9.3).

' (14) We require additional information to describe how the f unctional capability of all piping essential to the safe shutdown of the proposed plant will be demon-strated (Section 3.9,3),

(15) We require that the inservice testing program for pumps and valves comply with

'10 CFR 50.55a(g)(3) with respect to the selection of the earliest applicable l

edition and addenda (Section 3.9.7).

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)

, (16) We have not completed our review of the proposed environmental qualification of Class lE equipment (section 3.11).

a 1

i (17) We require additional information trum the applicant tn complete our review of l the overpressurization protection system for the case when the reactor coolant

! system is water solid (Secticn 5.2.2).

I 6

l (18) We require that the test program for high strength ferritic materials br j performed in conformance with Note 1 of ASME Code Case 1528-3 (Section 5.2.3). I I

e l (19) We require that each leakage detection syt, tem for the reactor coolant pressure l boundary be set to alarm in the control room on an increase in leakage of one j gallon per minute above background levels (Section 5.2.5).

t j (20) We require that the calibration of the sump monitoring system include the 4

capability for performing system operability checks with background leakage i for use in calibrating the radiation monitoring systems ($ection 5.2.5). I l

j (21) We require that the applicant provide information regarding the criteria that i f will be used for plugging degraced steam generator tubes (Section 5.4.2). l t

(22) We have not completed our revie. of the aspects of the residual heat removal 1

4 system dealing with (a) overpressurization protection, and (b) maintaining piping integrity in the event of an excessive cooldown rate during initial j reactor shutdown (Section 5.4.3).

I 1

1 (23) We require that the applicant assess the capability of achieving a cold shutdown 1

l condition using only safety grade systems, assuming that only offsite or j onsite power is available, and also assuming the most limiting single failure

{ (Section 5.4.3).

i !

l (24) We have not completed our confirmatory analysis of the containment response to l a postulated main steam line break (Section 6.2.1).

E j (25) We require additional information from the applicant regarding the calculated I peak pressure for the reactor cavity subcompartment due to a postulated loss of-4 coolant accident, to complete our review of the design of this subcompartment (Section 6.2.1).

(26) We require additional information from the applicant, in the areas of subcom-partment nodalization and the transient loadir.g on major components, to deter-

, mine the acceptability of the subcompartment analysis for use in the design of major component supports (Section 6.2.1).

1-10 i

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t (27) We require additional informatien from the applicant to justify its calculated hydrogen evolution rate used in the evaluation of the hydrogen recombiners (Section 6.2.4).

(28) We require that the lock-out feature, to prevent the spurious movement of certain motor-operated valves in the emergency core cooling system, be included for ene additional valve, Valve 1-8885 (Section 6.3.3). l (29) We require that position indication be provided in the control room to indicate bypass or inoperable status of critical manual valves in the emergency core cooling system (Section 6.3.3).

i (30) We have not completed our review of the analyses of postulated small break loss-of-coolant accidents (Section 6.3.4).

(31) We require that a modeling error, pertaining to the heat generation rate from metal / water reaction, be

solved for the emergency core cooling system per--

formance evaluation (Section 6.3.4).

(32) We have not completed our review of the instrumentation and control systems (Section 7.0).

(33) We require that the applicant meet the requirements of IEEE 5tandard 279-1971 with regard to the protection of electrical penetrations (Section 8.3.4).

(34) We require that physical restrictions be provided for the fuel Cask lifting devices to assure that the maximum Dotential for a cask drop is 30 feet (Section 9.1.3).

(35) We require that the reactor polar crane be designed to meet the single j failure criterion or that the applicant demonstrate by analysis that the consequences of a postulated drop of the reactor vessel head by the crane are !

acceptable (Section 9.1.3).

1 1

(36) We require that the component cooling water system be designed to prevent fuel damage or damage to the reacto 'lant system pressure boundary due to a postulated extended loss of cooi , to the reactor coolant pumps (Section 9.2.2).

(37) We have not completed our review of the fire protection system (Section 9.5.1).

3 (38) We require that the applicant provide the results of a reanalysis of the postulated feedwater line break accident by considering a Model F steam gener-ator without a feedwater flow restrictor, as proposed for Sundesert (Section 15.3.1).

1-11 i

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(39) We require that the applicant assess the sensitivity of operator action time, in terminating a postulated steam line break, on the long-term cooling effects on, and integrity of, the reactor pressure vessel (Section 15.3.2).

l (40) We require that the applicant demonstrate that the construction portion of the /

quality assurance program is being implemented in accordance with acceptable commitments (Section 17.5).

(41) We have not completed our review of the applicant's financial qualifications (Section 20.0).

1.9 Generic Issues The Advisory Committee on Reactor Safeguards periodically issues a report listing various generic matters applicable to light water reactors. Our discussion of these matters is provided in Appendix C to this report which includes references to sections of this report for more specific discussions that particularize the generic status for the proposed facility.

On January 1, 1978, the Commission's Office of Nuclear Reactor Regulation issued a report, NUREG-0410. "NRC Program fnr the Resolution of Generic Issues Related to Nuclear Power Plants," which discusses the status of the ongoing efforts for resolving the generic issues identified by the Advisory Committee on Reactor Safeguards as well as other generic issues identified by the Commission's staf f. While many of the other generic issues discussed in NUREG-0410 were considered during our review of Sundesert in accordance with our current review procedures, no specific attempt has been made in this Sundesert interim Safety Evaluation Report to assess the generic status of these other issues for the proposed facility. Should the review of the Sundesert application be reactivated, these other generic matters will be specifically considered. !

1.10 Standard Review Plan The Commission's Office of Nuclear Reactor Regulation issued a document NUREG-75/087,

" Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants" (Standard Review Plon),-dated $eptember 1975, as amended, to provide guidance ;

to staff reviewers who perform safety revlews of applications to construct or operate nuclear power plants. Our review of Sandesert has been conducted with the benefit of this guidance.

Tne Standard Review Plan identifies acceptance criteria appropriate to various subsections of safety analysis reports. At the beginning of our safety review of j the Sundesert application, we requested that the applicant document its conformance with the acceptance criteria described in the Standard Review Plan. The applicant provided this information in Table 1.3-2 of the Preliminary Safety Analysis Report.

1-12 j

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i We will report our evaluation of the conformance of the proposed plant design to the acceptance criteria of the Standard Review Plan in a future Safety Evaluation Report should the staff be requested to reactivate its review.

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a 1-13

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' 2.0 SITE CHARACTERISTICS Our evaluations of the suitability of the proposed site with respect to (1) demog-raphy and geography, (2) consideration of nearby industrial, military and trans-portation activities, (3) hydrology,. including certain hydrologic design criteria, (4) geology and seismology, including seismic input criteria, and (5) site meteor-ology, were presented in NUREG-0171, Early Site Review Report for the Sundesert Site," dated February 10, 1977. The following sections, which present our evalua-tion of the site characteristics performed during the construction permit stage of review, incorporate the applicable portions of the evaluations presented in NUREG-0171.

2.1 Geography and Demography The 7,040 acre Sundesert site is located in the extreme southeastern portion of Riverside County, California, about 5.5 miles west of the Colorado River. The site is located on the Palo Verde Mesa, 9.5 miles southwest of Ripley, California, 16 miles southwest of Blythe, California, and 50 miles north-northwest of Yuma, Arizona.

The Unit I containment will be centered at 33 degrees, 27 minutes, 7 seconds north latitude and at 114 degrees, 47 minutes, zerc seconds west longitude. The Unit 2 containment will be located 600 feet oue east of Unit 1. Figure 2.1 identifies the site location and characteristics of the area within 10 miles of the site.

The applicant has defined an exclusion area radius of 3,200 feet and a site boundary as shown in Figure 2.2. The applicant has acquired that portion of the exclusion area showr. as Parcel No. 5, including 100 percent of the mineral rights. The applicant proposes to acquire the 6,560 acres of land shown as Parcel No. 1, which includes the remainder of the exclusion area, from the U.S. Department of Interior, Bureau of Land Management, through an in-lieu property exchange. To this end, the Bureau of Land Management has designated lands, in the Coachella Valley region in California, which it is interested in acquiring bv exchange. The applicant has optioned the land designated by the Bureau of Land Management for the property j exchange. Based on the Bureau of Land Management's land exchange practices, the (

) applicant will acquire 100 percent of the mineral rights in about two-thirds of the land to be acquired from the Bureau of Land Management. However, the applicant will acquire from the Bureau of Land Management 100 percent of the mineral rights in all the land within the exclusion area.

Since the exchange of lands with Bureau of Land Management has not taken place yet, we cannot conclude at this time that the applicant has the authority to determine all activities within the exclusion ars, " required by Section 100.3(a) of 10 CFR Part 100. We will not be able to conclude on bi* aatter until the exchange of l lands has take' place. Therefore, this matter remains utstanding.

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JERSHIP - SUNDESERT SITE 2-3

The applicant has specified a lo population zone with a radius of three miles.

l The 1985 population within the three mile low population zone is estimated by the applicant to be 20 persons. The 1985 population within five miles of the site is estimated to be 489 persens and within 10 miles is estimated to be 1184 persons.

The estimated cumulative population distribution within 50 miles of the site for the year 1985 is shown in Table 2.1. The average population density out to 30 miles is about six persons per square mile.

The applicant states that the population center of more than 25,000 persons, as uefined in 10 CFR Part 100, closest to the proposed site area is the city of Yuma, Arizona. The 1970 population of Yuma was 29,007 and its location is approximately 50 miles south-southeast of the proposed site. Population projections do not indicate that any other area within f6 niles of the site will attain a population exceeding 25,000 by the year 2020, the approximate end-of plant life. Therefore, the population center distance of 50 miles is well in excess of the minimum popula-tion center distance of one-and-one-third times the low population zone radius of three miles, as required by 10 CFR Part 100.

Two distinct types of transient population are attracted to the area within a 50 mile radius of the proposed Sundesert site. The first type involves people pursuing recreational activities who visit the area primarily during the winter season. The second type involves transient farm workers employed on the area's irrigated farm lands.

Major concentrations of desert transient recreationists are located south, west, and east of the proposed site. Major concentrations of river oriented transient recreationists occur along the Colorado River Valley extending north by northeast, to south by southeast from the site. The section of the Colorado River Valley extending to the southeast contains the largest fraction of transient recreation-1sts. The estimated mean seasonal day recreational population within a 50 mile radius of the site is $638 for 1985. No transient recreationists are projected to be within the three mile low population zone during the plant lifetime.

Agricultural areas within a 50 mile radius of the proposed Sundesert site contain more than 300,000 areas of irrigated farmlands. The 1974 peak transient work force in the area was estimated to be approximately 3,500 during cultivation and harvest periods. During the lifetime of the proposed plant, however, there are no known plans for agricultural activities that would result in transient agricultural j workers within the three mile low population zone boundary.

Based on the applicant's designated low population zone and population center 1

distance, and upon the calculated radiologica?' onsequences of design basis acci-dents at the outer boundary of the low poptlation zone (see Section 15.5 of this report), and on our evaluation of the applicant's preliminary emergency plans (see Section 13,3 of this report), we conclude that the low population zone and population t

i 2-4

\ . _ - _ - - - _ . - _ _ . _

l TABLE 2.1

. ESTIMATED POPULATION DISTRIBUTION WITHIN A 50 MILE l i

RADIUS Of THE SUNDE5ERT SITE (Permanent Residents) i 1985 Sector 0-10 10-20 20-30 30-40 40-50 Total North 0 0 13 0 15 28 North-Northeast 45 1,060 72 1,023 2,307 4,507 11,986 542 174 55 12,963 Northeast 206 East-Northeast 238 574 25 895 24 -1,756 j East 165 0 0 94 6 265 ,

I East-Southeast IM 0 0 6 4 206 Southeast 280 2 0 0 4 286 34 70 0 336 S,378 3,818 4 South-Southeast

\

3 0 0 1,094 1,111 I South 14 19 10 28 0 58 South-Southwest 1 0 0 785 787 Southwest 0 2 0 0 54 4,110 4,164

,, Southwer.t 0 0 2 0 133 135 West 0 i

0 0 621 576 1,202 West-Northwest 5 0 0- 32 3,823 3,855 Northwest 0 0 0 0 0 North-Northwest 0 0 I

13,716 664 3,263 16,314 35,141 Total 1,184 1

2-5

I 1

center <tistances maat the guidelines of 10 CFP Dart 100 and, therefore, are acceptable.

We have also determined that the radiological consequences of design basis accidents at the boundary of the designated exclusion area are within 10 CFR Part 100 guidelines.

Final determination of the acceptability of the designated exclusion area with regard to the requirements nf Secti w 100.3(a) of 10 CFR Part 100 cannot be made until the applicant obtains title to those portions presently owned by the Bureau of Land Management.

2.2 Neart11 ndustrial, Transportation, and Military Facilities The site is comparatively remote from industrial and transportation facilities.

The separation distances of the site trom roads, railroads, pipelines, the Colorado i River, mines, and quarries are such that any hazardous materials potentially avail- l able it those locations would not pose a credible threat to the safe operation of )

the proposed plant. The nearest likely locations for hazardous materials are a Bureau of Reclamation rock quarry and State Highway 78, each located approximately l five kilometers from the site, to the north and east, respectively. We evaluated the effects on the proposed plant of potential explosions at these two locations by j using tne approach presented in Reguletory Guide 1.91, " Evaluation of Explosions l Fostulated to Occur on Transportation Rot.tes Near Nuclear Power Plant Sites." The I results of our analysis indicated that an explosion of ovet 14,000 tons of TNT would be required to induce even minor damage to the plant as it will be designed.

Since this is greatly in excess of the amounts of hazardous materials of equal or lesser damage potential that could be reasonably ex.pected at those locations, we conclude that the above facilities do not pose undue risk to the proposed plant.

A federal airway, V-135, serving north-south air traffic between 5,000 and 18,000 feet altituries, passes approximately six kilometers east of the site, and the near-est jet airways, J-50 and J-65, serving east-west traffic above 18,000 feet, pass I

18 kilometers to the north. These airways are sufficiently distant that the expecta- i tion of an impact at the site of an aircraft disabled while flying within the airways is less than 10~7- per year.

l Military aviat ion in the Colorado River Valley south of Blythe is due to traf fic to '

and from the Marine Corps Air Station at Yuma, 90 kilometers south of the site, This military traffic consists of several low-level training routes at 500 to 1500 feet altitude, and a Standaro Instrument Departure route and approach holding pattern at altitudes above 18,000 feet. All other military aviation in the general area is confined to restricted military airspace 20 kilometers or more from the site.

1 By agreement between the Nuclear Regulatory Commission and the Department of Defense (D00 Flight Information Publication AP/1B, Chapter 6, " Avoidance Locations"), the location of any plant for which a Commission operating license is pending is trans-mitted to t.he Department of Defense, which then issues appropriate instructions j l

2-6

that all low-level military aviation training routes avoid close passage of the site, commencing prior to the initial operation of the plant. The applicant has l performed an analysis which assumes that the loclevel routes will have.been moved f the minimum distance to comply with the directive, and has estimated the residual

probability density of aircraft impact at the site to be: :f the order of 10' per unit per year or less. We have reviewed this analysis and find it to be acceptable as a measure of low likelihood. W f urther believe there is a trend toward decreased military interest in the future in low-level flight training, such that future risk is unlikely to equal or exceed the risk as presently estimated.

Military aviation above 18,000 feet consists of holding patterns and a departurc route associated with the Marine Corps Air Station at Yuma. Since armed aircraft j

are included in the military aviation leaving Yuma, we required the applicant to i arrange with the Department of Defense for a repositioning of these flight paths, i

regardless of the probability of an accident. This repositioning will occur such

{ that armed aircraf t will be directed to pass approximately 15 kilometers to the

) west of the site. The holding patterns for unarmed fighter and attack aircraf t l pass five kilometers or more f rom the site. Assessed as a stochastic process, i.e. , assuming accident initiation at typical in-flight fighter plane peacetime

! loss rates with random impact probability over all areas within glide distance, the probability density of aircraft impact at the site from high altitude military traffic is in the order of 10 per unit per year or less.

We conclude that, subsequent to repositioning of the flight paths as indicated i above, the likelihood of aircraft impact at the Sundesert site is sufficiently l remote and need not, be considered as e design basis event.

I The nature and extent of activities, including shipment of potentially hazardous materials and aircraf t flights, which are conducted at nearby industrial, trans-portation and military facilities have been evaluated to determine if such activ-1 ities have the potential for adversely affecting plant safety related structures.

Based on the evaluation of information contained in the Preliminary Safety Analysis l

l Report, as well as information independently obtained by us, we conclude that the I facility is adequately protected and can be operated with an acceptable degree of safety with regard to potential accidents which may occur as the result of activities I at nearby industrial, transportation and military facilities.

i h

2.3 Meteorology To assure that safety related plant design and operating bases for a proposed plant

include metecrological parameters which are within Commission guic? lines, an evalu-

! ation is pe" formed of regional and local climatological information, including l extremes o, climate and severe weather occurrence, which may affect safe design and siting er a nuclear power plant. To determine that postulated accidental and routine operational releases of radioactive effluents from a proposed plant are l

, 2*7

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

within these guidelines, an evaluation is 91s0 performed of the atmospheric dif f usion l characteristics of the proposed site. 0,sr evaluation of the meteorological charac-teristics of the Sundesert site was performed in accordance with the procedures outlined in Sections 2.3.1 through 2.3.5 of the Standard Review Plan and is presented in the following sections. '

1 l

2.3.1 Regional Climatology The lower Colorado River Valley between California and Arizona experiences a desert-type climate. Skies are usually clear in the ar?a and the area receives over 90 percent uf possible sunshine. Throughout this arid region, the summers are long and hot with afternoon temperatures avereging over 100 degrees Fahrenheit from June into September; and on almost half of the days of the year, temperaturec will exceed 90 degrees Fahrenheit. Winters are mild with daytime temperatures averaging near 70 degrees Fahrenheit. Freezing temneratures occur only a few days each year.

The maximum and minimum dry-bulb temperatures selected by the applicant for general plant desion are 120 and 24 degrees Fahrenheit, respectively. Since a minimum I dry-bulb temperature of five degrees fahrenheit has been recorded at Blythe, California on January 6,1913 (U.S. Depar tment of Commerce, Environmental Data Service, 1964), we need to evaluate whether the minimum temperature proposed by the applicant is reasonable f* general plant design. Therefore, this matter remains outstahding.

The design temperature parameters to tie used for general plant heating, ventilation and air conditioning systems are as follows:

Temperature, degrees Fahrenheit l

l Dry-Bulb Wet-Bulb l Summer 112 78 Winter 32 ---

The design dry-bulb and wet-bulb temperatures for the summer months (June-August) ,

should be equalled or exceeded only one percent of the time in the site region. !

The design dry-bulb temperature for the winter months (December - February) should be equalled or exceeded 99 percent of the time in the site region. Based on com*

parisons with regional historical cata, we conclude that the design temperatures for the general plant hnting, ventilation and air conditioning systems are reason-able for the region.

i I

2-8

l l

l M applicant has provided wet-bulb temperature data to evaluate the performance of

he vet mechanical' draft cooling towers proposed for the ultimate heat sink.

! . Table 24 3.1-1 of the Sundesert Preliminary Safety Analyses Report lists these data which are based on 14 station years of data from Blythe, California, and Yuma, Arizona. Based on the criteria outlinea in Regulatory Guidc 1.27 " Ultimate Heat L

Sink for Nuclear Power Plants," these meteorological data are reasonable for evalu-ating the performance of the mechanical draf t cooling towers. As stated in Section 2.4.3 of this report, the applicant will provide the results of preopera-tional testing to confirm the significant design parameters for the particular to, sing tower selected.

Rainfall is sparse in the region averaging only three to four inches each year.

Relative humidities can be quite low, averaging around 15 percent on spring after-noons, However, during the summer months the water content of the air is higher than over most desert regions as air from the Gulf of California or the Gulf of Mexico flows irto the region.

Snow, glaze and hail are almost non-existent in the region. Therefore, we conclude that snow or ice need not be considered as design parameters for a nuclear plant in this region.

Although infrequent, thunderstorms, tornadoes, tropical cyclones, and dust storms can affect the site region. Thunderstorms occur less than 10 days annually, prin-cipally in late summer.

Between 1953 and 1974, only two tornadoes were reported within a 10,000 square mile area containing the site. Using the methods of Thom (1963), these data project a recurrence interval of 40,000 years for a tornado at the plant site. The applicant has selected the following values fo- the design basis tornado parameters for the Sundesert plaat:

r

,T or, Value Maxit our) 300 Rotat1L as/ hour) 240 Maximum ., Speed (miles / hour) 60 Total Pres. ore Drop (pounds / square inch) 2.25 Rate of Pressurc Drop (pounds / square inch /second) 1,2 i

These values conform to those described for Tornado Intensity Region 11 in Regula-tory Guide 1.76, " Design Basis Tornado for Nuclear Power Plants," and are, therefore, i acceptable for this region.

Tropic.al storms are also rare in the site region, with a storm entering the region only about once every 10 years. The " fastest mile" of wind recorded at Yuma, Arizona, has been 60 miles per hour (August 1973). The applicant has selected an 2-9

4 1

j ope nting b e is wind speed (defined as the " fastest mile" wind speed at a height of

( 30 feet with a return period of 100 years) of 80 miles per hour. Based on regional j data (U.S. Department of Commerce Environment Data Service,1976; Thom,1968), we l find the operating basis wind speed selected by the applicant for the site to be'

{ acceptable.

I e Dust storms are relatively common within the site region. Between 1940 and 1970, dt.st or blowing dust and sand reduced visibility to under seven miles about 45 hours5.208333e-4 days 0.0125 hours 7.440476e-5 weeks 1.71225e-5 months annually.

i j The applicant has established an ambient dust and salt monitoring program which l will provide: (1) airborne particle i,ize distribution and corresponding mass loading as a function of height (to at,sist in the design of building ventilation l systems and the cooling tower - circulating water - evaporation basin system); and (2) ambient background concentrations of airborne salts (to assist in assessing environmental impact due to cooling tower operation). (See Section 2.4.3 of this

! report regarding cooling tower operation.) The details of the program are described l in the Preliminary Saf ety Analysis Report starting on page Q372.13. We will review l the results of the monitoring program at the operating license stage of review.

I h

h We conclude that, except for the design basis minimum temperature for general plant design which we are still evaluating, the applicant has sufficiently described the

regional climatology and severe weather phenomena which are important to the safe i design and siting of the Sundesert f ac ility.

i i

{ 2.3.2 Local Meteorology 4

f The Sundesert site is located on the generally flat, sparsely vegetated Palo Verde Mesa. Within five miles of the site, elevations range from about 1300 feet above the proposed plant grade (375 feet above mean sea level) to about 150 feet below plant grade.

To assess the local meteorological characteristics of the Sundesert site, cliwato-logical data from Blythe, California (15 miles northeast of the site), Yuma, Arizona 1 (50 miles south-southeast) and parker, Arizona (50 miles northeast), and two years of data collected onsite were evaluated. These data are reasonably representative of conditions expected at the site and its vicinity. f I

In the site area, average daily maximum and minimum temperatures range between 107 degrees Fahrenheit and 81 degrees Fahrenheit in July, the warmest month, and between 67 degrees Fahrenheit and 41 degrees Fahrenheit in January, the coolest month. The extreme maximum temperature recorded in the area was 123 degrees Fahrenheit (Yuma, Arizona, September 1,1950); the extreme minimum temperature has been five degrees Fahrenheit (Blythe, California, January 6,1913).

I 1

1 2-10

i I

l l

Blythe receives .about 3.2 inches of rain annually. August has been the " wettest" l l

month with an average of 0.7 inches, while during May and June the combined average j is less than 0.1 inches. The maximum 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months rainfall recorded among these stations I has been 4.0 inches (Yuma, August 1909). Snowfall is a rarity, with a trace (less 0.01 inches) being the most ever observed. Heavy fogs (visibility of 1/4 mile or less) occur only one or two days each year. !

I For the two year period of June 1, 1975 through May 31, 1977, the wind flow over the site, as measured at the 10 meter (33 foot) level of the onsite meteorological ,

l tower, was from the southwestward direction about 28 percent of the time. Figure 2.3 1 shows the directional frequency of onsite winds. Winds were calm (windspeeds less f than one mile per hour) less than one percent of the time.

We conclude that the applicant has described the local meteorological conditions which are important to the safe design and siting of the Sundesert facility.

2.3.3 Onsite Meteorological Measurements Program A 260-foot high meteorological tower, erected onsite about 2500 feet north of the proposed location of the reactor f acility, became operational in May 1975. The parameters measured and the measurement levels on the tower are indicated in Table 2.2.

Data were recorded on magnetic tape with analog strip charts as a backup. Complete calibrations of the system were performed at four month intervals Section 6.1.3 of the Sundesert Environmental Report provides information regarding maintenance, calibrations, quality assurance, data handling and processing procedures, and the specific instrumentation used for the onsite program. We conclude that this onsite meteorological program meets the recommendations stated in Regulatory Guide 1.23, "Onsite Meteorological Programs" and, therefore, is acceptable.

The applicant has provided joint frequency distributions of wind speed and direc-tion by atmospheric stability class, defined by the vertical temperature gradient, based on data collected onsite from the meteorriogical tower for the period June 1, 1975 to May 31, 1977. The distributions were for wind speed and direction measured at both The 33-foot and 190-foot levels with the vertical temperature difference between the 33-foot and 190-foot levels. Joint data recovery rates for stability and wind at both levels were 97 percent.

i From an examination of these data and a comparison of the data with long-term data from Blythe, California, and Yuma, Arizona, we conclude that the data collected onsite are reasonably representative of the long-term meteorological conditions of 3

importance to making diffusion estimates.

2-11

l l

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! NORTH I'

7 '~ ~~ 'N PERCENT

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Onsite data at 10 meters (33 feet) above ground level, June 1,1975 through May 31, 1977. Bars show the direction from which the wind blows. Calms are those winds with hourly average speeds less than 0.45 meters per second (1.0 mile per hour).

3 FIGURE 2.3 DIRECTIONAL FREQUENCY OF WIND-SUNDESERT SITE 2-12 j

1 TABLE 2,2 I

SUNDESERT MElEOR0 LOGICAL INSTRUMENTATION Measured Parameter Elevation Above Ground Meters feet Wind Direction and Speed 10;58 33;190 10 33 Dry-Bulb Temperature Dewpoint Temperature 10 33 Dry-Bulb Temperature Gradient 58-10 190-33 4 15

. Solar Radiation Precipitation lh 4 NOTE: Redundant instrumentation exists for all wind and dry-bulb temperature measurements.

I l

i t

I i

2-13

L 3. 4 Short-Term (Accident) Diffusion Conditions We have calculated the meteorological dif fusion characteristics (relative concentration values) of the site, for evaluating postulated accidental releases of radioactivity from plant buildings and vents, by using the two years of onsite data provided by the applicant. The calculated short-term relative concentration values were determined for various time periods, following a postulated release, for both the exclusion area boundary (975 meter radius) and the outer boundary of the low population zone (4827 meter radius) by using the wind speed and direction measured at the 33-foot level and the temperature gradient between the 33-foot and 190-foot levels and by assuming a ground level release with a building wake factor of 1144 square meters.

a j Our current position for determining the short-term relative concentration values j

allows the use of either the direction-independent approach presented in Section 2.3.4 i

of the Standard Review Plan in unamended form, or the direction-dependent approach l outlined in draf t Regulatory Guide IJXX, " Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants," which was transmitted to the applicant by letter dated December 5, 1977.

In our determination of the short-term relative concentration values, we used a calculational model which considers the variability of meteorological conditions by

) direction and which is suitable for a site in a desert location. The model is a modified (for desert conditions) version of a model presented in draft Regulatory Guide 1.XXX.

Table 2.3 shows the 0-2 hour relative contentration values which on the average we estimate will be exceeded no more than 27 hours3.125e-4 days 0.0075 hours 4.464286e-5 weeks 1.02735e-5 months per year at the exclusion area boundary for each of 16 downwind directions. The south-southeast downwind sector has the highest relative concentration value for both the exclusion area boundary and the outer boundary of the low population zone. The values for this sector, which are listed in Table 2.4, were used in our evaluation of short-term accidental releases presented in Section 15.5 of this report. The 0-2 hour relative concen-tration value of 2.2 x 10 ~4 seconds per cubic meter for this maximum sector will occur less than five percent of the time (438 hours0.00507 days 0.122 hours 7.242063e-4 weeks 1.66659e-4 months per year) around the exclusion area boundary.

, The applicant calculated short-term relative concentration values using the proce-dures in Section 2.3.4 of the Standard Review Plan. This method uses a direction-independent model to estimate atmospheric diffusion conditions which occur no more

?.han five percent of the time (438 hours0.00507 days 0.122 hours 7.242063e-4 weeks 1.66659e-4 months per year) around the site at a distance equal to the minimum exclusion boundary distance. The applicant modified this '

model by using plume spread parameters which reflect plume meandering in a desert location. However, to allow for the enhanced effects of plume meander, it is necessary to also consider jointly the variation of meteorological conditions by i

direction and the variation of exclusion area boundary, if applicable, all of which are considered in the model in draft Regulatory Guide 1.XXX.

l 2-14 l

s l

l i

TABLE 2.3 SHORT-TERM RELATIVE CONCEN1 RATION VALUES BY DOWNWIND ~

DIECTION - SUNDE5ERT SITE ,

I The values are the 0-2 hour relative concentrations which we estimate will be exceeded no more than 27 hours3.125e-4 days 0.0075 hours 4.464286e-5 weeks 1.02735e-5 months per year at the exclusion area boundary (a 975 meter radius from the containment buildings) in the downwind direction indicated.

RelativeConcentratjons RelativeConcentratjons Downwind Sector (seconds / meter ) Downwind Sector (seconds / meter )

North 1.6 x 10'4 South 2.2 x 10'4

~4 ~4

North-Northeast 1.6 x 10 South-Southwest 2.2 x 10 Northeast 1.7 x 10'4 Southwest 1.7 x 10' E as t-Northea s t 2.0 x 10'4 West-Southwest 1.1 x 10'4 East 1.9 x 10'4 West 1.2 x 10'4

~4 1.3 x 10'4 East-Southeast 2.1 x 10 West-Northwest

~4 Southeast 2.2 x 10'4 Northwest 1.2 x 10 South-Southeast 2.2 x 10'4 North-Northwes t 1.3 x 10'4 TABLE 2.4 SHORT-TERM RELATIVE CONCENTRATION VALUES USED FOR ACCIDENT ANALYSIS - SUSE5ERT SITE I

The values are the short-term relative concentrations used to evaluate accidental releases from plant buildings and vents. The values are for appro-priate time periods following a release and are for the exclusicn area boundary and the outer boundary of the low population zone.

RelativeConcentrajions Time Period location (seconds / meter )

0-2 hours Exclusion Area 2.2 x 10'4

-5 0-8 hours Low Population Zone 3.3 x 10

-5 8-24 hours Low Population Zone 2.4 x 10 1-4 days Low Population Zone 1.2 x 10 -5 4-30 days Low Population Zone 4.8 x 10 -6 l

2-15

---. -. -_ . - - . .. - _. - - = _ - -- . -. -

A comoarison of our calculated values with those calculated by the applicant indi-cate that our relative concentration value for the exclusion area boundary is about 20 percent higher whereas our values for the low population zone boundary range from about 80 percent higher for the shortest time period (0-8 hours) to no change for the longest time period (4-30 days).

Our modified approach was based on the procedures in the Standard Review Plan which also require that we determine whether the lateral and vertical plume spread para-meters used in our calculational model are conservative for estimating relative concentrations. Based on atmospheric diffusion field data developed at two desert locations (. e National Reactor Testing Station, Idaho, and the Hanford Area, Washington) (Yanskey, et al. ,1966), we concluded that the overall plume disper-sion, including the effects of plume meander, for stable atmospheric conditions with low wind speeds (periods of poorest dispersion) can be less for a desert climate than the plume dispersion assumed in draft Regulatory Guide 1.XXX. Thus, in lieu of the plume spread and plume meander factors described in this draft guide, we used the lateral and vertical plume spread factors (which include plume meander) based on the field data from deserts. These were the same factors used by the applicant in its direction-independent analysis.

for comparative purposes, we calculated relative concentration values by using the unmodified draf t Regulatory Guide 1.XXX dif fusion parameters and by using the desert diffusion parameters in our modified model. The c6)culations were made for downwind distances corresponding to the Sundesert exclusion area boundary and low population Zone outer boundary. For wind speeds less than two meters per second ,

during stable conditions, individual one hour average relative concentration values calculated with the desert field data ranged from no change to a factor of about 1.5

)

times higher than those using the draf t regulatory gui6 dif fusion parameters.

Using actual Sundesert meteorological data to calculate diffusion estimates similar j to those given in Table 2.4, both sets of diffusion parameters produced about the

{

same 0-2 hour relative concentration values at the exclusion area boundary, whereas !

the desert parameters produced a higher value, by a factor of about 1.2, for a 0-2 hour relative concentration value at the low population zone distance. Therefor?, 1 we conclude that our modified model provides conservative alues. l 2.3.5 tong-Term (Routine) Diffusion Estimates Using the onsite 10 meter level wind data and the vertical temperature difference I data, we calculated average atmospheric dispersion conditio1s for the Sundesert site using our atmospheric dispersion model for long-term re%ases (Sagendorf and Coll,1976). This model is based on the " Straight-line Trajs ctory Model" described in Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport end Dis-persion of Gaseous Ef fluents in Routine Releases f rom tight-Watec-Cooled Reactors" As recommended in Regulatory Guide 1.111 and as discussed in Section 2.3.4 of this report, we used the vertical dispersion parameters developed from atmospheric 2-16

I I

r diffusion data irom the desert field tests. We assumed a ground level release, and included an i. stimate of the maximum increase in calculated relative concentration and relative deposition values which may occur due to the spatial and temporal I variation of the airflow not constoered in the straight-line trajectory model discussed in Regulatory Guide 1.111. l Table 2.5 lists the relative concentration and relative deposition values used to

estimate radiation doses as described in Section 5.1.4 of the Sundesert Draft Environmental Statement, NUREG-0405. We calculated the Table 2.5 values using only the first year (June 1975 - May 1976) of onsite data, in our calculations, we conservatively assumed that all onsite data denoted by the applicant as variable direction winds occurred during periods of calm winds, in the two year data set,

]

the applicant noted the wind speeds associated with the variable winds. Subsequent calculations of the routine dirpersion values with the two years of data, including variable wind speeds assigned in tne proper wind speed class, show that the values of Table 2.5 are greater than or within 10 percent of those calculated with the two years of data.

2.3.6 Conclusions The applicant has provided sufficient information for us to evaluate the regional and local meteorological conditions of importance to the design and siting of the Sundesert facility. The two years (June 1973 - May 1,977) of onsite meteorological data provide an acceptable base for tne calculation of reasonably conservative relative concentration values of annual and post-accident atmospheric dif fusion conditions. As stated in Section 2.3.1 of this report, we have not completed our evaluation of the minimum temperature proposed by the applicant for general plant design.

2.4 Hydrologic Engineering l 4 2.4.1 Hydrologic Description The proposed site for the Sundesert plant is located about 5.5 miles west of the Colorado River on Palo Verde Mesa overlooking Palo Verde Valley, and is about 16 miles southwest of Blythe, California. The proposed plant. grade will be about 375 feet above mean sea level. The Colorado River in the vicinity of the site is about 150 feet below the proposed plant grade.

The site is within the Colorado River watershed, of which approximately i82,200 square miles are upstream. The Colorado River is highly regulated with numerous i

dams t%at provide water supply, irrigation, recreation, power, and flood control.

The two largest and most significant dams are Hoover and Glen Canyon Dams. The site local drainage area is about 7.5 square miles and is characteristic of desert area basins. It consists of the very steep and barren Mule Mountains with many canyons termi ating in alluvial fans which merge, forming a bajada with numerous dry channels.

2-17

1 4

i i

i TABLE 2.5 i d

I

SUMMARY

OF LONG-TERM RELATIVE CONCENTRATION i AND RELATIVE DEPOSITION VALUES FOR TELECTED LOCATTDNS NEAR s

i THE SUNDESERT~51TE location Source RelativeConcentrat{ons RelativeDepgg)itions )

(seconds / meter )

(meters l Nearest Site Boundary A 3.0x10:5 3.8 x 10 8

! 0.98 kilometers 8 6.7 x 10 j

(0.61 mile) C 8.0 x 10 -5 8.5 1.0 xx 10 10 7 South southeast l Nea est garden t

A 1.2 x 10f 1.5x10l 4.8 kilometers i (3.0 miles)

B C

3.3 x 10 4.1 x 10 -6 4.0x10_l 5.0 x 10 9

{ Northeast 3

l Key: ' Nearest refers to that type of location where the highest radiation dose

is expected to occur f rom all appropriate pathways.

i

! Sources: A - Plant Vent, continuous releases y

j B - Gas decay tank, purge releases (15 purges / year 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months / purge) 4 C - Containment vessel, purge releases (?4 purges / year, 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months / purge) l i

J i

i i

1 t

i !

t 1

1 1

1 l

5 4

d 4

3 4

2-18

l 2.4.2 flood Potential The applicant assessed the potential of flooding the site from three sources: (1) a probable maximum flood on the Colorado River; (2) flonding due to seismic failures of upstream dams on the Colorado River coincident with a standard project flood; l

' and (3) flooding due to probable maximum precipitation on the local 7.5 square mile

' basin. Based on our review of the applicant's analyses, we agree with the appli-cant that the site is not subject to flooding from any credible events, including dam failures, on the Colorado River since, as noted previously, the site is about '

150 feet above the Colorado River and about 5.5 miles away. The design basis flood conditions will result from probable maximum precipitation on the loct.1 drainage basin.

The applicant has estimated a peak discharge flow of 33,000 cubic feet per second for a local probable maximum flood resulting from probable maximum precipitation on the local drainage basin. We independently evaluated the local probable maximum precipitation using " Probable Maximoa Thunderstorm Precipitation Estimates, South-west States" prepared by the National Weather Service, National Oceanic and Atmosphere Administration. We also assessed the values developed by the applicant for precipi-tation losses, runoff functions, and debris flow contribution to a local probable f

maximum flood and concluded that they are conservative. Based on these assess- l ments, we concur with the applicant's e:,timate of 33,000 cubic feet per second for the peak discharge of a local probable maximum flood.

5 The applicant has proposed a system of safety-related diversion dikes and channels to protect plant safety-related structures, systems and components from a local probable maximum flood. The site drainage system, including the roofs of safety-related structures, will be designed to carry rainfall, up to local probable maximum precipitation severity, away from the site without adversely affecting safety-related structures. Also, accesses to safety-related buildings will be one foot above plant grade. We have reviewed the proposed design criteria for the diversion dikes and channels, including erosion protection and freeboard, and conclude that they meet the recommendations of Regulatory Guide 1.59, " Design Basis floods for Nuclear Power Plants," and Regulatory Guide 1.102, " flood Protection for Nuclear Power Plants," and, therefore, are acceptable.

2.4.3 Water Supply The applicant estimate

Wt approximately 17,000 acre-feet per year of makeup f

water will be used for each unit. The applicant proposes to supply the makeup f water for Unit No. 1 (an average flow of 23.5 cubic feet per second) with irriga-tion return water, which is relatively high in salt content, from the Palo Verde irrigation District's Outfall Drain. In order to maintain the water allotment for the downstream users, the applicant has acquired 17,000 acre-feet of water per year l

l f rom the Metropolitan Water District of Southern California's allotment to the 2 43

i i

d l California coastal plain which it obtains via the Colorado River aqueduct. Instead i of actually using it, this amount of additional water, having a lower salt' content than the irrigation return water in the falo Verde Irrigation District's Outf all f Drain, will be allowed to pass through Parker Jam and down the Colorado River.

l l For Unit No. 2, the applicant proposes to reduce the amount of irrigation water to f the applicant's farm lands within the Palo Verde Irrigation District by a suf fi-t cient amount to provide the 17,000 acre-feet per year needed for makeup.

t The makeup water will be pumped to cooling towers which will provide the necessary f cooling of the plant for normal operation. Blowdown from the cooling towers will be to onsite evaporation ponds from which no releases will be made.

{

i j Separate safety-related cooling towers will supply emergency cooling water. Ne l have reviewed the applicant's proposed design bases for emergency water suppl,r and l conclude that these bases meet the criteria suggested in Regulatory Guide 1.27, I

j " Ultimate Heat Sink for Nuclear Power Plants," Revision 2, and, therefore, are f acceptable. In addition, at our request, the applicar.t has committed to provide the results of pre-operational testing to confirm significant design parameters of

the particular safety-related cooling tower selected. Additional discussion on the ultimate heat sink is provided in Section 9.2.3 of this report.

5 2.4.4 GroundWay Ground water at the site occurs under water table conditions from about 145 to 150 feet below the surface at the site and is directly related to surfac? water levels in the Colorado River. The hydraulic gradient is toward the Colorado River. Due to the depth of the water table, dewatering during construction will not De neces-sary, and the design basis grounr1 water levels will be below safety related struc-tures. Our assessment of the design bases ground water levels included the potential effects of the evaporation ponds on these levels at the site. We conclude that the I water in the ponds would not influence ground water levels to the extent that there l could be an adverse impact on safety-related structures. This conclusion is based on the distance of the ponds from safety related structures, the relatively small amount of water that would normally be in the ponds and the applicant's intent to place relatively impermeable liner material in the ponds.

We and the applicant independently analyzed the ef fects of a postulated spill of radioactive liquids at the plant site. For the purposes of this analysis, we determined that the most critical case would be the failure of the boron recovery tank (120,000 gallons) located in the radwaste building. Upon the postulated rupture of the boron recovery tank and building, the radioactive liquids would mix and travel with the ground water according to the hydraulic gradient of the ground water. The nearest well in the direction of ground water flow is located about I three miles from the radwaste building. In our analysis, we conservatively assumed 2-20 1

that the spill would enter the ground water beneath the radwaste building instanta-neously and conservatively determined that the spill would be diluted by a factor '

of approximately 480 and would have a travel time to the well of about 90 years.

We also determined that the resultant concentrations of all nuclides at the nearest well would be smaller than the concentrations listed in Appendix B, Table 11. to 10 CFR Part 20 (see Section 15.5.5 of this report for an additional discussion of this postulated release).

l 2.4.5 Conclusions I

Based on our independent review and analyses, we conclude that an adequate water supply can be assured for safety-related purposes, that adequate design bases for flooding have been provided, and tnat postulated spills of radioactive liquids will not result in radionuclide concentrations in excess of 10 CFR Part 20 limits at public water supplies or domestic wells. Therefore, we conclude that the proposed hydrologic design bases for the plant are acceptable.

2.5 Geology and Seismology The results of our review of the geologic and seismologic characteristics of the Sundesert site are presented in Section 2.5 of NUREG-0171. Since the completion of that review, no new information has evolved relating to geology and seismology of the site or which effects in any way the conclusions presented on geology and seismology in NUREG-0171. Therefore, the following evaluation of geology and seismology is the same as that reported in NUREG-0171 and is presented for the sake i of completeness. We subsequently requested the applicant to provide a commitment l to geologically map excavations to be marie for seismic Category I structures during the construction phase. Our evaluation nf the applicant's response is presented in Section 2.5.3 of this report. l l

2.5.I Regional Geolog The Sundesert site is located in the Sonoran Desert subprovince of the Basin and Range geologic and physiographic province. Basin and Range type structural geology and the San Andreas fault system (including subparallel major fault zones with similar characteristics) provide the distinguishing geologic characteristics of site region.- Within 200 miles of the site are located parts of the Great Basin and Mexican Highlands Transition Zone subprovinces and parts of the Colorado River Plateau, Salton Trough-Gulf of California, Peninsula Ranges, and Transverse Ranges provinces, The Sonoran Desert province includes the Mojave Desert of California and the Gila Desert of northwestern Mexico. This province is characterized by j subdued mountain ranges, usually less than 4000 feet in elevation, trending north-west, north and northeast. This subriand relief suggests a relatively stable crust.

l 2-21

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! The Basin and Range geologic province was involved in several orogenic events j ranging in age from Precambrain to Tertiary. The most recent diastropMsm to i affect the site region was the Laramaide orogeny which began in late Cretaceous and l 1

! continued into Tertiary time. A good description of the orogeny during the Ter- i i

l tiary time is ' presented by the applicant on pages 2.5-46 and 2.5-47 of Appendir A l to the Preliminary Safety Analysis Report under the heading Late Tertiary. Igneous activity, including volcanism and plutonism, was widespread in the Sonoran Desert

{ and Mexican Highland-Tras ition Zone during the Mesozoic. Volcanism occurred in '

j the Central Sonoran Desert Region, the Western Mojave, the Colorado Platec.u, and 1

the Salton Trough during Quaternary time. Continued crustal spreading along the

{ San Andreas f ault system is evidenced by extensive Quaternary and Holocene f ault j displacement which can be related to movement of the Pacific Plate relative to the North American Plate.

1 1

} The San Andreas fault system is the tectonic first order feature in Western North l America. The closest approach of this system to the site is approximately 40 l miles. The San Jacinto, Whittier, Elsinore, Garlock, and the Death Valley-furnace j Creek fault zones are approximately 75, 80, 170, and 200 miles, respectively, from

{ the site. Quaternary deformation is continuing in some areas of the site region.

l As a result, a number of active fault zones can be found in the region. All of the l active faults within the 200 mile radius of the site are not discussed here due to

] the dominant influence of the San Andreas fault zone and some smaller faults closer j to the site on the determination of the safe shutdown earthquake.

l The geologic evaluation ar1 tectonic implications of the San Andreas fault system have been discussed extensively by many authors. For the purposes of this raport,

only its relationship to the site area is addressed. The San Andreas fault system i is approximately 700 miles long and extends from the Mendocino Escarpment to the )

j Gulf of California. In Central California, the fault is basically a single, linear break displaying right lateral strike-slip displacement. Further to the southeast, the San Andreas fault has several elements. Still further to the south, the San l Andreas zone appears to terminate. As tne applicant describes, "At the south end I of the Salton Sea, the San Andreas Fault appears to terminate at an active spreading I center, transferring motion within the San Andreas system to the Imperial and San 1

1 Jacinto faults." Although the Sand Hills and Algodones faults lie along the pro-1 l jection of the San Andreas fault southeastward from the Salton Sea, they do not j appear to be active elements of the present San Andreas fault system. However, the l applicant has conservatively assumed the Sand Hills fault to be the element of the

} San Andreas fault. system closest to the site. ,

t Numerous small faults were found in the site region. The applicant canducted an intensive geologic investigation of all such features which were identified. None of the faults within 50 miles of the site, with the exception of those of the San Andreas fault system, have been associated with historic seismicity, although some show geologic evidence of Quaternary displacement. The Chuckwalla Mountain, Salton 2-22

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Creek and Sheep Hole faults and the Blythe Graben are considered to be capable faults. In addition to these faults, extensive investigations were conducted by the applicant along the Lost Trigo fault and in the Chocolate Mountains which lie adjacent to and northeast of the Salton trough and San Andreas fault zone, i

The Chuckwalla Mountain faults trena nortnwest for several miles. The closest approach of the faults to the site is about 25 miles and the faults are identified primarily as linears which parallel stream drainages. One of the linears aligns with an east-west trending fault which juxtaposes indurated Tertiary deposits with interbedded clay, silt, and sand deposits. Overlying younger alluvial fan surfaces and deposits appear undisturbed, but field relationships are not definitive enough to preclude Quaternary faulting.

The east west trending Salton Creek f ault separates the Orocopia Mountains from the l l Chocolate Mountains and is marked by a major change in geology between the two i l areas. Tertiary alluvial deposits are offset by the fault which has a mapped )

length of 12 miles and is located 38 miles from the site.

The Sheep Hole f ault, which trends northeast along the Sheep llole Mountains, dis-rupts Quaternary formations. Extension of this fault to the southeast is based on gravity data. A few earthquakes have been located near the northern end of this fault. Its length is about 40 miles and its closest approach to the site is 41 miles.

The Blythe Graben is a set of two parallel normal f aults spaced about 300 feet apart. It is a small arcuate structure which strikes approximately northwest, has a traceable len:th of 3-1/2 miles, and is 22 miles northeast of the site. The faults of the Blythe Graben offset Quaternary units and the last movements most likely occurred between 6,000 and 30,000 years ago. At present, the graben can be seen as a topographic depression in the alluvial surface. This structure is located to the southwest of the Big Maria Mountains and on strike with the general trend of the structural front of both the Big Maria and Little Maria Mountains.

The Blythe Graben coincides with a steep gravity gradient along the Little Maria and Big Maria Mountains. Although available data are inadequate to establish a direct structural relationship between the gravity gradient and the Blythe Graben, the coincidence of strike and location require that it be assumed that such a relationship exists. This gradient and another parallel to it, about four and a half to seven miles southeast of it (22 and 15 miles respectively from the site),

are interpreted as faults with large vertical separation. These faults would delineate a northwest-trending subsurface basin approximately coincident with McCoy wash.

To the southeast in the Dome Rock Mountains (approximately 30 miles from the site) are several northwest-trending faults which indicate separation up to two miles. l l

l 2-23 m m w w w e -o-

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These faults do not appear to distrub Plio-Pleistocene alluvial fan materials. The l steep gravity gradients noted in McCoy wash do not cut the Dome Rock Mountains. To the northwest, faulting was observed only in the older Tertiary fanglomerate, based on field reconnaissance and inspection of aerial photographs of the Palen Pass I 1

I area, but'no capable faulting was found. l l

l The Lost Trigo fault rune is a zone pproximately 1,000 to 2,000 feet wide contain-i ing numerous small faults, some of which are en echelon and others indicating dips f both to the east and west. This zone has a general north-south strike, has been traced for seven and a half miles and is located 15 miles southeast of the proposed l

Sundesert site along the western margin of the Dome Rock Mountains. Geologic l evidence indicates that this fault is not capable. The fault exposed in Hart Mine

wash offsets the Pliocene Bouse formation, a Plio-Pleistocene alluvial fan deposit, i and a Plio-Pleistocene fluvial deposit but is crosscut by an alluvial fan deposit i which is middle Pleistocene in age (estimated to be 500,000 to 1,000,000 years j old). l

! The Chocolate Mountains of California are immediately adjacent to and east of the a 4 l Imperial Valley-Salton Trough and the San Andreas fault system. To the north and J l

south of this range are the Orocopia and Cargo Muchaco Mountains, respectively.

J Previous mapping of this area, the Salton Sea Sheet, Geologic Map of California 1

(Jennings,1967) indicated the presence of numerous northwest-southwest and some l i

east west trending f aults. Some of the northwesterly trending faults were inferred l l

to be continuous for tens of kilometers. Some faults were shown to offset Qbater-nary units. Because of the proximity to the San Andreas fault system, the potential existence of a large throughgoing northwest-trending fault, which might be directly related to the San Andreas system and closer to the site than the Sand Hills fault, was assessed. l I

The geology of the Chocolate Mountains is not well known partly due to limitations _

on ground and aerial access to large areas of the Chocolate Mountain Aerial Gunnery !

1 Range, which is an active military practice range. In order to obtain more detailed '

mapping of this area, the applicant undertook a reconnaissance geologic mapping study utilizing newly acquired Landsat imagery and black and white aerial photo-graphs. This reconnaissance study was supplemented by ground field checks and extensive consultation with numerous experts on the geology of this region. As a result, the applicant has been able to generate a new updated map of this area. As a result of our review of this updated mapping, we conclude that the northwest-trending Tertiary or Quaternay faults in the Chocolate Mountains region southeast of the Salton Creek fault are discontinuous structures which cannot be directly related to the presently active San Andreas fault zone. Although there is evidence for the existence of some small capable faults along Salton Creek and on the western flank of the Chocolate Mountains, they have no influence on the determination of the safe shutdown earthquake for the Sundesert site.

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,,-_r-- - n.- + .,wa

l 2.' 5. 2 Tectonic Province and Regional Tectonics l

The proposed Sundesert site is located in the Basin and Range tectonic province.

As described by Eardley (1962), this province is characterized by an extensional stress regime which has resulted in block faulting with the mountains and interven-ing alluvium-filled valleys corresponding to up-lif ted and down-dropped blocks i

respectively. Throughout much of the province, the faults which marked the bound-ary between the up-lif ted and down-dropped blocks are now buried under alluvium eroded from the receding mountain fronts whirl % their identification difficult.

l The main tectonic event responsible for the development of the Basin and Range structure began in middle-Miocene time and continued into Pleistocene time (Eardley, 1962). However, a tensional stress regime conducive to strike-slip and/or normal faulting apparently persists to the present time in some parts of the province.

The Sonoran Desert region of the Basin and Range province, in which the Sundesert site is located, is characterized by broad and deep alluvial valleys and low-altitude mountains which are considered to be evidence that the area has experienced relatively little orogenic activity since the earlier stages of Be:b and Range development.

Northwest-trending right-lateral strike-slip deformation and northeast-trending lef t-lateral strike-slip deformation appear to be present in many parts of the western and southern portions of the Basin and Range province. Most of this deforma-tion was apparently initiated in Miocene time and is contemporaneous with the tectonic activity generally thought to be responsible for the formation of the

' typical 9asin and Range structural pattern (Hamilton and Myers, 1966). A limited number of earthquake focal mechanisms, displacements observed in historical surface faulting and observations of strain accumulation indicate that the present stress regime in the western and southern po 'tions of the Basin and Range province corre-sponds to extension, oriented between northwest-southeast and east-west.

If strike-slip faulting is the uominant mode of tectonic activity in the western and southern portions of the Basin and Range province, recent faulting could be more difficult to recognize than if normal faulting is dominant. However, if dip-slip displacement accompanies strike-slip movement, as is expected for most faults in the region, recognition of recent faulting would be facilitated.

Several faults have been identified by geologic investigation in the general vicinity of the site. As discussed in Section 2.5.1 of this report, some of these faults, such as the Chuckwalla Mountain f ault, the Salton Creek fault and the Blythe Graben, show geologic evidence of Quaternary fault displacement which is regarded as indicative that these faults are capable. However, the Chuckwalla Mountain fault, Salton Creek fault and Blythe Graben are not recognized to have associated seismicity.

The nearest of these faults to the Sundesert site is the Blythe Graben approximately 1 22 miles north of the site.

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West and southwest of the site, the tectonics are more strictly controlled by the interaction between the Pacific and North American Plates, This interaction mainly is represented by right-lateral strike-slip movement along faults in the San Andreas fault system, associated high seismicity and relatively recent (Quaternary) surf ace displacement. The southeast portion of the San Andreas fault system, where the fault system has its closest approach to the Sundesert site, splays into several strands which are in most areas buried under thick alluvium in the Salton Trough.

As noted in Section 2.5.1 of this report, the applicant has indicated that the San Andreas fault appears to terminate in an active spreading center at the south end of e Salton Sea which transfers motion within the system to more active strands further west in the Salton Trough. This spreading center would align with, and represents a continuation of, a series of such centers linked by transform faults which have been described further south in the Gulf of California (Atwater,1970 j and Anderson, 1971). I l

The existence of a spreading center and multiple stranding of the San Andreas f ault system in the Salton Trough tend to distinguish this region from areas further to the northwest where most activity is confined to a much narrower zone and where the largest earthquakes have occurred.

Within and bounding the Salton Trough, several northwest-trending fault strands are recognized including principally the Imperial, Calipatria, Brawley, Supers

ltion Mountain, Superstition Hills, and San Jacinto faults and, closer to the site, the San Andreas, Algodones, and San Hills faults. Based on seismicity, the most active of these appear to be the San Jacinto fault and the Imperial fault.

Northwest of the site, the Mojave Block is identified as an area bounded by the Garlock fault, part of the San Andreas fault, the eastern Transverse Ranges, and on the east by a less well-defined boundary, the Soda-Avawatz fault zone (Garfunkel, 1974). The Majave Block includes several northwest-trending, right-lateral, strike-slip faults which have undergone displacement in Quaternary time, such as the Helendale fault, the Lockhart fault, the Lenwood fault, the Camp Rock fault, the West Calico fault, the Pisgah fault and the Blackwater fault, These faults apparently do not represent through going structures and do not extend beyond the boundaries oftheMojaveBlock. Garfunkel (1974) suggested that this faulting has been produced by a distortion of the overall shape of the Mojave Block to accommodate lateral variations in crustal spreading between the area east and the area southwest of the Mojave Block. Because seismicity and faulting in the Mojave Block is lower in magnitude and rate of activity than in the San Andreas fault system, and because the Mojave Block comes no closer to the site than the San Andreas fault system, the largest earthquakes associated with the San Andreas fault system are expected to produce larger ground motions at the site than earthquakes in the Mojave Block.

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2.5.3 Site Geology The proposed site is located in the western part of the Palo Verde Valley on the Palo Verde Mesa west of the Colorado River in Eastern California. The site is i flanked on the west by the Mule Mountains, to the south by the Palo Verde Moun-l tains, and on the east by the Colorado River and the Dome Rock Mountains. To the north of the site is the continuation of the Palo Verde Valley and Mesa, in the I

site area (five mile radius), the Palo Verde Mesa is composed of a series of broad, gently sloping alluvial fans and fluvial terraces which slope 40 feet per mile to the east. The proposed site is situated on an alluvial fan surface and partly on a flat surface of the 70 foot terrace, one of two terraces above the present Colorado River level.

A north-south trending linear wash exists along the Pebble Terrace part of the Palo Verde Mesa. Reconnaissance geologic mapping by the California Division of Mines and Geology noted this lineation as a fault, but trenching of this feature revealed undisturbed sedimentary strata across the trend of the lineation. The lineation is due to a difference in erosion rate of the fluvial material and, tnerefore, is not a fault.

The section underlying the site has been investigated directly by borings and surface mapping and indirectly to basement rock, by gravity and magnetic analyses, by seismic refraction and by projections of units from surface mapping. The sub-surf ace investigation program included 51 drill holes with depths from 140 feet to 900 feet. Thirty-four of these drill holes were used for geological investigation while the others were used for foundation engineering assessment. Subsurface continuity of strata was based on correlation of drill logs and geophysical data, such as radiation logging, resistivity and potential measurements.

The section beneath the site area consists of Cretaceous plutonic and metamorphic I

basement rocks, overlain by Tertiary volcanic and fanglomeratic bedrock. These units are overlain by the Bouse Formation which is a Pliocene Marine sediment. ]

Surficial deposits at the site are Pliocene-Pleistocene alluvial deposits of the Colorado River, and Holocene alluvial and fluvial deposits and eolian sands.

Structure contour and isopach maps developed for the site area did not indicate the presence of any faulted stratigraphic units. Good correlations can be made in the site area using seven units, a silt lens, and four intra-unit clay horizons. To ,

the east and southeast, correlation becomes more difficult as the alluvial fan pinches out. Lateral variation within the units is common even over short dis-tances so correlation of detailed sub units is not feasible. Elevation changes are to be noted but no consistent anomalies are evident.

The applicant's seismic refraction survey and gravity and magnetic surveys indicat-ed no evidence of faulting. Displacement of sediments caused by vertical or lateral i

2-27

faulting could create sharp breaks or uiscontinuities to appear on the profiles, isometric drawings, and structure contour and isopach maps. The absence of such discontinuities is strong supportive evidence that there is no faulting beneath the site. No evidence of ground subsioence has been noted in the site area. There is no petroleum extraction and no mining activity or other man-made activities which would have any effect on the site.

At the construction permit stage of review, we require that an applicant provide a commitment to geologically map site excavations made for seismic Category I struc-tures. In response to this requirement, the applicant has provided the following commitment in a letter dated April 21, 1978:

" Site excavations made for Category I structures (including Category I buried pipelines) will be geologically mapped in detail, except that all excavations I for buried Category I pipelines will be inspected by a geologist, but only those areas where unexpected geologic conditions are encountered would be mapped. The mapped surfaces will include the excavation walls and floors. In addition, mapping will include geologic units or features exposed as a result of excavations made for other than Category I structures and pipelines if such mapping is required to adequately interpret the site geology.

The NRC Regulatory Staff will be notified shortly before completely mapping the soil units exposed in Category I structure excavations to permit schedul-ing of a Staff inspection." i l

We find this commitment acceptable with the exception that we require that'all excavations to be made for seismic Category I buried pipelines also be geologically I mapped regardless of the conditions encountered. Therefore, this matter remains )

outstanding. I 2.5.4 Surface Faulting j We have found no evidence to indicate that a potential exists for surface faulting at the site, The closest known capable fault is the Blythe Graben which is located 22 miles from the site and is discussed in Section 2.5.1 of this report.

2.5.5 Regional Seismicity The Sundesert site is located in an area of the Basin and Range province which apparently has experienced a relatively low level of historical seismic activity. q lt must be recognized, however, that the historical record in this area is short l compared to most areas of the United States, and that the population density in much of the Sonoran Desert area has historically been very low and remains low. A limited capability for en ihquake detection in this area has existed since the earliest seismograph stations were established in southern California in the late 2-28

1920's. The applicant estimates that the instrumented detection threshold since 1945, for earthquakes with epicenters in this area, is atout magnitude 4. (The size of earthquakes in the Western United States is typically classified by the units of magnitude on the Richter scale.) This detection and location capability has improved substantially in the past few yeara with installation of a dense seismograph network in the eastern Mojave desert, such that the current threshold level in the area is estimated to be as low as magr.itude 1.0.

4 1

Much of the earthquake activity in the Basin and Range province is concentrated near its eastern and western trargins as evidenced by the earthquake epicenters along the Wasatch Front and those in western Nevada and extending southward into California just to the east of the Sierra Nevada batholith.

Comparable zones of high seismicity are not apparent in the southern portion of the Basin and Range province in which the site is loc:ted. Exclusive of the Fort Yuma J earthquake, which is discussed in detail below, the earthquake reported nearest to the site occurred in 1943 about 30 miles southwest of the site and has an estimated magnitude from 4 to 4.5. The earthquakes reported nearest to the site, which are of magnitude 6 ano greater, were associated with the San Andreas fault system which l 1

approaches the site no closer than 35 miles. The largest earthquake in the histor-l ical record associated with these southern splays of the San Andreas fault system was the Imperial Valley earthquake of 1940, which had a magnitude of 7.1 and occurred '

on the Imperial fault approximately 60 miles southwest of the site.

o l During the course of our review, several questions were raised regarding an earth-1 1 quake which occurred in the vicinity of fort Yuma in 1852. Because this earthquake l i occurred so early in the history of southern California at a time when the area was virtually undeveloped, detailed information regarding this earthquake was not easily attainable. The main qt.estions raised were with regard to the date, loca-tion, and structural association of the Fort Yuma earthquake. Conflicting reports regarding these points exist in the published accounts for this earthquake. This is a problem which is generally encountered when one attempts to obtain information about earthquakes which occurred in a region prior to or during its early develop-l ment. Specific information to unequivocally determine the location of such an earthquake and demonstrate its structural association is usually not available.

I In this case, the applicant conducted a careful literature search and was able to identify the primary sources for the published reports on this earthquake. These sources consisted of diaries kept by two military officers stationed at fort Yuma, a report published in 1861 on the Colorado River expedition of 1857 and 1858 (lves, 1861), and two newspaper accounts of effects felt at large distances. In addition to the literature search on the fort Yuma earthquake, the applicant investigated reports of similar phenomena observed during more recent earthquakes in this area of the San Andreas fault system, such as the 1940 Imperial Valley earthquake and two earthquakes in 1915 and one in 1934 located in the Salton Trough. The applicant

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argued that geyser activity, ground cracking and liquefaction, which occurred 1

southwest of Fort Yuma during the 1852 earthquake, should be regarded as the pri-i mary indicator of proximity to the epicenter. The applicant further contended that i

the rock fall at Chimney Peak (Picacho Peak), which occurred at the time of the s

earthquake, should be discounted because the weathered condition of the Peak made is susceptible to rock falls at relatively low levels of motion. I 4

As a result of analysis of data gathered in the literature search and consideration 1

i of the history of earthquake activity in this area, the applicant concluded that:

j (1) The Fort Yuma earthquake occurred on November 29, 1852 at approximately noon.

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(2) The epicenter of the earthquake was located in the Salton Trough.

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l (3) The magnitude of the earthquake is estimated to have b o n between 6 and 7.

As a result of our review of data on the Fort Yuma earthquake and knowledge of seismicity and tectonics in the area, we have concluded that:

s (1) The date for the Fort Yuma earthquake determined by the applicant is accurate.

4 J

(2) It is reasonable to assume that the Fort Yuma earthquake was associated with structures of the San Andreas fault system.

8 (3) The Fort Yuma earthquake was probably no larger than other earthquakes which have occurred in this area of the San Andreas fault system.

i.

j Besides the arguments provided by the apphcant cited above, the prime data sup-i porting our conclusions are (1) the relatively high seismicity in the Salton Trough j

and virtual absence of seismicity to the northeast of this area, (2) the existence of several faults with Quaternary displacement within the Salton Trough and relative scarcity of evidence of recent fault displacement to the northeast of this area, and (3) the existence of major, plate bounding faults in the Salton Trough and lack of similar features to the northeast of this area.

2.5.6 Design Basis Earthquakes As already noted in Section 2.5.5 of this report, the historical record of seismic activity in the southern portion of the Basin and Range nrovince is poor. Because of this, it is necessary to rely primarily on the recognition of active faulting in i establishing the safe shutdown earthouake for the Sundesert site.

The majority of earthquakes which have occurred in the Basin and Range province can be reasonably associated with mapped faulting. In particular, what was probably the largest earthquake in the province, the Owens Valley earthquake of 1872, produced 2-30

surface ruptures at the time of the earthquake ($lemmons, 1967; Bonilla, 1967).

Many of the other large earthquakes in Pleasant Valley, Nevada, and the 1954 earth-quakes at Fairview Peak and Dixie Valley, Nevada, also are reported to have produced surface displacements. Because of this association between earthquake activity and faulting, according to the criteria of Appendix A to 10 CfR Part 100 it is not necessary to assume that earthquakes in the Basin and Range prevince can occur closer to the site than the faults with which they can be reasonably associated.

In connection with our geology and seismology review of the Palo Verde nuclear power plant site, it was determined that the largest earthquake in the Basin and Range tectonic province, which could nut reasonably be ass niated with faulting, has a magnitude of 4. The applicant has conservatively assumed a magnitude 5 earthquake could occur near the site, at a distance of five miles, in establishing the safe shutdown earthquake.

Except for the Sundesert site area and a few other scattered areas, only reconnais-sance geologic mapping has been conducted throt.g % ut much of southeastern California and most of the western half of the State of Arizona. The applicant has conducted state-of-the-art geologic investigations in the vicinity of the site. Based on the applicant's investigations and the results of reconnaissance mapping in the region, the fault nearest the site which is considered to be capable is the Blythe Garben, 22 miles from the site. As discussed in Section 2.5.1 of this report, the Blythe Graben has a traceable length of three and a half miles but can be inferred to be longer based on gravity measurements. Based on interpretation of the gravity data, the Blythe Graben has been inferred to be on the northeast side of a structural trough about 25 miles in length, whose southwest side is about 15 miles northwest of the site. Though the southwest side of the structural trough may be inferred to be related in the mechanism of its origin to the northeast side, the southwest side has not been assumed to be capable because of the lack of evidence of Quaternary f ault displacement on the southwest side of the trough. The applicant assumed a magnitude fi.5 earthquake could occur on the Blythe Graben 22 miles from the site.

Given the relatively short length (approximately 25 miles) of the structure and lacking evidence of associated seismicity, the applicant's assessment appears conservative when compared to existing correlations between earthquake magnitude and fault length.

Capable faulting is known to exist in the area of the San Andreas fault system southwest of the site. The San Andreas fault system extends from the Gulf of California on the southeast to Cape Mendocino on the northwest, a distance of about 700 miles. The length of the southern San Andreas fault system from the bend near the Garlock fault to the Gulf of California is about 300 miles. The southern part of the system has several splays.

The largest earthquake which has occurred on the San Andreas fault system was the 1906 San Francisco earthquake with an estimated magnitude of 8.3. An earthquake of 2-31

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estimated magnitude 8 occurred in 1857 at Fort Iejon near the intersection of i

Carlock fault and the San Andreas fault, producing surface displacements north and j ,

south of this intersection. This earthquake has been associated with the northern

}. portion of the San Andreas fault system since the geologic characteristics of the j faultsystemnearthisintersectionandthecharacteristicsofthefortTejon i

earthquake are more representative of those associated with the northern San Andreas j fault system. '

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a The largest earthquakes on the southern San Andreas fault system were slightly j larger than magnitude 7. These include the 1915 Baja California earthquake, the

} 1934 Baja California earthquake, and the 1940 Imperial Valley earthquake, all of i ,

magnitude 7.1, and the 1903 Baja California earthquake listed as magnitude 7 plus.

The fault strands in the San Andreas fault system clo'sest to the site are about 35 1 miles to the southwest in the Salton Trough. The applicant assumed a magnitude 8.5 earthquake could occur on these structures 35 miles from the site. This assumed earthquake is larger than any reported for California. Based on relations between magnitude and length of surface fault rupture during earthquakes, a magnitude 8.5 l corresponds to a surface rupture length of about 300 miles. Based on these con-siderations, an earthquake producing surface rupture along the entire length of the I suuthern San Andreas fault system; i.e., from the Gulf of California to the bend near the Garlock fault, could reasonably be expected not to exceed magnitude 8.5.

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Considering that (1) earthquakes in tne historical record for the southern San Andreas fault system have not had magnitudes exceeding about 7.1, (2) the largest earthquake in the historical record anywhere on the San Andreas fault system had a magnitude of 8.3, (3) total offset in the San Andreas fault system may be distrib-uted over multiple strands in the southern San Andreas system, and (4) the more I active strands within the Salton Trough are further to the southwest, the assump-tion of a magnitude 8.5 earthquake on northeast strands of the San Andreas fault system 35 miles from the site appears conservative.

The applicant has proposed to use the response spectra defined in Regulatory Guide 1.60,

" Design Response Spectra fut Nuclear Power Plants," to define the characteristics of the safe shutdown earthquake. The horizontal response spectra are to be normalized to 0.35g, and the vertical response spectra are to be normalized to 0.23g. Several different scenarios were evaluated in assessing the adequacy of a herizontal acceler-ation level of 0.35g for the safe shutdown earthquake:

(1) A magnitude 5.0 earthquake was assumed to occur near the site, beyond the l

region of intense geologic investigations conducted within five miles of the site. Based on empirical relations between magnitude, epicentral distance, and acceleration, the peak acceleration due to this earthquake would be expected to be between about 0.079 and 0.15g.

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(2) Historical earthquakes associated with mapped faulting in the Basin ano Mange province were assumed to occur on those faults at their closest mapped posi-tions to the Sundesert site. All such earthquakes had magnitudes less than 8.3, the estimated magnitude of the Owens Valley earthouake of 1872, and the associated faults are sufficiently distant from the sita so that the peak accelerations resulting at the site from such earthquakes would be expected to be less than 0.35g.

(3) A magnitude 6.5 earthquake, asscciated with the Blythe Graben, was assumed to occur 22 miles from the site. The peak accelerations calculated from acceleration-magnitude-distance relationships for this event are between about 0.lg and 0.25g.

(4) A magnitude 8.5 earthquake, associated with the San Andreas fault system, was assumed to occur 35 miles from the site. Peak accelerations for this event calculated from acceleration-magnitude-distance relationships are between about 0.19g and 0.35g.

(5) Effects at the site due to potential earthquakes in the Mojave Block were also ,

considered. As discussed in Section 2.5.2 of this report, peak accelerations at the Sundesert site from earthquakes in the Mojave Block are expected to be less than that for earthquakes associated with the San Andreas fault system.

Therefore, the horizontal acceleration level proposed for the safe shutdown earth-quake is as great as, or greater than, the peak accelerations which would be expected to result at the site due to any of the postulated earthquakes.

Trifunac and Brady (1975) developed empirical relationships between earthquake intensity and peak acceleration for both horizontal and vertical components of motion. By a comparison of the relationship for peak horizontal acceleration to

! the relationship for peak vertical acceleration, the peak vertical acceleration is seen to be somewhat less than two-thirds the peak horizontal acceleration. Based on this comparison, the vertical acceleration level of 0.239 proposed for the safe shutdown earthquake is as great as tne peak vertical acceleration which would be expected to occur at the site from an earthquake producing a peak horizontal accel-3 eration of 0.35g, i.e., a magnitude 8.5 earthquake occuring 35 miles from the site.

Therefore, we conclude that the applicant's proposed horizontal and vertical accel-l eration values of 0.35g and 0.23g, respectively, for the safe shutdown earthquake are acceptable for the Sundesert site As an additional check on the adequacy of the proposed safe shutdown earthquake, the applicant developed response spectra from strong motion time histories for four earthquakes recorded at firm-soil sites thought to be most representative of the conditions at the Sundesert site. For each of the earthquakes; i.e., the 1952 Kern 2-33 4

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County earthquake recorded at Taft, the 1940 Imperial Valley earthquake recorded at El Centro, the 1933 Long Beach earthquake recorded at Vernon, and the 1971 San Fernando earthquake recorded at Whittier Narrows; the horizontal and vertical components of strong motion were scaled using acceleration-magnitude-distance relationships The response spectra were determined and compared to the response spectra in Regulatory Guide 1.60 scaled to 0.35g (horizontal) and 0.23g (vertical).

In general, the response spectra in Regulatory Guide 1.60 envelope the response spectra for the real earthquake records with the exception of the El Centro spectra which slightly exceed the spectra in Regulatory Guide 1.60 at a few frequencies.

The vibratory ground acceleration values for the operating basis earthquake, which are taken to be one-half the vibratory ground acceleration for the safe shutdown earthquake, are consistent with the guidelines of Appendix A to 10 CFR Part 100.

Therefore, we find them acceptable.

2.5.7 Conclusions Based on our review of the geology and seismology for the proposed Sundesert site, we conclude that (1) there are no geological structures that would tend to localize earthquakes in the immediate vicinity of the sitt or cause surrace faulting at the site, (2) there are no known geologic features at the site which could represent a potential hazard due to solution activity and/or subsidence, and (3) the seismic l design bases are appropriately conservative for the earthquake potential at the site. Therefore, we conclude that the proposed Sundesert site is acceptable with regard to geology and seismology considerations. As stated in Section 2.5.3 of this report, we require that the applicant commit to geologically map all excavations to be made for seismic Category I structures, including buried pipelines.

2.6 Geotechnical Engineering The adequacy of the soils investigation program and the safety of foundations, as presented in the Sundesert Preliminary Safety Analysis Report, have been evaluated in accordance with the criteria outlined in Appendix A to 10 CFR Part 100 and Sections 2.5.4 and 2.5.5 of the Standard Review Plan. The following sections summarize our evaluation of these geotechnical engineering considerations.

2.6.1 Stability of Subsurface Materials and Foundations Site Conditions i

The plant site is situated on the Palo Verde Mesa, partly on an even, gently slop-ing alluvial fan surface and partly on a flat surfaced remnant of a river terrace, and is bounded on the north and south by two broad washes, 15 to 25 feet deep and one quarter to one-half mile wide. Elevations at the plant site range from 385 j feet above mean sea level in the northwest corner to 365 feet above mean sea level 2-34

in the southeast corner. The proposed final plant grade will be 375 feet above mean sea level.

To establish engineering properties of subs.rface soils beneath seismic Category I structures, fifty-eight rotary wash borings sere drilled within the proposed plant area to depths ranging from approximately 90 to 550 feet below the existing ground surface. In addition, four large diameter bucket auger borings were drilled near the centers of the proposed locations for major structures in order to obtain bulk samples for backfill evaluation studies. Two types of in-situ geophysical investi-gations were performed to evaluate the subsurface soil conditions at the site: (I) seismic velocity surveys were used to determine the propagation velocities of compressional and shear waves in the soils; and (2) borehole geophysical logging techniques were used to measure electrical, radiatiun and physical soil properties for evaluating stratigraphic correlations across the site. ,

The borings made within the site area show a subsurface profile consisting of approximately 380 feet of dense to very dense granular soils underlain by a hard, overconsolidated marine clay deposit. All borings in the immediate site area were terminated in the marine deposit at a depth of approximately 500 feet. A boring, located 1500 feet southwest of the site area, extended below the marine clay and encountered a cemented clayey gravel deposit at a depth of 740 feet and a dense fanglomerate at a depth of 860 feet. The granular soils present in the upper 380 feet of the soil profile in the site area consist primarily of fine to medium alluvial sands. Layers up to 40 feet thick, containing significant amounts of coarse sand and gravel, are present at depths of 50, 110, and 270 feet.

l Ground water levels in the site area have been monitored on a monthly basis since late 1974. Since monitoring of water levels began, seasonal fluctuations have been less tha'i one foot from an average ground water elevation of 230 feet above mean sea level. This level is approximately 100 feet below the bottom of the deepest I proposed excavation. Therefore, dewatering is not a construction requirement at the Sundesert site. Based upon well pumping test results, minimal drawdown would result from ground water withdrawal for potable and fire fightirg use. We con-clude, therefore, that subsidence or rebound of the ground surface related to ground water withdrawal will not be a problem. Section 2.4.4 of this report provides additional discussion on ground water.

Excavation and Backfill Seismic Category I structures will be supported on dense in-situ soils with the exception of the diesel generator fuel oil pumphouse,,the eastet portion of the control building (electrical tunnel) and buried service water pipelines and elec-trical ducts which will be supported on structural backfill. Structural backfill will also be used arotad most seismic Category I structures. Two general types of i

sands derived from onst'e excavations will be used for all backfill: (1) relatively 2-35

well graded silty sands encountered at shallow depths; and (2) uniformly graded clean sands encountered below 10 feet. The clean sands will be used for structural backfill while the silty sands will be used as random fill and as a thin cover layer at the site to orovide traf ficability for construction equipment.

The applicant states that the structural backfill to be used beneath seismic Category I structures 3nd the fill to be placed in the construction of the site drainage dike will be compacted to at least 95 percent of the maximum dry density as determined by the American Society for Testing Materials (ASTM) Standard D-1557, " Moisture Density Relations of Soils Using 10 Pound Rar 'er and 18 Inch Drop." However, we require that these structural backfill and fi materials be compacted to either (1) the applicant's above proposed criterion, or (2) an average of 85 percent but not less than 80 percent relative density as determined by ASTM Standard D-2049,

" Relative Density of Cohesionless Soils." whichever results in the higher in place dry densit! Therefore, this matter remains outstanding.

Foundation Stability Bearing capacity analyses were performed to estimate factors of safety against i

ultimate foundation failure under static and dynamic loads. All seismic Category 1 l structures will be supported on mat foundations at depths of from 2.5 feet for the l standby diesel generator fuel storage pumphouse to 46 feet for the service water cooling towers. The average foundation contact pressures will range from 1500 pounds per square foot for the pumphouse to 7000 pounds per square foot for the l containment structure. The foundation soil shear strength parameters used in the analyses are a friction angle of 35 degrees with zero cohesion. The minimum static bearing capacity factor of safety was 52.0 for the fuel oil pumphouse, and the minimum dynamic bearing capacity factor of safety was 3.7 for the same structure.

The results of the bearing capacity study show that the foundation soils ara com-petent and will adequately support the loads of the proposed structures.

Heave and settlement analyses were performed to est% ate the magnitude of vertical ground deflections at points beneath seismic Category I structures which will be caused by construction of the plant structures. The estimated maximum heave due to excavation was found to be 1.0 inch under the center of the containment structure.

Estimated total recompression settlements ranged from 0.22 inch at the corner of the standby diesel generator building to 1.73 inches at the center of the contain-ment structure. A system of instrumentation, consisting of borehole extensometers, mechanical rebound anchors, and settlement markers in structural concrete, will be established to monitor the heave and settlement during and following construction.

We will evaluate the m,easurements of the actual settlement following construction at the operating license stage of review.

Static and dynamic lateral earth pressure coefficients for active, passive and at rest conditions have been calculated for lateral earth pressure design purposes.

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Static active and static passive coefficients were calculated based on the Coulomo theory and are estimated to be 0.21 and 10.5, respectively. For very dense backfill,.

I the static coefficient of lateral earth pressure at rest is estimated to be 2.0 near the backfill surface due to comoaction ef forts but is estimated to decrease to 0,4 with depth. Lateral earth pressure due to earthquake loading were computed )

. using the Mononobe-Okabe modification of the Coulomb theory. Based on our review, {

we conclude that conservative methods have been applied to the calculation of lateral earth pressure coefficients. ,

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8ecause compaction of cohesionless soils can occur when they are subjected to strong ground motion, dynamic subsidence studies were conducted to evaluate this condition. Dynamic subsidence was calculated for both the upper 150 feet of dry j sand and the underlying 200 feet of saturated sand, using as dynamic input the acceleration time history records of four earthquakes scaled to a ground surface acceleration of 0.35g. The maximum total dynamic subsidence from these studies is estimated to be 3.1 inches distributed uniformly across the plant site. In order to monitor areal subsidence or rebound, a series of three primary benchmarks will be installed on outcrops of bedrock close to the site area. In addition, a network of secondary benchmarks arranged in a grid pattern will be installed in the plant area. The elevations of the secondary benchmarks will be surveyed on a bi-monthly basis during plant construction and annually after completion of the plant and during its service life.

An evaluation of the liquefaction potential of the deep saturated sands at the site wac ,arformed by considering the following aspects. The ground water table is at a depth of approximately 150 feet below the proposed plant grade. Results of labora-tory relative density analyses on undisturbed samples of the saturated sands indicate relative densities of from 80 to 100 percent. An analysis of the liquification potential, performed by comparing shear stresses induced by the four earthquakes scaled to a ground surface acceleration of 0.35g to the available shear strength of the soil obtained from strength curves developed by Seed and Idriss (1971), indicate factors of safety ranging from 1.5 to 3.1.

Based on our review of the soil conditions and studies outlined above, we conclude that there are adequate safety margins against sharp differential settlements due to dynamic subsidence and against liquefaction for ground acceleration levels up to 0.35g.

2.6.2 Stability of Slopes There are no natural soil or rock slopes in the vicinity of the site, since the natural ground surface slopes gently for several thousand feet in all directions.

The only man-made slope will be that associated with the site drainage dike which is discussed in Section 2.6.3 of this report.

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I l 2.6.3 Embankments A site drainage dike, 10,000 feet long, will be designed so that upland runoff due to precipitation will not inundate the safety-related facilities. The dike will be located immediately up-slope of the plant and will be a five foot high, compacted j fill, diversion structure. The dike will be designed with three to one horizontal-j to vertical slopes and a five foot wide crest which will all be covered by a six l inch layer of gravel. The upland side of the dike will have an 18 inch layer of l riprap over ti.e gravel and a filter cloth under the gravel to prevent migration of j fine soil. The dike will be designeo to be stable if subjected to: (1) the com-bined effects of the operating basis earthquake and one-half the probable maximum

( flood and (2) the combined effects of the safe shutdown earthquake and the 25 year f flood. Based on our review, we conclude that the proposed design of site drainage j dike is acceptable.

i i 2. 6. Conclusions i

i l Subject to the satisf actory resolution of the matter discussed in Section 2.6.1 cf l 1

this report relating to the density of structural backfill, we conclude that the I geotechnical engineering aspects of the proposed power plant are adequate to meet the requirements of 10 CFR Part 100 and are acceptable.

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3.0 DESIGN CRITERIA FOR STRUCTURES, SYSTEMS AND COMPONENTS 3.1 Conformance With General Design Criteria In Section 3.1.2 of the Preliminary Safety Analysis Report, the applicant provided l

a discussion of how the Sundesert plant will conform to each of the General Design Criteria for Nuclear Power Plants (Appenoix A to 10 CFR Part 50). We will report our evaluation of the conformance of the proposed plant design to the requirements of the General Design Criteria in a future Safety Evaluation Report should the staff be requested to reactivate its review.

L2 Classification of Structures, Systems and Components 3.2.1 Seismic Classification Criterion 2 of the General Design Criteria requires that nuclear power plant struc-tures, systems and components important to safety be designed to withstand the effects of earthquakes without loss of capability to perform their safety function.

These plant features are those necessary to assure (1) the integrity of the reactor coolant pressure boundary, (2) the capability to shut down the reactor and maintain

-it in a safe shutdown condition, or (3) the capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures compa-rable to the guideline values of 10 CFR Part 100.

Structures, systems and components that will be designed to remain functional if a safe shutdown earthquake should occur and which are designated as seismic Category I have been identified in Table 3.2.5-1 of the Preliminary Safety Analysis Report.

All other structures, systems and components that may be required for operation of the facility will be designed to other than seismic Category I requirements.

Included in this latter classification are those portions of seismic Category I systems which are not required to perform a safety function.

We have reviewed the applicant's seismic classification of the structures, systems and components important to safety and conclude that they have been identified in an acceptable manner and classified as seismic Category I items in conformance with Regulatory Guide 1.29, " Seismic Design Classification," except for the following items.

The applicant has classified the fuel building filtration system and the off-mat portion (away from the fuel storage area) of the fuel building, including the building isolation dampers, as non-seismic Category 1. The applicant will design l 3-1

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g the fuel storage portion of the fuel building as seismic Category I and will design 4 all of the fuel building to be able to safely withstand the effects of the design i basis tornado winds, including tornaco generated missiles.

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} To determine the potential effects of the applicant's proposed design in this area, t

we evaluated the radiological consequences of a postulated fuel handling accident 7 in the fuel building assuming no creoit for the filtration system. The results of this evaluation, which are presented in Section 15.5.2 of this report, indicate

that the calculated thyroid dose at the exclusion area boundary is sufficiently

high to warrant additional design provisions to assure that, at the operating i license stage of review, the calculatea consequences of a postulated fuel handling j

{ accident within the fuel building are well within the guideline values of 10 CFR

{ Part 100, Accordingly, we require that the fuel building and the building filtratior l system and isolation dampers be designea as seismic Category I. Therefore, this j matter remains outstanding. Additional discussions on the fuel building filtration j system are presented in Sections 6.5,3, 9.4.2 and 15.5.2 of this report.

} Subject to the satisfactory resolution of the above matter regarding the seismic a

j classification of the fuel building and its filtration system and isolation dampers, l l we conclude that structures, systems and components important to safety, that will i be designed to withstand the effects of a safe shutdown earthquake and remain l ll functional, have been classified as seismic Category I in conformance with the i Commission's regulations as set forth in Criterion 2 of the General Desig9 Criteria l and Regulatory Guide 1.29, and industry codes and standards and are acceptable.

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1 3.2.2 system Quality Group Classification 1

! l j Criterion 1 of the General Design Criteria requires that nuclear power plant systems 3

j and components important to safety be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety function to be i performed. I

, We have reviewed the applicant's classification system for pressure-retaining l components, such as pressure vessels, heat exchangers, storage tanks, pumps, piping, l ind valves in fluid systems important to safety, and the assignment by the applicant of safety classes to those portions of systems required to perform safety functions, j The applicant has applied the classification system of the American Nuclear Society j (Safety Classes 1, 2, 3, and Non-Nuclear Safety), which corresponds to the Commis-l sion's Quality Groups A, B, C and D in Regulatory Guide 1.26, " Quality Group l l Classifications and Standards," to those fluids containing components which are part of the reactor coolant pressure boundary and other fluid systems important to safety where reliance is placed on these systems to (1) prevent or mitigate the consequences of accidents and malfunctions originating within the reactor coolant pressure boundary, (2) permit shutdown of the reactor and maintain it in a safe l 3-2

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shutdown condition, anu (3) contain radioactive material. These fluid system I components have been identified in an acceptable manner and classified in conformance l with Regulatory Guide 1.26 in Table 3.2.5-1 of the Preliminary Safety Analysis Report. Piping and valves for these fluid systems have also been classified in an acceptable manner on system piping and instrumentation diagrams in the Preliminary Safety Analysis Report.

Fluid systems pressure-retaining components important to safety that are classified as Quality Group A, B or C will be constructed to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code as follows:

Component Code Quality Group ASME Section III, Division 1 A Class 1 B Class 2 C Class 3 Quality Group A components will comply with Section 50.55a of 10 CFR Part 50.

Quality Group B and C components will comply with Subsection NA-ll40 of the ASME Code.

Components that are classified as Quality Group D will be constructed to the following l codes as appropriate; ASME Boiler ano Pressure Vessel Code, Section 111, Divisions 1 or 2, American National Standards Institute (ANSI) Standard B31.1-1974 or 1977,

" Power Piping," and manufacturer's standards.

The basis for acceptance in our review has been conformance of the applicant's designs, design critet ia, and design bases for pressure-retaining components, such as pressure vessels, heat exchangers, storage tanks, pumps, piping, and valves in fluid systems, important to safety with the regulations as set forth in Criterion 1 of the General Design Criteria, the requirements of the codes specified in Section 50.55a of 10 CFR Part 50, Regulatory Guide 1.26, and industry codes and standards.

We conclude that fluid system pressure-retaining components important to safety will be designed, fabricated, erected, and tested to quality standards in conformance with the Commission's regulations, the applicable regulatory guides, and industry codes and standards and, therefore, are acceptable.

3. 3 Wind and Tornado Design The applicant has specified a design wind of 80 miles per hour, based on a recurrence interval of 100 years, and design basis tornado characteristics, based on Tornado Intensity Region 11 in Regulatory Guide 1,76, " Design Basis Tornado for Nuclear Power Plants." As stated in Section 2.3.1 of this report, we find these values to be acceptable for the region in which the proposed plant will be located.

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The applicant has also described the procedures that will be used to transform the design wind velocity and the tt.rnado wind velocity into pressure loadings on struc-tures which would be subjected ta these winds, and has further described how these loads will be combined with other applicable loads in order to design these structures.

We have not completed our review of these procedures and the methods for combining loads. Therefore, this matter remains outstanding.

3.4 Water Level (Flood) Design Our review has included the applicant's proposed design criteria and design basis for safety-related systems, structures and components, the adequacy of those criteria and bases, and the requirements to maintain the capability for a safe plant shutdown during a design basis flood.

The plant grade for the Sundesert plant will be located at a minimum elevation of 375 feet above mean sea level. Since the probable maximum flood is computed to be 312 feet above mean sea level and since normal ground water level is 234 feet above mean sea level, none of the seismic Category I structures and safety related systems will be subject to flooding since they will be located above the calculated probable maximum flood level.

l Access to structures will be located above grade. All penetrations through exterior walls below grade will be sealed to prevent any water inleakage.

As a result of our review, we conclude that the criteria and bases for the water level (flood) design are in accordance with Criterion 2 of the General Design Criteria, with regard to protection from damage due to the probable maximum flood, and Regulatory Guide 1.27, " Ultimate Heat Sink for Nuclear Power Plants," as it relates to protection of the ultimate heat sink from damage due to the probable maximum flood. Therefore, we conclude that the design criteria and bases are acceptable.

3.5 Missile Protection 3.5.1 Missile Selection and Protection Criteria The facility design requirements consider the possibility of missiles being generated from pressurized piping and vessels, rotating equipment, and tornadoes. Protection of safety-related components and structures will be provided by orientation and separation from missile generating sources, and by the use of adequate barrier or energy absorbing materials. Engineered safety features systems will be separated in a manner such that the failure of one train cannot cause the f ailure of its redundant train, or that the failure of any plant component which brings about the need for these engineered safety features systems does not render the safety system inoperative.

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l We have concluded, in Section 2.2 of this report, that nearby industrial, transpor-tation, and military facilities do not represent hazards requiring protection features in the facility design. We have categorized the other missiles considered by the applicant as (1) missiles generated by postulated failures of facility equipment, (2) missiles generated by postulated tornadoes, and (3) missiles generated by postulated failure of the turbine generator. The results of our review of the !

applicant's missile selection are as follows:

Facility Equipment Generated Missiles The potential sources of missile generation within containment identified by the applicant are: parts of the control rod drive mechanism; valve bonnets located in the region where the pressurizer will extend above the operating floor (pressurizer safety valves, motor-operated isolation valves in the relief line, air-operated relief valves and air-operated spray valves); temperature and pressure sensor assemblies connected to the reactor coolant system; and pressurizer heaters. The missile characteristics for the above components have been evaluated by the applicant to aid in its design of barriers to withstand the effects of missiles. In identifying ;

the potential sources for missile generation, the applicant did not consider the reactor coolant pump flywheel as a source because of the measures taken to assure the integrity of the reactor coolant pump flywheel during a postulated loss-of-coolant accident, as discussed in Section 5.4.1 of this report.

In assessing the effects of internally generated missiles in other areas of the facility (i.e., outside containmeit), the applicant states that missiles will be postulated and evaluated to assure '. hat the plant separation and barrier design is 1

adequate. However, the applicant states that it does not consider valves in high pressure systems and motor-driven pumps and fans outside containment to be potential sources of missiles.

We do not agree with the applicant's conclusion on this last matter and will require that the applicant also postulate and evaluate potential missiles generated within other regions of the facility from overspeed of rotating machinery and from failure of pressurized components and systems to assure the adequacy of the plant separatior features and barrier design. Therefore, this matter remains outstanding.

Subject to the satisfactory resolution of the above matter, we conclude that the selection criteria for facility equipment generated missiles conform with the Commission's regulations and applicable regulatory guides and are acceptable.

Tornado Missiles Criteria 2 and 4 of the General Design Criteria require that structures, systems I and components important to safety be designed to withstand the effects of natural phenomena su:h as tornadoes, including the effects o" missiles, without loss of 4

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capability to perform their saf ety funecion. Section 3.5.1.4 of the Standard Review Plan specifies the tornado missile spectra against which a nuclear power plant should be protected.

Revision I to Section 3.5.1.4 of tht 5*. ann d Review Plan contains the missile spectrum defined in the original Stanaard e view Plan section (Revision 0 spectrum) as well as alternate missile spectra (Revision 1 spectra). The applicant for any construction permit application docketed before June 1, 1977 has the option of selecting, for plant design purposes, the Revision 0 tornado missile spectrum in total or either of the two Revision i tornado missile spectra in total. The Sundesert application is in this category and the applicant has chosen the Revision 0 spectrum.

The Sundesert site is located in lornado Negion II, as defined in Regulatory Guide 1.76. The wall thickness, for structures and barriers used to orovide tornado missile protection, will be 24 inches thick and will be constructed of steel rein-forced concrete having a strength of 4000 pounds per square inch. The roof thickness of structures and barriers used to nrovide tornado missile protection will be 21 inches thick and will be constructed of steel reinforced concrete having a strength of 4000 pounds per square inch. These roof and wall thicknesses and concrete strength will provide adequate tornado missile protection for structures, systems and components important to safety for a plant located in Tornado Region II.

We have independently verified the applicant's assessment of the hazards from missiles generated by natural phenomena at the site. Based on the likelihood and severity of these phenomena and the protection provided against tornado generated missiles, we have determined that the probability of an accident at the proposed plant due to these phenomena having radiological consequences worse than the exposure g 'delines of 10 CFR Part 100 is less than 10-7 per year. We conclude, therefore, tiat the construction and operation of the Sundesert plant on the proposed site will not present an undue risk to the health and safety of the putilic from missiles generated by natural phenomena, including tornadoes.

Turbine Missiles The turbine generator for each unit of the Sundesert plant will be located in a peninsular orientation such that all safety-related systems and/or structures associated with that unit will be outside the potential low trajectory turbine missile strike zone for that unit's turbine generator. However, the turbine generator of each unit. will be adjacent to and parallel to the turbine generator of the other unit and, therefore, some safety-related systems and/or structures for one unit will be within the potential low trajectory turbine missile strike Zone for the other unit. Our evaluation of this arrangement is presented below.

In order for a low trajectory turbine missile to strike any safety-related equipment, the angle of the trajectory would have to be less than 10 degrees relative to a 3-6

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, plane parallel to the turbine operating floor and passing through the centerline of the turbine generator. The probability of a turbine missile having an angle of less than 10 degrees is 0.11. In addition, the only wheel that could generate a damaging missile would be the end wheel at the north end of each turbine generator as proposed for the Sundesert plant. The probability that the end wheel would I

f ail, given a turbine f ailure, 'is 0.02. The probability that a low trajectory missile would be generated as a result of destructive overspeed is taken to be 4.5 x 10 -5 based on historical data. Therefore, the overall probability that a lov trajectory turbine missile would strike and do damage to any safety-related equip-ment is conservatively estimated to be 5 x 10'I per year per turbine.

Based on information presently available, we conservatively estimate the risk from

~7 This high trajectory turbine missiles to be about 10 per year per turbine.

estimate is based on: (1) a conservative probability for high trajectory missile

~4 -3 generation of 10 per year per turbine, (2) a strike prubability of i.3 x 10 , j and (3) a penetration and damage probability conservatively assumed to be 1.

l Based on our review and on the results of our conservative calculations, we conclude ;

that the overall probability for turbine missiles damaging the plant and leading to l consequences in excess of 10 CFR Part 100 exposure guidelines meets the acceptance criteria of Section 2.2.3 of the Standard Review Plan. Therefore, we conclude that the plant'5 essential systems will be adequately protected against turbine missile damage.

3.5.2 Barrier Design Procedures i

i The applicant has provided information to describe the procedures that will be used in the design of structures, shields and barriers to resist the effects of missiles.

The procedures consider both local effects of the missile impact and the overall structural response of the impacted target.

We have not completed our review of the applicant's barrier design procedures.

Therefore, this matter remains outstanding.

3.6 Protection Against Dynamic Effects Associated With the Postulated Rupture of Piping The applicant has provided information concerning the break location criteria and the methods of analysis for evaluatino the dynamic effects associated with postulated breaks and cracks in high and moderate energy fluid system piping, within the containment and within other facility structures. The design criteria, methodology and bases for protecting essential systems from the effects of postulated piping failures both within the containment and within other structures have been descrioed.

These criteria and bases will be used in the design to assure that the functions necessary to maintain the capability nf a safe plant shutdown during any failure of high or moderate energy systems will be preserved.

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Design basis pipe breaks will be postulated to occur in all high energy piping 5

systems (or portions of systems) in accordance with Section 3.6.2 of the Standard Review Plan. High energy piping breaks will be analyzed for damage to essential f equipment due to pipe whip, jet impingement and environmental effects. Moderate 8

energy piping cracks will be analyzed for flooding, spray, and environmental effects only. The Sun 6 sert plant general arrangement and the layout of the high energy j systems will utilize combinations of physical separation, pipe enclosures, pipe whip restraints ar.d equipment shields.

1 1

j The p Snt design basis will include the ability to sustain a high energy pipe break l acrident coincident with a single active failure and retain the capability for a l 'A fe cold shutdown. Pipe motion subsequent to rupture and the pipe restraint !

, dynamic interaction will be analyzeo by the use of an elastic plastic lumped mass l

beam element model sufficiently detailed to reflect the strut .1 characteristics of the piping system.

1 lhe criteria to be used to define break location and configuration in the reactor

coolant loop are as described in Westinghouse Topical Report WCAP-8082-P-A, " Pipe j Breaks for the LOCA Analysis of the Westinghouse Primary Coolant Loop," which we j previously reviewed and found acceptable as documented by letter to Westinghouse,

dated May 22, 1974 Actual seismic loads for the Sundesert site will be included in the reactor coolant loop analysis. The stresses and cumulative usage factors resulting from these seismic events will be included in the analysis, which will be

presented in the Final Safety Analysis Report, to verify the design basis break j locations in the reactor coolant loop.

4 The applicant will provide protection against pipe failure in other areas of the j facility in conformance with the criteria contained in Branch Technical Position j APCSB 3-1, " Protection Against Piping Failures in Fluid Systems Outside Containment."

] At our request, the applicant has provided an analysis to show how the residual j heat removal system will be protected from the effects of postulated piping system j failures as an example of how the criteria of Branch Technical Position APC5B 3-1 l

will be implemented. The analysis included pipe whip. jet impingement, flooding, I

, and environmental effects resulting from postulated piping system failures in all l the high and moderate energy systems that cod d affect the residual heat removal j system. Based on our review, we find this analysis of the residual heat removal system acceptable. ;

Main steam and feedwater lines will exit the containment in a centralized location designated as the main steam valve house. The mair, steam and feedwater lines will j be routed straight through the main steam valve house and annulus building to the

, turbine building. The electrical tunnel, which will house one of the redundant safety related electrical cable trains, will be located directly below the floor of the main steam valve house. Since the safety-related electrical tunnel will be located adjacent to the main steam valve house, the applicant has modified the l

1 3-8 1

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l design criteria and bases in the following manner at our request. Vent areas will be included so that the structural capability of the main steam valve house will be maintained following a postulated nonmechanistic break equivalent to the full flow area of a single-ended rupture, without pipe whip or jet forces, in those " break exclusion areas" of tne main steam and feedwater system. Also, the safety-related equipment located inside the main steam valve house will be designed to withstand the environmental effects resulting from a postulated break in a feedwater line or a main steam line.

The applicant has proposed not to use tne turbine-driven auxiliary feedwater pump for normal plant startup and shutdown and will reserve its use for emergency shutdown only. On this basis, the applicant considers the piping system extending from the turbine-driven auxiliary f eedwater pump discharge to the check valves inside contain-ment as a moderate energy system. This is not acceptable. We will require that the piping system downstream of the turbine-driven auxiliary feedwater pump discharge be considered as a high energy system and that protection against postulated pipe breaks in this system be provided accordingly. Therefore, this matter remains outstanding.

Subject to the satisfactory resolution of the above matter, we conclude that the design criteria and bases for protection of essential systems and components from a postulated failure of piping both inside and outside the containment meet the requirements of the Standard Review Plan and staff technical positions and are acceptable.

3.7 5eismic Design The applicant has described the seismic design features for the proposed plant, 4 including the seismic input criteria, the system and subsystem seismic analyses to be performed, and the seismic instrumentation program to be utilized.

l j

We have not completed our review of tnese seismic design features. Therefore, this matter remains outstanding.

3.8 Design of Seismic Category I Structures The applicant has described the criteria to be used in the design of seismic Category l

I I concrete and steel structures and foundations, including the combination of I loads, materials of construction and the use of codes and standards.

We have not completed our review of the design criteria for these seismic Category I structures and foundations. Therefore, this matter remains outstanding.

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3.9 Mechanical Systems and Components 3.9.1 Design Transients and Analytical Methods The applicant has provided information concerning the design transients and the methods of analysis to be used in the design of seismic Category I components, all ASME Code Class 1, 2, 3 or core support components, including Class I component supports, reactor internals not specifically addressed by the ASME Code, and uther components not covered by the Code. The areas of our review included the following

, specific subjects:

(1) Transients which will be used in the design and fatigue analyses of $11 ASME Code Class 1 and core support components and of Class I component supports and reactor internals.

(2) Descriptions of all computer programs which will be used in analysis of ASME Code and non-Code items.

(3) Descriptions of any stress analysis programs which will be used in lieu of theoretical stress analyses.

(4) Descriptions of the analysis methods which will be used if the applicant elects to use inelastic stress analysis methods in the design of the above noted components.

Transients The applicant has provided a list of the transients to be used in the design and fatigue analysis of all Code Class 1 and core support components, as well as Class I component supports and other reactor internals not specifically addressed by the ASME Code.

All design transients have been specified and include startup and shutdown operations, power level changes, emergency and recovery conditions, switching operations (i.e.,

startup or shutdown of one or more coolant loops), control system or other system malfunctions, component malfunctions, transients resulting from single nperator errors, inservice hydrostatic tests, seismic events, and other transients that are contained in the Code required " Design Specifications" for the components of the reactor coolant pressure boundary. The transient conditions selected for equipment fatigue evalution are based upon a conservative estimate of the magnitude and frequency of the tempor6t m and pressure conditions resulting from these transients.

We have reviewed the list of tiansients and the number of events estimated for each transient presented by the applicant. Inis information conforms with the acceptance criteria outlined in Section 3.9.1 of the Standard Review Plan. The design transients have been properly categorized with respect tc the component operating conditions of design and, therefore, we find them to be acceptable.

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2 Computer Programs Used in Analysis Computer programs that will be used in the design of seismic Category I components and piping systems within the balance-of plant scope have been listed. Also included are descriptions of the programs, the extent of their application and the program verification which will demonstrate the validity and applicability of each program.

The computer programs which will be used for analyzing seismic Category I components and equipment within the scope of the nuclear steam supply system have also been listed. Their verifications and descriptions are contained in Westinghouse Topical Report WCAP-8?"2, " Documentation of Selected Westinghouse Structural Analysis l

Computer Codes." Although our review of WCAP-8252 has not been completed, we have completed our review of the analytical procedures described in that report. We find that the analyses, performed consistent with our requirements, will provide adequate assurance that the combined stresses and strains in the components of the l reactor coolant system and reactor internals will not exceed the allowable elastic design limits for the materials of construction as specified in Appendix F to Section 11I of the ASME Code, and that the resulting deflections or displacements of any structural elements of the reactor internals will not distort the reactor internals geometry to the extent that core cooling can be impaired. Therefore, we conclude that the applicant has acequately described each program with respect to the type of analysis to be performed and the specific components to which tre program will be applied and that sufficient information has been provided ja conform-ante with the criteria in Section 3.9.1 of the Standard Review Plan.

Experimental Stress A,.a ysis No experimental stress analysis is expected to be employed in lieu of analytical j methods for the design of balance-of plant equipment, components and piping systems l as well as seismic Category , systems or components within the scope of the nuclear steam supply system. However, extensive use will be made of measured results from prototype plants and variou< scale model tests for determining the dynamic response of the reactor internals. These tests are discussed in Section 3.9.2 of this report.

Inelastic Stress Analysis Only elastic analysis techniques, described in Section 3.7.3A of the Preliminary 1 Safety Analysis Report, will be ut ized in the qualification of seismic Category I ASME Code and non-Code equipment wi.nin the balance of plant scope. The design conditions and stress limits defined are applicable for an elastic system (and equipment) analysis. Inelastic analyses are not anticipated at this time. If and i when inelastic analyses are employed, the applicant has made a commitment to establish detailed design bases demonstrating maintenance of function and/or structural integrity, prior to implementation. We find this commitment to be acceptable, j 3-11

I 3.9.2 Dynamic Testing and Analysis A preoperation vibration and thermal effects test program will be conducted on ASME Code Class 1, 2 and 3 piping systems within the balance of plant scope under simulated transients for the normal and upset operating modes of the systems. During the preoperational and initial startup test program, if excessive vibration is visually observed on any ASME Code Class 1, 2 and 3 piping system, corrective support :,ystems will be designed and installed ana the effect of the modification will be incorporated in the pipe stress analysis. When instrumented testing is used, the selection of measurement stations in the test program will assure adequacy of qualifying the pipe systems. The applicant has made a commitment to provide a descripton of the tests in the Final Safety Analysis Report, including the test prerequisites, objec-tives and acceptance criteria, the design and supervision of the tests, and the corrective actions taken if excessive vibrations should occur.

A preoperational piping vibrational and dynamics effects testing program will be conducted for the reactor coolant loop / supports system during startup functional testing of the Sundesert plant. The purpose of these tests will be to confirm that the system has been adequately designed and supported for vibration as required by paragraph NS-3622.3 of Section III of the ASME Code. The tests will include reactor coolant pump starts and trips. If excessive vibrations are observed during these tests, the cause of these vibrations will be identified and necessary modifications will be made in the support system to reduce the vibrations.

The methods and procedures to be used in the design and qualification of the seismic Category I mechanical equipment are oiscussed in Sections 3.7.3, 3.9.3 and 3.10 of the Preliminary Safety Analysis Report. Loading combinations include operating as well as earthquake loading for evaluation by testing and/or analytical methods.

For those components designed in accordance with the ASME Code, the operating loads will be added to the operatinn basis eart U ke loads and evaluated against Service Limit B stress. For sv ..alc Category a components not designed in accordance with the Code, the stre level under the combined loadings will be limited to 75 percent of the minimum yield strength of the material per ASTM specification. General criteria for analysis of the safe shutdown earthquake, pipe rupture, and operating loads require that deformation of components be allowed only with no loss of safety function. Generally, stress limits will be set for the safe shutdown earthquake condition such that upper bound limit loads are not exceeded.

A dynamic response analysis will be made of reactor internals under operational flow transients and steady state conditions. Themainobjectiveofthisanalysis is to establish the characteristics of the forcing functions that determine the response of the structures. The most important forcing functions are those associated with flow turbulence and pump-relateo excitation. The effects of these forcing functions have been studied from test runs on models, prototype plants and in component tests.

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Test results from three Westinghouse plants will be used in support of the design l 1

of the reactor internals for Sundesert; the H. B. Robinson Plant, Unit No. 2 (Docket j No. 50-261), the Trojan Nuclear Plant (Docket No. 50-344) and the Sequoyah Plant, Unit Nos. 1 and 2 (Docket Nos. 50-327 and 50-328). The H. B. Robinson plant has l been established as the prototype design for three-loop plant internals. H. B. Robinson was instrumented and tested during hot functional testing. Sundesert is similiar f

to H. B. Robinson; the only significant dif ferences are the modifications resulting from the use of 17x17 fuel, replacement of the annular thermal shield with neutron shielding panels, and the thange to the inverted top hat support structure. The main structural differences in the internals between the 15x15 and the 17xl? fuel assembly are the guide tubes and control rod drive lines. The new 17x17 guide tubes are less susceptible to flow induced vibration.

The primary cause of core barrel excitation is flow turbulence, which is not affected by the upper internals. The vibration levels of the Sundesert plant internals ;

resulting from core barrel excitation are expected to be similar to those of the Trojan plant, since both plant designs include neutron shielding pads. The Trojan plant was instrumented and tested during hot functional testing, Results from the Trojan test as well as from scale mocel tests show that core barrel vibration of plants with neutron shielding pads is less than that of plants with thermal shields.

The Trojan results verify the adecuacy of the neutron pad core barrel and 17x17 guide tube designs and provide plant data applicable to Sundesert. The Sequoyah plant upper internals and the Sundestrt plant inverted top hat upper support structure are similar in design and their vibrational behreior is expected to bFW%.

The Sequoyah plant will be instrumented and tested outing hot functional testing and will provide vibrational data applicable to the Sundesert plant inverted top hat upper support structure.

The recommendations of Regulatory Guide 1.20, " Comprehensive Vibration Assessment Program for Reactor Internals During Preoperational and Initial Startup Testing,"

will be met by conducting the confirmatory preoperational examination for integrity.

This examination will include 35 locations with emphasis on such areas as (1) the major load bearing elements of the reactor internals relied upon to retain the core structure in place, (2) the lateral, vertical and torsional restraints to be pro-vided within the vessel, and (3) the locking and bolting devices whose failure could adversely affect the structural integrity of the internals. l 0> ring the hot functional test, tne internals will be subjected to greater than normal full flow conditions for at least 240 hours0.00278 days 0.0667 hours 3.968254e-4 weeks 9.132e-5 months . This will provide a cyclic 7

loading of approximately 10 cycles on the main structural elements of the internals.

A dynamic system analysis will be performed for the components of the reactor coolant system and reactor internals under postulated accident conditions. An acceptable analysis provides adequate assurance that the combined stresses i

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] and strains in the components of the reactor coolant system and reactor internals

will not exceed the allowable design stress and strain limits for the materials of l

construction, and that the resulting deflections or displacements at any structural i elements of the reactor internals will not distort the reactor internals geometry

to the extent that core cooling may be impaired. !

d t i i Dynamic system analysis of the reactor internals for blovdown loads resulting from a postulated loss-of-coolant accident will be based on tie time history response of the internals to simultaneously applied blowdown forcing functions. The blowdown

{ code evaluates the pressure and velocity transients for a maximum of 2400 locations 4

throughout the system. The pressure waves generated within the reactor would be

highly dependent on the location and nature of the postulated pipe failure. In the '

l case of the hot leg break, the vertical hydraulic forces would produce an initial upward lift of the ccre. A rarefaction wave would propagate through the reactor '

i t hot leg nozzle into the interior of the upper core barrel, and dynamic instability j or large deflections of the upper core barrel is a possible response of the barrel i i

during a hot leg break. This would result in a transverse loading on the upper l

core components as the fluid exits the hot leg nozzle. In the case of the cold leg break, the hydraulic forces would tend to cause the reactor core and lower support structure to move initially downward. The stresses due to the safe shutdown earth-

} quake will be combined with the blowdown stresses in order to obtain the largest

! principal stress and deflection.

l:

l The dynamic system analysis to be performed will take into account all related load j effects on the reactor internals and unbroken reactor coolant piping loops due to a i postulated loss-of-coolant accident and a safe shutdown earthquake. As discussed f in Section 3.9.3 of this report, these load effects will be combined by the l square-root-sum-of-the-squares method which we find acceptable for these Class 1 1 components.

i j The methods to be used for component analysis have been found to be compatible with

} those used for systems analysis and, therefore, are acceptable. The assurance of structural integrity of the reactor internals under loss-of coolant accident conditions, for the most adverse postulated loading event, provides added con-fidence that the design will withstand a spectrum of lesser pipe breaks and seismic j loading events. Accomplishment of the dynamic system analysis constitutes an j acceptable basis for satisfying the applicable requirements of Criteria 2 and 4 of

the General Design Criteria.

2 3.9.3 Loading Combinations, Design Transients and Stress Limits i

Components and systems, which are seismic Category I and which are ASME Code

[ Class 1, have been categorized with respect to " normal," " upset," " emergency" or

" faulted" plant conditions. The allowable stress limits have been defined and are applicable to stress results obtained by elastic analysis techniques. As stated in 3-14

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Section 3.9.1 of this report, design transients to be included in the above category have also been defined and found acceptable. Loading conditions and stress limits for ASME Code Class 2 and 3 components of seismic Category I fluid systems, which ,

are to be constructed in accordance with ASME Code Section 111, Subsections NC and ND, have been reviewed. These strest limits and design conditions are intended to assure that no gross deformation of the component occurs. The stress limits a*e applicable for an elastic system analysis. The applicant does not anticipate that inelastic deformation will be allowed on any ASME Code Class 2 and 3 components.

In case inelastic analysis is used, cetailed design bases, demonstrating maintenance of sither function and/or structural integrity, will be proposed prior to implementa-tion. The design criteria, design loading combinations (except for the methodology as discussed below), and stress limits for ASML Code Class 2 and 3 piping of seismic Category I fluid systems have been reviewed and found acceptable.

For Class I components and supports, the applicant proposes to combine the effects of the safe shutdown earthquake and a pestulated loss-of-coolant accident by the square-root-sum-of-the-squares methoo. We find this method to be acceptable for this load combination for Class 1 equipment. For combinations of effects from other applicable dynamic loads for Class 1 equipment, the applicant also proposes to combine them by the square root-sum-of-the-squares method for the balance-of-plant scope and does not address this matter for equipment within the scope of the nuclear steam supply system. We require that these additional load effects be combined by the absolute sum method. Therefore, this matter remains outstanding.

]

For Class 2 and 3 components and supports within the scope of the nuclear steam supply system, the applicant proposes to combine all effects from applicable dynamic loads, including the effects of the safe shutdown earthquake and a postulated loss-of-coolant accident, by the absolute sum method. We find this to be accept-able. Within the balance-of plant scope, the applicant proposes to combine all ef fects f rom applicable dynamic loads by the square-root-sum-of-the-squares methoc.

We require that all applicable load effects for Class 2 and 3 equipment be combined by the absolute sum method. Therefore, this matter remains outstanding.

However, we are currently reviewing, on a generic basis, the methodology for combining dynamic responses due to certain postulated loads. The recommendations resulting from our review of this matter will be applied to all plants, including Sundesert.

With respect to the postulated loss-of-coolant accident event, all related load effects, including any asymmetric pressure effects on the reactor vessel, its internals, its supports, unbroken piping, and on other primary loop component supports, must be evaluated. The applicant has made a commitment to provide, in the Final Safety Analysis Report, an analysis which will include the above mentioned loads. We find this commitment to be acceptable.

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All Class I components and supports will be designed and analyzed for the responses due to the design and service loads which result from plant and system design, normal, upset, emergency, and faulted conditions. The design and service loading combinations provided by the applicant have been reviewed and found acceptable.

1 The design and service limits for the loads specified by the applicant have also been reviewed and found acceptable. The applicant has committed not to use plastic component analysis in conjunction with elastic system analysis or with plastic system analysis, unless the deformations and displacements of the individual system members can be shown to be no larger than those which can be properly calculated by the analytical methods used for the system analysis. We find this commitment to be acceptable.

1 The linear component support buckling allowable load will not exceed 0.90 times the critical buckling strength of the support at a temperature based on ASME Section 13I, Appendix XVII. T+ the design of component supports results in loads in excess of 0.67 times the critkal buckling strength for the faulted condition, verification of the support functional adequacy will be established by analysis and/or testing l and will be documented in the Final Safety Analysis Report.

1 The applicant has made a commitment that the increased design stress limit identi- '

g fied in NF-3231.l(a) shall be limited to the smaller of the ultimate strength or twice the yield strength, unless otherwise justified by shakedown analysis. The applicant has taken exceptions to Regulatory Guide 1.124, " Design Limits and Loading Combinations for Class 1 Linear-Type Component Supports," regarding the critical, buckling strength for the faulted condition relative to the plate type supports. We require additional information to justify the exceptions taken.

Therefore, this mattter remains outstanding.

The reactor vessel supports will be analyzed for the ef fects of loss-of-coolant I accidents resulting from postulated breaks in the reactor coolant system piping, including breaks in the vessel cavity which would result in nonaxisymmetric pressure distributions on the internals and on the vessel exterior walls.

The applicant has made e commitment to prepare a detailed dynamic model, specifi-cally for the Sundesert plant, which will include the stiffnesses of the reactor vessel supports and the attached piping. Hydraulic forces which would be developed !

in the internals due to a break at the reactor vessel nozzle will be included in the model. These forces are characterized by time-dependent forcing functions on the vessel and core barrel. In the derivation of these forcing functions, the fluid-structure (or hydro elastic) interaction in the downtomer region between the core barrel and the vessel will be taken into account.

The analysis will include loads due to internal reaction, reactor cavity pressure, and the loop mechanical forces. These loads will be applied simultaneously, in a nonlinear elastic dynamic time history analysis, in the model of the vessel, its 3-16

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i supports and internals. The results of this analys.is will provide the dynamic i l

loads for the reactor vessel supports and vessel time history displacements. The maximum stresses in the supports will be calculated and evaluated against the stress criteria of Appendix F to Section III of the ASME Code.

The criteria to be used in the metnous of analysis which 'the applicant will employ in the design of all seismic Category 1 ASME Code Class 1, 2, 3 and core support ,

components, component supports, and non-Code items are in conformance with I established technical positions and criteria which are acceptable to the staff.

The use of thete criteria in defining the applicable design transients, the computer codes used in analysis, and the analytical methods and experimental stress analysis methods to be used provide assurance that the stresses, strains and displacements calculated for the above n.oted items will be as accurate as the current state-of-the-art permits and are adequate for the design of these items.

During our review, we requested that the applicant provide a commitment that where ASME Code Service Limits C and D are used in the design of piping essential to safe shutdown of the plant, piping deformations would be evaluated to assure that sufficient dimensional stability is maintained to enable the system to deliver its rated flow. Although the use of Service Limits C and D provides adequate assurance that structural integrity will be maintained, the ASME Code does caution that large deformations could occur at areas of structural discontinuity when stresses are at the levels permitted by these two limits. The applicant's response to this concern was not totally acceptable. We require that the applicant describe how f unctional capability of all piping essential to the safe shutdown of the Sundesert plant will i be demonstrated. Appendix D to this report provides guidance on an acceptable way of demonstrating this functional capability. Therefore, this matter remains outstanding.

Subject to the satisfactory resolution of the above matters relating to the (1) methodology for combining load effects and (2) demonstration of funct'ional cap-ability of all piping essential to the safe shutdown of the plant, we conclude that the design load combinations and associated stress and deformation limits specified l for ASME Code Class 1, 2, 3 and core support components, component supports and l

non-Code items are in conformance with the acceptance criteria of Section 3.9.3 of the Standard Review Plan and constitute an acceptable basis for design in satisfying the applicable requirements of Criteria 1, 2 and 4 of the General Design Criteria.

3.9.4 Control Rod Drive systems i

The control rod drive mechanism pressure housings will be Class 1 components designed to meet the stress requirements for normal operating conditions defined in Section 111 of the ASME Code. The information provided by the applicant includes design criteria, testing programs and a summary of the method of operation of the control rod drives.

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i j in the dynamic analysis, the ( 'clic stresses due to dynamic loads and deflections

, will be coeined with the stresse; inposed by the loads from component weights,

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hydraulic forces and thermal gradients for the determination of the total stresses of the control rod drive system. for normal operating conditions, Section 111 of

{

i the ASME Code will be used. All pressure boundary components will be analyzed as l Class I components under Article NB-3000. The use of these criteria provides

reasonable assurance that the system will function reliably when required, and

{ forms'an acceptable basis for satisfying the mechanical reliability provisions of l Criterion 27 of the General Design Criteria.

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! 3.9.5 Pump and Valve Operability Assurance Program The component operability assura w program is intended to assure the operability j j of ASME Code Class 1, 2 and 3 active valves and the ability of active pumps to '

function under plant conditions where their operation is relied upon for plant l

j shutdown or for mitigating the consequences of an accident. The program has been j evaluated with respect to test and analytical methods and combinations thereof.

l The test program will include prototype testing, both under simulated test

! conditions in the shop and in-situ conditions after installation. j l

, A listing has been provided in the Preliminary Safety Analysis Report of active j Class 1, 2 and 3 valves and pumps identified by system and active function. Class 1 4 2 and 3 pumps will be designed and analyzed according to the rules of ASME Code section III, Subsections NC and NO. Performance of these analyses with the code 1 allowable stress limits assures that critir.al parts of these pumps are not damaged ;

{ during the short duration of the faulted condition and that reliability of the j j pumps for post-faulted condition operation is not impaired by the seismic event. !

in addition to the post-faulted condition operation, it is necessary to assure that

] the pumps function throughout the safe shutdown earthquake. The seismic loading j would cause only a slight increase in torque necessary to drive the pumps at the

constant design speed. Therefore, the pumps would not shut down during a safe j shutdown earthquake, and would operate at the design speed despite the safe j shutdown earthquake loads. The pump motors will be independently qualified for

. operation for the maximum seismic event. All vital auxiliary equipment will be j qualified by meeting the requirements of IEEE Standard 344-1975, " Guide for Seismic j Qualification of Class 1 Electrical Equipment for Nuclear Power Generating ;

j Stations." If the testing option is chosen, sinusoidal or sine beat testing will l be justified by satisfying one or more of the following requirements to demonstrate i that the multifrequency response is negligible or that the input is of sufficient

magnitude to conservatively account for this ef fect. !

(1) fhe equipment response is basicalty due to one mode, i

(2) The sinusoidal or sine beat response spectra envelop the floor response spectra in the region of significant response.

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(3) The floor res. m se spectra consist of one dominant mode and have a peak at this frequency.

The functional ability of active pumps after a faulted condition will be assured since only normal operating loads and steady-state n0ZZle loads would exist. The post-faulted condition ability of the pumps to function under these applied loads 1 will be proven during the normal operating plant conditions for active pumps.

Safety-related active valves must perform their mechanical motion during and/or following an accident. The safety-related valves will be subjected to a series of tests prior to service and during the plant life. Prior to installation, several tests will be performed. These will include shell hydrostatic tests in conformance l l '

with ASME Code Section 111 requirements, backseat and main seat leakage tests, disc hydrostatic tests and operational tests to verify that the valve will open and close. Cold hydrostatic tests, hot functional tests, periodic inservice inspections and periodic inservice operations will be performed in-situ to verify and assure the functional ability of the valve. These tests guarantee reliability of the valve for the design life of the plant, in addition to these tests and analyses, representative valves of each design type will be tested for verification of opera-bility during a simulated plant faulted condition event by demonstrating operational capabilities within the specified limits.

The operability of the valves during a faulted condition will be demonstrated by satisfying several criteria which include:

(1) All active valves will be designed to have a first natural frequency which is greater than 33 Hertz.

(2) The actuator and yoke of the valve system will be statically deflected and cycled under faulted condition loads.

(3) Electrical motor operators, limit switches and pilot solenoid valves necessary for operation will be qualified in accordance with IEEE Standard 344-1975,

" Guide for Seismic Qualification of Class 1 Electric Equipment for Nuclear Power Generating Sations," and Regulatory Guide 1.100, " Seismic Qualification of Electric Equipment for Nuclear Power Plants."

The component operability assurance program for ASME Code Class 1, 2 and 3 active valves and pumps provides adequate assurance of the capability of such active components to (1) withstand the imposed loads associated with Service limits A, B, C and 0 without loss of structural integrity, and (2) perform necessary " active" functions (e.g. , valve closure or opening, pump operation) under accident con ti-tions and conditions expected when plant shutdown is required. The specified component operability assurance test program constitutes an acceptable basis for satisfying applicable requirements of Criteria 1, 2 and 4 of the General Design Criteria and, therefore, is acceptable.

3-19

3.9.6 Inservice Testing of Pumps and Valves The applicant has proposed an inservice testing program that will include baseline preservice testing and periodic testing to assure that all ASME Section 111 Code Class 1, 2 and 3 pumps and valves will be in a <tata ,f operational readiness to perform their safety function throughout the life of the plans. The proposed test program will be based on the ASME Cooe,Section XI, 1974 Edition, through the Winter 1975 Addenda, dated December 31, 1975. To date, the Commission has not accepted the Winter 1975 Addenda of Section XI of the Code. We require that the Sundesert inservice testing program for pumps and valves comply with 10 CFR 50.55a(g)(3) with respect to the selection of the earliest applicable edition and addenda. Therefore, this matter remains outstanding.

3.10 Seismic Qualification of Seismic Category 1 Instrumentation and Electrical \

Equipment !

Instrumentation and electrical components required to perform a safety function will be designed as seismic Category 1, Seismic requirements established by the seismic system analysis will be incorporated into equipment specifications to assure that the equipment purchased or designed will meet seismic requirements l equal to or in excess of the requirements for seismic Category I components, either by appropriate analysis or by qualification testing.

l The applicant has proposed a seismic qualification program that will be implemented i for seismic Category I instrumentation and electrical equipment, and associated supports for this equipment, to provide assurance that such equipment can be expected to function properly and that structural integrity of the supports will not be impaired during the excitation and vibratory forces imposed by a safe shutdown earthquake and the conditions of post accident operation. The seismic qualification program described by the applicant is consistent with IEEE Standard 344-1975, " Guide for Seismic Qualification of Class 1 Electric Equipment for Nuclear Power Generating Stations," which we find to be acceptable.

The applicant has taken exception to certain recommendations in Regulatory Guide 1.100, " Seismic Qualification of Electric Equipment for Nuclear Power Plants," which are presently under discussion between the Commission's staff and the Nuclear Power Engineering Committee of the Institute of Electrical and Electronic Engineers. The applicant has, however, committed to implement the agreed upon resolutions of these exceptions when this issue is finally resolved.

We find this commitment to be acceptaole.

3.11 Environmental Design and Qualification of Mechanical and Electrical Equipment The design bases for engineered safety features equipment will include the expected environmental conditions resulting from postulated accidents to assure that this equipment can perform its intended safety function when required to do so.

3-20

._. , __ ~

I We reviewed the applicant's estimated chemical and radiation environment to which engineered safety features equipment will be exposed during a postulated design basis accident. Following a postulated loss-of-coolant accident, the chemical ,

environment inside the containment structure will consist of a spray solution of l.5 percent boric acid, with sodium hydroxide added to result in a pH of 10.5 for the first two hours and a pH of 8.5 after two hours. This chemical environment reflects the chemical composition of all fluids and additivies that will be added to the containment environment in the course of the accident.

The applicant has stated that the engineered safety features equipment to be located within the containment and exposed to the containment atmosphere will be qualified for the radiation environment as recommended in Regulatory Guide 1.4, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-of-Coolant Accident for pressurized Water Reactors." The engineered safety features equipment ;

to be located within the containment or within other areas of the facility and exposed to recirculation water will be qualified for the radiation environment as recommended in Regulatory Guide 1.7. " Control of Combustible Gas Concentrations in Containment following a loss of-Coolant Accident." The engineered safety f eatures equipment to be located uutside the containment and not in contact with recircu-lation water will be qualified for a radiation environment resulting from direct radiation f rom the containment or, where applicable, direct radiation f rom the ,

systems containing recirculation water.

The applicant indicates that the radiation level inside the containment under design basis loss-of-coolant accident conditions plus the integrated exposure over 7

~

40 years of normal operation will be 4 x 10 rads. This radiation level is I

consistent with the source terms of Regulatory Guides 1.4 and 1.7.

i t

We have determined that the proposed chemical and radiation environments are comparable with those of similar plants recently reviewed and approved. Accordingly, we conclude that the applicant's r*.emical and radiation source terms that define ,

the envit a ental conditions to be used in the design of the engineered safety features mechanical and eiectrical equipment are appropriate for the postulated design basis loss-of-coolant accident.

1 The applicant has also discussed its proposed environmental qualification program for Class lE equipment, We have not completed our review of the information pro-vided by the applicant. Therefore, this matter remains outstanding.

1 3-21

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4.0 REACTOR 4.1 Summary Description Criterion 10 of the General Design Criteria requires that the reactor core and associated systems be designed to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences. We have reviewed the information provided in the Preliminary Safety Analysis Report in support of the proposed reactor design. Our evaluation is contained in the following sections.

The reactor design for each unit of Sundesert will be essentially the same as the design reviewed and found acceptable for Koshkonong, Unit Nos. I and 2 (Docket Hos.

STN 50-502 and STN 50-503). Like Koshkonong, the Sundesert plant will use fuel assemblies with a 17x17 fuel rod array.

The nuclear steam supply system for each unit will be designed to operate at a maximum thermal output of 2785 megawatts with sufficient margin to allow for transient operation and instrument error without causing damage to the core and without exceeding the pressure settings of the safety valves in the coolant system. The core thermal power level will be 2775 megawatts.

The core will be cooled and moderated by light water, at a pressure of 2250 pounds per square inch absolute, in the reactor coolant system. The reactor coolant will contain soluble boron for neutron absorption. The concentration of the boron will be varied as required to control relatively slow reactivity changes including the effects of fuel burnup, Additional boron, in the form of burnable poison rods, will be employed to establish the desired initial reactivity. ,

4.7 Mechanical Design 4.2.1 Fuel 4

Description The fuel assemblies proposed for Sundesert will consist of 264 fuel rods, 24 guide thimbles, and one instrumentation thimble arranged in a 17x17 array. The instru-mentation thimble will be located at the center of the assemblies and will facil-itate the insertion of neutron detectors. The guide thimbles will provide channels for inserting various reactivity controls. The fuel rods will contain uranium i

dioxide ceramic pellets hermetically clad in Zircaloy-4 tubes. The fuel assembly structure will be held together by Zircaloy thimble tubes and the stainless steel l

l 4-1

fuel assembly nozzles at the top and bottom. Alignment and transverse spacings will be maintained by eight spacer grids separated uniformly along the axis of the assembly.

All fuel rods will be internally prepressurized with helium during final welding to reduce cladding compressive stresses during service. The level of prepressurization is designed to preclude cladding flattening. The specific level of prepressurization will be dependent upon the planned fuel burnup and will be determined prior to establishing Technical Specifications.

The 17xi7 array fuel assembly design proposed for Sundesert is identical to the assemblies used or to be used in the Trojan Nuclear Plant (Docket No. 50-344), the J Farley Plant, Unit Nos. I and 2 (Docket Nos. 50-237 and 50-249), the Beaver Valley Power Station, Unit No. 1 (Docket No. 50-334), the Salem Nuclear Generating Station, Unit No.1 (Docket No. 50-272) and the Surry Power Station, Unit Nos. I and 2 (Docket Nos. 50-280 and 50-281). The 17x17 fuel assembly design is only a slight modification of the previously used Westinghouse 15x15 fuel assembly design. Those mechanical aspects of the Sundesert and Trojan 17x17 fuel assembly design which differ from the previously used 15x15 design are indicated in Table 4.1. The )

differences are essentially geometric and will result in a lower linear power density and other increased safety margins for the 17xl7 type fuel assembly, l

TABLE 4.1 FUEL MECHANICAL DESIGN COMPARISON j I

WESTINGHOUSE ;

DESIGN PARAMETER SUNDE5ERT TROJAN TYPICAL OPERATING FUEL l FUEL ASSEMBLY:

Rod Array 17x17 17xl7 15x15 Number of Fuel Rods 264 264 204 Fuel Column Length, inches 144 144 144 Number of Spacer Grids 8 8 7 l Number of Guide Thimbles 24 24 20 inter-rod Pitch, inches 0.496 0.496 0.563 Average Thermal Output (4 loop), kilowatts per foot 5.44 5.44 7.0 FUEL PELLETS: I Density (theoretical),

percent 95 95 94 Fuel Weight / Unit length (per rod, not assembly),

pounds per foot 0.364 0.364 0.462 FUEL CLADDING:

Ootside Radius, inches 0.187 0.187 0.211 Thickness, inches 0.0225 0.0225 0.0243 Radius / Thickness ratio 8.31 8.31 8.68 4-2

The evaluation of the Westinghouse fuel mechanical design is based upon mechanical tests, in-reactor operating experience, and engineering analyses. Additionally, the in-reactor performance of the fuel design will be subjected t,o the continuing surveillance programs of Westinghouse and individual utilities as discussed below under Fuel Surveillance. These programs provide confirmatory and current design performance information.

Thermal Performance in our evaluation of the thermal performance of the reactor fuel, we assume that l

densification of the uranium oxide fuel pellets may occur during irradiation. The initial density of the fuel pellets and the size, shape, and distribution of pores within the fuel pellets influence the densification phenomenon. Briefly stated, in-reactor densification (shrinkage) of oxide fuel pellets: (1) may reduce gap conductance, and hence increase fuel temperatures, because of a decrease in pellet diameter; (2) increases the linear heat generation rate because of the decrease in pellet length; and (3) may result in gaps in the fuel column, as a result of pellet length decreases, which produce local power spikes and the potential for cladding creep collapse. l The engineering methods to be used to analyze the densification effects on fuel thermal performance have been previously submitted in Westinghouse Topical Report WCAP-8218, " Fuel Densification Experimental Results and Model for Reactor Applica- i 1

tion." The methods include testing results, mechanical analyses, thermal and j hydraulic analyses, and accident analyses. The results of our evaluation and .

approval of the methods for use in licensing are given in a staf f report, " Technical Report on Densification of Westinghouse PWR Fuel," dated May 14, 1974; additional information on densification methods can be found in NUREG-0085, "The Analysis of t Fuel Densification," dated July 1976.

Recently, Westinghouse has changed its fuel design criterion for fuel rod internal pressure, as described in Westinghouse Topical Report WCAP-8963, " Safety Analysis for the Revised fuel Rod Internal Pressure Design Basis." This new criterion would allow the internal pressure to exceed external system pressure provided that the pressure of the highest burnup rod will be limited to a value below that which could cause the diametral gap to increase due to outward clad creep during steady state operation. The applicant also proposes to use this new criterion. Based on our review, we find the criterion acceptable, as documented by letter to Westinghouse, dated May 19, 1978, provided the pressure is also limited so that extensive departure from nucleate boiling propagation cannot occur. At the operating license stage of review, we will assure that this aspect of the criterion is also met.

The applicant will also utilize a revised version of the Westinghouse thermal performance code, as described in WCAP-8720, "Impeoved Analytical Models Used in Westinghouse Fuel Rod Computations," for its safety analyses. This code contains a revision of the earlier Westinghouse model for fission gas release to account for 4-3

increased gas release at high burnup, and revised models for helium solubility, fuct swelling, and densification.

The revised Westinghouse code is under review. Although our review is not complete, we find that the revised model is not significantly different than the previous model which we found acceptable with some modification to account for increased gas release at high burnup. Since we project that our review of the revised model will be completed prior to the operating license review phase for Sundesert, any modifi-cations to the model resulting from our review of WCAP-8720 can be taken into account for Sundesert at that time. On this basis, we find the use of the revised thermal performance code acceptable at the construction permit stage of review. At the operating license stage of review, we will assure that the Sundesert safety analyses are performed with an acceptable version of the thermal performance code.

The applicant will also use a revised Westinghouse cladding flattening model described in Westinghouse Topical Report WCAP-8377, " Revised Clad Flattening Model," which, for a given fuel region, predicts initial flattening time and the flattened rod frequency for pressurized rods containing relatively stable fuel. This revised model was based on an analysis of the results of television examinations of irradi-ated fuel rods which indicated that the original flattening model, described in Westinghouse Topical Report WCAP-7982, " Fuel Densification Penalty," significantly underpredicted the time and frequency of collapse. The "COLLAP" computer code is used to perform the analysis. We have reviewed Topical Report WCAP-8377 and found the revised model, as well as the input data used i, the model, to be acceptable for use in safety analysis, as documented in a letter to Westinghouse, dated February 14, 1975.

Mechanical Performance Although limited operating experience exists on 17x17 fuel assemblies, substan-tially all of the in reactor operating experience with Westinghouse fuel rods and assemblies is applicable to the Sunde>ert fuel design since the 17x17 fuel assembly is only a slight mechanical extrapolation of the 15x15 fuel assembly. The current use of the similar fuel rods and 15x15 assemblies has yielded operating experience that provides confidence in the acceptable performance of the fuel assembly design.

The range in design parameters for which in-reactor experience is specifically applicable has been tabulated in Table 4.2.

The assemblies referred to in Table 4.2 have been irradiated for up to six years and have peak exposures of 30 gigawatt days per metric tonne, totaling more than 70 million megawatt hours of power generation.

In addition to the in-reactor experience with the 15x15 fuel assemblies, verifica-tion tests have also been performed on 17x17 fuel assemblies with the 12-foot core length proposed for Sundesert. The results of the completed tests have been 4-4

TABLE 4.2 RANGE OF DESIGN PARAMETER EXPERIENCE PARAMETER RANGE OF POWER REACTOR EXPERIENCE Fuel Rod Array 14x14, 15x15, and 17x17 Rods per Assembly 179 to 264 Guide Thimbles per Assembly 16 to 24 Assembly Envelope, inches 7.76 to 8.43 Inter-rod Pitch, inches 0.563 to 0.463 Plenum Length, inches 3.27 to 6.69 Prepressurization, pounds per square inch, absolute 14.7 to over 400 Diametral Cap, inches 0.0065 to 0.0075 Spacer Grids / Assembly 7 to 9 fuel Column Height, inches 120 to 144 reported in Westinghouse Topical Reports WCAP-8278, " Hydraulic Flow Test of the 17x17 Fuel Assembly," and WCAP-8236, " Safety Analyses of the 17x17 Fuel Assembly for Combined Seismic and Loss-of-Coolant Accident." In addition, seismic effects i

md vertical loads resulting from double-ended hot and cold leg pr mary coo lant pipe breaks on the 17x17 fuel assembly design during a postulated loss-of-coolant accident where analyzed in Westinghouse Topical Report WCAP-8236. As a result of our review of these two reports, we had concluded that the reported results provide an acceptable basis for demonstrating the functional reliability of the 17x17 fuel assecblies, as documented in a letter to Westinghouse, dated June 12, 1974.

However, Westinghouse subsequently postulated a new asymmetric horizontal hydraulic load caused by a postulated pipe break within the biological shield. Westinghouse informed us in a letter, dated March 1,1976, that, although the experiments and calculational techniques supplied in WCAP-8236 may be applicable in assessing the adequacy of the fuel assembly to withstand this additional load, the adequacy should f be reviewed on a plant-by plant basis.

The applicant has stated in the Preliminary Safety Analysis Report that the above asymmetric loads will be included with the other loads cited above in assessing the design adequacy of the Sundesert fuel assemblies. Westinghouse has performed

$1milar analyses in support of several applications for operating licenses which utilize Westinghouse eight grid 17x17 fuel and has shown that these fuel assemblies will be able to withstand this additional hydre.ulic load. At the operating license stage of review, we will assure that this additional load is adequately considered in the Sundesert fuel assembly analysis. The effects of this additional load can be analyzed for Sundesert either on an individual plant basis or within a bounding calculation for a group of plants.

t The consideration of fuel rod bowing in the 17x17 design was previously analyzed by Westinghouse and documented in Topical Report WCAP-8346, "An Evaluation of Fuel Rod Bowing." The topical report described an analysis of rod bowing based upon delibera-l l

tion of the potential mechanisms causing fuel rod bowing. The analysis appeared 4-5

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

i i

)

4 j rigorous and compatible with the available oa w.

t The methodology of the topica' 3

report was approved with the requirement that observations of fuel rod bowing in

] modified fuel assemblies (rod off-bottom) substantiate this methodology. Subsequent observations, however, indicated that tne magnitude of rod bow was underpredicted.

j Consequently Westinghouse has reassessed its analysis in light of this new infor-

} mation and has documented its findings in Topical Report WCAP-8691, " Fuel Rod l: Bowing." In this report, Westinghouse has documented its rod oowing experience I

i wh uh, to date, is basea upon the inspection of 26 different regions of fuel (about 25,000 fuel rods) including more than seventy 15x15 assemblies at burnups beyond j 27,000 megawatt days per ton of uranium. This experience has demonstrated the f exposure (burnup) dependence of rod bowing.

i d

j We issued an interim safety evaluation on WCAP-8691 in April 1976. In this evalu-j ation, we accepted the burnup-dependent approach to rod bow used by Westinghouse I

( with modifications to account for extensions to the 17x17 design and for an increase '

in rod bow from as measured values (cold dimensions) to those in the reactor (hot t dimensions). Additional discussion on rod bowing is presented in Section 4.4.1 of ,

1 this report. !

i

!~

Limitations on power rate changes will af fect pellet / cladding interaction, which is j

being reviewed as a generic item. The Westinghouse 17x17 fuel rod design proposed '

for Sundesert incorporates features that will reduce cladding strain due to pellet /

f cladding interaction, when compared with the Ibx15 design. These features include (1) pellet chamfering, (2) rod prepressurization, (3) lower linear heat rating, and (4) smslier cladding diameter-to-thickness ratio. Based on the available experi-

) mental and commercial reactor data, thede design features should result in a reduction j of pellet / cladding interaction failures or delay of the failures to later in the j

fuel design life. While the failure thresholds are probably lower at high burnup than at low burnup, the fuel duty is also less severe. Our review of the consequences of pellet / cladding interaction failures nas so far not resulted in the identifi-4 I cation of any safety problems. Therefore, no restrictions are Currently warranted.

If any safety issues are identified on this matter in the future, appropriate restrictions on power rate changes will be implementtd.

4 t

We have reviewed the safety aspects of fuel rod falltres due to waterlogging in a l recent survey published in NUREG-0303, " Evaluation of the Behavior of Waterlogged

}

F uels in LWR's. " The evaluation was based on available information and included

) (1) results of tests in the capsule driver core at SPERT and the Japanese test i

g reactcr NSRR, and (2) observations of waterlogging failures in test and commercial reactors.

It was concluded that (1) operating restrictions to reduce pellet /

j cladding interactions also reduce the potential for waterlogging failures during i

I transients, (2) tests to simulate accident conditions produced the worst water-logging failures, and (3) there is no apparent threat from waterlogging failures to j

the overall coolability of the core or to safe reactor shutdown. We will continue i

) to monitor the waterlogging test programs and study this phenomenon generically.

h l

46

The results of f uel assembly f retting and wear tests for 17xl? f uel assemblies are reported in Westinghouse Topical Report WCAP-8278, f or a 7 grid assembly, and in a letter, dated May 15, 1975 from Westinghouse, to V. Stello, Nuclear Regulatory Commission, for an 8 grid assembly. Tnese tests indicated that f uel rod wear under both normal and transient operating conditions was within the Westinghouse predicted values and that, even for fuel rods with deliberately damaged grid cells, the wear was within acceptable limits. These tests, which simulated actual in-reactor conditions, also showed that no anomalous vibrations were observed or could be induced. Therefore, no modification to the 17xl7 fuel assembly design due to wear considerations is required.

Topical Report WCAP-8278 also presented results for fretting wear at contact points between the control rods and thimble tuDes. Contact is usually observed in two locations; (1) at the top nozzle for fully withdrawn control rods, and (2) in the dashpot transition section for inserted rods. In both regions, the observed wear was significant but was stated to be within the design limits. Because of exces-sive guide tube wear experienced in ani.ther pressurized water reactor fuel design, this area is currently under review with all pressurized water reactor vendors.

Fuel Surveillance Performance of the fuel during operation will be indirectly monitored by measure-ment of the activity of the primary coolant for compliance with Technical Specifi-cation limits, Westinghouse has proposed a fuel surveillance program for several plants that will use the 17x17 fuel assemblies. This program includes lead assem-blies in the second fuel cycles for Surry Unit Nos. I and 2 and the initial core i i loadings f or Trojan Unit No. 1, Beaver Valley Unit No. 1, Farley Unit No. 1, and Salem Unit No. 1, A summary of this program is given in Topical Report WCAP-8691 on fuel rod bowing.

The Surry units each have two lead burnup 17xl? fuel assemblies. One of the lead assemblies in each unit has removable rods. These assemblies were carefu11y measured prior to insertion and will be examined between cycles for dimensional changes, fretting corrosion near the spacer grids, tuel rod bowing, axial gamma distribution, cladding defects, and surface deposits, inspections after two cycles in Unit No. I and after the first cycle in Unit No. 2 have revealed no anomalies.

The other four reactors included in tne surveillance program (Trojan, Beaver Valley, farley, and Salem) will each have an initial core loading of 17x17 fuel assemblies.

Except for Beaver _ Valley, each core will include one removable fuel rod assembly.

Two of the f our will be examined as part of the 17x17 f uel assembly surveillance program selected on the basis of the first two reactors to actually reload fuel.

The surveillance program includes visual examination (100 percent television scanning) of the initially loaded (first core) fuel assemblies to be removed during the first three refueling outages. If any anomalies are detected, further examination will be performed using the removable fuel rod assemblies.

4-7

i f

f 5

{ Fuel Design Conclusion 1

j On the basis of our review of (1) the 17x17 fuel design analysis, including the

) analytical technl ques, (2) the Technical Specifications that will limit off gas and

{ effluent activity iMm operating clants, and (3) the confirmatory results from

{ out-of pile tests and irro,Mi;tal assemblies, we conclude that there is reasonable l assurance that the cladding 'stogrity of the Sundesert 17xl? fuel will be maintained.

4 4

j 4.2.2 Reactor Vessel Internals j The components of the reactor internals are divided into three parts: (1) the l lower core support structure (including the entire core barrel and neutron shield l pad assembly), (2) the upper core support structure, and (3) the incore instrumenta-i tion support structure. The reactor internals support the core, maintain fuel alignment, limit fuel assembly movement, maintain alignment between fuel assemblies

{-

and control rod drive mechanisms, direct coolant flow past the fuel elements, direct coolant flow to the pressure vessel head, and provide gamma and neutron

shielding.

1 l The lower core structure will be supported at its upper flange by a ledge in the '

reactor vessel head flange. Its lower end will be restrained in its transverse movement by a radial support system to be attached to the vessel wall. The neutron j shield pad assembly will consist of four pads that will be bolted and pinned to the f outside of the core barrel. These pads will be constructed of Type 304 stainless

) steel and will be located azimuthally to provide the required degree of vessel protection. Specimen guides, in which material sample specimens can be inserted l and irradiated during reactor operation, will be attached to the pads. Radial and i axial expansions of the core barrel will be accommodated, but transverse movement 4

of the core barrel will be restricted by constraints in the design. In the event j of an abnormal downward vertical displacement of the internals following a hypo-l thetical failure, energy-absorbing devices will limit the displacement after they j contact the vessel bottom head.

) Vertical loads from weight, earthquake acceleration, hydraulic loads, and' fuel l assembly preload will be transmitted through the upper core plate via the support

! columns to the top support plate assembly and then to the reactor vessel head.

! Transverse loads from the coolant crossflow, earthquake acceleration, and possible j vibrations will be distributed by the support columns to the top support plate and I

upper core plate.

j The evaluation of the dynamic analysis of the reactor internals is discussed in Section 3.9.2 of this report. The main objective of the design analysis is to satisfy allowable stress limits, to assure an adequate design margin, and to establish

deformation limits that are primarily concerned with the functioning of the components.

Both high and low cycle fatigue stresses will be considered in the determination of the allowable amplitude of oscillation.

4-8

As part of the evaluation of design loading conditions, extensive testing and inspection of raw materials will be performed during the range of activity from component installation to plant operation.

The design bases for the mechanical design of the reactor vessel internals components f include the following:

l (1) The reactor coolant flow through the core shall be such that heat transfer performance requirements will be met for all modes of operation, and thst ro l

leakage from the flange will occur during reactor operation.

(2) The pressure vessel will be shielded from radiation damage and the required ductility will be maintained during all modes of operation.

(3) The core internals will withstand the mechanical loads resulting from a safe shutdown earthquake, an operating basis earthquake, and pipe rupture (including d asymmetric loads).

(4) The core and the internals will be adequately supported and the core will be intact with acceptable heat transfer geometry following transients arising from abnormal operating conditions.

(5) Following a design basis accident, the plant shall be capable of being shut down and cooled in an orderly fashion 50 that the fuel cladding temperature is kept within specified limits.

f l For normal operating conditions, downward vertical deflections of the lower core support plant will be negligible. For a loss-of-coolant accident plus a safe shutdown earthquake condition, the limiting deflection values have been specified.

The criteria for the core drop accident have been based on analyses that determine the total downward displacement of the internal structures following a hypothetical core drop resulting from the loss of the normal core barrel supports.

' The allowab'e stresses for normal operating conditions will be in conformance with Section 111 of the ASME Code. Both static and alternating stress intensities have been considered. The allowable stress limits during the design basis accident used for the reactor internals are in accordance with the 1974 Edition of the ASME Code, Subsection NG, and the criteria for faulted conditions.

We conclude that the design procedures and criteria proposed by the applicant f or the design of the reactor internals are in conformance with the acceptance criteria of Section 3.9.5 of the Standard Review Plan and constitute an acceptable basis for satisfying the applicable requirements of Criteria 1, 2, 4 and 10 of the General Design Criteria.

4-5 a

All the major material for the reactor vessel internals will be Type 304 stainless steel except for the bolts and dowel pins, which will be Type 310 stainless steel, l

and radial support key bolts, which will be Inconel 750. The materials to be useJ for the construction of comoonents of the reactor internals have been identified by specification and found to be in conformance with the requirements of Section Ill of the ASME Code and the applicable portions of Parts A, B, and C of Section II of the ASME Code.

The materials for reactor vessel internals, which will be exposed to the reactor coolant, have been identified and all of the materials are compatible with the l expected environment, as proven by ext.ensive testing and satisfactory performance.

General corrosion on all materials is expected to be negligible.

l The controls to be imposed on reactor coolant chemistry will provide reasonable assurance that the reactor vessel internals will be adequately protected during )

operation from conditions which could lead to stress corrosion of the materials and loss of component structural integrity.

The controls that will be imposed upon reactor internal components constructed of austenitic stainless steel satisfy the recommendations of Regulatory Guide 1.31,

" Control of Stainless Steel Welding," Regulatory Guide 1.34, " Control of Electrostag Weld Properties," and Regulatory Guide 1.44, " Control of the Use of Sensitized l Stainless Steel." Material selection, fabrication practices, examination procedures, and protection procedures to be performed in ac . dance with these recommendations will provide reasonable assurance that the austenitic stainless steel used for .

reactor internals will be in a metallurgical condition which reduces the suscepti-bility to stress corrosion cracking during service.

Conformance with the requirements of the A5ME Code and with the recommendations of the above regulatory guides constitutes an acceptable basis for meeting the appli-cable requirements of Criteria 1 and 14 of the General Design Criteria.

4.2.3 Reactivity Control Systems Functional Design The functional design of the reactivity control systems for the Sundesert plant has been reviewed to confirm that the systems will have the capability to shut down the reactor with appropriate margin during normal, abnormal, and accident conditions.

The reactivity control systems reviewed include the control rod drive system and the chemical and volume control system. The scope of review included layout drac ings and descriptive information for the systems and for the supporting systems that are essential for operation of the systems, 4-10

The applicant has agreed to submit a complete failure mode and effects analysis for the control rod drive system. On the basis that this system is similar in design to the previously approved systems for SWESSAR/RE$AR-35 (Docket No. STN 50-495) and l

Beaver Valley Unit 1 (Docket No. 50 334), wa find this commitment acceptable for the construction permit stage of review. At the operating license stage of review, i

our acceptance of the final design for the control rod drive system will be based l

on an approved failure mode and effects analysis.

With regard to the vulnerability of the control rod drive system to common mode failures, the applicant referenced Westinghouse documentation intended to demon-strate that acceptable safety criteria would not be exceeded even for an anticipated transient without scram event. Our review of these documents is presented in Section 15.4 of this report. As stated in Section 15.4, any changes necessary to meet the limits specified as a result of our review, can be incorporated in the design of the Sundesert plant prior to the completion of construction.

The applicant has stated that protection will be provided for the control rod drive system from the effects of high or moderate energy pipe breaks. Also, failure of the control rod drive mechanism cooling system will, in the worst case, result in an individual control rod trip or a full reactor trip. We find these design bases to be acceptable.

The accident analyses for the Sundesert plant were based on a conservative rod drop time of 2.3 seconds to dashpot entry, or approximately 85 percent of the rod cluster travel. This time was based on tests conducted at the Westinghouse Test Engineering Laboratory in the D-loop test facility. The applicant has committed to a technical specification which will require testing of the full length rod to drop times of less than or equal to 1.9 seconds. We find this limiting condition of operation to be within the bounds of the analyzei conditions and acceptable for the construction permit stage of review.

The chemical and volume contr' system will be designed to provide reactivity control for all normal modes of reactor operation. The rate of boration will be sufficient to take the reactor from full power operation to a one percent shutdown margin in the hot condition in less than 90 minutes with no rods inserted. This capability meets the applicable requirements of Criterion 26 of the General Design Criteria and, therefore, is acceptable. Additional discussion on the chemical and volume control system is presented in Section 9.3.4 of this report.

Based on our review, we conclude that the applicant's designs, design criteria, and design bases for the reactivity control systems and their supporting systems are in conformance with the Commission's regulations as set forth in the General Design Criteria. Therefore, we conclude that the proposed designs of the reactivity control systems are acceptable.

4-11

1 i

l Structural Materials i

j The propertles of the materials to De used in the control rod system were reviewed from the standpoint of adequate performance throughout the design life of the j plant, or the component. The properties of the materials to be selected for the control rod system must be equivalent to those given in Appendix 1 to Section 111 i of the ASME Code or Part A of Section 11 of the ASME Code, except that cold worked

[ austenitic stainless steels shall have a 0.2 percent of fset yield strength no

] greater than 90,000 pounds per square inch to minimize the probability of stress j corrosion cracking occurring in these systems, i

i The properties of the materials to De selected for the control rod system compo-nents, which will be exposed to the reactor coolant, will conform to Appendix 1 of f

Section Ill of the ASME Code or Part A of Section II of the ASME Code and the t

j requirement that the yield strength of cold worked stainless steel shall n.t exceed

, 90,000 pounds per square inch.

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The controls to be imposed upon the austenitic steel of the system conform to the l recommendations of Regulatory Guide 1.31 and Regulatory Guide 1.44. Fabrication

f. and heat treatment practices to be performed in accordance with these recommendations l provide added assurance that stress corrosion cracking will not occur during the design life of the component.

j The compatability of all materials to be used in the control rod system with the f reactor coolant, satisfies the criteria for Articles NB-2160 and NB-3120 of Section III l of the ASME Code. Cleaning and cleanliness control are in accordance with ANSI f Standard N45.2.1-1973, " Cleaning of Fluid Systems and Associated Components During l Construction Phase of Nuclear Power Plants" and Regulatory Gwide 1.37, " Quality Assurance Requirements for Cleaning of Fluid Systems and Associated Components of l

Water-fooled Nuclear Power Plants." Section C.3 of Regulatory Guide 1.37 recommends

that the water quality for final flushes of fluid systems and associated components I

be at least equivalent to the quality of the operating system water. With regard l

j to this recommendation, the applicant states the the dissolved oxygen content of l the water cannot be maintained at reactor coolant levels during flushing of open j systems. We concur with the applicant and conclude that the higher oxygen content i of the final flush is a practical necessity and is acceptable to us.

Conformance with the codes, standards and regulatory guides indicated above consti- j

. tutes an acceptable basis for meeting the applicable requirements of Criterion 26 ;

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of the General Design Criteria.

I 4.3 Nuclear Design l We have reviewed the nuclear design of the Sundesert plant. Our review was based l

on information supplied by the applicant in the Preliminary Safety Analysis Report 1

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and in response to our requests for additional information. Our review was con-ducted within the guidelines provided in Section 4.3 of the Standard Review Plan. i The nuclear design of the Sundesert plant is essentially identical to that of the l Koshkonong plant which has been previously reviewed.

4.3.1 Design Bases The General Design Criteria which are applicable to the nuclear design bases are as follows. Criterion 10 requires that the core be designed with appropriate margin to assure that fue' design limits are not exceeded. Criterion 11 requires a nega-tive prompt feedbe.ck coef ficient. Criterion 12 requires that power oscillations either be not presible or be detected and suppressed by the control system.

Criteria 13 and 20 require a control and monitoring system and a protection system which automatically initiate a rapid reactivity insertion to prevent exceeding fuel design limits in normal operation or anticipated transients. Criterion 25 requires that the protection system be designed so that a single malfunction or single operator error will not cause violation of fuel design limits. Criterion 26 requires that the control sytem be designed so that shutdown is assured when the most reactive rod is stuck out of the core and that a chemical shim system is provided which is capable of bringing the reactor to cold shutdown. Criterion 27 requires that the control system, in combination with the engineered safety features, control reac-tivity changes during accident conditions. Criterion 28 requires that reactivity accident conditions be limited 50 that no damage to the reactor coolant system boundary occurs.

We have reviewed the design bases presented by the applicant in the Preliminary Safety Analysis Report and conclude that they comply with the applicable General b Design Criteria. Therefore, we find that the design bases to be acceptable.

4.3.2 Design Description The P ~1iminary Sefety Analysis Report contains the description of the first cycle fuel loadin9 whicq will consist of three different enrichments and will have a first cycle of approximately one year. The enrichment distribution, burnable poison distr Bution, soluble poison concentration and plutonium content as a func-tion of core exposure are presented. Values given for the delayed neutron fraction and prompt neutron lifetime at beginning and end of the cycle are consistent with l those normally used and are acceptable. Our evaluations of the design aspects are presented in the ruosequent paragraphs.

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! Power Distribution i

The design bases affecting power distribution for the core are:

1 l (1) The peaking factor in the core will not be greater than 2.32 during normal l operation at full power in order to meet the initial conditions assumed in the loss-of-coolant accident analysis. l (2) Under abnormal conditions, including maximum overpower, the peak power will l not produce fuel melting.

t (3) The core will not operate, during ncrmal operation or anticipated operational occurrences, with a power distribution that will cause the departure f rom nucleate boiling ratio to fall below 1.3 (usin; .;., Westinghouse W-3 correlation !

with modified spacer effect),

i The applicant has described the manner in which the core will be operated and how the power distributions will be monitured to assure that these limits are met. The 1

{ core will be operated in the constant axial offset control mode which has been

} shown to result in peaking factors less than 2.32 for constant power and load j following operation.

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Two types of instrumentation systems will be provided to monitor the power dis-tributions in the core; (1) excore detectors which will monitor core power, axial l offset and azimuthal tilt, and (2) incore detectors which will provide the capabil-l ity for performing power distribution measurements. These systems are used

! successfully in operating reactors supplied by Westinghouse and we find their use i ]

acceptable for Suncesert. !

l Reactivity Coefficients The reactivity coefficients are expressions of the effect on core reactivity of changes in such core conditions as power, fuel and moderator temperature, and boron f

concentration. These coefficients vary with fuel burnup and power level. The l applicant has presented values of these coefficients in the Preliminary Safety j Analysis Report and has evaluated the uncertainties in these values.

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1 We have reviewed the calculated values of the reactivity coefficients and have concluded that they adequately represent the full range of expected values. The

applicant has presented comparisons of measured and calculated values for the f various coefficients. Based on these comparisons, the applicant concludes that the j accuracy of current methods is:

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(1) + 0.2 percent. change in ef fective multiplication factor f or the Doppler defect,

-5 change in effective multiplication factcr per degree Fahrenheit for (2) + 2 x 10 the moderator coefficient, l

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(3) 1 50 parts per million for the critical boron concentration with burnup, and

-6 change in effective multiplication factor per part per million for (4) 15 x 10 I the boron worth.

t We have reviewed the reactivity coefficients used in the safety analyses and con-clude that they conservatively bound the expected values, including uncertainties.

Further, the coefficients will be measured as a part of the startup physics testing to assure that actual values are within those used in the analyses.

Control Requirements To allow for changes in reacthity due to reactor heatup, load following, and fuel burnup with consequent fission product buildup, a significant amount of excess reactivity will be designed into the core. This excess reactivity will be con-trolled by a combination of soluble boron and control rods.

Soluble boron will be used to control reactivity changes due to: (1) the moderator defect from ambient to operating temperatures; (2) equilibrium xenon and samarium buildup; (3) fuel depletion and fission product buildup (that portion not con-trolled by lumped burnable poison); and (4) transient xenon resulting from load following.

Full length regulating rods will be used to control reactivity changes due to: (1) the moderator defect from hot zero power to full power; and (2) power level changes (Doppler).

Burnable poison rods will be used for racial flux shaping and to control part of the reactivity change due to fuel depletion and fission product buildup.

The applicant has provided data to show that adequate control will exist to satisfy the above requirements with enough additional control rod worth to provide a hot shutdown effective multiplication factor less than the design basis value of 0.982 3

during the initial and equilibrium cycles with the most reactive control rod stuck out of the core. In addition, the chemical and volume control system will be capable of shutting down the system, by adding soluble boron, and maintaining it shut down in the cold xenon and samarium free conditions at any time in core life.

These two systems will satisfy the reouirements of Criterion 26 of the General Design Criteria.

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) Comparisons have been made between calculated and measured control rod bank worths in operating reactors and in critical experiments. These comparisons lead to the

{ conclusion that bank worths may be calculated to within + 0.2 percent of the effec-l tive multiplication factor. In aadition, bank worth measurements will be performed as a part of the startup test program to assure that conservative values have been used in safety analyses.

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j Based on these comparisons, we conclude that the applicant has made suitably con-servative assessments of reactivity control requirements and that adequate reactivity will be provided to assure shutdown capability.

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1 i Provision will be made in the design for the use of part length control rods.

' I j However, Westinghouse has informed us that the use of part length rods has not been l 4

l completely analyzed. Therefore, use of part length control rods is currently not permitted on any operating reactor. Until an analysis that is acceptable to us is j completed, use of these rods will be prohibited.

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5 Control Rod Patterns and Reactivity Worths 4

j The full length control rods are divided into two categories; shutdown rods and regulating rods. The shutdown rods will always be completely out of the core when f the reactor is at operating conditions. Core power changes will be made with regulating rods which are nearly out of the core when it is operating at full j power. Regulating rod insertion will be controlled by power-dependent insertion j limits which will be established to assure that:

(1) There will be sufficient negative reactivity available to permit rapid shut-3 down of the reactor with adequate margin.

! (2) The worth of a control rod that might be ejected will not be greater than that j- which has been shown to have acceptable consequences in the safety analyses.

j We have reviewed the calculated rod worths and the uncertainties in these worths, j and conclude that rapid shutdown capability will exist at all times in core life

assuming the most reactive control rod assembly is stuck out of the core.

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Stability I.

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1 The stability of the Sundesert core to xenon induced spatial oscillations is dis-l cussed in the Preliminary Safety Analysis Report. The overall negative reactivity J

(power) coef ficient will provide assurances that the reactor will be stable against

total power oscillation. We also conclude that sustained radial or azimuthal j oscillations are not possible. This conclusion is based on measurements on the 1

H. B. Robinson plant (Docket No. 50-261), an operating reactor of the same dimensions as Sundesert, which showed stability against these oscillations.

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Unstable axial oscillations are predicted to occur af ter about 12,000 fr.egawatt days per ton for this core. The applicant has shown that axial oscillations may be controlled by the regulating rods to prevent reaching any fuel safety limits.

Criticality of Fuel Assemblies Criticality of f uel assemblies outside the reactor w. il be precluded by adequate design of fuel transfer and storage facilities. The applicant presents information on calculational techniques and assumptions used to assure that criticality is avoided. We have reviewed this information and the criteria which will be employed and find them to be acceptable.

i Vessel Irradiation Values are presented for the estimated flux in various energy ranges at the mid-height of the pressure vessel inner boundary. Core flux shapes calculated by standard design methods were used as input to ! transport theory calculations to obtain a 10 value of 2.9 x 10 neutrons per square centimeter per second for the flux of neutrons having energy greater ttan one million electron volts at the v9ssel Dout.J .',

I9 This results in a fluence of 2 9 x 10 neutrons per square centimeter for a forty year vessel life with an 80 percent use factor. We conclude that acceptable values for the vessel fluence have been presented.

4.3.3 Analytical Methods A summary description of the methous used in the nuclear design of the Sundesert reactor is presented in the Preliminary Safety Analysis Report. Comparisons be-tween calculat 4 n and experiment are also given which permit evaluation of uncer-tainties in the calculations. Based on our review, we conclude that the methods used are state of the art and are acceptable.

4.3.4 Summary of Evaluation and Conclusions l

The applicant has described the computer programs and calculational techniques used l

to predict the nuclear characteristics of the reactor design and has provided examples to demonstrate the ability of these methods to predict experimental results.

We conclude that the information presented adequately demonstrates the abil,cy of these analyses to predict reactivity and physics characteristics of the Sundesert plant.

l To allow for changes of reactivity due to reactor heatup, changes in operating con-ditions, fuel burnup, and fission product buildup, a significant amount of excess reactivity will be designed into the core. The applicant has provided substantial information relating to core reactivity balances for the first cycle and has shown that means will be incorporated into the design to control excess reactivity at all 4-17

times. The applicant has shown that sufficient control rod worth will be available to shut down the reactor with a subcritical margin of at least 1.8 percent of the effective multiplication factor in the hot condition at any time during the cycla with the most reactive cnntrol rod stuck in tha fully withdrawn position.

On the basis of our review, we conclude that the applicant's assessment of 'eactiv-ity control requirements over the first core cycle is suitably conservative, and that adequate negative worth will be provided by the control system to assure shutdown capability. Reactivity control requirements will be reviewed for additional cycles as this information becomes available during plant operation. We also conclude that nuclear design bases, features, and limits have been established in conformance with the requirements of Criteria 10, 11, 12, 13, 20, 25, 26, 27 and 28 of the General Design Criteria and are, therefore, acceptable.

4.4 Thermal and Hydraulic Design 4.4.1 Thermal-Hydraulic Design Criteria and Design Bases The principal safety criteria, presented in the Preliminary Safety Analysis Report, to be used for the thermal and hydraulic design of the reactor are as follows:

(1) Fuel damage (defined as penetration of the fission product barrier; i.e. , the f uel rod clad) should not occur during normal operation and operational tran-sients (Condition I scents) or any transient conditions arising from f aults of moder;te frequency (Condicion 11 events), except possibly for a very small number of rods. The number of rods damaged. during these transients shall be small enough to be within the capability of the plant cleanup system and con- i sistent with the plant r.esign bases.

(2) The reactor will be designed 50 that it can be brought to a safe state follow-ing a transient aris'ng from an infrequent fault (Condition III event) with only a small fraction of fuel rods camaged (see above definition for fuel damage) although sufficient fuel damage might occur to preclude resumption of operation without considerable outage time.

(3) The reactor will be designed so that it can be brought to a safe state and the core kept subcritical with acceptable heat transfer geometry following tran-sients arising from limiting faults (Condition IV events).

These criteria will be implemented through the following proposed thermal-hydraulic design bases on departure from nucleate boiling ratio, hydrodynamic stability and l fuel temperature:

(1) Departure From Nucles.te Boiling Rat: 0- Considering plant parameter uncertain-ties, there must be at least a 95 pertep,t probability that departure from nu::leate boiling will not occur on the limiting fuel rod during Condition 1 I

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I and Il events. The departure from nucleate boiling ratio limit for the correla-tion is established based on the variance of the correlation such that there is a 95 percent probability with a 95 percent confidence level that departure from nucleate boiling will not occur on the limiting rod when the calculated departure from nucleate boiling ratio is at its limit.

(2) Hydrodynamic Stability - Modes of operation associated with Condition I and 11 events shall not lead to hydrodynamic instability.

(3) Fuel Temperature - During modes of operation associated with Condition I and Il events, the maximum fuel temperature, for at least 95 percent of the peak linear heat generation fuel rods, will not exceed to melting temperature of uranium dioxide with a 95 percent confidence level. The melting temperature of s cradiated uranium dioxide is 5080 degrees Fahrenheit and decreases with burrup by 58 degrees Fahrenneit per 10,000 megawatt days per metric tonne of uranium.

Both the Sundesert thermal-hydraulic design criteria and the design bases use the classification of events in ANSI Standard 18.2, " Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants." According to this classi-fication, the three pump loss of-flow transient is a Condition ill event and, therefore, the applicant does not consider this transient to be subject to the design bases stated above. We consider this transient to be an anticipated opera-tional occurrence, as defined by the General Design Criteria, and require that it meet the Condition !! design bases. However, based on our review, this requirement will not result in any change in the Sundesert design since the analyses in Chapter 15 of the Preliminary Safety Analysis Report indicate that the three pump loss-of-flow transient does meet the above design bases.

Departure From Nucleate Boiling In order to implement the design basis; i.e., that there is a 95 percent probability that departure from nucleate boiling does not occur on the limiting rod, each operating parameter, nuclear and thermal parameter, and fuel fabrication parameter was treated conservatively and, where statistical data were available, a fixed value of that parameter was used such that there was a 95 percent probability of not having a more adverse value of that parameter. We have found this procedure acceptable on previous applications and on that basis conclude that it is also acceptable for Sundesert.

The departure from nucleate boiling correlation which was used in implementing the design basis for the departure f rom nucleate boiling ratio is the Westinghouse W-3 critical heat flux correlation. This correlation and the associated departure from nucleate boiling ratio limit of 1.30 for R grid fuel designs, have been found acceptable by us, for the W-3 correlation, analysis indicates that a minimum 4-19

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departure from nucleate boiling ratio of 1.30 corresponds to at least a 95 percent i

4 probability at a 95 percent confidence level of not experiencing the critical heat j flux.

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j A significant parameter that influences the thermal-hydraulic design of the core, including the departure from nucleate boiling ratio, is rod-to rod bowing within fuel assemblies which has been observed in operating reactors. Only limited experi-l mental data are available on the extent of rod bowing in the 17x17 fuel design.

However, acceptable methods to evaluate this influence are available based on data f obtained with the 15x15 fuel design. We have also determined that other design margins exist to offset the presently imposed penalty on operating reactors due to g rod bow. The Sundesert thermal-hydraulic design does not include the effects of

{ possible fuel rod bowing beyond the manufacturing tolerances. However, the applicant j has committed to do the following for the final design stage and provide the results

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1 in the Final Safety Analysis Report:

(1) Fully define the gap closure rate for prototypical bundles; i

j (2) Determine by appropriate experiments the departure from nucleate boiling

! effect that bounds the effect of gap closure; and I t

(3) include the effects of rod bowing in the final design and analysis.

/ J j We find this commitment to be acceptable,

, Hydrodynamic Stability i

In steady-state, two phase heated flow in parallel channels, the potential for hydrodynamic instability always exists. For years, Westinghouse has used the HYDNA

, code to predict the inception of hydrodynamic instability for its reactors. The HYDNA code assumes that the core consists of parallel closed channels. Westinghouse

] performed experiments intended to demonstrate that flow in parallel open channels is more stable than in parallel closed channels. Westinghouse's experimental data were provided in Topical Report WCAP-7240, "An Experimental Investigation of the Effect of Open Channel Flow on Thermal-Hydrodynamic Flow," which is referenced by the applicant. This report does not describe the HYDNA code nor the details of its i

use in reactor calculations. We reviewed the topical report and concluded that,

, while the experimental data is useful as background information, it alone is not j sufficient to support a conclusion that the HYDNA code conservatively predicts the onset of flow instability in the core. We also concluded that the experiments described by Westinghouse in support of the hydrodynamic design are not st.fficient to justify that the design basis is satisfied.

' While we have not accepted the analysis of the hydrodynamic stability for Sundesert because the HYDNA model has not been prnvided for our review, we nevertheless 4-20

recognize that the thermal-hydraulic characteristics for Sundesert can reasonably be expected to exhibit characteristics similar to other open channel pressurized l_

water reactor designs. Based upon the thermal-hydraulic characteristics of similar l

pressurized water reactors and past operating experience with these designs, we J conclude that the Sundesert reactor will contain sufficient margins with respect to hydrodynamic stability without the need for design changes and, therefore, the proposed design is acceptable for the construction permit stage of review.

j We are performing a generic study of the hydrodynamic stability characteristics of pressurized water reactors which will include an evaluation of the analytical methods to be used for Sundesert. Therefore, modifications to the analytical methods that may result from this study will be applicable to Sundesert. In this event, the applicant has provided a commitment to comply with the generic resolu-tion of this matter. We find this commitment to be acceptable.

Fuel Temperature The design basis regarding fuel temperature will be implemented through the reactor protective system overpower trip setpoints which will assure that the calculated fuel centerline temperature will not exceed 4700 degrees fahrenheit. Therefore, we conclude that the fuel temperature design basis will be met.

4.4.2 Thermal-Hydraulic Analytical Methods i l

l For Sundesert and other recently reviewed Westinghouse designed reactors, the THINC-IV computer code has been used to calculate core thermal-hydraulic perform-ance characteristics. This code considers cross-flow between adjacent assemblies in the core and cross-flow and thermal diffusion between adjacent subchannels in the assemblies.

The THINC-IV program is described in Westinghouse Topical Reports WCAP-7956, "THINC-IV An Improved Program for Thermal and Hydraulic Analysis of Rod Bundle Cores," and WCAP-8054, " Application of the THINC-IV Program to PWR Design." We have reviewed these reports and conclude that the use of the THINC-IV code is acceptable as documented by letter to Westinghouse, dated April 19, 1978.

We also reviewed the input assumptions used in the THINC-IV code for the Sundesert design and questioned the treatment of postible crud buildup in the core. Crud deposition in the core and an associated change in core pressure drop and flow have been observed on some pressurized water reactors. Regarding crud buildup in the core, the applicant has stated that: (1) operating experience from several Westinghouse reactors indicates very low levels of crud buildup on the core; (2) some margin for uniform crud buildup is included in the clad surface roughness factor used in the analysis; and (3) significant changes in core pressure drop and flow would be observed during periodic core flow measurement. We have reviewed 4-21

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, this information and the list of instrumentation to detect significant changes in

, core flow and conclude that it adequately treats our concerns relative to crud deposition in the core. At the operating license stage of review, we will assure i that the technical specifications provide appropriate considerations for detection t

j and actions relevant to significant crud deposition.

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j We also questioned the effect of a radial pressure gradient at the core exit on the

} thermal-hydraulic design. This matter was first raised as a result of our review of the Westinghouse 1/7 scale hydraulic tests which showed a radial pressure gradient in the upper plenum. The thermal-hydraulic design analyses assume a uniform core

} outlet pressure distribution, although the effects of the core exit radial pressure gradient can be explicitly accounted for in the thermal-hydraulic design calculation.

In response to our question on the radial pressure gradient, the applicant referenced j

a sensitivity study on radial pressure gradient provided in Topical Report WCAP-8054. l

[ The study compares the minimum departure from nucleate boiling ratio for a uniform pressure gradient with the minimum departure from nucleate boiling ratio for a

, cosine pressure distribution, having a five pounds per square inch increase at the 4

core center. This comparison showed that the effect on the minimum departure from i nucleate boiling ratio was negligible.

4 l A similar sensitivity study performed by our consultants, Battelle Pacific Northwest 3 i Laboratory, by using the COBRA-IV code, showed a small ef fect of pressure gradient !

j on the minimum departure from nucleate boiling ratio. The results of these independ- 1 i

l ent calculations, which are in substantial agreement, show that the effect is small j

relative to the THINC-IV design calculations using the Westinghouse WRB-1 departure l from nucleate boiling correlation for conditions which could exist during normal j operation and anticipated transients. However, similar results have not been j provided for the W-3 R grid departure from nucleate boiling correlation which will J

be used for the Sundesert design. we are currently pursuing this aspect with )

Westinghouse on a generic basis, althougn we do not anticipate that the comparative

] results for the W-3 R grid correlation will dif fer significantly with the compara-j tive results already presented for the WRB-1 correlation. In response to our i

a concern, the applicant has provided a commitment to comply with the generic resolu-tion of this matter. We find this commitment to be acceptable.

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] 4.4.3 Thermal-Hydraulic Design Comparison 1 I

4 q A thermal-hydraulic design comparison and fuel design comparison between the Sundesert j and Koshkonong designs are provided in Tables 4.3 and 4.4, respectively. This j comparison indicates that the thermal-hydraulic designs for the two plants are j essentially identical, a

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l TABLE 4.3 THERMAL-HYDRAULIC DESIGN COMPARISON ,

THERMAL-HYDRAut1C DESIGN PARAMETER SUNDE5ER1 K0$HKON0NG l

Core Power (thermal megawatts) 2775 2775 l

l Minimum System Pressure (pounds per square inch absolute) 2200 2200 Reactor Coolant System flow (106pounds per hour) 107.5 107.5 Coolant Inlet Temperature (degrees Fahrenheit) 554.8 554.8 1 l

l Enthalpy Rise Factor I.55 1.55 Departure from Nucleate Boiling Ratio Correlation W-3 (R grid) W-3 (R grid)

Departure from Nucleate Boiling Ratio Correlation Limit (95 percent probability with 95 percent confidence) 1.30 1.30 Minimum Departure from Nucleate Boiling Ratio for Design Transients >1.30 >1.30 Average Heat Flux (British thetmal units per hour per square foot) 189,000 189,000 Overall Peaking Factor 2.32 2.50 Maximum Heat flow for Normal Operation (British thermal units per hour per square foot) 440,300 474,300 TABLE 4.4 FUEL DESIGN COMPARISON SUNDE5ERT K05HKONONG Fuel Assembly Design 17x17 17x17 193 193 Number of Assemblies Heat Transfer Surface Area (square feet) 48,600 48,600 Average Linear Heat Rate (kilowatts per foot) 5.44 5.44 Peak Linear Heat Rate for Normal Operation (kilowatts per foot) 12.6 13.6 Maximum Linear Heat Rate at Overpower Trip (kilowatts per foot) 18.0 18.0 Peak Fuel Temperature at 18 kilowatts per foot (degrees Fahrenheit) >4700 >4700 4-23

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i i 4.4.4 Conclusions We have reviewed the thermal-hydraulic design of the core, including the design criteria, design bases, methods of analysis, and the steady-state analysis of the

{ core thermal-hydraulic performance. Based on our review, we conclude that the design criteria and design bases are adequately conservative and that the proposed l thermal-hydraulic design will have adequate margin to meet the criteria and bases.

l In the event that the analytical methods still under review are determined not to be conservative during the final design review, appropriate restrictions on opera-4 tion can be established at the operating license stage. Therefore, we conclude that the proposed preliminary design is acceptable.

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l 5.0 REACTOR COOLANT SYSTEM 5.1 Summary Description The reactor coolant system for each unit of the Sundesert plant will include the reactor vessel, the control rod drive mechanism housings, the steam generators, the reactor coolant pumps, a pressurizer, and the interconnecting piping and v11ves associated with those components, The reactor coolant system will consist of three coolant loops connected in parallel to the reactor vessel, with each loop containing a steam generator and a reactor coolent pump. The pressurizer will be connected to one of the three loops.

During operation, the reactor coolant system will transfer the heat generated in the core to the steam generators where steam will be produced to drive the turbine-generator. Borated demineralized water will be circulated in the system at a flow rate, pressure and temperature consistent with achieving the design thermal-4 1

hydraulic performance of the reactor core. The water will also act as a radiation shield, and neutron moderator and reflector.

The proposed reactor coolant system design for Sundesert is essentially the same as that previously reviewed and found acceptable for Koshkonong (Docket Nos. STN 50-502 and STN 50-503).

5.2 Integrity of the Reactor Coolant Pressure Boundary 5.2.1 Compliance With Codes and Code Cases Components of the reactor coolant pressure boundary, as defined by Section 50.55a of 10 CFR Part 50, have been properly identified and classified as ASME Section 111, Class I components in Table 5.2.1 of the Preliminary Safety Analysis Report. These components within the reactor coolent pressure boundary will be constructed in accordance with the requirements of the applicable codes and addenda as specified by Section 50.55a of 10 CFR Part 50.

The applicant has made a commitment that no ASME code cases considered unacceptable to the Commission will be applied in the construction of pressure-retaining ASME Section 111, Class I components within the reactor coolant pressure boundary (Quality Group Classification A). In the event the use of new ASME Council approved code cases are planned, authorization will be requested of the Commission prior to their. spphcotion in the construction of Section III, Class 1 components. The applicint has also indicated it trill comply with Regulatory Guide 1.84, " Code Case Acceptability - ASME Section III Design and Fabrication," and Regulatory Guide 1.85,

" Code Case Acceptability - ASME Section III Materials."

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j We conclude that construction of the components of the reactor coolant pressure

, boundary in conformance with the ASME code and approved code cases and the Commis-sion's regulations provides adequate assurance that component quality will be i commensurate with the importance of the safety function of the reactor coolant i

j pressure boundary and, therefore, is acceptable, i

4 j 5.2.2 Overpressurization Protection i

The function of the pressure relief system is to prevent overpressurization of the j reactor coolant pressure boundary under the most severe transients and to limit the I

reactor pressure during normal operatioral transients. Overpressure protection for

{ the reactor coolant pressure boundary will be accomplished by utilizing three j spring-loaded safety valves and three power-operated relief valves to be located on s

i the pressurizer. The steam release from these valves will discharge to the pres-l surizer quench tank through a common header from the pressuri7er. The reactor j coolant system safety valves, in conjunction with the steam generator safety valves i and the reactor protection system, will protect the reactor coolant system against i

j overpressure above 110 percent of the design pressure (2500 pounds per square inch)

I following a complete loss of steam flow to the turbine, l

1 j The applicant has referenced Westinghouse Topical Report WCAP-7769, " Overpressure j Protection for Westinghouse Pressurized Water Reactors," Revision 1, as the basis

for the design requirements of the overpressure protection system for the Sundesert j plant. In this report, the overpressure analyses were performed for two major l cases. The first case considered a complete loss of steam flow and assumed that l main feedwater flow was maintained with no credit taken for reactor trip. This j analysis was performed as a conservative method of sizing the pressurizer safety valves based on mar,imum surge rate. The second case involved a complete loss of j steam flow with a simultaneous' loss of all feedwater and with credit for reactor j trip. This analysis was performed to verify the adequacy of the sizing method.

l The assumptions used in these analyses included no credit for onore Mn of reactor 1

j coolant system power-operated relief valves, steam line power clief valves, steam f dump system, reactor coolant system pressurizer level contNI system, and pressur-i izer spray. The plant was assumed to be operating at a power level corresponding

to the engineered safeguards design rating of 2910 megawatts thermal, which is equivalent to approximately 104 percent of the rated nuclear steam supply power of

2785 megawatts thermal.

d b

The results of the analyses show that for the complete loss of steam flow transient j with a simultaneous loss of all feedwater, and credit talien for reactor trip on l low-low steam generator water level, the peak pressurizer safety valve flow capacity a

would be 86 percent of rated. While our generic review of WCAP-7769 has not been completed, the margin for overpressure predicted in this report, when applied to the Sundesert plant, would be acceptable. Also, the analyses in this report were performed for a four loop, 3423 megawatts thermal plant which is a more restrictive l

5-2

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case for overpressure protection than for a three loop plant such as Sundesert.

l The applicant has justified the use of this topical report by showing that the ratio of available pressurizer safety valve capacity to peak surge rate into the pressurizer during the sizing transient is greater for a three loop plant (e.g.,

'1.21 for Sundesert) than the ratio for the four loop plant (1.056). Therefore, we conclude that for Sundesert the proposed design of the pressure relief system during power operation is acceptable.

There have been reported incidents of reactor vessel overpressurization in pressur-ized water reactors during startup and shutdown, when the reactor coolant system is water solid, in which the limitatinns of Appendix G to 10 CFR part 50 have been exceeded. The applicant has recognized this potential source of overpressurization and has proposed an automatic pressure contral feature for the reactor coolant system to maintain pressures within allowable limits during low temperature opera-tion. This feature will be provided by incorporating an independent actuation logic to each of the two pressurizer power-operated relief valves. The system logic will continuously monitor reactor coolant system temperature and pressure condith:M whenever plant operation is at a temperature below the reference nil ductility temperature, and will actuate a signal to open the power-operated relief valves when required to prevent pre' ure-temperature conditions from exceeding allowable limits. This system will be testable and the electrical power supply to the control circuit of the power-operated relief valves will be independent of offsite power. The applicant has stated that the Appendix G limits will be main-tained for faults of moderate frequency (Condition Il events) during low temperature operation assuming water solid operation. The transient events considered included inadvertent startup of a charging / safety injection pump or startup of a reactor coolant pump, and energization of all pressurizer heaters curir.w startup.

However, the applicant has not provided sufficient information to demonstrate that the proposed pressure control feature will provide the overpressure protection capability for the reactor vessel when the reactor coolant system is water solid.

Therefore, we require the following additional information in order to complete our review:

I (1) The results of analyses which identify the most limiting overpressure transient and which demonstrate that the system will provide the required pressure relief c 9acity assuming the most limiting single failure. .

(2) #;i evaluation which shows that the system would remain functional af ter an operating basis earthquake in orcer to take credit for not meeting seismic Category I requirements for the entire protection system. i l

(3) A discussion of how the reactor coolant system pressure control system will be j enabled when the plant is brought into a shutdown or startup mode of operation.

1 5-3

If automatic, the discussion should show that the potential for inadvertent actuation during power operation with the automatic system would be no greater than with a manual procedure.

Therefore, the above matter remains outstanding.

Additionally, since the pressurizer puwer-operated relief valves are normally designed for steam flow conditions, at the operating license stage of review we will require that the applicant show that these valves are also capable of $

relieving water with a sufficient flow rate to satisfy Appendix G limits in the event of an overpressurization occurrence while the reactor coolart system is water ;

solid.

Based on our review, and subject to the satisfactory resolution of the above matter relating to the proposed pressure control feature for the overpressure protection system when the reactor coolant system is water solid, we conclude that the pressure relief system conforms to all applicable regulations, regulatory guides and staff positions and is acceptable.

5.2.3 Reactor Coolant Pressure Boundary Materials j

The composition and mechanical properties of the materials specified for the reactor coolant pressure boundary are required to conferm to the requirements identified in Appendix 1 to Section III of the ASME Code, an.1 specifi~f in detail in Parts A, B, and C of Section II of the ASME Code. The materials to .e used for the construction of the reactor coolant pressure boundary must be compatible with the reactor coolant.

Since corrosion and stress corrosion cracking can be irduced by impurities in the reactor coolant, we require that the chemistry of the reactor coolant be monitored and controlled. Therefore, sampling and chemical analysis for chlorides, fluoride %

and oxygen must be performed on a scheduled basis.

The water chemistry in the reactor coolant system will be rigorously controlled to prevent the intrusion of aggressive species. In particular, the maximum permissible oxygen and chloride concentrations are 0.10 parts per million and 0.15 parts per l million, respectively. The use of hyorogen overpressure will limit the level of oxygen during operation, The effectiveness of these controls has been demonstrated I by both laboratory tests and operating experience.

The controls to be imposed on reactor coolant chemistry are in conformance with the i

recommendations of Regulatory Guide 1.44, " Control of Sensitized Stainless Steel,"

and provide reasonable assurance that the components of the reactor coolant pressure 1

boundary will be adequately protected during operation from conditions that could lead to stress corroston of the materials and loss of structural integrity of a a

i 5-4

1 i

component. Conformance with the recommendations of Regulatory Guide 1.44 consti-tutes an acceptable basis for satisfying the applicable requirements of Criteria 14 and 31 of the General Design Criteria. l All of the ferritic low alloy and carbon steels to be used in principal pressure retaining applications will be provided with corrosion resistant cladding on all surfaces which will be exposed to the reactor coolant. The cladding material will have a chemical composition that has the equivalent corrosion resistance of J. Types 304 and 316 austenitic stainless steel alloys, I

The reactor coolant pressure boundary f ei . ic materials with cladding have been l identified and all of the materials are compatible with the expected environment, as proven by extensive testing and satisfactory performance. General corrosion of all materials is expected to be negligible.

The thermal insulation to be used on the reactor coolant pressure boundary will

' either be made of reflective stainless steel or made of compounded materials which The compounded yield low leachable chloride and/or fluoride concentrations.

materials will be silicated to provide protection of austenitic stainless steels against stress corrosion which may result f rom accidental wetting of the insulation a

by spillage, minor leakage or other contamination from the environmental atmosphere.

The thermal ins.i a ion will be compatible with the materials of construction of the reactor coolant pressure boundary and will be in conformance witt the recommenda-tions of Regulatory Guide 1.36, " Nonmetallic Thermal Insulation for Austenitic Stainless Steels."

In order to prevent sensitization, wrought austenitic stainless steel material to be used for the reactor coolant pressure boundary will utilize one of the following conditions during fabrication:

(1) Solution annealed and water quenched; or i '

2 (2) Solution annealed and cooled through the sensitization temperature range within less than approximately five minutes.

Sensitization of the heat affected zones of welded components will be avoided by l

i control of the welding parameters and welding processes. For these contrris, the 1 l

a heat input and associated cooling rate are of primary importance.

). The controls to be imposed upon components constructed of austenitic stainless l

steel and to be used in the reactor coolant pressure boundary will conform to the recommendations of Regulatory Guide 1,31, " Control of Stainless Steel Welding,"

j Regulatory Guide 1.34 " Control of Elt1ctroslag Weld Properties" and Regulatory I

+

5-5 i

w x a e

i a

4 Guide 1.44, " Control of the Use of Sensitized Stainless Steel." Material selec-l tion, f abrication practices and examination procedures performed in accordance with f these recommendations will provide reasonable assurance that the austenitic stain-

{

1ess steel in the reactor coolant pressure boundary will be in a metallurgical

{ condition which minimizes susceptibility to stress corrosion cracking during service.

Conformance with these regulatory guides constitutes an acceptable basis for meeting

}

i the applicable requirements of Criteria 1 and 14 of the General Design Criteria.

Assurance of adequate fracture toughness, under test, normal and transient condi-l tions, of ferritic materials in the steam generators, auxiliary pressure vessels and tanks, will be provided by compliance with the requirements for fracture tough-

]

ness testing included in NB-2300 of Section III of the ASME Code and Appendix G to i

10 CFR Part 50.

I For low alloy ferritic maiarials with specified minimum yield strengths greater i than 50,000 pounds per square inch, the applicant states that a test program is being conducted to demonstrate compliance with Appendix G to Section Ill of the ASME Code.

SA-533 Class 2 steel, having a minimum specified strength of 70,000

)

pounds per square inch, is one of the materials that might be used as plate material in the pressuri2er, steam generators and reactor vessel (other than the core region). We require that this test program on high strength ferritic materials, including SA-533 Class 2 steel, be performed to also be in conformance with Note 1 of ASME Code Case 1528-3 (which has been incorporated into the 1977 Edition of Appendix G to Section Ill of the ASME Code). Therefore, this matter remains outstanding.

1 1

j The fracture toughness tests required by the ASME Code, as augmented by Appendix G to 10 CFR 50, provide reasonable assurance that adequate safety margins against non-ductile behavior or rapidly propagating fract re will be established for all

{ pressure retaining components of the reactor coolant pressure boundary under rpera-j ting, testing, maintenance, and postulated accident conditions.

s i

t The controls to be imposed on welding preheat temperatures are in conformance with the recommendations of Regulatory Guide 1.50, " Control of Preheat Temperature for i i

Welding Low Alloy Steels," as modified by the procedures described in Westinghouse Topical Report WCAP-8577, "The Application of Preheat Temperatures After Welding of l Pressure Vessel Steels." We have reviewed the procedures in WCAP-8577 and found j I

them to be an acceptable alternative to the recommendations of Regulatory '

! Guide 1.50 as documented by letter to Westinghouse, dated June 18, 1976, j l

i I The controls to be imposed on electroslag welding of ferritic steels will also be in conformance with the recommendations of Regulatory Guide 1.34.

I Compliance with the provisions of the ASME code and tammission regulations and con-formance with the recommendations of the regulatory guides constitute an acceptable i

),

5-6 ,

1

basis for satisfying the applicable requirements of Criterion 31 of the General Design Criteria.

5.2.4 Inservice inspection Program i

Criterion 32 of the General Design Criteria requires that components which are part of the reactor coolant pressure boundary be designed to permit periodic inspection and testing of important areas and features to assess their structural and leaktight integrity. Inservice inspection programs are based on Section XI of the ASME Code.

The actual inservice inspection and testing program for Sundesert will reflect the final design of the plant and a detailed description of the program will be pro-vided in the Final Safety Analysis Report. At this, the construction permit stage of review, we evaluated tha provisions for accessibility and inspectability of the components that will be included in the design of those portions of the reactor coolant system boundary that will be subject to inspection.

The design of the reactor coolant system for the Sundesert plant will incorporate provisions for access for inservice inspection in accordance with Section XI of the ASME Code. To assure that no deleterious defects develop during service, accessibility for selected welds and weld heat-af fecte f7es will be provided to permit periodic inspection. Methods will be developed to facilitate the remote inspection of those areas of the reactor vessel not readily accessible to inspec-tion personnel. The applicable Edition and Addenda of Section XI of the ASME Code I will be determined in accordance with the rules of 10 CFR Part 50, paragraph 50.55a(g) and will be addressed in the Final Safety Analysis Report. l The conduct of periodic inspections and hydrostatic testing of pressure retaining components of the reactor coolant pressure boundary, in accordance with the require-ments of Section XI of the ASME Code, will provide reasonable assurance that evidence of structural degradation or loss of leaktight integrity occurring during service will be detected in time to permit corrective action before the safety functions of a component are compromised. Compliance with the inservice inspec-tions required by the ASME Code constitutes an acceptable basis for satisfying the requirements of Criterion 32 of the General Design Criteria.

5.2.5 Reactor Coolant Pressure Boundary leakage Detection Our review of the reactor coolant pressure boundary leakage detection systems included the proposed leakage detection methods for continuous monitoring of both identified and unidentified leakage rates, intersystem leakage, leakage detection system sensitivity and response times, seismic capability of the systems, indicators and alarms, and system testability.

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4 4

i l The three primary leak detection methods for the facility will be containment sump 4

j level and inflow monitoring, containment atmosphere particulate radioactivity j monitoring, and containment atmosphere gaseous radioactivity monitoring. A fourth method available for indirect indication nf leakage will be the monitoring of containment pressure, temperature, and humidity.

s 1

l Identified leakage will be determined by monitoring the rate of drainage collection d

in the reactor coolant drain tank, which will be equipped with connected leakoff f paths to such components as valve stems, reactor coolant pump shaft seals, and J

j pressurizer relief valves.

Intersystem leakage detection methuds will include sampling and monitoring for i radioactivity in the secondary side of the steam generator. Additional intersystem leakage detectior capability will be available by means of periodic testing of the emergency core cooling system components as discussed in Section 6.3.5 of this report. We find these capabilities to be acceptable. At the operating license stage of review, we will confirm that provisions have been included to monitor systems connected to the reactor coolant system pressure boundary, such as the residual heat removal system, for signs of intersystem leakage.

, Containment sumps will De used to collect and determine the rate of unidentified leakage. The sump level instrumentation will be capable of detecting the leakage rate within an accuracy of one gallon per minute within on hour, which is in i accordance with the criteria of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," and is, therefore, acceptable.

The containment airborne radiation monitoring systems for gas and particulates will be designed to detect primary system leakage of one gallon per minute in less than one hour, in conformance with the recommendations of Regulatory Guide 1.45, provid- I ing there is no prior reactor coolant leakage within the containment (i.e., the existing reactor coolant activities are low). The applicant has state'd that, with an assumed prior leakage of 0.3 gallons per minute, the response time of the particulate and the gas detectors for additional leakage will be greater than one hour for an increase in the leak rate of one gallon per minute. Based on a recognition that such prior leakige can affect the sensitivity of the gas and particulate mor,itors, we are currently evaluating the practicality of the require-mants specified in Regulatory Guide 1.45 regarding the sensitivity response times l

of the containment airborne radiation monitoring systems. The conclusions result-I ing from this evalua'. ion will be applied to the Sundesert plant during the operat-ing license stage cf review. At this, the construction permit stage of review, we conclude that the sensitivity of the gas and particulate monitors proposed for Sundesert is acceptable.

The applicant has stated that the leakaoe detection systems will be capable of per-forming their functions following an operational basis earthquake and that the 5-8 E_ __ _ _ _ _ _ _- - -- - - --

airborne particulate monitoring system will be capable of performing its function after the safe shutdown earthquake. These criteria are consistent with the require-ments of Section 5.2.5 of the Standard Review Plan and, therefore, are acceptable.

Indicators and alarms for each leakage detection system will be provided in the control room, procedures for converting various indications to a common leakage eaulvalent will be available to the operators and will be described in the Final Safety Analysis Report. The applicant has stated that alarms will be provided for containment high radiation, containment high sump level, and containment high pressure. However, we require that ear.h leakage detection system also be set to alarm in the control room on an increase in leakage of one gallon per minute above the background level determined at the time of calibration. Therefore, this matter remains outstanding.

The applicant has stated that the leakage detection systems will be oesigned to permit testing and instrument calibration, and that the containment radiation monitoring systems will have a radioactive source (" check Source") built into the system to permit test and calibration during operation. With regard to testing and calibration of the sump monitoring system with background identified leakage, we require that the applicart include the capability to perform system operc.bility checks with background leakage so that the directly measured quantity of flow obtained from the sump can be used to calibrate the radiation moniturir.g systems.

Therefore, this matter remains outstanding.

The Sundesert technical specifications addressing the limiting conditions for operation on reactor coolant system leakage will permit a maximum of one gallon per minute unidentified leakage and ten gallons per minute identified leakage. These specifications are in conformance with the recommendations of Regulatory Guide 1.45 and are, therefore, acceptable.

We conclude that the leakage detection system to be provided will include suffi-ciently diverse leak detection methods, with adequate sensitivity to measure small leaks ar.d to identify the leakage sources within practical limits, sod will include suitable control room alarms and readouts. Our evaluation of sensitivity response time requirements for the teakage detection system with the existence of prior leakage will be performed at the operating license stage of review.

The systems to be provided to detect leakage from components of the reactor coolant pressure boundary will furnish reasonable assurance that structural degradation, which may develop in pressure-retaining components of the reactor coolant pressure boundary and result in coolant leakage during service, will be detected on a timely basis, so that corrective actions can be made before such degradation could become sufficiently severe to jeopardize the safety of the system, or before the leakage s could increase to a level beyond the capability of the makeup systems to replenish the coolant loss.

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l 4.

Subject to the satisfactory resolution of the above matters relating to the (1)

} setting of an additional alarm in the control room to indicate leak rates in excess j of one gallon per minute above background levels, and (2) capability to perform l operability checks with back0round leakaps- to calibrate the radiation monitoring i systems, we conclude that the propnsed design of the leakage detection systems for i

the reactor coolant pressure boundary conform to the requirements of Criterion 30

{ of the General Design Criteria and the recommendations of Regulatory Guide 1.45 and are acceptable.

i j 5.3 Reactor Vessel l 5.3.1 Reactor Vessel Materials i

l l

Criterion 31 of the General Design Criteria requires that the reactor coolant j pressure boundary be designed with suf ficient margin to assure that, when stressed l under operating, maintenance, testing, and postulated accident conditions, the boundary will behave in a non-brittle manner and the probability of rapidly propa-gating fracture is minimized. Criterion 32 of the General Design Criteria requires that the reactor coolant pressure boundary be designed to permit an appropriate material surveillance program for the reactor pressure vessel.

I We have reviewed material specifications for the reactor vessel and closure studs.

Thetir adequacy for use in the construction of such components was assessed on the g basis of their material, mechanical, and physical properties and the effects of f irradiation on these materials. Our review of the corrosion resistance and fabric-ability of the materials, and of the welding contrels and procedures for low alloy

and austeni',1c steel welds, is presented in Section 5.2.3 of this report.

I.

f The guidelines specified for the fracture toughness requirements for the ferritic

)

d materials of the reactor vessel and the closure studs will comply with Appendix G and Appendix H to 10 CFR Part 50. The ferritic materials will be ordered and, as i modified by our position as stated in Section 5.2.3 of this report regarding SA-533 i

i Class 2 steel material, will be tested in compliance with the acceptance criteria j of the Edition and Addenda of the ASME Code as required by Section 50.55a of 10 CFR j Part 50.

I Appendix G to Section III of the ASME Code will be used in calculating the reactor vessel oressure-temperature limitations. The procedures for monitoring and evaluating i

t irradiation induced changes in the reactor vessel beltline region will be in accordance with the American Society for Testing Materials ( ASTM) Standard E-185-73, " Standard f

Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels."

Fracture toughness tests performed in accordance with the ASME Code and with Appen-

, dix G to 10 CFR Part 'd provide reasonable assurance that adequate safety margins against the possibility of non-ductile behavice or rapidly propagating fracture can be established for the reactor vessel. The use of Appendix G to the ASME Code as a t

5-10

guide in establishing safe operating procedures, and use of the results of the fracture toughness tests performed in accordance with the ASME Code and Commission regulations, will provide adequate safety margins during operating, testing, main-tenance, and postulated accident conditions. Compliance with these code provisions and Commission regulations constitutes an acceptable basis for satisfying the requirements of Criterion 31 of the General Design Criteria.

A materials surveillance program periurmeu in accordance with Appendix H to 10 CFR Part 50 will provide information on material properties and the effects of irradia-tion on material properties so that (1) changes in the fracture toughness of the material in the reactor vessel beltline caused by es.posure to neutron radiation can be properly assessed, and (2) adequate safety margins against the possibility of vessel failure can be provided. Compliance with ASTM Standard E-185-73 and Appen-dix H to 10 CFR Part 50 assures that the surveillance program will monitor radiation induced changes in the fracture toughness of the reactor vessel material and constitutes an acceptable basis for satisfying the requirements of Criteria 31 and 32 of the General Design Criteria.

5.3.2 Pressure-Temperature Limits Appendix G and Appendix H to 10 CFR Part 50, describe the conditions that require pressure-temperature limits and provide the general basis for these limits. These appendices specifically require that pressure-temperature limits must provide safety margins at least as great as those recommended in Appendix G to Section III of the ASME Code, during heatup, cooldown, and test conditions. Appendix G to 10 CFR Part 50 requires additional safety margins whenever the reactor core is critical (except for low-level physics tests).

The pressure-temperature limits to be imposed on the reactor coolant pressure boundary during the following operations and tests are reviewed to assure that they will provide adequate safety margins against non-ductile behavior or rapidly propa-gating failure of ferritic components, as required by Criterion 31 of the General Design Criteria:

(1) Preservice hydrostatic tests; (2) Inservice leak and hydrostatic tests; j l

l (3) Heatup and cooldown operations; and l

(4) Core operation.

l Actual operating limit curves will be determined after fracture toughness tests and l l

other required tests have been performed on the actual material that will be used.

Limit curves will be calculated in accordance with the requirements of Appendix G 5-11

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to 10 CFR Part 50, and Appendix G to Section 111 of the ASME Code (Code Edition and j Addenda will be determined in accordance with the rules of 10 CFR Part 50, j Section 50.55a. and will be addressed in the Final Safety Analysis Report), and by l Using the Westinghouse suoplied curves to predict the shift in the reference nil ductibility temperature as a function of copper content and neutron fluence (for j reutrons with energy greater than 1,000,000 electron volts), as an alternative to the recommendations of Regulatory Guide 1.99, "Effect of Residual Elements on l Predicted Radiation Damage to Reactor Vessel Materials," Revision 1.

i 3 We have reviewed this alternative method of calculating the shift in the reference nil ductility temperature and find that it is acceptable, for the controlled composi-tion (0.12 percent copper, 0.17 percent phosphorus and 0.05 percent vanadium,

} maximum content) of the reactor vessels to be used at the Sundesert plant, until a f fluence of 1 x 10 0 neutrons per square centimeter has been experienced by the j inner surface of the reactor vessel. At the operating license stage of review, we will require that, when this fluence limit has been reached, the recommendations of j Regulatory Guide 1.99 be followed for calculating the shift in the reference nil l ductility temperature.

f

+

The pressure-temperature limits to be imposed on the reactor coolant system for all operating and testing conditions, to assure adequate safety margins against non-

ductile or rapidly propagating failure, are in conformance with estelished criteria, j codes and standards acceptable to the staff. The use of operating limits based on I these cr Meria, as u fined by applicable regulations, codes, and standards, provides reasonable assurance that non-ductile or rapidly progagating failure will not occur, and constitutes an acceptable basis for satisfying the applicable require-1 ments of Criterion 31 of the General Design Criteria.

i j 5. 3. 3 Reactor vessel Integrity 9

, In Sections 5.2.M, 5.3.1 and 5.3.2 of this report, we have reviewed all factors contributing to the structural integrity of the reactor vessel, including design, materials of construction, fabrication methods, inspection requirements and operating conditions. As a result of our review, we conclude that there are no

{ special considerations that make it necessary to consider potential reactor vessel j failure for this plant.. The bases for our conclusion are that the design, materials, 1 fabrication, inspection, and quality assurance requirements for the reactor vessel will conform to applicable Commission regulations and regulatory guides, and to the rules of Section llI of the ASME Boiler and Pressure Vessel Code. The stringent ,

j fracture toughness requirements of the Commission's regulations and Section 111 to 1 the ASME Code will be met, including requirements for surveillance of the reactor vessel material properties throughout service life. Also, operating limitations on temperature and pressure will be established for this plant in accordance with Appendix G to Section 111 of the ASME Code, and Appendix G to 10 CFR Part 50.

s.

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We conclude that the integrity of the reactor vessel is assured because the vessel:

(1) Will be designed, analyzed and fabricated to the high standards of quality required by Appendix G to 10 CFR Part 50, and by the A$ME Code and any pertinent Code Cases; (2) Will be made from materials of controlled and demonstrated high quality; (3) Will be subjected to extensive preservice inspection and testing to provide assurance that the vessel will not fail because of material or fabrication deficiencies; (4) Will be operated under conditions and procedures and with protective devices that provide assurance that the reactor vessel design conditions will not be exceeded during normal reactor operation or during design upsets in operation, and that the vessel will not fall under the conditions of any of the postulated accidents; (5) Will be subjected to periodic monitoring and inspection to demonstrate that the high initial quality of the reactor vessel has not deteriorated signifi-cantly under service conditions; and (6) May be annealed to restore the material toughness properties if this becomes necessary.

5.4 Component and Subsystem Design d

5.4.1 Reactor Coolant pumps lhe three reactor cor sant pumps will be sized to deliver flow at rates which equal l or exceed the requirvd flow rate under normal and transient operating core condi-tions. Sufficient pump rotat-lonal inertia will be provided by a flywheel, in j

conjunction with the impeller and motor assembly, to provide adequate flow following I

an assumed loss of pump power.

Each reactor coolant pump will be a vertical, single stage, centrifugal, shaft seal pump. The motor assembly for each pump will consist of a vertical solid shaft, squirrel cage induction type motor, an oil lubricated double Kingsbury type thrust bearing, two oil lubricated radial bearings, and a flywheel.

Criterion 4 of the General Design Criteria requires that structures, systems, and components of nuclear power plants important to safety be protected against the l

effects of missiles that might result from equipment failures. Because flywheels have large masses and rotate at speeds of approximately 1200 revolutions per minute during normal reactor operation, a loss of integrity of the pump flywheel could l

4 5-13 I

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result in high energy missiles and excessive vibration of the reactor coolant pump assembly. The safety consequences could be significant because of possible damage to the reactor coolant system, the containment, or the engineered safety features.

The probability of a loss of integrity of the pump flywheel can be minimized by the ]

use of suitable material, adequate design, preservice testing and inservice inspec- !

tion. The applicant states that the flywheels will be (1) fabricated from SA-533, l Grade B, Class 1 steel, (2) produced by a ,srocess that will minimize flaws and improve fracture toughness, and (3) cut, machined, finished, and inspected la accordance with Section III of the ASME .:cde and Regulatory Guide 1.14, " Reactor Coolant Pump Flywheel Integrity," Revision 1. The flywheel material will be drop weight tested in accordance with ASTM Standard E-208-69, " Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels," and Charpy V-notch impact tested in accordance with ASTM Standard E-23-73, " Notched Bar Impact Testing of Metallic Materials," to demonstrate that the fracture toughness is in confortnance with Regulatory Guide 1.14. The inservice inspection program will be performed in accordance with the recommendations of Regulatory Guide 1.14 and the applicable Edition of Section XI of the ASME Code in accordance with the rules of 10 CFR Part 50, paragraph 50.55a (the actual Code Edition and Addenda used will be addressed in the Final Safety Analysis Report).

The integrity of the reactor coolant pump flywhe(1 will be prcvided by designing it to 125 percent of the normal synchronous speed of the motor (approximately 1500 revolutions per minute). The lowest cesign operating temperature is specified to be 110 degrees Fahrenheit. The applicant has stated that the reference nil ductil-ity temperature will be no higher than 10 degrees Fahrenheit. Thus, the normal operating temperature of the pump flywheel will be at least 100 degrees Fahrenheit above the reference nil ductility temperature, which satisfies the acceptance

)

criteria for fracture toughness of Regulatory Guide 1.14 and, therefore, is i acceptable.

Based on our evaluation, we conclude tnat the applicant is in conformance with '

Regulatory Guide 1.14. Conformance with the recommendations of Regulatory Guide 1.14 constitutes an acceptable basis for satisfying the requirements of the applicable portions of Criterion 4 of the General Design Criteria.

The potential for the pump flywheel to become a missile, in the event of a rupture in the pump suction or discharge sections of the reactor coolant system piping, is under generic study by the staff. To demonstrate that the integrity of the fly-wheel will be maintained during a postulated loss-of-coolant accident, the appli-cant has referenced the analysis performed by Westinghouse in Topical Report WCAP-8163, " Reactor Coolant Pump Integrity in a LOCA," which we are currently reviewing. In conjunction with our review, the Electrical Power Research Institute has performed scaled pump tests with single- and two phase flow intended to veri fy vendor analytical techniques and predictions on reactor coolant pump overspeed.

5-14

The Electric Power Research Institute has contracted Combustion Engineering, CREARE, and the Massachusetts Institute of Technology to perform experimental and analytical work on two phase flow reactor coolant pump performance. The pump tests were performed on a 1/5 scale test loop at Combustion Engineering, and a 1/20 scale test loop at CREARE. We are reviewing the test results. The Electric Power Research Institute plans to complete the program in late 1978. In addition, Westinghouse, together with Framatome, and the French Atomic Energy Commission, are conducting a research program on pump testing and modeling using a 1/3 scale model of a Westing-house reactor coolant pump. The test and analysis efforts are expected to be completed by late 1978.

We are following the development and performance of these programs. If the results i of the generic investigation indicate that additional protective measures are warranted to prevent excessive pump overspeed or to limit potential consequences to safety related equipment, we will determine what modifications, if any, are neces- l sary to assure that an acceptable level of safety is maintained.

5.4.2 Steam Generators l

l Each of the three steam generators will be a Westinghouse Model F (a newer model steam generator), vertical shell and U-tube evaporator with integral moisture separating equipment. On the primary side, the reactor coolant will flow through the inverted U-t Wes, entering and leaving through nozzles to be located in the hemispherical bottom head of the steam generator. Steam will be generated on the shell side, will flow upward and exit through the outlet noZ21e at the top of the steam generator. Feedwater for steam generation will enter the steam generator through a nottle at an elevation above the top of the U-tubes. ,

1 The tube and tubesheet boundary will be designed for the reactor coolant side design pressure and temperature to minimize the transfer of activity, generated within the core, to the secondary side. The steam generators will provide a heat sink for the reactor coolant system and will be located at a higher elevation than

' the reactor core to improve natural circulation for decay heat removal.

The materials to be used in Class 1 and Class 2 components of the steam generators will be selected and the components will be fabricated in accordance with codes, standards, and specifications acceptable to the staff. The steam generator pres-l sure retaining parts will be designed and fabricated to meet the requirements of Section I!! of the ASME Code. The pressure boundary materials will comply with the fracture toughness requirements of Article NB-2300 of Section III of the ASME Code.

The primary side of the steam generator will be designed to ASME Class 1, in con-1 formance with our requirements, The secondary side pressure boundary parts of the a

steam generator will be designed, manufactured and tested to ASME Class I criteria which exceed our requirements (ASME Class 2) for the secondary side.

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i a

1 i fhe design and fabrication of the steam generator will also include access for the 4

inservice inspection considerations of Section XI of the ASME Code. Welds and weld l

4 joints required to be inspected during service will be designed to permit satis-j factory inspection, as well as' access to the joint for inspection. In addition, the steam generator will be designed to provide access for tube inspections in l j accordance with the recommendations of Regulatory Guide 1.83, " Inservice Inspection j of Pressurized Water Reactor Steam Generator Tubes."

j The onsite cleaning and cleanliness controls to be used during fabrication of the ;

j steam generators will conform to the recommendations of Regulatory Guide 1.37, )

j " Quality Assurance Requirements for Cleaning of Fluid Systems and Associated Compo-j nents of Water-Cooled Nuclear Power Plants." Interface requirements on the con- l j trols to be placed on secondary conlant chemistry are in agreement with established j staff technical positions. Conformance with applicable codes, standards, staff positions, and regulatory guides constitutes an acceptable basis for meeting the applicable requirements of Criteria la,15, and 31 of the General Design Criteria.

l Recent operating experience with some pressurized water reactors, including Westinghouse plants, has revealed a phenomenon associated with steam generator tube 7

deformation in the form of a reduction in tube diameter (i.e., the phenomenon known 3 as tube denting). Tube denting 4 term which describes a group of related j phenomena resulting from corr > the carbon steel in the crevices formed

{ between the tubes and the 1 L plates or tubesheet in a steam generator.

l l Denting was first discovered in g ril 1975 during an inspection of a steam j generator for Unit No. 2 of the Surry plant.

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Since then, Westinghouse has conducted a comprehensive research program in order to determine the cause and extent of the problem and to establish corrective action to be implemented in all new Westinghouse plants. As a result of the findings of this program, the applicant will incorporate the following steam generator design and fabrication modifications: (1) the crevice between the tubesheet hole and insertad

tube will be eliminated by full depth expansion of the tube - the tube expansion j

] and subsequent positive contact pressure between the tube and tubesheet will pre-clude a buildup of impurities f rom f orming in the crevice region and, therefore, reduce the probability of crevice boiling; (2) the inconel 600 tube material will j be thermally treated for resistance to stress corrosion - the treatment will pro-

{ vide an improved metallurgical structure, associated with grain boundary precipi-tate morpholc s which provides increased margin with respect to stress corrosion performance; (3) the tube support plates will be manufactured from ferritic stain-less steel material which has been si;own in laboratory tests to be resistant to corrosion in the operating environment; and (4) the tube supp wt plates will also be designed with broached tube holes rather than drilled holes - the broached tube hole design promotes high velocity flow along the tube, sweeping impurities away from the supnort plate locations.

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As a result of our review, we conclude that the above mentioned steam generator l design and fabrications modifications will provide appropriate features for minimi2ing the potential for the phenomenon of tube denting and, therefore, are acceptable. In addition, operating experience, verified in numerous steam gene-rator inspections, indicates that the tube degradation, such as thinning associated with phosphate water treatment, is not occurring where only all volatile treatment l has been utilized. The incorporation of the described steam generator modifica-tions, close adherence to all volatile treatment chemical specification for second-ary water and rigorous secondary water monitoring, together with the limiting of condenser in-leakage, provide reasonable assurance of satisfactory performance of

l. steam generator tubing.

1 Additional assurance will be provided by the high quality of the condenser system design. Corrosion of the condenser will be minimized through the use of titanium tubes, aluminum bronze tube sheets, and cupper nickel clad carbon steel water boxes. In addition, the condenser will t.e designed with provisions for cathodic protection by means of sacrificial anodes or impressed current. The cooling tower water treatment system will inhibit corrosion by preventing scaling and fouling.

The condenser will also be designed with provisions for mechanical cleaning equipment.

The applicant has stated that the bases for plugging degraded steam generator tubes will be discussed in the Final $afety Analysis Report. Regulatory Guide 1.121,

" Bases for Plugging Degraded PWR 5 team Generator Tubes," describes a method for providing acceptable bases. Since the criteria for plugging degraded tubes may have a bearing on the design of a tteam generator, we require that the applicant either commit to the recommendations of Regulatory Guide 1.21 or provide its proposed criteria and bases for plugging degraded tubes for our review. Therefore, this matter remains outstanding.

Criteria 1 and 32, of the General Design Criteria require that components which are part of the reactor coolant pressure boundary or other components important to safety be designed to permit periodic inspection and testing of critical areas for structural and leaktight integrity.

The components in the steam generator are classified as ASME Code Class 1 and 2, depending on their location in either the primary or secondary coolant systems, respectively. The steam generator will be designed to permit inservice inspection of the Class I and 2 components, including individual tubes. The design aspects that provide access for inspection and the proposed inspection program will follow j the recommendations of Regulatory Guice 1.83 and will comply with the requirements of Section XI of the ASME Code (Cooe Editioa and Addenda will be determined in accordance with the rules of 10 CFR Part 50, Section 50.55a(g) and will be I addressed in the final Safety Analysis Report).

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Conformance with Regulatory Guide 1.83 and Section XI of the ASF Code constitutes an acceptable basis for meeting the applicable requirements of Criteria 1 and 32 of the General Design Criterie.

l 5.4.3 Residual Heat Removal,5ystem The residual heat removal system will be designed to remove core decay heat and provide long-term core cooling following the initial phase of reactor cooldown.

The scope of our review of the residual heat removal system included piping and

-instrumentation diagrams, plant arrangement drawings, and design performance speci-fications for essential components. The review also incuded the applicant's pro-posed design criteria and design bases for the residual heat removal system, the analysis of the adequacy of the criteria and bases, and the conformance of the design to these criteria and bases.

The residual heat removal system will consist of two parallel flow trains each consisting of a heat exchanger, a pump, and the associated piping, valves, and instrumentation necessary for Operational control. The system components will be designed to seismic Category I and Saf ety Class 11 requirements in conformance with the recommendations of Regulatory Guide 1.29, " Seismic Design Classification" The inlet lines to the residual heat removal system will be connected to the hot legs of two of the reactor coolant system loops and the return lines will be connected to the cold legs of all three reactor coolant system loops. The residual heat removal system suction lines will be isolated from the reactor coolant system by two motor-operated valves in series located inside the containment. Each discharge line will be isolated from the reactor coolant system by two check valves located inside the containment and by a normally open motor-operated valve located outside the containment. Thus, the residual heat removal system will incorporate two independent and redundant barriers whenever the reactor coolant system pressure will be above the residual heat removal system design pressure. We find these features to be acceptable.

However, as a result of our review, we have determined that the residual heat removal system design will not be capable of performing its shutdown cooling f unc-tions with only onsite or offsite electrical power available, if the most restric-tive single active component failure is assumed. Since one of the two isolation valves in the suction lines of each of the two trains will be powered from the same source, failure of this electrical source would prevent operation of both residual heat removal trains. In the event of this type of postulated failure, the appil-cant has proposed the use of operator action outside the control room to restore power to the affected residual heat removal isolation valve by connecting a temporary power supply. Based on a preliminary design of the temporary power supply arrangement, the applicant has estimated that it would take less than two hours for an operator to perform all the necessary actions. While limited operator 5-18 l

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action outside the control room after the single failure would be considered accept-able as stated in the revision to Section 5.4.7 of the Standard Review Plan (see discussion ir, the following paragraph), final acceptance of the applicant's proposed design concept cannot be made without the more detailed information normally avail-able at the operating license stage of review. At the operating license stage of

( review, we will assure that the residual beat removal system is desigr,ed to have ,

l the capability of performing its shutdown cooling function with only onsite or offsite power, ass. ming the most restrictive single active failure. ,

I We have recently Jublished a revision to Section 5.4.7 of the Standard Review Plan which requires tLat a plant have the capability for transferring heat from the j reactor to the environment from normal operating conditions to cold shutoown condi-I tions using on'y safety grade systems, m uming that only offsite or onsite power is available and also assuming the most limiting single failure. As a result, we ranoire tb the applicant address how the design of the Sundesert plant will conferm with each of the following requirements:

4 (1) The capability to take the plant to e cold shutdown condition by using only safety grade equipment, with only onsite power or only offsite power, assuming a single failure. Limited manual actions outside the control room can be considered to compensate for a single failure. The plant design must provide f or safety grade dump valves, operators, and power supplies so that manual action should not be required except to meet the single failure criterion.

4 (2) In support of item (1) above, the capability to achieve the following by using only -afety grade equipment with an assumed single f ailure:

(a) Cold shutdown condition in less than 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

(b) Depressurization of the reactor coolant system. Identify the manual ,

actions inside and outside containment that must be performed to achieve this. Discuss the capability of remaining at hot standby until the

' manual actions, or repairs if required, to achieve a cold shutdown are complete.

(c) Boration of the reactor coolant system to cold shutdown. Identify the manual actions inside and outside containment that need to be performed to initiate and maintain boration.

(3) Provisions for collection and containment of residual heat removal pressure relief valve discharges.

(4) A seismic Category I auxiliary feedwater suppij for at least four hours at hot Shutdown plus the time necessary to achieve cooldown to activate the residual heat removal system, based on the longest time (for only onsite and offsite power and assuming the worst single failure).

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l (5) In addition to the above items, discuss the tests that will be performed to demnnstrate the mixing of the added barated water and the cooldown under natural circulation conditions assuming a single failure (i.e. , a steam generator atmospheric dump valve fails to open). A procedure for cooling down by using natural circulation must be available to the operator.

Therefore, the above matter remains outstanding.

Overpressure protection for the residual heat removal system will be provided by the use of pressure relief valves in the system discharge lines and inlet lines.

The applicant states that the check valves, that will provide pressure boundary separation between the residual heat removal system and the reactor coolant system, will be designed for Zero leakage and that periodic testing will be conducted to verify ttGt the check valves do not leak. In the event that the check valves do leak, the pressure relief valves in the discharge lines of the residual heat removal system will be capable of relieving 20 gallons per minute of possible back leakage through the check valves, We find these design features to be acceptable.

The pressure relief valve in each inlet line to the residual heat removal system will be designed to relieve the combired flow of three charging pumps at the relief valve pressure seving of 450 pounds per square inch gauge for the purpose of preventing overpressurization of the residual heat removal system while the plant is in the startup or shutdown mode, in crder to evaluate this design basis, we have requested additional information, which has been submitted by the applicant, to justify the relief flow capacity of this pressure relief valve. We have not completed our review of the information submitted. Therefore, this matter remains outstanding.

Following shutdown and return to reactor power operation, the residual heat removal suction lines inside the containment wnuld normally have a section with water trapped between the two isolation valves. This condition could lead to a potential overpressire problem in the short section of pipe between these two valves subsequent to a post dated loss of-coolant accident or a steam line brejk inside containment.

The high temperatures inside the containment af ter the postulated accidert could heat up the water in this short section of pipe causing a pressure increase high enough to compromise valve and piping integrity.

The applicant has analyzed this potential problem by including the following assump-tions: (1) the trapped water would initially be sealed at a temperature of 300 degrees Fahrenheit and.a pressure of 400 pounds per square Inch; (2) due to natural convection currents, the volume of water octween the isolation valves would cool to containment ambient temperature; and (3) a postulated loss-of coolant accident or a steam line break occurs at'this time with resulting saturation temperatures of 270 degrees Fahrenheit and 264 degrees Fahrenheit, respectively. The applicant, there-fore, concluded that the maximum pressure reached in the pipe section, which will 5-20

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De designed for 2500 pounds per square inch, would be less than 400 pounds per square inch. We note, howeser, that no consideration was given to expected valve seat leakage across the isolation valve nearest the reactor coolant system. Addi-tional analysis and information will he required to show that the result of a loss of residual heat removal valve or piping integrity is of no safety consequence, or that valve leakage is not a problem. However, resolution of this issue can be conducted during the operatinr' license stage of review since the design modifica-tions (e.g., addition of a t- 'ulve to relieve potential overpressure), it required, should not be signif. it and can be accomplished at that time.

The applicant also evaluated the pntential for exceeding the allowable cooldown rate of the residua'l heat removal system and the reactor coolant system during the shutdown cooling mode assuming a loss of the nonsafety grade instrument air system which will control the outlet and bypass valves of the residual heat removal heat exchanger. The applicant has stated that the positioning of the heat exchanger outlet and bypass valves to the fully open and fully closed position, respectively, would have no ef fect if this f ailure were to occur af ter the first two hours of residual heat removal operation since this is the normal positioning for the valves for the remainder of the cooldown. However, if the failure were to occur during the initial two hour period of residual heat removal operation, operator action would be required to prevent exceeding the technical specification limit on cool

down rate. We requested additional information to demonstrate that the piping integrity of the residual heat removal system can be maintained for the resulting stresses caused by the excessive cooldown rate duria9 the first two hours.

in response to our request, the applicant provided additional information on this matter. We have not completed our review of the additional information. Therefore, this matter remains outstanding.

The applicant has committed to meet Regulatory Guide 1.68, " Initial Test Programs for Water-Cooled Reactor Power Plants", Revision 1, relating to the preoperational l and initial startup test programs. We find this commitment to be acceptable.

i Based on our review of the residual heat removal system, and subject to the satis-factory resolution of the matters discussed above concerning (1) the relief flow capacity of the pressure relief valve on the inlet line, (2) loss of instrument air j

during the initial two hours of residual heat removal operation, and (3) the require-l ments to achieve a cold shutdown using only safety grade systems, we conclude that the prcposed design of the residual heat removal system conforms to all applicable regulations, regulatory guides and staff positions and is acceptable.

5. 5 Loose Parts Monitoring System The applicant has committed to install a loose parts monitoring system. We find this commitment to be acceptable.

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4 Recently, prototype loose parts monitoring systems have been developed and are presently in operation or being installed at a number of plants. As a result of a

! study we conducted on the installation of, and experience with, loose parts monitoring

{ systems in operating plants, we have identified the following aspects for a loose 4

parts monitoring system which we will use to assess the acceptability of the specific l system to be provided for Sundesert when we review the detailed information to be submitted in the final Safety Analysis Report:

(1) The description of the loose parts monitoring system shall include the loca-e tion of all sensors and the method for monitoring them. A minimum of two i sensors will be required at each natural collection region. For example, in a pressurized water reactor, two sensors should be included at the top and at the bottom of the reactor vessel and at each steam generator primary coolant inlet.

4 j (2) The description of the monitoring equipment shall include the levels and the 7 basis for the alarm settings. In addition, the manufacturer's sensitivity j specifications for the equipment shall be provided. Anticipated major sources i

of internal and external noise shall be identified along with the plans for f minimizing the effects of these sources on the ability of the monitoring f equipment to perform its intended function.

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(3) The loo e parts monitoring system will be required to function af ter any i,

i seismic event for which plant shutdown is not required. A description of the j 4

precautions to be taken to assure the operability of the system after an j operating basis earthquake shall be provided.

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l (4) The loose parts monitoring system shall be shown to be adequate for the normp'.

) operating environment, including temperature, humidity, radiation and vib a-tion, by either analysis, or test, or combined analysis and test.

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(5) The loose parts monitoring system must be operational and capable of recording !

vibration signals for signature analysis at the time of initial startup test-

) ing. A detailed discussion shall be provided of the operator ttaining program, planned operating procedures, and record keeping procedures for the operation

of the system.

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6. 0. ENGINEERED SAF ETY FEATURES <

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6.1 Design Considerations l

Systems and design features which will be provided to function during or following a postulated accident at the facility to prevent and reduce the release of fission l products are called engineered safety features. These engineered safety feaures will be designed to retain, within acceptable leakage limits, the fission products which might be released from the reactor fuel, to mitigate the damage to the fuel cladding and other fission product barriers, to provide for the protection of the public and station personnel, and to provide for fission product removal and cleanup within the plant structures. This section describes our review of the containment systems, the emergency core cooling system, the control room habitability system, and the atmospheric cleanup systems.

Certain of these systems, in addition to serving as engineered safety features, will have functions for normal plant operation. Systems and components desig-nated as engineered safety features will be designed to be capable of assuring safe shutdown of the reactor under the adverse conditions of the various pos-tulated design basis accidents described in Section 15.0 of this report. They will be designed, therefore, to seismic Category I standards and must function even with a complete loss of offsite power. Components and systems will be j

provided with sufficient redundancy so that a single failure of any active com-ponent or system will not result in the loss of the capability to perform the safety function. The instrumentation and control systems and emergency power systems will be designed to the same seismic and redundancy requirements as the systems they serve.

6.2 Containment Systems The containment systems for each unit of the Sundesert plant will include the containment structure, containment heat removal system, containment isolation system, containment combustible gas control system, and provisions for contain-

! ment leakage rate testing. The hydrogen recombiners in the combustible gas control system will be shared by each unit.

6.2.1 Containment functional Design The containment structure will be a cylindrical, carbon steel lined, reinforced concrete structure with a net free volume of 2,268,000 cubic feet. The containment structure will house the nuclear steam supply system, which includes 6-1

l the reactor vessel, reactor coolant piping, reactor coolant pumps, pressurizer, and steam generators, as well as certain components of the plant's engineered safety feature systems. The containment structure will be designed to withstand internal pressurization resulting from postulated high energy pipe breaks inside containment and external pressurization due ta inadvertent actuation of the containment heat removal systems. The containment structure will be designed for an internal pressure of 48 pounds per square inch gauge and a temperature of 280 degrees Fahrenheit. The containment structure will be designed for an external differential pressure of two pounds per square inch. Our evaluation of the acceptability of the proposed containment design is presented in the following discussions.

Containment Analysis The applicant has analyzed the containment pressure responses for postulated accidents in the following manner.

The mass and energy release rates were us in conjunction with the Stone &

Webster LOCTIC computer code to perform the containment pressure response analysis.

The data for mass and energy release to the containment following a primary system rupture were calculated by using the following Westinghouse codes: SATAN-V for the blowdown period; WREFLOOD fo- the reflooding period; and FROTH to calculate post-reflood steam boil-of f f rom the core and steam generators. These analytical methods conservatively maximize steam flow to the containment so as to maximize containment pressure. The methods and assumptions used to calculate the mass and energy release rates are documented in Westinghouse Topical Report WCAP-8264,

" Westinghouse Mass and Energy Release Data for Containment Design " which we have reviewed and found acceptable as documented by letter to Westinghouse dated March 12, 1975.

The applicant has analyzed a number of postulated reactor coolant system pipe breaks, along with consideration of various single f ailures, in order to identify the containment design basis loss-of-coolant accident. The containment design basis loss-of-coolant accident was identified as the postulated double-ended rupture at the pump suction of the reactor coolant system, which resulted in a peak calculated pressure of 39.8 r.ounds per square inch gauge. The loss of one of the two containmert spray trains and full emergency core cooling system operation were conservatively assumed for the evaluation.

We performed a confirmatory analysis of the containment pressure response to a postulated double ended rupture at the pump suction of the reactor coolant system using the CONTEMPT-LT H00 26 computer code. Our analysis was based on the mass and energy release, containment structural heat sink, and containment heat removal systems performance data provided by the applicant. Conservative condensing heat transfer coefficients to the structures inside the containment were used. Our 6-2

analysis of the pressure response resulted in a peak calculated pressure of 39.5 pounds per square inch gauge. Therefore, the design pressure of 48 pounds per square inch gauge provides a 20 percent margin above the peak calculated pres-sure, which exceeds the 10 percent margin we require for a construction permit application.

The applicant has analyzed a spectrum of postulated main steam line break acci-dents to determine the containment pressure and temperature transients. Mass and energy release data for the spectrum of steam line breaks considered were cal-culated using the MARVEL code described in Westinghouse Topical Report WCAP-8843,

" MARVEL, A Digital Computer Code for Transient Analysis of a Multiloop PWR System."

The MARVEL code simulates the primary and secondary systems of a pressurized water reactor, including the power excursion which may occur in the core follow-ing a postulated main steam line break. For a postulated main steam line break, the code calculates the heat flow from the core and intact steam generators into the primary system, and the heat flow from the primary system into the steam generator with the broken line. The primary system heat flow produces additional steam which is added to the containment. No liquid entrainment is assumed in the flow from the break so that only steam flows from the break. This assumption permits the secondary liquid to remain in the steam generator until it is boiled by heat flow from the primary system and, therefore, maximizes the energy release.

6 Using this assumption, approximately 100 x 10 British thermal units would be added to the containment for a double ended break at full power, which is the 6

worst case. This amount of heat energy is about 35 x 10 British thermal units more than would be added to the containment if entrainment were considered and provides a margin of conservatism in the analysis. The analysis includes addi-tional steam from the intact steam generators, before closure of the isolation valves, and the unisolable steam in the steam lines and turbine plant piping.

Feedwater flow is added to the affected steam generator, based on runout flow,

, before isolation. Hence, no credit is taken for any feedwater flow reduction during the valve closure period. The unisolable feedwater mass is added to the initial steam generator inventory, We have not completed our review of the MARVEL code. Based on our review to date, however, we believe that the mass and energy release data obtained for Sundesert from this code are conservative for containment analysis. If upon the completion of our review, we determine that these data are not conservative, we will require the applicant to perform new analyses based on the conclusions resulting from our review.

We have not completed our confirmat.ory analysis of the containment response to the postulated main steam line break. Therefore, this matter remains outstanding.

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f The applicant has committed to qualify the safety-related equipment to be located

] inside containment to the most severe environmental conditions based on the

{ results of acceptable calculations performed for the Sundesert plant. We find j this commitment to be acceptable.

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9 The applicant has calculated the maximum external differential pressure on the j containment structure due to inadvertent operation of the containment heat removal

! systems. For this analysis, the applicant has assumed the containment is cooled f to a temperature of 55 degrees Fahrenheit resulting in an external differential i pressure of 1.8 pounds per square inch. The containment was initially assumed to be at the maximum temperature of 105 degrees Fahrenheit with a dewpoint of 95 i degrees Fahrenheit which corresponds to the limiting Technical Specification value proposed by the applicant for containment atmosphere relative humidity.

l The calculated value is below the proposed design value of two pounds per square

inch differential. Based on our review of the applicant's calculation, we con-clude that the design external differential pressure for the containment is j acceptable.

f j Containment Subcompartment Analysis I I The applicant has analyzed the pressure response of subcompartments inside the containment to the postulated high energy line breaks identified in l Table 6.2.1-16 of the Preliminary Safety Analysis Report.

t i The blowdown rates from postulated primary system ruptures within containment

, subcompartments were calculated using the SATAN 4 code. This code uses the j

modified Zaloudek correlation to calculate flow when the break fluid is subcooled and the Moody slip flow model to calculate flow when the break fluid is saturated.

Stagnation conditions at the break are approximated by removing the momentum flux option from the SATAN-V code. This method is also documented in Topical Report WCAP-8264 which, as stated prevfously, has been found acceptable. .

j The flow calculation for a postulated broken feedwater line in the steam genera-j tor cavity was performed assuming constant pressure at both ends of the break.

The Henry-Fauske subcooled critical flow correlation was used for the feedwater pipe side of the break, and the Moody slip flow crrelation was used for the saturated flow from the steam generator vessel side. This method is conservative since the pressure would be reduced below the constant value assumed by the j applicant soon after the rupture which would reduce the flow rate to a value less j than that calculated by the applicant.

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! The mass and energy release data obtained from the above analysis were then used I

with the Stone & Webster THREED code and the RELAP-4 code to perform the subcom-partment analysis. The THREED code accounts for the flow of a steam-water-air S

mixture through the vents but does not consider the inertial effects which may be 6-4 3

i significant when the vent flow is subsonic. The RELAP-4 code used by the appli-cant considers inertial effects but does not include the effects of air mixing with the flowing steam-water mixture. The results from the code predicting the higher differential pressures are repurted in the Preliminary Safety Analysis Report.

A comparison of the applicant's results with the iesults of our confirmatory analysis are provided in Table 6.1. As can be seen from Table 6.1, the appli-cant's calculated differential pressures for the steam generator and pressurizer subcompartment analysis are slightly higher than our calculated values. The applicant has also added a 40 percent margin to the peak calculated differential pressures to be used in the structural analysis. We, therefore, find the steam generator and pressurizer compartment analyses acceptable for use in the struc-tural design of those subcompartments.

For the reactor cavity subcompartment analysis, however, we calculated higher dif ferential pressures across the reactor cavity walls for the majority of the subcompartment nodal volumes. As a result, we require that the applicant provide additional information to justify the proposed design of the reactor cavity subcompartment. Therefore, this matter remains outstanding.

We have also reviewed the applicant's subcompartment analysis for use in the major component supports design evaluations. In this regard, the applicant has not provided (1) sufficient justification for the nodalization of the subcompart-ments and (2) the transient loading on the major components, such as the reactor vessel and steam generators, for use in the design of component supports. We, therefore, cannot conclude that the results of the applicant's subcompartment analysis are acceptable for use in the design of the component supports. As a result, this matter remains outstanding.

Minimum Containment Pressure Evaluation for Emergency Core Cooling System Analysis Appendix K to 10 CFR Part 50 of the Commission's regulations requires that the effect of the operation of all installed pressure reducing systems and processes be included in the emergency cora cooling system evaluation. For this evaluation, it is conservative to minimize the containment pressure since this will increase the resistance to steam flow in the reactor coolant loops and reduce the reflood rate in the core.

Following a postulated loss-of-coolant accident, the pressure in the containment building will be increased by the addition of steam and water from the primary reactor system to the containment atmosphere. After initial blowdown, heat transfer from the core, primary metal structures, and steam generators to the emergency core cooling system water, will produce additional steam. This steam, together with any emergency core cooling system water spilled from the primary 6-5

TABLE 6.1 SUBCOMPARTMENT PRESSURE DIFFERENTIALS Peak Calculated Pressure (pounds per square inch Postulated Pipe location Break Description differential) Nodes Applicant From Reactor Cavity 5taff h

150 square inch Limited 67.9 Displacement Rupture in 68.3 2 13

, 1.8 5.2 4 13 Pump Discharge 14.8 16.4 6 13 95.0 107.1 7 13 148.8 132.8 8 13 11.3 18.0 9 13 1.9 7. 4 10 13 Steam Generator Single Ended Rupture

> Compartment Below 35.2 27.7 3 39

& in Pump Discharge 26.5 24.2 Operating Floor 7 39 25.9 24.1 12 M 24.7 23.9 16 39 23.7 23.7 18 39 l 23.1 22.4 26 39 Pressurizer Double Ended Rupture in Cubicle 18.9 17.8 2 Spray Line 18.8 5

17.9 3 5 18.9 17.9 4 5

system, will flow through the postulated break and into the containment. This i energy will be released to the containment during the blowdown phase and later during the emergency core cooling system operational phases; i.e., reflood and post-reflood.

Energy removal from the containment atmosphere will occur by several means. Steam condensation on the containment walls and internal structures will serve as a passive heat sink that will become effective early in the blowdown phase. Subse-quently, the operation of the containment sprays and fan coolers will remove steam from the containment atmosphere. When the steam removal rate exceeds the rate of steam addition from the primary system, the containment pressure will decrease from its maximum value.

The emergency core cooling system containment backpressure calculations were performed with the Westinghouse emergency core cooling system evaluation model described in Topical Report WCAP-8327, " Containment Pressure Analysis Code (C000)."

We have previously reviewed this model and concluded that it was acceptable for the evaluation of the containment backpressure, as documented by letter to Westinghouse, dated May 30, 1975, subject to the review of the plant-dependent input parameters used in the analysis. We have reviewed the Sundesert plant parameters used for the analysis of the containment pressure for the emergency cooling system evaluation and find them to be suitably conservative. We, therefore, conclude that the containment pressure analysis for the emergency core cooling system evaluation is acceptable and meets the requirements of Appendix K to 10 CFR Part 50.

Summary and Conclusions We have evaluated the containment system functional design for conformance with the General Design Criteria and, in particular, Criteria 16 and 50. Based on our review, and subject to the satisf actory resolution of the matters concerning (1) the containment response to postulated main steam line breaks, for which we have not completed our review, (2) the design of the reactor cavity subcompartment, for which we require additional information, and (3) the acceptability of the containment subcompartment analysis for use in the design of major component supports, for which we require additional information, we conclude that the proposed containment functional design meets the requirements of the applicable General Design Criteria and is acceptable.

6.2.2 Containment Heat Removal Systems The containment heat removal systems for the Sundesert plant will consist of the containment spray system. the containment atmosphere recirculation system, and the low head safety injection / residual heat removal system.

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i lhe function of the containment heat removal systems is to return the containment

pressure to a low value following a postulated break in either the primary or j secondary system piping inside the containment. To accomplish this function, heat will be transferred from the containment atmosphere to the spray water and

to the reactor plant component cooling water by the containment spray system and the containment atmosphere recirculation system, respectively. In addition, heat will be transferred from the water in the containment sump region to the reactor j_ plant component cooling water, during the recirculation mode of safety injection, I

l via the residual heat removal system heat exchangers.

{ The containment spray system serves as an engineered safety feature and will not 1

be used for normal plant operation. The containment spray system will consist of l l

) two redundant trains. The system will be safety grade (Quality Group B and

. seismic Category I), and all active components will be located outside of the

containment structure. The applicant has also committed to design the contain-9 ment sump to meet the guidelines of Regulatory Guide 1.82, " Sumps for Emergency j Core Cooling and Containment Spray Systems."

f The operation of the containment spray system will be automatically initiated by f the containment depressurization actuation signal, which will be initiated by j high containment pressure. The containment depressurization actuation signal j will start the spray pumps and open the spray isolation valves. The spray water

{ will be discharged into the containment upper region through spray nozzles arranged i

on headers. The containment spray pumps will initially take suction from the refueling water storage tank. When a predetermined level is reached in the refueling water storage tank, a low level signal will automatically transfer the

l. pump suction to the containment sump and close the suction valves from the refuel-l ing water storage tank.

1 q The applicant's evaluation for the available net positive suction head for the

) containment spray pumps is consistent with the guidelines of Regulatory l Guide 1.1, " Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal System Pumps," and is acceptable. The results of this evaluation i show that the available net positive suction head for the containment spray pumps t

] in the injection and recirculation modes are 50 feet and 3.7 feet, respectively.

t j The required net positive suction head for the spray pump is 1.5 feet. Therefore,

! suf ficient net positive head will be available for both the injection and recir-culation modes of operation.

4 1

A' The safety related portion of the containment atmosphere recirculation system j will consist of one cooling coil assembly and one fan on each electrical train.

The containment atmospheric recirculation system serves as an engineered safety q feature system and will be designed as seismic Category 1. During normal plant operation, the containment atmospheric recirculation system will be cooled with nonnuclear safety chilled component cooling water, and normal fans will be used

, to move air over the cooling (. oils. Upon receipt of a safety injection signal, a

6-8 I

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e

(1) the chilled component cooling water will be isolated from the cooling coils and reactor plant cooling water flow will be initiated to the coils and (2) the normal f ans will be stopped and the engineered safety feature fans will be started.

The residual heat removal system heat exchangers will utilize reactor plant cooling water on the shell side to remove heat from the containment during the recircula-tion phase following a postulated accident. During the injection phase, the tube side flow will consist of water from the refueling water storage tank. During the recirculation phase of emergency core cooling, the hot water from the sumps will be circulated through these heat exchangers, thus providing heat removal from the containment.

Based on our review of the containment heat removal systems, we conclude that the proposed system design is in conformance with the requirements of Criteria 38, 39, and 40 of the General Design Criteria and the recommendations of Regulatory Guides 1.1 and 1.82, and, therefore, is acceptable.

6.2.3 Containment Isolation System lhe containment isolation system witi be designed to automatically isolate the containment atmosphere from the outside environment in the event of an accident.

Double barrier protection, in the form of closed systems and isolation valves, will be provided to assure that no single active failure will result in the loss of containment integrity. The containment isolation provisions will be safety grade (ASME Code,Section III, Class 2 and seismic Category I) and will be missile protected.

Diverse parameters will be sensed to assure that containment isolation will n .;u r under all postulated accident conditions.

Based on our review, we have determined that the cottainment isolation provisions f or the lines penetrating containment conform to the equirements of Criteria 54, 55, 56 or 57 of the General Design Criteria, as appropriate. There are certain containment penetrations whose isolation provisions do not satisfy the explicit requirements of Criteria 55 and 56, but still conform with these criteria since the isolation provisions are acceptable on "some other defined basis" as permitted by Criteria 55 and 56. These penetrations are discussed below:

(1) Residual Heat Removal System Suction Line from the Reactor Coolant System The explicit requirement of Criterion 55 is that each line that is part of the reactor coolant boundary and penetrates primary reactor containment have two containment isolation valves, one inside containment and one outside contain-ment. The containment isolation valves must be either locked closed or capable of automatic isolation.

The applicant's proposed isolation provisions for the residual heat removal system suction lines will consist of two normally closed, motor operated valves in series inside containment. The valves will also be interlocked to 6-9

i ..~

1 j prevent them from being inadvertently opened. Since the residual heat i removal system suction lines have no post-accident safety function, they

{ will remain isolated following a loss of-coolant accident. The lines will

' also connect to the closed, emergency cere cooling system outside contain- j 1

'i

ment. In view of the above system design considerations, we conclude that l the normally closed, system isolation valve closest to and inside the containment and the closed engineered safety feature grade system outside containment (in lieu of an isolation valve located outside containment) l

] provide an acceptable alternate to the explicit isolation requirements of l 1 Criterion 55, as permitted by the "other defined basis" of that criterion, ;

j and that the proposed design, therefore, complies with Criterion 55.

l l !

(2) Engineered Safety Features Sump Suction Lines (Residual Heat Removal System pump Suction Line and Containmeht Spray Pump Suction Line) I I

The explicit requirement of Criterion 56 is that each line that connects l

directly to the containment atmosphere and penetrates primary reactor con-tainment have two containment isolation valves, one inside containment and l one outside containment. The containment isolation valves must either be I locked closed or capable of automatic isolation.

The containment sump suction lines are part of the emergency core cooling system and the containment heat removal systems, and must be opened f ollow- l ing a postulated loss-of-coolant accident to satisfy their post-accident l functional requirements, which is to permit long term cooling of the reactor core and the containment atmosphere. As a result, automatic isolation of these lines is not desirable and remote manual isolatinn capability will be provided.

The applicant's prcposed isolation provisions for the containment sump suction lines will include a single isolation valve in each line since system reliability is greater with only one isolation valve. In' addition, the valves will be located outside containment since they would be submerged following a loss-of-coolant accident if located inside containment. The i

closed engineered safety feature grade system outside containment serves as {

the second containment isolation barrier (in lieu of an isolation valve l located inside containment). In view of these system design considerations, l we conclude that the isolation provisions for these lines provide an accept- )

able alternate to the explicit isolation requirements of Criterion 56, as '

permitted by the "other defined basis" of that criterion and that the pro-posed design, therefore, complies with Criterion 56.

(3) Emergency Core Cooling System Saf ty Injection Lines The applicant's proposed design to meet the containment isolation provisions of Criterion 55 for certain emergency core cooling system safety injection lines will consist of a check valve inside containment and a remote manual valve outside containment. A remote manual isolation valve will be provided 6-10

(in lieu of an automatic isolation valve) because the lines, which are part of the emergency core cooling system, have a post-accident safety function and should not close on a containment isolation signal. We conclude that l

the isolation provisions for these lines provide an acceptacle alternative i

to the explicit requirements of Criterion 55 of the Generai Design Criteria regarding the actuation provisions for the valves outside containment, as permitted by the "other defined basis" provisions of that criterion, and that the proposed design, therefore, complies with Criterion 55.

(4) Reactor Coolant Pump Seal Injection Lines The applicant's proposed design to meet the containment isolation provisions of Criterion 55 for the reactor coolant pi ip seal injh tion lines will consist of a check valve inside containment v.a a remote manual isolation valve outside containment. Spurious valva closure of the seal injection l

)

lines could cause reactor coolant pump seal damage and the subsequent loss i of reactor coolant. Therefore, the remote manual isolation valves will not receive a signal to close from the engineered safety features actuation signal so that they will not be subject to spurious valve closure.

The reactor coolant pump seal injection system, which is not an engine.ered safety feature system, will be connected to and will be of the same quality (i.e., Quality Group B or better and seismic Category 1) as the emergency 4 core cooling system. The reactor coolant pump seal injection system, which f will he a closed system outside containment, will receive flow from the high 1 head safety injection pumps. We conclude that the isolation provisions for each of these lines, consisting of remote manual actuation of the isolation valve outside containment and a closed system outside containment, provide an acceptable alternate to the explicit isolation requirements of Criterion 55, as permitted by the "other defined basis" of that criterion, and that the proposed design, therefore, complies.with Criterion 55.

Our review of the containment isolation system has also included a review of the containment purge system, which will be used to reduce airborne radioactivity in the containment and allow personnel entry. The applicant has stated that the containment purge system will nnt be used during the modes of reactor operation when containment integrity is required, but only during the cold Ghutdown and refueling modes of operation. We find this commitment to be acceptable.

Based on our review, we conclude that the containment isolation system design f conforms to Criteria 54, 55, 56 and 57 of the General Design Criteria and meets the acceptance criteria of Section 6.2.4 of the Standard Review Plan and, l therefore, is acceptable.

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l j 6.2.4 Combustible Gas Control System i

a j Following a postulated loss of-coolant accident, hydrogen may accumulate inside

] the containment as a result of (1) chemical reaction between the fuel rod clad-

} ding and the steam resulting from evaporization of emergency core coo'ing water, j (2) corrosion of construction materials .by the spray solution, and G ) radiolytic l decomposition of the cooling water in the reactor core and the containment sumps.

i l

j in order to mitigate the consequences of hydrogen accumulation in the containment, j a redundant combustible gas control system and a backup purge system will be i provided. The combustible gas control system will consist of two 100 percent j capacity hydrogen recombiners located outside of containment. The hydrogen j recombiner system will be shared by the two units. Both of the redundant 100 s

percent capacity recombiners will be portable and will be accessible to either ,

unit. Each recombiner will have a minimum capacity of 50 standard cubic feet per I 4

minute. The system will be protected from the effects of tornadoes, external missiles, pipe whip and jet impingement, and will be designed as seismic I ,

Category I. A hydrogen analyzer, which will also be designed as seismic l l Category 1, will be installed and connected to the piping downstream from the containment isolation valves in each recombiner suction line.

4 I

j The applicant has specified, in the Preliminary Safety Analysis Report, the design

criteria for the hydroger recombiners, whi: h we find acceptable, In addition, f the applicant has committed to use either hydrogen recombiners that we have

previously found acceptable, or provide in the Final Safety Analysis Report a ,

l complete description of the hydrogen recombiners selected along with the test )

l results which demonstra'e the functional capability of the recombiner system. We e

f find this commitment to te acceptable.

4 j The applicant has performed an analysis of the post-accident production and accumulation of hydrogen within the containment that is consistent with the j guidelines of Regulatory Guide 1.7, " Control of Combustible Gas Concentrations in

Containment following a Loss-of-Coolant Accident," and Branch Tecnnical Position CSB 6-2, of the same title. For this analysis, the applicant assumed operation of only one recombiner, actuated ten days after a postulated loss-of-coolant l

l accident whtn the hydrogen concentration is calculated to be approximately 3.5 i volume percent. The results of the analysis indicate that the containment hydrogen concentration would be maintained below the lower flammability limit of four volume percent, which meets our acceptance criteria.

4 We have reviewed the applicant's analysis and are unable to accept the results since the applicant has not provided su'ficient justification for calculated rate of hydrogen evolution, which would result from the corrosion of construction 4

materials inside containment. As a result, additional information is required f or the hydrogen evolution rate in order to com,Cete our review. Therefore, this matter remains outstanding.

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Based on our review, and subject to the satisfactory resolution of the matter regarding the calculated hydrogen evolution rate, we conclude that the proposed design of the combustible gas control system is in conformance with Regulatory Guide 1.7 and applicable staff posif: ws and is acceptable.

[

6.2.5 Containment Leakage Testing Program We have reviewed the applicant's containment leak testing program, as presented in l

Section 6.2 of the Preliminary Safety Analysis Report, for compliance with the containment leakage testing requirements specified in Appendix J to 10 CFR Part 50.

Such compliance will provide adequate assurance that containment leak-tight integrity can be verified throughout service lifetime and that the leakage rates will be periodically checked during service on a timely basis ta maintain leakage within the specified limits. Maintaining containment leakage witnin limits provides reasonable assurance that, in tne event of any radioactivity release within the l containment, the loss of the containment atmosphere through leak paths will not be in excess of the limits specified for the facility.

The ap,licant has provided a discussion of the test procedures and acceptance criteria for the containment integrated leak rate (Type A) test. The applicant has identified those systems penetrating containment that will be vented and drained to the containment atmosphere during the test 50 that the accident differential pressure '

will exist across the containment isolation valve. Justification was provided for cach system not vented and drained for the Type A test.

The applicant has also provided a list of all containment isolation valves and identified the local leak testing (Type C test) that will be done for these valves, for those isolation valves that are Type C leak tested, the applicant has indicated the direction in wnich the valve will be tested. Paragraph 111.C.1 of Appendix J to 10 CFR Part 50 allows testing in the direction opposite to that in which the valve performs its safety function if it can be shown that the test will result in equivalent or conservative leakage results. The applicant has committed to provide in the Final Safety Analysis Report the justification for any planned reverse direction testing. We find this commitment to be acceptable.

We have reviewed the applicant's provisions for containment leak testing and find that they will comply with the requirements of Appendix J to 10 CFR 50 and, therefore, conclude that they are acceptable. We will review the details of the leak testing program at the operating license stage of review.

6.3 imergency Core Cooling System 6.3.1 Design Bases Criterion 35 of the General Design Criteria and Section 50.46 of 10 CFR Part 50 l

require that an emergency core cooling system be provided which can perform its safety function assuming a single failure.

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The emergency core cooling system for Sundesert will be designed to provide emergency core cooling during those postulated accident conditions where it is assumed that mechanical failures occur in the reactor coolant system piping which result in a loss of coolant from the reactor vessel greater than the available coolant makeup capacity to be provided by normal operating equipment. The emergency core cooling system will also be designed to protect against the following postulated accidents:

(1) inadvertent opening of a steam generator relief or safety valve or a steam line break; (2) feedwater system pipe break; (3) rod cluster control assembly ejection; (4) inadvertent opening of a pressurizer safety or relief valve; and (5) steam gene tor tube rupture.

The y rgency core cooling system for the Sundesert plant will be similar in design, size, anu :opou ty to that of the Kosh<onong (Docket Nos. STN 50-502 and STN 50-503) and Shearon Harris (Docket hos. 50-400, 50-401, 50-402 and 50-403) plants, all of which will be designed for a core output of 2775 thermal megawatts. The design bases for the system are to (1) prevent fuel and cladding damage that would interfere with adequate emergency core cooling and (2) mitigate the ef fects of clad-water reaction, for any size break up to and including a double-ended rupture of the largest primary coolant line. The applicant has stated that the level of performance for core cooling will be met even with minimum engineered safeguards available, /

such as the failure of a diesel generator to start or component failure of the electrical supply bus. The Sundesert plant emergency core cooling system will be designed to withstand a safe shutdown earthquake and will have the required number, diversity, reliability, and redundancy of components such that, during a postulated accident, no single active component failure during the short term of an accident or no single active or passive failure during the long term of an accident will result in inadeauate cooling of the reactor core.

The boric acid injection portion of the emergency core ccoling system will be designed to control the reactivity insertion accompanying the rapid cooldown following any single 5. team line rupture or spurious relief valve lifting. Control of the reactivity insertion will be accomplished by injection of high concentration boric acid solution into the reactor coolant system. Protection will be provided for a range of steam line ruptures up to and including the doubled ended circumferential rupture of the largest pipe in the steam system.

6.3.2 Design Description The emergency core cooling system will consist of two independent subsystem trains.

Each series aligned train will consist of a nigh head safety injection / centrifugal charging pump, a low head residual heat removal pump, a heat exchanger, and associated valves and piping in the flow path to provide a sufficient capacity of borated water flow to satisfy the design bases. Power sources for actuation of the emergency core cooling system components will be supplied from separate emergency buses and separate diesel generators in the event of a loss of offsite power.

6-14

l Following a postulated loss-of-coolant accident, the emergency core cooling system will operate initially in the passive scrumulator mode and the active high head Injection mode, then in the active low head injection mode, and finally in the l

l recirculation mode. In the injection mode of operation, the pumps in each train of the emergency core cooling system will initially take suction f rom the ref ueling water storage tank and deliver flow to the reactor through three cold leg connections.

For the recirculation mode of operation, the pumps in each train will switch over to take suction from the contairment sumo.

Each of the three accumulators will have a total volume of 1450 cubic feet with a nominal volume of 1000 cubic feet of borated water and a nominal volume of 450 cubic feet of nitrogen gas at a minimum pressure of 600 pounds per square inch j go.me. The numinal boric acid concentration will be 1950 parts per million. Each tank will be connected to one of the cold legs of the reactor coolant system with two check valves in series. A normally open motor-operated gate valve will also be located in the lines between each accumulator and the cold leg piping. Adminis-trative procedures identified in the technical specifications will require that the a power to the motor operators of these normally open accumulator isolation valves be disconnected by removal of the breaker from the circuit during reactor power operation.

I Upon actuation of a safety injection signal, the high pressure injection mode of operation will consist of the ope.ation of two high head safety injection / centrifugal charging pumps (each rated at 150 gallons e minute at a design head of 5800 feet) ;

which will provide high pressure injection of boric acid solution from the boron injection tank (with a nominal concentration of 21,000 parts per million of boron) into the reactor coolant system. These pumps will automatically be aligned to take 4

suction from the refueling water storage tank which will have a minimum boron i concentration of 2000 parts per million.

j The boric acid injection portion of the emergency core cooling system will consist j of the boron injection tank, boron injection surge tank, boron injection recircula-

! tion loop, charging pumps, and the associated valves and piping. The boron injec-tion tank will contain 900 gallons of a boric acid solution (21,000 parts per million) and will be connected to the reactor coolant system by means of a loop 4

from the refueling water storage tank, through the high head safety injection /

centrifugal charging pumps, and to the boron injection tank inlet. The boron l

injection tank outlet will be connected, through a common manifold pipc, to pipes

' to be connected to each of the three reactor coolent cold legs. The boron injection surge tank will contain 75 gallons of the same concentration of boric acid as the boron injection tank and will be used to supply surge capacity for the boron injection tank recirculation loop. During normal operation, the boric acid solution will be recirculated by the two recirculation pumps continuously in a closed loop consisting of the boron injection tank and boron injection surge tank. This will be done to maintain mixing and prevent stratification. A safety injection signal will automatically stop the recirculation pumps and close the valves in the recirculation r

lines.

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Low pressure injection will be provided by two residual heat removal pumps (each j rated at 3000 gallons per rninute at a design head of 270 feet) which will .take

{ their suction from the refueling water storage tank. Upon actuation of the low-level setpoint from the refueling water storage tank following a postulated loss of-coolant i

accident, a changeover will be made from the injection mode to the recirculation mode.

As stated by the applicant, this will be initiated automatically, as described j below, and will be completed manually by operator action from the control room.

Protection logic will be provided to automatically open the two engineered safety

] features sump suction isolation valves in each train upon receipt of a refueling water storage tank low-level signal coincident with a reactor trip signal generated

{

l by a safety injection signal. This automatic action would align the two residual j heat removal pumps to take suction from the containment sump and to deliver coolant j

directly to the reactor coolant system cold leg piping. The low head safety l Injection / residual heat removal pumps would continue to operate during this changeover j

from the injection mode to the recirculation mode. The two charging pumps would j

contir.ue to take suction from the refueling water storage tank, following the above l automatic action, until manual operator action is taken to align these pumps in series with the residual heat removal pumps. Also, means will be provided to reduce the poter,tial for boron precipitation subsequent to a postulated loss of-coolant accident. Procedures will exist which will require manual switchover to

! hot leg recirculation, i

'[

6.3.3 Design Evaluation i

3 As discussed in Section 6.3.2 of this report, the emergency core cooling system will include the piping, valves, pumps, heat exche.Pgers, and associated instru-

[ mentation and controls to be used to transfer heat from the core following a

{ postulated loss-of-coolant accident. The scope of our review of the emergency core cooling system for the Sundesert plant included piping and instrumentation diagrams, k

1 equipment layout drawings, failure modes and effects analyses, and design specifica-tions for essential components. The review also included the applicant's proposed

! I design criteria and design bases for the emergency core cooling system and the l manner in which the design will conform to these criteria and bases. Specifically, we evaluated the system's ability to withstand a single active failure during the j

short term or a single active or passive failure during the long term following a

}.

postulated loss-of-coolant accident.

Although the applicant does not recognize spurious valve repositioning as a credible event which could affect the system's capability of performing its intended safety l function, the applicant has provided a commitment that the design of the system will provide for the removal or power to the valve operators for certain valves which are required to remain in their normal Operating position in the event of a loss-of-coolant accident. These valves are:

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(1) Accumulator isolation valves (to remain open): 1-8808A, 1-8808B and 1-8808C (2) Valve between the hot leg and the low head safety injection / residual heat l

removal pump discharge (to remain closed): 1-8889 (3) Valves between the hot leg and the high head safety injection / centrifugal charging pump discharge (to remain closed): 1-8884 and 1-8886 t

We find this commitment to be acceptable for the above valves. However, based on our review of the p 4ing and instrumentation diagrams for the Sundesert riant, we I

j I

~

have identified an additionE motor operated valve required either to have power locked out to remain closed during plant power operation, or to be included as a failure mode in submitted analyses. This would be Valve 1-8885 in the high head safety injection / centrifugal charging pump discharge line to the cold leg injection header which must remain closed to prevent unplanned dilution of the boron concentra-tion of the cooling water. Therefore, this matter remains outstanding, j i During our review, we requested that the applicant identify any manual (band-wheel) valve which could jeopardize the operation of the emergency core cooling system if inadvertently left in the wrong position. The applicant provided a list of critical valves and stated that they will be administratively controlled, have a local position indicator, and have locking provisions. We find this acceptable. However, we also require that a system-level indicator be provided on the main board in the control room to indicate bypass or inoper6ble status of these valves, in conformance with the recommendations of RegJlatory Guide 1.47, " Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety Systems." Therefore, this matter remains a outstanding.

l The applicant has stated that a detailed f ailure mode and ef f ects analysis for the {

emergency core cooling system has been performed and is described in the Westing-house Topical Report WCAP-8977, " Failure Mode and Effects Analysis of ECCS for a Westinghouse Type 312 3-Loop Pressurized Water Reactor." We have not completed our review of this topical report. However, based on the similarity of the Sundesert emergency core cooling system to previously accepted plant designs, we conclude that the proposed design is acceptable at this stage of review. At the operating license stage of review, we will assure that the failure mode and effects analysis i for Suncesert is consistent with our conclusions resulting from the review of WCAP-8977.

We also reviewed the potential for valve motors and associated valve control /

actuation systems to become submerged within the containment following a postulated l loss-of-coolant accident. The applicant has stated that all emergency core cooling system valves and control / actuation systems will not be vulnerable to flooding. l The applicant indicates that, following a postulated large break loss-of-coolant accident, the maximum water level height would be 6.9 feet above the floor inside

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the containment. This height will be below the proposed location of the emergency j core cooling system components which we find acceptable. At the operating license

{ stage of review, we will evaluate the results of the analysis reflecting the as-built 4

plant conditions.

4 i

With regard to long-term recirculation cooling, we requested the applicant to j}

evaluate the emergency core cooling system performance by applying a single failure a

analysis which includes consideration of passive as well as active failures in the

[ emergency core cooling system (assuming no prior failure during the short-term j phase). As part of the passive failure evaluation, the applicant addressed the j concern of post-accident water leakage from emergency core cooling system components, i

such as a failed pump seal or valve stem packing, which could degrade more than one k subsystem. The maximum 'eakage from a passive failure has been determined by the f applicant to be less t' an 50 gallons per minute for the case of a sudden pump shaf t f seal fai. ore. Valve packing leaks are considered to be less severe than the pump seal failure. We conclude that the maximum leakage rate determined by the applicant for assessing passive failures is acceptable.

I i

i In order to assure that post accident leakage, from emergency core cooling system components located outside the containment, does not degrade other safety systems, the applicant has committed to incorporate a leak detection system with the following i

appropriate design features. Detection of leakage from emergency core cooling l

system components will be accomplished by monitoring the sump levels in the engineered f safety feature areas, including the sump in the charging pump cubicle. The leakage

}

to be collected in these sumps will be detected by instruments sensitive enough to

( initiate, by alarm, operator action to isolate the leak prior to adversely affecting j other systems by flooding. The system design will be based on assuming no operator i

action for 30 minutes prior to isolation of a leak. The leak detection system will

{ identify the faulted emergency core cooling system train and the leak will be j

i isolable. Also, the level sensors, power supplies, and alarms will meet the criteria enumerated in IEEE Standard 279, " Criteria for Protection Systems- for Nuclear Power i

Generating Stations," with the exception that this detection system need not meet the single failure criterion. We conclude that the applicant's commitment to f incorporate the above design features into the leak detection system is acceptable.

j 14anual switchover to hot leg recirculation for long-term cooling following a postulated 4 loss-of-coolant accident will be employed as the means to prevent excessive boron l p recipi ta tion. The applicant estimates that the hot leg recirculation will be f initaited about 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months af ter the start of the accident. Based on previous calcula-1 4

tions for plants of similar design, we conclude that this time frame is acceptable.

) Confirmation of the time available for manual switchover will be made at the operating

} license stage of review when the final design is available.

4 i 1 During our review of this matter, we determined that the injected fluid from hot leg recirculation may not have enough head to pass through the core to assure I

i e

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adequate dilution flow after a hot leg break. Thus, an additional path may be Since this aspect required, such as simultaneous hot and cold leg recirculation.

is administrative in nature and since the Sundesert plant system design is amenable to simultaneous hot and cold leg injection, we conclude that the details of the procedures to prevent boron precipitation may be addressed at the crerating license stage of review. l 6.3.4 Performance Evaluation l.

The applicant has submitted an evaluation of the emergency core cooling system performance pursuant to the requirements of 10 CFR Part 50.46 of the Commission's regulations. The results of analyses for postulated small and large break lost of-coolant accidents have been provided in Section 15.6.5 of the Preliming Safety Analysis Report. The large break loss-of-coolant accident analpis was limited to a spectrum of three double-ended guillotine breaks with r6 charge coef ficients (Moody multiplier) of 0 4, 0.6, and 0.8. To supplement the large break analysis, l

the applicant has referenced Westinghouse Topical Report WCAP-8853, " Westinghouse l ECCS-Three Loop Plant (17x17) Sensitivity Studies," which covers other break sizes.

types, and locations, and demonstrates that the double-ended cold leg break of the reactor coolant system is the worst case for this type plant. We previously reviewed this topical report and found it acceptable as documented by letter to Westinghouse, dated April 1, 1977. In this topical report and in the applicant's analyses, the upper head temperature of the reactor vessel was assumed to correspond to the hot leg temperature which is conservative. All of the analyses were performed with the previously approved version of the Westinghouse evaluation model, described in

.l Topical Report WCAP-8622 " Westinghouse ECCS Evaluation Model October 1975 Version," ,

I and assumed a Model D steam generator in the plant design. I The results of the analyses submitted by the applicant identify the worst break as I

' the double-ended cold leg guillotine break with a discharge coefficient (Moody multiplier) of 0.6. The calculated peak clad temperature was 2185 degrees Fahrenheit which is within the acceptable limit of 2200 degrees Fahrenheit as specified in g

In addition, the calculated maximum local l Section 50.46(b)(1) of 10 CFR Part 50.

metal / water reaction of 8.3 percent and total core-wide metal / water reaction of less than 0.3 percent are well below the allowable limits of 17 percent and one percent, respectively, as specified in Sections 50.46(b)(2) and 50.46(b)(3) of 10 CFR Part 50. Coolable core geometry will be preserved and long-term cooling will be maintained. The analyses were performed based on an assumed total peaking

' factor of 2.10, 102 percent of the engineered safety features nuclear steam supply system power of 2910 megawatts thermal, and 102 percent of a peak linear power density of 11.66 kilowatts per foot.

In Amendment No. 3 to the Preliminary Safety Analysis Report, the applicant Due to this Cesign submitted a revised steam generator design (new Model F).

change, an additional analysis was provided for the limiting break (double-ended

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j cold leg guillotine rupture with a discharge coefficient of 0.6) using the Model F j steam generator parameters. The results show the peak clad temperature to be 2172 degrees fahrenheit, the maximum local metal / water reactor to be 8.04 percent, and j the total core-wide metal / water reactor to be less than 0.3 percent. These results also comply with the acceptance criteria of Section 50.46(b) of 10 CFR Part 50. !

j Appendix K to 10 CFR Part 50 also requires that the combination of emergency core cooling subsystems to be assumed operative shall be those available assuming the most severe single failure. The worst single f ailure in the emergency core cooling i system was assumed to be the loss of one safety injection train. The minimum backpressure in the containment following a postulated loss-of-coolant accident was j based on maximum containment cooling (all containment safeguard systems operating).

This set of assumptions is conservative since it maximizes the peak cladding tempera-ture. Our evaluation of the minimum containment backpressure analysis is presented in Section 6.2.1 of this report. As stated in that section, we conclude that the analysis is acceptable.

The applicant also included an analysis of postulated small break loss-of-coolant accidents for three break stres specific to the Sundesert plant. The break sizes evaluated were three-inch, four-inch, and six-inch breaks. The applicant has stated that the analysis was performed with the previously approved October 1975 version of the Westinghouse evaluation model and also made reference to Westinghouse Topical Report WCAP-8970, " Westinghouse Emergency Core Cooling System Small Break October 1975 Model," which we have found acceptable as documented by letter to Westinghouse, dated June 8, 1978. The three-inch pipe break was identified as the limiting small break with a calculated peal clade temperaturo of 1648 degrees Fahrenheit, maximum j local metal / water reaction of 2.27 percent, and a total core wide metal / water reaction of less than 0.3 percent, in order to justify that the worst small break has been identified, we requested additional information from the applicant. We also requested information pertain-ing to post-loss-of-coolant accident decay heat removal capability for the small break analysis, The applicant has provided information in response to these requests, but our review of the responses has not been completed. Therefore, this matter remains outstanding. I We have recently been advised by Westinghouse and the applicant that the heat generation rate from metal / water reaction used for all previous Westinghouse plant 3

emergency core cooling system calculations, including Sundesert, was in error. We I require resolution of this modeling error for Sundesert. Therefore, this matter remains outstanding.

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i 6.3.5 Tests and Inspections 1

The applicant will demonstrate the operability of the emergency core cooling system by subjecting all components to preoperational tests, periodic testing, and in-service I testing and inspections. The preoperational tests will be performed in conformance j with Regulatory Guide 1.68, " Initial Test Prof ams for Water-Cooled Reactor Power l l Plants," Revision 1, and Regulatory Guide 3.79, "Preoperational Testing of Emergency Core Cooling Systems for Pressurized Water Reactors," Revision 1. For the low head safety injection recirculation test, the applicant has committed to perform the test with the low head safety injection pump taking suction from the engineered safety features sump instead of the refueling water storage tank to verify that 1 adequate net positive suction head will be available to the pump. I The applicant has also proposed a modified test program to verify the operability of the accumulator check valves at higher temperatures. Regulatory Guide 1.79 recommends that check valve operability be verified by performing a core flooding flow test under hot operating conditions and slowly decreasing the reactor coolant system pressure and temperature until the accumulator check valves operate. The i applicant's test program will use the high head safety injection pumps and the leakage test lines for the check valves to provide flow through the check valves.

The applicant has stated that this method would allow the check valves to be nearer to the operating temperature. Accordingly, we find this procedure to be acceptable.

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l I The applicant plans to perform routine periodic testing of the emergency core cooling system components and all necessary support systems with the plant at power. Valves that are required to operate in the event of a loss-of-coolant accident will be tested through a complete cycle and pumps will be tested individ-ually on their miniflow lines. As indicated above, test lines will be provided to l

) perform periodic tests on emergency core cooling system check valve operability.

The applicant has stated that tests will be performed to verify that each of the check valves in series can independently sustain differential pressure across its disc, and to verify that the valve is in its closed position. Lines in which the

! check valves are to be tested include the residual heat removal pump cold leg, the j hot leg injection lines and the high head safety injection / charging pump cold leg injection lines.

Periodic emergency core cooling system testing will also include a visual inspec-tion of pump seals, valve packings, flanged connections, and relief valves to detect leakage. The applicant has stated that the emergency core cooling system components will be designed and fabricated to permit inspection and in-service tests in accordance with Section XI of the ASME Code. For the long-term core cooling operation (normal and post-loss-of-coolant accident), we will require that the applicant provide assurance, during the operating license stage of review, that the low head safety injection / residual heat removal pumps will operate for the time i period required to fulfill that function.

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As a result of our review, we conclude t'nat the proposed tests and inspections of the Sundesert emergency core cooling system design are acceptable.

6.3.6 Conclusions We have reviewed the drawings, component descriptions, design criteria, performance analyses, and the proposed testing of the Sundesert emergency core cooling system. !

Based on our review, and subject to the satisfactory resolution of the matters discussed in Section 6.3.3 of this report relating to (1) removal of electrical power from emergency core cooling system motor operated valves that may degrade the intended performance of the emergency core cooling system and (2) position indica-tion in the control room of critical emergency core cooling system manual valves, l

we conclude that the proposed system conforms to the Commission's requirements as set forth in the General Design Criteria, regulatory guides, and staff technical j positions. Subject to the satisfactory resolution of the items discussed in I Section 6.3.4 of this report relating to (1) our incompleted review of the small break performance evaluation and (2) an error in the emergency core cooling system evaluation model, we also conclude that the proposed system meets the acceptance ;

criteria of Section 50.46(b) of 10 CFR Part 50.

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6.4 Cnntrol Room Habitability We reviewed the control room ventilation system and control building layout and ,

structures, to assure that plant operators will be adequately protected against the l effects of accidental release of toxic and radioactive gases, and that the plant c5n 59 safely operated or shut down under design basis accident conditions. Our evaluation was based on the acceptance criteria described in Section 6.4 of the Standard Review Plan.

Relevant portions of the control room ventilation system are described in this section, while the evaluation of the entire ventilation system is prssented in Section 9.4.1 of this report.

6.4.1 General Description The control room for each unit of the facility will be housed in separate build- j

, ings. The control buildings and the ventilation systems for the control rooms will be constructed to seismic Category I requirements and will be protected against tornado generated missiles. Each control room will be equipped with two remote air ;

intakes, separated from one another by about 800 feet, on opposite sides of the control building and will be away from the reactor buildings by more than 200 feet.

Each air intake will be equipped with redundant quick-acting chlorine detectors, radiation monitors, and two positive shutoff butterfly valves (isolation dampers) l in series, i

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The isolation dampers will meet the single failure criterion, and will be able to be operated manually. Each control room ventilation system will be equipped with two redundant standby filtration systems, each containing a four-inch charcoal bed.

I During normal operation, fresh air will be drawn throsgh the air intakes and the air Inside the control area will be recirculated. A positive pressure of 0.125-inch f

water gauge will be maintained in the control room. Vestibule type doorways will be provided to maintain pressurization in the control room envelope. An acceptance test will be conducted to verify the level of pressurization.

When one of the remote air intakes detects radioactivity or chlorine, the isolation dampers in that intake will be automatically closed, fresh air will be drawn

'through the other intake to maintain pressurization, and the supply and return air in the control area will be automatically diverted through the standby filtration system. In the event that both air intakes are closed, portable individual breath-ing apparatus will be availatde for the control room operators.

6.4.2 Radiation Protection Provisions The applicant proposes to meet Criterion 19 of the General Design Criteria by using two feet of concrete shielding for the walls and roof of the control room, and by installing redundant charcoal filters for use in recirculation of the control room air during an emergency. The standby filtration system for the control room is discussed in Section 6.5.4 of this report.

We determined that, following a postulated design basis accident, such as a loss of coolant, both the calculated thyroid and whole body doses to the control room personnel will be 1.0 rem or less, based on a total leakage rate of 0.2 percent per day from the reactor building. These calculated radiation doses are well within the guideline values of Criterion 19. We, therefore, conclude that the proposed radiation protection provisions for the control room are acceptable.

6.4.3 Toxic Vapor Protection Provisions We reviewed the provisions for protection of the control room a.;ainst toxic vt.pors

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using the guidelines in Regulatory Guide 1.78, " Assumptions for Evaluating the Habitability of a huclear Power Plant Control Room during a Postulated Hazardous Chemical Release," and Regulatory Guide 1.95, " Protection of Nuclear Power Plant Control Room Operators Against an Accidental Chlorine Release, Revision 1."

No toxic chemicals will be stored in the control room buildings. Aqueous ammonia and sulfuric acid will be stored in the turbine buildings, which are physically separated from the control room buildings and air intakes. Compressed chlorine, up to two containers of 30 tons each, will be stored on site 488 meters from the nearest air intake. A few other less toxic chemicals will also be stored onsite.

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Anhydrous ammonia is infrequently shipped via Highway 95 within five miles of tr.e site.

The applicant proposes to install redundant, sensitive and automatic, seignic Category I detection and isolation equipment for toxic chemicals which could produce vapor concentrations inside the control room in excess of their respective toxicity limits. The ventilation system will be designed with fast damper closure times such that the chlorine vapor concentration inside the control room will not exceed the toxicity limit for a two-minute exposure. Adequate time will be available for operators to don portable individual breathing apparatus which will be provided in the control room.

The applicant proposes to take the following exceptions or clarifications to Regulatory Guide 1.95:

(1) Regulatory Position C.3 maximum allowable chlorine inventory in a single container stored onsite. In lieu of the inventory restriction guidelines of this position, the applicant performed an evaluation of the control room habitability using the general design considerations of Regulatory Guide 1.78, and proposes to design the control room such that the chlorine concentration inside would not exceed 15 parts per million within two minutes of detection.

We conclude that, since this proposed design criterion meets the acceptance criteria of Regulatory Guide 1.95, and since Regulatory Position C.3 was established on the basis of the general considerations of Regulatory Guide 1.78, l this exception is acceptable.

l (2) Regulatory Position C d - immediately after control room isolation, the emer-l gency recirculating charcoal filter, or equivalent equipment designed to remove or otherwise limit the accumulation of chlorine within the control room, should be operated. The applicant indicates that the charcoal filters to be used are not specifically designed to remove or limit chlorine accumula-tion, but that the filters will be used during an emergency for recirculation.

We determined that, although the four-inch charcoal bed is not specifically designed to remove chlorine, it has been demonstrated that this type of filter is capable of adsorbing chlorine (B. A. Soldano, " Activated Charcoal Filter and The Adsorption of Chlorine," Furman University,1974). Accordingly, we conclude that this clarification is acceptable.

(3) Maintenance, testing, and calibration of chlorine detectors - the applicant proposes to conduct a channel check of the chlorine detection system at least once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and a function test at least once every 31 days, instead of an operational ctuck at one week intervals as recommended. The applicant also proposes to conduct a calibration check at least once every 18 months instead of six months, as recommended. We determined that the proposed surveil-lance program would not be adequate to assure proper operation of the chlorine 6-24

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detection system as a result of our previous reviews for other applications.

The frequency of testing and calibration of the detectors will not affect the design and construction of the facility and, therefore, need not be resolved at the construction permit stage of review. However, we will require, at the operating license stage of review, that the applicant comply with the testing and calibration schedule for chlorine detectors, as described in Regulatory P

Guide 1.95, unless an acceptable alternative is proposed.

We determined that, for an accidental release of ammonia or chlorine, the toxic l vapor concentration inside the control room would not exceed their respective toxicity limits when the affected air intake is closed. The likelihood of both air intakes being contaminated simultaneously by toxic vapors is very low. If required, j the control room operators would have enough time to don breathing apparatus. The other toxic materials would not pose significant hazards to control room habitability.

We, therefore, conclude that the proposed toxic vapor protection provisions for the control room are acceptable.

6.5 Engineered Safety Feature Atmosphere Cleanup Systems 6.5.1 Summary Description The engineered safety feature atmosphere cleanup systems for the Sundesert plant will consist of process equipment and instrumentation to control the release of ,

l radioactive materials in gaseous effluents (radioactive iodine and particulate matter) following a design basis accident. In the Sundesert plant, there will be three filtration systems designed for this purpose: the annulus building filtra-tior, system, the fuel building filtration system, and the control room standby filtration system. In addition, the containment spray system will be designed for the removal of elemental iodine from the post-accident containment atmosphere, thereby reducing its release following a postulated loss-of-coolant accident.

6.5.2 Annulus Building Filtration System The function of the annulus building filtration system will be to collect and process the leakage of radioactive materials from the following potential sources:

(1) leakage from the engineered safety feature equipment located in the annulus building following a postulated loss-of-coolant accident and (2) leakage from the containment building into the atmosphere of the annulus building following a postu-lated loss-of-coolant accident or a fuel handling accident inside containment. The system will be designed to maintain a slight negative pressure in the annulus l

building in the event of a loss-of-coolant accident.

The annulus building filtration system will have redundant trains. Each train will have a design capacity of 30,000 cubic feet per minute of air and will include the following sequential components; heating coil, prefilter, high efficiency particu-late air filter, carbon adsorber, high efficiency particulate air filter, and fan.

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The equipment and components will be designed to Quality Group C Lnd as seismic Category I and will be located in a seismic Category I structure.

Based on our review, we have determined that the annulus building filtration system will be designed in accordance with the guidelines of Regulatory Guide 1.52, " Design,

Testing, and Maintenance Criteria for Atmospheric Cleanup System Air Filtration and Absorption Units of Light-Water-Cooled Nuclear Power Plants," and will be capable of controlling the release of radioactive materials in gaseous effluents following a postulated design basis accident. We, therefore, find the proposed system to be acceptable.

. 6.5.3 Fuel Building Filtration System i

The function of the fuel building filtration system will be to process potentially radioactive air prior to release to the environment after a postulated fuel handling accident, The system will be designed to maintain a slight negative pressure in the fuel building in the event of a fuel handling accident.

The fuel building filtration system will have redundant trains. Each train will have a design capacity of 8000 cubic feet per minute of air and will include the following sequential components: demister, heating coil, prefilter, high ef ficiency particulate air filter, carbon adsorber, high efficiency particulate air filter, and fan. The equipment and components will be designed to Quality Group C. However, the applicant has classified the equipment and components of the filtration system, including the portion of the building that will house the system, as non seismic Category I. We will require that the equipment and components of the fuel building filtration system be designed as seismic Category I, in conformance with the guide-lines of Regulatory Guide 1.52, and that the system be located in a seismic Cate-gory I structure (see also Sections 3.2.1, 9.4.2 and 15.5.2 of this report).

Subject to the satisfactory resolution of the above matter, we conclude that the fuel building filtration system will be designed in accordance with the guidelines of Regulatory Guide 1.52 and will be capable of controlling the release of radioactive materials in gaseous effluents following a postulated design basis accident. l 1

6.5.4 Control Room Standby Filtration System The function of the control room standby filtration system will be to process potentially radioactive air in the control room af ter a postulated basis accident to permit operating personnel to remain in the control room following the postu-lated accident. The control room standby filtration system will have redundant trains which will be operated to pressurize the control room following a postulated design basis accident. Each train will have a design capacity of 1500 cubic feet per minute of air and will include the following sequential components; demister, prefilter, heating coil, high ef ficiency particulate air filter, carbon adsorber, 6-26

high efficiency particulate air filter, and fan. The equipment and components will be designed to Quality Group C and as seismic Category I and will be located in a seismic Category I structure. .

Based on our review, we have determined that the control room standby filtration system will be designed in accordance with the guidelines of Regulatory Guide 1.52 and will be capable of maintaining a suitable control room environment following a j postulated design basis accident. We, therefore, find the proposed system to be acceptable.

6.5.5 Containment Spray System We reviewed the containment spray and spray additive systems to determine their fission product removal effectiveness in the event of a postulated design basis accident. Our evaluation was based on the acceptance criteria described in Section 6.5.2 of the Standard Review Plan.

The containment spray system will have two parallel and separate flow paths, each consisting of a spray pump, three spray headers, spray nozzles, an engineered safety feature pump, and associated piping and valves. Following a postulated loss-of-coolant accident, borated water from the refueling water storage tank will be mixed with a sodium hydroxide solution, and the resultant solution will be sprayed over the containment atmosphere. The spray system will become effective 85 seconds after the accident is initiated. When the refueling water storage tank water level becomes low, the spray solution will be taken from the containment sump by the engineered safety feature pump. Sodium hydroxide will continue to be mixed with the spray solution until the pH of the sump water reaches 8.5.

The spray nozzles will be of the SPRACO 1713 type, will have no internal moving parts, and will have an orifice diameter of 3/8 inch. The spray will cover 90 percent of the containment cross-sectional area at the containment bend line and 70 percent of the total containment volume. Air mixing within the containment will be achieved by the containment ventilation system. The spray system will be capable of operating continuously for at least 30 days after an accident.

Pre-operational testing of the containment spray pumps and eductors, periodic testing of safety valves and spray pumps, periodic sampling of the refueling water and sodium hydroxide solution, and periodic checking of the water level in the refueling water storage tank will be conducted. Technical specifications of the limiting conditions for operation of the spray and additive systems are given in Section 16.0 of the Preliminary Safety Analysis Report.

We have determined that the containment spray system meets the redundancy require-

'ments of an engineered safety feature and the single failure criterion for the active components. The concept upon which the proposed system is based has been 6-21 l - - - - - - _ _ . - _ , _ _ _

Demonstrated to be effective for iodine removal and retention under post-accident conditions. Therefore, we conclude that the proposed system design is acceptable.

The proposed pre-operational tests, post-operational testing and survelliance, and proposed limiting conditions for operations for the spray system will provide adequate assurance that the iodine scrubbing function of the containment spray system will meet the effectiveness assumed in the accident evaluation.

The values of the parameters we used for the evaluation of the iodine removal

. effectiveness of the containment spray system following a postulated loss-of-coolant accident are given in Table 15.4 of this report, We have also reviewed the proposed methods and procedures for controlling the chemical composition of the solutions to be sprayed and recirculated within the containment after a postulated loss-of coolant accident, to assure that adverse chemical reactions or inadequate mixing of solution will not occur. l l

We determined that, with the proposed methods and procedures, the pH of the spray j solution during the injection phase and the initial pH of the sump water during the recirculation will be at least 8.0 but no more than 11.0. The methods and procedures for controlling the pH of the spray solutions have been found adequate since they provide assurance that the pH of the spray solutions will be kept at a level which minimizes the possibility of stress corrosion cracking of mechanical systems a '

components. Accordingly, we conclurie that the proposed pH control system is acceptable. I

6. 6 Inservice Inspection of Engineered Safety Features Criteria 36, 39, 42, and 45 of the General Design Criteria require that the emer-gency core cooling system, containment heat removal system, containment atmosphere cleanup system, and cooling water system be designed to permit appropriate periodic inspection of important component parts to assure system integrity and capability, i l

l To assure that no deleterious defects develop during service in ASME Code Class 2 system components, selected welds and weld heat-affected zones will be inspected prior to reactor startup and periodically throughout the life of the plant. In addition, ASME Code Class 2 and 3 systems will receive visual inspections while the systems are pressurized in order to detect leakage, signs of mechanical or structural distress, and corrosion. An augmented inservice inspection program will be developed for high energy fluid system piping between containment isolation valves or, where no isolation valve is used inside the containment, between the first rigid pipe connection to the containment penetration or the first pipe whip restraint inside the containment and the outside isolation valve. The program will be included in l the Final Safety Analysis Report. The applicable Edition and Addenda of Section XI of the ASME Code will be determined in accordance with the rules of 10 CFR Part 50, paragraph 50.55a(g) and will be addressed in the Final Safety Analysis Report.

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Compliance with the inservice inspections required by the ASME Code and staff technical positions constitutes an acceptable basis for satisfying the applicable requirements of Criteria 36, 39, 42, and 45 of the General Design Criteria. Therefore, we conclude that the proposed program is acceptable.

l 6.7 Engineered Safety Features Metallic Materials The mechanical properties of the materials to be selected for the engineered safety features will satisfy the requirements of Appendix 1 to Section III of the ASME i

Code and the applicable portions of Parts A, B, and C to Section II of the ASME

! Code, and our position that the yield strength of cold worked stainless steels shall be less than 90,000 pounds per square inch, The controls to be imposed on the pH of the containment spray water and the emergency core cooling water, following a postulated loss-of-coolant accident, are adequate to assure freedom from stress corrosion cracking of the austenitic stainless steel components and welds of the containment spray and emergency core cooling systems throughout the duration of the postulated accident to completion of cleanup. The controls to be imposed on the use and fabrication of the austenitic stainless steel for the systems satisfy the recommendati ons of Reguletory Guide 1.31, " Control of Stainless Steel Welding," and Regulatory Guide 1.44, " Control of the Use of Sensitized Stainless Steel " Fabrication and heat treatment practices perforrr.ed in accordance with these recommendations provide added assurance that stress corrosion cracking will not occur during the postulated accident time interval.

The control of the pH of the spray and cooling water, in conjunction with controls on selection of containment materials, are in accordance with the recommendations l of Regulatory Guide 1.7, " Control of Combustible Gas Concentrations in Containment following a Loss-of-Coolant Accident," and provide assurance that the spray and

! cooling water will not cause serious deterioration of the containment. The controls j to be placed on concentrations of leachable impurities in nonmetallic thermal insulation used on austenitic stainless steel components of the engineered safety features are in accordance with the recommendations of Regulatory Guide 1.36,

Nonmetallic Thermal Insulation for Austenitic Stainless Steel."

Conformance with the codes and regulatory guides mentioned above, and with the requirements on (1) the allowable maximum yield strength of cold worked austenitic stainless steel, and (2) the minimum levet of pH of containment spray and emergency core cooling water, constitute an acceptable basis for meeting the applicable requirements of Criteria 35, 38 and 41 of the General Design Criteria.

6.8 Organic Materials l We reviewed the proposed criteria for coatings to be used inside the containment to l l

I determine their suitability for design basis accident conditions. We also reviewed i

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the organic materials to be used to determine any potential adverse interaction l with the engineered safety feature equipment.

The applicant states that the l>rotective coatings to be selected will meet the >

recommendations of Regulatory 3uide 1.54, " Quality Assurance Requirements for Protective Coatings Applied tr Water-Cooled Nuclear Power Plants," with the follow-ing exceptions:

(1) Documentation for coatings to De used on manufactured equipment, such as pumps, motors, pipe hangers, and support , will be kept in the vendor's files instead of the applicant's files, and a certificate of compliance will be furnished by the vendor. The vendor's files will be audited periodically to verify that complete documentationw ill be maintained. We determined that this alternative procedure of documentation is consistent with the quality assurance requirements of Appendix B to 10 CfR Part 50 and is acceptable.

(2) Small components and equipment surfaces may not be coated with qualified paints. However, process specifications will control the application of i

paints to those structures and components which may come into contact with the ~

containment spray and cooling systems. The applicant estimates that these surfaces will be small relative to the total area to be covered by qualified paints. At the ccnstruction permit stage of review, it is impractical to require an applicant to specify the type and quantity of all unqualified paint on small components. In addition based on reviews of similar plants, we determined that the quantity of unqualified paint on small components will be small and will not pose a safety problem. Therefore, we conclude that this exception is acceptable.

The applicant states that the organic materials inside the containment will consist principally of paints, coatings, and insulation, and that there will be no signifi-cant quantity of other organic materials. At the operating license stage of review, we will require that the applicant specify the total amount of paint and organic materials used which do not meet the recommendation of Regulatory Guide 1.54 and which will he exposed to the containment atmosphere.

In summary, the proposed criteria for containment coatings have been evaluated as to their suitability to withstand a postulated design basis accident environment.

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The coatings to be chosen by the applicant will be qualified under conditions which take into account the postulated design basis accident conditions. No adverse interactions (under design basis accident conditions) between the decomposition products and the engineered safety features have been identified. We conclude,

therefore, that the proposed criteria for the coatings are acceptable.

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7.0 INSTRUMENTATION AND CONTROLS The applicant has described the design criteria and bases for the instrumentation and control systems, including the reactor protection system, the engineered safety features actuation system, safety-related display instrumentation and control systems not required for safety. The applicant has also described the proposed criteria for the physical independence of the instrumentation and control systems.

The above systems are similar to the systems we previously reviewed for the Koshkonong plant (Docket Nos. STN 50-502 and STN 50-503). At the time the applicant announced suspension of its activities on Sundesert (see Section 1.1 of this report), our review of the Sundesert instrumentation and control systems had not progressed to the point where we could present our conclusions. Therefore, this matter remains outstanding.

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1 8.0 ELECTRIC POWER SYSTEMS l

1 8.1 General Discussion 1 l

l The Commission's General Oesign Criteria, including Criteria 17 and 18; the staff r l

technical positions noted in Table 8-1 of the Standard Review Plan; the institute of Electrical and Electronics Engineers (IEEE) standards, including IEEE l Standard 308-1974, " Criteria for Class lE Electric Systems for Nuclear Power Gener- l ating Stations"; and applicable regulatory guides, including Regulatory Guide 1.6,

" Independence Between Redundant Stardoy (Onsite) Power Sources and Between Their Distribution Systems," Regulatory Guide 1.9, " Selection of Diesel Generator Set Capacity for Standby Power Supplies," Regulatory Guide 1.32, " Criteria for Safety-Rela *ted Electric Power Systems for Nuclear Power Plants," Regulatory Guide 1.41, "Preoperational Testing of Redundant Onsite Electric Power Systems to verify Pmer ,

Load Group Assignments," and Regulatory Guide 1.75, " Physical Independence of Electric Systems," Revision 1; were utilized as the bases for evaluating the adequacy of the electric power systems for the Sundesert plant.

i 8.2 Offsite Power System The offsite power system will be the preferred source of power for the plant. This system will include the grid, transmission lines, transformers, switchyard components and associated control systems provided to supply electric power to safety-related equipment and other equipment. The electrical grid will be the source of energy for the offsite power system. The safety function of the offsite power system (assuming that the onsite power systems are not available) will be to provide sufficient capacity and capability te assure that the specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary will not be exceeded and to assure that core cooling, containment integrity and other vital functions will be maintainea in the event of postulated accidents. The objectives of our review were to determine that the offsite power system (1) will satisfy the criteria set forth in Section 8.1 of this report, and (2) can reliably perform its design functions during plant normal operation, anticipated operational occurrences, and accident conditions, i

The offsite power system proposed for Sundesert plant will include (1) four physi-cally independent 500 kilovolt lines to the 500 kilovolt switchyard, and (2) two physically independent 500/69 kilovolt circuits which will connect the 500 kilovolt switchyard to the 69 kilovolt switchyard. Both switchyards will be arranged in a breaker-and-a-half configuration. Two out of the four transmission lines will enter the 500 kilovolt switchyard on separate rights of way. The physical separa-tion between the four transmission lines will be such that no single event (such as i

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a tower f alling or a line breaking) ecan simultaneously remove all four lines from service. Each circuit will be provided with primary and backup protection for isolation of components in the event of failure or abnormal conditions. Two separate and independent 120 volt direct current systems, independent of the generating station batteries, will be provided for operation of primary and backup protective devices of the switchyard breakers. The station's two turbine-generators will provide power at 25 kilovolts through the corresponding main transformers to the 500 kilovolt switchyard. l The onsite Class IE power distribution buses of each unit will be supplied by three independent 69 kilovolt circuits emanating from the 69 kilovolt switchyard.

Each 69 kilovolt circuit will be capable of immediately providing power to safety-related equipment following a postulated loss of-coolant accident and will have sufficient capacity and capability to supply the load connected to the respective bus during normal or abnormal operating conditions.

The applicant has conducted a grid stability analysis and determined that the electrical grid w remain stable after (1) a loss of the largest single supply to the grid, (2) removal of the largest load from the grid, and (3) simulation of a three phase fault at the Sundesert 500 kilovolt bus with the resulting outage of

, two critical transmission circuits.

We have reviewed the applicant's grid stability analysis and conclude that the results of the analysis provides reasonable assurance that the ability of the applicant's grid to provide offsite power to the Sundesert plant will not be impaired by the loss of the largest single supply to the grid or by the loss of the largest load on the grid. This satisfies our requirements set forth in Section 8.1 of this report and is acceptable.

The design of the offsite power system will include provisions for periodically testing the capability of the system to start and operate all required safety-j related loads in conformance with the requirements of Criterion 18 of the General Design Criteria.

l Conclusion

, Our review of the offsite power system for the Sundesert plant covered single-line diagrams, preliminary physical layout drawings and descriptive information for the system. The review also included the applicant's proposed design criteria and design bases for the offsite power system and its analyses of the adequacy of those criteria and bases. On the basis of our review, we conclude that the design of the offsite power system satisfies our requirements set forth in Section 8.1 of this report and, therefore, is caceptable.

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6.3 Onsite Power Systems 8.3.1 Alternating Current Power System The onsite Class IE alternating current power system will serve as standby to the offsite power system. The safety function of the Class lE alternating current power system (assuming that the offsite power system is not available) will be to provide sufficient capacity and capability to assure that the specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary will not be exceeded and to assure that core cooling, containment integrity and other vital functions will be maintained in the event of postulated accidents.

The objectives of our review were to determine that the onsite Class IE alternating current power system will (1) have the required redundancy, meet the single failure criterion, and have the capacity, capability and reliability to supply power to all required safety loads, and (2) be designed in accordance with tne criteria set forth in Section 8.1 of this report.

The onsite alternating current power system will be comprised of the normal and Class 1E (essential) systems. The normal system, which will supply nonsafety loads, will be electrically and physically separated from the essential system and its normal power source will be from the main turbine generator which will supply power to the primary winding of the unit station service transformers via the 25 kilovolt isolated phase bus lead. The alternate (offsite) power source will be the transmission system network through the 69 kilovolt switchyard.

The onsite Class lE alternating current power system will be comprised of two redundant and independent 4160 volt Class lE distribution systems with their 480 volt load centers, the 120 volt uninterruptible alternating current power system and the standby power supplies (diesel generator units). The design will meet the single failure criterion requirements by providing two completely redundant and independent load groups supplied by independent power sources and associated distribution (division) systems. Three essential bus transformers will be provided and will constitute the preferred power source for the two essential 4160 volt buses; one of these transformers will be connected on the primary side but will be unloaded and used as a spare for either of the essential buses. The safety loads i for the facility will be distributed among the two Class lE 4100 volt buses in such a manner that the operation of any one division is all that will be required to meet minimum safety requirements.

The standby alternating current power will be supplied by independent diesel generator sets to be connected to each of the 4160 volt essential buses for each unit. The capacity of each diesel generator unit will be sufficient to meet the engineered safety features demand in the event of a postulated loss-of coolant accident coincident with loss of the preferred power source. The design of the standby power system will comply with the recommendations of Regulatory Guides 1.6 and 1.9. The standby diesel generators will be started (1) on a loss of voltage 8-3

to the respective 4160 volt bus to which each generator will be connected, (2) on a sustained degraded voltage condition, (3) by a safety injection signal, or (4) manually. If the preferred power supply is not available, the standby diesel generators will be automatically connected to the 4160 volt essential buses and sequentially loaded. Each diesel generator and its auxiliary systems will be housed separately in a seismic Category 1 installation. The suitability of each diesel generator to be selected as a standby power source will be confirmed by prototype qualification tests and preoperational tests. As a result of our review of the applicant's reliability qualification testing program for the diesel gener-ators, we conclude that the qualification program complics with the recommenda-tions of Branch Technical Position EICSB 2, " Diesel-Generator Reliability Qualification Testing," and, therefore, is acceptable.

Four redundant and independent divisions of 120 volt Class lE uninterruptible power system will supply control and instrument power to engineered safety features and other safety related systems. Each essential division will be fed from a battery and battery charger through a static inverter, or from a 480 volt essential bus through a regulating transformer when the inverter is out of service.

To satisfy the requirements of Criterion 18 of the General Design Criteria, the design of the alternating current power system will include provisions for periodi-cally testing and inspecting all safety-related equipment and systems to verify that the voltage profiles at the safety-related buses will be satisfactory for full lead and no load conditions on the system and for the range of grid voltages.

Conclusion Our review of onsite Class 1E power systems for the Sundesert plant covered single-line diagrams, preliminary physical layout drawings, functional logic diagrams and descriptive information for those systems and auxiliary systems that are vital to the proper operation of the onsite Class IE power system and its connected loads.

The review also included the applicant's design bases and their relation to the proposed design criteria for the onsite Class lE system and for the vital supporting systems and the applicant's analyses of the adequacy of those criteria and bases.

On the basis of our review, we conclude that the design of the onsite Class lE power system satisfies our requirements set forth in Section 8.1 of this report and, therefore, is acceptable.

8. 3. 2 Direct Current Power System The direct current power system will provide motive or control power to safety-related equipment. Batteries and battery chargers will be used as the power sources for the direct current power system, and inverters will be used to convert direct current from the direct current distribution system to alternating current instru-mentation power as required. The objectives of our review were to determine that 8-4

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the Class lE direct current system will (1) have the required redundancy, meet the single failure crM erion, and have the capacity, capability and reliability to supply power to all required safety loads, and (2) be designed in accordance with the criteria set forth in Section 8.1 of t.his report.

The direct current power system will be comprised of the Class lE direct current power system and the normal direct current power system. The normal direct current power system will consist of two 125 volt and one 250 volt direct current power systems to serve the nonsafety-related direct current loads.

The Class lE direct current power system will be comprised of four redundant and independent 125 volt batteries. Each battery system will include a battery, two battery chargers (one spare) and a main distribution panel. There will be no bus ties between any divisions and each division will be nhysically separated from the others, which complies with the recommendations of Regulatory Guide 1.6. Under

' normal conditions the battery chargers will provide primary direct current supply and the batteries will be floating on the bus. Each battery system will also supply power to an inverter which will convert the direct current power to 120 volt alternating current power for vital instrumentation and protection systems.

Each battery will be sized to carry all connected loads for two hours upon the loss of normal supply during postulated loss-of-coolant accident conditions or forced shutdown. Each battery will be located in its own ventilated room in a seismic Category I building. Each battery charger will have enough power output capacity l for the steady-state operation of the connected loads required during normal or l i

off-normal operation, while maintaining the battery in a fully charged state.

Battery chargers and distribution panels associated with a given battery will be located in close proximity to the battery in each division.

Each distribution panel will be equipped with a battery output / input ammeter, bus voltmeter, low bus voltage relay, critical low bus voltage relay, high bus voltage relay, low battery charger current relay, high battery charger current relay, high battery current relay, and bus voltage transducer. All these relays will be con-nected to annunciator points in the control room. The bus transducers will provide continuous bus voltage monitoring in the control room for each distribution panel.

To satisfy the requirements of Criterion 18 of the General Design Criteria, the design of Class lE direct current power system will include provisions for periodic inspection and testing to assess continuity of the system and the condition of its I components.

Conclusion Our review of the Class IE direct current power system for the Sundesert plant covered single-line diagrams and descriptive information for the system and 8-5

l 1

for those supporting systems that are essential to the operation of the direct current power system. The review also included the applicant's proposed design criteria and bases and its analyses of the adequacy of those criteria and bases.

On the basis of our review, we conclude that the design of Class lE direct current power system satisfies our requirements set forth in Section 8.1 of this report, including the single failure criterion, and, therefore, is acceptable.

8. 3. 3 Physical Independence of Electrical Power Systems The applicant has provided the criteria for physical separation of electrical equipment to preserve the independence of redundant equipment. In addition, the applicant states that the design of the electrical power systems will meet the recommendations of Regulatory Guide 1.75, " Physical Independence of Electric Systems."

We have reviewed the proposed design criteria for physical independence of elec-trical power systems a.d conclude that they meet the criteria outlined in Section 8.1 of this report and, therefore, are acceptable. As stated in Section 7.0 of this report, we have not completed our review of the proposed criteria for the physical independence of the instrumentation and control systems.

8.3.4 Electrical Penetrations The electrical penetration assemblics will be designed and constructed in accordance with IEEE Standard 317-1976, " Electric Penetration Assemblies in Containment Structures for Nuclear Power Generating Stations." In the course of our review, we requested the applicant to provide a description indicating how the overload protection of the penetration circuits will meet the recommendations of Regulatory Guide 1.63, "Electrice.1 Penetration Assemblies in Containment Structures for Light-Water-Cooled Nuclear Power Plants."

It vTs not clear from the description presented in response to our request whether the penetration circuits will be designed for overicad protection in accordance with IEEE Standard 279-1971, " Criteria for Protection Systems for Nuclear Power Generating Stations." Therefore, we require that the applicant meet the following requirements of IEEE Standard 279-1971 with regard to the protection of the electri-cal penetrations:

(1) The system shall, with precision and reliabi ity, automatically disconnect power to the penetration conductors when currents through the conductors exceed the preset limits; (2) All source and feeder breaker overload and short-circuit protection systems shall be qualified for the service environment, including seismic. The elsmic qualification for non-Class lE circuit breaker protectior, systems 8-6

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.should, as a minimum, assure that the protection systems will remain operable during an operating basis earthquake; (3) The circuit breaker protection system trip setpoints shall be compatible with the capability for test and calibration. Provi<'ons for test under simulated fault conditions should be provided; l (4) No Single failure shall cause excessive currents in the penetration conductors which will degrade the penetration seals; and t

(5) Signals for tripping source and feeder breakers shall be independent physi-cally separated and powered from separated sources.

Therefore, the above matter remains outstanding.

1 8.3.5 Environmental Qualification of Class lE Equipment The environmental qualification of Class 1E equipment for Sundesert is discussed in Section 3.11 of this report.

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l 9.0 AUXILIARY SYSTEMS l .

We have reviewed the design bases fur the auxiliary systems, including their safety-related objectives, and the manner in which these objectives will be achieved.

i The auxiliary systems evaluated in this section which are necessary for safe plant J

shutdown include: the service water system; the reactor plant component cooling water system; the ultimate heat sink; portions of the chemical and volume control system; the safety-related ventilation systems; and the diesel generator auxiliary

)

systems, 1 Our review of the proposed auxiliary systems necessary to assure safe handling of fuel and adequate cooling of the spent fuel included: the new and spent fuel y

storage facility; the fuel pool cooling and purification system; the fuel handling system; and portions of the fuel building ventilation system.

We have reviewed the vent and drain systems whose failure would not prevent safe )

shutdown but could indirectly be a potential source of radiological release to the environment.

The results of our evaluation of the above systems are presented in the following sections. The status of our review of the fire protection system is presented in j Section 9.5.1. Except for the ultimate heat sink (Section 9.2.3) and the fire q l

protection system (Section 9.5.1), none of the above auxiliary systems will be shared between the two units of the proposed facility.

1 We have also reviewed those auxiliary systems whose f ailure would neither prevent l 4

safe shutdown nor result in potential radioactive releases. These include the pressurizer relief tank, potable and sanitary water system, demineralized water makeup syste.m compressed air systems, and the non-safety related ventilation i systems, ine acceptability of these systems was based on our review which deter-mined that: (1) where the system interfaces or will connect to a seismic i l

i Category I system or component, seismic Category I isolation valves will be provided to physically separate the non-essential portions from the essential system or component, and (2) the f ailure of non-Seismic Category I systems, or portions of systems, will not precluce the operation of safety-related systems or components located in close proximity. Therefore, we find them to be acceptable.

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9,1 Fuel Storage and Handling 9.1.1 New and Spent Fuel Storage The new and spent fuel storage facility will be designed to store new and spent fuel in racks under water in the fuel pool. The fuel pool for each unit will provide storage capacity for one-third of a core of new fuel plus three and a half cores of spent fuel using a 16 inch center-to-center spacing. This spacing will be sufficient to maintain an effective multiplication' factor of less than 0.95 should the pool inadvertently be flooded with unborated water.

The fuel pool will be a stainless steel lined, reinforced concrete pool located in the seismic Category I portion of the fuel building. The new and spent fuel storage racks will be designed to provide protection against damage to the fuel and to prevent fuel assemblies from being stored in other than prescribed locations.

The new and spent fuel storage racks and the fuel pool will be designed to seismic Category I requirements, and will be protected from tornado missiles. The facility l

will be oesigned to prevent the cask handling crane from traveling over, or in the '

vicinity of, the fuel storage areas, thereby precluding damage to the stored fuel fii iiie cvant of a dropped cask. These design features are in conformance with the recommendations of Regulatory Guide 1.13. " Spent fuel Storage Facility Design Basis," and Regulatory Guide 1.29, " Seismic Design Classification," including the positions on seismic design, missile protection and compatibility with the handling of the fuel cask in the, fuel pool areas.

Based on our review, we conclude that the design criteria and bases for the new and spent fuel storage facility are in conformance with the requirements of Criterion 62 of the General Design Criteria ano the above mentioned regulatory guides. There-fore, we conclude that the proposed cesign of the new and spent fuel storage facility is acceptable.

9.1.2 Fuel Pool Cooling and Purification Systems l The fuel pool cooling and purification systems will be designed to maintain the

)

quality and clarity of the water in the fuel pool and to remove the decay heat gen-erated by the stored spent fuel assemblies.

E The fuel pool cooling system will consist of two separate 100 percent capacity 4

redundant trains and will be designed to Quality Group C and seismic Category I i

r.auirements in conformance with the recommendations of Regulatory Guide 1.13. 4 l

Each train will include one pump and one heat exchanger. The cooling water pumps l will be powered from separate essential buses. The safety-related reactor plant component cooling water system will provide cooling water to the fuel pool cooling a

heat exchangers. Makeup water to the pool will be provided by permanently installing a connection betweet, the primary grade water system and the refueling water storage 9-2

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tank via the fuel pool purification system ine seismic Categoq l s r . ke water system will serve as a backup supply of water via temporary connectiun; Inc. fuel pool cooling water system piping will be arranged su that t.he fuel pool cannut be .

Inadvertently drair.ed to uncover the unrad fue). ,

1 1

The applicant states that the fuel pool cooling s> stem will L2e designed to aU nmmodate the heat load f rom a total of 3-2/3 cores of normal ref ueling t>atches plus one fail core emergency unload. For the maximum nnrmal heat load conditions ( 3-2/3 cures of spent fuel in the fuel pool), one 0 4the t wo f uel pool couling t rains will maintain the pool temperature below 132 degrees Fahrenheit. During " abnormal" heat luad conditions (3-2/3 cores of normal ref ueling batcnes, p tus one f ull core emeruency unload), (1) one of the two f uel pool cooling trains will t>e capable of maint a i n i ng the pool water temperat ure below It ' *gr( e Fahrenhei t , and ( 2) two oper at ing trains will be able to maintain the pool water temperature Deluw ill degrees fahrenheit.

These pool water temperatures are acceptatile even with the loss of one fuel pool cooling train. 4 I

Based on our review, we conclude that *.he design criteria and nases f or the fuel pool cooling and purification ~ systems . ire in conf ormance with t he requirements o' Criterion 61 of the General Design Criter a and the recommendations of Reaulatory i

Guide 1.13, including the recommendatinns on the seismic design, missile protection j and availability of the assured makeup systems. We, therefore, find the proposed design of the fuel pool cooling and purification system to be acceptable.

11.3 fuel Handling System The f uel handling system in conjunct ion with the f uel storage area will provide a safe and effective means of transporting and handling fuel from the time it reaches the plant in an unirradiated condition until it leaves the plant after post-irradiation cooling.

The spent fuel cask loading area will be located away from the spent fuel pool and will be connected to the fuel pool by a transfer canal. The loading area will be separated from the fuel pool by a sufficient distance such that an inadvertent drop of a spent fuel cask over the loading area cannot damage the spent fuel pool.

l Unacceptable damage to stored fuel, due to a spent fuel cask drop, will also be prevented by plant physical arrangement. to limit the travel of the spent fuel cask

! to an area which will contain no sa#ety-related equipment or stored f uel, eacept

{ f or the redundant saf ety-related fuel pool cooling piping which will be routed i

beneath the path of travel of the cask. However, in order to provide protection for the piping, the portions of the floor above the fuel pooling cooling piping will be designed to safely withstanii the loads resulting from a postulated cask drop.

9-3

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The applicant has concluded that, in the event of an accidental drop of the spent

! fuel cask, unacceptable damage to the fuel within the cask would not occur since i

[ the height of the drop will be limited to a maximum of 30 feet. We concur with the

^

applicant that the proposed design of the spent fuel task would not result in unac-l ceptable damage to the fuel if the accidental drop can be limited to a maximum of

. 30 feet. However, since this maximum drop height is based on administrative controls i to be placed on the cask along the cask travel path, we require that physical f

j restrictions be provided to the cask lifting devices, such as mechanical stops or a

electrical interlocks, to assure that the maximum potential for a cask drop is 30 l feet. Therefore, this matter remains outstanding.

i j We have also reviewed the proposed design of the reactor polar crane which will be

{ used inside containment to rerncve and replace the reactor vessel head for refueling operations. The applicant does not propose to provide a single failure proof f

j reactor polar crane. The applicant has also not provided an analysis of a postulated '

l accidental drop of the reactor vessel head by the crane to demonstrate that the consequences of such a drop are acceptable. As a result, we require that the i reactor polar crane be designed to meet the single failure criterion un'ess the j applicant can demonstrate by analysis that the consequences of a postulated drop of i the reactor vessel head by the crane are acceptable. Therefore, this matter remains f outstanding.

4 4 Based on our review, and subject to the satisf actory resolution of the above matters 1

} relating to (1) physical restrict' ions on the spent fuel cask lifting devices and l (2) the single failure criterion for the reactor polar crane, we conclude that the l design criteria and bases for the fuel handling system conform to the requirements i of Criterion 61 of the General Desion Criteria and the recommendations of Regulatory 4

j Guide 1.13 and are acceptable.

I j- 9.2 Water Systems

! 9.2.1 Service Water Systa 1

i e The service water system will supply cooling water to the reactor plant auxiliaries.

1 t

s The reactor plant auxiliaries to be cooled by the service water system :.re the I i

j reactor plant component cooling water heat exchangers, standby diesei generator l i

i cooling water heat exchangers, control building water chiller condensers, and l various unit coolers for safety related systems. The service water system will f also be capable of supplying emergency backup water to the auxiliary feedwater l system, spent fuel pool makeup, and the seismically designed portion of the fire protection system.

] The service water system will consist of two independent safety-related trains.

, Each train will contain one half capacny normal service water pump and one full capacity standby service water pump which will pump water from an individual cooling tower basin through the plant components and return the water to the cooling tower s

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to dissipate the heat. The system will be designed so that one of the two redundTnt trains with its full capacity standby service water pump in operation during a postulated design basis accident will meet the minimum engineered safety feature l

recoirements. The ennling t owers whic h .ill serve as t he ult imate heat sink for t.he seriime wp .r system, are discussed in Section 9.2.3 of this report.

Essential p'rt:ons of the system will be designed to Quality Group C and seismic Category I requirements, and will be des:gned to withstand adverse environmental occurrences, such as tornadoes and floods, Each train will be powereo from a separate essentiai alternating current bus except that the normal service waterr pumps will be powered from non-essential power supplies since they are not required to achieve a safe shutdown,

)

Baseo on our review, we conclude that design criteria and bases for ti* service water system are in conformance witn the requirements of Criterion 44 of the Gene,JI

Design Criteria regarding the ability to transfer heat from safety-related components to the ultimate heat sink and the ability to meet the single failure criterion. We also cone.lude that they are in conformance with the requirements of Criteria 45 and 46 of the General Design Criteria regarding the provisions in the system design for per iodic tests and inspections, including functional testing and confirmat;on of heat transfer Capabilities, We, therefore, Conclude that the proposed design of the service water system is acceptaDie.

9. E 2 Reactor Plant Component Cesling Water System J

lhe reactor plant component cooling water system will provide an intermediate cooling loop for removing heat f rom reactor plant aux.iliary systems and transf erring it to the service water system. The system will consist of two independent closed loop flow paths, Each flow path will contain one half capacity normal pump and one f ull capacity standby pump which will pumo water through the r eactor plant component l cooling water system heat exchangers, where the system heat load will be transferred to the service water system. The cooled water will then be circulated to the plant

omponents and will be returned to the pump suction. The system will be designed l 0 that one of the two redundant ficw patos with its full capacity standby pump in operation during a postulated de>! 9n basis accident will meet tne min..num engineereo safety feature requirements.

Essential portions of the reactor plant component cooling water system w be designed to Quality Group C and seismic Category I requirements, and wil' 'e pratected 1 to withstand adverse environmental occurrences, such as tornadoes and fic 25. Each train will be powered from a separate essential alternating current bus. The l l

non-saf ety related portions of the systen supplies will be automatically isolated during a postulated design basis accident.

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The proposed design of the reactor plant component cooling water system will provior a single supply and return line from train A of the system to the Coolers of one reactor coolant pump and another single supply and return line from the train B of the system to the rnnlark nf t he remainiraj two nf t he t hree rPactor Toolant pumps.

[ach of these lines will contain motor-operated valves for containment isolation, lhe motors, seals, and bearings of the re3Clor coolant pumps will require continuous cooling for pump operation. Inadvertent failure or closure of any one of the above motor operated valves would terminate the component cooling water flow to one or two reactor coolant pumps, thus, potentially leading to fuel damage or breach of the primary system barrier which could result from a pump locked rotor or pump seal failure.

We, therefore, required the applicant to nodify this portion of the reactor plant component cooling water system so that the following criteria will be met:

(1) A single failure in the component cooling water system shall not result in f uel damage or damage to the reactor coolant system pressure boundary caused by an extended (30 minute) loss of cooling to the reactor coolant pumps. A single failure includes operator error, spurious actuation of motor-operated valves, and loss of component cooling water pumps.

(2) A moderate energy leakage crack or an accident that is iritiated from a failure in the component cooling watar piping shall not result in excessive fuel I

damage or a breech of the reactor coolant system pressure boundary when an extended (30 minutes) loss of cooling to the reactor coolant pumps occurs. A single active failure shall be considered when evaluating the consequences of this accident. Moderate energy leakage cracks should be determined in accordance with the guidelines of Branch Technical Position APL5B 3-1, " Protection Against Postulated failures in a fluid System Cutside Containment."

]n response to our above requirement, the applicant proposes that it will either (1) requalify the reactoc 'lant pumps for extended optration without component cooling water supply, and provide saf ety grade instrumentation to detect the loss of component cooling water to the reactor coolant pumps and to alert the operator in the control room, or (2) modify the design of the reactor component cooling water system to assure that excessive fuel damage or a breach of the reactor coolant pressure boundary will not occur as a result of a single active f ailure, a moderate energy line break, or an accident initiated from a failure of the component cooling water system.

We have evaluated the applicar.'_'s proposai and reouire the following modifications for the proposal to be acceptable:

(1) The applicant has not specificed in its first option the length of time that the reactor coolant pump will be requalified without component cooling water.

We require that the requalificaticn be for 30 minutes.

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(2) for the second option, we require that a single active failure in the component i cooling water system shall not result in fuel damage or damage to the reactor coolant system boundary, as originally stated in our position. We find the remaining portion of the secnnn nptinn to be consistent with our position and, therefore, acceptable.

Therefore, the above matter remains outstanding.

Subject to the satisf actory resolution of the at ove matter, we conclude inat the design criteria and bases for the reactor plant component cooling water system conform to the requirements of Criteria 44, 45 and 46 of the General Design Criteria and are acceptable.

9.2.3 Ultimate heat Sink The" ultimate heat sink will be desigr.ed to remove and dissipate the heat from the service water system for all mooes of normal operation and tc permit the safe shutdown of the plant under postulated accident conditions even with the loss of offsite power. It will consist of two 100 percent capacity wet mechanical draf t cooling towers with basins and two pumphouses all to be shared by the two reactor units in the facility. Each cooling tower will be designed to dissipate the maximum cooling load to permit the safe shutoown of one unit following a postulated design basis accident in that unit and the concurrent cooldown of the second unit. Each pumphouse will contain the service water pumps from one of the redundant service water trains. Our evaluation of the service water system is discussed in Section 9.2.1 of this report.

The ultimate heat sink cooling tower basins will be below grade and each will contain sufficient ' water for 30 days of operation without the need for makeup water. The cooling tower structure, tower basins, and all tower components will be designed to seismic Category I requirements and will be protected against tornadoes and tornado generated missiles, in conformance with the recommendations of Regulatory Guide 1.27. " Ultimate Heat Sink for Nuclear Power Plants." The cooling tower fans will be powereJ by the essential buses.

The applicant has submitted the results of an analysis which demonstrate that the ultimate heat sink will be capable of dissipating the heat from both units for 30 days in the event of an accident in one unit and the simultaneous shutdown and cooldown of the other unit. The analysis takes into account fission product decay heat, rejected heat f rom plant auxiliary systems, and plant sensible heat. Additianal discussion on the adequacy of the ultimate heat sink performance is presented in Section 2.4.3 of this report.

Based on our review, we conclude that the design criteria and bases for the ultimate heat sink conform with the recommendations of Regulatory Guide 1.27. Therefore, we conclude that they are acceptable.

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i j- 9.2.4 Condensate Storage f acilities I

j The condensate storage facilities will consist of the demineralized water, auxiliary

! feedwater, and condensate storage tW s. Only the auviliary feedwater storage tank, and associated piping and valves, is safety-related since it will serve as a

{ reliable water source to the auxiliar y f eedwater system for safe plant shutdown.

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i The auxiliary feedwater storage tank will be located in the annulus building. It will be designed to Quality Group C and seismic Category I requirements and will be protected from tornadoes, missiles, and flooding. The auxiliary feedwater storage tank for each unit of the plant will be sized to provide sufficient capacity to

satisfy the requirements of the aLxiliary feedwater system. This will include

{ maintaining the plant at hot shutoown for at least two hours and cooling down the l plant for an additional five hours, at a rate of 50 degrees Fahrenheit per hour, to l approximately 350 degrees Fahrennelt at which time the residual heat removal system j will be operated. Safety grade levet indication will be provided for the auxiliary i feedwater storage tank and high/ low level alarms for the tank will be provided in a

j the control room and the remote shutdown area. During normal operation, low levels f in the auxiliary feedwater storage tank will be made up by water from the nonsafety

( grnde condens&tf! storage tank.

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j Based on our review, we conclude that the condensate storage facilitias will be l designed to meet the requirements of Section 9.2.6 of the Standard Review Pla1 and, j therefore, the proposed design is acceptable.

9. 3 process Auxiliaries j 9.3.1 Compressed Air Systems i

Ihe compressed air systems will be designed to provide both instrument air and j service air to the plant. Instrument air reciuired inside the containment wi'.1 be l provided by a separate instrument air system to be located inside co6tainment, with a connection to the plant instrument and service air system for backup.

i j The compressed air n. + ens are classified as non-safety related except the portions

, that will penetrate containment walls , including isolation valves, which will DO

] designed to Quality Group B and seismic Category I requirements. All air operatto valves in safety-related plant sys ten's will be designed to f ail to a safe positicn

$ upon loss of instrument air.

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j Based on our review, we find that the design criteria and bases for the compressed l air systems are in conformance with the recommendations of Regulatory Guide 1.26, i " Quality Group Classifications and Standards for Water , Steam , and Radioactive-i Waste-Containing Components of Nuclear Power Plants," and Regulatory Guide 1.29, "Seirmic lesign Classification," with regard to quality group classification and sei,mic design of the safety-related portions of the systems. We also find that 9-8 r-.- -- g-- - -

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the systems will be designed to protect the safety function of plant safety-related l

l systems in the event of a 1sss of air supp)y. Therefore, we find the proposed design of the compressed air systems to be acceptable.

L 9.3.2 Process Sampling System de process sampling system will be designed to provide representative samples of radioactive, as well as non-radioactive, fluid streams from Systems and components throughout the plant for chemical and radiochemical analysis. The locations for obtaining representative samples from these systems and components are listed in Table 9.1. The process sampling system will consist of piping, fittings, isolation and throttling valves, sample ccolers, sample flasks, sample sinks, gaseous sample pumps, and instrumentation. The seismic design and quality group classification of sampling lines and components will appropriately conform to the classification of the system to which each sampling line and component is connected (acceptable classification of systems and components is discussed in Regulatory Guides '.20 And 1.29).

TABLE 9.1 PROCESS SYSTEM SAMPLE LOCATIONS Nuclear Steam Supply System Reactor Building Sumps Demineralized Water Makeup System Makeup and Purificatio,, System Chemical Addition and Boron Recovery System Service Water System Core Flood Tanks Auxiliary Boiler Reactor Coolant Drain Tank Steam Generator Blowdown Component Cooling Water System Spent fuel Pool Our review of the proposed process sampling system included the provisions for f

sampling all principal fluid process streams associated with plant operation and the location of sampling points, as shown on piping and instrumentation diagrams. j As a result of our review, we have determined that the applicant's proposed design for the process sampling system is in conformance with applicable regulatory guides, as well as with industry standards. Therefore, we conclude that the proposed I

system is acceptable.

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. L 3. 3 Vent and Drain Systems The "ent and drain systems will consist of the aerated vent and drain system, and the gaseous vent and drain system. The gaseous vent and drain system will collect fluids from areas housing components in radioactive systems which will contain hydrogen. The aerated vent and drain system will collect fluids from floor and 1

% uipment drains and vents from equipment and tanks which will contain degassed potentially radioactlve liquids. All radioactively contaminated vents and drains will be routed to the radioactive liquid and gaseous waste systems, which are 9-9 )

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discussed in Section 11.2 of this rejiort, for cleanup of radioactive effluents.

The non-racioactive vents and drains will be directly discharged to the atmosphere and to the floor drainage system.

Sun.ps for drain collection will te provided in the containment, annulus building.

engineered safety features area, solid waste building, fuel building, and turbine building, fach sump will be provided with two full capacity sump pumps and level alarms. The system will be designed to preclude inadvertent transfer of contaminated fluids to a non contaminated drainage system or flooding of equipment in safety-related areas due to backflow from the drainage systems. The vent and d*ain systems are classified as non-safety related en ept the portions that penetrate containment walls, including the isolation valves, which will be designed to Quality Group B and seismic Category I requirements.

i Based on our review, we conclude that the design criteria and bases for tha vent and drain systems are suf ficient to protect safety related areas and components from flooding and to prevent the inadvertent release of radioactive fluids to the environment. Therefore, we conclude that the design Criteria and bases for these systems are acceptable. i 9.3.4 Cnemical and Volume Control System The chemical and volume control system will be designed to control and maintain reactor coolant inventory in the primary system and also to control the boron concentration in the reactor coolant through the pi.; cess of makeup and letdown. i The system will purify the primary coolant by demineralitation. Portions of this

{

system will also supply high pressure injection of borated water into the reactor i I

coolant system for emergency boration. Centrifugal charging pumps will serve as safety injection pumps when the emergency core cooling system is required to function.

l The emergency core cooling system is discussed in Section 6.3 of this report.

l The chemical and volume control system will also maintain seal water injection flow to the reactor coolant pumps and provide means for filling, draining, and pressure

{

testing the reactor coolant system. The portions of the chemical and volume control

)

system required for safe shutdown of the reactor will be designed as seismic Category 1, l will meet the single failure criterion, and will be powered from essential buses. l Based on our review of the design criteria and bases for the chemical and volume control system, we conclude that the system will be designed to meet its intended safety function and, therefore, is acceptable.

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9.4 Air Conditioning, Heating, Cooling and Ventilation Systems Heating, ventilating, and air conditioning systems will be provided in the control building, fuel building, annulus building, standby ciesel generator building, service water pumphouse, solid waste building, and turbine building, lhe ventilation systems in the control building, annulus building, service water pumphouse and I standby diesel generator building will be safety-related. The outdoor air isolation dampers of the fuel building ventilatinn system will be Quality Group C. The remainder of the ventilation systems, although not safety-related, will either provide a suitable environment for personnel and equipment, or prevent the spread or release of airborne radioactive materials to the atmosphere.

9.4.1 Control Building Heating, Ventilation and Air Conditioning Systems The control building heating, ventilation and air conditioning systems will be designed to remove heat, ventilate and maintain personnel comfort in the building and the electrical tunnels. The contrni building heating, ventilation and air conditioning systems will include the following subsystems: the control room envelope area air conditioning system; the essential switchgear area air conditioning system; the control building chiller room air conditioning system; and the chilled water system. Each of the above systems will have 100 percent redundancy and will be designed to seismic Category I requirements. The outside air intakes will also be tornado missile protected.

The control room envelope area air conditioning system will be designed to maintain the control roc- invelope area within the environmental limits required for operation of plant controls and for uninterrupted safe occupancy of required manned areas during all operational modes, including postulated design basis accident conditions.

The system will be designed to maintain the control room under positive pressure.

Redundant radiation monitors and chlorine gas detectors with alarms in the control room will monitor the outside air supply. A safety injection signal, a high radiation signal or a chlorine gas signal will automatically isolate the outside air supply to the control room, isolate the exhaust air from the outside atmosphere and place the system in the recirculation mode and start the essential standby filtration j units.

The essential switchgear area air conditioning system will consist of two full capacity air conditioning units. The system will supply air to the essential switchgear rooms, and to each of the four battery rooms and the respective battery I charger rooms. Individual full' capacity exhaust fans will be provided for each of the four battery rooms to assure that the maximum hydrogen concentration inside the f battery room will be below two percent.

The control building chilled water system will consist of two 100 percent capacity centrifugal compressor water chillers. Each chiller will supply essential chilled 9-11

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f water to one of the redundant air conditioning units of the control building heating, cooling and air conditioning systems. The seismic Category I service water system j will provide cooling water to remove heat from the condensers of the chillers.

!. Each chiller room will have its own air conditioning unit and exhaust fan for cooling and ventilating the equipment area.

t 3

} We have reviewed the design criteria and bases for the control building heating, l cooling and air conditioning systems and conclude that they meet the requirements 1 l 5

4 set forth in Criterion 19 of the General Design Criteria, with regards to the

capability to operate the plant from the control room during normal and postulated accident conditions, and that they meet the single failure criterion. We also f conclude that the control building chilled water system is in conformance with Criterion 44 of the General Design Criteria regarding the ability to transfer heat from safety related components. Therefore, we conclude that the proposed design of j these systems is acceptable.

f

{ 9.4.2 Fuel Building Ventilation System j The function of the fuel building ventilation system will be to maintain a suitable

] environment for equipment operation and to limit potential radioactive releases to l the atmosphere during normal operation and postulated fuel handling accident conditions.

l 4 l

! i The non-safety portion of the a stem will normally provide heating, ventilation and 4 j air conditioning functions to the fuel building and will consist of two half capacity

} air handling units and two half capacity exhaust fans. During a postulated fuel handling accident or in the event of high airborne radiation levels within the fuel building, redundant Class lE radiation monitors, to be located at the inlet of the normal exhaust fans, will automatically close the outside air isolation dampers and divert the exhaust air through two redundant fuel building filtration units prior f to discharge to the atmosphere.

E i The portions of the fuel building ventilation system which will contain the fuel building filtration units and the outside air isolation dampers are safety related i and will be designed to Quality Group C standards. However, they will not be l I

designed to seismic Category I requirements. Also, the safety-related fuel building l filtration units will be located in the non-seismic Category I portion of the fuel !

building. Therefore, we require that the applicant modify the design criteria and a

bases for the fuel building filtration system to provide seismic Category I fuel building filtration units and outside air isolation dampers for this system in conformance with the recommendations of Regulatory Guide 1,29. (See Sections 3.2.1, 6.5.3 and 15.5.2 of this report for additional discussion on the fuel building filtration system.) As stated in Section 3.2.1 of this report, we also require that all portions of the fuel building be designed as seismic Category I.

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$ubject to the satisfactory resolution of the above matter, we conclude that the design criteria and bases for the fuel building ventilation system meet the recommen-dations of Regulatory Guide 1.29 and are acceptable.

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9.4.3 Annulus Building Ventilation System r

The function of ae annulus building ventilation system will be to maintain a suitable environment for equipment operation and to limit potential radioactive releases to the atmosphere during normal operation and postulated design basis accident conditions.

I 1

The non-safety portions of the system will normally provide heating, ventilation and air conditioning functions to the annulus building and will consist of two half capacity air handling units and two half capacity exhaust fans. The radiation monitors in the exhaust line will alarm in the control room in the event of high airborne radiation levels in the building. Under those conditions, the building exhaust air will be diverted manually through one of the two annulus building filtration units prior to discharge to the atmosphere. In the event of a postulated design basis accident, a safety injection signal will automatically isolate the normal exhaust isolation dampers and divert the exhaust air from the engineered safety feature areas, through one of the annulus building filtration units, to l l assure a negative pressure in these areas. Unit coolers will be provided for each j engineered safety feature area. The exhaust air isolation dampers, the annulus building filtration units, and the unit coolers in safety-related areas will be designed to Quality Group C and as seismic Category I, in conformance with the l recommendations of Regulatory Guide 1.29, and will be powered from essential buses.

The safety-related reactor plant component cooling water system will supply cooling water to the engineered safety feature area unit coolers.

Based on our review of the design criteria and bases for the annul n building ventilation system, we conclud: that the proposed system will be designed to meet l

the single failure criterion, conforms to the guidelines of Rego atory Guide 1.29 l and, therefore, is acceptable.

I l 9.4,4 Standby Diesel Gen 9rator Building Ventilation System The function of the standby diesel generator building ventilation system will be 10 maintain a suitable environment for equipment operation during normal and post-accident conditions. The safety-related portions of the system will consist of two half capacity supply air fans, together with exhaust dampers for each of the two diesel j

generator aoms, and one full capacity air handling unit in each of the two electrical equipment "ooms. This safety-related sub-system will be designed to Quality Group L and as 'eismic Category I, in conformance with the recommendations of Regulatory Guide 1. 9, and will'be powered from an essential bus associated with the diesel it will se ie. The safety-related service water system will provide cooling water to the el .trical equipment room air handling units.

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Based on our review of the design criteria and bases for the standby diesel generator building ventilation system, we conclude that the proposed system will be designed j

to meet the single failure criterion, conforms to the guidelines of Regulatory

] Guide 1.29 and, therefore, is acceptable.

9.4.5 Service Water Pumphouse Ventilation System i

4 The function of the s evice water pumphoure ventilation system will be to maintain j

a suitable environm,.;. for equipment operation during normal and post-accident

} conditions. The stem will consist of one full capacity air handling unit and one

} exhaust air damper for each of the two service water pumphouses. The system is safety related, will be designed to Quality Group C and as seismic Category 1, in f conformance with the recommendations of Regulatory Guide 1.29, and will be powered j from essential buses. The safety-related service water system will provide cooling j water to the air handling units. ;

J l

.a Based on our review of the design criteria and bases for the service water pumphouse ventilation system, we conclude that the proposed system will be designed to meet

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the single failure criterion, conforms to the guidelines of Regulatory Guide 1.29 '

} and, therefore, is acceptable.

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5 9.5 Other Auxiliary Systems 9.5.1 Fire Prntection System '

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j Our current guidelines for fire protection systems are set forth in Branch Technical ,

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Position APC58 9.5-1,' " Guidelines for fire Protection for Nuclear Power Plants."

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During our review of the fire protection system, we requested the applicant to j {

conduct a reevaluation of the proposed Sundesert fire protection provisions, and to i

compare the provisions, in detail, with the guidelines in Branch Technical Position APC58 9.5-1.

I 5

The applicant provided additional information in response to our request, but our f review of this information has not been completed. Therefore, this matter remains outstanding, i

} 9.5.2 Communication System 1 i i The communication system will include all components for intra plant communications and offsite communications with the plant.

The scope of our review of the communications system for the Sundesert plant included verification that offsite equipment will be capable of providing for notification of personnel and for implementation of evacuation procedures, and verification that onsite communications will be adequate in the event of an emergency. As a result of our review, we have determined that the applicant's proposed design criteria and 9-14

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Dases for the communication system and the requirements for all conditions of plant operation are acceptable. l The basis for our acceptance has been conformance of the applicant's proposed j design criteria and design bases for the communication system and necessary auxiliary l supporting systems to applicable staff positions presented in Section 9.5.2 of the l Standard Review Plan, and the ability of the system to provide effective communica-I tions between plant personnel in all v"tal areas for a full spectrum of accident or incident conditions and under maximum potential noise levels.

Therefore, we conclude that the proposed design of the communications system is acceptable.

9.5.3 Lighting Systert The lighting system will include all components necessary to provide adequate lighting during both emergency and normal operating conditions.

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The scope of our review of the lighting system for the Sundesert plant included an assessment of the adequacy of the energency power sources and verification that the design will adequately consider accident conditions. As a result of our review, we have determined that the applicant's proposed design criteria and design bases for the lighting system and the requirements for adequate lighting during accident conditions are acceptable.

The basis for our acceptance has been conformance of the applicant's proposed j design and design criteria for the emerger,cy lighting system and necessary auxiliary l supporting systems to applicable staff positions presented in Section 9.5.3 of the Standard Review Plan.

Therefore, we conclude that the proposed design of the lighting system is acceptable.

9.5.4 Diesel Generator Auxiliary Systems The diesel generator auxiliary systems will include the diesel generator fuel oil storage and transfer system, the diesel generator cooling water system, the diesel gene or air sivling system, the diesel generator lubrication system, and the diesti generator combustion air intake and exhaust system.

l Diesel Generator Fuel Oil Storage and Iransfer System The diesel generator fuel oil storage and transfer system will be designed to provide fuel oil storage and transfer capability to allow operation of each diesel generator for at least seven days.

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i The fuel oil system will consist of two separate and independent trains, one for each diesel generator. The principal components of each train will include one j main fuel oil storage tank, two fuel oil transfer pumps, a day tank with 1100 l gallons of fuel oli storage capacity, and the associated piping and valves required !

j to conrect the equipment. The underground storage tanks, filters, valves, and i

j piping will be designed to seismic Category I. Quality Groe;, C and Safety Class 3 i requirements, will be protected from tornada missiles, and will be designed to resist the loading imposed by the probable maximum flood. All other components _of l the fuel oil system will be located inside the diesel building and will be designed to seismic Category I and Safety Class 3 requirements, j i

l Based on our review of the design criteria and bases for the diesel generator fuel

! oil storage and transfer system, we conclude that the proposed system design will have adequate capacity, will meet its designated safety function, conforms with our I 1

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l acceptance criteria in Section 9.5.4 of the Standard Review Plan and, therefore, is f acceptable, i

1 l Ciesel Generator Cooling Water System

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1 j The diesel generator cooling water system will be an integral part of the diesel 1 1

generator and will be designed to maintain the temperature of the diesel within a safe operating range. The cooling water system for each diesel generator will 1

consist of one expansion tank, one full capacity generator-driven jacket water pump, une alternating current motor-driven circulating heater pump, an electric immersion heater, a thermostatic valve, a jacket water cooler, a lubricating oil l J

cooler and associated piping and valves to interconnect the equipment.

l The heat of the closed cooling water system will be rejected to the service water j system through the jacket water cooler. The expansion water tank will be sized with sufficient capacity to allow for leakage from pump shaft seals and valve stems, and f rom other leakage sources, during a period of seven days of. continuous diesel generator operation at maximum rated load. During cold weather, the electric

immersion heater will maintain the cooling water at the normal diesel generator starting temperature to increase the generator's reliability for starting on the first try. The lubricating air cooler will use the ccoling water system as a cooling medium f or the lubrication oil system described below. The components of the diesel generator cooling water system necescary to cool the diesel generator i during operation and all piping and valves in the system will be designed to seismic

, Category I and Safety Class 3 requirements.

4 Based on our review of the design bases and criteria for the diesel generator ccaling water system, we conclude that the proposed system design will have adeouate j capacity, will meet its designated safety function, conforms with our acceptance

criteria in Section 9,5.5 of the Standard Review Plan and, therefore, is acceptable.

i 9-16 1

Diesel Generator Air Starting System The diesel generator air starting system for each diesel will be separate and inde-l

' pendent from the other air starting system. Each air starting system will consist of two air compressors, two air dryers, two air receivers, two engine cranking devices and associated piping and valves to interconnect the equipment. The air receivers for each diesel will be capable of holding sufficient air for five starts of a cold engine, and will be connected to the engine cranking devices through a filter and solenoid operated starting valves.

Components of the diesel generator air starting system needed to start the emergency diesels, namely the air receivers, cranking devices, and associated piping and valves, will be designed to seismic Category I and Safety Class 3 requirements, Based on our review of the design bases and criteria for the emergency diesel starting system, we conclude that the proposed system design will have adequate capacity, will meet its designated safety function, conforms with our acceptance l criteria in Section 9.5.6 of the Standetrd Review Plan and, therefore, is acceptable.

Diesel Generator Lubrication System The diesel generator lubrication system will be an integral part of the diesel gen-erator. The system will be designed to lubricate and to circulate lube oil through the engine for cooling when the engine is operating.

The lubrication system for each diesel generator wil, be separate and independent of the other lubrication system and will consist of one full capacity engine-driven lubricating oil pump, a lubricating oil cooler, lubricating oil filters and strainer.

and associated piping and valves to interconnect the equipment. An independent circulating loop containing an alternating current motor-driven circulating heater pump and an electric heater will supply warm lubricating oil to the engine bearing 5 and other necessary components, when the generator is idle, to improve the reliability of the generator starting on the first try. All components of the diesel generator i

lubrication system, including piping and valves, will be designed to seismic Category 1 and Safety Class 3 requirements. l Based on our review of the design bases and criteria for the emergency diesel I engine lut rication system, we conclude that the proposed system design will have adequate capacity, will meet its designated safety functio.., conforms with our acceptance criteria in Section 9,5.7 oi the Standard Review Plan and, therefore, is acceptable.

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t Diesel Generator Combustion Air Intake and Exhaust System i

$ The diesel generator combustion air intake and exhaust system will be designed to i

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< supply fresh air for combustion to the eme gency diesels and to disDose of the l engine exhaust to the atmosphere.

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The combustion air intake and exhaust system for each diesel will be separate and 1

independent from the other intake and exhaust system and will consist of an air l intake filter, air intake silencer, expansion joints. exhaust silencer, and cssociated l piping and valves to interconnect the eouipment. The components of the diesel generator combustion air intake and exhaust system, including piping and valves, k will be designed to seismic Category I and Safety Class 3 requirements.

i i

. The proposed arrangement and locatinn of the combustion air intake and exhaust I j structures will preclude the possibilit.y of fire extinguishing agents and other unwanted gases from being drawn into the air intakes and will also preclude the possibility of exhaust gas recirculation with the air intakes.

Based on our review of the design base:, and criteria for the diesel generator com-l bustion air intake and exhaust system, we conclude that the proposed system desigo l will have adequate capacity, will meet its designated safety function, conforms with our acceptance criteria in Section 9.5.8 of the Standard Review Plan and, i

x therefore, is acceptable.

t i

j Conclusion j As a result of our review of the design criteria and bases for the diesel generator j auxiliary systems, we conclude that thei proposed designs assure that these systems l

will meet their designated safety functions, will have the needed capacity and,

{ therefore, are acceptable.

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a 9-18

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10.0 STEAM AND r0yFR,[2tjvER5!0N SYSTEM 10.1 Summary Description The steam and power conversion system #or each unit will be of conventional design, similar to those of previously appruved piants using the Westinghouse nuclear steam supply system. The steam and power conversion system will be designed to remose heat f rom the reactor coolant system by generating steam in three steam generators and by converting the heat energy in the steam to electrical energy by means of a turbine generator. A condenser will transfer unusable heat in the cycle to the circulating water system and the heat will be rejected to the atmosphere tnrough cooling towers. The entire system will be designed for the maximum design heat generation from the nuclear steam supply rystem.

The system will be designed with a load following capability. In the event of a loss of external load, the system will be capable of rejecting the heat energy.

without reactor or turbine trip, through the turbine bypass system directly to the condenser and through atmospheric dump valves. In the event that the condenser is not available, the reactor will automatically trip and the excess heat energy will be dissipated through the main steam safety valves, if required.

10.2 Turbine-Generator The proposed Westinghouse turbine generator for each unit will consist of a tandem arrangement, on a single shaft, of a double-flow, high pressure turbine and three identical double-flow, low pressure turbires driving a direct-coupled generator at 1800 revolutions per minute. An electro-bydraulic control system will control the speed and synchroni2ation of the turbine generator by positioning the steam control valves to regulate the flow of steam to tre turbine.

The turbine control system will be designed to trip the turbine under the following conditions: turbine overspeed, condent.er high pressure, excessive thrust bearing wear, low bearing oil pressure, low fluid pressure in the electro-hydraulic control system, direct current power failure to the electro-hydraulic control system, Safety injection signal, steam generator high-high level, low pressure turbine j

exhaust causing high temperature, reactor trip, generator breaker trip, and manual j

turbine trip.

The Overspeed protection will be provided by three redundant and diverse systems.

j electro-hydraulic control system will rapidly close the main control valves and the intercept valves when the turbine speed reaches 103 percent of its rated speed.

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in addition to the overspeed protection control and trip functions to be provided by the electro-hydraulic control system, a diverse method of tripping will be provided by an independent.overf requency relay which will trip the turbine generator when the gener9 tor f reque1Cy reachas !!! percent Of its rated value. As a backup to these electrical overspeed control and trip functions, a mechanical overspeed protection system will be provided, which will close all turbine valves when the turoine speed reaches ill percent of its rated value.

Based on our review of the proposed turbine generator overspeed protection system design, we conclude that the system can perform its designated functions and, therefore, is acceptable.

10.3 Main Steam Supply System l The steam produced in the three steam gererators will be routed to the high pressure turbine by three main steam lines up t o t he common header. Each main steam line will contain one atmospheric relief valve, a set of safety valves and one main steam isolation valve. The portions of the main steam lines from the steam genera-tors, through the containment and up to and including the main steam isolation valves, will be Safety Class 2 and seismic Category 1.

The main steam isolation valves will be designed to c!ose in five seconds upon receipt of a main steam isolation signal. The valves will be designed to stoo .

I steam flow from either direction. In the event that a steam line break occurs i upstream of a main steam isolation valve concurrent with a failure to close of a main steam isolation valve on an unaffected steam generator, blowdown of the unaffected steam generator will be preverted by the closure of the non-seismic Category I turbine stop walves and turbine bypass valves. Based on our review of these non-seismic Category I valves, as presented in NUREG-0138, "Staf f Discussion of fifteen Technical issues listed in Attachment to Novembat 3, 1976 Memorandum from Director, NRR, to NRR Staff," we conclude that the design and past perf ormance of these valves will De compatible with the accident conditions for which they will be called upon to function and will serve as an acceptable backup to a single failure in the safety grade main steam isolation valve, i

Seismic Category I saf ety valves and a power-operated atmospheric relief valve will be provided for each steam generator immediately outside the containment ]

structure upstream of the main steam isolation valve. The power-operated atmospheric l relief valve will be electro-hydraulic ally operated, powered f rom the Class IE power j system, and will have the capability to te manually controlled from the main control board or auxiliary shutdown panel.

Based on our review, we conclude that the design criteria and bases for the main steam supply system, up to and including the main steam isolation valves, are in conf ormance with the single failure criterion and the recommendations of Regulatory 10-2

Guide i.29, " Seismic Design Classification, related to seismic design, and include acceptable isolation valve closure time requirernents. Therefore, we conclude that the proposed design is acceptable.

10,4 Other features of Steam and Power Conversion System t( 4.I Main Condenser System The main condenser system, which is designated as nonnuclear datety cibs, will t)e l designed to condense and deaerate steam frtm the lon pressure turbine exhautts and the turbine bypass system. The main condenser system will include all components and equipment f rom the low pressure tort.ine exhaust to the Connections and interfacet

' with tne main condensate system and other systems.

l Ihe scope of our review of the main Londenser system included layout drawings, piping and instrumentation diagrams, and desc riptive inf ormat ion f or the main condenser system and supporting systems that are essential to its operation. As a result of our review, we hava tiermined that the applicant's proposed design criteria and bace- % the main condenser system, and the requirements for preclud-inq malfuncticos or failures of safety related equipment due to a rupture of the n.ain condenser, are acceptable.

Ite b n's 'or our acceptance has been conformance of the applicant's proposed d d gn criteria and design bases for the main cundenser system and its supporting systems to applicable staf f technical positions and industry standards ptvsented in Section 10.4.2 of the Standard Review Pian.

Theref ore, we conclude that tne propnsed design of the main condenser system is acceptable.

l(.4.2 Main Condenser Evacuation System The main condenser evetuation system will be designed to establish and maintain condenser vacuum by removing noncondensable gases from the condenser shells. The system will be designed to Quality Group D in accordance with the recommendations of Regulatory Guide 1.26, " Quality Group Classifications and Standards for Water ,

Steam , and Radioactive-Waste Containing Components of Nuclear Power Plants."

Since the system is not safety-related,-it is classified as nonseismic. The main condenser evacuation system will con:,ist of two steam jet air ejectors. Air and noncondensables f rom the air ejectcr exhausts will be continuously monitored by a radiation detector prior to release the environment.

The scope of our review included the design provisions incorporated in the main condenser evacuation system to monitnr and control releases of radioactive materials in gaseous effluents in accordance with Criteria 60 and 64 of the General Design Criteria, As a result of our evaluation, we find that the design criteria and 10-3

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i design bases of the proposed main condtnser evacuation system are in confermance

' .with the applicable regulations and regulatory guides, Therefore, we conclude

, that the proposed desion is acceptable.

10.4.3 lurbine Glarxi Sealing System i

i e

The turbine gland seallng system will be designed to control radioactive steam j

leakage from, and air inleakage into, the turbine. The components of the system will be designed to Quality Group D in accordance with the recommendations of 4

Regulatory Guide 1.26. Since the system is not safety-related, it is classified as nonseismIC. Ihe turbine gland sealing system will Consist of a series of labyrinth-type seal rings, a steam supply system (main steam), and a condenser.

Steam will be supplied to the sealing rings f rom the main steaa supply system.

The condenser will (1) maintain a slight vacuum in the system, (2) return seal leakage to the main condenser as condensate, and (3) exhaust the noncondensables to the main condenser evacuation system.

As a result of our evaluation, we find that the design critor# 3 and design bases for the proposed turbine gland sealing system are in conformance with Regulatory Guide 1.26. Theref ore, we conclude that the proposed design is acceptable.

10.4.4 Turbine Bypass System The turbine bypass system, which is not safety-related, will be designed to permit 100 percent external load rejection without reactor or turbine trip. Ine turbine bypass system will include all components and piping from the branch connection at the main steam system to the main condensers.

l The scope of our review of the turbine bypass system included layout drawings,

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piping and instrumentation diagrams, and descriptive information for the turbine bypass system and auxiliary supporting systems that are essential to its operation.

As a result of our review, we have determined that the applicant's proposed design criteria and bases for the turbine bypass system, and the requirements for safe operation of the system during normal, abnormal, and accident conditions, are acceptable.

The basis for our acceptance has been conformance of the applicant's proposed design criteria and design bases for the turbine bypass system and its supporting systems to applicable staff technical positions presented in Section 10.4.4 of the Standard Review Plan.

Therefore, we..:onclude that the proposed design of the turbine bypass system is acceptable.

10-4

l l 10.4.5 Ci-culating Water System 1

The Circulating water system will be designed to remove the heat rejected from the i main condenser to the atmosphere via mechanical draft cooling towers. The system t

will not be required to maintain the reacto: in a safe shutdown condition or miti-gate the consequences of accidents.

Flooding, as a result of a postulated failure of the circulating water system at an expansion joint, will not prevent safety-related equi ~ <nt f rom perf orming its intended safety function. The turbino tu: icing floor will be 3.1 feet above plant grade. The fastener * , aunch will hold the siding on the turbine building, will be designed to release the siding by the time the water level has reached 2.6 feet above the turbine building floor. 1 Fit will allow flood water to drain into the yard area around the turbine building. There will be no safety-related struc-tures, systems, or components inside the turbine building or in the yard area in the vicinity of the turbine building. In addition, there will be no direct passage-ways at or below ground level connecting the turbine building with safety-related t aeas, such as

nc annulus building or control building. A solid wall, designed to sithstr ,d the maximum flood water level inside the turbine building, will be piu.id20 between the turbine building and the health physics area and will serve as e barrier to water flowing toward the annulus building.

Flood protection walls will be also provided at the entrances of the below grade cable tunnel, the main steam pipe tunnel and the auxiliary pipe tunnel to prevent flood water from flowing into safety-related areas through these tunnels.

We have reviewed the adecuacy cf the applicant's proposed design criteria and bases for the circulating wr;er system during normal, abnormal, and accident conditions. Since a far.ure in the system will not affect safe plant shutdown, we conclude that the design criteria and t,ases of the circulating water system are

, acceptable.

10.4.6 Condensate and Feedwater Systems The condensate and feedwater systems will be designed to return condensed steam from the condenser to the steam generators and to automatically control the water i

levels in the steam generators during steady-state and transient conditions.

l The condensate and feedwater systems were reviewed to assure that a failure in these systems would not result in the loss of any essential equipment and would not affect safe Llant shutdown. They were also reviewed to assure that adequate isolation wi'1 be provided from these systems where they connect to seismic Cate-gory I systens.

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the portion of the condensate and feedwater systems otending f rom and including l the cutermost feedwater isolation v' alves to the steam generator inlets will be l designed to seismic Category I. requirements. There will be two seismic Category 1

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feedwater isoletina valves ia eae sv ety " elated feedwater supply line. The l feedwater isolation valves will be designed to isolate feedwater supply to the a

4 steam generators witnin five seconds *ollowing a feedwater isolation signal, We have reviewed the design bases dnd criteria for these systems and conclude that a system failure will not affect safe plant snutdown, and that the condensate and

feedwater isolation systems will meet single failure criterion. Therefore, we-tenclude that the proposed system design is acceptable.

? l 10.4.7 Auxiliary Feedwater System l l l

3 l The auxiliary feedwater system will be designed to supply an independent source of a

C water to the steam generators to remove reactor sensible and decay heat in the

{ event of a loss of the main feedwater supply.- The auxiliary feedwater system will d

be designed to function automatically in the event of malfunctions such as loss of l power, loss of feedwater, main steam line or feedwater line break, or other accidents.

j The auxiliary feedwater system wili be designed to seismic Category I requirements f and will be located in a tornado and missile protected structure.

l The major components of the auxiliary feedwater system will include one 100 percent

{ capacity steam turbine-driven pump and two 50 percent capacity motor-driven pumps. !

j The turbine-driven pump will receive steam from the main steam lines upstream of the main steam line stop valves and exhaust to the atmosphere. The turbine-driven pump train will be capable of supplying the required auxiliary feedwater to the steam generators independently of onsite or of f site alternating current power i supplies, since the alternating current powered, motor-operated valves at the !

) steam supply lines and the auxiliary feecwater discharge lines of the turbine-4 driven auxiliary feedwater train will be designed to normally be fully open and l

j to remain open in case of a loss at alternating current power.

l

( The motor-driven auxiliary feedwater pumps and associated valves and instrumenta*

1 tion will be powered from the emergency diesel generators in the event of a loss j of offsite power. These pumps will normally take suction from the seismic Category 1, j tornado and missile protected auxiliary feedwater storage tank. For a long-term backup water supply to the auxiliary f eedwater system, the seismic Category 1 )

service water system will be connected to the pump suction with nnrmally closed, I f manually-operated isolation valves.

?

1 We have reviewed the adequacy of the proposed design criteria and bases of the auxiliary feedwater system necessary for safe operation of the plant during normal, abnormal, and accident conditions, We conclude that the proposed system design 10-6 4

. m~ .,-.-- --.<_.w.. ~ _. _ _ -- ,,-,.- , , ~ _ 4 m . - - - - - . ,

conf orms with the diversity requirements of Branch Tecnnical Position APC5810-1

" Design Guidelines for Auxiliary Feedwater System Pump Drive and Power Supply Diversity for Pressurized Water Reactor Plants," and that the system will have suf f ic ient f lev ibility 9nd radundecy, inc luding the capability of tne system to withstand the combination of a single active failure and high energy line breaL Therefore, we conclude that the proposed system design is acceptable.

The Sundesert plant will use a steem generator design which incorporates a top feedwater ring. A number of events, referred to as water hammers, have occurred with this type of steam generator in which a pressure wave has propagated through the feedwater pipes as a result of uncovering the feedwater rings in the steam generator. As a result we have requested the applicant, and the applicant has committed, to perform tests acceptable to us to verify that an unacceptable feedwater hammer will not occur on Sundesert. The tests will be performed prior to the plant reaching 100 percent power to check out the operating procedures f or refilling )

the steam generator with the auxiliary feedwater system following a turbine trip with a loss of offsite power.

10.5 Steam and Feedwater System Materials The mechanical properties of materials to be selected for Class 2 and 3 components of the steam and feedwater systems will satisfy Appendix 1 to Section III of the ASME Code, and Parts A, B or C of Section 1[ of the ASME Code. The fracture toughness properties of ferritic materials will satisfy the requirements of the provisions of the Summer 1977 Addendum to Section III of the ASME Code; Subsection NC-2300 for Class 2 components; and Subsection ND-2300 for Class 3 components.

The controls to be imposed upon austenitic stainless steel will be in accordance {

with the recommendations of Regulatory Guide 1.31, " Control of Stainless Steel Welding," and Regulatory Guide 1.44, " Control of the Use of Sensitized Stainless Steel." Fabrication and heat treatment practices performed in accordance with these recommendations provide reasonable assurance that. stress corrosion cracking will not occur during the design life of the plant. The controls to be placed upon concentrations of leachable impurities in nonmetallic thermal Insulation to be used on austenitic stainless steel components of the steam and feedwater systems are in accordance with the recommerdations of Regulatory Guide 1.36, " Nonmetallic I Thermal Insulation for Austenitic Stainless Steel."

The onsite cleaning and cleanliness controls to be used during fabrication satisfy l

t the recommendations of Regulatory Guice 1.37, " Quality Assurance Requirements for Cleaning of Fluid Systems and Associated Components of Water-Cooled Nuclear Power Plants," and the requirements of ANSI Standard N45.2.1-1973, " Cleaning of fluid Systems and Associated Components During Construction Phase of Nuclear Power Plants." The precautions to be taken in controlling and monitoring the preheat 10-7

and interpass temperatures during welding of carbon and low-alloy steel components I conform to the recommendations of Regulatory Guide 1.50, " Control of Preheat Temperature for Welding tow-Alloy Steel" and to the modifications described in Westinghouse Topical 9eport WCAD-B577. "The 6pplication of Preheat Temperatures Af ter Welding of Pressure Vessel Steels," which we have found acceptable.

Conformance with the codes, stancaras, and regulatory guides mentioned above consti-tutes an acceptable basis for assuring tt.e integrity of steam and feedwater systems, and for meeting the applicable requirements of Criterion 1 of the General Design Criteria. Therefore, we find the materials and fabrication procedures for the steam and feedwater systems to be acceptable.

10-8 I - -- - - - - - - - - _ _ _ _ _ _ _ - - - - - - - - - - - - _ _ _ _ __

- . . - _ . - , -- - - -- . _ -.= - .- -

4 1

1 i

11.0 RADI0 ACTIVE WASTE MANAGEMENT l

11.1 Summary Description l

I Each unit of the Sundesert plant will have its own radioactive waste management systems. The radioactive waste management systems will be designed to provide for controlled handling and treatment of liquid, gaseous and solid wastes. The liquid radioactive waste treatment systems will process wastes from equipment and floor drains, sample wastes, decontamination and laboratory wastes, regenerant chemical wastes, and laundry and shower wastes. The gaseous radioactive waste treatment s>3tems will provide holdup capacity to allow decay of short lived noble gases st'ipped . rom the primary coolant and will provide for treatment of ventilation ;

I exhausts through high efficiency particulate air filters and charcoal adsorbers as necessary to reduce releases of radioactive materials to "as low as is reasonably The achievable" levels in accordance with 10 CFR Part 20 and 10 CFR Part 50.34a.

solid radioactive waste treatment systems will provide for the solidification, packaging and storage of radioactive wastes generated during station operation prior to shipment ofrsite to a licensed facility for burial. ]

I In our evaluation of the liquid and gaseous radioactive waste treatment systems, we have considered; (1) the capability of the systems for keeping the levels of radio- j activity in effluents "as low as is reasonably achievable" based on expected rad- f waste inputs over the life of the plant, (2) the capability of the systems to maintain releases below the limits in 10 CFR Part 20 during periods of fission product leakage at design levels from the fuel, (3) the capability of the systems to meet the processing demands of the plant during anticipated operational occur-rences, (4) the quality group and seismic design classification applied to the equipment and components and structures housing these systems, (5) the design features that will be incorporated to control the releases of radioactive materials in accordance with Criterion 60 of the General Cesign Criteria and (6) the potential for gaseous release due to hydrogen explosions in the radioactive gaseous waste system.

l In our evaluation of the solid radioactive waste treatment systems, we have con-i sidered; (1) system design objectives in terms of expected types, volumes and activities of waste processed for offsite shipment, (2) waste packaging and conform-ance to applicable Federal packaging regulations, and provisions for controlling potentially radioactive airborne dusts during baling operation, and (3) provisions for onsite storage prior to shipping.

l In our evaluation of the process and effluent radiological monitoring and sampling systems, we have considered the systems' capability; (1) to monitor all normal and potential pathways for release of radioactive materials to the environment, (2) to 1

11-1 ,

l

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

control the release of radioactive materials to the environment, and (3) to monitor performance of process equipment and detect radioactive material leakage between systems.

l Our evaluation included the determination of the quantities of radioactive materials l that will be released in liquid and gaseous effluents and the quantity of radio-active waste that will be shipped offsite to a licensed burial facility, in making these determinations, we have considered waste flows, activity levels and equipment performance, consistent with expected normal plant operation, including anticipated operational occurrences for an assumed 30 years of normal plant operation.

4 The estimated releases uf radioactive materials in gaseous effluents were cal-l culated using the PWR GALE Code described in NUREG-0017, " Calculation of Releases l of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors (PWRs)", dated April 1976. The principal parameters used in these calcula-tions are given in Table 11.1. The gaseous source terms are given in Table 11.2.

The source terms were used to calculate the individual and population doses in accordance with,the mathematical models and guidance contained in Regulatory Guide 1.109, Revision 1, " Calculation of Annual Average Doses to Man from Routine !

Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix 1." Meteorologic factors in the dose calculations were deter-mined by using the guidance in Regulatory Guide 1.111, Revision 1, " Methods of Estimating Atmospheric Transport and Dispersion of Gaseous Effluents from Routine Releases f rom Light-Water-Cooled-Reactors. " The calculated individual and popula-tion doses are given in Table 11.3.

We evaluated potential radwaste system augments based on a study of the applicant's system designs, the doses to the population within 50 miles of the reactor, an interim value of $1,000 per total body man rem and $1,000 per man-thyroid-rem for reductions in dose by the application of augments, and the cost of pot,ential rad-waste system augments as presented in Regulatory Guide 1.110, " Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors." The principal parameters used in the cost-benefit analysis are given in Table 11.4.

Based on our e,aluation, as described in the following secticns, we conclude that the proposed liquid and gaseous radioactive waste treatment systems for the Sundesert plant will be capable of maintaining releases of radioactive materials in liquid l and gaseous eifluents to "as low as is reasonably achievable" levels in accordance with 10 CFR Part 50.34a, Sections 11.A, II.B. II.C and 11.0 of Appendix 1 to 10 CFR Part 50. Therefore, we find the proposed liquid and gaseous radioactive waste treatment systems to be acceptable. We also find the associated process and effluent radiological monitoring and sampling systems and the solid raditactive waste treatment systems to be acceptable.

11-2

TABLE 11.1 PRINCIPAL (ARAMETERS USED IN ESTIMATING RELEASES OF RADI0 ACTIVE MATERIAL IN FRt0ENTS F WM 7 NTE W T NUCLEAR PLANT, UNIT N05, 1 AND 2 Reactor Powcr Level 2958 megawatt thermal Plant Capacity Factor 0.80 Operating Power Fission Product hource Term 0.12 percent Primary System:

Mass of coolant 400,000 pounds Letdown rate to makeup and purification system 75 gallons per minute Shim bleed rate 1 gallon per minute Leakage rate to secondary system 100 pounds per day Leakage rate to auxiliary area 160 pounds per day Frequency of degassing (cold shutdowns) 2 times per year Secondary System:

Steam flow rate 1.28 x 107 pounds per hour Mass of steam in each generator 5700 pounds Mass of liquid in each generator 9500 pounds Mass of secondary coolant 1.3 x 106 pounds pate of steam leakage to turbine building 1700 pounds per hour Containment Building Volume 2.42 x 106 cubic feet Frequency of Containment Purges (high volume) 24 times per year Turbine Building Leak Rate 5 gallons per minute lodine Partition factors:

Steam generator internal partition 0.01 Primary coolant leak to auxiliary area 0.0075 Condenser / vacuum pump (volatile species) 0.15 Iodine Decontamination Factor for Ventilation Systems:

Charcoal adsorbers 10 Particulate Decontamination Factor for Ventilation Systems:

High efficiency particulate air filters 100 Dynamic Adsorption Coefficients for Charcoal Beds:

Xenon 330 cubic centimeters per gram Krypton 18.5 cubic centimeters per gram 11-3

_ TABLE 11.2 CALCULATED RELEASES OF RA010 ACTIVE MATERIALS IN GASEOUS EFFLUENTS FROM $UNDESERT NUCLEAR PLANT, UITIT N05. 1 AND 2 RELEASE (Curies) PER YEAR PER UNIT Waste Gas Condenser Processing Building Ventilation Vaccuum Nuclides System Reactor Aux i lia ry Turbine Pump Total Kr-85m a a 2 a 1 3 Kr-85 210 a a a a 210 Kr-87 a a 1 a a Kr-88 a 1 4 1

a 2 7 Xe-131m a 2 a a a 2 Xe-133m a 5 a a a 5 Xe-133 a 370 29 a 18 420 Xe-135 a 5 4 a 2 11 Xe-138 a a 1 a a Total 1 660 I-131 a 8.8(-4)D 1-133 3.7(-2) 5.8(-4) 2. 3 (-2 ) .. 6.1(-2) a 3.l(-5) 5.8(-2) 8.2(-4) 3.6(-2) 9.5(-2) j Co-60 c 3.l(-5)

Co-58 2.7(-2) c c 2.7(-2) c 6.2(-5) 6.0(-2) c fe-59 c 6.2(-6) c 6.0(-2)

Mn-54 6.0(-3) c c 6.0(-3) c 1.8(-5) 1.8(-2) c Cs-137 c 3. l(- 5) c 1.8(-2)

Cs-134 3.0(-2) c c 3.0(-2) c 1.8(-5) 1.8(-2) c Sr-90 c 2.5(-7) c 1.8(-2)

Sr-89 2.4(-4) c c 2.4(-4) c 1.4(-6) 1.3(-3) c c 1.3(-3) l Total Particulates

! 1.6(-1) l C-14 8 - - - -

8 H-3 590 590 - - -

A r-41 - 1200 25 - - -

25 a = less than one Curie per year for noble gases, less than 10~ Curies per year for iodine.

b = exponential notation; 8.8(-4) = 8.8 x 10 ~4 c = less than one percent of total 11-4

4 i

i TABLE 11.3 CALCULATED DOSES TO A MAXIMUM INDIVIDUAL AND THE 50-MILE POPULATION

~

FROM SUN 6ESERT NUCLEAR ELANT1, UNIT NOS. 1 AND 2

,1NDIVIDUAL DOSES Liquid Effluents:

Dose to total body from all pathways

Dose to any organ f rom all pathwyas a

Noble Gas Ef fluents (at site boundary 0.61 miles south-southeast):

Gamma dose in air D.96 millirad per year per unit Beta dose in air 2.2 millirad per year per unit i Dose to total body of an individual 0.61 millirem per year per unit '

Dose to skin of an individual 1.9 millirem per year per unit Radiolodines and Particulates Dose to any organ from all pathways .

0.62 millirem per year per unit (at a garden, 3.D miles noetheast)

POPULATION DOSES Liquid Effluents:

Dose to total body from all pathways

Dose to thyroid frsm all pathways Gaseous Effluents:

Dose to total body from all pathwr.js 0.91 man-rem per year per unit Dose to thyroid from all pathways 1.2 man-rem per year per unit There are no routine radioactive effluents discharged to the hydrosphere, and, thus, no dose commitments are presented.

11-5

l l

TABLE 11.4 PRINCIPAL PARAMETERS USED IN THE C051-BENEFIT ANALYSIS FOR SUNDEg er NUCLEAR PLAN 1, UN!T N05. 1 AND 2 1

Labor cost correction factor, Feder31 Power Commission Region VII]* 1.2 Indirect cost factor

I.625 Capital recovery factor ** 0.1463 l

From Regulatory Guide 1.110. " Cost-Benefit Ar.alysis f or Radwaste Systems f or Light-Water-l Cooled Nuclear Power Reactors.'

r ..

l Value furnished by the San Diego Gas and Electric Company l

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11.2 Radioactive Waste Treatment. System Description and Evaluation 11.2.1 Liquid Radioactive Waste Treatment Systems The liquid radioective waste treatment systems will consist cf process equipment l and instrumentation i.ecessary to collect, process, monitor anu recycle radioactive liquid wastes. The liquid radioactive waste treatment systems will be designed to collect and process wastes based on the crigin of the waste in the plant and on the espected levels of radioactivity. All liquid waste will be pivcessed on a batch basis to permit optimum control, Tritium concentration in thi reactor coolant will be controlled by a vaporizer to permit tritium releases as a gas.

4 A schematic diagram of the liquid radioactive waste treatment systems is given in figure 1. The liquid radioacti4e waste treatment syst.ms will consist of the j aerated warte system, the turbine building and regenerant caemical waste system, e d the detergent waste system, in addition, for horon control, the boron recovery system will p*ocess a portion of the flow (shim bleed) from the makeup and purifica-tion system along with the wastes collected in the reactor coolant drain tank.

Steam generator blowdown will be flashed, with the steam being sent to the feedwater heaters and the liquid being sent to the condenser hotwell prior to cleanup by the condensate demineralizers.

Baron Recovery System A letdown stream of approximately 75 pallons per minute of primary coolant will be removed f rom the reactor coolant system f or processing through the makeup and purification system. The letdown stream will be cooled through the regenerative and letdown heat exchangers, reduced in pressure, filtered, processed through one of the two mixed bed demineralizers (in the lithium borate forn.) . and sent to the makeup tank. A third domineralizer, which will contain cation resins, will be used intermittently (approximately 10 percent of the time) for lithium and cesium control, Boron concentration will be controlled during core. life by feet and bleed operation l to the buron recovery system, and at the end of the core life by anion deborating demineralizers in the boron recovery system. The shim bleed flow to the boron recovery system will be approximately 1,440 gallons per day.

Processed along with the shim bleed will be the liquids collecteu in the reactor plant drains. The liquids from these drains will be collected inside containment by tt.. reactor coolant drain tank and outside containment by the pr' mary drains transfer tank. The shim bleed and liouids from the eactor plant drains will be processed through a 150 pallon per minute degasifier, and collected in ten 150,000 gallon boron recovery tanks. Liquids collected in these tanks will be processed in batches thiough a 95 gallon per minute boran eve orator and rollected in one the two 12,000 gallon boron test tanks. If further purification is required, the l

contents will be processed through a 35 cubic foot boron demineralizer and filtered.

11-7 I

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The system will be designed so that no liquid waste from the boron recycle system will be released to the Palo Verde Outfall Drain.

Aerated Waste System The aerated waste system will process wastes fron equipment urains and floor arains in the auxiliary building resin sluice water, decant water, refueling operation wastes, sample drains, laboratory wastes, and spent fuel pool leakage. The wastes will be collected in one of the three ? 000 alica waste evaporator feed tanks anc processed through a 25 gallon per m'nu'e x e evaporator. The distillate will be collected in two 12,000 gallon waste test tanks for d vsis Fol!nwing analysis, the evaporator distillate can be further processed in a 35 cuoic fmt ooit: m g demineralier and filtered, or recycled te u e primary grade water tanks for teuse in the plant. The system will be designed so that none of the distillate will be discharged to the Palo Verde Outfali Drain.

Turbine Building and Regenerant Chemical Waste System l Wastes f rom condensate demineralizer regeneration and other turbine building ctr nical wastes of high conductivity will be collected in two 20,000 gallon high conductivi - 6 waste storage tanks. The liquid waste will be processed through a 25 gallon per minute regenerant chemical evaporator, and collected in a 550 gallon regenerant I chemical distillate tank for analysis. f he waste will either be f urther heated in the waste evaporator of the aerated waste system, er recycled to the condensate system The system will be designed so that none of the liquid waste will be discharged to the Palo verde Outfall' Drain.

Detergent Waste System l Detergent wastes, which will consist, of hot shower wastes, laundry wastes, and decontamination solutions containing surfectants, w':1 be collected in a 4,000 gallon waste drain tank. The waste will te processed a two gallon per minute j

waste concentrator of the solid radioactive waste treatment system. fhe distillate will be sent to the waste evaporator feed tanks and the concentrates will be solid-ified and buried offsite.

Conformance with Regulations and Brunch Technical Positions lhe liquid radioactive waste treatment systems will be located in the annulus tuilding which will be designed as se hmic Category 1. The proposed design para-meters of principal components considered in the liquid radioactive waste ratment l

system evaluation are listed in Table il 5. We find the applicant's proposed design for the liquid radioactive waste treatrient systems to be in accordance with Branch Technical Position i,$B 11-1, Revision 1, " Design Guidance for Radioactive Waste

.4nat, .- Svster, Installed in L ight-Water-Cooled Nuclear Power Reactor Plants.'

11-9

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

e TABLE 11.5 DESIGN PARAMETERS OF PRINCIPA' COMPONENT 5 CONSIDERED IN THE Ei!A' UATION OF LIQulD AND'GASE005 RADIOACTIVE WA51E

~

TREATMENT (,' $T E MS CAPACITY CCMPONENT NUMBER EACH i LIQUID SYSTEMS Waste Evaporatur Subsystem

Feed Tank 3 25,000 gallons Evaporator 1 25 gallons per minute Test Tank 2 12,000 gallo,s Demineralizer 1 150 gallons per minute l

l Regenerant Chemical Subsystem

Feed 'ank 2 20,000 gallons Esap;rator i 25 gallons per minute Dissillate Tank 1 550 gallons laundry Drains subsystem

Feed Tank I 4,000 gallons Evaporator 1 2 gallons per minute vaporizer Subsystem *

. Feed Tank 1 12,000 gallons

{ Vaporizer i 2 gallons per minute

GASE0us SYSTEMS s

a j Radioactise Gaseous Waste System

Degasifier i 150 gallons per minute Charcoal Bed Adsorbers 2 13,000 pounds charcoal I. r-

! Qaality group and seismic design are in accordance with Bra Technical Position,

! ETSB 11-1, Revision 1.

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

The system design wii 6 also include measures intended to control the release of -

racioactive materials due te p_' i.sl overflows from indoor and outdoor storage tanks. Tank levels will ba men:tored eitner locally or in the control room and

high level alarms will be activated should preset levels be exceeded. Overflow provisions', such as sumps, dikes, and e erftew lines, will permit the collection and subsequent processing of tank overflow. he consider these provisions to be capable of controlling the release of radioactise materials to the environment.

The proposed liquid radioactive waste treitment systems will be designed so that, during normal operation, there will be no releases of liquid radioactive effluents to the environment. Accordingly, no doses due to liquid effluents are expected, and we conclude that the liquid radioactise wn:te treatment systems will reduce liquid radioactive effluents to "as low as is reasonably achievable" levels ', i accordance with 10 CFR Part 50.34a and Appentiix 1 to 10 CFR Part 50. We have also determined that the liquid radioactive waste treatment systems will be captale of reducing the release of radioactive materials in liquid ef fluents to concem ~ations below the guideline values of 10 CFR Part 20. during periods of fission prod et leakage from the fuel at design levels.

< Based on these findings, we conclude that the proposed design of the liquid radio-

' active waste treatment systems is acceptable.

. 2. 2 Caseous Radioactive Waste Treatment Systems The gaseous radioactive waste treatment systems will consist of the radioactive gaseous waste system and the plant ventilation systems. The radioactive gaseous 1 waste and plant ventilation systems will be designed to collect, stor , process, monitor, recycle, and/or discharge potentially radioactive gaseous wa,.es which will be generated during normal operat*on of the plant. The systems will consist of equipment and instrumentation necessary to reduce releases of radio. active gases and particulates to the environment. The principal sources of gaseous waste will be the effluents from the gaseous waste management system, condenser vacuum pumps, and ventilation exhausts from the reactar containment, annulus and turbine build-ings, and fuel handling area.

i b

The principal system for treating gaseous wastes will be the process gas portion of l

the radioactive gaseous waste system. The process gas portion of the radioactive j

gaseous waste system will collect and store hydrogenated fission product gases stripped from the primary coolant letdown and gases stripped from the reactor plant l gaseous (hydrogenated) drains by processing the gases through charcoal bed adsorbers and a high efficiency particulate air filter. The process vent portion of the radioactive gaseous waste system will collect and treat low activity aerated streams, including the gas from the reactor plant aerated vents header, by processing the gases through high efficiency particulate air filter and a charcoal adsorber.

Ventilation exhaust air frum the containment will be processed through high efficiency 11-11

e t

1 l particulate air filters and charcoal aosorbers prior to release to the environmeat.

' ventilation exhaust air from the turbine ouilding, main condenser vacuum pump t

sanaust, annulus building, and fuel building will be released to the environment

' without treatment, Ventilation exhaust air from the solid waste building will be ;

j processed through a high ef ficiency pertica!a'.e air filter prior to release to the i

{ environment. The radioactive gaseous waste system and ventilation treatment systems

! are shown schematically in Figure 11.2.

1 l

i Radioactive Gaseous Waste System j Ihe radiotsctive gaseous waste system will consist of two subsystems: the process gas portion and the process vent portion.

l 4

t he process gas portion of the radioactive gaseous waste systam will tv designed to i

collect and process gases stripped from the primary coolant and from the hydro-genated liquids collected in the reactor coolant drain tank and primary drains I

'ransfer tanir. These gases will consist mainly of hydrogen and water vapor and will contain relatively small quantities of radioactive gases. The gases will be j aried and processed in a charcoal adsorber train that will adsorb iodine from the 4

j trocess stream, and selectively delay the release of krypton and xenon isotopr-Ihe carcoal adsorber train will consist of two ambient-temperature charcoal bed

adsorbers, in series, each containing 13,000 pounds of charcoal. On the basis of the charcoal dynamic' adsorption coefficients listed in Table 11.1 and a flow rate j through the system of 0.65 standard cubic feet per minute, we calculated that the i

i decay times provided will be 12.8 days for xenon and 7.2 days for krypton. The

processed gases will be treated by a high ef ficiency partculate air filter and j released t the environment via the ventilation vent. A rac + ion monitor will

automatically terminate the discharge of radiation levels which exceed a predeter-j mined value.

i l

The process vent (aerated) portion of the radioactive gaseous waste system will be e

i designed to collect the low activity aerated gas streams from the reactor plant aerated vents header, degasifier flush vent, dilution air a +ist blowers (from l combustible gas control) and the process gas adsorption b u s h .. These streams will be treated by high efficiency particulatt tir fil.ers and charcoal adsorbers

[

4 orice to release to the environment.

i.

1 Containment Ventilation System I

4 Radioactive gases will be released Inside the containment when primary system components are opened or when primar" .ystem leakage occurs, During normal opera-tion, the gaseous activity will be seaied within the containment but will be re-j leased during containment purges. In accordance with the guidelines in NUREG-0017, we assumed that the containment will be purged 24 times per year. Prior to purging, f the containment-atmosphere will be recirculated through high efficiency particulate a

1 11-12 h

(NO TRE ATMENT)

TURBINE BUILDING > TURBINE BUILDING EQUIPMENT UENTS MAIN CO 8N TURBINE BUILDING EQUIPMENT VENTS E ORS PRIMARY SYSTEM DEG ASSING A A CVCS DEGASSIFIER 4

- A y VENTILATION VENT REACTOR G ASEQUS .

O R AINS SYSTEM r V V 1 r

CH ARCO AL BED TO VOLUME ADSORBERS CONTROLTANK CONTAINMENT BUILDING 10,000 CFM

J6

+ A C - A C 7 VENTILATION VENT w

CONTAINMENT __ _

l I I

I (NO TREATMENT) V y VENTILATION VENT ANNULUS BUILDING

VENTIL ATION VENT FUEL BUILDING s

SOLID WASTE BUILDING > VENTIL ATION VENT

CFM = cubic feet per minute HIGH EFFICIENCY A -PARTICULATE AIR FILTER C -CHARCOAL ADSORBER FIGURE 11.2 GASEOUS WASTE TREATMENT SYSTEMS, SUNDESERT NUCLEAR PLANT

i l

1 i air filters and charcoal adsorbers for particulate and iodine removal. In our evaluation,~we assumed that following recirculation, the containment will be purged j through high efficiency particulate air filters and charcoal adsorbers. We assumed f radionuclide removal during the recirculation phase to be based on the design flow

{ rate of 10,000 cubic feet per minute. system uoeration for 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months , a mixing j efficiency of 70 percent, a particulate decontamination factor of 100 for high

efficiency pirticulate air filters, and an iodine decontamination factor of 10 for i

3 charcoal adsorbers. The containment porge exhaust monitors will automatically

) isolate the purge system upon receipt of a radioactivity level above predeter-a l

mined value. i i

i t

Ventilation Releases From Other Buildiny i

j Radioactive materials will be introduced into the plant atmosphere due to leakage

{ from equipment transporting or handling radioactive materials. In accordance with l

{ fhe guidelines in NUREG-0017, we estimated that 160 pounds of primary coolant per l day will leak to the auxiliary area with an iodine partition f actor of 0.0075.

Small quantities of redionuclides will be released to the turbine building atmos-j phere based on an estimated (NUREG-0017) leakage of 1700 pounds per hour of steam.

l Dur calculations assumed that effluents from the annulus building, fuel building.

{ and the turbine building will be released directly to the environment without l

j treatment, and that the effluents from the solid waste building will be processed l

{

through a high efficiency particulate air filter prior to release to the environment.

$ Main Condenser Vacuum Pump $

i 1 Of f gas f rom the main condenser vacuum pumps will contain radioactive gases as a

( result of primary to secondary leakage. In accordance with the guidelines in f NUREG-0017, we assumed a primary to secondary leak rate of 100 pounds per day in

{ our evaluation. Noble gases and iodine will be contained in the leakage from the f steam generator and will be released to the environment through the main condenser i vacuum pumps in accordance with the partition factors listed in Table 11.1. The 1

d vacuum pump exhaust will be released directly to the environment without treatment.

)

l Conformance with Regulations and Branch Technical Positions l

t The proposed design parameters of the principal equipment i'1 the gaseous radio-

] active waste treatment systems are listed in Table 11.5. We find the applicant's proposed gaseous radioactive waste treatment systems to be in conformance with

{ Branch Technical Position ETSB 11-1, Revision 1. In addition, the radioactive j gaseous waste system will be located in the annulus building which will be a seismic

Category I structure.

i i

a 11-14 i

1 1

i

.1 1

The radioactive gaseous waste system will be designed to prevent a hydrogen explo-l

.sion. Instrumentation with automatic alarm and control f unctions will be provided

(

l to monitor the concentrations of the appropriate gas in the portions of the systems having the potential for containing " pler.ive mixtures. We find the applicant's f

i prcposed radioactive gaseous waste astem capacity and design criteria, along with the design provisions incorporated to reduce the potential for hydrogen esplosions, i to be acceptable. ,

t We have determined that the proposed gasecas radioactive waste treatment and plant ventilation systems will be capable of rec:ucing the release of radioactive materials in gaseous effluents to approximately 060 Curies per year per unit for noble gases, 0.061 Curies per year per unit for iodine-131,1200 Curies per year per unit for tritium, eight Curies per year per unit for carbon-14, and 0.16 Curies per year per

. unit for particulates. The applicant estimated a release of approximately 350 l Curies per year per unit for noble gase s, 0.063 Curies per year per unit f or

' iodine-131, 1200 Curies per year pei unit for tritium, eight Curies per year per

' unit f or carbon-14, and 0.16 Curies per year per unit for particulates. The dif-ference in the two noble gas source terms is due to the fact that we assumed four shutdown purges of containment aer yeai ar.d 20 purges at power, whereas the applicant assumed two shutdown purges per year ard two purges at power. Our calculated values for the annual releases of radicnuclides in gaseous effluents from each unit are given in Table 11.2.

Using tne source terms given in Table 11.2, we calculated the total body dose to be !

less than 10 millirads per year per unit for gamma radiation and an organ dose to be less than 15 millirem per year per unit for radiciodine and radioactive partic-ulates, which are in accordance with the guideline values in Sections II.B and ll.C of Appendix 1 to 10 CFR Part 50. The calculated doses are given in Table 11.3.

i The calculated total body and thyroid coses from gaseous releases to the population within a 50 mile radius of the plant, when multiplied by $1,000 per total body j

mar rem and $1,000 per man-thyroid-rem, resulted in cost-assessment values of $910 per year per unit, and $1200 per year per unit, respectively. Potential radwaste The l

system augments were selected from the list given in Regulatory Guide 1.110.

most ef fective augment considered was the addition of a charcoal /high ef ficiency I

particulate air filtration unit to the main condenser evacuation system.

l The calculatec annualized cost of $16,200 for the above augment exceeds the reduc-tion in the cost assessment value of $910 per year per unit for the total body man-rem dose and the reduction in the cost-assessment value of $1200 per year per unit for the man-thyroid-rem dose. We conclude, therefore, that there are no cost-effective augments to reduce the cumulative population dose at the favorable cost-benefit ratio, and that the proposed gaseous radioactive waste treatment systems meet the requirements of Section ll.D of Appendix I to !0 CFR Part 50.

11-15

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

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1 i

I We conclude that the gaseous radioactive waste treatment systems will be capable of reaucing releases of radioactive materials in gaseous effluents to "as low as is reasonably achievable" levels in accordance with 10 CFR Part 50.3Aa and Appendix !

4

{ to 10 CFR Part 50. We have also determined that the proposed gaseous radiodClive l

! waste treatment systems will be capable Of reuucina the release of radioactive '

materials in gaseous effluents to concentrations below the guideline values of 10 CFR Part 20 during periods of fission oroduct leakage from the fuel at design I,

j levels.

i l

}

t Based or, these findings, we conclude tnat the proposed design of the gaseous radio-i active waste treatment systems is acceptaole.

} 11.2.3 Solid Radioactive Waste Treatment Systems e

i ,

! The solid radioaClive waste treatment systems ill be designed to process two I

general types of solid wastes: " wet" solid w'stes waich require colidification f prior to shipment, and " dry" solid wastes which requ ne packaging, and in some l

e cases compaction, prior to shipment to a licensed burial facility.

4 j Wet solid wastes will consist mainly of spent filter cartridges, demineralizer j resins, reverse osmosis concentrations, and evaporator bottoms which contain radio- l l active materials removed from liquid streams during processing. Wet solid wastes will be combined with a urea formaldehyde solidification agent and a catalyst in a -i 50 cubic foot container to form a solid m.1trix. The container will be subsequently

sealed and placed in a shield, as required, for offsite shipment.

! l 1

i Dry solid wastes, which will consist mainly of ventilation air filtering medium

(charcoal), contaminated clothing, paper, rags, laboratory glassware, and tools, j will be packaged in 55 gallon drums.

i Wet Solid Wastes 1

i e

l The principal sources of spent demineralizer resins will be one 35 cubic foot fuel l

i pool demineralizer, two 35 cubic foot cesium removal ion exhangers, one 35 cubic j foot condensate polishing demineralizer, two 30 cubic foot mixed bed demineralizers, j 1

one 20 cubic foot cation bed demineralizer, and four 70 cubic foot thermal regener- l ation deminerdlizers. Spent and contaminated resins will be collected in a spent

resin hold tank. When the resin is to be packaged, it will be slurried in batches j to a mixer package and a shipping container for solidification. On the basis of j

j our evaluation, we have determined that approximately 15,000 cubic feet of wet solid wastes, containing approximately 1700 Curies of activity, will be shipped

offsite annually from each unit. The principal radionuclides in the solid wastes will be long-lived fission and corrosion products, namely, cesium-134, cesium-137, cobalt-58, cualt-60, manganese-54 and iron-55.

l 4

11-16

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

Concentrated wastes from the reactor coolant bleed evaporators and liquid radwaste evaporators will be pumped to the evaporator bottoms tank for holdup prior to solidification. j l

Dry Solid Wastes Dry solid wastes will be packaged in 55 gallon drums. Compressible wastes, such as clothing, paper, and rags, will be compressed prior to packaging.

During the compacting operation, the air flow in the vicinity of the compacter will be exhausted by a fan through a high efficiency particulate air filter of the solid radwaste building exhaust system to reduce the potential for airborne radioactive dusts. We estimate that the dry solid wastes will total 4,100 cubic feet per year per unit with a total activity content of five Curies.

Conformance with Regulations and Branch Technical Positions l l

The solid radioactive waste treatment systems will be housed in the solid radwaste building and will conform to the design, construction, and testing criteria of Branch Technical Position ETSB 11-1, Revision 1 " Design Guidance for Radioactive Waste Management Systems Installed in Light-Water-Cooled Nuclear Power Reactor l Plants." The solid radwaste building will be designed to the seismic criteria of Section V and the quality assurance criteria of Section VI of Branch Technical Position, ETSB 11-1, Revision I. In addition, the solid radioactive waste treatment systems will incorporate a process control program and provide for waste storage in accordance with Branch Technical Position, ETSB 11-3, " Design Guidance for Solid Radioactive Waste Management Systems Installed in Light-WateryCooled Nuclear Power Reactor Plants." Storage facilities will include an area in the solid radwaste l building for approximately 50 shipping containers (50 cubic feet each) of high level waste and eighty 55 gallon drums of low level waste. We find the storage capacity adequate for meeting the demands of the station for normal operation.

On the basis of our evaluation of the solid radioactive waste treatment systems, we conclude that the systems will accommodate the wastes expected during hormal opera-tions, including anticipated operational occurrences.

The packaging and shipping of all wastes will be in accordance with the applicable requirements of 10 CFR Parts 20 and 71 and 49 CFR Parts 170-178.

Based on these findings, we conclude that the proposed design of the solid radioac-tive waste treatment systems is acceptable.

I 11-17 8

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11.3 Process and Effluent Radiological Monitoring Systems The process and effluent radiological monitoring systems will be aesigned to provide information concerning radioactivity levf15 in systems throughout the plant, inoicate radioactive leakage between systems, monitor equipment per formance, and monitor and control radioactivity levels in plant dischargts to the environs. Our review of these systems was based on the recommended guidelines provided in Regulatory Guide 1.21, Revision 1, " Measuring, Evaluating and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in liquid and Gaseous Ef fluents f rom Light-Water-Cooled Nuclear Power Plants."

Table 11.6 provides the proposed locations of continuous monitors. Monitors on certain effluent release lines will automatically terminate discharges should radioactive levels exceed a predetermined value. Systems which are not amenable to continuous monitoring, or for which detailed isotopic analyses are required, will be periodically sampled and analyzed in the plant laboratory.

We have reviewed the proposed locations ard types of effluent and process monitors to be provided. Based on the plant design and continuous monitoring locatiens and intermittent samplion locations, we conclude that the monitoring provisions for all i normal and potential release pathways will be adequate for detecting radioactive material leakage to normally uncontaminated systems and for monitoring plant pro-cesses which affect radioactivity releases. On this basis, we consider the monitoring and sampling provisions to meet the requirements of Criteria 50, 63 and 64 of the General Design Criteria and the guidelines of Regulatory Guide 1.21.

Therefore, we conclude that the proposed oesign of the process and effluent radio-logical monitoring systems is acceptable.

11.4 Conclusions In our evaluation, we have calculated releases of radioactive materials in liqtid and DJseous ef fluents for normal operation including anticipated operational occur-rences based on expected radwaste inputs over the life of the plant. We determined that the applicant's proposed design of the liquid and gaseous radioactive waste treatment systems satisfies the design objectives of Appendix I to 10 CFR Part 50.

Therefore, we conclude that the liquid and gaseous radioactive waste treatment systems will reduce radioactive materials in effluents to "as low as is reasonably achievable" levels in accordance with 10 CFR Part 50.34a and, therefore, are acceptable.

We have considered tha potential consequences resulting from reactor operation with a one percent operating power fission product source term and determined that, under these conditions, the concentrations of radioactive materials in liquid and gaseous effluents in unrestricted areas will be a small fraction of the guideline values specified in 10 CFR Part 20.

11-18

)

i TABLE 11.6 PROCESS AND EFFLUENT MONITORING LOCATIONS STREAM MONITORED LIQdl0*

l Component Cooling System Reactor Coolant System l Auxiliary Condensate System Service Water System Steam Generator Blowdown Liquid Waste Vaporizer **

s GASEOUS

Process Vent Effluent Process Gas Effluent **

Containment Purge Air Exhaust **

Steam Jet Air Ejector Exhaust T

All liquid and gaseous effluent streams will be monitored in accordance with the guidelines of Regulatory Guide 1.21. 1 These monitors will alarm and automatically terminate the release when the radiation level exceeds a predetermined level.

11-19

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. . - - = .

4 i

i We have considered the capabilities or the raoloactive waste management systems to 3 meet the anticipated demands of the plant due to anticipated operational occur-f rences and conclude that the liquid, gaseous, and solid radioactive waste treatment

{ system capacities and design flexibilities'will be adequate to meet the anticipated 4

, needs of the plant.

4 k

j We have reviewed the applicant's quality assurance provisions for the radioactive j

waste management systems, the quality group classification used for system compo-

! nents, the seismic design applied to the design of the gaseous waste processing i

a system, and the seismic design applied to the design of structures housing the radioactive waste management systems and have determined that the proposed design of these systems and structures housing these systems meet the acceptance criteria j as set forth in Branch Technical Position, ETSB 11-1, Revision 1.

l We have reviewed the provisions incorporated in the applicant's design to control j the releases of radioactive materials in liquids due to inadvertent tank overflows

! and conclude that the measures proposed by the applicant are consistent with our j acceptance criteria as set feath in Branch Technical Position, ETSB 11-1.

4 j in our review of the process and effluent radiological monitoring systems, we i determined that the systems will include provis-lons for (1) sampling and monitoring

(

all normal and potential effluent discharge paths in conformance with Criterion 64 of the General Design Criteria, (2) automatic termination of ef fluent releases and y assuring control over releases of radioactive materials in effluents in conformance with Criterion 60 of the General Design Criteria and Regulatory Guide 1.21, (3) f sampling and monitoring plant waste proce 5 streams for process control in conform-ance with Criterion 63 of the General Design Criteria, (4) sampling and conducting analytical programs in conformance with the guidelines in Regulatory Guide 1.21,

].

and (5) monitoring process and effluent streams during postulated accidents. The review included piping and instrumentation diagrams and process flow diagrams for j the liquid, gaseous, and solid radioactive w ste treatment systems and ventilation f systems, and the location of monitoring points relative to effluent release points.

Therefore, we conclude thit the applicant's proposed radiological process and

} effluent monitoring systeos are acceptable.

Based on the foregoing evsluation, we conclude that the proposed radioactive waste treatment and monitoring sistems are acceptable. The basis for acceptance has been

conformance of the applicant' designs, design criteria, and design bases for the j radioactive waste treatment and monitoring systems to the applicable regulations

! and guides referenced above, as well as to staff technical po.sitions and industry f standards.

(

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11-20

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12.0 RADIATION PR0fECTION The Sundesert Preliminary Safety Analysis Report provides information on the occupa-tional radiation protection aspects of the plant design and on the health physics program. The basic considerations to be employed in the radiation protection design and program are set forth and include the radiation protection review process to be used in the design phase, to assJre that occupational radiation exposure will be as low as is reasonably achievable. Radiation source terms associated with.each of the plant systems are described. The shielning, ventilation, radiation monitoring systems and other radiation protection features to be incorporated into the plant design are explained and occupational radiation exposures to plant personnel are .

estimated. In addition, the health physics program is described. This information, .

and the responses to our requests for information, were ceviewed against the accept- I

]

an:e criteria in Sections 12.1, 12.2, 12.1, 12.4 and 12.5 of the Standard Review l

Plan.

lhe criteria used to determine acceptability of the radiation protection program are that doses to personnel will be maintained within the established limits of 10 CfR Part 20 and that the design and program features are consistent with the guidelines of Regulatory Guide 8.8, "Information Relevant to Maintaining Occupa-tional Radiation Exposures as low as Practicable (Nuclear Power ReactorsJ." Those portions of the new guidelines in Revision 2 to Regulatory Guide 8.8, which are dppIiCable to a construction permit review, have been addressed for sundesert. The remaining guidelines of Revision 2 will be addressed during the operating license stage of review. In eesponse to our requests for information, the applicant has described the improvements in design made for the purpose of assuring that occupa-tional radiation exposu es will be as low as is reasonably achievable.

1 The applicant has considered means to keep e>ternal and internal radiation expo-I sures to personnel, including both individual and total man-rem doses, as low as is reasonably achievable. Shielding will be designed to control radiation exposure such that, (1) doses to operating personnel and the general public will be less than those required by the applicable sections of 10 CFR Part 20 and 50; and (2) individual and total doses to plant personnel will be as low as is reasonably achievable and the applicant will follow, where appropriate, the guidance of Regu-latory Guide 8.8.

l On the basis of our review, we have concluded that the radiation protection program will provide reasonable assurance that doses to personnel will be less than the limits established by 10 CFR part 20, and maintained as low as is reasonably achiev-able, consistent with the guidelines of Regulatory Guide 8.8. Therefore, we conclude 1

12-1

that the Sundesert radiation protection program is acceptable. Details of our evaluation are discussed in the following sections.

12.1 Assuring That Occupational Radiatinn_Fxp g res Are As low As is Reasonably Achievable The San Diego Gas and Electric Company's management is committed to maintaining radiation exposures to its employees as low as is reasonably achievable. For the protectio') of its employees, the applicant subscribes to the as low as is reason-ably achievable philosophy set forth in Regulatory Guide 8.8, "Information Relevant to Maintaining Occupational Radiation lxp0sures as low as Practicable (Nuclear Power Reactors)," and Regulatory Guide 8.10, " Operating Philosophy for Maintaining Occupational Radiation Exposures as Lc+ n is Reasonably Achievaole," in the design and operation of all facilities utilizing radioactive materials or radiation sources.

The applicant's design guidelines with regard to keeping occupational radiation exposures as low as is reasonably achitvable include:

(1) Minimize contact maintenance and inspection of the steam generators; l (2) Minimize field run piping, and crud traps such as piping elbows; 1

(3) Minimize lengths of piping for radioactive materials; and (4) Provide sufficient space around valves and equipment for portable shielding.

The applicant is committed to a continuing radiation protection review process throughout the design phase for the purpose of assuring that occupational radiation exposures will be as low as is reasonably achievable. Changes and improvements resulting from such reviews have included:

(1) Rearrangement of pressure relief openings in the steam generator cubicles, in order to reduce radiation streaming to areas potentially accessible to personnel; (2) Rerouting of the residual heat removal system piping to provide for improved shielding; and (3) Improved !ayout of piping for 1adioactive fluids at containment penetrations, to provide for improved shielding.

Based on the information provided in the application and in the responses to our requests for information, we conclude t hat the applicant plans to design, con-struct, and operate the Sundesert plant in such a manner that occupational radia-tion exposures will be maintains 1 as low as is reasonably achievable, in accordance with Regulatory Guides 8.8 and 8.10 and with 10 CFR Part 20 and will be consistent with the acceptance criteria in Section 12,1 of the Standard Review Plan.

12-2

l l

l 12.2 Radiation Sources The applicant has provided design and expected concentration values of fission and l corrosion product activities in the reactor coolant system and in other reactor systems, and radiation source terms for eight energy groups for principal sources l within the containment. The core source oescription is similar to that given in RP-8A, " Radiation Shielding Design and Analysis Approach for Light Water Plants," a j topical re port by Stone & Webster, which we have previously reviewed and approved j as documented by letter to Stone & Webf-ter, dated April 4, 1975. For the source calculations, the assumptions included catration at 2958 megawatts thermal, a failed fuel fraction of 0.01, and a primary to secondary leak rate of 1370 pounds I per day, Estimated radioactivity leakage rates into the containment structure, turbine building, annulus building, and fuel building are provided, as well as radioactive airborne concentrations of the principal nuclides.

Fission products inventories and concentrstions in the radwaste system are esti- l mated on the basis of system design parameters; use is made of prior experience by j the applicant and the architect-engineer with similar plants. The assumptions and I procedures used by the applicant in estimating radiation source terms fave l been evaluated; and we conclude that the resulting estimates are reasonable and consistent with the acceptance criteria in Section 12.2 of the Standard Review Plan.

12.3 Radiation Protection Design features l

The applicant has addressed in detail the application of Sections 3.a through 3.r i in Regulatory Guide 8.8, Revision 1, to tt,e design of the Sundesert plant. Examples 4

l include:

(1) Individually shielded cubicles for radioactive equipment requiring servicing; (2) Selection of best valve products available, considering valve type, seat materials, and servicing conditions; i

(3) Routing of pipes and ducts containing radioactive fluids so as not to pass through occupied areas; (4) Routing of piping systems to avoid unnecessary sharp bends, pockets, and low points; (5) Permanent pipe flushing connections in liquid waste systems; and (6) Remote handling equipment for changing filters and placing them into shipping containers.

12-3

l Shielding wall thicknesses will be determined using basic shielding data. The shielding requirements are consistent with the American National Standard Institute (ANSI) Standard N101.6-1972, " Concrete Radiation Shields," and Regulatory Guide 1.69, " Concrete Radiation Shields for Nuclear Power Plants;" the shielding approach and methods used are similar to those described in Topical Report RP-8A. We consider the assumptions used for these shielding calculations to be conservative, and the proposed models and codes acceptable.

The applicant has provided six radiation zones as a basis for classifying occupancy and access restrictions for various areas. On this basis, maximum design dose rates are established for each Zone and used as input for shielding of the respective zones. For example, maximum design radiation levels in operating areas where l personnel are expected to be working without access restriction will be less than l two millirem per hour.

B Radiation protection concepts directed to keeping personnel exposures below regula-

, tory limits will be used throughout the design. To the extent practicable, major radiation sources will be in individually labyrinthed, shielded cubicles. Pipes l

and ducts will be routed through high-zoned, low access areas when practicable; shielding will be provided for pipe trenches and penetration.

i In order to maintain occupational radiation exposures as low as is reasonably

{

achievable, the design objectives of the ventilation system incluoe: ;

i (1) Assuring that concentrations of airborne radioactivity levels will be less than those given in Appendix B,10 CFR Part 20, Column 1, Table 1 in restricted areas, and less than those given in Appendix B, 10 CFR Part 20, Column 1, Table 2, in administrative areas; (2) Provision of containment atmosphere filtration capable of limiting iodine concentrations to the maximum permissible concentration within approximately 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months , and a containment purge system capable of reducing airborne radio-activity to acceptable levels prior to aad during extended personnel occupancy of the containment; l

l (3) Operation of the fuel building ventilation system in a once-through mode if airborria radioactive levels exceed a predetermined level; (4) Provisions to terminate filtering ventilation flow within the annulus, solid waste, and fuel taildings in the event of a high radioactivity alarm; and (5) System design such that filters containing radioactivity C. ue maintained with minimal doses to pe y nnel.

12-4

i The design features of the ventilation system are consistent with the acceptance I

criteria in Section 12.3 of the Standard Review Plan and, therefore, are acceptable.

The area radiation and airborne radioactivity monitoring instrumentation systems i and programs will be designed to:

(1) Measure radiation levels in, and exnaust from, areas where personnel are l expected to remain for extended periods of time, and where variable radiation levels may occur; (2) Be able to measure radiation levels from below design levels to the maximum dose rate for anticipated occupational occurrences; (3) Provide emergency power supplies in the event of power failure or postulated accidents; (4) Provide for instrument calibration routinely and after any maintenance work I performed on the instrument; and I

(5) Provide, for each detector, a local audible and visual alarm, variable alarm set points, and a local readout device; and a readout device in the control room equipped with alarms to indicate equipment malfunction, as well as alarms to indicate alert and high radiation levels.

l The objectives and location criteria for these monitoring systems are in conformance ]

with the acceptance criteria in Section 12.3 of the Standard Review Plan and, j therefore, are acceptable. '

12.4 Dose Assessment The estimates of annual man-rem exposure are based on conservatively assumed radia-l tion sources, design shielding, calculated design dose rates and manpower levels, l and take into account expected functions and occupancy times, and a working year of 2000 hours0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months . Based on expected dose rates and occupancy times, and taking into acccunt available experience from operating plants, the applicant estimates the average annual occupational radiation exposure at the Sundesert plant to be about 100 man-rem per unit for non-maintenance personnel, and a total of about 500 man-rem per unit for all personnel. The applicant projects that, because of significant design improvements, the major sources which have caused occupational radiation exposure at plants operating today will not cause significant doses at Sundesert.

the applicant estimates the total radiation exposure to construction workers on Unit 2, as a result of Unit 1 operation during Unit 2 construction, to be about 11 man-rem. The bases for the applicant's exposure estimates are reasonable, and l consistent with the acceptance criteria in Section 12.4 of the Standard Review Plan. l

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12-5 !

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12.5 Hea'th Physics Procram E

j lt is the applicant's policy that, througn radiation protection policy and prac-

} tices, occupational radiation exposure will be maintained within the limits of l 10 CFR Part 20 and as low as is reasonaoly ac.hievable, consistent with the provi-j sions of Regulatory Guide 1.8, "Persc 1 5 election and Training," and Regulatory

} Guides 8.8 and 8.10.

i 4

The Health, Safety, and Chemistry Supervisor, who will be responsible for the i administrative supervision of the Health Physics Program, will report directly to I

the Plant Superintendent. He will also be responsible for:

l (1) Directing the radiation protection program, to assure that radiation doses are l as low as is reasonably achievable; 4

l l (2) Training of personnel on radiation principles and procedures; 4

l (3) Planning and scheduling of monitoring and surveillance, and maintaining cur-l rent records; l i '

i l (4) Assuring adequacy of radiation protection equipment provided, and maintenance of adequate records; and 4 l l (5) Assuring adequacy of radiation and contamination control, receipt and handling !

of radioactive materials, and necessary decontamination procedures.

?

The health physics facilities will include office areas, a radiation laboratory, counting room, a whole-body counting facility, instrument and equipment storage l i rooms, per'onnel and equipment decontamination areas, change rooms, laundry facil-5 ities, anc cess-control check points. The laboratory will be furnished with appropri radiation instrumentation and facilities for routine analyses required l

} for pers" ,el protection, surveys and related health physics functions. Nec e s sa ry

] protectivt clothing and equipment will be provided in the clothing storage area and j in the personnel decontamination facilities.

l Based on the information provided in tne application, and the responses to our I

! requests for information, we conclude that the applicant plans to implement a health physics program that will meet the objectives of Regulatory Guide 8.8 and maintain occupational exposures as low as is reasonably achievable, in conformance

with the acceptance criteria of Section 12.5 of the Standard Review Plan, i

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l 13.0 C000C1 01 OPERATIONS l 13.1 Organizational Structure of Applicant; The San Diego Gas and Electric Company is responsible for the oesign, construction, quality assurance, testing, and operat ion of the Sundesert Nuclear Plant, Unit Nos. I and 2. Westinghouse Electric Corporation will be responsible for the ndClear steam supply systems. Stone & Wecster En,ineering Corporation will act as tre architect-engineer and constructor for the plant.

The engineering and design for the plant is the responsibility of the Vice-President, Project Management Division Peporting to him is the Sundesert Project Manager who has overall responsibility for the Sundesert plant. The Manager, Nuclear Department, who reports functiunally to the Sundesert oroject Manager, will be responsible for the operation of the plant. Quality assurance aspects are discussed in Section 17.0 of this report.

The proposed station organization for the operation of the Sundesert plant will consist of approximately 136 persons under the direction of the Plant Superintend-ent. Reporting to the Plant Superintendent will be: an Assistant Plant Superin-tendent; a Health, Safety and Chemistry Supervisor: and a Security Supervisor.

Reporting to the Assistant Plant Superintendent will be: an Operations Supervisor with a staff of approximately 60 persons responsible for the operation of the plant; a Maintenance Supervisor with a staff of approximately 38 persons respon-sible for the maintenance of electrical, mect,&1ical, and instrument and control systems; and a Plant Nuclear Engineer with a staff of approximately four persons responsible for plant performance and nuclear engineering. Reporting to the Health, Safety, and Chemistry Supervisor will be: a chemist with a staff of approximately four persons responsible for plant chemistry and radiochemistry; and a Health Physicist with a staff of approximately eight persons responsible for the health physics program. The Security Supervisor will be in charge of the plant security guards. The shift crew composition for tre operation of each unit will consist of six persons including one senior licensed operator and two licensed operators.

l The applicant has described its minimum qualification requirements for plant person-f nel. These qualification requirements meet those described in Revision 1 to Rece-latory Guide 1.8, " Personnel Selection ano Training," and are, therefore, acceptable.

The proposed technical support for the operation of the Sundesert plant will be provided crimarily by personnel reporting to the Manager, Nuclear Department.

I i

l 13-1 ;

1

v/e conclude that the proposed plant organization, the proposed qualifications of personnel, and the proposed plans for offsite technical support, are acceptable for the contruction permit stage of review, We f urther conclude that the San Diego Gas and Electric Company is technically qualified to design and construct the Sundesert Nuclear Plant, Unit Nos. I and 2, based on the following:

(1) Our review of the applicant's corporate and technical organization:

(2) The technical resources as embndied in the numbers and technical experience of I personnel assigned and available to the project; (3) The applicant's participation in nther similar nuclear projects (i.e., San Onof re Nuclear Generating Station, Unit Nos.1, 2 and 3);

(4) The Qualify Assurance Program as ilescribed in Section 17.0 of this report; and C

(5) The exchange of technical informdtion experienced in our meetings and corre-i spondence during the course of this review.

1 13.2 Training Program l The Supervisor of Nuclear Operations will have responsibility for the initial training program. Following plant startup, the Plent Superintendent will assume i responsibility for replacement and requalification training.

l The applicant states that a training program will be established to provide plant personnel with sufficient knowledge and operating experience to start up, operate, and maintain the plant in a safe and ef ficient manner.

The applicant also states that the program will fully meet the recommendations of Regulatory Guide 1.8 and American National Standard Institute (ANSI) Standard N18.1-1971, " Selection and Training of Nuclear Power Plant Personnel." The program will be developed by the San Deigo Gas and Electric Company with principal assistance from the Westinghouse Electric Corporation.

Training for personnel wno will De !icensed will include: Basic Nuclear Course; Operating Plant Observation; Simulator Operation; Design Lecture Series; Plant Systems Lecture Series; Health Physics and Radiochemistry Course; and Onsite and On-The-Job Training.

Maintenance and technical staff personnel will receive specialized training in their particular fields. Station personnel will also receive training in security 13-2

_ _. . _ . _ _ . - . - . _- m -. . _ ._m . _. ..

and emergency plans, administrative procedures and radiation protection, as l appropriate.

I 1he information submitted relative tc the training program is satisfactory at the construction permit stage of review to give reasonable assurance that qualified individuals will be available for the preoperational test program, for operator licensing, and for fuel loading.

13.3 Emergen:y Planning We have evaluated the applicant's preliminary plans for coping with emergencies utilizing the acceptance criteria set forth in Section 13.3 of the Standard Review 1 i

Plan. The material reviewed cover s that tro ough Amendment No. 12 to the Preliminary i I

Safety Analysis Report and is primarily contained in Section 13.3 of the Preliminary Safety Analysis Report and in the responses to our requests for information.

We find that the applicant's preliminary plans are acceptable with respect to the criteria in Section 13.3 of the Standard Review Plan. In particular, we have determined that the plans:

(1) Conform to the requirements of 10 CFR Part 50, Ap)eadi.- E, Part 11; (2) Provide consistency with the facility design features, analyses of postulated accidents, and the characteristics of the proposed site location; and (3) Provide reasonable assurance that appropriate protective measures can be taken in the event of a serious accident within and beyond the s ite boundary. l 1

The bases for these findings are summarized below.

The plant Superintendent has t,een designated as the individual responsible for control of emergency efforts at the Sundesert facility. In the event of his ab- i l sence, a specific line of succession is provided, including the Shift Supervisor, i to assure that continuous coverage is provided by a senior plant staff member. To i

the extent required, the onsite emergency organization can be augmented via resources available from the applicant's corporate staf f, local service organiza-tions, California and Arizona State agencies, and the Federal government. Both primary and backup notification systems will be available for assuring prompt and effective communication during emergencies. Therefore, we conclude that the appli-cant has met the requirements in subsection II.A. of Appendix E to 10 CFR'Part 50 regarding organizational structure and means of notification for coping with emergencies.

The applicant has made significant progress towards establishing arrangements with the appropriate local, State, and Federal Agencies for assistance in dealing with 13-3

i i

j emergencies. Meetings have been held between the applicant and representatives of the agencies listed in fable 13.1. In addition. letters of intent to participate

]

l in the planning process with the applicant have been obtained from the Riverside i County Office of Emergency Services wnich will be the lead agency responsible for the coordination of local support plans, and from the California State Office of j Emergency Services which is the agewy de<.ignated by the Governor in Executive j Order B-6-75 as having responsibility- for radiological emergency planning at the j State level. Also, the Arizona Atomic Energy Commission has offered, in writing, to cooperate with the applicant and otner local, State and Federal agencies in the ,

development of offsite emergency response plans. A schedule, which we find com- l patible with the applicant's tentative fuel lot. ding date, has been developed for I f the preparation of the various emergency plans and procedures whic'. will be re- j j quired in support of the Sundesert facility.

1 4

! As part of our review, we have consulted with the Commission's Office of State l

l 4 Programs regarding the emergency preparedness responsibilities of the various i

Arizona and California State and local agencies as they relate to the picposed i Sundesert facility. We find it relevant that State and local agency represent- !

l atives from both California and Arizona have already participated in the Federally j sponsored training programs in such areas as radiological response planning, coor-

{ dination, and operations.

1 i

)

l Therefore, we conclude that the applicant has met the requirements in subsection i 11,8 of Appendix E to 10 CFR Part 50 regarding contacts and arrangements with f government agencies for emergency planqing.

l l In plant instrumentation will provide identification and assessment information for l radiological emergencies. This information together with protective action levels will be used for initiating onsite and/or offsite protective measures. Two emer-j gency operations centers will be provided by the applicant, one of which will be l located at the visitors center and the other at an as yet unidentified location.

{ In any case, the onsite direction of the emergency effort could be accomplished l from the control room which is designed for continuous occupancy during the course I cf an accident.

! Onsite measures for plant personnel will include respiratory protection, isolation t

l f

of ccntaminated areas, and evacuations of in plant areas. Additional onsite measures, j 'ecessitated by the applicant's obligations arising from emergencies having potential i o ' site consequences, have been established as part of the preliminary planning l r cess. These measures include the assessment of the riagnitude of the release and j the potential offsite effects. Also included are measures to assure that local, State, and Federal agencies are alerted, provided with technical information regarding the facility status, and provided with advice regarding offsite protection actions, 13-4

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

TABLE 13.1 ORGANIZATIONS AND AGENCIES - EMERGENCY PLANNING FOR THE SUNDESERT NUCLEAR FACILITY l

1. U.S. Department of Energy

2. Arizona Atomic Energy Commission ,

3. Arizona Division of Emergency Services )

4 California Office of Emergency Services

5. California State Highway Patrol

6. California Department of Transportation (CALTRANS)

7. Riverside Lounty Office of Emergency services

8. Riverside County Department of Health

9. Riverside County Sheriff's Office

10. Imperial County Office of Emergency Services

11. Maricopa County Civil Defense and Emergency Services

12. Yuma County Emergency Services 1 13. Yavapai County Emergency Services 14 Blythe Police Department

15. Blythe Fire Department

16. Palo Verde Fire Department l

l l

+

13-5

i i

l i

! tor of fsite communications and logistical support of emergency activities, an

Emergency Operations Center will be designated by the responsible local agency; the j most likely location being the Sheriff's office in Blythe, California. Potential

] measures to be tMan to protect t he pohlic haalth anel safety in the plant environs i will include prompt public notification, environmental monitoring, area access control,

) evacuation including relocation and reassembly, and public health maintenance. The j specific response actions to be provided by the numerous agencies involved in support l;

of the aforementioned measures to protect the public, are detailed in the Preliminary 3 Safety Analysis Report.

5 j Therefore, we conclude that the applicant has met the requirements in suosection j ll.C. of Appendix E to 10 CFR Part 50 regarding the measures to be taken in the j event of an emergency.

{

i 1 1

The proposed design of the plant includes a first aid room and decontamination l

j facilities in the health physics area of each unit. Additionally, a medical treat-j' ment room is proposed to be located in the administration building for the care of injured personnel, 11 the onsite facilities are insufficient for the required treatment, the injured individual will be moved to a hospital. Arrangements will

be made with a local ambulance service which maintains two such vehicles for trans-j porting ill or injured persons. Backup capability will be available from the local l fire fighting units which operate rescue trucks that could be used for such emer-j gency purposes. Additionally, private vehicles would be available if necessary.

Therefore, we conclude that the applicant has met the requirements in I j subsection 11.0 of Appendix E to 10 CFR Part 50 regarding cnsite provisions for i 1

treating emergencies.

4 i

i Preliminary arrangements for the receipt and treatment of injured plant staff have been pursued with the Palo Verde Community Hospital in Blythe, California. Several j meetings have already been held with the hospital administrator and the board with j respect to the potential requirements of the Sundesert plant. Plans will be developed j with the aid of the applicant's medical consultant for hospital improvements. I l

l Considerations will be given to specialized facilities for handling radiological j emergencies include a designated area with standard emergency room equipment, valved piping and holding tanks, showers, portable radiation monitors, stripable

floor paint, and temporary flore coverings. Although specific arrangements have I l not as yet been made with another medical center for extensive treatment of potential

! . radiation overexposure victims, it appears feasible that such arrangements are l likely with the Scripps Memorial Hospital in La Jolla. We consider that there j should be no impediment to reaching an acceptable agreement since the hospital has a similar agreement with General Atomics for the emergency treatment of its employees.

Therefore, we conclude that the applicant has met the requirements in subsection j II.E. of Appendix E to 10 CFR Part 50 regarding offsite provisions for treating j emergencies.

13-6

All Sundesert plant personnel involved as members of the onsite emergency response team will receive training in the follewing areas; Health Physics, First Aid, Dacontamination Practices, and Emergency Practices. All other members of the plant staf f will receive either extensive training in nuclear science, including health Preliminary physics and emergency actions, or plant indoctrination in these areas.

plans also include provisions for training local support services personnel in radiation protection practices and emergency procedures. The physician retained by the applicant will also receive training at the Oak Ridge REAC/T5 program relating to the treatment of radiation injuries. Periodic drills will also be conducted to test the adequacy of the emergency procedures, to test the emergency equipment, and to assure that emergency organization personnel are familiar with their duties.

Offsite support agencies will be invited to participate in the annual drill involv-ing the entire plant. Therefore, we conclude that the applicant has met the require-ments in subsection II.F. of Appendix E to 10 CFR Part 50 regarding training programs for coping with emergencies.

The various plant features to assure evacuation capability, in addition to design features to permit safe shutdown, include radiation emergency alarms, evacuation and fire alarms, self powered emergency lighting, diverse communications systems, and adequate evacuation routes. Capability for facility reentry following an accident will be supported by the establishment of two emergency operations centers equipped with emergency supplies, radiation monitoring instrumentation, and communi-cations equipment. Either of these centers could be used for directing emergency efforts towards reentry and recovery. Therefore, we conclude that the applicant has met the requirements in subsection II.G. of Appendix E to 10 CFR Part 50 regarding facility features to assure the capability for plant evacuation and facility reentry.

An additional criterion identified in Section 13.3 of the Standard Review Plan deals with the consistency of the preliminary emergency planning with respect to the facility design, site location, and potential accident consequences. This aspect of our review focussed primarily upon the extent to which the emergency plans for the Sundesert facility should provide for the implementation of protec-tive and related measures beyond the site boundary. In this regard, information was provided by the applicant in the Preliminary Safety Analysis Report in the form of projected dose levels resulting from the most serious design basis accident as a function of both time following onset of the postulated release and of distance from the point of release. This information shows that the projected inhalation doses for thyroid exposure could exceed the Protective Action Guide level of 25 rem in EPA-520/1-75-001, " Manual of Protective Action Guides and Protective Actions for Nuclear Incidents," for which the U.S. Environmental Protection Agency recommends evacuation as a protective action, at a distance of six miles from the plant within eight hours. On these grounds we required the applicant to include provisions for protective actions in his plans extending at least to this distance. The applicant's preliminary plans and planning objectives reflect a commitment to this end.

13-7

In addition, we have also performed an assessment of the potential efficacy of evacuation as a protective action within this region and for the postulated accident.

This assessment shows that the sequence of events, including accident identification, notification of authorities, warning of the publir, and actual evacuation by the public, could reasonably be accomplisr4d within a time frame such that significant benefit to individual members of the public would be gained via reduction or elimination of the radiological exposure which otherwise might result if such measures were not taken. Based in part upon these results, as well as upon the preliminary plans and arrangements oescribed above, we consider that there is reasonable assurance that appropriate protective measures can be taken in the event of a serious accident within and beyono the site boundary.

We will evaluate the applicant's final plans for coping with emergencies, to be submitted in accordance with 10 CFR Part 50.34(b), at the operating license stage of review.

13.4 Review and Audit The applicant has committed to a program for the review and audit of plant operations that will meet Section 4 of American National Standard Institute (ANSI) Standard N18.7-1972, " Administrative Controls for Nuclear Power Plants." This program meets the recommendations in Regulatory Guide 1.33, " Quality Assurance Program Requirements (Operation)." A detailed review of the applirant's program will be performed during the operating license stage of review.

13.5 Plant Procedures All safety related operating, maintenance and testing activities will be conducted in accordance with approved, written procedures. Procedures will conform to the guidelines of ANSI Standard N18.7-1972 and Regulatory Guide 1.33. The preparation of these procedures will be completed at least six months prior to fuel loading.

The information submitted in the Preliminary Safety Analysis Report relative to the plant procedures is satisfactory for the construction permit stage of review _

13.6 Industrial Security The applicant has submitted information regarding its plans for developing a secu-rity program for the Sundesert facility. The applicant states that the security program will conform with the guidance contained in Regulatory Guide 1.17, " Protection of Nuclear Power Plants Against Industrial Sabotage," and ANSI Standard N18.17-1973,

" Industrial Security for Nuclear Power Plants." In addition, the applicant has committed to emphasizing physical security throughout the design process and imple-menting such design features as compartmentation of vital equipment wherever practical to augment or replace administrative security controls.

I i 13-8 l

l

On February 24, 1977 the Commission puolished new requirements for the physical protection of nuclear power plants against a:ts of sabotage (10 CFR Part 73.55).

This new rule does not require applicants for construction permits to demonstrate compliance at this stage but does require such demonstration at the operating license stage. The applicant has submitted a plan which describes, in , preliminary manner, a security program which will De designed to comply with the new rule. As a result of our review of the applicant's preliminary plans for physical security, we conclude that a satisfactory planning base has been described upon which a complete security program can be developed to demonstrate compliance with the new regulations and to pro.ide an acceptable level of physical protection to this site at the appropriate tin.e. We will continue tc work with and provide guidance to the applicant to assure compliance with 'O CFR Part 73.53 at the operating license stage of review.

l 13-9 i

14.0 INITIAL TEST PROGRAMS We have completed the review of the information provided on the initial test pro-gram in support of the construction permit application for the Sundesert plant.

Our review wat conducted in accordance with Section 14.1 of the Standard Review Plan and included:

(1) Evaluation of the scope of the applicant's test program including the respon-

' sibilities and qualifications of participating organizations, the general testing objectives, the divisions between major phases of the test program, the administrative controls gcVerning the test program, and the extent to which the test program would verify the functional adequacy of the facility; (2) Evaluation of the testing proposeo for unique or first-of-a-kind design fea-tures for the facility; (3) Evaluation of the applicant's plans relative to utilization of Pagulatory Guide 1.68, "Preoperational and Initial Startup Test Programs for Water-Cooled Power Reactors," and other regulatory guidcs applicable to testing in the formation of its test program; (4) Evaluation of the applicant's plans for review and utilization, where applic-able, of operating experience from other reactors in the development of its test program; (5) Evaluation of the applicant's test program schedule to estabilsh that suf-ficient time for testing is planned and that the schedule is compatible with the schedules for the hiring and training of plant personnel; (6) Evaluation of the applicant's plans to utilize plant operating and emergency procedures to the extent practicable during preoperational testing; and (7) Evaluation of the applicant's plans to augment the station staff, as neces-sary, during the test program.

i On the basis of this review, we have determined that the San Diego Gas and Electric Company has committed to conduct a comprehensive initial test program for the facility. We conclude that the program described by the applicant is acceptable l

and will provide for further verification of the functional adequacy of the facility.

l 14-1

15.0 ACCICENT ANALYSES 15.1 General

15. 1 Classification of Events The applicant has performea and we have reviewed safety analyses to evaluate the capability of the Sundesert plant to withstand normal and abnormal operational

' transients and a broad spectt um of postulated accidents without undue risk to the health and safety of the public. Ihe events considered inCluce all relevant types discussed in Regulatory Guide 1.70, " Standard Format and Content of Safety Analysis Por' orts for Nuclear Power Plants," Re- Mon ? The postulated events have been Cidssified Dy the applicant in accord nCs with the following Criteria:

(1) Condition '- Normal Operation and Operational Transients--events which are expected frequently or regularly in the course of power opera-tion, refueling ms mtenance, or maneuvering of the facility.

(2) Condition II: Faults of Moderate Frequency--events that at worst result in a 3

reactor trip with the facility being capable of returr to operation.

(3) .cndition III: Infrequent Faults--events that are very infrequent during the life of the facility and may result in fuel damage which could preclude the resumptien of operation.

(4) Condition IV: Limiting Faults--events which are not expected to occur, but are postulated because their consequences would include the potential for release of significant amounts of radioactive material.

The applicant's classification of events analyzed is itemized in Table 15.1.

/

15.1.2 Input Parameters for Transient and Accident Analyses We have reviewed the assumptions and parameters employed in the transient and accident analyses. The mathematical models and methods used by the applicant have been previously reviewed and found acceptaDie by the staff unless otherwise noted in this report. Reactor protection system trip set points and the assumed trip delay times used in the analyses are tabulated in Table 15.2. And, as stated in Section 4.2.3 of thls report, the insertion rod time used in the transient analyses was 2.3 seconds to reach 85 percent of the rod travel. These values are suitable provided that they remain conservative with respect to tha set points finally 15-1

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

TABLE 15.1 I

l CATEGDRIES OF TYPICAL TRANSIENTS AND FAULTS l

l 1

l Condition I

(. Reactor startup

! Re6Ctor shutdown l Refueli.'O operations Condition II Uncontrolled control roa assembly bank withdrawal while the reactor is suberitical

! or at power, Partial loss of forced reactc r coolant flow Startup of an inactive reactor coolant 1000 l

! Turbine trip Loss of normal feedwater Loss of offsite power Uncontrolled boron dilation Control rod assembly e ;11gnment Excessive load increase Excessive heat removal due to feedwater system malfunctions Accidental depressurization of the reactor coolant system Accidental depressuri2ation cf the main steam system Inadvertent operation of the emergency core cooling system during power cperation Condition III Improper loading of a fuel assembly Complete loss of forced reactor coolant flow Single control rod assembly withdrawal at full power l Waste gas decay tank rupture Loss of reactor coolant from small break Condition IV Control rod ejection fuel handling accident Steam generator tube rupture Major secondary system pipe rupture Reactor coolant system rupture (LOCA)

Single reactor coolant pump locked rotor 15-2 t__________ _ _ _ . _

1 l

TARif }) 2 TRIP POINTS AND TIME DELAYS TO TRIP ASSUMED IN ACC1 CENT ANAlv5ES Limiting Trip ,

Point Assumed Time Delay Trip 8 anc ti on in Analyses _, (Seconds),

Pov 'r Range High Neutron Fica, High Setting !18 percent 0.5 Pcwer Range High Neutron Flux, Low Setting 35 pe :eit 0.5 i 1

Hign Neutron Flux, P-8 74 percent 0. ,

Osertemperature Delta T variable (see figure 15.0-1 of Sundesert Preliminary Safety Analysis Report) 6.0*

Overpower Delta T Variable (see Figure 15.0-1 of Sundesert Preliminary Saf ety Analysis Report) 6.0*

High Pressurizer Pressure 2410 pounds per souare inch gauge 2.0 Low Pressurizer Pressure 1845 pounds per square irich gauge 2. 0 Low Reactor Coolant Fio.

(From Loop Flow Detectors) 87 percent loop flow 1.0 Reactor Coolant Pump Undervoltage Trip 68 percent naminal 1.5 Turbine Trip Not applicable 1.0 Low-Low Steam Generator 37.5 percent of narrow range Level level span 2. 0 High-High Steam Generator 93 percent of narrow range 2.0 Level--Produces Feed *ater level span Isolation and Turbine Trip f

" Total time delay (including resistance temperature detector bypass lono fluid transport delay, ef fect of bypass loop piping thermal capacity, resistance tem,9 Lure detector time response, and trip circuit channel electronics delay) from the time the temperature difference in the coolant loops exceeds the trio set point until the rods are free to fall.

I 15-3

(

4 4

i i

1 l implemented, fully accounting for all sensor and process delays and uncertainties.

! Taerefore, at the operating license stage of review, we will require that the i applicant justify the instrument errors and time delen assumed in the analyses.

l The events initiated at full power were assumed tt, start at the nuclear steam j supply system thermal design power of 2785 megawatts plus an allowance of two percent for errors in the steady-state core power. Also, the appli: ant has provided

{ analyses which assumed a nucle e steam supply system thermal power output of 2910 j megawatts plus an allowance for er:or where the engineered safety features equipment I

would be involved. These latter analyses were performed for the following postulated i

r events: loss of nonemergercy alternating current power to the station auxiliaries; l loss of normal feedwater flow; feedwater system pipe break; steam generator tube i

rapture; and loss-of-coolant accidents resulting from postulated small and large j reactor cool nt ystem <;ipe ruptures. The assumed nuclear steam supply system

! thermal ;mwer 6 put 'evels used in the Sundesert transient and accident analys6s dre in aC % rda v e with the recommendations of Regulatory Guide 1.49, " Power Levels f of Nucleai Pcwer dants," and are, therefore, acceptable.

i The appli. ant has selected the most adverse conditions of core life with respect to l

4 reactivi' / coef ficients (moderator temperature coef ficient and the Ocupier co+ fi-f cient), c mtrol rod worths, and lordi power peaking factors. The applicant has I stated thit no credit was taken for norsafety grade systems to mitigate the con-I t sequences of any postulated accident presented in Chapter 15 of the Preliminary

, Safety Analysis Report. Furthermore, for some transients, analyses were performed i bod. with and without nonsafety-relatec wtrol system operation to determine the worst case. The above assumptions useo in ie analyses are acceptable, a

At our request, a failure mode and effects analysis of the various plant responses j to the Chapter 15 events was submitted by the applicant. The protection sequence h

4 diagrams for each event provide the following: (1) the safety systems reoufred to function to provide the sa bly actions necessary for mitigating the consequences of j the transient or accident during any plant operating state; (2) all operator actions

! required until the reactor achieves the final stabilized condition; and (3) safety-1 related information readouts and controls to be utilized by the operator to analyze and control the transient or accident. We have reviewed these protection sequence 1 diagrams and found them acceptable for the construction permit stage of review.

] In the analyses of the transients and postulated accidents, the applicant takes no j credit for manual actions to terminate the transient or accident until 10 minutes i after the event is initiated. This assumption is generally applied to Westinghouse J

i 1

plants. We are developing a generic position regarding the credit that can be taken for manual actions as a function of the information available to the operator

. to properly diagnose the transient, the anticipated number of manual actions, and i the severity of the event. At the operating license stage of review, we will a

1 1

1

! 15-4 3

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

require that the applicant reevaluate the transients and postulated accidents using the criteria that will develop f rom our generic review of this matter, 1

l Fina'iy, the effects of a new fuel rod pressure design criterion have been considered I

'sr the safety analysis of Condition III and IV events. These effects are addressed in Westinghouse Topicai Report WCAP-8963, " Safety Analysis for Revised f uel Rod Internal Pressure Design Bases." which we nave reviewed and found acceptable as documented by letter to Westinghouse, dated May 19, 1978. We have concludtd that I the increased fuel rod pressure will not resJ t in a significant number of addi-tional departure from nucleate boiling events d ving Condition III and IV events. l Based on the above evaluation, we conclude that the input parameters used in the l analyses of the various transients and postulated accidents are suitably conserva-tive for the construction permit staQe of review.

15 l.3 Analytical Techniques The analytical techniques used in the evaluation of the various transients and postulated accidents for Sundesert, hich are described in various topical reports incorporated by ref erence. are normally reviewed on a generic basis. The principal dnalytical techniques used for the specific events evaluated in the following sections of this report are described in the following Westinghouse Topical Reports: j (1) WCAP-7213, "The TURTLE 24.0 Diffusion Depletion Code." 1 (2) WCAP- 688, "An Evaluation c' t.ne Rod Ejection Accident in Westinghouse Pressur aed Water Reactors Using Spacial Kinetics Methods."

(3) WCAP-7907, "LOFTRAN Code Description."

(4) WCAP-7903, "FACTRAN - A FORTRAN IV Code for Thermal Transients in a Q F el Rod."

(5) WCAP-7956, "THINC-IV - An Improved Program for Thermal And Hydraulic Analysis of Rod Bundle Cores."

(6) WCAP-7979, " TWINKLE - A Multi-Dimensional Neutron Kinetics Computer Code."

(7) WCAP-9226, " Reactor Core Response to Excessive Secondary Steam Releases."

(8) WCAP-9230, " Report on the Consequences of a Postulated Main Feedline Rupture."

Our review of Topical Reports WCAP-7213, WCAP-7588, WCAP-7956 and WCAP-7979 has been completed and the analytical techniques in these reports have been found acceptable as documented in letters to v/estinghcuse, dated July 25, 1974, 15-5

August 28, 1973. April 19, 1978 and July 29, 1974, respectively. The status of our

. review cf the analytical techniques presented in the remaining topical reports is as follows:

4 l (1) WCAP-7907 and WCAP-7908, LOFTRAN and F ACTRAN - The LOFTRAN computer program is described in WCAP-7907 and is used to evaluate various systen transients and l accidents. This code has been under review by us for some time. The calculations

} performed for Sundesert have been based upon a version of LOFTRAN that contains some modi'ications to the documented version in WCAP-7907. The modifications j to the code were submitted by letter, dated May 26, 1978, from Westinghouse, j and include the addition of multi loop capability and an improved pump model

! tc allow calculation of pump coastdown transients. The FACTRAN computer i program is described in WCAP- 7'JOS and 's used to perform f uel rod heatup f calculations through a single plaqe.

i l Our review of these reports has progressed to the point that there is reasonable 4

]

assurance that the conclusions based on these analyses will not be appreciably

{ altered by the completion of the analytical review. If the final approval of j

LOFTRAN and/or FACTRAN indicates that any revisions to the analyses are required,

] this information shall be included in the Sundesert Final Safety Analysis

{ Report.

l The LOFTRAN code submittal did not discuss the verificatiJn o' the code rela-1

tive to experimental evidence. It is likely that we Wi6 tquire that certain

] plant transient tests be performed for code verifica; ion. Ttase tests will be required during startup for Sundesert unless suitable inf o+ mation is available j from another source.

t l (2) WCAP-9226 and WCAP-9230, Main Steamline and Ieedline Breaks - Westinghouse has

{ recently submitted two topical reports which present its analysis methods and l sensitivity studies for postulated main steamline and feedline breaks. This l information is contained in WCAP1 226 for steamline breaks and WCAP-9230 for l feedline breaks.

I

The review of these topical reports was,recently initiated and is scheduled i for completion during 1979. This schedule will allow the applicant to revise l the analysis and to modify the design or proposed opet ag requirements, if l necessary, in the Sundesert Final Saf ety Anr. lysis Report.

I i

{ Generic submittals for methods integration have not M n received for the steam i generator tube rupture and reactor coolant pump lock;s rotor or broken shaft accidents j and for the various transients analyzed as Condition 11 and 111 events in Chapter 15 of the Sundesert Preliminary Saf ety Analysis Report. We have requested Westinghouse to submit a detailed description of the analysis methods used for these accidents and transients. The requested information should be submitted on a schedule that 15-6

l will allow the applicant to revise the analysis and modify the design, if necessary, in the Sundesert Final Safety Analysis Report.

Based on previous acceptable analyses for plants similar to Sundesert, on compari-son with other industry models, on independent staf f audit calculations, and on previous startup testing experience, we conclude that the analytical methods for Suncesert are acceptable for the construction permit stage of review.

15.2 Aoderate Frequency Transients A number of plant transients can be expected to occur with moderate frequency during a plant lifetime as a result of equipment malfunction or operator error in the course of refueling or power operation. Such transients meet the criteria of Condition II in the evaluation and classification presented by the applicant.

We have compared the applicant's Condition II events listed in Table 15-1 of this report to typical anticipated moderate frequency events normally considered for safety reviews as specified in the Standard Review Plan. We agree with the appli-cant's listing of events under Condition 11 except for the " Complete loss of forced Reactor Coolant Flow" event. The applicant classified this event as a Condition fil infrequent event. According to the Standard Review Plan, this event should be considered as a Condition II moderate frequency event and we have evaluated the consequences accordingly.

Our basic acceptance criteria for the review of the moderate frequency transients are as follows:

(1) Pressure in the reactor coolant and main steam systems shall not exceed 110 percent of design pressure (Section III of ASME Boiler and Pressure Vessel Code).

(2) Clad integrity shall be n.uintained by assuring that the minimum departure from nucleate boiling ratio of 1.30 throughout the transient will satisfy the 95/95 criterion. (The 95/95 criterion proviues a 95 percent probability, at a 95 percent confidence level, that no fuel rod in the core experiences a departure from nucleate boiling.)

(3) Transients will not lead to more serious plant conditions (assuming other independent faults have not occurred).

All of the transients which are expected to occur with moderate frequency.can be giouped according to the following plant process disturbances: increase in heat removal by the secondary system; decrease in heat removal by the secondary system; core flow decrease; core reactivity increase; reactor coolant inventory increase; and reactor coolant inventory decrease. The results of the analyses for these 15-7

I 1

j transients, which are discussed in the following sections, show that the above acceptance criteria have been met.

4 15.2.1 Increase in Heat Removal by the Secondary System i

j An unplanned increase in heat removal by the secondary system that F >6

{ ed to occur with moderate frequency can either be caused bu 1

pressure regulator malfunctions, or by an excessive ir

flow or by the inadvertent opening of a steam ger' ..

All of these postulated transients have been r evalu-ated by the applicant using mathematical mode' '

Topical

Reports WCAP-7907, WCAP-7908, and WCAP-795r 3

} The results of the analyses presented i y .,afety Analysis Report 5

indicate that the most limiting event o core thermal margins was the

{ increase in feedwater flow. Similt , limiting event for pressure in the j reactor coolant system was the excessive int 5 in second:ry steam flow. The I

results of the analyses indicate that no significant pressure excursion would occur i for this category of events. The results of the analyses for the transients evalu-

{ ated also show that cladding integrity would be maintained since the minimum departure l from nucleate boiling ratio did not decrease below 1.30 and the maximum pressure l within the reactor coolant and main steam systems did not exceed 110 percent of the l design pressures.

1 1 15.2.2 Decrease in Heat Removal by the Secondary System j A number of plant transients can result in an unplanned decrease in heat removal by 5

1 the secondary system. Those that might be expected to occur with moderate frequency )

} are turbine trip, loss of external load, steam pressure regulator malfunctions,

! loss of condense ~ vacuum, loss of nonemergency alternating current power to the i

1 station auxiliaries, and loss of normal feedwater flow. All of these postulated 4

4 transients have been reviewed. The transients were evaluated by the applicant ,

1 i using a mathematical model described in Westinghouse Topical Report WCAP-7907.

lt 1

It was found that the most limiting, with regard to core t ermal margins and pres-sure within the reactor coolant and main steam system, was the loss of normal j feedwater flow caused by a loss of offsite power.

l

For the loss of feedwater event analysis, no credit was taken for the pressurizer 4

and the steam generator power-operated relief valves. The results of the analysis

, of the loss of normal feedwater flow transient show that cladding integrity would i

be maintained since the minimum departure from nucleate boiling ratio did not

decrease below 1.30 and the maxinwm pressure within the reactor coolant and main

] steam systems did not exceed 110 percent of their design pressures, e

15-8 i_ - - ._- -

-15,2.3 Decrease in Reactor Coolant System Flow Rate Several types of plant occurrences can result in an unplanned decrease in reactor coolant flow rate. The ones that mignt be expected to occur with moderate fre-I quency during the life of the plant are a partial loss-of-coolant flow caused by reactur coolant pump trip (s), or a complete loss of forced reactor coolant flow that may result from the simultaneous loss of electrical power to all pumps. rsr the partial loss of forced reactor coolant flow transient evaluation, two cases have been analyzed; a loss of two pumps with three loops in operation and a loss of one pump with two loops in operation. The two cases analyzed for the complete loss of forced reactor coolant flow were the loss of three pumps with three loops in operation and the loss of two pumps with two loops in operation. All of the tran-sients were evaluated by the applicant using mathematical models described in Westinghouse Topical Reports WCAP-7907, WCAP-7908, and WCAP-7956.

These postulated transients have been reviewed and it was found that the most limiting event with regard to core thermal margins and pressure within the reactor coolant system was the complete loss of reactor coolant flow transient. The results of the analysis of the complete loss of reactor coolant flow transient show that cladding integrity would be maintained since the minimum departure from nucleate boiling ratio did not decrease below 1.30 and the maximum pressure within the reactor coolant and main steam systems did not exceed 110 percent of the design pressures.

15.2.4 Core Reactivity Insertion There are a number of transients that may occur with moderate freque ~y which can cause unplanned core reactivity insertions. The types of transients analyzed in this category are: startup of an inactive reactor coolant loop at an incorrect (lower) temperature, which would result in increased core flow and a lower inlet temperature and thereby an increase in core reactivity; uncontrolled control rod assembly bank withdrawal from a subcritical or low power startup condition, or at power; and an uncontrolled boron dilution incident. The mathematical models used in the evaluation of core reactivity insertion events are described in Westinghouse Topical Reports WCAP-7907, WCAP-7908, VCAP-7956, and WCAP-7979.

We have reviewed the uncontrolled control rod assembly bank witdrawal transients initiated from suberitical or at low power startup conditions and at power condi-tions. For the initiation of this transient at subcritical or low startup power conditions, the results yield a large margin to departure from nucleate boiling and show that the departure from nucleate boiling ratio would always be greater than 1.30. The more significant decreases in core thermal margin (departure from nucleate boiling ratio) occur from an uncontrolled control rod assembly bank withdrawal at power, A complete range of at power transients was evaluated. Parameters investi-gated were initial power levels ranging from low (10 percent) to full power, a 15-9

- _ _ - _ ~ ._. .

4 I

l a

j range of reactivity insertion rates, ord consideration of minimum and maximum fuel j and moderator reactivity feedback effects. The results show that fuel cladding j integrity would be maintained since the minimum departure from nucleate boiling j ratio did not dacrease halow 1.30. Tne high neutron fluw and over-temperature delta T trip channels will provide adecuate protection over the entire range of possible reactivity insertion rates.

j Reactivity can also be added to the core by adding primary grade water to the j reactor coolant system via the makeup portion of the chemic,.1 and volume control j system. Various chemical and volume control syftem malfunctions, which could Icad j to an unplanned boron dilution incident, have been reviewed. The applicant has

{ analyzed postulated boron dilution transients, starting from plant conditions of 4

tartup, power operation (automatic and manual.), hot standby, cold shutdown, and i

t refueling. For these transients, the following additional acceptance criterion was 1

imposed. From the time an alarm makes the operator aware of unplanned moderator

dilution, the following minimum time intervals must be available before a complete i

j loss of shutdown margin occurs:

j (1) During refueling: 30 minutes.

\

h j (2) During startup, cold shutdown, hot standby, and power operation: 15 minutes.

1

As a result of the above requirements, Valves FCV-1138, FCV-Il4A, 8439, 8441, and l 8454 in the chemical and volume control system will be locked closed during refuel-ing operations to minimize the potential for baron dilution through the chemical l

and volume control system to acceptably low values.

4 4.

We requested the app icant to evaluate the potential for a boron dilution incident 4

caused by dilution soirces other than the chemical and volume control system. The applicant identified aid evaluated the most limiting alternate source of uncon-trolled boron dilution during refueling. This would be the inadvertent opening of

) a valve in the boron thermal regeneration system causing depleti u of boron in the j reactor coolant system.

a

?

l 1

The results of the events analyzed show that an operator would have 84.4 minutes to j take corrective action if the incident were to occur during refueling; 16.5 minutes j if at cold shutdown; 16.7 minutes if at hot standby or startup; and about 56 minutes 4

if at power in manual or automatic control mode. All of these predicted time j intervals meet the above acceptance criterion. The most severe unplanned boron dilution incident would occur at power and would result in a minimum departure from nucleate boiling ratio of 1.43 and reactor coolant and main steam system pressures 4

of less than 110 percent of design.

3 The apolicant has stated that, with the proposed Sundesert design, a baron dilution event with the sodium hydroxide tank as the source, cannot occur. We will review d

, 15-!0

the details of the final design at the operating license stage of review to confirm this design feature.

The results of an analysis for the startup of an inactive reactor coolant loop at an incorrect temperature show this transient to be relatively mild when compared to the other reactivity insertion events discussed above. The minimum dQorture from nucleate boiling ratio would remain well above 1.30 throughout the transient.

In summary, the most limiting transient with regard to core thermal margins, result-ing from an unplanned core reactivity anomaly, would be the uncontrolled control rod assembly bank withdrawal at power conditions.

15.2.5 Decrease in Reactor Coolant Inventory An event which can result in a decrease of reactor coolant inventory with an expected moderate frequency is an inadvertent opening of a pressurizer safety or relief valve. This transient aas evaluated for three-loop and two-loop operation with conservative input parameters used in the analysis. The mathematical model used in the transient evaluation is described in Westinghouse Topical Report WCAP-7907.

The results of the analysis show that the minimum departure from nucleate boiling ratio would remain above 1.30 throughout the transient, 15.2.6 Increase in Reactor Coolant Inventory Events that can result in an increase of reactor coolant inventory with an expected moderate frequency are inadvertent operation of the emergency core cooling system, and chemical and volume control system malfunctions. Both of these postulated transients have been reviewed, and the limiting event (spurious emergency core cooling system operation) assumed inadvertent borated water injection into the cold legs by two charging pumps while at power. The transients were evaluated by a mathematical model described in Westinghouse Topical Report WCAP-7907. The results of the inadvertent operation of emergency core cooling system during power operation show that the minimum departure from nucleate boiling ratio would not decrease from the initial value throughout the transient, thus assuring that cladding integrity would be maintained. The pressure in the reactor coolant system would dec ease due to the injected boron causing a decrease in reactor power.

s 15.2.7 Rod Cluster Control Assembly Misalignment Rod cluster control assembly misalignment accidents, including a dropped full-length

' assembly, dropped full-length assembly bank, and a misaligned full-length assembly, have been analyzed by the applicant.

The analysis was performed using the TURTLF code (described in WCAP-7213) to deter-mine X-Y peaking factors. The THINC IV code (described in WCAP-7956) was then used 0

15-11

l 2

l to calculate the departure from nucleate boiling ratio. For the transient respoise

} to a dropped assembly or assembly bank, the LOFTRAN code (described in WCAP-7907) t was used.

4-

{ Misaligned rods will be detectable by: (1) asymmetric power distributions sensed l by excore nuclear instrumentation or core exit thermocouples, (2) rod deviation t

alarm, and (3) rod position indicators. A deviation of a rod from its bank by about 15 inches, or twice the resolution of the rod position indicator, would not cause power distributions to exceed design Ifmits. Additional surveillance will be required to assure rod alignment if one or more rod position channels are out of service.

i l

t In the event of a dropped assembly, the automatic controller may return the reactor j to full power. The results of the analysis indicate that a departure from nucleate boiling ratio of less than 1.30 would not occur during this event. For the case of dropped assembly groups, the reactor would .;e tripped by the power range negative l

i neutron flux rate trip without core damage.

d For cases wnere an assembly group is inserted to its insertion limit with a single l assembly in the group stuck in the fully withdrawn position, the results of the 1 analysis indicate that a departure from nucleate boiling ratio of less than 1.30 A

would not occur.

i i

j We have reviewed the calculated estimates of the expected reactivity and power I

distribution changes that accompany postulated misalignments of representative

} assemblies.

Based on our review, we conclude that the values used in the analyses j

conservatively bound the expected values including calculational uncertainties and, i therefore, are acceptable.

1 f

15.2.8 Summary and Conclusions The transients which may be expected to occur with moderate frequency during the j' lifetime of the facility were analyzed and the results show that they meet the l

a acceptance criteria. Therefore, we conclude that the Sundesert plant design with j

regard to these transients is acceptable. The most limiting analysis with regard

! to core thermal margin is the uncontrolled control rod assembly bank withdrawal at f power. Similarly, the most limiting transient with respect to pressure within the i

i reactor coolant system and the main steam system is the loss of external load, i

i 15.3 Infrequent IncidratLand ~tulated Accidents We have rr iewed the applicant's analyses of postulated events which are expected l to rance from infrecuent incidents to limitina nostulated accidents with regard to the O cility design bases. These events have been classified by the applicant to be Cc idition III and IV events, are itemized in Table 15.1 and are discussed in the J

15-12

i sections below. The specific acceptance criteria used in evaluating the consequences l of these postulated events are as follows: !

t (1) Pressure in the reactor coolant and main steam systems shall be maintained ;

below 110 percent of the design pressures. !

L (2) The potential for core damage shall be evaluated on the following bases. If the minimum departure from nucleate boiling ratio remains above 1.30 for the l postulated event, then it can be concluded that no fuel damage will occur. If the departure from nucleate boiling ratio falls below this value, fuel damage (rod perforation) should be assumed unless it can be shown, based on an accept-able fuel damage model, that no fuel failure results. If fuel damage is calculated to occur, it should be of sufficiently limited extent so that the core will remain in place and geometrically intact with no loss of core cool-ing capability.

(3) Any activity release shall be such that the conservatively calculated doses at the site boundary are within the guideline values of 10 CFR Part 100.

15.3.1 Feedwater System Piping Breaks The analyses and effects of feedwater line break 3 inside containment, during various modes of operation, and with or without offsite power, have been reviewed. The applicant has stated that sensitivity studies were performed to determine the limiting feedwater line break. The results of these studies, which are docun.+nted in WCAP-9230, show that a double-ended rupture located between the steam generator and the main feedline check valve is the limiting break. Since the feedwater line rupture has the potential of reducing the capability of the secondary system to remove the heat generated by the core, an auxiliary feedwater system will be provided to assure that adequate feedwater will be available to remove decay heat, to prevent overpressurization of the reactor coolant system, and to prevent uncovering of the reactor core. The results of an anal" sis presented in the Preliminary Safety Analysis Report indicate that the assumed auxiliary feedwater capacity of 380 gallons per minute will be suf ficient to allow liquid in the reactor coolant system to cover the core at all times and will also be sufficient to prevent overpres-surization of the reactor coolant system.

The mathematicel models used in the accident evaluation are described in Westinghouse bM N;;rt: "P " *;7, WCAP-7908, and WCAP-7956. The results of the analysis of the postulated feedwater line break accident show that the pressures in the reactor coolant system and the main steam system would remain below 110 percent of the respective design pressures, Since the departure from nucleate boiling ratio would remain above 1.30 throughout the accident, no fuel rod failure would result.

15-13

Although the results presented are acceptable for the major feedwater line break considered, the analysis does not represent the applicant's proposed steam genera-tor design. The analysis was performed for a Westinghouse Model D steam generator with a feedwater flow restrictor whereas the proposed design for Sundesert incorpo-rates a Westinghouse Model F steam generator without a feedwater flow restrictor.

Therefore, we require that the feedwater system piping breaks be reanalyzed with the Model F steam generator design proposed for Sundesert. As a result, this matter remains outstanding.

15.3.2 Spectrum of Steam Piping Failures Outside of Containment The analyses and effects of steam line break accidents outside containment, during various modes of operation and with or without offsite power, have been reviewed.

The postulated accident which resulted in the most severe consequences was a 1.4 square foot steam line rupture analyzed at zero thermal power and with offsite power available. In the analysis it was assumed that the most reactive rod cluster control assembly was stuck in its fully withdrawn position and that a single failure in the engineered safety features occurred concurrent with the accident. The single failure was in the high head safety injection system. The mathematical models used in the evaluation are described in the Westinghouse Topical Reports WCAF7907 and WCAP-7956.

l The results of the analysis of the postulated steam line break accident show no expected fuel damage and no loss of core cooling capability. The maximum pressure within the reactor coolant and main steam systems did not exceed 110 percent of the design pressures. The applicant has stated that sensitivity studies were performed in determining the effect of initial reactor coolant flow on fuel thermal margins following a main steam line break accident. These studies, which are documented in WCAP-9226, show that the results are relatively insensitive to reactor coolant flow.

However, the long-term cooling effects on the primary system pressure and the pressure vessel temperature following a steam line break have not been evaluated in detail. The applicant has recognized this potential concern but only states that the pressure vessel integrity will be maintained for a flaw depth of less than one inch with no further propagation and that if a flaw does propagate, it will be arrested within 75 percent of the wall thickness. We require that the applicant assess the sensitivity of operator action time, in term'inating a postulated steam line break, on the long-term cooling effects on and integrity of the pressure vessel. Therefore, this matter remains outstanding.

Based on our review, and subject to the satisfactory resolution of the above matter concerning long-term cooling ef fects on pressure vessel integrity following a postulated steam line break, we conclude that the proposed Sundesert plant design

'Is acceptable with regard to the consequences of postulated steam line breaks.

15-14

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

15.3.3 Reactor Coolant Pump Rotor Seizure !

The analyses and effects of an instantaneous seizure of a rotor and of an instan-taneous break of a shaf t of a reactor coolant pump during any allowed mode of operation have been reviewed. The mathematical models used in the evaluation are I described in Westinghouse Topical Reports WCAP-7907 and WCAP-7908. It was found j that the more limiting of these events was the instantaneous seizure of a reactor coolant pump rotor. The locked rotor transient evaluations were performed for both two and three loop operation. The most limiting transient with regard to thermal margin was a locked rotor with three loops in operation. l The results of this analysis show that less than 10 percent of the fuel rods would experience departure from nucleate bulling and that the peak clad temperature would be 1939 degrees Fahrenheit. This assures that the fuel damage would be minimal and that no loss of core cooling capability would result. The radiological consequ.oces .

of this event are discussed in Sectinn 15.5 of this report. The maximum calculated reactor coolant system pressures for initial two loop and three loop operation were 2623 pounds per square inch absolute and 2607 pounds per square inch absolute, respectively. Thus, the results of the analyses show that the maximum pressure within the reactor coolant system would not exceed 110 percent of its design pres-sure (2750 pounds per square inch absolute). Also, the results show that the maximum pressure within the main steam system would not exceed 110 percent of its design pressure. i We conclude that the proposed Sundesert plant design is acceptable with regard to the consequences of a possible seizure of the rotor or break of the shaft of a reactor coolant pump.

15.3.4 Spectrum of Piping Breaks Within the Reactor Coolant Pressure Boundary The applicant has performed analyses of the performance of the emergency ;vic cooling system in accordance with the requirements of Section 50.46 of 10 CFR Part 50, by considering a spectrum of postulated break sizes and locations. Our evaluation of these analyses is contained in Section o.3.4 of this report.

15.3.5 Inadvertent Loading of a Fuel Assembly Into an Improper Position Comparisons of the results of calculations of the power distribution for the normal fuel loading pattern and five cases of fuel assembly and burnable poison misload-ings are presented by the applicant. These represent the spectrum of probable inadvertent improper loadings. With the exception of a case involving an inter-l change of region 1 and 2 assemblies near the center of the core, the resultant distortion of the power distribution would be detectable by the instrumentation provided. In the excepted case, the distortion of the power distribution is sufficiently small such that the increase in the total peaking factor would be 15-15

i l'

j approximately the uncertainty in the measurement of this value and hence causes no

! safety problems.

4 Incore instrumentation using movable fission chamber detectors will be provided i that would detect the loading mistake. A power distribution measurement using this 1

l system is required by the technical specifications to determ;ne if misloadings exist. .Thermocouples at the exits of approximately one-third of the fuel assem- j j blies could also provide an indication of a loading mistake. In most cases, an 5

improperly loaded fuel assembly would cause a quadrant power tilt that would be l 4 1 detected by the excore nuclear instrumentation. In addition to the instrumentation

{ l l system to detect misloadings, strict administrative controls will be provided to ;

prevent such an event, j l

j We conclude that an improperly loaded fuel assembly or burnable poison cluster that

would ceu w a significant safety problem could be detected with the instrumentation

! to be provided.

5 i i I j 15.3.6 Rinture of a Control Rod nrive Mechanism Housing (Rod Cluster Control Assembly Ejection)

The mechanical failure of a control rod mechanism pressure housing would result in

, the ejection of a rod cluster control assembly. For assemblies initially inserted, the consequences would be a rapid reactivity insertion together with an adverse core power distribution, possibly leading to localized fuel rod damage. Althougn j mechanical provisions have been made to make this accident extremely unlikely, the t

j applicant has analyzed the consequences of such an event.

1 i

j Methods used in the analysis are reported in WCAP-7588. This report c?monstrates i that the model used in the accident analysis is conservative relative to a three-dimensional kinetics calculation.

i

! The applicant's criteria for gross damage of fuel are a maximum clad temperature of l 2700 degrees Fahrenheit and an energy hposition of 200 calories per gram in the

} hottest pellet. The energy criterion is more conservative than that proposed in 1

Regulatory Guide 1.77, " Assumptions used for Evaluating a Coatrol Rod Ejection l

! Accident for Pressurized Water Reactors."

i j

j Four cases were analyzed: beginning-of-cycle at 102 percent and zero power, and end-of-cycle at 102 percent power and zero power. The highest clad temperature,

! 2420 degrees Fahrenheit, was reached in the zero power end-of-cycle case and the highest energy deposition, 177 calories per gram, was reached in the beginning-of-j life full power case. The analysis also shows that less than 10 percent of the fuel would go through departure from nucleate boiling and less than 10 percent of the fuel volume at the hot spot would melt. Analyses have been performed to show that I

the pressure pulse produced by the rod ejection would not stress the reactor coolant

+

15-16 i

system boundary beyond faulted limits. Further analyses have shown that a cascade effect (i.e., ejection of additional rods) is not credible.

The ejected rod worths and reactivity coefficients used in the analysis have been I

reviewed and have been judged to be conservative. Also the assumptions and methods of analysis used by the applicant are in accordance with or more conservative than those recommended in the Regulatory Guide 1.77. Therefore, we conclude that this analysis is acceptable.

15.3.7 Summary and Conclusions On the basis of our review of the results of the analyses performed for infrequent incidents and postulated accidents for the Sundesert plant, and subject to satisfactory resolution of the items listed below, we conclude that the conse- I

(

quences meet the acceptance criteria and, therefore, are acceptable.

i l

(1) We require a reanalysis of a postulated main feedwater line rupture to reflect ,

a Model F steam generator without a feedwster flow restrictor which is pro-posed for Sundesert (Section 15.3.1).

(2) We require a detailed evaluation of long-term cooling effects on pressure 1

vessel integrity following a steam line break (Section 15.3.2). 1 15.4 Anticipated Transients Without Scram i A numoer of plant transients can be affected by a failure of the scram system to function. For a pressurized water reactor, the most important transients affected include loss of normal feedwater, loss of electrical load, inadvertent control rod withdrawal, and loss of normal electrical power.

In September 1973, we issued a report WASH-1270, " Technical Report on Anticipated Transients Without Scram for Water-Cooled Power Reactors" establishing acceptance criteria for anticipated transients without scram. In response to the requirements of Append'x A to WASH-1270, Westinghouse submitted an evaluation of anticipated transir.nts without scram in Topical Report WCAP-8330, " Westinghouse Anticipated Transients Without Trip Analysis," which the applicant has incorporated by reference.

l On December 9, 1975, we issued our " Status Report on Anticipated Transients Without Scram for Westinghouse L. actors" in which we noted that certain additional analyses and justification of the Westinghouse analysis model, presented in WCAP-8330, are needed and that certain changes in typical Westinghouse plant designs are indicated.

These comments were forwarded to Westinghouse and to the applicants proposing a Westinghouse nuclear steam supply system, including the applications for the Koshkonong g ocket Nos. STN 50-502 and STN 50-503) and Shearon Harris (Docket Nos.

50-400, 50-401, 50-402 and 50-403) nuclear plants which are similar in design to the Sundesert plant.

15-17

]

l

! We have completed a reevaluation of the probability of and the potential consequen-j ces associated with anticipated transients without scram and have recently published i a more definitive report, NUREG-0460, " Anticipated Transients Without Scram for I Light Water Reactors," Volumes 1 and 2, dated June 16, 1978. This report was

! issued by the Division of Systems Safety in the Office of Nuclear Reactor Regulation.

l The report states that, considering the expected frequency of occurrence of j transients, the reliability of current reactor scram systems necessary to meet

! safety objectives has not been demonstrated. Therefore, the report recommends that t

i means of reducing the consequences of anticipated transients without scram events

! should be provided.

1 1-j In our view, any changes necessary to meet the limits specified in WASH-1270 and l NUREG-0460 can be incorporated in the design of the Sundesert plant prior to com-i pletion of construction. We will review this matter further at the operating l license stage of review and during any continuing generic review of the matter.

I i 15.5 Radiological Consequences of Accidents We evaluated the effectiveness of the engineered safety feature equipment for l reducing the offsite radiological consequences of postulated accidents. The j evaluations were based on the acceptance criteria described in Sections 15.4.8, i 15.6.5, 15.7.2, 15.7.4, and 15.7.5 of the Standard Review Plan and are presented in f the following sections. The postulated accidents evaluated included control rod j ejection, loss-of-coolant, fuel handling accident, and spent fuel cask drop. Table i 15.3 summarizes the calculated doses for these accidents. We have also evaluated i

g postulated radioactive releases due to liqu i d tank failures and the results of this j evaluation are presented in Section 15.5.5 of this report.

1 i

On the basis of our experience with the evaluation of steam line break, steam generator tube rupture, and reactor coolant pump locked rotor accidents for pres-j surized water reactor plants of similar design, we have concluded that the radio-l logical consequences of these accidents can be controlled by limiting the permis-

! sible radioactivity concentrations in the primary and secondary coolants. These f limits can be established such that the calculated consequences are within the i exposure guidelines of 10 CFR Part 100. We will include appropriate limits on the f

e primary and secondary coolant activity concentrations in the technical specifica-I tions to be issued with the operating license.

1 i

4 j The radioactive waste gas adsorber tanks will be designed to seismic Category I

! requirements. Therefore, the total failure of these tanks is sufficiently improt-i able and that 10 CFR Part 100 guideline doses are applicabir Our calculations l indicate that the offsite doses resulting from the failure of these tankt would be well within 10 CFR Part 100 guidelines. Appropriate technical specifications will

) be placed on the maximum activity that can be stored in any one tank at any time 1

1 15-18

TABLE 15.3 CALCULATED RADIOLOGICAL CONSEQUENCES OF DESIGN BASIS ACCIDENTS Low Population Zone Exclusion Area Two Hour Course of Accident Dose Accident Dose (rem) (rem)

Thyroid Whole Body Thyroid Whole Body loss-of-Coolant

170 3 163 1 Fuel Handling:

Inside fuel Building ** 63 <1 9 <1 Inside Containment (with filters) 16 <1 2 <1 Cask Drop <1 <l <1 <1 Rod Ejection (estimattd):

Release Through Containment Only 13 <1 2 <1 Release Through Secondary System Only 28 <1 4 <1 Includes the added dose due to primary coolant leakage from the emergency core cooling system into the annulus building.

- No credit is given for the filters in the fuel building since they are not 'esi( ated as seismic Category I.

15-19

such that a single failure of any active component, including the lifting or stick-ing of a safety or relief valve, will not result in radiological consequences that would exceed small fractions of 10 CFR Part 100 guideline doses.

15.5.1 toss-of-Coolant Accident The containment model used to describe the dose mitigating effects of the engi-neered safety features for the Sundesert plant includes a low leakage containment structure surrounding the reactor and a sodium hydroxide additive injection system acting in conjunction with the containment spray system. The purpose of the sodium hydroxide injection system is to increase the iodine removal capability of the spray following a postulated loss-of-coolant accident. The assumptions we used in eva cating the radiological consequences of this accident are listed in Table 15.4, and are based on the recommendations of Regulatory Guide 1.4, " Assumptions Used for Evaluating the Potential Radiological Consequences of a toss-of-Coolant Accident for Pressurized Water Reactors." The results of the calculation indicate that the ,

potential radiological consequences are within the guideline values of 10 CFR Part 100.

As part of the loss of-coolant accident analysis, we have also evaluated the radio-logical consequences of leakage of containment sump water, which is circulated by the emergency core cooling system outside the containment, into the annulus build-ing after the postulated accident. Af ter a loss-of-coolant accident, this water would be circulated into the annulus building to be cooled. If leakage should develop, part of the iodine in the sump water would become airborne and could exit to the outside atmosphere. As discussed in Section 6.5.2 of this report, a redun-dant, safety grade (iltration system will be provided in the annulus building to process the potential airborne radioactivity from the leakage of the sump water.

The applicant has estimated a leakage rate of about 4328 cubic centimeters per hour from components of the emergency core cooling system. We assumed that the sump

~

water contains a mixture of iodine species as fission products in conformance with the values in Regulatory Guide 1.7, " Control of Comoustible Gas Concentrations in Containment f ollowing a loss-of-Coolant Accident." Based on the estimated leakage of the containment sump water and on the assumed concentration of iodine fission products in the water, we calculate that the contribution tc the offsite doses are small, and when added to the calculated loss-of-coolant accident doses, result in total doses, as shown in Table 15.3, that are still within the exposure guideline values of 10 CFR Part 100. In the calculations, we have allowed a margin for uncertainties to assure that the 10 CFR Part 100 guideline values will be met at the operating license stage of review. '

15.5.2 Fuel Handling Accident Two postulated fuel handling accidents have been evaluated; one occurring inside the fuel building and the other occurring inside the containment. The assumptions 15-20

TABLE 15.4 i LOSS-OF-COOLANT ACCIDENT ASSUMPTIONS AND INPUT PARMETERS I

Power Level (megawatts thermal) 2958 Operating Time (years) 3 Fraction of Core Inventory Available for Leakage (percent):

Iodine 25 Noble Gases 100 Initial lodine Composition in Containment (percent):

91 Elemental 4

Organic 5

Particulate Containment Volume (cubic feet):

Sprayed Region 1.6 x 106 Unsprayed Region 6.8 x 105 Containment Spray System Ef fectiveness:

Decontamination Factor, Elemental Iodine 100 Removal Coefficient for lodine (per hour) 10 Elemental Particulate 0.5 0

Organic Containment Leak Rate, Direct, Unfiltered (weight percent per day):

0-24 hours 0.2 1-30 days 0.1

- Atmospheric Dispersion factors (seconds per cubic meter):

~

Exclusion Boundary, 0-2 hours 2.2 x 10 4 Low Population Zone 0-8 hours 3.3 x 10~5

~

- 8-24 hours 2.4 x 10 5 1-4 days 1.2 x 10~5

~

4 30 days 4.8 x 10 6 15-21

! we used in evaluating the radiological consequences of these accidents are listed l in Table 15.5 and are based on the recommendations of Regulatory Guide 1.25, l " Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel i Handling Accident in the Fuel Handling and Storage Facility for Boiling and i Pressurized Water Reactors." The applicant has indicated that fuel handling in the l

fuel building will be permitted only when the filtration system is in operation.

However, as stated in Sections 3.2.1, 6.5.3 and 9.4.2 of this report, the

{

j filtration system and the portion of the fuel building which will house the

{ filtration system are not classified as seismic Category I. We have evaluated the radiological consequences of the postulated fuel handling accident in the fuel f

. building without giving credit for iodine removal by the filtration system sincJ it i is not classified as seismic Category I. Although the calculated offsite doses for

, the postulated accident, as shown in Table 15.3, are within the guideline values of l

10 CFR Part 100, we require that additional design provisions be incorporated to assure that, at the operating license stage of review, the calculated consequences of this postulated fuel handling accident are well within the guidelines.

Accordingly, we require that the proposed filtration system and the fuel building be designed as seismic Category _I (see also Sections 3.2.1, 6.5.3 and 9.4.2 of this report).

Subject to the satisfactory resolution of the above matter, we conclude that the provisions to be incorporated for mitigating the consequences of a postulated fuel handling accident in the fuel building are in conformance with our requirements as stated in Sections 3.2.1, 6.5.3 and 9.4.2 of this report and are acceptable.

The applicant indicates that fuel handling insioe the containment will be allowed only if the containment is isolated or when filters are in operation. Although the specific design has not been chosen, one of the following options will be imple-mented to mitigate the consequences of a fuel handling accident inside the contain-ment so that the calculated doses will be within Commission guideline values.

(1) Provide either fast shutting isolation valves or early detection capability in conjunction with an analysis justifying that the travel times will allow the dampers to shut prior to a significant release, or l

(2) Provide a filtration system in accordance with the recommendations of Regulatory Guide 1.52, " Design, Testing, and Maintenance Criteria for Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration ano Adsorption Units of Light-Water-Cooled Nuclear Power Plants."

We evaluated the radiological consequences of a fuel handling accident within the containment assuming that the proposed filtration system (second) option will be in operation daring refueling. The calculated doses, which are listed in Table 15.3, are well within the exposure guideline values of 10 CFR Part 100. We conclude that 15-22

I TABLE 15.5 FUEL HANDLING ACCIDENT ASSUMPTIONS AND INPUT PARAMETERS Power Level (megawatts thermal) 2968 Number of fuel Rods Damaged 264 Total Number of Fuel Rods in Core 41,448 Power Peaking Factor of Damaged Fuel 1.65 Shutdown Time (hours) 100 Inventory Released From Damaged Rods (fodine and noble gases) 10 (percent)

Pool Decontamination Factor:

Iodine 100 Noble Cases 1 Jodine Fraction Above Pool (percent):

Elemental 75 Organic 25 Filter Efficiencies for Iodine Removal (percent):

Inside Frel Building - elemental 0 organic 0 Inside Containment - elemental 90 organic 30 3 Atmospheric Dispersion Factor (seconds per cubic meter):

~

Exclusion Boundary, 0-2 hours 2.2 x 10 4

~

Low Population Zone 0-8 hours 3.3 x 10 5 15-23

i i

J the proposed tiltration system option for mitigating the consequences of a refuel-I trg accident inside the containment would meet our acceptance criteria and, there-fore, would be acceptable, l

We have also reviewect the proposeu containment isolation system (first) option for use in refueling operations inside the containment. Based on our review, we con-

! clude that this option would also be acceptable since timely isolation of the

containment can be effected for a postulated refueling accident with a proper design to limit the radioactivity releases to acceptable levels. If this option is l selected by the applicant, we will review the isolation system layout in detail at I

a the operating license stage of review to assure that the provisions fo mitigating i the consequences of refueling accidents will meet our acceptance criteria.

1 i

15.5.3 Cask Drop Accident 4

i I

Table 15.6 lists the assumptions and input parameters we used in the evaluation of a

the radiological consequences of a spent fuel cask drop accident. The fission i product inventory in the fuel element gap was determined in a manner similar to that for the fuel handling accident. $ince the accident can occur in an area where i safety grade filters are not available, an untreated puff release was assumed.

t 4

As shown in Table 15.3, the calculated doses are well within the exposure guideline t

values of 10 CFR Part 100. In this calculation, we have allowed an adequate margin l

{ for uncertainties to assure that the doses will be well within the guidelines at I the operating liceae stage of review.

1 4

15.5.4 Rod Ejection Accident i

j We previously performed an evaluation of the radiological consequences of a control l rod ejection accident for a plant similar to Sundesert, i.e., RESAR-3S (Docket No.

STN 50-545). Our evaluation for RESAR-35 is presented in NUREG-0104, ASafety

) Evaluation Report Related to the Preliminary Design of the Standard Reference j System RESAR-35," dated December 1976. Using the doses calculated for RESAR-35, we estimated the resultant doses at the Sundesert site for this postulated accident by

taking into account the plant and site specific parameters for Sundesert. Two cases were considered in this evaluation: all radioactivity releases are through the containment only, and all release:; are through the secondary coolant system only.

Table 15.7 lists the assumptions and input parameters we used in the dose l calculations, i

The estimated thyroid and whole body doses, which are given in Table 15.3, are well within the exposure guideline values of 10 CFR Part 100. If a reevaluation of this accident at the operating license stage o review results in doses that would exceed tre guideline values, appropriate limits on the primary-to-secondary leakage and the setpoint for containment spray actuation can and will be set.

15-24

l TABLE 15.6 CASK OROP ACCIDENT ASSUMPTIONS AND INPUT PARAMETERS Power Level (megawatts thermal) 2958 Number of Fuel Assemblies Damaged 10 Power Peaking Factor of Damaged Fuel 1.65 Cooling Per od (days) i 150 Atmospheric Dispersion Factors (seconds per cubic meter):

Exclusion Boundary, 0-2 hours 2.2 x 10 4

~

Low Fopulation Zone, 0-8 hours 3.3 x 10 5 15-25

1 TABLE 15.7 ROD EJECTION ACCIDENT ASSUMPTIONS AND INPUT PARAMETERS 1

Power Level (megawatts thermal) 2958 Percent failed fuel 10 Atmospheric Dispersion Factors (seconds per cubic meter):

Exclusion Boundary, 0-2 hours 2.2 x 10 4

~

Low F?oulation Zone, 0-8 hours 3.3 x 10 5

~

CASE 1, ALL_ RELEASES THROUGH CONTAINMENT Inventory Released From Failed Fuel (iocine and noble gases) 10 (percent)

Plateout of Iodine (percent) 50 ..

Containment Leak Rate, Direct, Unfiltered (weight percent per day):

0-24 hours 0.2

> 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 0.1 CASE 2, ALL RELEASES THROUGH SECONDARY SYSTEM Fuel Melt After Accident (percent) 0.25 Inventory Released From Failed Fuel (percent):

Noble Cases 100 lodine 25 , ,

5 Time to Reach Pressure Equalization Between Primary and Secondary 40 Systems (minutes) 15-26

l 2

i

/

15.5.5 Pos,ulated Radioactive Releases Due to Liquid h Failures l l

The consequences of failure of components located outside the reactor containment, l whick could result in releases of liquids containing rad wactive materials to the l etai runs, were evaluated. Considered in our evaluation were (1) the radionuclide inventory in e n n component assuming a one percent operating power fission product l '

source term (the amount of fission product inventory in the core at full power that 15 released to the primary coolant). (2) a component liquid inventory equal to 80 l l

percent of its design capacity, (3) the mitigating effects of plant design, includ-ing overflow lines and the location of storage tanks in curbed areas designed to retain spillage, and (4) the effects of site geology and hydrology.

The applicant has incorporated provisions in the design to retain releases from ,

liquid overflows as discussed in Section 11.2.1 of this report. In the event of a j i

spill, we evaluated the consequences by postulating that the liquid flows directly to ground water beneath the site and is transported to the nearest potable water well which is located 2.9 miles southeast of the proposed plant location.

Based on our evaluation, the potential tank failure resulting in the greatest quantity of activity release to the environment is the failure of the 150,000 gallon boron recovery tank of the boron recovery system. In our evaluation, we l i

have determined that the liquid transit time for the leakage to the nearest user I would be 90 years and that the liquid would have an overall dilution factor of 480 in transit (see Section 2.4.4 of this report). Considering the leakage dilution and transit time, the calculated radionuclide concentrations 11 the well result in values that are small fractions of the guideline values of of 10 CFR Part 20, Appendix B, Table II, Column 2, for unrestricted areas. Based on the foregoing evaluation, we conclude that the provisions incorporated in the applicant's design to mitigate the effects of component failures involving contaminated liquids are acceptable.

l l

15-27 l

l

)

16.0 TECHNICAL SPECIFICATIONS The technical specifications in an operating license define certain 'eatures, characteristics, and conditions governfrg operation of a facility tha' cannot be F'nal techni-changed without prior approval of the Nuclear Regulatory Commission.

cal specification will be developed and evaluated at the operating license stage.

However, in accordance with Section 50.34 of 10 CFR Part 50, an applicant for a construction perV is required to include preliminary technical specifications in the Preliminary Safety Analysis Report. The regulations require an identification and justification for the selection of those variables, conditions or other items which are determined as a result of the preliminary safety analysis and evaluation to be probable subjects of technical specifications for the facility, with special attention given for those items which may significantly influen:e the design.

We have rcnewed the proposed technical specifications presented in Section 16.0 of the Preliminary Safety Analysis Report with the objective of identifyfog those items that would require special attention at the construction permit stage, to l

assure that any significant change in design to wy; ort the final technical specifications will not be necessary. The proposet, technical specifications are

' similar to those being developed or in use for plants of design similar to the proposed facility. We have not identifier, any items which require special i attention at this stage of our review.

On this basis, we have concluded that the proposed technical specifications are acceptable.

I l

l 16-1 1

17.0 QUALITY ASSURANCF 17.1 General The description of the quality assurance program for the Sundesert Nuclear Slant, Unit Nos. I and 2 is contained in Section 17.0 of the Preliminary Safety Analysis Report.

Our evaluation of the quality assurance progran ciscription is based on our review of the Prelkinary Safety Analysis Report to deter.'ine if the San Diego Gas and Electric Company and its principal contractors, Stone & Webster Engineering Corpo-ration, and Westinghouse Electric Corporation, comply with the requirements of Appendix B to 10 CFR Part 50, and the applicable guidance listed in Table 17-1.

The appli a nt has delegated to the above principal contractors the work of estab-lishing and impitmenting portions of the quality assurance program commensurate with their scope f.f work for design, procurement, fabrication, and construction of the safety-related items covc ed by the quality assurance program.

Westinghouse will supply the nuclear steam supply systems while Stone & Webster g will serve at architect-engineer and constructor. The applicant retains ultimate ,

responsibility for the implementation of all activities and functions required for the design, construction, and operation of the Sundesert plant.

17.2 San Diego Gas and Electric Company

'he San llego Gas and Electric Company will be responsible for the desip, construc-v ar. , and operation of Sundesert Nuclear Plant, Unit Nos. I and 2. The President of the company is responsible to the Board of Directors for the overall management of company operations and for the establishment of policies including quality assurance for nuclear power plants. He has delegated responsibility and authority for all aspects of nuclear power plant design, procurement, construction, cperations, and quality assurance to the Senior Vice-President - Operations, as shown in Figure 17-1. The senior Vice President - Operations has delegated to the Vice-President - Project Management, responsibilities for licensing, design, procare-ment, construction, and preoperational testing of Sundesert plant; and has delegated to the Manager - Quality Assurar.ce the responsibility for establishing the quality assurance program for the plant and the authority to assure its implementation.

17-1 s _ - - _ _ _ _ _ _

TABLE 17.1 REGULATORY GUIDANCE APPLICABLE TO

~ ~VOXH TY ASSURANCE PROGRAMS 1.

Regulatory Guide 1.28, " Quality Assurance Program Requirements (design and Construction)."

2.

Regulatory Guide 1.30, " Quality Assurance Requirements for the Installation, Inspection and Testing of Instrumentation and Electric Equipment."

3.

Regulatory Guide 1.37, " Quality Assurance Requirements for Cleaning of Fluid Systems and Associated Components of Water-Cooled Nuclear Power Plants."

4.

Regulatory Guide 1.38, " Quality Assurance Requirr,ments for Packaging, Shipping, Receiv-ing, Storage and Handling of Items for Water-Cooled Nuclear Power Plants."

5. Regulatory Guide 1.39, " Housekeeping Requirements for Water-Cooled Nuclear Power Plants."

6. Regulatory Guide 1.58, " Qualification of Nuclear Power Plant Inspection, Examination and

, Testing Personnel."

l 7. Regulatory Guide 1.64, " Quality Assurance Requirements for the Design of Nuclear Power

! Plants."

8. Regulatory Guide 1.74, " Quality Assurance Terms and Definitions."

9. Regulatory Guide 1.88, " Collection, Storage, and Maintenance of Nuclear Power Plant Quality As,arance Records."

! 10.

Regulatory Guide 1.94, " Quality Assurance Requirements for Installation, Inspection and Testing of Structural Concrete and Structaral Steel During the Construction Phase of Nuclear Po er Plants."

11. Reguli 'ory Guide 1.116, " Quality Assurance Requirements for Ins tallation Inspection and Testit of Mechanical Equipment and Systems."

12. Regulatory Guide 1.123, "Qu(!;ty Assurance Requirements for Control of Procurement of Items and Services for Nuclear Power Plants."

13. ANSI Standard N45.2.8, " Quality Assurance 0; ireme r's for Installation Inspection and Testing of Mechanical Equipment and Systems.

14.

ANSI Standard N45.2.12, " Requirements for Auditing nf Quality Assurance Programs for Nuclear Power Plants."

15. ANSI Standard N45.2.13, " Quality Assurance Requirements for Procurement of Items and $

Services for Nu: lear Power Plants."

6 17-2

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17-3
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The Senior Vice-President - Operations, is responsible for assuring that the Manager -
l Quality Assurance has sufficient independence an f authority to fulfill his responsibil-ities. The Senior Vice-President - Operations implements his responsibilities by I
approving the quality assurance plan, which sets forth the requirements of the
[ quality assurance program, and maintains continuing involvement in quality assurance activities by frequent meetings with the Manager - Quality Assurance, by reviewing quality assurance audit reports and reports of corrective action, and by having f management audits of the quality assurant.e program implementation conducted at f lea,t annually.
i Each nuclear power plant is assigned a Projectr4anager (Figures 17-1 and 17-2) by the Vice-President - Project Manacement. The Project Manager is responsible for j directing the day-to-day nuclear project activities. He also has the authority to stop unsatisfactory work of the prime contract (rs, i
I The Manager - Quality Assurance reports directly to the Senior Vice-President -
l Operations and is responsible for assuring eff.ective implementation of the quality
) assurance program. This organf2ational relati]nship provides the Manager - Quality l Assurance with sufficient authority, organizational freedom, and independence from j undue influence or responsibilities of project costs and schedules so that he can l
effectively direct and control the quality as!urance program. He is responsible for verifying that activities affecting quality performed Dy the applicant and
} prime contractors have been performed in accordance with quality assurance program l requirements. He maintains liaison with quality assurance organi2ations of prime contractors which provides him with a curt <n~. status of quality-related and other activities affecting the applicant's pir., orojects. He communicates directly with
~~the Project Manager and with appropriau mar 'gement levels in prime cantractor I~~
organizations to identify quality-relatec problems; initiate, recommend, or provide l solutions; and to verify implementation of solutions to the quality-related problems.
He and his staff develop quality assurance procedures and documentation, provide j
trairing and indoctrination, and perform QLality audit functions internally within one ampany as well as externally. The Quhlity Assurance Department consists of a j
< home office Qualit: ieering staff and a site Quality Compliance Section as j shown on Figure 17-L e Manager - Quality Assurance and his staff also have the ,
j authority to stop un 4tisfactory site work during construction and preoperational
, testing.
i 1
j The San Diego Gas and Electric Company has delegated to the architect-engineer and constructor, Stone & Webster Engineering Corporation, and to the cupplier of the nuclear steam supply systems, Westinghouse Electric Corporation, the administration and implementation of large portions of the quality assurance program associated 1
with the design, procurement and installation of safety-related items associated
with the plant. Oroanizational and programmatic arrangements with Stone & Webster and Westinghouse are subsequently summa *ized in Sections 17.3 and 17.4 of this 1 report. i I
17-4
SEWlOR VICE PRESIDENT OPE R ATIONS I
VICE PRESIDENT PROJECT MANAGEMENT MANAGER gg,C; ________________
OUAug p - _____ _ _
l CONSULTANTS .
L_______J OUALITY r------------]
I OUALITY ENGINEERING l COMPLIANCE I
' ' I MECH ANIC AL lg MECHANICAL l

PROJECT SYSTEMS ELECTRICAL ,
ORGANIZATION ! CIVIL /STRUCTU R AL CIVIL /STRUCTUR AL 1 AUDITS l INSTRUMENTATION & CONTROL l RECORDS MAN AGEMENT l WELDING /NDE g
DESCRIBED IN FIGURE 13.1-2 OF THE QA SYSTEMS RECORDS ELECTRICAL !
PRELIMINARY SAFETY ANALYSIS REPORT l l l 1 SITEg
ORG ANIZATIONS SUBJECT To L_____ -
INTERNAL AUDIT BY OUALITY ENGINEERING FIGURE 17.2 SAN DIEGO GAS AND ELECTr'IC COMPANY NUCLEAR POWER PLANT PROJECT ORG ANIZA'IION
we conclude that the structure and responsiblity of the applicaht's organizational arrangement for the Sundesert plant provide the necessary independence and authority to effectively implement the quality assurance program and to effectively control the quality assurance functions of the principal contr,ctors.
The policies utilized to meet the requirements of Appendix 0 to 10 CFR Part 50 are contro' led by procedures contained in the applicant's manuals described in Topical Report 50QAPD-1 (Revision lA), " Quality Assurance Program Description," and by procedures and manuals referenced in Topical Report WCAP-8370 (Revision 8A),
" Quality Assurance Plan - Nuclear Energy Systems Divisions," and Topical Report SWSQAP-1-74A (Revision B), " Standard Nuclear Quality Assurance Program," of Westinghouse and Stone & Webster respectively. The Westinghouse and Stone & Webster quality assurance programs are considered extensions of the applicant's quality assurance program and, as such, are subject to review, audit and approval by the applicant. Subcontractors providing safety-related items and services are required to implement quality assurance programs whi-b are consistent with the applicant's program. The structures, systems and co- i.its that are subject to the quality assurance programs have been identified in Table 3.2.5-1 of the Preliminary Safety Analysis Report.
The applicant has committed to comply with the guidance of the applicable regu-latory guides and ANSI standards in Table 17-1 with some exceptions, which the Commission's staff hn found acceptable, as described in Topical Report SDQAPD-1.
Our acceptance of this topical report is documented in a letter to the applicant, dated December 6, 1976.
The applicant's quality assurance program :s revised at least annually to assure that it is kept current. The applicant rerforms audits on Westinghouse and Stone &
Webster to assure that their programs are kept up-to-date and ef fective.
s In our review, we have evaluated the applicar.t's quality assurance program for compliance with the Commission's regulations and applicable regulatory guides and industry standards. Based on our review, we conclune that the quality assurance program contains the necessary provisions, requirements and controls for compliance with Appendix B to 10 CFR Part 50 and applicable regulatory guides and industry sta7dards and, therefore, is acceptable for the design, procurement and construction of the Sundesert plant.
17.3 Westinghouse Electric Corporation Westinghouse Nuclear Energy Systems Divisions maintain responsibility for providing '
the nuclear steam supply systems for the Sundesert plant. Figure 17-3 shows the organization of the divisions of Westinghrase which provide nuclear plant services and equipment. The Westinghouse divisions operate under an Executive Vice-President who reports to the President, Westinghouse Power Systems. The Pressurized Water Reactor Systems Division is the lead division for design and procurement.
17-6
W'ESTINGE ~1USE POWER SYS1"'iS
. COMPAN\'
l PRESIDENT THE NES QUALITY ASSURANCE COMMITTEE IS COMPOSED OF THE QUAitTY ASSURANCE MANAGERS FROM EACil OF THE NES DIVISIONS AND THE EXECUTIVE ASSISTANT TO THE NES EXECUTIVE NUCLEAR ENERGY VICE PRESIDENT.THE COMMITTEE'S CHAIRMAN SYSTEMS IS THE PWR PRODUCT ASSURANCE MANAGER.
EXECUTIVE VICE PRESIDENT EXECUTIVE (NES) NES QUALITY ASSISTANT TO THE ASSURANCE NES EXECUTIVE COMMITTEE VICE PRESIDENT WATER REACTOR DIVISIONS d (WRD) a I
WESTINGHOUSE WRD PWR NUCLEAR NUCLEAR EAR NEL NUCLEAR MARKETING SYSTEMS EQUIPMENT SERVICE I ION EUROPE DIVISION DIVISION DIVISIONS DIVISION (WN E) (WRDM) (PWRSD) (NED) (NSD) (N FD)
ELECTRO- SPECI ALTY TAMPA PENSACOLA MECH ANICAL METALS DIVISION DIVISION DIVISION DIVISION (EMD) (PD) (SMD) (TD)
FIGURE 17.3 WESTINGHOUSE NUCLEAR ENERGY SYSTEMS (NES) ORGANIZATION (WATER REACTORS)
Each Westinghouse division contains an organization specifically responsible for quality assurance and for quality control which reports at a management level to assure independence consistent with Criterion 1 of Appendix B to 10 CFR Part 50.
Quality assurance management in each division is free of prime responsibility for schedule or cost. It has the authority to: stop work pending resolution of matters relating to quality, and has the freedor to identify quality-related problems; initiate, recommend or provide solutions through designated channels; verify imple-mentation of solutions; and control further processing, delivery or installation of nonconforming items. We find that Westinghouse has adequately defined the respon-sibilities of the organizations performing quality-related activites and that they are acceptable.
The Exec.:tive Vice-President has (stablished a Quality Assurance Committee which includes the Quality Assurance and Reliability Managers of each Westinghouse division.
The Product Assurance Manager of the Presrurized Water Reactor Systems Division is l Chairman of the Quality Assurance Committee. The Quality Assurance Committee
! audits each division annually to assess the scope, implementation and effectiveness l Of the division's quality assurance program.
l The quality assurance program applies to all safety-related structures, systems and components within the Westin0 house scope of work, Westinghouse has also committed to comply with the guidance of the applicable regulatory guides and ANSI standards in Table 17-1 with some exceptions, which the Commission's staff has found acceptable, as described in Topical Report WCAP-8370. Our acceptance of this topical report is ,
documented in a letter to Westinghouse, dated December 31, 1974.
Sinca each Westinghouse division has a different scope of work, each division manager amplifies the common Westinghouse quality assurance policy, as necessary, for local application. Each division documents and implements a quality assurance program which assures that safety-related items meet the applicable requirements of Appendix B to 10 CFR Part 50. Each division manager authorizes, reviews and approves the quality assurance program for his division.
A matrix which relates the procedures of the various manuals to the applicable quality assurance criteria of Appendix B to 10 CcR Part 50 is given in Topical Report WCAP-8370. Based on our review of this matrix, we conclude that each criterion has been specifically included in written procedures within the Westinghouse quality assurance program.
The quality assurance program includes provisions for the control of design informa-tion. Design inputs are reviewed, and analyses are accomplished in accordance with applicable codes, standards and regulatory requirements. Knowledgeable groups within Westinghouse, including quality assurance and reliability personnel, independently review drawings and equipment specifications prior to issuance.
17-8
To provide control of PLrchaseo items and services, Westinghouse evaluates the quality control systems of each prospective supplier of safety-related items.
Quality assurance engineers review purchase requisitions, purchase orders and subsequent change notices. Westinghouse reviews and retains supplier documentation which demonstrates acceptable quality. Audits and feedback af discrepancy data are used by WestingSouse quality assurance engineers to measure ,upplier performance.
Westinghouse b ecutes a comprehensive audit program which provides its management j
with information on the effectiveness of the quality assurance program. Westinghouse audits activities affecting quality at Westinghouse and at supplier facilities.
Audit areas include all quality related procedures and operations. Trained personnel, not having direct responsibilities in the area being audited, conduct the quality assurance audits in accordance with defined procedures and checklists.
In our review, we have evaluated the Westinghouse quality assurance program for compliance with the Ccmmission's regulations and applicable regulatory guides and industry standards. Based on this review, we conclude that the Westinghouse quality assurance program contains the necessary provisions, requirements and controls for compliance with Appendix B to 10 CFR Part 50 and applicable regulatory guides and industry standards and, therefore, is acceptable for the nuclear steam supply systems for the Sundesert plant.
17.4 Stone & Webster Engineering Corporat_fon 1
The Stone & Webster Engineering Corporation is responsible for architect-engineering and construction management for the Sundesert plant. The Stone & Webster organiza-tion responsible for design, procurement, and construction activities is shown in figure 17-4. The President has established quality assurance policies for the corporation, and has delegated the authority for the development of the quality assurance program to the Vice-President of Quality Assurance who is responsible for the overall quality assurance program. As Figure 17-4 shows the Vice-President of Quality Assurance is independent of, anj has organizational authority equal to, the l other Stone & Webster Vice-Presidents.
i l
The Manager of Quality Assurance, who reports to the Vice-President of Quality Assurance, is responsible for administering and managing the quality assurance program for Stone & Webster procurement and construction activities. The Chief Engineer, Engineering Assurance Divisinn of the Engineering Department, is respon-sible for administerine, and managing the quality assurance program for engineering and design. As Figure 17-4 shows, the Chief Engineer has the organizational independence and authority to perform his functions in accordance with his assigned responsibilities.
Major quality assurance activities which are carried out by the Quality Assurance and Engineering Assurance organizations of Stone & Webster are:
17-9
. m m.- m .._.m.. ...__.........._.~_..m__m._ . _ .. ..- . _ ._... - . ....._.._ ,_. .m _ _ _ , . . . _ . . _ . . . _ , - -m- .m m .m I
PRESIDENT I .
-l 1 i q VICE PRESIDENT VICE PRESIDENT SENIOR VICE PRESIDENT QUALITY ASSURANCE VICE PRESIDENT VICE PRESIDENT AND MAN AGER AND ENGINEERING AND DIRECTOR OF PROJECTS MANAGER I I OF CONSTRUCTION MANAGER QUALITY MANAGER ASSURANCE OF PURCHASING l I I i i >
CONSTRUCTION ENGINEERING CHIEF ENGINEER D ISiON E IN R' FIELD PROCUREMENT MANAGER Q C DIVISION p NCE O C DIVISION I I l PROJECT STAFF FIELD PROCUREMENT MANAGER SPECI Alls TS Q C SITE OC O E I I .
l l QUALITY SYSTEMS DEPARTMENT O A NDT DIVISION SERV'CES GROUP DIVISION RESIDENT PROJECT MANAGER ENGINEE R I
I I I SUPT OF PROJECT O A CONSTRUCTION DESIGN ENGINEERING COORDINATOR l l l CONSTRUCTION RESIDENT J-lE LD SUPERVISORS ENGINEER PURCHASING AGENT "
l FIGURE 17.4 STONE & WEBSTER ENGINEERING CORPORATION ORGANIZATION FOR QUALITY ASSURANCE l
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(1) The review and approval of Stone & Webster design, procurement, manufacturing, inspection, construction, and test documents; and (2) Inspections and audits within Stone & Webster, at supplier's facilities, and at the construction site.
Quality Assurance and Engineering Assurance personnel have the authority and freedom to: identify quality-related problems; initiate, recommend, or provide solutions; and control further processing, delivery, or installation of a nonconforming item until proper disposition of the deficiency or unsatisfactory condition has been approved. We conclude that the Stone & Webster organization is structured such that an individual re,sponsible for coordinating the directica and control of the quality assurance and quality control function and that personnel performing quality assurance functions in tne organization have sufficient authority and organizational freedom to perform their critical functions effectively and without reservations.
The quality assurance program applies to all safety-related structures, systems, and components within the Stone & Webster scope of work. Stone & Webster has also l committed to comply with the guidelines of the applicable regulatory guides and ANSI standards in Table 17-1 with some exceptions, which the Commission's staff has found acceptable, as described in Topical Report SW5QAp-1-74A. Our acceptance of this topical report is documented in a letter to Stone & Webster, dated December 31, 1974.
The quality assurante program for engineering includes quality assurance review of applicable engineering instructions, procedures, specifications, and drawings to !
assure that the quality-related requirements are clearly, accurately, and adequately stated. The program requires that design work be verified or reviewed by individuals within the engineering organization not responsible for originating the design and that a determination be made that the engineering specifications, procedures, instructions, and drawings comply with reculatory requirements and design bases.
For procurement control, quality assurance measures provide for the review of procurement documents to assure that the stated quality-related requirements are adequate for supplier qualification and for approval of the supplier's quality assurance program. The program provides for inspection, surveillance, and audit of the suppliers' safety-re'ated structures, systems, and components to assure compli-ance to procurement requirements.
During construction, the program provides for onsite quality assurance involvement including inspection, nondestructive testing, retention of records, and processing of deficiencies, nonconformances, and design changes. The cuality control engineers, inspectors, and nondestructive testing personnel are organizationally separate and independent from the construction organization.
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b The quality assurance program provides for a comprehensive system of detailed audits to be perfcrmed by the Stone & Webster quality assurance organization. The audits encompass the review and evaluation of all quality-related activities associ-ated with the quality assurance program and involve procedures, work areas, hardware, activities, and records. The program requires that the audits be conducted in accordance with pre-established procedures by qualified personnel not having direct responsibilities in the area being audited. The results are documented and distributed to the appropriate levels of management.
In our review, we have evaluated the Stone & Webster quality assurance program for compliance with the Commission's regulations and applicable regulatory guides and l industry standards. Based on this review, we conclude that the Stone & Webster quality assurance program contains the necessary quality assurance provisions, requirements, and controls for compliance with Appendix B to 10 CFR Part 50 and applicable regulatory guides and industry standards and, therefore, is accpetable for the design, procurement, and construction of the Sundesert plant.
17.5 Implementation of the Quality Assurance Program The Commission's Office of Inspection and Enforcement has examined the applicant's quality assurance program manuals and procedures for Sundesert plant to determine the applicant's conformance to the quality assurance program described in Topical Report SDQAPD-1. On the basis of this inspection, the Office of Inspectio.s and Enforcement concludes that the quality assurance program implementation is consistent with the. status of the project, with the exception of t, hat portion related to construction activities as discussed below.
Because of uncertainties re w ording approval of the project by the State of Cali-fornia, the applicant had reduced construction related pro,,ect activities some time ago. As a consequence, development of the organization and procedures for control ,
of site construction activities has not progressed beyond the initial phases.
As a result, a meaningful assessment regarding the construction phase of the appli-cant's quality assurance program cannot be made at this time. Therefore, this matter remains outstanding.
17.6 Conclusions In our review, we have evaluated the quality assurance programs and quality assurance organizations of the San Diego Gas and Electric Company, Stone & Webster Engineering Corporation and Westinghouse Electric Corporation to determine compliance with the Commission's regulations and applicable regulator', c d r and industry standards.
Based on our review, we conclude that these organizations have described acceptable quality assurance organizations, and that their quality assurance programs (1) comply with Appendix B to 10 CFR Part 50 and applicable guides and standards, and 17-12
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I (2) are acceptable for the design, procurement and construction of the Sundesert plant. l l
Subject to the satisfactory resolution of the outstanding issue discussed in Section 17.5 of this report regarding implementation of the construction related j I
activities in the quality assurance program, we conclude that the applicant's i
quality assurance program implementation is consistent with the status of the j
~~project.~~
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18.0 REVIEW BY THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS Due to the applicant's announced suspension of work on Sundesert, the application for the proposed facility is not currently being reviewed by the Advisory Committee on Reactor Safeguards. Should the applicant subsequently request us to complete the review of the application, then the Committee will be requested to review the application.
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19.0 C0KMON DEFENSE AND SECURITY The applicant states that the activities to be conducted will be within the juris-diction of the United States and that all of the directors and "rincipal officers of the applicant are citizens of the United States.
The applicant is not owned, dominated or controlled by an alien, a foreign corpora-tion or a foreign government. The activities to be conducted do not involve any restricted data, but the applicant has agreed to safeguard any such data that might become involved in accordance with the requirements of 10 CFR Part 50. The applicant will rely upon obtaining fuel as it is needed from sources of supply available for civilian purposes, so that no diversion of special nuclear material for military purposes is involved. For these reasons, and in the absence of any information to the contrary, we find that the activities to be performed will not be inimical to the common defense and security.
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20.0 FINANCIAL QUAllFICATIONS The Commission's regulations which relate to financial data and information required to establish financial qualifications for an applicant for a facility construction permit are Section 50.33(f) of 10 CFR Part 50 and Appendix C to 10 CFR Part 50. To assure that we have the latest information to make a determination of the financial qualifications of an applicant, it is our current practice to review this informa-E tion during the later stages of our review of an application.
Due to the applicant's announced suspension of work on Sundesert, we are not currently reviewing the financial qualifications of the applicant. In the event that the applicant subsequently requests us to complete the review of the application, we will resume our review of financial qualifications and will report the results of our evaluation in a future Safety Evaluation Report.
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~~1.0 CONCLUSION~~
S Our conclusions regarding the findings required by the Commission's regulations to be made prior to issuing a construction permit will be reported in a future Safety Evaluation Report should the staff be requested to reactivate its review.
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APPENDIX A CHRONOLOGY OF RADI0t0GICAL SAFETY REVIEW OF SUNDESERT NUCLEAR PLANT, UNITS 1 AN'J_2 November 30, 1976 Letter from applicant transmitting application consisting of General Information and Environmental Report.
March 4, 1977 Letter f rom applicant transmitting Preliminary Safety Analysis Report.
March 9,1977 Letter to applicant acknowledging receipt of Preliminary Safety Analysis Report.
March 11, 1977 Letter from applicant revising page 6 and 6a of the beneral
.Information portion of the application, clarifying participation by other utilities in tne Sundesert plant.
March 23, 1977 Letter to applicant stating that the application and Environmental Report were found acceptable for detailed review.
April 12, 1977 Application and Environmental Report docketed.
April 22, 1977 Letter to applicant concerning standard format for meteorological data on magnetic tape.
April 26, 1977 Letter to applicant stating that the Preliminary Safety Analysis Report was found acceptable for detailed review and requesting additional information.
April 26, 1977 Preliminary Safety Analysis Report docketed.
May 10, 1977 Meeting with applicant to discuss the results of the acceptance review for Preliminary Safety Analysis Report (Summary of Meeting issued May 19, 1977).
May 13, 1977 Letter from applicant transmitting financial information on prospective participants for the Sundesert plant.
June 13, 1977 Submittal of Amendment No. I to the Preliminary Safety Analysis Report containing a portion of the responses to the request for additional information, dated April 26, 1977.
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June 16, 1977 Letter to applicant concerning ser.urity clearance for review of classified Sandia safeguards reports.
July 1, 1977 Submittal of Amendment No. 2 to the Preliminary Safety Analysis Report containing additional responses to the request for additional information, dated April 26, 1977.
July 1, 1977 Submittu of Amendment No. 3 to the Preliminary Safety Analysis Report, concerning revised steam generator design (Model F).
July 28, 1977 Letter from applicant providing response to Question 031.10 which was omitted from Amendment No. 2 to the Preliminary Safety Analysis Report.
August 4, 1977 Letter to applicant concerning schedule for the radiological safety review of Sundesert.
August 12, 1977 Letter to applicant requesting additional information for the review of the Preliminary Safety Analysis Report.
August 17, 1977 Meeting with applicant to discuss the Preliminary Safety Analysis Report review schedule (Summary of Meeting issued September 6, 1977).
August 26, 1977 Submittal of Amendment No. 4 to the Preliminary Safety Analysis Renort containing additional responses to the request for additional information, dated April 26, 1977.
August 29, 1977 Letter to applicant transmitting information on fire protection functional responsibilities.
September 23, 1977 Submittal of Amendment No. S to the Preliminary Safety Ar.alysis Report containing a portion of the responses to the request for additional information, dated August 12, 1977.
September 28, 1977 Meeting with applicant to discuss September 23, 1977 responses and oustanding responses to the request for information, dated August 12, 1977 (Summary of Meeting issred October 19, 1977).
October 6, 1977 Submittal of Amendment No. 6 to the Preliminary Safety Analysis Report containing additional responses to t'he request for additional information, dated August 12, 1977.
October 11, 1977 Letter to applicant concerning implementatica of Regulatory Guide 1.97.
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l October 21, 1977 Letter from applicant transmitting an application for withhold-ing proprietary information from public disclosure submitted in response to Question 222.1, dated August 12, 1977.
October 25, 1977 Letter to applicant providing information on physical security assessment models subject to the requirements of 10 CFR 73.55(a).
November 1, 1977 Letter from applicant transmitting additional information relative to responses to Questions 010.16, 010.19, and v12.16 provided in Amendment No. 5 to the Preliminary Safety Analysis Report.
November 7, 1977 Letter to applicant requesting additional information on Preliminary Security Plan.
November 14, 1977 Letter to applicant requesting additional information for the review of the Preliminary Safety Analysis Report.
o November 17, 1977 Meeting with applicant and Houston Lighting & Power Company to discuss the implementation of Regulatory Guide 1.97 (Summary of Meeting issued December 6, 1977).
November 18, 1977 Submittal of Amendment No. 7 to the Preliminary Safety Analysis Report containing remaining responses to the requests for additional information, dated April 26, 1977 and August 12, 1977.
December 5, 1977 Letter to applicant requesting additional information for the review of the Preliminary Safety Analysis Report.
December 6, 1977 Letter from applicant concerning Regulatory Guide 1.97.
December 6, 1977 Submittal of Amendment No. 8 to the Preliminary Safety nalysis Report containing an update of regulatory guide positions.
December 15, 1977 Meeting with applicant to discuss the request for additional informa-tion, dated December 5, 1977 (Summary of Meeting issued January 9, 1978).
December 19, 1977 Submittal of Amendment No. 9 to the Preliminary Safety Analysis Report containing a portion of the responses to the request for additional information, dated November 14, 1977.
December 23, 1977 Letter to applicant concerning a request for additional information for the review of Preliminary Safety Analysis Report.
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January 5,1978 Submittal of Amendment No.10 to the Preliminary Safety Analysis Report containing additional responses to the request for additional information, ;ated November 14, 1977.
January 9, 1978 Letter to applicant concerning revised request for additional informa-tion for the review of Dreliminary Safety Analysis Report, dated December 5, 1977.
January 10, 1978 Letter from applicant concerning Implementation of Regulatory Guide l 1.97.
1 January 13, 1977 Letter from applicant transmitting the revised Sundesert Preliminary Physical Security Plan.
January 20, 1977 Submittal of Amendment No. 11 to the Prelimine y Safety Analysis Report containing additional responses to the requests for additional information, dated November 14, 1977 and December 5, 1977.
January 26, 1977 Letter to applicant requesting additional information for the review of the Preliminary Safety Analysis Report.
February 1, 1978 Submittal of Amendment No. 12 to the Preliminary Safety Analysis Report consisting of additional responses to the requests for addi-tional information, dated November 14, 1977 and December 5, 1977.
February 17, 1978 letter to applicant concerning schedule for the radiological safety review of Sundesert provided by letter, dated August 4, 1977.
February 27, 1978 Submittal of Amendment No. 13 to the Preliminary Safety Analysis Report consisting of additional responses to outstanding requests for additional information.
February 27, 1978 tetter to applicant concerning staff position regarding the design of the fuel building for Sundesert.
March 2,1978 letter to applicant concerning implementation of Regulatory Guide 1.97.
March 7,1978 Letter from applicant concerning staff p'<ition regarding the design ;
cf the fuel building for Sundesert.
March 20, 1978 Submittal of Amendment No. 14 to the Preliminary Safety Analysis Report containing remainir.g responses to outstanding requests for information.
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l March 21,1978 Letter frcm applicant submitting additional proprietary information in response to Question 222.1, dated August 12, 1977.
March 31, 1978 Letter to applicant concerning site excavation mapping and inspection.
i April 18, 1978 Letter to applicant concerning staff position regarding the design of the fuel building and associated systems for Sundesert.
April 18, 1978 tetter to applicant transmitting Revised Intrusion Detection Systems Handbook.
April 25, 10 ci Letter from applicant providing additional information in support of responses provided to various requests for iaformation.
April 27, 1978 Meeting with applicant to-discuss applicant's plans for Sundesert.
l May 1, 1978 tetter from applicant concerning implementation of Regulatory Guide -l 1.97.
i May 5, 1978 Letter to applicant concerning information on trairing security personnel.
May 8, 1978 Letter from applicant advising that the Board of Directors had announced suspension of all work on Sundesert and requesting that the Commission issue the Safety Evaluation Report.
May 22, 1978 Letter from applicant concerning schedule tor issuance of the Safety Evaluation Report.
June 12, 1978 Letter to applicant concerning information on physical protection systems.
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( June 27, 1978 Letter to applicant concerning schedule for issuance of the Safety l Evaluation Report, September 5, 1978 Letter from applicant affirming its request of May 8, 1978 that the Commission issue the Safety Evaluation Report.
l October 16, 1978 Interim Safety Evaluation Report issued.
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APPENDIX B BIBLIOGRAPHY FOR THE SUNDESERT PLANT SAFETY EVALUATION REPORT NOTE:
Documents referenced in or used to prepare this Safety Evaluation Report, excluding those listed in the Preliminary Safety Analysis Report, may be obtained at the source stated in the Bibliography or, where no specific source is given, at most major public libraries. Correspondence between the Commission and the applicant (Preliminary Safety Analysis Report, Environmental Report, and application) and Commission Rules and Regulations and Regulatory Guides may be inspected at the Commission's Public Docueent Room, 1717 H Street, N.W., Washington, D.C. 20555.
Correspondence between the Commission and the applicant may also be inspected at '
the Palo Verde Valley District Library,125 West Chanslorway, Blythe, California 92555 and at the San Diego Law Library,1105 Front Street, San Olego, California 92101. Specific documents relied upon by the Commission's staff and referenced in this Safety Evaluation Report are listed as follows:
Meteorology
~~1. American Society of Heating, Refrigerating and Air-Conditioning Engineers,1972,~~
~~" Handbook of Fundamentals," New York, N.Y.~~
~~2. Baldwin, J. L. ,197.>, " Climates of the United States," U.S. Department of Commerce, Environmental Data Service, Washington, D.C.~~
~~3. Crutcher, H. L., and R. G. Quay 1 , 1974, " Mariners Worldwide Climatic Guide to Tropical Storms at Sea - NAVAIR 50-1C-61," Naval Weather Service Environmental Detachment, Asheville, N.C.~~
~~4. National Severe Storms Forecast Center,1975, " Listing of Tornadoes for the Period 1953 - 1974," National Oceanic and Atmospheric Administration, Kansas City, Mo.~~
~~(Unpublished).~~
~~5. Orgill, M. M., and G. A. Sehmel, 1976, " Frequency and Diurnal Variation of Dust Storms in Contiguous U.S. A. ," ASnospheric Environment, Volume 10, pages 813-825,~~
6. Sagendorf, J. F., and J. T. Goll, 1976, NUREG-0324, "X0QD0Q, Program for the Meteorological Evaluation of Routine Effluent Release at Nuclear Power Stations, (DRAFT)," U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Washington, D.C.
~~B-1~~
~~7. Thom, H. C. S. ,1968, "New Distributions of Extreme Winds in the United States.~~
Journal of the Structural Olvision,'.' Proceedings of the American Society of Civil Engineers - July 168, pages 1787-1801, 8.
U.S. Department of Commerce, Environmental Data Service, 1964 (1972 Reprint),
"Climatography of the United States No. 86-4, Climatic Summary of the United States -
Supplement for 1961 through 1960, California," National Climatic Center, Asheville, N. C.
9.
U S. Department of Commerce, Environmental Data Service,1975, "Climatography of the United States No. 20, Climate of Parker, Arizona," National Climatic Center, Asheville, N.C.
10.
U.S. Department of Commerce, Environmental Data Service, 1976, " Local Climatological Data, Annual Summary with Comparative Data - Yuma, Arizona," National Climatic Center, Asheville, N.C.
~~11. Yanskey, G. R., E. H. Markee and A. P. Richter, 1966, "Climatography of the National Reactor Testing Station," I00-12048, Air Resources Field Research Office, Idaho Falls, Idaho.~~
Hydrologic Engineering 1.
Buresu of Reclamation, 1973, " Design of Small Dams," Second Edition, U.S. nepartment j of Interior.
2.
Chow, V. T. ,1964, " Handbook of Applieo Hydrology," McGraw-Hill Book Company, New York, New York.
~~3. Chow, V. T., 1959, "Open-Channel Hydraulics," McGraw-Hill Book Company, New York, New York.~~
4.
Department of the _ Army, Corps of Engineers,1971, " Additional Guidance for Riprap Channel Protection," Engineer Technical letter No. 1110-2-120, May 14, 1971.
5.
Department of the Army, Corps of Engineers,1970, " Hydraulic Design of Flood Control Channels," Engineer Manual 1110-2-1601, July 1970.
6.
U.S. Weather Eureau (now U.S. National Weather Service, NOAA), 1973, " Draft Report on Probable Maximum Thunderstorm Precipitation Estimates for the Southwest States,"
U.S. Department of Commerce, March 30, 1973.
Geology, Se_ismology and Geotechnical Engineering
~~1. Anderson, D. L., 1971, "The San Andreas Fault," Scientific American, Volume 225, pages 53-68.~~
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