Text

Application for Renewal of License R-71
	ML19290E213
Person / Time
Site:	05000142
Issue date:	02/28/1980
From:	; CALIFORNIA, UNIV. OF, LOS ANGELES, CA
To:	;
Shared Package
ML19290E212	List: ML19290E211; ML19290E213;
References
	NUDOCS 8003050468
	Download: ML19290E213 (250)
	v • d • e

O APPLICATION FOR A CLASS 104 LICENSE FOR A RESEARCH REACTOR FACILITY GENERAL CONTENTS Appendix I Financial Qualifications of the Applicant Apoendix II Environmental Impact Appraisal Appendix III Argonaut Safety Analysis Report (ASAR)

Appendix IV Emergency Resoonse Plan Appendix V Technical Specifications Appendix VI Operator Requalification Program 8003050 /7#

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APPENDIX I FINANCIAL QUALIFICATIONS OF THE APPLICANT CONTENTS Chapter 1 General Description . . . . . . . . . . . . . . . I/1-1 Chapter 2 Estimated Annual Cost of Operations . . . . . . . I/2-1 Chapter 3 The Estimated Cost of Permanently Closing Down the Reactor. . . . . . . . . . . . . . . . . I/3-1 Chapter 4 Annual Maintenance of Moth-Ball State . . . . . . I/4-1 Attachment A: Estimated Demolition Cost . . . . . . . . . . I/A Attachment B: Financial Statement . . . . . . . . . . . . . I/B I/i

APPENDIX II ENVIRONMENTAL IMPACT APPRAISAL CONTENTS Chapter 1 Facility, Environmental Effects of Co n s t ruc ti on . . . . . . . . . . . . . . . . . . I I /1 -1 Chapter 2 Environmental Effects of Facility Operatior. . 11/2-1 Chapter 3 Environmental Effects of Accidents. . . . . . 11/3-1 Chapter 4 Unavoidable Effects of Facility Construction and Operation . . . . . . . . . . . . . . . 11/4-1 Chapter 5 Alternatives to Construction and Operation of the Facility. . . . . . . . . . . . . . . . . . II/5-1 Chapter 6 Long-Term Effects of Facility Construction and Ope ra ti on . . . . . . . . . . . . . . . . . . . I I /6-1 Chapter 7 Costs and Benefits of Facility and Alternatives II/7-1 Attachment A The Environmental (TLD) Program. . . . . . . II/A-1 1

II/i

APPENDIX II ENVIRONMENTAL IMPACT APPRAISAL LIST OF FIGURES PAGE Figure II/2-1 Film Badge Locations Nuclear Energy Laboratory - 1st Floor. . . . 11/2-2 Figu:e II/2-2 Film Badge Locations Nuclear Energy Laboratory - 2nd Floor. . . . 11/2-3 Figure II/2-3 Film Badge Locations Roof View Overlooking Stack. . . . . . . . . . II/2-4 Figure II/A-1 Roof View Overlooking Stack. . . . . . . . . . II/A-5 Figure II/A-2 Average Quarterly Dosimeter Readings . . . . . II/A-6 LIST OF TABLES Table II/2-1 History of Annual Releases . . . . . . . . . . II/2-5 11/11

APPENDIX III ARGONAUT SAFETY ANALYSIS REPORT ( ASAR)

_ CONTENTS Chapter 1 Introduction and General Description. . . . . . III/1-1 Chapter 2 Supporting Facilities . . . . . . . . . . . . . III/2-1 2.1 Subcri tical Assembly . . . . . . . . . . . III/2-1 2.2 Neutron Genera tor. . . . . . . . . . . . . III/2-1 2.3 Activation Analysis Laboratory . . . . . . III/2-1 2.4 Heat Transfer Laboratory . . . . . . . . . III/2-1 2.5 To kama k La bo ra to ry . . . . . . . . . . . . I I I /2-1 2.6 Auxiliary Laboratory Areas . . . . . . . . III/2-1 Chapter 3 General Plant Description . . . . . . . . . . . III/3-1 3.1 Si te location . . . . . . . . . . . . . . . III/3-1 3.2 Site Geology . . . . . . . . . . . . . . . III/3-1 3.3 Site Hydrology . . . . . . . . . . . . . . III/3-1 3.4 Site Seismology. . . . . . . . . . . . . . III/3-1 3.5 Site Meteorology . . . . . . . . . . . . . III/3-2 3.6 Population Distribution. . . . . . . . . . III/3-3 Chapter 4 Plant Arrangement . . . . . . . . . . . . . . . III/4-1 Chapter 5 Nuclea r Reactor . . . . . . . . . . . . . . . . III/5-1 5.1 Re a c to r Co re . . . . . . . . . . . . . . . I I I / 5 5 5.2 Reactor Cooling System . . . . . . . . . . III/5 8 5.3 Reactor Instrumentation and Control. . . . III/512 5.4 Reactor Fuel Handling. . . . . . . . . . . III/5-15 5.5 Reactor Waste Disposal Control . . . . . . 111/5 15 Chapter 6 Comparison Tables . . . . . . . . . . . . . . . III/6-1 6.1 Comparisons with Similar Facility Designs. III/6-1 6.2 Comparison of Final and Preliminary Design . . . . . . . . . . . . . . . . . . III/6-1 Chapter 7 Identification of Agents and Contractors. . . . III/7-1 Chapter 8 Design Bases, Accidents, and Consequences . . . III/8-1 Chapter 9 Requirements for Further Technical Information . . . . . . . . . . . . . . . . . . III/9-1 Chapter 10 References . . . . . . . . . . . . . . . . . . III/10-1 Attachment A - Estimation of Effects of Assumed Large Reactivity Additions . . . . . . . . . . . . III/A Attachment B Radiation Doses Resulting From Release of Fission Products into Atmosphere. . . . . . . III/E III/i

APPENDIX III ARGONAUT SAFETY ANALYSIS REPORT (ASAR)

LIST OF FIGURES PAGE Figure III/1-1 Area Map - West Los Angeles. . . . . . . III/1-2 Figure III/4-1 Aerial View of Central Campus. . . . . . III/4-2 Figure III/4-2 UCLA Campus Map. . . . . . . . . . . . . III/4-3 Figure III/4-3 Nuclear Energy Laboratory - 1st Floor. . III/4-4 Figure III/4-4 Nuclear Energy Laboratory - 2nd Floor. . III/4-5 Figure III/4-5 Reactor Building - Elevation Sections. . III/4-6 Figure III/5-1 Reactor - Longitudinal Section . . . . . III/5-2 Figure III/5-2 Reactor - Tranverse Section Through Core Center. . . . . . . . . . . . . . . III/5-3 Figure III/5-3 Reactor - Horizontal Section at Beam Tube Level . . . . . . . . . . . . . . . III/5-4 Figure III/5-4 Fuel Plate . . . . . . . . . . . . . . . III/5-6 Figure III/5-5 Typical Fuel Cluster . . . . . . . . . . III/5-7 Figure III/5-6 Fuel Boxes and Coolant Connections . . . III/5-9 Figure III/5-7 Cooling Systems Piping Diagram . . . . . III/5-ll LIST OF TABLES Table III/1-1 Chronol ogy . . . . . . . . . . . . . . . III/1-4 Table III/1-2 Reactor Annual Use . . . . . . . . . . III/l-5 Table III/1-3 Reactor Activity . . . . . . . . . . . . III/1-5 Table III/3-1 Habitable Area & Population Within R Miles of the Reactor Site. . . . . . . . III/3-4 Table III/5-1 General Cooling System Characteristics . III/5-10 Table III/6-l@) Comparison Table - General . . . . . . . III/6-2 Table III/6-1(b) Primary Coolant. Nuclear Data. . . . . . III/6-3 Table III/6-1(c) Reactor Characteristics. . . . . . . . . III/6-4 Table III/6-2 Training Reactor Characteristics . . . . III/6-5 Table III/7-1 Identification of Agents . . . . . . . . III/7-2 Table III/7-2 Organization Relations . . . . . . . . . III/7-3 III/ii

APPENDIX IV EMERGENCY RESPONSE PLAN CONTENTS Chapter 1 Emergency Response and Accident Assessment Organization and Procedures. . . . . . . . . . . . . . IV/1-1 1.1 Types of Emergencies IV/1-1 1.2 Procedures During On-Duty (Working) Hours . . . . IV/1-2 1.3 Emergencies During Off-Duty Hours . . . . . . . . IV/1-7 Chapter 2 Emergency Equipment. . . . . . . . . . . . . . . . . . IV/2-1 2.1 Radiological and Emergency Equipment Available for use on the UCLA Campus. . . . . . . . . . . . IV/2-1 2.2 Equipment Available . . . . . . . . . . . . . . . IV/2-1 2.2.1 Non-Portable Equipment . . . . . . . . . . IV/2-1 2.2.1 Portable Equipment . . . . . . . . . . . . IV/2-2 2.2.3 Respiratory Equipment. . . . . . . . . . . IV/2-2 2.2.4 Prctective Clothing. . . . . . . . . . . . IV/2-2 2.2.5 Miscellaneous. . . . . . . . . . . . . . . IV/2-2 Chapter 3 Notification Methods . . . . . . . . . . . . . . . . . IV/3-1 Chapter 4 Notification Information . . . . . . . . . . . . . . . IV/4-1 Chapter 5 Emergency Response Training %nd Planning . . . . . . . IV/5-1 Chapter 6 Review, Updating, and Distribution of Emergency Response Plan. . . . . . . . . . . . . . . . . . . . . IV/6-1 Chapter 7 Implementation of Emergency Response Plan. . . . . . . IV/7-1 Attachment A Emergency Procedure. . . . . . . . . . . . . . . . . IV/A-1 Attachment B Chart of Organizational Relations. . . . . . . . . . IV/B-1 Reactor Emergency Call List. . . . . . . . . . . . . IV/B-2 Attachment C Radiation Accident Procedure: UCLA Emergency Medicine Center. . . . . . . . . . . . . . . . . . . IV/C-1 C.1 Background. . . . . . . . . . . . . . . . . . IV/C-1 C.2 Definition of Radiation Accident Cases. . . . IV/C-1 4/C.2.1 Radiation Exposure . . . . . . . . . IV/C-1 4/C.2.2 Internal Contamination . . . . . . . IV/C-1 4/C.2.3 External Contamination . . . . . . . IV/C-1 4/C.2.4 Contaminated Wounds. . . . . . . . . IV/C-2 IV/i

C.3 Referral of Radiation Accident Cases. . . . ..I /C-2 C.4 Notification of Hospital Personnel. . . . . . IV/C-3 C.5 Transport of Radiation Accident Victims . . . IV/C-3 C.6 Management of the Contaminated Patient. . . . IV/C-3 C.7 Radiation Injuries not Involving Contamination . . . . . . . . . . . . . . . . IV/C-5 C.8 Specific Therapy for Internal Contamination . IV/C-5 Attachment D D.1 General Procedures in Radiation Accidents . . IV/D-1 D.2 Contaminated and/or Injured Personnel in Radiation Areas Without Assistance. . . . . . IV/D-1 IV/ii

APPENDIX IV EMERGENCY RESPONSE PLAN LIST OF FIGURES Figure IV/B-1 Chart of Organizational Relations . . . . . . . . . IV/B-1 LIST OF TABLES Table IV/D-1 Time Dose Table. . . . . . . . . . . . . . . . . . . IV/D-2 Table IV/D-2 Distance (d) versus Dose Table . . . . . . . . . . . IV/D-3 IV/iii

APPENDIX V TECHNICAL SPECIFICATIONS CONTENTS Fo rwa rd . . . . . . . . . . . . . . . . . . . . . . . . . . V/ i Chapter 1 Definitions . . . . . . . . . . . . . . . . . .

1.1 Safety Channel . . . . . . . . . . . . . . V/1-1 1.2 Reactor Safety System. . . . . . . . . . V/1-1 1.3 O perabl e . . . . . . . . . . . . . . . . . V/1 -1 1.4 Channel Check. . . . . . . . . . . . . . . V/1-1 1.5 Channel Tes t . . . . . . . . . . . . . . . V/1 -1 1.6 Channel Calibration. . . . . . . . . . . . V/1-1 1.7 Unscheduled Shutdown . . . . . . . . . . . V/1-1 1.8 Reactor Shutdown . . . . . . . . . . . . V/1-1 1.9 Reactor Operating. . . . . . . . . . . . . V/1-2 1.10 Reactor Secured. . . . . . . . . . . . . . V/1-2 1.11 heasuring Channel. . . . . . . . . . . . . V/1-2 1.12 Reportable Occurrence. . . . . . . . . . . V/1-2 1.13 An Experiment. . . . . . . . . . . . . . . V/1-2 1.14 Experiment Facilities. . . . . . . . . . . V/1-3 1.15 Control Rod. . . . . . . . . . . . . . . . V/1-3 1.16 Readily Available on Call . . . . . . . . . V/1-3 1.17 Rod Drop Time . . . . . . . . . . . . . . V/1-3 1.18 Drop-Rod Scram . . . . . . . . . . . . . . V/1-3 1.19 Full Scram . . . . . . . . . . . . . . . . V/1-4 1.20 Inhi bi t. . . . . . . . . . . . . . . . . . V/1 -4 1.21 Safety L imi ts . . . . . . . . . . . . . . . V/1 -4 1.22 Safety System Margin . . . . . . . . . . . V/1-4 Chapter 2 Safety Limits and Limiting Safety System Settings 2.1 Safety Limits of Reactor Operation . . . . V/2-1 2.1.1 Applicability . . . . . . . . . . V/2-1 2.1.2 Objective. . . . . . . . . . . . . V/2-1 2.1.3 Specifications . . . . . . . . . . V/2-1 2.1.4 Bases. . . . . . . . . . . . . . . V/2-1 2.2 Limiting Safety System Settings. . . . . . V/2-2 2.2.1 Safety Channel Set Points. . . . . V/2-2 2.2.1.1 Applicability . . . . . . V/2-2

2. 2.1. 2 Objective . . . . . . . . V/2-2 2.2.1.3 Specification . . . . . . V/2-2 2.2.1.4 Bases . . . . . . . . . . V/2-2 V/iii

Chapter 3 Limiting Conditions for Operation . . . . . .

3.1 Reactivity Limitations . . . . . . . . . V/3-1 3.1.1 Shutdown Margin. . . . . . . . . V/3-1 3.1.2 Excess Reactivity. . . . . . . . V/3-1 3.1.3 Experiments. . . . . . . . . . . V/3-1 3.1.4 Control Rods . . . . . . . . . . V/3-1 3.2 Control and Safety Systems . . . . . . . V/3-2 3.2.1 Scrais Time . . . . . . . . . . . V/3-2 3.2.2 Measuring Channels . . . . . . . V/3-2 3.2.2.1 Bases. . . . . . . . . V/3-2 3.2.3 Safety Channel s. . . . . . . . . V/3-3 3.2.3.1 Bases. . . . . . . . . V/3-3 3.3 Radiation Monitoring Systems . . . . . . V/3-4 3.4 Engineered Safety Features . . . . . . . V/3-5 3.4.1 Safety High Level Radiation Monitor . . . . . . . . . . . V/3-5 3.4.1.1 Specification. . . . . V/3-5 3.4 .1. 2 Basis. . . . . . . . . V/3-5 3.4.2 Containment . . . . . . . . . . V/3-5 3.4.2.1 Specification. . . . . V/3-5 3.4.2.2 Bases. . . . . . . . . V/3-5 3.5 Limitations on Experiments . . . . . . . V/3-6 3.5.1 Experiments . . . . . . . . . V/3-6 3.5.1.1 Applicability. . . . . V/3-6 3.5.1.2 Objective. . . . . . . V/3-6 3.5.1.3 Specification. . . . . L'3-6 3.5.1.4 Bases . . . . . . . . V/3-6 3.6 Fu el . . . . . . . . . . . . . . . . . . V/ 3-8 3.6.1 Applicability. . . . . . . . . . V/3-8 3.6.2 Objective. . . . . . . . . . . . V/3-8 3.6.3 Specifications . . . . . . . . . V/3-8 V/iv

3.6.4 Bases . . . . . . . . . . . . . . . V/3-8 3.7 Primary Wa ter Quality. . . . . . . . . . . V/3-9 3.7.1 Applicability . . . . . . . . . . . V/3-9 3.7.2 Objective . . . . . . . . . . . . . V/3-9 3.7.3 Speci fica tions. . . . . . . . . . . V/3-9 3.7.4 Bases . . . . . . . . . . . . . . . V/3-9 3.8 Radioactive Releases . . . . . . . . . . . V/3-10 3.8.1 Airborne Stack Release Limit. . . . V/3-10 3.8.2 Dose in Unrestricted Areas. . . . . V/3-10 3.8.3 Liquid Effluent Releases. . . . . . V/3-il

?." Radiological Environmental Monitoring. . . V/3-12 3.10 Bases for Environmental Specifications . . V/3-14 Chapter 4 Surveillance Requirements . . . . . . . . . . . V/4-1 4.1 General. . . . . . . . . . . . . . . . . . V/4-1 4.2 Safety Channel Calibration . . . . . . . . V/4-1 4.3 Reactivity Surveillance. . . . . . . . . . V/4-1 4.4 Control and Safety System Surveillance . . V/4-1 4.5 Radiation Monitoring System. . . . . . . . V/4-1 4.6 Engineered Safety Features . . . . . . . . V/4-2 4.6.1 Safety High Level Stack . . . . . . V/4-2 4.6.2 Containment . . . . . . . . . . . . V/4-2 4.7 Reacto r Fuel . . . . . . . . . . . . . . . V/4-2 4.8 Primary Water. . . . . . . . . . . . . . . V/4-2 Chapter 5 Design Features . . . . . . . . . . . . . . . . V/5-1 5.1 Rea cto r Fuel . . . . . . . . . . . . . . . V/5-1 5.2 Control and Safety Systems . . . . . . . . V/5-2 5.2.1 Power Level (normal channels) . . . V/5-2 5.2.2 Long Power Level Channel . . . . . V/5-2 5.2.3 Count Rate (start-up channel) . . . V-5-2 5.2.4 Neutron Source. . . . . . . . . . . V-5/2 V/v

5.3 Rod Control System . . . . . . . . . . . . V/5-3 5.3.1 Shim (control) Rods . . . . . . . . V/5-3 5.3.2 Regul ating Rod. . . . . . . . . . . V/5-3 5.4 Cooling System . . . . . . . . . . . . . . V/5-4 5.4.1 Primary Coolant System. . . . . . . V/5-4 5.4.2 Secondary Cooling System. . . . . . V/5-4 5.5 Containment System . . . . . . . . . . . . V/5-5 5.5.1 P hys i ca l Fea tu re s . . . . . . . . . V/ 5-5 5.5.2 Emergency Sequence. . . . . . . . . V/5-5 5.5.3 Exhaust Duct Monitor (" stack monitor). . . . . . . . . . . . . . V-5-5 5.6 Fuel S tora ge . . . . . . . . . . . . . . . V/5-6 5.6.1 New Fuel. . . . . . . . . . . . . . V/5-6 5.6.2 Irradiated Fuel . . . . . . . . . . V/5-6 Chapter 6 Administrative Controls . . . . . . . . . . . . V/6-1 6.1 Organization . . . . . . . . . . . . . . . V/6-1 6.1.1 Structure . . . . . . . . . . . . . V/6-1 6.1.2 Responsibility. . . . . . . . . . . V/6-1 6.1.3 S ta f fi n g . . . . . . . . . . . . . . V/ 6- 1 6.1. 4 e, election and Training of personnel V/6-2 6.1.5 Review and Audit. . . . . . . . . . V/6-2 6.1.5.1 Composition and Quali fications . . . . . . V/6-2 6.1.5.2 Charter and Rules. . . . . V/6-2 6.1.5.3 Review Function. . . . . . V/6-2 6.1.5.4 Audit Function . ... . . V/6-3 6.2 P rocedures . . . . . . . . . . . . . . . . V/6-5 6.3 Experiment Review and Approval . . . . . . V/6-6 6.4 Rey;"ed Actions . . . . . . . . . . . . . V/6-7 6.4.1 Action to be Taken in Case of a Reportable Occurrence . . . . . . . V/6-7 6.5 Re po r ts . . . . . . . . . . . . . . . . . . V/ 6-8 6.5.1 Operating Reports . . . . . . . . . V/6-8 6.5.2 Special Reports (reportable occurrences). . . . . . . . . . . . V/6-8 V/vi

6.6 Reco rds . . . . . . . . . . . . . . . . . . V/6-10 6.6.1 Records to be Retained for a Period of at least Five Years. . . . . . . V/6-10 6.6.2 Records to be Retained for at least One Requalification Cycle or for the Length of Employment of the Individual, Whichever , Smaller. . V/6-10 Chapter 7 ALARA (10 CFR 50.36a) . . . . . . . . . . . . . V/7-1 Chapter 8 Re fe ren ce s . . . . . . . . . . . . . . . . . . . V/8-1 V/vii

APPENDIX V TECHNICAL SPECIFICATIONS LIST OF FIGURES PAGE Figure V/6-1 Organizational Relations . . . . . . . . V/6-2 LIST OF TABLES Table V/3-1 Radiological Environmental Monitoring System . . . . . . . . . . . . . . . . . V/3-13 V/viii

APPENDIX VI OPERATOR REQUALIFICATION PROGRAM CONTENTS Chapter 1 Exi s ti ng Program . . . . . . . . . . . . . . . . . . . VI/1 -1 Chapter 2 Complementary Programs and Alternatives. . . . . . . . VI/2-1 Attachment A Content and Scheduling . . . . . . . . . . . . . . . VI/A

APPLICATION FOR A CLASS 104 LICEllSE FOR A RESEARCH REACTOR FACILITY Based on Code of Federal Regulations, Title 10, Part 50 to U.S. Nuclear Regulatory Commission R. R. O'Neill , Dean School of Engineering ane'. Applied Science University of California Los Angeles February 1980

APPLICATION FOR A CLASS 104 LICENSE FOR A TRAINING REACTOR FACILITY Date of Application: March 30,1980 To: U.S. Nuclear Regulatory Commission Washington 25, D.C.

Attention : Robert W. Reid, Chief Operating Reactors Branch #4 Division of Operating Reactors Application Based On: Code of Federal Regulations Title 10, Part 50 Section 50.33

a. Name of Applicant: The Regents of the University of California

b. Address of Al plicant: Berkeley, California Location of Reactor: Los Angeles 90024, California

c. De/ription of business or occupation of applicant: Non-profit edu-catioc al institution

d. (1) and (2) not applicable (3) (i) State where organized: State of California Principal location of business: Los Angeles 90024, California (ii) Names, addresses and citizenship of directors and princi-pal officers:

2 Officers of the Regents The Honorable Edmund G. Brown, Jr. Herbert M. Gordon Governor of California Treasurer of The Regents President of The Regents 615 University Hall State Capitol Berkeley, California 94720 Sacramento, California 95814 (41 5) 642-3251 (916) 445-2841 Donald L. Reidhaar Robert 0. Reynolds General Counsel of The Regents Chairman of The Regents 590 University Hall 11661 San Vicente Blvd. , Suite 306 Berkeley, California 94720 Los Angeles, California 90049 (415) 642-2822 (213) 879-1956 Miss Marjorie J. Woolman DeWitt A. Higgs Secretary of The Regents Vice Chairman of The Regents 689 University Hall 1800 Home Tower Berkeley, California 94720 707 Broadway (415) 642-0505 San Diego, California 92101 (714) 420-1803 The Regents of tFa University Regents Ex Officio The Honorable Edmund G. Brown , Jr. Cheryl F. Biles Governor of California President of the Alumni Association State Capitol of the University of California Sacramento, California 95814 14462 Deerfield Avenue (916) 445-2841 Tustin, Cali fornia 92680 (714) 552-5011 Mike Curb Lieutenant Governor of California Forrest A. Plant State Capitol, Room 1028 Vice President of the Alunni Association Sacramento, California 95814 of the University of California (916) 445-9533 455 Capitol Mall Sacramento, Cali fornia 95814 Leo T. McCarthy (916) 444-3910 Speaker of the Assembly State Capitol, Room 3164 David S. Saxon Sacramento, California 95814 Pre,ident of the University (916) 445-8995 714 University Hall Berkeley, California 94720 Wilson Riles State Superintendent of Public Instruction 721 Capitol Mall Sacramento, California 95814 (916) 445-4338

3 Appointed Regents Edward W. Carter (1982) William A. Wilson (1988) 550 South Flower Streed 10475 Bellagio Road Los Angeles, California 90071 Los Angeles, California 90024 (213) 620-0150 (213) 270-3181 William K. Coblentz (1980) Gregory Bateson (1988)

Bank of America Plaza Suite 3100 Esalen Institute 555 California Street Big Sur, California 93920 San Francisco, California 94104 (415) 391-4800 Vilma S. Martinez (1990) 28 Geary Street, 6th Floor DeWitt A. Higgs (1982) San Francisco, California 94108 1800 Home Tower (415) 981-5800 707 Broadway San Diego, California 92101 Verne Orr (1988)

(714) 236-1551 Orr Enterprises 1930 Eleventh Street Glenn Campbell (1984) Santa Monica, Cali fornia 90404 Hoover Institution (213) 450-6425 Stanford, California 94305 (415) 497-2056 John F. Henning (1989) 995 Market Street, Suite 310 William French Smith (1986) San Francisco, Cali fornia 94103 515 South Flower Street, Suite 4753 (415) 986-3585 Los Angeles , California 90071 (213) 4SS-7236 Stanley K. Scheinbaum (1989) 240 Bentley Circle Robert 0. Reynolds (1986) Los Angeles, California 90049 11661 San Vicente Blvd., Suite 306 (213) 472-9541 Los Angeles, California 90049 (213) 879-1956 Yori Wada (1980) 1530 Buchanan Street Dean A. Watkins (1984) San Francisco, Cali fornia 94115 3333 Hillview Avenue (415) 931-8720 Stanford Industrial Park Pal o Al to , Cal i fo rni a 94304 Renee P. Turkell (1979)

(41 5) 493-4141 Placement Center University of California, Los Angeles Joseph A. Moore (1990) 405 Hilgard Avenue 351 California Street Los Angeles, Cali fornia 90024 San Francisco, California 94104 (213) 825-9641 (415) 392-5248 Yvonne Brathwaite Burke John H. Lawrence, M.D. (1988) 450 North Roxbury Drive, 8th Floor 220 Glorietta Boulevard Beverly Hills , California 90212 Orinda , Cali fornia 94563 (213) 391-4800 (415) 849-0197

4 Lee B. Wenzel George David Kieffer 1545 Wilshire Blvd., Suite 800 1888 Century Park East, 21st Floor Los Angeles, California 90017 Los Angeles, California 90067 (213) 483-1961 (213) 556-1500

5 (iii) Foreign R:;1ationships: The applicant is in no way owned, controlled, or dominated by an alien, a foreign corporation, or foreign government.

(4) Agent: The applicant is not acting as the agent or representative of another in filing this application.

The applicant is the principal pari.y.

e. Class of license applied for:

Class 104 License.

Use to which the facility will be put:

The reactor and its supporting laboratories will be used for the education of senior undergraduate and graduate students in nuclear engineering and related sciences. In addition to formal courses and demonstrations, the reactor will be used to support research at the M.S. and Ph.D. levels.

Period of time for which license is requested:

Twenty (20) years, or until March 30, 2000.

Other licenses applied for in connection with this facility:

Special Nuclear Material: (1) 4700 gms U-235 (irradiated),

(2) 4700 gms U-235 (fresh),

(3) Pu-239 as a 2 Curie, Pu-Be neutron source.

f. Financial qualifications of the applicant:

This item is treated in Appendix I " Financial Qualifications".

g. Deleted

h. Not applicable

i. Not applicable

j. No restricted data or defense information is contained in this application or in any material offered in support of this application.

6 Section 50.30

e. Exempt

f. Environmental consideration This application includes an Environmental Impact Appraisal for the UCLA Research Reactor (Appendix II).

7 Section 50.34

a. Not applicable

b. Final Safety Analysis Report, (FSAR)

An FSAR, the Argonaut Safety A_nalysis R.eport (ASAR) is included with this application.

(1) Environmental monitoring results are discussed in the Technical Specifications, Appendix V, Section 7.0.

(2) See the ASAR, Appendix III.

(3) See the E_rvironmental n I_mpact A_ppraisal (EIA), Appendix II.

(4) No structural weaknesses (earthquake vulnerability) have ever been identified. The biological shield was augmented in 1963 when the reactor power was increased to 100 kwt. Al uminum primary coolant lines, embedded in concrete beneath the reactor core and shield, were replaced (by-passed) by new lines in 1971 because of (external) corrosion problems.

The originally planned PuBe start-up was replaced by a RaBe source prior to initial operation. The RaBe source was renlaced in 1976. Ventilation stack monitoring problems (type of monitor and calibration) were prevalent until 1975.

The present monitor, a 4 liter, flow-throuah ion chamber, is believed to be quite satisfactory.

(5) Safety questions raised during the Construction Permit stage are unknown at UCLA today.

(6) (i) An organization chart is provided in the Technical Specifications, Figure V/6-1. Principal responsibilities are designated. " Demonstrated Ability" is the most common personnel qualification at intermediate and higher administrative levels.

(ii) Not applicable.

(iii) Not applicable.

(iv) Plans have been replaced by Technical Specifications, Appendix V, and (implicitly) Procedures.

(v) An Emergency Response Plan is included in this application, Appendix IV.

(vi) Technical Specifications are included in this application, Appendix V.

(vii) Not applicable.

8 (7) Technical Qualifications of Applicant The staff operating group of the Nuclear Energy Laboratory presently consists of a manager, reactor supervisor, instructional coordinator, two technicians, and a secretary. A resident hcalth physicist is assigned to the NEL. The reactor supervisor, instructional coordinator, and health tyS icist are licensed SR0's and one of the technicians is a licensed R0. Except for the secretary, all have a minimum of 5 years experience with the reactor.

Via a training program, several students become licensed RO's ear:i year and serve to augment the operating staff.

Staff time is distributed over the reactor and other laboratory operations. The effective full-time equivalent staff for reactor operations is estimated to be approximately 4.8 man-years per year (see Financial Qualifications, Annual Operating Cost).

Currently, the Nuclear Energy Laboratory is administratively responsible directly to the Dean of the School of Engineering and Applied Science, UCLA. One of the seven departments within the School is currently known as the Depart. ment of Chemical, Nuclear and Thermal Engineering. Strong ties have always existed between the NEL and this Department, particularly with the Nuclear subgroup of the Department (the " Nuclear Faculty").

The Director of the NEL and the faculty members of the Reactor Use Committee (review and audit group) are drawn from the Nuclear Faculty. The Nuclear Faculty now includes an ACRS member, a consultant to the ACRS, and (currently) an ACRS fellow on sabbatical leave. The knowledge and experience of the Nuclear Faculty is available to guide and advise the NEL in operational and safety matters.

(8) Operator Requalification Training This item is shown in Appendix VI.

c. Physical Security Plan.

This item is submitted under separate cover as a single document.

d. Nat applicable.

9 The Appendices are enclosed and are hereby incorporated into this license application.

I. Financial Qualifications of the Applicant 10 CFR 50.33(f)

II. Environmental Impact Appraisal 10 CFR 50.30(f)

III. Final Safety Analysis Report (ASAR) 10 CFR 50.34(b)

IV. Emergency Plan 10 CFR 50.34(b)(6)

V. Technical Specifications 10 CFR 50.34(b)(6)(vi)

VI. Operator Requalification Program 10 CFR 50.34(b)(8)

Additionally, the following document, although separately filed, is hereby incorporated into the license 'oplication.

VII. Physical Security Plan 10 CFR 50.34(c)

10 CERTIFICATE The applicant or any official executing this certificate on behalf of the applicant certify that these applications are prepared in conformity with Title 10, Code of Federal Regulations, Parts 50 and 70, and so solemnly swear (or affirm) that all information contained herein, including any supplements attached hereto, is true and correct to the best of our knowledge and belief.

On - 4 y0 , before the undersigned, a fictary"Pu lic for the State of California, personally appeared 7I( '/I 1 ,

known to me to be the person whose name is subscribed to the within instrument, and acknowledged that he executed the same.

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University of California, Los Angeles

APPENDIX I FINANCI AL QUALIFICATIONS OF THE APPLICAf!T FOR THE UNIVERSITY OF CALIFORNIA AT LOS ANGELES _

TRAINING REACTOR LICENSE NO. R-71 DOCKET NO. 50-142 February, 1980

APPENDIX I FINANCIAL QUALIFICATIONS OF THE APPLICANT CONTENTS Chapter 1 General Description . . . . . . . . . . . . . . . I/1-1 Chapter 2 Estimated Annual Cost of Operations . . . . . . 1/2-1 Chapter 3 The Estimated Cost of Permanently Closing Down the Reactor. . . . . . . . . . . . . . . . . I/3-1 Chapter 4 Annual Maintenance of Moth-Ball State . . . . . . I/4-1 Attachment A: Estimated Demolition Cost . . . . . . . . . . I/A Attachment B: Financial Statement . . . . . . . . . . . . . I/B I/1

FINANCIAL QUALIFICATIONS OF THE APPLICANT 1.0 GENERAL DESCRIPTION The University of California is a land grant college that is financially supported by:

(a) appropriations from the California State Legislature; (b) contracts and grants; and (c) fees.

There are nine campuses of the University and several laboratories.

This application pertains to the Nuclear Research Reactor on the Los Angeles campus. To the extent that the Legislature supports the University of California, those monies are distributed to the campuses, and the portion received by UCLA is further distributed to the various Colleges, Schools, and Departments. Direct support of the Nuclear Reactor derives principally from the operating budget of the School of Engineering and Applied Science (see page 19 of the attached UCLA Annual Report - Attachment B). UCLA provides further indirect support in administrative, surveillance, and maintenance functions. Except for the direct support of a Resident Health Physicist (via the Office of Research and Occupational Safety),

it is difficult to estimate the dollar cost of the other indirect support. These latter indirect costs are not included in the following.

The UCLA School of Engineering and Applied Science supports, from its annual state funded operating budget, a b oad range of academic programs in pursuit of the University's teaching and research mission.

The UCLA Nuclear Reactor is one such program. Periodically, these programs are subjected to academic review by the faculty of the School. Bas:d on these reviews, reconnendations are made to the Dean for the continuing financial support. Subject to the availability of funds from the State of California, continuing positive recommendations by the faculty, and continuing programatic need, the Nuclear Reactor will be funded at the levels indicated below.

In addition to the SEAS appropriated support, the Nuclear Energy Laboratory derives funds from direct support of specific contracts and grants, and from fees charged for reactor services. The NEL does not regularly issue annual reports of a fiscal nature, however, the approximate distribution of fund sources in the last three years is shown below. The academic year ends on June 30.

Academic Year Ending: 1978 1979 1980 (est)

SEAS Appropriation 131187 127636 151735 Reactor Earnings 9170 11130 21000 Other Income 71675 55923 67180 TOTAL 212032 194689 239915 I/1-1

2.0 ESTIftATED ANNUAL COST OF OPERATIONS The estimated annual cost of operating the UCLA Nuclear Reactor is a prorated portion of the cost of operating the Nuclear Energy Laboratory. These prorated costs are estimated in the following table.

Director (1/2 time) no charge Administrative Manager (2/3 time) 23512 Secretary (2/3 time) 10443 Miscellaneous (1/4 time) 8817 Reactor Supervisor (full time) 35266 Technicians (1 equivalent) 26981 Student operators (200 hours0.00231 days 0.0556 hours 3.306878e-4 weeks 7.61e-5 months / year) 1215 Health physicist 35266 (a) Total salary, wages, & benefits 141500 (b) Supplies and expenses 15000 (c) Equipment and facilities 10000 Travel 500 TOTAL 167000 Of the full-time staff of seven people, the Health Physicist is supported by the Office of Research and Occupational Safety (distinct from the School of Engineering and Applied Science).

The twenty year cost of operating the reactor, in 1980 dollars, is estimated to be 3.34 million dollars.

I/2-1

3.0 THE ESTIMATED COST OF PERMENENTLY CLOSING DOWN THE RE_ ACTOR The estimates below are based upon the following major as 'tions.

A. The moth-balling option will be chosen initially.

B. Moth-balling will include fuel removal and disposition, removal or decontamination of piping and process equipment external to the core, leveling the shield tank to the top of the reactor, providing cover for same, general decontamination, and decontamination or removal of ancillary facilities such as the rabbit system and the exhaust ventilation system.

C. That moth-balling can be accomplished in 18 months with a 6-man UCLA labor force including health physics surveillance.

D. The core contents of the reactor will remain undisturbed (except for fuel removal) for another 18 months (minimum). The reactor room will remain a controlled radiation area under a possession-only license, and the health physics surveillance, security maintenance, and demolition planning will continue through this 18 month interval with a 2-man labor force.

E. Core removal and demolition will commence no sooner than 3 years after the last reactor run. The cost of this last phase has been estimated by Rockwell International Corporation. Their letter, attached to this Financial Statement (see Attachment A),

includes further cetails of the assumptions used to derive their cost estimate.

None of the foregoing should be regarded as a scheduling commitment by UCLA. The plan is partially designed to retain a force of experienced radiation workers for a period plausible for fuel removal and shipment. Other arrangements are possible and might be employed. In particular, the high demolition cost versus the low annual maintenance cost suggests that the room could be used indefinitely as a controlled radiation area housing instructional or experimental facilities appropriate to such an area (sigma piles and subcritical assemblies).

The costs associated with each phase of the hypothetical shutdown are as follows.

Mothballing Shipment of 24 irradiated fuel bundles at $1000 per bundle $ 24000 Other shipping costs 20000 6 man years $25600 / man-year 153600 Miscellaneous supplies and expenses 5000 Coordination and administration @l5E of the direct cost 30400 TOTAL Mothballing Cost $ 233300 I/3-1

Ultimate Demolition See attached notes by Rockwell International $ 308000 Add: Demolition planning, supervision, and health physics surveillance 3.0 man years 106000 TOTAL Demolition Cost S 414000 I/3-2

4.0 ANNUAL MAINTENANCE OF MOTH-BALL STATE One man-year equivalent per year for maintenance of a controlled area (radiation monitoring, key control, lock maintenance).

Per Year S 35400 All of the forecast figures are in 1980 dollars (current salary cost, etc.) and no attempt has been made to i-troduce adjustments for future inflation.

1/4-1 e

APPErlDIX I FIfiATICIAL OUALIFICATI0tlS OF THE APPLICA'q Attachment A Estinated Demolition Cost from letter from J. H. Brindley, Manager Contracts and Proposals Rockwell International Energy Systems Group Canoga Park, California 91304 dated - October 30, 1979 I/A

Energy Systems Group 8900 De Soto Avenue Canoga Park. CA 91304 Telephone (213)3411000 TWX 910 4941237 Rockwell Ta&n 181017 International October 30, 1979 In reply refer to 79ESG-10473 Mr. Neill C. Ostrander Manager Nuclear Energy Laboratory UCLA - 2567 Boelter Hall Los Angeles, California 90024

Dear Mr. Ostrander:

Based upon our visit to the Argonaut Reactor Facility, discussions with the operations personnel and a review of the facility drawings, we have prepared an estimate of cost elements associateu with the ultimate disposal of the reactor.

It is our understanding the estimate is required primarily to comply with Item f-3 of the general information section of 10CFR 50.33. Your letter to me in which you request the estimate, specifically limits the estimate to cost elements relating to: materials handling, concrete demolition, shipping, and burial .

In preparing our estimate, we have made the assumptions as follows.

1. Prior to dismantling, the reactor will be mothballed for a period of 3 years to allow short-level isotopes to decay.

2. Reactor fuel and the startup source will be removed and dispositioned.

3. All extrane .us materials, experiments, instrumentaiton, and structures not directly supporting reactor operation will be cleared from the reactor room.

4. A facility dismantling plan will be prepared during the 3-year mothball period.

5. A radiological survey and an assessment of the contanination and induced activity in the reactor structures and supoorting systems will be made and included in the dismantling plan.

6. There were no major spills, lea'.s, or other contamination of reactor structures during the reactor operating history which would require extensive demolitior. and disposal of radioactive concrete and soil.

I/A-1

79ES1-Ad473 Oc tober 30, 1979 Page 2

7. Access to the reactor facility for trucks and demolition equipment and facility services such as cranes, light, electric power, etc., will be as presently available.

8. The dismantling sequence generally will be:

a. Drain water, disconnect water cooling and purification systems. Remove this equipment from reactor room.

b. Remove concrete blocks, survey for activation products, and if clean as expected for outer blocks, transport offsite for storage or demolition. Contaminated blocks will be decontaminated, if practical, or will be shipped to bu rial .

c. Remove graphite blocks, package, and ship to burial.

d. Remove water shield tank, cut up, package, and ship to burial. Remove bad shielding, survey and package, and ship to burial as necessary.

e. Remove activated concrete in inner areas of biological shield using jackhammers, scabblers, and hoe ram.

Package rubble for shipment to burial .

f. Demolish biological shield to floor level,

g. Remove beam parts and other penetrations from bioshield.

Package and ship contaminated materials.

h. Remove drain and cooling system piping. Pack and ship.

i. Decontaminate facility walls, crane, etc.

j. Pave excavation to existing floor level.

Cost for these above activities are developed using the following rates:

Radiation workers' labor $40 per hr Packaging costs $15 per cu. yd.

Shipping costs $1,000 per 20 T load Burial costs $6 per cu. ft I/A-2

79ESG-10473 October 30, 1979 Page 3 Labor required for removal of blocks, graphite, support structure, shield tank and water system, and decontami-nation of facility is estimated at 2,000 radiation workers' hours $80,000 Costs for burial, packaging, and shipment of the core contents and thermal column assuming a packing factor based on experience with such materials is estimated at $10,000 Cost for burial, packaging, and shipment of contaminated concrete from blocks, structures (above floor level),

and the pedestal foundation, process pit, and buried drains $35,000 Costs for demolition of massive concrete structures and excavatiores of drains, pipe, and support structure $40,000 Cleanup of facility and repaving of reactor room excava-tion costs $20,00')

Decontamination support materials: solvents, kimwipes, gloves, plastic sheet, enclosure structure, ducts, fil-ters boxes $15,000 Rental equipment: compresso , fork lif t $10,000 Sol purchase: saws, scabbler, chipping hamer, vacuum cleaners $10,000 Base Cost $220,000 Contingency 15 33,000 Contractor profit, overhead 25E 55,003 To tal S303,000 1/A-3

79ESG-10473 October 30, 1979 Page 4 These costs reflect our experience in decomiissioning similar facilities and are estimates for your budgetary and planning purposes only; hopefully, they will satisfy your requirements for relicensing the argonaut reactor.

Enclosed is your Argonaut Reactor Manual.

Very truly yours, ROCKWELL If1 TERT 1ATI0t1AL CORPORATIO!1 Energy Systems Group z 'n

.,I' #'l ;.x, , i . ._

J. H. Brindley Manager .

Contracts and Proposals cef:1030 Enclosure I/A-4

APPErlDIX I FIflAflCIAL QUALIFICATI0!iS OF THE APPLICA!iT Attachment B Finar.cial Statement from University of California, Los Angeles FIriAf1CIAL REPORT 1978-1979 I/B

University of California, Los Angeles FIN ANCIAL REPORT 1978-79 i

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UNIVERSITY OF CALIFORNIA, LOS ANGELES

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. I am pleased to submit herewith the financial

, report of the Los Angeles campur for the year ended June 30, 1979.

_ The records and accounts of the University from hich this report is prepared are maintained in accordance with the general principles recommended in College and University Business Administration and the American Institute of Certified Public

, Accountants Industry Audit Guide.

The accounts of the University are audited on an annual basis by the firm of Deloitte, liaskins and Sells whose report is transmitted to The Regents.

Copies of the report have been furnished to the Administrative Officers of the campus, an? ill be made available as appropriate to other i :sonnel 0

% LC . J James W, llobson Administrative Vice Chancellor

s FINANCIAL REPORT FOR THE YEAR ENDED JUNE 30, 1979 l

UNIVERSITY OF CALIFORNIA LOS ANGELES J

N A F R A T I v1: ;Um1ARY r The financial statemento includod in this report provide an accounting for the fundc a"ailabi to the University of California, Lou Angeles campus, for use i. n carryinu out the major objectives of the insti? : tion--teaching, re. < arch, and public service. This annual report covers these activitier and related optrationo during the 1978 ,9 fiscal year (a special year since i t is the Sath anniversary of the nove to Westwood of the Los Angelen carpus) , and ; hows the statun of the campus tinancial re,ources at June 3], 1979.

The !!niversit' of California recordr tinancial transactions into its ledgers in accordance with generally accepted prin-ciplea of accounting for educational institutions using concepts of fund accounting, Eund accaunting doeu not tell how nuch the University r:a d e or lo ;t , but what resourcen it started the year with, what resourc. were received, what were spent, and what rennined at the end of the year, all in the traditional purpose a r o u p i n '; 5 - Current, Plant, and Loan Funds.

A summary of operations and financial condition in contained in the following conmenta, with additional detail provided in several schedules within the report CURRENT F U t: D 5 The basic operating funds of the University are the current funds, Current fund balances are separated into those which are restricted by donors or grantors and those which are unrestricted. Restricted fundo nay be expended only for the purpose indicated by the donor 3r grantor whereas the use of unrestricted fund: 15 determined ty the University to achieve its educational goals Por stateaent purposes and to provide consistency with prior years, the unrestricted funds have been 3eparated into General and Designated fund categories.

Current fund receipts, available as the princi ial source of funds for financina Uaiversity operatione in 1978-79, totaled S490.0 million, representing a 7.04 increase over the pre-ccding year. The Federal and 3 tate Governments continued as the major source of these funds, providing 18.2" and 36.21 respectively, along with Teaching Hospitals, providing another 20.21 A significant rate of crowth was experienced during the year although no range adjustnant was nade in salaries. The total expenditures of $478.R million represented a 9.4%

increase over 1977-78, li reconciliatian of the difference 1

between campus receipts and expenditures iu presented on the last page of this report.

A summary comparison of income between the current and preceding fiscal year is shown below (for nore detail, see l Current Funds Receipts schedules, Pages 6 and 1 11 l

l l Comparison of Receipts by Pund Source (in Thousands)

Tota 1 1977-78 1978-79 Amount Percent Amount Percent State Government $ 176,726 '8.6 U 177,407 36,2 Federal Government H2,055 l' 9 R4,600 17 3 Tuition and Peei 37,157 8.1 39,044 8.0 Teaching Hospitals 87,7H0 19.. 98,755 20.'

Educational Activities 30,135 6.6 37,950 7 7 Auxiliary Enterprises 17,370 3.8 20,045 4.1 Private Gifts, Grants 15,666 3 ,. 4 18,439 3 t1 Local Government 1,993 o4 g,476 g,9 Other sources 4,630 1.0 4,627 0.9 DOE LaboratorS ( G E t; 12) 4 a t. g 1 o a c,; q o S .; s 7 p,c 100.0 $ 4 soy 7 ? _ 100.0 Additional sources of funds, described as transfers within the University system, are shown in the sunnary of Changes in rund Balances Utilizing the availablo current funds receipts, the total ex-penditures for financing University operations in 1978-79 are compared by function with the expenditures of the prior fiscal year:

Comparison of Expenditures by Function (in Thousands)

Tota 1 1977-78 1978-79 Amount Percent Anount Percent Instruction $129,924 7 9 . ~/ 0137,745 28.8 Research 70,358 16,1 77,571 16.2 Public Service 8 ,17 ^. 1.0 B,952 1.9 Academic Support 49,626 11.3 57,167 11.9 Teaching Hospitals 93,6'O 'l.4 103,806 21.7 Student Services 14,964 3.4 15,980 3.3 Institutional Support 17,948 4.3 l'3,475 4.1 Operation and Mainte-nance of Plant 20,427 4 ., 7 20,224 4.2 Student Financial Is i d 14,567 3.3 17,067 3.6 Auxiliary Enterprises 13,735 3 .,1 16,238 3.4 DOE Laboratory (GEN 12) 4,334 1.0 4,546 0,9 S437,684 100.0 S478,771 100.0 2

In carrying out the functions of the University, salary and wage payments to faculty, research personnel, administrative, hospital and other personnel accounted for 0282.8 nillion in 1978-79, which represented 59.1% of total campus funda expended.

For further detail, see additional Current Funds 1: x p e n d i t u r e schedules in this report, pages 7, 8, and 14 - 34.

PLANT FUNDS Plant funds are funds designated for expansion of the physical facilities of the campus, and for improvement of existinq facilities. Funds are expanded for new construction, initial equipping of new buildings and additions, for renovation, re-modeling and alteration of existing structure;, and for studies and surveys related to planning for the physical plant of the campus. A summary of plant activity for fiscal 1978-79 is nhown below compared to a summar/ of plant activity for fiscal 1977-78.

1977-78 19 M-74 1 Increase Total Expenditures for year 013,409,355 028,973,549 116.2 Construction in Progress 26,999,107 48,905,850 31,1 Increase in Plant Assets 6,615,70: 7,171,537 8.4 Included in construction in progress are the following: HSC s-LevelExpansion Project 05,773,000; HSC South Parking Structure, 09,024,000; Factor Building, 511,682,300; Jerry Lewis Neuro-muscular Research Center, $2,522,700; Re;idential .3uite and Related Parking, 02,064,700; and Westwood Plaza South Mall and Parking Terrace, 03,468,200, plus other smaller projects.

Capitalized as investment in plant is the purchase of the Land-fair Apartments at a cost of $2,500,000. Also included in the amount capitalized are the following: School of Dentistry Completions of Shell Space, $867,300; HSC Heliport, 5397,971; HSC Modular Units, $1,166,406; and Modernization of Hospital Operating Rooms, $567,800, and other projects completed in 1978-79.

LOAN FUNDS Outstanding student loans as of June 30, 1979, totaled

$34,926,651, an increase of 2.1% over the $34,221,452 outstand-ing amount in the preceding fiscal year. Unloaned funds at that date amounted to 5 . 0 'h of loan fund balances available.

Of the outstanding notes receivable at June 30, 1979, 025,313,149 or 72.5% were provided under Federal programs.

3

TABLE OF C O ti T E !4 T S Page Schedule A Current Funds Receipts (Comparison with b Prior Year) 7 B Current Funds Expenditures by Function (Comparison with Prior Year)

B C Current Funds Expenditure.' by Fund Source (Comparison with Prior Year) 9 D Balance Sheet 11 E Summary of Changes in Current Fund Balances 13 F Current Funds Receipts A

14 G Current Funds Expenditures by Uniform Classification Category Current Funds Expenditures by Department 19 H

Current Funds Expenditures by Fund Source 32 I

1 35 J Reconciliation of Difference between Current Funds Receipts and Expenditures

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tOS ANG! l L5 SCHE DUU A CURRENT FUNDS RECEIPTS 1978 79 Cut t ent t u m ts e m e pt s, a . ) .iat o as ibe p inc rat mu - i i h e th t a +i u) U'-

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t mo u t u .: egor m.e ce of the .e fmich.Ufmi>'a 5e % t e , r i ac ,y . ( : un mi hit anmh ' h) 2 ',. 6 i t ) .d* r *L is STATE GOVERNMENT 36.2 %

TUITION 81)% E D UC A TlO N Al.

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1977-78 1978-79 } 1977-78 1973 79 inccc a w Increav TinMn & F us S 37.157 S 39.094 31 80 $ 1.937 82.055 84.G09 1/.9 17 3 2.554 3.1 F eder al Govet onent (E =ctud ng GEf417) 17G.72G 177.407 38.6 30 2 t>31 04 St.ite Gover nnvnt 1.993 4.4 / G 0.4 0.9 2.483 124 G L oc.d Goverinnent Private Giit ,. Gr ants 15.6G3 18.430 34 3.8 , 1,7 /1 17.1 j

and Conte acts l 30.135 37.950 GG 7.7 7.31 b ?59 S. des and Services of Educational Actmties Sde, and S rvices of 17.370 20.045 38 41 2.t; / 5 15.4 Aux:bary Enter pr ises l Sate; arid Sorvices of 8/.780 93. / L5 19.2 20 2 10.0 / b 12.5 Teadnng Ho .pitals Other Sources 4.G36 4.627 10 09 (U) (07 DOE Laborator y (GE N 12) 4.356 4.570 1.0 0.9 214 49 S457,876 $489.972 100 0 100.0 S32,09G 7.0 TOTAL CAMPUS 6

LOS ANGELLS SCHEDULE B CURRENT FUNDS EXPENDITURES BY FUNCTION 1978-79 T h.. g eatest vot a rn. . o' ex;wntia i , $13 7 7 m n omtm ; to tx- mr " s t ru, t mo e s,po s+ .nd n er ew m s ?H F ,, of tk.* e n t ;l c ur r en t f and e x pe rd ru . s T h , i. ,n g er v. a . fJ: >ml in s e 1.9 f..#. tu n g hosturais, wth $1012 ma.nin, or

71. 7% rf the turien tund *

pen 1 run ' < 1 r e .,* s tb v ' i S 7 7 G m , m r ,r s ev u t h pet son + el, .irt+ r c o a t n 3tA v .1 s t rw r r - '

1 GJ t Sm n y and w ir psynsnts to ta ulty,

, , m to t o ul 51, / U rn a nu e tu'i rm 19 70 7 9, o' LU 1 % o f I"e tuf f c 3mpas f unth e xper +d T his r. ornpe t s t o $ 2 719 m .'n > i 62< "i1911 73

\ INSTRUCTION

,-,3 3 OPER ATION & MAINTENANCE ACADEMIC -

OF PHYSICAL PL ANT SUPPO R T llM

- -AUXILIARY E NTERPRISES INSTIT UTIO N AL %

3 WPPOR T STUDENT SE RVICES 4.1% 3E

' ' - PUBLIC SE RVICE #~

STUDENT AID 3.3%

1D%

RESEARCH T E ACH ING llNCLUDING GEN 12) HOSPIT ALS 17.1 % 21,7%

) e lotal (in Thousands) Percent of Total Dolla r Percent of 1977 78 1978-70 1977 78 1978-79 Increase Ir. crease hstruction S129,924 S13/,145 29.7 28.8 S 7.321 10 FMe.u ch 70,358 77,5'1 I G.1 1G 2 7,2 ' 3 10.2 Publ-c Ser vice 8.172 8,952 1.9 1.0 780 9.5 Acad. mic Suppor t 49,G20 L /,167 11.3 11.9 7,541 15.2 Teachmq Hospitals D3,G29 103 806 21.4 21.7 10,177 10.3 Stu&nt Sennces 14,904 15,980 3.4 33 1,01 G G7 Institu tional Suppor t 17,948 19,475 4.1 4.1 1,52/ 8.5 Opeution and Mainte 20,427 20,224 47 42 (203) (1.01 nance of Plant Student Financol A<d 14;G7 17,0G 7 33 3G 2,500 17.2 Auxihary Enterpr e,es 13,735 1G.233 3.1 3.4 2,503 18.2 DOE Laixnatory (GEN 12) 4.334 4,54 G 1.0 09 212 4.9 TOTAL CAMPUS $437,G34 $478,771 100.0 100.0 $41,087 9.4 7

[:),

LOS ANGELES SCHEDULE C CURRENT FUNDS EXPENDITURES BY FUND SOURCE 1978 79 ces n r .nt tu mes em .d u 7; a m m , oso - , , - , o > o<s w .. m e , s o w o .o ,4 55 5 m u.an p owird 3') 2 % of the toui. ect t', to. A 7v -

ar t t e s a 1 -i mNr 5 7J L , i m < r l ', -t ;

. 1*

are im m* p r souro-s of fure ng, alor. g w 6 f eding N r ou h. s cru il SW 6 N a. 19 . ' ; t +. < >

tatue, 'm turemt f unds rye,4,ts of StJ0 0 ny!..m , e- J t m t.o. r u o' M id E m > <

$ 11.. ' r r i rrawh a y e.d cN tra nstr's of f und,.

GENERALFUNDS E NDOYM EN TS gyg 12%

OTHE R 57 ATE OTHER N 93 2.3% -

s' _ ~ PRIVATE OlFTS

.- 3jg F EDER AL GOVE RNMENT - ~ -

- AUXILIARY ENTERPRlGES ONCLUDING GEN 12b 15 A %

\ ,

3.4%

N STU DE P(T F E ES Ca%

\

E DUCATION AL ACTIVITIE S TEACHING HOSPIT AL 72% 102 %

Total On Thousands) Percent of Total Dollar Percent of 1977 78 1978 79 j 1977-78 1978-79 increa se i ncre.ise Rmo: aces Utihred Sute of Cahtornia:

General Funds $175,65 7 S182,448 40 1 38.1 $ G.701 39 Appropriations & Con.racts G.054 5,4:9 4 1.1 (595) (9 8:

Uneted States of Arrenca:

Gr ant s 51,094 55,572 11.7 11.0 4,478 88 Contracts 14,00G 13.343 3.4 2.8 ( 1,5G3) (10.5)

S:udent Fees and Tuition 28.942 33.101 06 G.9 4.159 14.4 Teach.nq Hospitals 84,122 91,G33 19 3 19.8 9,910 11.7 Educat:onal Activit.es 28,484 37,433 G.5 7.8 8,940 31.4 Auxihary Enter poses 13,554 10.270 31 3.4 2.716 20 3 Povate Gif ts Grants 14,250 16,301 33 34 2,051 14 4 Endn ,rnents 7,554 8.455 1.' 1.8 901 11.9 Local Governmont 2,039 4,544 0.5 09 2,505 12 2.9 Other Sources 6,094 G.001 1.4 1.4 SG7 9.3 Major Department of Energy 4.334 4,540 10 1.0 212 4.9 (GE N 12)

S437,G84 S478,771 100.0 100.0 $41,087 9.4 8

L. ANO L) %

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.'s J p.f,.

\

s h[Vk'>

\

v' s 9

31

LD$ A3XL.lb 1978-79 C a lo.h i t u..e t.Xf L:.e tTLt t s sY Ft*J su. e6 t N t,ed t.l e l Cur revit t send v U I , t c 3 p,o g y yp S4Lartes 'a her sees Total snecatrtitcJ -

est r n te1 a nd w.=e e n--- t v'ndituren s

Ira i, r e r a 64t.) RAL fL*?S Irla t rur t t un 5 9), % 9, I s 1 . wi . u.y , i t .1 -- , . , )i $ 1, t . * # . ' ,

l*J , b l ! s ) , n 5 ) , l .1 itesea r s t, 4,915,. '

9, v l et ,

2 - b ,'+ d f , b 3. I 193,9J. 16 > ,12*

Nolic serv ice 5,T 3 u l , .n *, - --

..,6, m --

A(4deuit s u p p.; r t k,,e *.h. 19,0e , tW e> ~ 12, 'vle= , h e l 11,112,19' 11.4, le Teac hing hospital m 4,54>,924 e,Wi,92, - 1,la; d,WJ,1.' ~

5tudent servlien 3,9. ) ,i .1 3, v

l, > = ~ i , ts ; . . > i

6,tus,114 u,$15 Ine r t t ut tan.1 as, po r r 14,o45,65 44,di$,1b2 ~

! ! ,d li, * + t>, * % 3, 5 33 1 i i i. t Uperailsn amt maintenance .t pla s 13,911,53 i n ,91.; e - 1, N , h / i d , *62 , ti le ),*17, .1 Etudent f inanc ial a tJ 6 t 2,s 13 ta;f ?) -- --

't lig ? $ --

Totat ---

is. l., o

i :4 1 x 2, 4a

v')9 -

11 2 V ,- L. 0 ,

m 1- i ,m lL.4 .,

41L.*1---

9 4

--T '.1 T 10 . A.. a >sra Inst r uc t ion 1 ,in,'2n I. 94,111 --

o ,' r , t* t l ',6 )l t F i .h ,97 Mesearch .!!,J.* 211, j = - 211,' e - -

Public se rv i.a .c , 1 )) so e , l l 7 -

1.3 2, V 7 *.n> ~

9 t ,m '

Acad e mic suppc t t - ,1 1. ' . ,1 1. . --

I ?2 -

htudent se r v i c e s 9,4, .. 9,. ,, - , 'd l t , l e 's * . ..y* 1,8cel 1

',99) 47,9'43 -- .3,72, em -

In6t it ut ions. e uit h t t,de n t t i aa nc ia l a1J . 23 , ,*25 -- -

7.- 21 -

A4A111 dry e n t e r ;. r 1 Se s ._ , 32 ; t e -

m ,/ li lla,= < _ _ --

Totat ,t,i ,siv ,i, gu -- is 4 - t . :gg. i ., . , _ _

> t Ak Ai. t ,V t n .w.r '

orant=

5 f%b 1 r LDLf1M b Yk1 i N g F,9k4,h * ,-%3 t I-M %GP4 I'c Il )44 I **

I [, " I ) , 'e I I , Q $ h lt i (,9 g -

Pub i t e on rv n e 40 ,9 - # 't', vie * ,1 4- ) e, n+ -

A ddNIC SubPJff 1.43F,Y79 ~

I , 4s l' , 'O

h 9 } } , ') I '

j),*hI **

Te a c h i na, fr.6 pi t al s . 6

,lau 19..rS *l,**i --

St ude n t se r v ic e > 291 70 --

c)2 *'b ' t , 'c 31 . ,5.) --

Ina,t i t ut ier.dl t opo r t tv , i ~ s ,25i .9,5 5 , elJ --

'tuacot f l e.a nc 1 A 1 Aid

- N6 2 , } % -

H2,L s -- 6 %' , ) *- -

%4 L 11ary enterprlww e 4 --

t,wt . g1 >_ --

Tatal ii,n-,199 --

2 92 - ,19't ), e , il ? , eg 4

(. e.t r a c t s ins t r w t loa 147, * ~

1. ' ,9 : I ) 51 ' , tr i e --

fWhfalih 1 2 ' ' 'O' ' , 'f ** *-

a 4 H ) t- M , e $[ ),2 )h %, J M -*

!%t.13 c se rv 't e L 4, L t's ~

' 19, 0 0 'f il e : .1 1 j fl.l it -

M ade131C Su ppvi r t ' ,9 w 'i -- '

',9 } l t F , f' t .' IG t P --

Institut inal support 17, ~

v', --

9'. --

Student f i c.a nc i a l aid _ ey * . --

g .. --

.t ,_ . 7 Iut41 1 3. . ,3 d **

I ! ,D ',DS> b\4 7 ;/0, . y ) -

retal tederat .~v r m e, _ t y i,,, --

4,s i ,se u 'i,iri 3i ac ..n _-

T

' \ s\ ,

(~D \ U \ W'\\,-

\\W f &e. a n

W3 A;6.r.Lt S 197(6-79 t -M',' F N EXPr %1TJio d 51 Fi r N ; '- w i.e.o 21, 6 urrent tu .= : 1. I r i t. !L-s.al rte- .. t 6er vs.

T.'.. i r ro r .1tM m -t r i t . t a 4 m...s e att r

re ster 5 H C I A;. it ATE. AP Psi. .P P I ET 4 A'.._ ."R n :

1.ne t I wo ! ion 4 4 . ) ~ , ,4o. , ;j e -

elesee r .:n 1 -- 4 i

. 1 % .r , 's ; . -

f%ti t< msev ic e n. ..n -

09,, , , , t!.s --

Acadeele m W port ,jl; --

i s /j. a e -

a t u4 esi t serv 1(en i e -- e , r.o ,1 ; , --

Ins t itut ional suppo rt >

..u. - i -. , -

Ope rat Lon anJ LainteNam e ui ,. i a c . , l a.i -

,tr ,e -

b l ud e r.t fihAD(lal aid _-_3.1 . -

1 & .

.__- . __ a .a- ,

-_ . . - - . ~

Totai . , ;l --

, , _. g ,3,; $ . ,

13. AL v o W.'a 9.1 Ir s t r u<. t ton 1* , --

> r,t > ,t.. .i -

neaearch i ;3 --

Put i le serv ce ij. 4 -.

4 s+ 1 ,,. ,jn, --

Ar a.tes ic supy r t 3, 4., --

9. f 1, , . -

l e a4: h t y hospitais e ~ --

_ 3.- is.,r* ,4 -

Student ne rv ic e s 11> 73 -

,i 3 ,><a -

I ns t it ut isnal su;v<.r t r_ -

w ;, -

v;wration and mal,tena,te f , c- t l i ,17 - -

l 3,1 ; - ~4, t,91, -

htudent f inanc ial 413 . -- - -

Ava111ery enterprin s see x+,

---.._-. ._-2

- i 1._.-

Total - - . . -

t... '4

- - _ _ - . -_x , ; _ t _ .

13e -e i Fili v AT F t. tis g o r. A.. ' >s *sA '. 3 Intraustion i, , 1 24 --

1, ,)34 L, '.,, -

he me a rc h i ,

1 ; ,1 , *;i y; , g , t .3 Pubite service '

3,=,, -

p, , 1 1, __

A ademic s ur ;m r t i ,ju te . ,

r ., ; --

Te4c h i ng 'mapitals . -

6li,;5 m .i . - --

St dent se rv i< c s --

i lr t' 4 -

i ins t i t ut ional w;y t t 1, , e -

a, * . # 1,i 4 -. ~

ope ration and e.41 r t e na'. e >+ r 4 mi i -

!il.1i --

312,14 -

htudent i knanc ial al. ' --

i * --

31. e4 --

Auxiliary er.ter;rtats __

. a, ,3, r Total 1- 4. _1_,, e , . , i - 44 .,a

! .i 1_u _ -

- _r - ft, >,. q>.;. ,

E M= >=Mt3T 6 L- ,

Instra. flan 1, 3,_i# 6,17 . , ; ; ., 3 3,, ;g, _

hesear c h , i f- . ,. j s ],, . ,. 3,n 33) j, s ;g. 3,, , ,g3 ,

Pu blic settice . .> j , ,e _ . ,

Ac adem ic s up po r t I,lli, *' l lla e ..i, d i ,14, M --

le ac hing hospitals 11d -

,,914 -

.,914 --

5tudect se r v ic e s 6.',31. h . 7, ~3 14 -

,33. n4; e -

Inst!'utional s up;x r t 1,i ,37 3,1! ,9;) +, - ;4 j, ,

5, Ope ra t ion a nd a41 n t e n4 v e of ;la, J!r,941 lin .17 ni 9, $ ;e 1; , e 3 -

h t uder. t financial did 91 ,5,1 23 it; ++., '.'1 -- 4 4,,Mi -

Aemilla ry ente r pr ises 44 *4 7e -- --

.i 4

.1L'l l es L _ _ . _ ._-

Total 4,. 4,M t; o th 7.1 3 3 , > _ , , - ., g,, -, ,,1,,, .-y -t-(

\ x, Q[

,.N

.\

( .- \ \N',

cQ.,, \

+ . v. e ,hV a KNN - G.* -

'U) 9

\) '

Q \;w '

e 33

L' - S A.O . F L F S 19?3 74

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

ne ars! tm: F u* st' ' 'ii t. s ey tr s '. k. 'edule

.-_m_ _ . _ _ _ . _ . . - ,

to risert i rren u.._.

41 e r t .m t r*cr l e u.

'o

.' ei r e s t r_q c t e d s = tl. ci i- . g. = F- e t t r :. . '**.

S A1.t s A*.D sF p W T ; t s : 5 FD it!' N Ai A ' Tit iA e

i tint rw t i ori s *'.9,* * ,4~ -

11,'14 1

sesear h t is9,+ 1

', 8 ; e4 Putlte se r vi c e ',41 ,3;a ,ve ,1 ! -

" +4 ^1.

8 4, '

11,11' ?le 11 ;6 4( ade nt e eig ps rt l' 2 . *?. < , '3\ -

,56 4 i 4,

Student serviess '

a-1 i 41 5.

Instituticaal supN rt IT,it? 1 1, '- 4 '. ~4 ='i Ope rat t or. n a,* in.i l n t er e ic e of pict 4-) 44 - .i S t 2d nt f i na n. 141 aid , l _s . ,1% - -

,!4..

Tutal t*,5 ,q t',4t 7, *

i e

, ,1?1 . 1, 4 S Ap s A 's s! R\ !i5 ;'l A'. ) ! !ARY F N!' Pk s ke n e.s r t5 * ,

< w --

S t s u. ..t t i e.a m: t a l Stj r e e 8

Aa n i l l a ry erterprises $4'A< l '. e 3E i

=

9 2. 1 1 L TGt4I !

, j i I & J 2 3 S ALf 5 AND 5>$4 't3 vF TF A: .

Nr;'s s

-1, " 4 P --

Acateg{( a..p ; 0 f f I*ach l eg h.. spi t a 's ,* b e (4,7 **

, 14, *

6

, , , ! ?, e

,Fi

'o'ai Sh ,s 4 ,o'( *

, 4, r

, 1 ~_ , d,*

U r T..F. R . x ! .F ,' '

Instru *iw a :4' -

e i kenc a r c h

1 ,

Pn51Lr serv 1 c * *1 4 1 - 44' i

4 5 'a A =dem!< *arrert 19, t:14 e' 5 * * * > t, , l' I,it! ett Stuaent services -

e > a . ,r<

' ;1 .9, Inst It ut tora sepp it 5

' *), *s ape ra t 1< n and w 4n ere. , *nt 1 . * ,174 - i e e 5todert tir4: cant aid ir 4 a e u-

.- 1 1

atat ;,y .,\ ,\'.,u --

_t _ t ,y gt' , f_ '

R_ rM_ E_ 'r. _V._S R,ae4rch + d 1, i 8 -

e,1 ? ' -

Pu b li t. merv:ce - - -

Acadeetc suppcrt - -

student services ! ,T1 2' "4 19 '. --

I f.S i ll he t l U11a k S Op p b rf $ I' '

l $ ', 3 0' - 1 g 8 Operation and ma i nt eaa nc e .f ,. an' li g *.8 _ s t, t ' 4 -

Total

Ft* **^*^^ '" ' Y, h} : ~~

, g , _-2. , , _

MAJcp Dt P ARM NT N F N> p t.Y 1 A p: FA; r.ir .7 <%

4 J

> * =} + -_

,L 1 J.

3 J J ,..) .

%1 F t, ' S j ukJ 1 Ws %3* t,'18 It>I 41 4 4J r Te CC IUPd5 f Rp e nd l 1 u f r t. 4.I w 4 d4 m; t Ih ] M,,

w 7 N .- . .4 J_, .,

41Q

.J. .- _

A 2 1. . ...A J

L. .

4_

..g s.

ex , * %.

\ d [\ . .\~g $

c os s. u-\ a. ' d.;

ew4. . N

-, -Q'Q

(' 3$ ,\

{c&,s. . -n,)s.. (nt.'o' A (dJJ y'W

_6s\\ j% t x g(p AqN f6 1

L,5 A.% '.F L E S 197A-79 R! r w- ' (I g T I sg ry p 'yp ygg gr g ps gygg C hPl c Fi \M Rf r f"* $ A%'D F) F %n g 7 'at y g

1. bl' E 11' x 197n 79 O

TOTAL LRMENT FL NDS RE M I P*' S (45 ', o ? r '15 4 99,97 341 TOT AL C; R REN* P i:%DS Lu1\SITsues 77

-.1..L.M.1.a *. *?* L.1s 1

. .2 _1 *.

u!)1FEf%LE A . e 'sI c 1 a _? 1

_L, !I_s' Transfer of Camrus : uf rent Funds Reeetit_a_to Systeaw!de -

I n.11 re ct Cost Reishurse w r.t - 9t el (

11, ,4. e 14,621,151 De bt Re t i rew nt 3, r 7 5,0 7 1,6)M,7*)

F ducat ion F ee e , 374 e s2,8 V General Fan.is - incemdtted 1,'il2 ell 6?h 592 Other Fu n <. s -

S t ad e r.t l ean Fam4 97- a l' ,rR.

Plant Fundo I, u . ', l e ! 21?

Flant Princ f ral 6 Re ctse- <

- P >.1, e ' )

Fayme nt s ( 8 Prttui a l 6 I r.t e r < = r - 1st ( 41, $ e 3 Uneuter.ded Balances cf urrear to . > r ApprotrLetion la Soh ne ueat Years t

44) 3, 5 'n , 55 A g e nc y - BF O ' Pr ogr am 1, ' + ),2 ? s,, e9 Other 7l. ip : u

-.s. t s;s e _s E

--h1 .a. 2 i.- 1 J .,.r e.1_71 T ra ns f e r of %v e r eirne l de Tu-ds d aqu e f o r E rp e n d ! ' . e .

Eudww r;t Principal Appr.,priarci 99,19 w A,ee',;12 E ndw ee nt I n c oce Arp 3p t ! .a t ei ',ve,e: s;s,pcw Addit ional Allv(ati r of eneraa ri nds =,71' 2: e, up ,.'4 (

Ad di t t or.41 A lle.. 4 t i on <

F .b r a t t 2,41 See inc * -- ' ', 41 '

Short-Ters Irvestmer ' ,v s.* a9 3,..a,oas Indirect Cast P e t e% r sew at t. f t.- J Contract end Geart A d t.i n t a t r a t i cm i , 15, *J 1 I ,215, w )

Other yrq p%

< 1:1953 -1.--

[

- l ..- s.4 l 31---.

e 6.6 .._4 1 16%- 7 7l.

CIF FEl&M.E . 19 7st vi ir l.

' .1 = L'1. ' . .

y_

$ _i l'. _t .t_._

,..\g

,,- , in s'a t

y

\ %.

9 c. .y 9

O

APPENDIX II ENVIRONMENTAL IMPACT APPRAISAL FOR THE UNIVERSITY OF CALIFORNIA AT LOS ANGELES TRAINING REACTOR LICENSE NO. R-71 DOCKET N0. 50-142 This section deals with the environmental effects which can be attributed to the operation of the UCLA Argonaut research reactor since the issuance of amendment 10 to the UCLA facility operating license R-71 on February 5, 1976. It will also address potential future environmental effects.

February 1930

APPENDIX II

$ ENVIRONMENTAL IMPACT APPRAISAL CONTENTS Chapter 1 Facility, Environmental Effects of Construction. . . . . . . . . . . . . . . . . . II/l-1 Chapter 2 Environmental Effects of Facility Operation . 11/2-1 Chapter 3 Environmental Effects of Accidents. . . . . . . II/3-1 Chapter 4 Unavoidable Effects of Facility Construction and Operation . . . . . . . . . . . . . . . . . II/4-1 Chapter 5 Alternatives to Construction and Operation of th e Fa c i l i ty . . . . . . . . . . . . . . . . . . I I / 5-1 Chapter 6 Long-Term Effects of Facility Construction and Ope ra ti on . . . . . . . . . . . . . . . . . . . I I /6-1 Chapter 7 Costs and Benefits of Facility and Alternatives 11/7-1 Attachment A The Environmental (TLD) Program. . . . . . . I./A-1 II/i

APPENDIX II ENVIRONMENTAL IMPACT APPRAISAL LIST OF FIGURES PAGE Figure II/2-1 Film Badge Locations Nuclear Energy Laboratory - 1st Floor. . . . . II/2-2 Figure II/2-2 Film Badge Locations Nuclear Energy Laboratory - 2nd Floor. . . . 11/2-3 Figure II/2-3 Film Badge Locations Roof View Overlooking Stack. . . . . . . . . . II/2-4 Figure II/A-1 Roof View Overlooking Stack. . . . . . . . . . II/A-5 Figure II/A-2 Average Quarterly Dosimeter Readings . . . . . II/A-6 LIST OF TABLES Table II/2-1 History of Annual Releases . . . . . . . . . 11/2-5 4

II/ii

1.0 FACILITY, ENVIRONMENTAL EFFECTS OF CONSTRUCTION The plant housing the reactor is located in the northwest wing of Boelter Hall at the University of California, Los Angeles. The reactor is located at the Nuclear Energy Laboratory in a 2 story, reinforced concrete struc-ture with a floor area of approximately /5 x 49 ft. and a height of 27 feet.

Subsequent construction, completed in 1963, has surrounded the reactor room on the north, east, and south sides with additional laboratory space that provides a buffer zone between the reactor room and adjacent but un-related facilities. On the west side, first floor laboratory spaces and the second floor control room intervene between the reactor room and the exterior building wall.

The third floor (roof) of the reactor blilding is bridged by new construc-tion at the fifth, sixth, seventh, and eighth (roof) levels. The void region between the third and fifth floors is a limited access region which contains a small structure housing air conditioning and water deminerali-zation equipment.

There are no exterior conduits, pipelines, electrical or mechanical struc-tures or transmission lines attached to the nuclear reactor facility other than utility service facilities which are similar to those required in other campus facilities, especially laboratories. Heat dissipation is accomplished by a shell and tube heat exchanger contained within the pri-mary water dump tank. The secondary system utilizes Los Angeles city water which passes through the tube side of the heat exchanger to extract heat from the primary water. The secondary water effluent is monitored by a Nal detector and passes through a 225 gallon (10 minute delay) tank prior to rejection. The primary and secondary cooling systems are located in the process pit below the reactor floor level immediately north of the reactor. Radioactive gaseous effluents consist primarily of argon-41.

The exhaust is monitored. Both solid and liquid radioactive wastes are generated through the irradiation of samples to be used on campus either for neutron activation analysis or for radioisotopic tracer analysis.

Liquid wastes are released to the sanitary sewer in concentratior.s that do not exceed the limits specified in 10 CFR 20, Appendix B, Table I, column 2.

Solid wastes are packaged in DOT approved containers, and are transferred and shipped to the UCLA State Licensed radioactive waste holding area.

With other UCLA radioactive wastes, these dr *s are shipped via commercially licensed carrier for burial on approved site.

II/1-1

2.0 ENVIRONMENTAL EFFECTS OF FACILITY OPERATION The UCLA Researcn reactor has a maximum licensed power output of 100 KWt.

The environmental effects of thermal effluents of this order of magnitude are negligible. The waste heat is rejected to the storm drain through a double-pass tube and shell heat exchanger located in the process pit of the reactor building. The maximum secondary outlet temperature runs approxi-mately 300F above the city water supply temperature af ter the reactor has reached equilibrium at full power. This waste water flowing at 22 gallons per minute through the storm drain (an underground conduit) has negligible effect on the environn _nt.

The reactor room (high bay) at the UCLA Nuclear Energy Laboratory (NEL) is completely surrounded by an NEL radiation controlled area for which entry requires a special security pass, unique key, health physics quali-fication by examination and mandatory personnel dosimetry. The unrestricted area (uncontrolled) available to the University population and general public begins at the laboratory concrete wall outside perimeter. Measured levels of direct radiation (beta, gamma. and neutron) in this uncontrolled area are not detectable above background (s 0.04 0.03 mrem / hour) with calibrated GM monitors during full power reactor operation (100 kilowatts thermal).

The room in which the reactor is lacated is continuously monitored for gemma-ray fields. Two GM gamma detectors are mounted on the north and south balcony wall of the reactor room. The alarm set point of these moni-tors 5 mR / hr. The typical readings during full power reactor operations are approximately 1 mR/hr. A continuous particulate air monitoring system samples the facility exhaust stack (and intake duct) at the third floor (3000 level). The limit of sensitivity for the system is between 2 x 10-12 and 2 x 10-13 pCi/ml. The system for collection of particulates is more conservative than detailed in ANSI 13.1 - 1969 and the particulate filters are counted routinely on a batch basis.

The system includes an environmental upwind sample which parallels the exhaust sample and is used for background subtraction from the exhaust sample to eliminate natural radon and thoron daughters collected as a result of this type of sampling system. These filters do not vary by more than a factor of 3 from one another. Typical background readings range from 7.8 to 21.2 c/m, which corresponds to approximately 2.5 x 10-12 pCi /ml .

Film badges are located at various positions both inside and outside the reactor toom. The film badges are placed in the locations shown in Figures 11/2-1, 11/2-2, and 11/2-3. The highest annual readings (1976 to present) for each location are also shown in these figures.

The principal radioactive gaseous effluent monitored in the building exhaust stack is identified as argon-41. The actual concentration of this gas is determined by a 4.3 liter ion chamber which has been calibrated in microcuries per milliliter vs. ion current. This data is recorded on a strip chart recorder whenever the reactor is running and the data periodically integrated using a compensating polar planimeter to obtain the total curie discharge.

II/2-1

IDENTIFIER BADGE PU43ER HIGHEST MF4JAL DOSE TYPE OF FILM a

x 219 0 mrem -----

b x 2048 0 mrem -----

c x 1965 1200 mrem 3,y , x-ray * [*flote - this badge includes x-ray radiation from the d

x 230 0 nRem -----

Tokamak Laboratory]

x" 1581 0 mRen -----

f x 220 0 nRen ----

x9 191a O mRen -----

km x h

218 110 mrem 8,y i

h x 265 350 mrem S only x3 203 425 mrem S only k

x 820 0 mrem -----

x 302 0 mrem -----

LOCATI0tl AflD MEASUREf1EilT GUIDE TO FIGURES II/2-1, 11/2-2, and 11/2-3

h

'~~~'

~~ ^ = $ -_ .= ._. . . __ f '

N 10008 1 0 T

t W X , , . , 34 16 T 4 sq ,,

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l.

E F 1005 1000 IOOOA

_ xQ

= = _ = . . ,

REACTOR ROOM 3

] o iOO4 TOKAMAK 1003

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=>41 .

=t 1567 1561 15 41 I 1540 j 1549

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NUCLEAR ENERGY LABORATORY - 1ST FLOOR (GROUND LEVEL)

Wd FIGURE II/2-1 FILM BADGE LOCATIONS f'

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Table 1 shows a five year history of annual ArW releases and reflects the increased use factor of the reactor.

Table 11/2 History of Annual Releases

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Year Curies 1975 23 1976 33 1977 47 1978 58 1979 65.5 The two-year environmental (TLD) survey program, undertaken in accordance with Part 2.c(3) of the R-71 Facility License (as modified in connection with Amendment 10, effective 6 March 1976), was completed in June,1978.

The quarterly average results at 20 dosimeter locations failed to demon-strate a clear dependence of radiation level upon either range or direc-tion from the ventilation stack. Ranges extended to roof tops of buildings isolated and remote from the reactor stack, and directional effects had been expected because of the prevailing wind (dominantly from the south-west during the daylight hours). These results, an interpretation, and the reasons behind the interpretation are described in Attachnent A.

Briefly, eleven of the locations involved dosimeters mounted upon concrete walls or parapets . The time average values (usually 8 quarters) at these locations ranged from 8.4 to 10.9 mr/ quarter. Only one other dosimeter location fell within this range--a dosimeter located at the top of the stack, supported on a wire screen in the center of the air stream) indi-cated an average value of 10.8 mr/ quarter. No other dosimeter indicated an average value greater than 6.0 mr/ quarter. (See Attachment A "The Environmental (TLD) Program").

Eighteen environmental routine wipe tests are made weekly (as well as many non-routine) at the most probable contaminated areas both inside and out-side the controlled areas. No activity above background statistics (none

> twice background) has ever been detected outside of the controlled area.

The detection limit at 95% confidence (10 minute count times , 60 minute background times) is between 5 x 10-7 and 5 x 10-e pCi/cm 2 (10: efficiency for beta counting).

Another possible airborne effluent is tritium (3H). 3 11 is produced through the neutron capture by deuterium nuclei present in the cooling water.

The only way that the tank water is released to the environment is through the evaporation of the water. The reactor primary water has not been changed since January of 1977 and losses through evaporation have been nil since no makeup water has been added during this period of time (3 years).

Assay of the primary water in January of 1980 disclosed a tritium concen-tration of 0.00326 .0005 pCi/mi for a total inventory of 2.47 x 10-3 Ci in the 200 gallons of primary water. During this 3 year period, the reactor generated 65.3 x 103 kilowatt hours which averages to 3.8 x 10 8 Ci/kwt.

Therefore, under the assumption that the reactor was running at its full II/2-5

licensed power of 100 KW and that all of the tritium being made is escaping into the exhaust stack, the maximum stack concentration would be 1.6 x 10-10 pCi/cm3 . This is three orders of magnitude below MPC for soluble.

tritium as stated in appendix B table II Col.1 of 10 CFR 20.

Another area of concern is the generation of high and low-level radioactive wastes. The storage or reprocessing of spent fuel elements is not a major concern at the UCLA Research Reactor because our typical annual U235 burn-up is approximately .3% of our excess reactivity. During the course of activation ~ analysis experiments and isotope production runs, the Reactor Laboratory generates an average of 0.4 m 3 of low-level radioactive waste annually. The main constitutents of this waste are short-lived isotopes such as Na-24, Al-28, Cl-38, Mn-56, La-140, Eu-152, Eu-154, Dy-165, Au-193. These wastes are shipped to authorized disposal sites in approved containers.

11/2-6

3.0 ENVIRONMENTAL EFFECTS OF ACCIDENTS Accidents ranging from failure of experiments to the largest core damage and fission product release considered possible result in doses of only a small fraction of 10 CFR Part 100 guidelines and are considered negligible with respect to the environment. The UCLA Reactor has been subjected to experimental vibration. The results were reported by C. B. Smith at the Winter fleeting of the American fluclear Society, November,1968, in a paper titled " Vibration Testing and Earthquake Response of fluclear Reactors".

11/3-1

4.0 UNAVOIDABLE EFFECTS OF FACILITY CONSTRUCTION AND OPERATION The unavoidable effects of construction and operation involves the materials used in construction that cannot be recovered and the fissionable material used in the reactor. No adverse impact on the environment is expected from either of the unavoidable effects.

11/4-1

5.0 ALTERNATIVES TO CONSTRUCTION AND OPERATION OF THE FACILITY There are no suitable or more economical alternatives which can accomplish both the educational and the research objectives of this facility. These objectives include the training of students in the operation of nuclear reactors, the production of radioisotopes, its use as a source of neutrons for neutron activation analysis, and also its use as a demonstration tool to faniliarize the general public with nuclear reactor operations.

11/5-1

6.0 LONG-TEPJi EFFECTS OF FACILITY CONSTRUCTION AND OPERATION h The long-term effects of a research facility such as the UCLA Nuclear Energy Laboratory are considered to be beneficial as a result of the contribu-tion to scientific knowledge and training. This is especially true in view of the relatively low capital costs involved and the minimal impact on the environment associated with a facility such as the UCLA Nuclear Energy Laboratory.

11/6-1

7.0 COSTS Afl0 BEtiEFITS OF FACILITY AftD ALTERf1ATIVES The cost for a facility such as the UCLA fluclear Energy Laboratory is on the order of $1 million with very little environmental impact. The benefits include, but are not limited to:

(a) education of students and public-(b) research (activation analysis and production of short-lived isotopes); and (c) training.

Some of these activities could be conducted using particle accelerators or radioactive sources, but these alternatives are at once more costly and less efficient. There is no reasonable alternative to a nuclear research reactor of the type presently used at the fiuclear Energy Laboratory for conducting the broad spectrun of activities indicated above.

11/7-1

APPEf1 DIX II EtiVIRONMEllTAL IMPACT APPRAISAL Attachment A The Environmental (TLD) Program II/A

THE ENVIRONMENTAL (TLD) PROGRAM ,

In the spring of 1975, questions were raised by the Nuclear Regulatory Commission regarding the argon-41 content of the ventilation air exhausted from the UCLA reactor building. These questions ultimately related to the radiation levels associated with the exhaust air, and the possible exposure of individuals to radiation emitted by the decay of the argon-41.

The reactor room is ventilated by once-through air and the argon-41 arises from the slow expulsion and/or migration of air from the reactor core into the ventilation system (whose flow rate is now approximately 14,000 CFM).

It should be under:tood that the presence of argon-41 in the ventilation air is a consequence of normal reactor operations and not a result of any unique, peculiar, or abnormal occurrence.

Prior to 1975, the reported argon-41 releases were apparently based upon a stack dilution factor that had been neither discussed with the commission, nor incorporated in the 1971 Technical Specification of the R-71 license.

To correct this matter, UCLA applied for an amendment to the Technical Specifications. The amendment, known as An'endment 10, was granted February 5, 1976, and included, among other things, a concentration reduction factor based upon considerations of stack dilution, occupancy, and annual averaging fa ctors . Attendant to the granting of the amendment, UCLA agreed to under-take an environmental survey program of two years duration, to measure the radiation levels occasioned by the ventilation exhaust air.

To implement this agreement, UCLA procured the services of the Radiation Detection Company of Sunnyvale, California, to provide and read Thermo-luminescent Dosimeters (TLD's). Twenty-two pairs of dosimeters were supplied and changed quarterly commencing March 10, 1976 and continuing until the last set was collected in March of 1978.

In granting Amendment 10, the Nuclear Regulatory Commission recogni::ed a calculated stack dilution factor based upon the wake cavity mixing model described in Appendix D of Draft Regulatory Guide, Docket No. RM-50-2, February 20, 1974. Although of uncertain applicability to the complexity of the UCLA roof-top geometry; it was believed to be appropriately con-servative in view of the many predictive uncertainties.

In the summer of 1976, Mark Rubin, a graduate student under the direction of Professor W.E. Kastenberg, undertook an experimental study of the behavior of the reactor ventilation system plume. The observations, made under the prevailing southwesterly wind condition, used sulfur hexafluoride (SFc) as a tracer which was injected at a known rate into the ventilation system at the reactor room. Air samples, taken downwind of the stack, were analyzed chromatographically to obtain measured dispersion. Using the occupancy and reactor utilization factors of Amendment 10 the results indicated that public exposure (concentrations) would range from 1% to 10% of those esti-mated from the wake cavity mixing model.

II/A-1

The reduced exposures (relative to those calculated by the wake cavity mixing model) were attributed to a combination of augmented turbulence and plume rise. That is,the model uses an atmospheric stability parameter that depends on solar insolation and wind speed. Turbulence introduced by varying roof elevations, parapets, roof structures, and vented court yards, is not readily introduced in the model idealization. The high ejec-tion velocity (40 ft/sec) of the exhaust is not recognized in the model but can be . visualized and photographed with the aid of smoke flares. These matters are also discussed by Rubin

and some examples of plume photography are available at UCLA.

The foregoing provides the background for a discussion of the TLD results.

In describing the exhaust, the plume is an observable physical reality and, in many respects, is a more appropriate concept than a mixed wake.

Further, the TLD results are anomalously high, and the work of Rubin tends to support the interpretation to be made here.

To describe the placement of Thermoluminescent Dosimeters, the roof areas in the vicinity of the release point (stack) are shown schematically in Figure II/A-1. The stack rises above a wind screen that protects an unrelated cooling tower.

The roof area, bounded by the walls surrounding the wind screen, was defined as Region 1 in the Safety Analysis that accompanied the Amendment 10 request.

The maximum occupancy factor for that region was taken as 0.05. The roof areas to the north and south of Region 1 were defined as Region 2; the occupancy factor of those areas was specified as 0.10 (maximum). The rationale behind these distinctions was given in the Safety Analysis and will not be reviewed here.

Figure II/A-2 presents the average quarterly dose at each of the 20 dosimeter locations. By average, we mean the sum of the quarterly readings (of the two year progran) divided by the number of quarterly readings. In general, eight quarters were observed, but a few dosimeters were lost to birds and possibly to curious individuals.

The perimeter of Region 1 is indicated by the dashed line countour on Figure II/A-2.

Dosimeters were placed so as to be unobtrusive or not easily accessible.

For example, dosimeters were placed along the south surface of the north wall by leaning down over the ninth floor parapet. These dosimeters had an unobstructed view of the plume for all wind directions ranging from west, through north, to east. Dosimeters were placed on the west surface of the west wall by leaning over the eighth floor parapet. The parapet cap, approximately 18 inches thick, shielded the dosimeters from a direct view of the stack. These west wall dosimeters could directly view a plume only under rare easterly wind conditions. The statistical frequency of winds from the eastern quadrant is approximately 13% during the hours 7:30 a.m. to 5:30 p.m.

Rubin, Mark, " Atmospheric Dispersion of Argon-41 from the U.C.L. A. Nuclear Reactor," Masters Thesis, 1976.

II/A-2

The foregoing example illustrates the variability of geometric placement (orientation and wind direction) that might be expected to influence the dosimeter readings. For the dosimeters of the example, it was expected that the west-facing dosimeters would show lower irradiation levels than the south-facing dosimeters. Figure II/A-2 indicates a converse conclusion.

In fact, there is virtually no discernible pattern (distance or direction) to the raw dosimeter data.

A marked effect is apparent in relating dosimeters to the type of strac-ture upon which they were mounted. Eleven dosimeters were mounted upon massive concrete parapets or structures. For these dosimeters, the readings ranged from a minimum of 8.4 to a maximum of 10.9 mR per quarter. The average value was 9.8 mR/ quarter. Nine dosimeters were mounted on non-concrete structures (sheet metal vents , rain gutters, wood); of these nine, one was centered in the air stream at the top of the stack and indicated 10.8 mR/ quarter. Another, located on the roof of Pauley Pavilion, was regarded as a control because of its considerable distance and generally upwind direction from the reactor stack. The averaged value at this loca-tion was 0.2 mR/ quarter. Values for the remaining seven ranged from 3.3 to 6.0 mR/ quarter with a mean of 4.7 mR/ quarter.

The dosimeters mou1ted on concrete structures are indicated by shading in Figure II/A-2. Although some bias might be introduced if concrete-mounted dosimeters were preferentially located closer to the stack than other dosimeters, the figure indicates that values seen by remote dosimeters, so mounted (#18,19, and 16) do not differ appreciably from other concrete-mounted dosimeters.

It is difficult to escape the conclusion that concrete is a source of measur-able background radiation. The remote SW point (#16) is located on a parking structure, at a distance of approximately 610' from the reactor stack.

Winds from the northeast are expected only about 12% of the daylight hours.

The remote NE point (#18) is on the physics building, at a distance of approximately 780' from the stack. This is a downwind location under pre-vailing wind conditions (71% of the daylight hours). Initially chosen as a sample point in the Rubin work, it was abandoned because of the high dispersion at that range.

It should also be recognized that dosimeters mounted on non-concrete struc-tures are not necessarily remote from concrete and hence not necessarily irmiune to radiation from nearby concrete. Thus, even if the readings of concrete-mounted dosimeters are rejected, the remaining data are not free of ambiguous interpretation.

For the purpose here, and for the text summary, we shall reject all data from concrete-mounted dosimeters. For the average exposure rate in Region 1, we shall take the average value of the 6 dosimeters lying within 100 feet of the stack including the one on the stack. This rate, as seen by the dosimeters, is 6.0 mR/ quarter. Adjusting to a yearly basis, and incor-porating a 5% occupancy factor yields 1.2 mR/ year.

For the rate in Region 2, we shall use the same group but delete the stack dosimeter and add the dosimeter #15 (the meteorological station 172' ENE II/A-3

of the stack). The average rate, 4.77 mR/ quarter, adjusted to an annual rate and correcting to 10% occupancy is 1.9 m?/ year.

From the previous two paragraphs, the interpreted dosimeter data indicates that Region 2 conditions are limiting and that reactor operations lead to an expected public exposure of not more than 1.9 mR/ year.

II/A-4

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APPENDIX III ARGONAUT SAFETY ANALYSIS REPORT (ASAR)

FOR THE UNIVERSITY OF CALIFORNIA AT LOS ANGELES.

TRAINING REACTOR LICENSE NO. R-71 DOCKET NO. 50-142 February 1980

APPENDIX III ARGONAUT SAFETY ANALYSIS REPORT (ASAR)

CONTENTS Chapter 1 Introduction and General Description. . . . . . III/1-1 Chapter 2 Supporting Facilities . . . . . . . . . . . . . III/2-1 2.1 Subcritical Assembly . . . . . . . . . . . III/2-1 2.2 Neutron Generator. . . . . . . . . . . . . III/2-1 2.3 Activation Analysis Laboratory . . . . . . III/2-1 2.4 Heat Transfer Laboratory . . . . . . . . . III/2-1 2.5 To k ama k La bo ra to ry . . . . . . . . . . . . I I I /2 - 1 2.6 Auxiliary Laboratory Areas . . . . . . . . III/2-1 Chapter 3 General Plant Description . . . . . . . . . . . III/3-1 3.1 Si te Loca tion. . . . . . . . . . . . . . . III/3-1 3.2 Si te Geology . . . . . . . . . . . . . . . III/3-1 3.3 Site Hydrology . . . . . . . . . . . . . . III/3-1 3.4 Site Seismology. . . . . . . . . . . . . . III/3-1 3.5 Site Meteorology . . . . . . . . . . . . . III/3-2 3.6 Population Distribution. . . . . . . . . . III/3-3 Chapter 4 Plant Arrangement . . . . . . . . . . . . . . . III/4-1 Chapter 5 Nuclea r Reactor . . . . . . . . . . . . . . . . III/5-1 5.1 Rea c to r Co re . . . . . . . . . . . . . . . I I I / 5 5 5.2 Reactor Cooling System . . . . . . . . . . III/5 8 5.3 Reactor Instrumentation and Control. . . . III/512 5.4 Reactor Fuel Handling. . . . . . . . . . . 171/5-15 5.5 Reactor Waste Disposal Control . . . . . . III/5-15 Chapter 6 Comparison Tables . . . . . . . . . . . . . . . III/6-1 6.1 Comparisons with Similar Facility Designs. III/5-1 6.2 Comparison of Final and Preliminary Design . . . . . . . . . . . . . . . . . . III/6-1 Chapter 7 Identification of Agents and Contractors. . . . III/7-1 Chapter 8 Design Bases, Accidents, and Consequences . . . III/8-1 Chapter 9 Requirements for Further Technical Information . . . . . . . . . . . . . . . . . . III/9-1 Chapter 10 References ............ . . . . . III/10-1 Attachment A- Estimation of Effects of Assumed Large Reactivity Additions . . . . . . . . . . . . III/A Attachment B Radiation Doses Resulting From Release of Fission Products into Atmosphere. . . . . . . III/B III/i

APPENDIX III ARGONAUT SAFETY ANALYSIS REPORT (ASAR)

LIST OF FIGURES PAGE Figure III/1-1 Area Map - West Los Angeles. . . . . . . III/1-2 Figure III/4-1 Aerial View of Central Campus. . . . . . III/4-2 Figure III/4-2 UCLA Campus Map. . . . . . . . . . . . . III/4-3 Figure III/4-3 Nuclear Energy Laboratory - 1st Floor. III/4-4 Figure UI/4-4 Nuclear Energy Laboratory - 2nd Floor. . III/4-b Figure 111/4-5 Reactor Building - Elevation Sections. . III/4-6 Fige e III/5-1 Reactor - Longitudinal Section . . . . . III/5-2 Fi,.re III/5-2 Reactor - Tranverse Section Through Core Center. . . . . . . . . . . . . . . III/5-3 Figure III/5-3 Reactor - Horizontal Section at Beam Tube Level . . . . . . . . . . . . . . . III/5-4 Figure III/5-4 Fuel Plate . . . . . . . . . . . . . . . III/5-6 Figure III/5-5 Typical Fuel Cluster . . . . . . . . . . III/5-7 Figure III/5-6 Fuel Boxes and Coolant Connections . . . III/5-9 Figure III/5-7 Cooling Systems Piping Diagram . . . . 1I1/5-11 LIST OF TABLES Table III/1-1 Chronology . . . . . . . . . . . . . . . III/1-4 Table III/1-2 Reactor Annual Use . . . . . . . . . . . III/1-5 Table III/1-3 Reactor Activity . . . . . . . . . . . . III/1-5 Table III/3-1 Habitable Area & Population Within R Miles of the Reactor Site. . . . . . . . III/3-4 Table III/5-1 General Cooling System Characteristics . III/5-10 Table III/6-l@ Comparison Table - General . . . . . . . III/6-2 Table III/6-1(b) Primary Coolant. Nuclear Data. . . . . . III/6-3 Table III/6-1(c) Reactor Characteristics. . . . . . . . . III/6-4 Table III/6-2 Training Reactor Characteristics . . . . III/6-5 Table III/7-1 Identification of Agents . . . . . . . . III/7-2 Table III/7-2 Organization Relations . . . . . . . . . III/7-3 III/ii

ARG0 NAUT SAFETY A_NALYSIS R_EPC.U (ASAR)

1.0 INTRODUCTION

AND GENERAL DESCRIPTION The Argonaut Safety Analysis Report has been prepared for submission to the U.S. Nuclear Regulatory Commission in support of reapplication for an Operating License. The application is made by the Regents of the University of California for the continued operation of the reactor, licensed as R-71, at the Los Angeles Campus.

The plant housing the reactor is located in the northwest wing of Boel ter Hall at the University of California, Los Angeles. The 400-acre campus is located on a coastal plain and is approximately five miles east of the Pacific Ocean and fifteen miles west of the Los Angeles civic center. To the south of the campus is a business and shopping district, and to the north, west and east are residential areas. A map of the general area is shown in figure III/1-1.

The reactor is located at the Nuclear Energy Laboratory in a 2 story, reinforced concrete structure with a floor area of approximately 75 x 90 ft.

and a height of 27 feet. The construction of the reactor facility began in 1959, with the assistance of a $203,350 grant from the US Atomic Energy Commision, through the efforts of the founding Director, Dr. Thomas E. Hicks. This grant was disbursed in construction and reactor equipment.

Subsequent construction, completed in 1968, has surrounded the reactor room on the north, east, and south sides with additional laboratory space that provides a buffer zone between the reactor room and adjacent but unrelated facilities. On the west side, first floor lat' oratory spaces and the second floor control room intervene between the reactor room and the exterior building wall .

The third floor (roof) of the reactor building is bridged by new con-struction at the fif th, sixth, seventh, and eighth (roof) levels.

The void region between the third and fifth floors is a limited access regior which contains a small structure housing air conditioning and water Jemineralization equipment.

The nuclear reactor is an Argonaut type; water-coole ' and moderated, graphite reflected, 93% enriched uranium thermal reactor, that is currently licensed for a maximum core thermal power of 100 kw. By special amendment, the reactor has operated in the past for brief periods of up to 500 kw. It appears that the reactor could safely operate up to 1,000 kw with modifications to the shielding, the cooling system, and special provisions for reducing argon-41 emission.

Historically, the UCLA reactor eached criticality on October 21, 1960, a t 6 : 54 p.m. After a program o r low-power testing at 10 watts, the reactor went to its then licenad power of 10 kw in February of 1961.

The reactor was modified slightly, license amendments were approved, and in October of 1963, the reactor reached its present licensed full thermal power output of 100 kw. The chronology of these and other III/1-1

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events is shown in Table III/1-1.

The reactor generates no electricity and is used primarily for activa-tion analysis, class instruction, student experiments and faculty, staff, and student research. To provide this flexibility, the reactor has three vertical irradiation holes (1.9" ID), a 78 cubic foot removable graphite thermal column with a one cubic foot irradiation volume, two 6" ID and four 4" ID horizontal beam ports, and a 3,000 gallon shield tank. A pneumatic transfer system (" rabbit") provides sample irradia-tion in the west vertical port with rapid transfer to a counting laboratory.

The variety of irradiation ports has provided great flexibility in the kinds of experiments that can be conducted with the reactor. The fast and thermal flux is maximized in the vertical ports, the thermal to fast flux ratio is maxinized in the thermal column and a neutron and/or y beam may be extracted from the horizontal beam ports. Tabl e III/1-2 gives a brief description of the annual reactor use from 1973 to 1979.

Variations from year to year are attributed to research demand, changes in technology, random maintenance requirements, class scheduling, and enrollments .

In recent years, it has become convenient to describe reactor usage in terms of research, class instruction, and maintenance. This dis-tribution of reactor hours into these categories is shown in Table III/1-1 The research category can include individual research projects as well as miscellaneous service irradiations, but the greatest number of research hours are for activation analysis (both y-ray spectro-scopy and delayed neutron counting), and fission track dating projects.

Graduate students from other departments and schools are usually par-ticipants in these projects. Class instruction generally means under-graduate instruction, but it also includes health physics and reactor operator training for undergraduates and graduate students. Mainte-nance includes only those hours for which the reactor is operated to perform maintenance tests or calibration work required by the Technical Specifications.

III/1-3

Table III/1-1 CHRON0 LOGY (actual month and year if events have occurred; estimated month and year if events have not occurred)

a. Start of design work 2/59

b. Completion of preliminary desigi 6/59

c. First contract signed for purc'iase of major component component 4/59

d. Start of construction at site 10/59

e. Delivery of core vessel to site 11/59

f. First containment shell testing at design pressure None 9 Construction essentially completed 1/60

h. First criticality 10/60

1. First operation at reactor's design power 2/61

j. Plant ready for planced operation 3/61 k 100kw operation 3/63

1. First shipment of irradiated fuel for chemical processing 6/80 III/1-4

Table III/1-2 REACTOR ANNUAL USE Year Number of Runs Megaviatt Hours 1973 76 13.8 1974 76 14.8 1975 91 11.9 1976 82 13.1 1977 106 15.9 1978 132 20.3 1979 149 29.0 Table 111/1-3 REACTOR ACTIVITY Activity Hours per Year 1973 1974 1975 1976 1977 1978 1979 Research 145 177 146 158 188 244 411 Class Instruction 46 28 39 27 88 60 34 Maintenance 12 52 31 23 14 36 1 TOTAL 203 257 216 208 290 340 446 III/1-5

2.0 SUPPORTING FACILITIES

2.1 SUBCRITICAL ASSEMBLY The subcritical facility is used to verify nuclear reactor core physics for Engineering 135AL and 135BL on fuel / moderator systems which are much smaller than critical reactor assemblies. The facility contains two natural uranium subtritical assemblies. One is a 5 foot by 5 foot by 6 foot rectangular graphite assembly resembling early US plutonium production reactors; the other is a 5 foot diameter, 5 foot high cylin-drical natural uranium / heavy water assembly resembling the Canadian Deuterium (CANDU) power reactors.

2.2 NEUTRON GENERATOR A neutron generator located between the two subcritical assemblies is used to " power" the two subcritical assemblies or to provide high energy (14 Mev) neutrons for activation analysis. The neutron gene-rator operates at a steady state of up to 2 x 1011 neutro ns/ s eco nd ,

provides a modulated neutron beam in various wave shapes (sinusoidal, sawtooth, etc.) and can also be pulsed.

2.3 ACTIVATION ANALYSIS LABORATORY This 13boratory contains several Nal scintillation detectors, two Ge-Li crystal systems and at least two multichannel analyzers with computer linkage.

2.4 HEAT TRANSFER LABORATORY This laboratory, comprised of approximately 1,000 square feet, houses several experimental-hybrid devices for use in heat transfer studies for both light water reactors (LWR's) and liquid metal fast breeder reactors (LMFBR's).

2.5 TOKAMAK LABORATORY The Tokamak laboratory is housed directly east of the reactor high bay. The Tokamak laboratory under the direction of Dr. R.J. Taylor is funded by the Department of Energy. Two Tokamaks, the Microtor and the Macrotor provide the principal tools for research in the areas of plasma physics, first wall impurity studies, rf heating studies, and scaling laws.

2.6 AUXILI ARY l ABORATORY AREAS Included as supporting facilities are a fuliy equipped machine shop, an electronics shop, a small chemical laboratog, a photographics laboratory and an undergraduate laboratory. These are used to ful-full most reactor maintenance requirements, to fabricate experimental

Facil i ties housed at the NEL, aside from the reactor, that are licensed by the State of California in some cases. None are federally licensed.

III/2-1

equipment, to develop prints and movies for research papers and to support laboratory needs for nuclear related engineering courses.

111/2-2

3.0 GENERAL PLANT DESCRIPTION 3.1 SITE LOCATION The reactor is located on the 400 acre campus of the University of California at Los Angeles. It is housed within the Nuclear Energy Laboratory (NEL) in a specifically designed and constructed reinforced concrete building.

3.2 SITE GE0 LOGY The UCLA campus is situated on a coastal plain, and is approximately 400 feet above sea level . The coastal plain consists of a terraced alluvial fill, 200 feet deep at the reactor site, overlying sedimen-tary rock of rather recent origin. The coastal plain lies at the base of the Santa Monica Mountains which are approximately 2000 feet high. The most important formation in these mountains is Santa Monica slate, an old sedimentary layer 2000 feet thick. Overlying this slate stratum are several more recent sedimentary layers.

3.3 SITE HYDROLOGY No deep wells have been drilled on the campus of UCLA or in the vicinity of the campus. The water table is estimated to lie 200 feet below the surface of this area.

Surface runoff water is collected in concrete-lined storm drains which empty into the ocean. This drainage sys .em has been adequate to pre-vent any flooding of the campus by heavy winter rains. The maximum rainfall in any 24-hour period during the last 75 years was ten inches.

It is barely conceivable that runoff from the watershed area north of the campus could flood Westwood Boulevard and the area to the west of the reactor site. However, the reactor core lies about ten feet above this level, and a rair. fall equal to the largest ever recorded would not flood the reactor.

3.4 SITE SEISf0 LOGY Southern Ca'ifornia is seismically active. The nearest major faul t to the reactor site is the Inglewood fault running in a north-westerly direction about two miles east of the campus. In Southern California ,

the region from the Mohave Desert to beyond the off-shore islands is traversed by a series of active faults. These faults extend from 20 to 50 to many hundreds of miles in length, and the trend is gene-rally between north and west. However, they are only roughly paral-lel, and in certain instances a major fault zone is divided into two or more well defined faults. In general, these faults are from five to twenty miles apart and apparently extend to depths of 15 or more miles below the surface.

Earthquakes have occurred in California for a long time in the geo-logic past, and it is extremely probable that they will recur from time-to-time in the future. In the southern coastal section, shocks III/3-1

of large magnitude were recorded in 1769, in 1812, and in 1857,1933, 1952, and 1971.

The Uniform Building Code, representing the accumulated wisdom of the engineering profession in this field, takes specific account of the earthquake hazard. Virtually none of the structures built according to the specifications of this Code have suffered any damage from earth-quakes. In particular, neither the Engineering Building nor other campus structures suffered any structural damage in the severe Tehachapi earthquake of 1952 or the San Fernando earthquake of 1971.

The reactor building, which is structurally independent of any other building, conforms to the specifications of the Uniform Building Code.

Since the reactor lattice spacing is optimized for minimum critical mass, any structural cearrangements which might result from a severe earth shock would reduce reactivity. Loss of water due to a breach of the primary coolant loop will scram the reactor due to loss of moderation. Water seepage, if any, into the graphite will reduce reflector savings. In addition, microswitches are located between the shield blocks and connected to the reactor instrumentation and control system and will scram the reactor in the event of a large earth tremor.

3.5 SITE METEOROLOGY The Los Angeles Basin has two distinct seasons: a summer season which starts in April and may extent into October, and a winter season which lasts from November through March.

Winds in the Los I.ageles Basia are influenced by two factors. The first factor is the local alternation between a sea breeze blowing toward the land and up in the mountain valleys during the day, and a land breeze or down-valley wind during the night. The seco'id fac-tor is the effect of prevailing westerly winds.

In the summer, the Los Angeles Basin is on the fringe of the belt of prevailing westerlies and receives light winds. The pressure gradient responsible for these winds tends to reinforce the sea breeze each day and to oppose the land breeze at night, causing a net trans-port of air from west to east. Nevertheless, this flow is reversed many times during the summer, and the reversed flow tends to contri-bute to the accumulation of impurities (smog).

In the winter, the belt of westerlies shifts southward and brings the Los Angeles Basin under strong west winds nost of the time. Tem-perature inversions are usually absent or quite high under these con-ditions. Any impurities are rapidly mixed through a deep layer of atmosphere, and advection carries the impurities away in a relatively short time. This general winter condition is frequently interrupted when a high pressure center develops over North America and particu-larly over the Great Basin. The prevailing westerlies over the Los Angeles Basin may then be displaced by northeast winds, and any III/3-2

temperature inversion sinks to the surface and disappears, and impurities are carried seaward at the surface. These two conditions, either the prevailing westerlies or the east winds from a continental high, characterize most of the Southern California winter. It sheuld be noted that under both conditions deep vertical mixing is possible because convection and turbulence are not limited by an inversion.

This, in combination with stronger winds, greatly reduces the inci-dence of air pollution in winter.

3.6 POPULATION DISTRIBUTION The UCLA campus is located in Westwood, Los Angeles (see FigureIII/1-1) nr the foot of the Santa Monica mountain range. The mountains to the north are diffusely populated, whereas commercial and multiple dwelling regions predominate to the south. The creas to the east and west are predominately single family, residential, but with localized areas of multiple family dwellings, small shopping centers, and even light industry.

The general picture is that of a developed foothill community, geographically limited on the north by the mountains, to the southwest by the Ocean, and having a population density distribution that is strongly azimuth-dependent, and within which marked radial fluctuations exist.

The 1970 census reported a Los Angeles (city) population of about 2.81 million, a growth of approximately 13% over the prior decade.

Assuming a continued 13% growth, the 1980 population is approximately 3.2 million distributed over an area of 452 square miles. Because of the Santa Monica Mountains, the Hollywood Hills, and other areas of low population density, it will be assumed here that the bulk of the population is confined to an area of about 300 square miles at an average population density of 10,000 per square mile.

In the following, three types of areas are distinguished:

Normal Population Density Area 10,000 per square mile Low Population Density Area 2,500 per square mile Zero Population Areas (Ocean) --

The (mountainous) low density area is bounded on the south by Sunset Boulevard, unbounded (in the area treated) on the west, bounded on the north by Muhlholland Drive, and on the east by the circle of 5 mile radius centered at UCLA. The eastern bounoary is arbitrary, but the population density does grow hetween Coldwater Canyon and the Hollywood Freeway, Highway 101.

Based upon these assumptions, the population within a specified distance from the reactor site is shown in Table III/3-1.

III/3-3

Table III/3-1 Habitable Area and Population Within R Miles of the Reactor Site Area (Square Miles) By Density Type Total R (Miles) flormal Low Population 2.5 10.3 9.3 126,250 5.0 41.4 35.9 503,750 7.5 110.0 50.9 1,227,250 10.0 199.9 68.6 2,170,500 III/3-4

4.0 PLANT ARRANGEMENT The location of the building and its relationship to the campus is indicated aerially and schematically in Figures III/4-1 and III/4-2, respectively.

The Reactor Facility lies within, and is surrounded by areas used and controlled, by the Nuclear Energy Laboratory security and radia-tion protection requirements. The Reactor Facility is essentially the Reactor Building as constructed in 1959, whereas the surrounding spaces are of that or later date.

The Reactor Facility, shown in plan in Figures III/4-3 and III/4-4 and in section in Figure III/4-5 is a 2 story, reinforced structure approximately 75 x 49 feet in plan and about 27 feet high. The west face of the building opens onto an access court and faces and connects with Engineering Unit I via a structural steel bridge at the second floor level. The reactor room is surrounded on the north, east, and south sides by space controlled by the Nuclear Energy Laboratory and shared with the Tokamak Fusion Laboratory.

There is a clear space of two stories above the reactor room, and three stories of building above the clear space. The clear space, open on the east and west, provides a cross ventilation duct (approxi-mately 60 feet wide by 26 feet high) to a patio area at the same level, and east of, the reactor room roof. The immediate reactor room roof area is controlled by a security chain-link fence, and supports a small structure that houses air conditioning and water demineraliza-tion equipment (reactor related).

Major columns for the Reactor Facility rest on 10 x 10 foot poured, reinforced concrete footings. All exterior walls are load bearing and are of 12-inch thick, reinforced concrete, faced with Roman brick where exposed. The interior wall separating the reactor high bay from the remainder of the building is also load bearing and is of reinforced concrete,18 inches thick. All other interior walls are nonbearing, 8 -inch, cement block, curtain walls.

The covered walkways on the second and third floor levels on the west face of the building and the equipment room on the building roof are framed with structural steel beams and girders. Walls for these areas are formed of aluminum louvers. The building roof is a poured con-crete slab, 6 inches thick and supported over its span by five rein-forced concrete beams approximately eight feet on centers. The roofs of both the equipment room and the covered walkway on the third floor level are of corrugated steel panels covered with black mastic and crushed rock.

The reactor nigh bay is a room which is 43 x 49 feet in plan and is two stories (27 feet) high. A balcony skirts this area about its perimeter at the second floor level. A ten-ton traveling bridge crane serves the high bay and can reach any point in the room with the excep-tion of the area under the overhang of the balcony. The high bay floor is an integral part of the building ground floor slab which III/4-1

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is of 3000 psi, reinforced concrete poured on a crushed rock bed resting on undisturbed grade.

A process equipment pit lies directly north of the reactor. This pit is 9 x 12 feet in plan, exclusive of its western partial exten-tion which contains two horizontally mounted retention tanks. The pit proper contains the reactor dump tank (acting also as the pri-mary heat exchanger), the primary circulating pump, the primary de-ionizer, and associated valves and plumbing. Four floor drains in the high bay area drain to a sump in the southeast corner of the pro-cess pit. A drain beneath the reactor core also feeds this sump.

Waste fluids collected in this sump are pumped to one of two 225 gallon retention tanks in the west end of the process pit. The waste is monitored for radioactivity and is then dumped to the city sewer system in compliance with 10 CFR 20. The entire pit is lined with a 1/4-inch coating of Liquid Tile. Four interlocking, steel framed, concrete covers, each one-foot thick, cover the pit area. These covers lie flush with the working floor when in place.

Directly east of the process pit is a 5 x 6 matrix of 30 galvanized steel-lined fuel storage pits buried in concrete. Each pit is 78 inches deep and is stepped once, 30.5 inches from the pit bottom. Below the step, the pit inside diameter is eight inches. Above the step, the pit inside diameter expands to 10.25 inches. The step acts as a barrier to streaming radiation and as a support for the pit plug.

Each pit plug is approximately four feet deep and is made of noured concrete.

The reactor high bay area is served by a separate ventilating system (approximately 8000 cfm). A negative pressure is maintained in this area, relative to its surroundings. Thus, air flows , due to leak-age around door seals, etc. , from the remainder of the building into the high bay. The reactor is interlocked so that it cannot be started unless the high bay fans are m.

The air supply system (fan and conditioning equipment) for the reactor room, is located in the small equipment room above the reactor. The exhaust fan is located on the eighth floor (roof) above the reactor.

A high level GM radiation stack and area monitor will automatically shut down the air supply to the reactor room, and dampers in the intake and exhaust system will isolate the structure if the radiation level exceeds values given in the Technical Specifications.

Exhaust air from the high bay is monitored for argon-41 by an air flow-through ionization chamber. This detector reads out in the control room on a chart recorder and will sound an alarm if airborne activity in the exhaust exceeds permissible levels. This alarm will initiate an orderly shut-down of the reactor according to standard operating procedures. At this time the high bay exhaust and supply fans will ae shut off and motor driven dampers will close the supply and exhaust ducts until the cause of the high atmospheric activity has been located and corrected.

III/4-7

Area radiation monitors within the reactor high bay indicate radiation levels, and if excessive levels arise, the room can be evacuated by operator action.

III/4-8

5.0 NUCLEAR REACTOR A reactor which is used for student instruction must be designed so that safety is insured without exercising greater restraint on the activities of students than is normally advisable in a university labora tory. This necessitates:

A. that the total available excess reactivity Le limited B. that the reactor have a high degree of demonstrated inherent safety C. that it be limited to low-power operation.

These requirements are met in the UCLA Argonaut reactor by combining a water-moderated, plate-type fuel section with a graphite system for maintaining a fixed geometrical arrangement.

There is no credible way in which the fission products of this reactor can be made to escape, and the amount of contained fission products will be relatively small since it is currently limited to a maximum steady state thermal power of 100 kilowatts. Nevertheless, because of the reactor location on the campus, it is housed in a biological shield with a minimum number of penetrations.

This reactor is of the same general type as the Argonaut Reactor and is similar to the University of Florida Training Reactor and the University of Washington Training Reactor. The basic element of the reactor is a rectangular prism constructed of graphite bars. The fissionable material is introduced into the graphite prism in the form of aluminum-uranium alloy plates in six aluminum boxes, each of which contains a small amount of water to ensure an undermoderated system. The object of this construction is to have the convenience of a solid moderator-reflector and the safety of the water plate arrange-ment with a relatively large negative temperature coefficient.

The fuel is contained in MTR type fuel plates assembled in bundles.

These fuel bundles are contained in six aluminum boxes set in a two-slab array in a 5-ft prism of graphite bars. The control rods are the swinging-arm type similar to those used on CP-3 and CP-5, and the Florida and Washington reactors. Four cadmium vanes protected by magnesium shrouds operate within the spaces between the fuel boxes.

These are moved in or out by mechanical drives and may be disconnected by means of electromagnetic clutches to allow the blades to fall by gravity into the reactor. The drives, located outside the reactor shield for accessibility, are connected to the blades by means of steel shafts. The arrangement of the reactor parts is shown in figures 111/5-1, III/5-2 and 111/5-3.

The prism of graphite is surrounded by a biological shield of both conventional and heavy (magnetite) aggregate concretes which has been designed with an adequate factor of safety against seismic forces for a Zone 3 earthquake area. Access to the ends and top of the III/5-1

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.c::: m , ~

- ,. s c-

-hD,7 -$+ NEENs L7

[ '

' 9 'h - . M,

), o. :x re ::u

- .s u ica

.4 W6% $

IO':"y[{jr S h , g '[,m twi.i. :..,it ,2

' 9p?j+1:n-G, g.pyg y'5,w '

.j l}

"C.:?ff- ((M ;*

li U33 3.TI ) U '? 5* l#?'4 ! # ! ' I

d~! , iWitt. ; ..vs EE G AE.T gen e!'E STE!s3[:

s>XI : a ;ovi.t Ul% I{ Ilt ,;

FIGURE III/5-3 REACTOR -

HORIZONTAL SECTION AT BEAM TUBE LEVEL 111/5-4 \Qh

reactor is provided by removable concrete blocks cast to fit the openings. These blocks, weighing up to ten tons each, have pick-up lugs so that they may be handled by means of the overhead crane.

Embedded micro-switches are provided in the monolithic shield which will scram the reactor above one watt or inhibit start-up below one watt in the event that the blocks are not properly seated or other-wise displaced. The biological shield is adequate to maintain most of the reactor room as a Radiation Area. However, the end of the thermal column and the reactor top are designated High Radiation areas.

The measured radiation at the surface of these areas (gamma plus neutron) does not exceed 200 mrem per hour at 100 kilowatts steady state operation.

5.1 REACTOR CORE The reactor core consists of 24 buncles of fuel plates contained in six water-filled aluminum boxes surrounded by reactor-grade graphite.

Four cadmium control blades, protected by magnesium shrouds, move between the fuel boxes.

The fuel plates are in the form of the MTR type (see Figure 111/5-4).

The 93% enriched uranium-aluminum alloy is 0.040 inches thick and is clad with 0.015 inch thick aluminum with a total thickness of 0.070 inches. These plates are 25-5/8 inches long, 2-7/8 inches wide.

Each plate contains approximately 13.0 grams of uranium-235.

These plates are bolted into bundles of 11 plates as shown in Figure III/5-5. In each fuel box there is space for four fuel bundles of 11 plates each. When fully loaded, the six fuel boxes contain 264 plates with a total of approximately 3.6 kilograms of uranium-235.

The calculated cold clean critical mass of the reactor is 3.2 kilo-grams of U-235. In order to adjust the fuel loading to achieve the specific excess k desired for operation of the reactor, aluminum dummy plates may be substituted for fuel plates in assembling the fuel bundles.

An estimate of the worth of a single plate lies between 10 and 15 cents, which should allow sufficient flexibility so that no special or partial plates will be required for adjusting reactivity.

Since heat-transfer considerations are of minor importance for this reactor, a number of different fuel elements could be considered.

It is desirable, however, to use a structure which closely resembles those used in the early Borax reactors, since th, behavior of Borax reactors during power excursions has been experimentally demonstra-ted. The use of metallic fuel plates of high thermal conductivity minimi es the extrapolation of these data so that there is a greater degree of confidence in the calculations of the results in th unlikely event of an excursion.

Plates of 93% enriched uranium-aluminum alloy jacketed in aluminum have been selected for the initial loading because:

A. nuclear characteristics are satisfactory for the purpose III/5 5

.A i

1 l

! I 1 l l

l l I I I l l l I

l I

I AL l l

U-AL l l %'%

LENGTH OF PLATE 25,625" l l I WIDTH OF PLATE 2.845" 1 0.005 l l THICKNESS OF PLATE 0.070"1 0.003 l l l l THICKNESS OF CLADDING 0. 015 "

THICKNESS OF U- AL 0.040" WEIGHT OF U-235 - 13. O grams \

\ l PER PLATE

~

NJ URANIUM ENRICHMENT 937 g FIGURE 111/5-4 FUEL PLATE 111/5-6

9

. n -

@ NO -

1 l l-

. i l

i I

i

~'

gas ,

}_,

FIGURE 111/5-5 TYPICAL FUEL CLUSTER III/5-7

B. proven fuel plates of this enrichment are available.

The six type-1100 aluminum fuel plate boxes have inside dimensions of approximately 5 x 6 x 48 inches high (figure III/5-6). The plates rest on a supporting member,11-1/2 inches above the bottom of the box, which centers the fuel vertically in the reactor and provides for a water reflector above and below the plates. The aluminum boxes are connected at the bottom by means of an aluminum duct through which the cooling water is supplied. The tops of the boxes are connected by aluminum overflow lines connected to manifolds that return water to the dump tank. To prevent air-lock, the manifolds are separately vented to the dump tank. Each box rests on 1/4-inch steel strips resting on the concrete floor.

The top of each box is closed by a plug which extends upward through the graphite that forms the base of the vertical thermal column.

The upper part of the plug consist of 8-3/8 inches of graphite on top of four inches of lead. The lower part of the plug consists of a flanged section which fits to the top of the fuel box, and an alumi-num diaphragm to keep water vapor from the fuel boxes from getting into the graphite space where it might condense.

5.2 REACTOR COOLING SYSTEM The cooling of the reactor is accomplished by a primary and a second-ary cooling system. The primary system, using demineralized water transfers the heat generated by the fuel to the secondary system by means of a shell and tube heat exchanger contained within the pri-mary water dump tank. The secondary system utilizes Los Angeles city water which passes through the tube side of the heat exchanger to extract heat from the primary water. The secondary water effluent is monitored by a NaI detector and passes through a 225 gallon (10 minute delay) tank prior to rejection. The primary and secondary cooling systems are located in the process pit below the reactor floor level immediatel, north of the reactor. The general specifications for the cooling system are given in Table III/5-1. The general primary and secondary coolant piping diagram is shown in Figure III/5-7.

The reactr moling system safety features include:

A. A graphite rupture disc set to burst at one to five psi above normal operating conditions, hence draining the water from the reactor B. A dump valve which provides quick drainage of the water from the fuel boxes into the dump tank, rendering the core sub-critical by a margin of about 525. The dump valve is spring loaded open, closed by air pressure and will open upon failure of the air supply and any full scram signal.

The primary flow rate is metered by a turbine flow meter for visual readout at the console, and is auto-controlled by an orifice and dif-ferential pressure cell that transmits a signal to a two-mode III/5-8

t FUEL BOXES h

/g CONTROL BLADE

/ SHROUDS (TO DUMP TANK) 4

%[

)/

% -;)

,/

(tm)

>'//

.v v TO DUMP TANK HEAT EXCHANGER TO CORE FIGURE 111/5-6 FUEL BOXES AND COOLANT CONNECTIONS 111 / 5 - 9

Table III/5-1 GENERAL COOLING SYSTEM CHARACTERISTICS

Primary Water Flow Rate 16.0 GPM Primary water to core 100 F Primary water from core 0 142 F Secondary Water Flow Rate 22.5 GP:1 Seconda y water to heat exchanger 57 F Secondary water from heat exchanger 87 F Primary Purification Systen Flow Rcte 0.5 GP!'

Primary System Capacity:

Dump tank 200 gal Core 36 gal Total 236 gal

The temperatures shown are typical of 100kw operation as the reactor mass approaches thermal equilibrium.

The temperatures are also subject to seasonal variations.

III/5-10

FIGURE 111/5-7 COOLING SYSTEM 3"

w -

,e)

^

r pd '

DUMP VALVE 3/4" MAKE-UP ORIFICE METER

.q % [ ; M ;

FILTER DEMINERAllZER n 9 1

[

__ _ SECONDARY

._ WATER X __3 DUMP TANK HEAT EXCHANGER CENTRIFUG AL PUMP DRAIN

reverse-acting controller on the console. The controller, in turn, positions the normally closed flow control valve as required, to main-tain constant primary flow. The valve is interlocked with a scram signal to the dump valve so that when the dump valve opens, the flow control valve reverts to the normal or closed position and stops the flow of water to the fuel boxes. The dump valve scram circuit sig-nal will override any other signal transmitted to the flow control valve. In addition, a loss of coolant flow will scram the reactor on a signal from the differential pressure cell.

A sight glass is provided on the north wall of the reactor room floor so that water level in the fuel boxes can be checked isually. Along-side the sight glass is a standpipe incorporating . liquid level detec-tor which is interlocked to the safety system to prevent reactor operation when the water level is outside the preset limits.

Vent lines connect the tops of the fuel boxes and the standpipe with the storage tank. These lines provide an air-return path during dumping and filling operations.

On the high pressure side of the primary system (upstream of the flow control valve) a portion of the coolant (1/2 gpm) is diverted to a purification circuit consisting of a filter and two cartridge-type demineralizers. A rotameter is provided on the inlet to the deminerali-zer for the purpose of flow regulation. A conductivity meter con-tinuously monitors the exit water of the demineralizer and actuates an alarm when it is necessary to replace the cartridges. A deminerali-zer water make-up connection is provided on the discharge of the demineralizer so that water may be added to the system from the building demineralized water system located on the third floor above the reactor.

5.3 REACTOR INSTRUMENTATION AND CONTROL The reactor may be operated in either manual or automatic modes.

Control is effected at the console with the necessary information provided to the reactor operator by meters and recorders. The recorders also provide an historical record of each reactor run. During opera-tions; linear power, logarithmic power, ventilation effluent activity, and secondary effluent water activity; are continuously recorded on strip charts. Twenty-four temperatures (primary and secondary water at inlet and outlet, 3 fuel points, and 17 graphite moderator-reflector points) are cyclicly recorded by a 24 point recorder. Reactor power, period, control rod positions, coolant flow rates , and area radia-tion levels are displayed by console meters.

The linear recorder-controller senses an error signal (between actual power and set-point power) which is amplified and used to drive a control blade (the reg rod) in the automatic mode. The primary flow is automatically controlled.

The foregoing instruments and controls constitute the principal ele-ments of the system during normal operations. Other displays and alarms that are used in start-up, or indicate reactor status, are III/5-12

described in subsequent paragraphs.

The reactor is controlled by means of four semaphore-type cadmium control blades of the type successfully employed in the CP-3 and CP-5 reactors. Through common usage, these blades have become known as

" rods", and the latter name will be used in the following description.

Three of the four rods are used as shim-safety rods and the fourth is used as a regulating rod. Rod motion is controlled from the con-trol console. In case of a scram (reactor trip), all four rods are dropped simultaneously by de-energizing the electromagnetic clutches between the rod shafts and the motors. For normal reactor shutdown, the rods may be driven down simultaneously or individually. However, for additional safety, the rods can only be withdrawn individually.

The reactor may be controlled either automatically in response to power-level demand settings or manually by the operator. The auto-matic regulating system is not designed to follow large changes in reactivity demand. Step demand changes in excess of approximately 7 percent cause the servo to be disenga~ged and the control mechanism to shift to manual operation.

In addition to instrumentation, administrative controls, and quali-fied operators; electronic interlocks and automatic protective systems exist to provide for safe operation of the reactor.

The reactor safety system acts automatically at three different levels:

A. full scram B. drop-rod scram C. inhibit.

There are additional alarms and lights that call for operator response, and the operator may manually initiate either type of scram. Inhibit preve ts the withdrawal of any control rod.

The drop rod scram is effected by removing electric power from the magnetic clutches that couple the rod drives to the control rod shafts.

The dump valve is opened by removing air pressure from the dump valve via a solenoid valve. Loss of electric power or air pressure will automatically induce full scram.

The various conditions that induce a full scram are:

A. power failure B. dump valve open C. power level in excess of 125 kw D. reactor period less than 3 seconds E. shield closures open (above 1 watt)

III/5-13

F. manual scram Items A and B act as ini Lits which must be cleared prior to start-up. The drop rod scram is more general and is induced by the fol-lowing conditions that do not demand full scram.

A. loss of primary pump power B. loss of primary coolant flow (less than 10 gpm)

C. loss of high bay exhaust fan power D. low core water level E. low shield tank water level F. reactor console key in "off" position G. logic condition (system failure)

H. high secondary effluent activity (more than 8 x 10-6pCi/ml of iodine-132 in effluent)

I. drop rods (manual switch)

Items A through F act as inhibits which must be cleared when starting up the reactor.

Other inhibits which prevent the withdrawal of control blades are:

A. low neutron count rate ( < 2 per second), bypassed at power greater than 0.022 watts B. reactor period less than 6 seconds C. shield closure open (below one watt, scram above one watt)

D. log N amplifier not in operate mode The automatic power level controller cannot be engaged under an inhibit condition, and will drop-out if an inhibit condition arises.

A console-mounted visual display panel provides the operator with a constant indication of the reactor status. The display is ener-gized when operating power is applied to the console through the power 0N switch. There are generally two lights for each displayed condition, one red or orange and the other green. The display lights are controlled by means of relay switches or solid state logic. When a condition is not satisfied, a red or orange light representing this parameter is activated. A green light signifies that the condition is satisfied.

111/5-14

The visual display is augmented with an audible horn to comprise an annunciator. The description of the annunciator is divided into two sections A. alarm light and horn B. alarm light only.

The conditions causing alarm light and horn are the full and drop rod scrams. Also included are:

A. high primary coolant temperature B. high radiation C. high argon-41 The following conditions are displayed by an alarm light only:

A. low conductivity in the demineralizer system B. BF3 detector not protected C. reactor sump flooding (6" below process pit floor)

D. any inhibit (with the exception of closures below 1 watt) 5.4 REACTOR FUEL HAtlDLING Approximately 3.6 kg of U-235, 93% enriched, is distributed in 264 fuel plates, each containing approximately 13 g of U-235. When the fuel is in use, the plates are assembled into 24 bundles, or fuel handling units (FHU), of 11 plates each. The loading of fuel into the reactor or any changes in the fuel loading is directed by the members of the staff who are qualified as operators and who are com-pletely familiar with the behavior of the reactor. A special hot cell is set up for visual inspection of fuel bundles when they are removed from the reactor.

Irradiated fuel is removed from the reactor in a 6 inch thick steel jacketed, lead cask by means of the 10 ton overhead crane. Irradia-ted FHU's can be stored in the fuel storage pits during reactor overhaul.

5.5 REACTOR WASTE DISPOSAL C0f1 TROL The amount of radioactive waste issuing from this reactor is rela-tively small. No irradiated fuel processing will be done by the University. All spent fuel elements will be shipped intact to DOE or approved commercial processing facilities licensed to receive and reprocess such fuel.

Since the major function of the reactor is education and activation analysis, and no large scale research or testing programs are planned, III/ 5-15

the production of low level " side-stream" radioactive waste is small.

Solid wastes such as contaminated glassware, gloves, paper, shoe covers, air filters and ion exchange resins, etc. , are packaged in D0T approved containers and transferred to the California State By-Products License under the control of UCLA's Research and Occupational Safety Office.

Waste containers are conspicuously labeled by the criginator to indi-cate the type and amount of waste contained. All waste is monitored by health physics personnel prior to trans fer to the State license.

This conforms with 10 CFR 20 and the " University of California Manual for Radiation Safety", Third Edition, July 1958.

Liquid waste is handled in accord.with the Manual for Radiation Safety and with 10 CFR 20.303. Such wastes arise infrequently from decontami-naticn solutions and draining primary or shield tank water. The waste from the decontamination sink and floor drains passes to a sump in the process pit and is pumped from the sump to a holding tank for analysis. If necessary, it can be diluted to acceptable concentra-tions prior to release to the sanitary sewer in conformance with appro-priate government regulations.

III/S-16

6.0 COMPARISON TABLES 6.1 COMPARISONS WITH SIMILAR FACILITY DESIGNS This section provides a tabular summary in sufficient detail of the principal similarities and differences between the following Argonaut type research reactors : University of California, Los Angeles; University of Florida; and the University of Washington. This infor-mation ~1s presented in Table III/6-1.

6.2 COMPARISON OF FINAL AND PRELIMINARY DESIGN A preliminary collection of reactor specifications and characteris-tics contained in a report entitled, UCLA Training Reactor Hazards Analysis by R.D. MacLain is considered to be the UCLA PSAR. Tabl e III/6-2 describes the modifications or changes that have occurred since the PSAR was written. The table is essentially the Training Reactor Characteristics table found in the PSAR with an additional column sunnarizing present characteristics.

III/6-1

I TABLE lil/6-1(a) COMPARISON TABLE - GENERAL REACTrA tii!VPSITY OF C/i!FDPN!A, tJ41VERSITY OF FLORIDA L24IVERSITY OF MSHlt0 TON LOCATION LOS 46ELES SEATTLE, EHitETON, T195 LOS 4 6ELES, CALIFORNIA M DI GAINESVILLE, FLDRIDA, 32S11 LICENSE to. E-71 R-50 R-73 DOCrET to. 50-142 548; 50-133 Od G BY REGENTS OF Tit tJilVERSITY OF COLLEGE OF D61NEER!t6 (#v!VERSITY OF W H!t0 TON CALIFORNIA 0FERATO BY PAJCLEAR ENERSY LABORATORY UCLA DEPARTTflT OF PAILE/# urb. SCIDCE, tarLEAR ENE M Y DEFART'iENT U OF FLORIDA COLLEGE Or DCINEERitC, U OC W PEACTOR KTEROGEtE005, THERt% LICHT , EXCEPT ORIG!tvi FUEL HAS 20".

TYPE M TER DFICHO REFLECTD, h( t0DEPATO, GRArHITE D61CHO LF#41tJ1 GDEPAL PAJCLEAR COPP., (REACTOR SYSTU4) LCNETT, STREISSC7JTH & ZEJ%

DES!CPC BY GD4EFAL FAXIIAR DrA. (PRINCJPAL (DNSULT44T TD RE/4 TOR SYSTDi) G.C. FULTDN, ARCHjTECT TO THE STATE ARCHITECT STAT (TON & STDC $ ELL,/#CHITECT BOARD OF W4 TROL LBUILDif6)

/4'F (PIACTCA)

COOPER C0tGTDUCTION CO.-BUILDif6 JENTOFT & FORBES-CONTRACTOR CONSTRUCT!Cid TJNES BROTHERS CDtGT. CD. & LOUIS BY C. DJA4 Itg.-BUILDjf6 A.I.-FUEL Aff ATTICS-REACTOR CEff0ftNTS THE t%RTIN CO. (FUEL SET)

ELEMENTS UfD Tl't) HONEWELL- SYLCOR PAELEAR-FUEL ELUiE'iTS CONTRL15, ELECTRONICS 60NEWELL-miTROLS

e ORIGINAL lb DESISNO F0WER tmvm 100+ (19G3) *(1906) *(19C7)

OPERAT!'6 RJWER rpPJVi FJWER .19$/FT, 23.12$/c 035 LLINEAR, SPECIFIC POWER)

OPERATI'6 VARit&_E,13-16 PEGAWATT ms VAFIABLE,15 TO 20 fu ees PtR YEAR VARIABLE SCHE 3JLE PER YEAR ACTIVATION MiALYSIS, REACTOR TRAIN!f6 40 QUCATION OF PAILEAR REACTOR OPERATOR TFA!N1'0, FHI!CIP/i 16E OF OPEFATOR TRAINING, 40 DIATION D6R. /40 SCIENTISTS, REACTOR CLASS DC0rGTRAT10NS, REACTOR OF PAELEAP Dr#. & ST@DiTS OPECATORS, ACTIVATION ArviYSTS '. LASER ISOT3DE PRCOITION, tv'A EXF'S, RESE/eCH OPESATit6 3 PEP SH:FT, 1 SHIF' 4 PART TIME 2 PER SHIFT,1 SHIFT FORT %LLY 2 P/6T 2PERSHirT,1 shirt STAFF STJDD4T OEFATORS TIME OPERATDRS STATUS:

DATE Fl:ST CRITICA_ 10/60 1/59 4/61 EJTE FULL 2/61&l')/E3 4/59, W 62 4/61,5/C7 FM.a IJo, lh SOLID FUEL:

F ' ELDCl! *

A Sh/J1 FLAT PLATE (MTR TYPE)

B FUEL ' '

COMPOSITION 13.4 WT.I LP/i /410Y C) FUEL DIMD6 t 0NS *

(<AT) 24"x2"x.0W 23.5" x 2.3" x .00" D) CLADD!t6 *

t%TU I AL 117) titfil'Af t i El CLADDI'6 . .

THICC.ESS .015IN.

FJ TffE OF *

SUEWaSUBLY PARAi1EL FtATES BCLTD TOGETHER G) # GF ELUit.TS FER *

A{SE%Y 11 hi PLATE *

crectorc 25-5/3x2-7/3x.'170IN.

li S1-ASSE 5 Y (FHJ) *

DIMENS!9tG p Atu.It6 Rit0) X J) # OF *

FHJ'S 24

K) ARFuri- 4 ELEJU,TS IN EACH OF 6 BCxES PU.~ OF SUB- APPAt0ED IN 2 PARALLIL POC ASSE.TLY *

L) LIFETIT lt0EFINITE N THOC or t%fAJAL, USl?6 hArc TOOLS, TRAf6FER REFUELING CASK, PfJiCTE MIRROR 111 / 6 - 2

f TABLE lil/6-1(b) PRIMARY COOLANT FLU!D y *

C{PCLEATION A1 EIRECTION OF FLOk Urgry;3s * .

B) FLOH FORC D FE O , GRAVITY RETIAN

ItCUCED BY

C) PDMi.

Fy)W RATE 16 C/ti I/) Grti 23ceM DJ MEAN VEL. 1100 C@E 1.3 avsEC 4 evSEc 2.1 ovsEC El AVERiLL

) CDFE T b HEAT MTEFFTCHdATER TUEsE-SHELL S.S. * *U TUBE PR]'%RY, SitrLE PASS

DISSIPATION KAT EXCWJ6ER II)UBLE-PASS SEC0f0ARY-(1 Plt RATED)

KTIOD SECDPE/RY (TEE)

T ANS OF

PRIP%RY SYSTDi: CDPJTIPUQUS h GF"4 *( % tt/ MIN)

PLRIFICATION BYPASS T)PU CARTR!DGE FILTER THEN TIR'JDOilNERALIZERS /JO RETJRN TO DLf1P TANK SHIELDTAQg: It0EPDODJT, CQP TIMIX13, 3.1 GPH, FILTER NO 4 PARA LEL DDiltLRALIZER CARTRl[CES NUCLEAR DATA FutL L0cen A) COLD CLEN4 CRITICAL r%ss 3.2KoU235 .

3,79 gg tp5 B) PEyc FFESH FUEL Loeiro 3.6 xs U235 .

3,q39 ga p9 C) EXCESS K, FRESH L0eac 2.3%>dx *

F

^gux a lh THEwYaux 1.5 x 10Nc2SEC 1.4 x 1912,gg32

' 3cc NsT ELE <1.5x10%cdSEC

ll C) PEAK 5.0x10evc?SEC THEv;L

$NE Flux 1.0 x 10kc8 SEC 3.0 x 10DC$ SEC REACTIVITY CNFF IC!DiTS IJTEP)

A) Ttm. -0.48 x 10',%'

dk (-0.74 U F)

-1.0 x Id 3 dd C %? h/d e -0

+b F(M),

B) v0!D -0,lrA x 10-4,5 dd%c!D 14.

-0.21 i d d% Kp w!D . :db' JT (GRAry!TE) w!D

(-bu% wla) .

C) t%SS CQ;[FICIENTS IP 0.33"t-x/x(1 cran >) * .

D) Co'E

0.01% ddce U23d ExCEn 2.3%

x/K (3.'34)

sana FolsoNs tac *

N s 6.6MCIP4A ur To 5 Cl SB-BE 401 CU RBE SB-BE (2 Ci fbSE IF NEEDGI f

Ill/6-3

f TABLE lil/6-1(c) REACTOR CHARACTERISTICS s' 4 2.55 x 10 SEC (K AS m D )

St.FC IP x 13 A . 5SEC AjS'J C -,

pjJEC s

B ASSif O B EFF

b *WQ BIOLOGICAL C0tCRETE COfCRETE SHIEL3 CONCRETE A) PW UAL '

s RNi BAR B) j SE L A) FWER FAULLEE BI PWiyi FULL SCRXi J)) POO FALLURE B) PWOAL (DROP ROD & (FULL SGPM1 BUTTOP.I C) 90RT OR LOP 0ER CJ l Q F 2 Lh ( S) SGPW1 C) 90RT pl0D ' > SEC WATER) PER10Q (LESS Tg 3 EC) D) HIGH (2SAFETYCHAjtELi D) PDE [ RE- D) HIGH FLUX

lb M E) OPLt.

FLUX (R)WER > O m E) CLDSURES SULTS IN LOW rl LEVEL, 'O PO t{ FLOW DLIP VAuVE OPENABOVE1 WATT F Dff VALVE SCP/et. INTERLOCh5WITH PtitP. KEY OPEN TURNED OFF WITH 4 OR FORE BLADES lf

FULL SCAN 1 WITH HATER DfP ,

DROP RCD A) KEY TURND OFF B) LOSS OF A) IF OtLY l BLADE UP = PWi C'UTCKS A) TmNy 0FF B) LOW FLCM SCrWt HIGH BAY VENTILATION LOSS OF OFF B) LDSS OF DIUJT!tf OR VENT FAN FHl W Y C LOW CORE LEVE1.

PRIP%RY PtfiP FWER D)C)LOSS OF F%HER C) LOSS OF PC Rff RMR L PRI%RY D LOSS PRI%RY FLCU E)1LOW PR!mRY CDOLANT FLOW E) LOW CORE D) PC FLOW DROPS TO 30 Gm FRO O SHIELD TANK MTER LEVEL WATER LEVEL F) LOW SHIELD TNX Qm E) PC WATER LEVEL DPODS A y F) QROP ROD BUTPJf4 G) DlLUTION WATER LEVEL G) HIGH SEC0f0ARY 4 INCHES (STILL MOVE FUEL PLATESi FAN (AIREMiAUST)

N TOR StVLL DROP - PU/ABLE LEVEL SWITCH (FFgUEfg&

t>oX Cl/CL I-(D IN lh0)

PflATIONf F}

Gi PO SCRN1. WEEKLY SMFLE #erLYZED H) DROP ROD BUTTON I LDGtC Cor01T10N lt441 BITS A) NEUTRON START-UP 40URCE CDurlT A) $#E AS SCR#1 BAR B) }SOr2D A) PER!m < 10 sEg,. B) 1 1E LESS IHAN 2 CeS B) PERIOD LESS WATER ABOV( l m. PC TD1P HIGiER Ci 10 DROP IN EUTRON

- OR ltt(T ' N C) br3' CTS /SEC THAN b $EC CJ CLOSURE OPEN (BELOW l W) D) LOG-N A'1FtlFIER OWEER HlW Q.TAGE, EITTR OF 2 tf/

PCT IN OPEFATE t0DE SuoPLIES DRIVE LYD)E )4 CPS LOGIC It411WITH-

- C##OT BIT OF UP DRArv OR FORE BLADES S'tiLTNEOUSLY F 1 SECDfD PERIOD UP IVE G PU4 H) WIDE rat 0C LOG N +

STARTUPJ OWrEL ALMYS IN OPERATION.

ItO! BIT IF CALIBRATE OR TRIP TEST SWITCHES POT IN OPERATE PODE (FOR UP DRIVE)

ALAPli3 A) HIGH PR]"/DY con.tJtiT EXIT tut- A) 1NY*AUDIBLEALARM B) AWOF A) plGH m (Ll WT & i 2) HIGH AFEA 5 eaEA fTWITORS " ALAR't LIW @ 2b rf/ lfdrCTSlk'TPYEXITLET

) F B) HIGH HAid PERATURE - l'm 5tm TH &

RADI AT 10t4 m - AUDIBLE tuat & Llw S13MR/ RA ATI IfEAS REACTO RADIO- FLCOR ' .5 F / RASSIT SOUTH HIGH BAY 2 m/ m. TWO OWtLS 3 il m/m ALSO = LOSS HR OF GREEN REACT A TOD EX1T >

ACTIVE STOPEE > 33 fE m [VACUATIONAudiSIRENS.

RAEBIT i C 7g!R M 41 IN r0 FAIL" ' IMT l' OWtEL P0DULE t0T '5m/m Hl3H APGON STACK ' X 41 ft IN AIR RECEIVit0 SIGrA. rROM AREA FUNITOR C) ## SCRAM OR RCD IU CHAMEER C) 2 CUPS WATER IN PC PIT SLf1P

AgDIBLE ALA? IN (DOOL RW1, LIGiT DJ ALAPfi LEVEL, ST/EK ErFLUENT ARJUST/GLE ACCORDlt0 TO FOO OF OPERATION 41Ri S- Al LDSS OF POO OR OPEN CIROJIT IN A) PRIMRY Cor TIVITY LIGHT OftY A) LOW FHIMQffCDOLW!T R T!VITY ' 1 X llof ts B) EVACUATION SIREN

LOSS OF GP EN LIGHT ' ~ ,97 Otti B) H13H D[TECTOR IN HIGH FUJx 0. WATTS B) PC SOLU BRIDGE P0NITORS D, M ItLET FLUX ' 2)HATTS C) I ilBITS CJ #N IPOlBIT APO OUTLET RESISTIVITY ALARM LIGHT ADJJSTACLE ON INSTR'JtE!!' C) A!R CDrotTiofER TRIP BY SIREN

RD LIGiT D) B PROPORTIOffL C0tfiTER WlEN EFEnGlZED ms RED EXTEr0 D RA L) GIT. AUTur% TIC P/ CUTCPS OFF 8 M.W, E) FAST PERIOD If4'IBIT LIGHT %

SWITCHES F) FAST FWER If44! BIT LISHI

& SWITCHES POT IN OPERATE ltelBIT PODE Tc0 ALAP!1 Li WATER FLOW GO Gm a TTS em Fb>cm RARM

SCFW1 AT OP /M/E TO AFTER D SECO'O MRN!r0 LIGHT Bi LDSS OF LIGHTS

FAILURES IN RAD PON STAT 0BY BATTERY POEP PACK C) AUDIBLE ALAR;i & R D LIGHT IF BACK DOOR 15 OPEtc. RD NO GREEN LIGITS PONITOR OUTER ACCESS DOORS TO FACILITY 111/6-4

TABLE 111/6-2 TRAINING REACTOR CHARACTERISTICS

mt 110 lom TrE ttTtROCDIOUS, THEMR PO CR 13 t;! 1Y O rLu< trc (AT 13 0;) 1 x 13lW C sEC 1.5xIPVCCEC EXCESS FEACT!v!TY (TIO! SPEC Ll"IT) 0Z: AT M 2.32 i AT RCot TDr ExC'S3 PLACTIVITY INSTR 1rD 1.5% o r.T 000rt Ter 1.E. ' AT R001 TElr CLEki COLD CFITICAL MSS 32W Gt l'-235 EFFECTI't ProrrT LIUTR0r: LIFET!!E 1.4 x 134stC 2xid tEC d.!FOGt WATE9 VOID COEFFIClC T -0.1C% c/T voln -0.1E4% # sctn TerErATurt CotrFICic.T -0.f48 x 10 4 %M .431 x 1 N i U-235 mss CarriCIENT +0.31" c.C L-235 r%s- +.3" c5 L-27.,

START-UP SCTCC 2 CURIE PU EE C.(> fi Cl fA I PIFLICTORS CfAPSITE (1.E7 G'/CC) 101 RAT:it IM Ar0 GPAPHITE ILAYED FIUToor. FPACTla: Oh .(IT5 FLEL PLATES FUEL 93* ErEICKD, U-IL Adm FUEL LC C !rr, 3,4'6.2 at l'-2.35 3,55C e4 E-235 PLATE ThlC01Ss 0.073IN.

w,Tt? o A IL 0.1371t:.

ntf41rm o wATtP RATIO (VOL) 0.51 FEAT CD 'IT13; 13.46.T7C-[L COOLA!.T 4,9 FLO: lJGr" li rfn TerErATurr, it. ISR 1rff TurErv,T71, or 11'PF Idi Cu rsat ruoEs e, Sun.ctr:s vnz, cm:w ru r JEER. 3 sefEm 1 REr.utATIOr.

Ir.:EnI0r. Tut 0. D sEC (C aCutATtD) 0.5SEC(trxurE:)

Pr:e.'At Trt T stC (r Ir;ru0 lI .i tuoE wTs, sAFETv 3 rCDc 1.5" c = 4.5" : 3 R0m 1.f* - 4.T BLATC bETh, PfGliATif.G 1 POD 3.C" c = 3.fJ c 1 RODS 1%,

TCit - 5.1". T0uL s 5.I REACTIVITY /CITICG FATE,1A(. 0. C :/IEC . E /;g; PAP?/ fit, SHIELD (Cor.CTITE)

O!XS, CEf.TEr, fFT.OIN. CAST,r%rtrTIE 5!K S, SF>1 ELD TN.E D C C FT. 8 IN. CAST, OCDINAT'i SIDES, TIES %L COLtf t. DC ( FT. 8 IN. CAST, P%GriTITE I'IDDLI CAST CDrKTITE ELIO $

K A E C0rt 5 n. l') It:. ma;ETITE EtoCis

rtu: 30" oc DorATE" PAFC1 Elf; EIC5 3 n. 4 IN. MGIETITE BLCC)$

Ex5EP!iEf.TAL FACILITIE!

THEF!%L COLIJf;, HDPID31 5 n. x 5 n. x 4 n.11 is, tu C0 it.. x 52 IN. x 43 it.. LONa PervcLE ThE9%L COLLlt., WRTICAL P'OVISION FOR INSTt41AT10t4 -

SHIELD TEST Tna. 5 n. x 5 n. x 14 n. E tr. Hio.

EXPERI'Et.TAL HDLES 2 - tF12DrCAL, C IN. DIA s.TEr.

5 - tatacAL, 4 IN. DINETEP 4 - For!3rJfi, 4 IN. CI A*ETEP 3 wrmCAL, lh it:. DIREER 3 - VERTIC1,1-7/8 IN. Ol#ET T.

ExPERITf.TAL K4.ES, TFEPlt. COL. 15 FU0VAttr GRAPHITE STR!fCEP:

FCIL SLOT 5 11 - I'.#1Z0r.TAL, 3/0 IN. X 1/2 Ir..

IC-vteTICx,3/3IN.x1In.

111 / 6 - 5

7.0 IDENTIFICATI0t1 0F AGErlTS AtiD C0f1 TRACTORS This section identifies the prime agents and/or contractors responsible for fundamental design, engineering, construction and operation of the UCLA reactor. Table III/7-1 identifies these agents from the formulation of the design to the operation of the reactor.

Table III/7-2 depicts the current organizational management of the reactor operations.

III/7-1

Table III/7-1 IDENTIFICATION OF AGENTS T'ESIGNER M0mE pp7pg PRINCIPAL CCNSULTANT

MWM 39gy GENERAL AMERICAN N'JCLEAR f% CHINE ENGlNEERIt6 FOUNDARY ARCHITECTS I

STANTON S STOCKWELL ENGINEERING STRUCTURE: NEL MECHANICAL DONALD LEVINE C DOUGLAS MC CANN CONSTRLCTION BCELTERlNEL CONTROL R00lf CONSOLE 1M & 2nd LEVELS ELECTRONICS h' ALLS OF HIGH BAY LOUIS C. DUNN, HONEYWELL, INC. INC.

PLUMBING, ELECTRICAL, VENTILATION, AIR-CONDITIONING, CCNCRETE SURROUNDING 0F REACTOR JONES BROTHERS CONSTRUCTION CO.

OPERATION N'JCLEAR ENERGY LABORATORY SCFOOL OF ENGINEERING AND APPLIED SCIENCE UNIVERSITY OF CALIFORNIA, LOS NEELES III/7-2

TABLE III/7-2 BOl#D OF REGEt4TS (LICE *CEE)

UCLA CHANCELLOR

\

EXECUTIVE ADMINISTRATIVE ASSOCIATE VICE CHANCELLOR VICE CHANCELLOR VICE CF%tCELLCR (ACADEMIC) RESPONSIBLE OFFICER (RESEARCH)

DEN 4, SCFOOL OF ASSISTNiT 4 --- ENGINEERING NO VICE CHAtJCELLOR COffe1ITY SAFETY V APPLIED SCIENCE ,

DIRECTOR RESEARCH RAL'ATICt1 REACTOR USE DIRECTOR, PfJCLEAR Ato OCCUPATIONAL - - - SAFETY COr1AI TTEE ENERGY LABORATORY SAFETY COtHITTEE RADIATION SAFETY g,EER NILEAR ENERGY LABORATORY F 4 4 REACTOR OPERATIONS PHYS C ST t

FUNCTIOt#d. RESPONSIBILITIES EXPERIMENTAL USE APPPOVAL STAFFIt6 OVERALL SAFETY BUDGET NRC CO'4U11 CATIONS TECHNICAL CHA*6ES UNIVERSITY PJBLIC

^

PREPARE LICENSE AMDJDMENTS ORGANIZATIONAL RELATIONS 111 / 7 - 3

8.0 DESI6a BASIS, ACCIDENTS, AN:L:0NSE00ENCE h Calculations pertinent to DBA's aild Consequences were presented in the 1960 " Hazards Analysis" (Reference I) that attended the original license application. These are reproduced in the original form and attached herewith as Attachment A and B.

Attachment A (titled Appendix B in " Hazards Analysis") treats with step changes of reactivity (Ak/k < 0.006) and limits the maximun licensed excess reactivity to 0.006 to asoid prompt criticality.

Amendment 7, approved in 1966, increased the maximum allowable excess reactivity to 0.023 with the restriction that no single exposure cavity would contain an excess reactivity of 0.006. The excess reactivity limit of 0.023 was justified on the basis of SPERT and BOPAX experimental results (see Attachment A). The provisions of Amendment 7 remain in effect today.

Attachment B (titled Appendix C in " Hazards Analysis") treats with Radiation Doses Resulting from Release of Fission Products into (the) Atmosphere. The release is not causally related to a specific accident, and from the SPERT and BORAX experiments, one can only state that the calculations attempt to suggest a scale of events that might follow a catastrophe of unknown cause. The calculation of fission product inventory is based upon a steady state equilibrium inventory at 10 kwt, and certain assumptions concerr.ing leak rate from the building.

The consequential dose calculations were apparently unreviewed in the approval of Amendment 3 (1963) that increased the maximun licensed power level to 100 kwt. They were reviewed by the Division of Licensing and Regulation in processing the application for Amendment 7 (referred to above). In view of the current restriction of the UCLA Reactor operating hours to 5% of the year, the maximum average power is now 5 kwt, a factor of two less than the 10 kwt used in the original calculations.

Because the basis of the earlier calculations are not exceeded in the present application, those representations and actions are herewith incorporated in the Technical Specifications.

111/ 8-1

9.0 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION This section does nct apply to this facility. The reactor has been at full power (100kw) operation since 1963 without any undue risk to the students, staff and faculty of UCLA and to the general public in Los Angeler and adjacent areas. There are no evident require-ments for further technical information.

III/9-1

10. REFERENCES

1. R.D. MacLain, UCLA Training Reactor Hazards Analysis, Final Report, Department of Engineering, University of California, Los Angeles, Report No. 60-18 March 1,1960.

2. AMF Atomict and General Nuclear Engineering Corporation, Start-Up

_ Report AMF-6NE Educator Reactor University of California Los Angeles, Cali fornia , January,1961.

3. American Society of Mechanical Engineers, Nuclear Reactor Plant Data, Research and Test Reactors, Volume 2, March 1959, McGraw-Hill Book Company, Inc. New York, New York.

4. C.E. Ashbaugh, Nuclear Energy Laboratory Report, UCLA Nuclear Energy La bo ra to ry , 1977.

5. A.L. Babb, University of Washington Training Reactor Final Hazards Summary Report, College of Engineering, University of Washington, Sea ttle, Washington , May,1960.

6. Personal communication with Pat Miller, Reactor Supervisor, University of Washington.

7. Personal communication with N.J. Diaz, Reactor Supervisor, University of Florida.

111/10-1

APPENDIX III ARGONAUT SAFETY ANALYSIS REPORT (ASAR)

Attachment A Estimation of Effects of Assumed Large Reactivity Additions from UCLA TRAINING REACTOR HAZARDS ANALYSIS, Final Report, R. D. MacLain UCLA Department of Engineering Report #60-18, UCLA-NEL 2 March, 1960 (there titled Appendix B)

III/A

ESill1ATION OF EFFECTS OF ASST 1!ED LARGE REACTIVITY ADDITIONS It has been demonstrated repeatedly in the Ibrax and SPEllT reactors that water-cooled, water-moderated reactors of suitable design may have a very substantial self-protection against the ef fects of reactivity accidents, even in the absence of corrective action by the reactor centrol system. This self-protection is provided by the negative steam-void coefficient of reactivity and the negative temperature coefficient of reactivity, both of which can result in iniportant reactivity reductions as the reactor power rises.

The UCTil has been designed with a high degree of self-protection of this type. In this appendix est imates are made of the behavior of the reactor under various hypothetical con-ditions of excess reactivity addition with no corrective action by the control system.

%e characteristics of the UCTil which detemine its behavior during power transients resulting from large reactivity additions are quite similar to, but not identical with, those of the Ibrax I reactor. Its behavior can be predicted most reliably by utilizing the Ibrax I data with simple correction factors to convert them to the UCTfl conditions.

He significant quantitative characteristics of the UClit and the Ibrax I reactor are compared in Table 14-1.

'l illt.E i:- 1 C0\lP A1:ISON Di l'C1R ANil UOlt A X I Cil 4 R 4 CT EltI S11 CS CHARACTERISTIC UCTR BORAX l Fuel plate " meat" 13.4 w/o U-Al alloy 18 w/o U- Al alloy 90% enriched fully-enriched Fuel plate cladding 1100 aluminum 1100 aluminum

" Heat" thickr.ess 0.04d inch 0.020 inch Cladding thickness 0.015 inch 0.020 inch Coolant-channel thickness 0.137 inch 0.117 inch Core volurre (approx. ) 71 liters 100 liters Void coefficient of -0.18% A/% coolant -0.29% k/% coolant reactivity (calculated) void void Temperature coefficient of -0.0095 k/'C (estimated) -0.01; k/ C reactivity (roon temperature)

Effective proert-neutrcn 1.4 x 10-4 sec 0.6L x 10~4 sec lifetime (calculated)

Power ratic in core, 1.03 1.82 max in.um average III/A-1

In addition to the quantitative dif ferences, the UCTIt dif fers from Ilorax I in that the n aximum coolant water level is only a few inches above the upper ends of the fuel plates (instead of about 4 f t) and the coolant wat er, once it has been ejected forcibly from the core by a power excursion, cannot fall or flow back int o the core. .

I:ricci of 0.E INeess peact isits An excess reactivity of 0.U; kerr will 1 e available in the reactor if its tenperature is abnormally low (nearly freezing).

a The addition of all this excess reactivity will cause the reactor to operate at power such that the reactivity losses associated with the temperature increase and the voida formed will equal the initial excess reactivity.

If the reactivity is added slowly, af ter the reactor is critical, the power will approath such an equilibrium level slowly as the reactivity is added. I f the reactivity is added suddenly, when the reactor is initially subcritical or at very low power, the power will at first rise exponentially with a peri J not shorter than 0.8 see which is the asynyt ot i c period corresponding to the full excess reactivity of 0.67 k,tr. Many experinent s wit h t he Borax reactors have demonst rated that for periods of this order of roagnitude, the transition from the exponential power rise to the equilibrium poacr level (in which excess reactivity is balanced by temperature and steam void coef ficients) is a snooth one involving little it can l e said that the mapiitude or no power overshoot. On the basis of this experiem e, of the power excursion which results f r om the 0.tM reactivity addition will not depuni greatly on whether the reactivity is added suddenly or relatively slowly and in neither case will it approach a level which would cause a fuel plate to hu m out .

In order to compute the power lesel at which the reactor will operate after the addi-tion of the 0.69 excess reactisity discussed in the foregoing, it is necessary to know the water-tenperature coef ficient of reactivity. %e relatise inportance of the two modera-tors, graphite and water, in determining the ef f ective neutron tenperature introducc un-certaint ies in the theoretical computation of this coef ficient. The coef ficient cannot, however, have an absolute magnitude less than that of the water-Jensity coef ficient of react.ivity referred to a temperature scale, i.e., the coef ficient computed on the assumption that:

0Neff _ bLeff bP JT SA ST where p refers to the water density and T to the tenperature. Oi the assunption that this n.inin,um value is the true value, u Iise of water tenperature from near O'C to 80'C would reduce reactivity b 0.u? L,,f.

L e capacity of the reai tor-coolant systen, u such that i f the out side air tmperature were O C and the averarc wat e r tmpe rature in t he reactor were BU'C, energy would he rmoved at the rate of 365,000 IrIE/hr or 107 Lw. Under these conditions the reactor water-inlet te:rperature would be 60*C and ti e exit temperature, coincidentally, would he 100 C. It is, '

there fore, concluded that if the full availanle excess react nity of 0.6% kerr were added to the reac t o r on a cold day with the coolant syst en, operat ing, the reactor would c;erate e at an equilibrium power level a!out ten tin es hipler (100 kw) t han it s normal maxin.um with little or no net st eaa pro <hu t ion. Irfore r eu h i n:. tl.e equilibriun. power , when the w at er in the coolant system would he heat ed t o the equilibriun. value, the reac t or would ope r at e ut a smie hat higher power level and son.e net steam protiuction might eccur. I f t he cool an t were not flowing during the tire of exc es . react n it y addition, t he quilibr iun power level III/A-2

would be quite low and equal to the heat losses. In no case would the power level approach a high enough value to justify any fear of fuel-plate burnout.

Masimum Jolerable SudcJen Reactiv,ity Addition in order to assess the safety factor which exists between the normal excess reactivity available in the reactor and the excess reactivity necessary for a serious power excursion, it is useful to estimate the value of excess reactivity which, i f suddenly inser ted and not removed by the control system, would raise the maximum temperature in the hottest fuel plat e to the rnelting point. Such an excursion would damage the reactor core but would not result in any substantial release of fission products.

'Ihe first step in the procedure is the estimation of the exponential period corresponding to the excess reactivity which would have characterized a power excursion of similar ef fect in linrax 1. 'lhe estimate requires that (1) a relationship be established between the maxi-mum tenperature of the fuel plate and the energy release of the excursion and (2) the energy release be related to the period of the excursion.

For the case of power excursions of short period, with reactor water at saturation tem-pe r a t u re , it is shown in lieference I that the maximum fuel-plate ten:peratur e rise is, to within experimental error, proportioeal to the maxicum energy release of the power excursion.

'lhe proportionality was determned t e l e const ant 24.4 F per W -sec.

Measurevents of the same type with cold reactor water (the case directly applicable to the U CTH ) sho.ed a simi-lar relationship lut with a protor tionality c onstant o f onl y abou t 10 F per MW- sec (Ile f-crence 3). 'lhe differente is nte an unreasonable one since the subcooled water represent s a more effective heat s in). than the saturated water. lloweser, the experiments with the saturated wat er were carried to .short periods in the runye of interest whereas the subcuoled experiments were lir..i t ed to longer periods. 'l her e fo r e, note conservative saturated water data will be used.** 'lo r ai se the ma x i n.ur' t er pe ra t u r e o f t he fuel plate frm. the t mpe ra-ture of boiling water to the melting point o f aluminum, a tenperature change of approximately 10#f 100F F, would require a power excursion with a total enern release of 24 4 F M M -dec or 41 M % -sec.

According to the dat a of fleference 3, replot ted in Figure D-la "subcooled" po*er excur-sion of reciprocal period 150 sec-1 would give an energy release of 41 M W -sec in additibn to the energy necessary to raise the fuel plate t emperature to t he saturation temperatur e of water. It is, therefore, concluded that a power excursion of period at least as short av 1/150 sec (6.7 millisec) could have been tolerated by llorax 1 with subc ooled water without melting at the hottest point in the fuel plates.

Ihperinents of the Ibrax and SPERT t ypes have not been made with reactors having widely di f ferent neu t ron li fetimes. 'Ihe reneral evidence o f the experimen t s, howeve r, s up.

port s the supposi t ion that of the three related variables--neut ron li fetin e, exiess reactivity, and exponent ial period--which charact erize the neutron physic s o f a power ex-cursion, it is the exponential period which determines the total energy release and the ternperatures attained. The excess reactivity and neutron li fetime have large ef fects only as they jointly det ermine the period. *lhis supposition is consistent, for exa:rple, with

'Actually, the energy data of he ference 1 eere revised in Beference 2 because of later and better calibrations of the instrumentation. 1he numbers above are taken froa the later (more pessin i st ic) data.

If subcooled data were used. the case directly applicable to UCTh. this analysis mould indicate that step reactivity additions 2.4 times as large as those discussed here mauld not d am a ge t ? reactor.

III/A-3

that the total energy transferred to the coolant water during a power excursion is many times the amount which would vaporize enough water to c<rripensate for the excess reactivity, and that the actual react ivity reduct ion which occurs during the excursion is noch larger than the initial excess reactisity. 'lhe ext ension of t he Ibrax resnits to the UCIll is made on the hasis of this evidence.

It is convenient, first, to t reat only the ef fect s of t he slightly great er fuel-plate spacing and the slight ly lower void coefficient of reactiuty of the UC111 rela t ive to t he Ibrax 1. Information will also le drawn from the Ibrax 11 experinent s. 'lhe Ibrax 11 r e-actor di f fer ed from Ibrax 1 in t ha t the coolant-channel t hickness was greater in the ratio 0.264 in.

0.117 in, 2.26 and that the calculated void coef ficient of reactivit y was lower in t he ratio.

0.1 5 k,rr/i soid 1 2.4 0*416 0.245 k,rr$ soid lbth of these dif ferences would he expected to cause a higher energy release per fuel plate in Ibrax 11 than in llorax 1 for a power excursion of gnen period. 'lhe neasurecents n,ad" with sulevoled water at periods down to 23 millisce showed that the energy release per fuel plate in Ibrax 11 was letacen 1.7 and 2.0 times that of Ibrax I, with the smaller ratio applying to the shorter periods (lleference 2). 'lhe r e for e , it seerr.s qui te conservat ive to assume, in the case of any two reactors. (1) an 1 (2), of the Ibrax t ype having a ratio of fuel plate spacings, S1/h2, and a ratio of soid c<u f ficients of reactivity, C /C1 , 2that the ratio of energy release per f uel plate for a suhc ooled poacr excursion of gisen perto.1 will he no greater than E2/El 52/S1 or E2 /lil = C 1/C2 whichever is the lurrer. For t he UClit and Ibrax 1 the ratiov are Cm, 0.24 St1 ~ 0,137 - 1*17 _ 1.33 Sm, O.117 Cy t O.18 It is concluded, there for e, t ha t a Ibiax r eactor having a coolant-channel thickness and a void coef ficient of reactisity equal to t hose of t he UCTit would release not more than 1.33 times as noch energy per fuel plate as Ibrax 1. The limiting nonmelting period for sus b a reactor would he that which in Ibrax I pase an energy release of 41/l.33 : 31 MW - sec . 'lhe period obtained from Figure 11-1, correspon ling to a total energy release of 31MW-sec, is 8.3 millisec, in coinparing the lehavior of dif ferent fuel plates, it must he recognized that the t otal energy release o f t he pow"r excursion can no lonrer he considered as a definitise variable ircause a large f rac tion of t he tot al energy released is stored in the fuel plate during the important stare of tie reactor shut-down. For example, a reactor composed of fuel plates at harh heat c apacit y undouhtedly will experiem e a larget tot al enern re-lease, but not necessar ily a haphez maximum ten.perature, during a power excursion of gnen period, than a reactor having plates of low heat cupacity.

1 rom examination of the Ibrax results, it seems clear that two distinct phases of the react or shut -down pr ocess occur con 3ecutively and that both may he important in determining the maxicum center tem}cratur e of a fuel plate. The first phase covers the interval before '

an in.portant amount of hoiling occurs at the fuel-plate surface. During this interval, the heat loss to the wates is sn all and ti.e inportant consideration is evident ly the ratio of fuel-plate surface teftperature (wlii r h de ter ma ries tir start of hoiline) t o c en t e r t err.pe r a-ture. For periods in the range under consideration, this terrperature ratio is theoretically not far from unity (0.70 minimum for a 10-mi l l i _s et period in Ibrax !). Experin,entally, the III/A-4

temperature ratio was unity for periods down tn 5 millisec in the Ibrax 1 measurements.

Since the total ef fect is sn.all and since the terrperature ratio for Borax and l'CTH fuel plates should not be much dif f er ent , t i.e thinner cladJing will tend to balance the effect of the poorer " meat" tonduc t ivi t y. It is concluded, therefore, that there will be no im-portant dif ference in fuel plate performance dur ing this initial phase of the excursion.

1he second phase of the [mer exc ur sion Lepins when a signi ficant rate of Loiling is estallished at the plate surface. Heactivity and consequently generation are reduced at a rate which must be a function of the rate at which heat can le trans ferred into the Loiling water. At. the sane tine, the transfer of heat into the water removes heat f r orr. the fuel plate and limits its temperature rise. 'lhe irrportant characteristic of the plate during this phase of the excursion is the heat flux which it can supply to the water for a given temperature di f ference between the plate center and sur f ace. A figure assuned to be roughly indicative of the relative performance or nerit of fuel plates during this phase is tir ratio of heat flux to tenperature dif ference under steady-state conditions. This ratio (figure of nerit) will overenphasize the dif ference letween fuel plates since the tem-perature distrilution in the plate will te n. ore peaked during a steady-state conduction than during conduction when the general tenperature level is rising. 1he ratio of these figures of nerit for lerax I and for the l'C1H is

'lleat 11ux' aTc-s l'C111 = 0.82

,lle a t I'l u s 6T<-s Ib ra x A consersative procedure would le to apply the alose factor to the permissible total enerry of excursion on the Forax I cur se. At the sau tine, hu*eser, the di f ference in gross rmximum to average po*er ratio for the two s cact ors should le taken into account since it is the ten.perat ure of the hot t es t point in the Lottest fuel plate which is being considered, lhe power ratio for the two reac toin is Max

^ " ' borax

_ l.82 _ 1.1,3.

Max 1.63 AV" UC1H

'lhe ccxr.Lination of these two factors reduces the pern.issible equivalent energy of the borax-type excursion to 31 x 0.82 x 1.12 = 28.4 Mw -a c 1he corresponding exponential perio! f ron I igure 15-1 is 4.1 rni ll is ec. It is, the r e f or e ,

conc luded that tim l'CTH w ill tolerat e a pm er euursion ut per mi at least as short us G.1 rrillisec withcut the meltinr of any part of any fuel plate. 1he excess reactivity corres-pending co this period is 2.3" L egg.

Succeshise_ Power Excurhite It is typical of the Borax and SPERT reactors, unless the excess reactivity is removed by external means, that an initial power excursion which terminates itself by exielling water fron. the reactor core will be followed by sulsequent excursions as the water falls and flows back into the core. An exception to this behavior occurs when the initial III/A-5

excursion is violent enough to cause a permanent loss of reactivity by throwing a large anount of water completely out of the reactor tank. In the UCTil the total quantity of water in the core is small, the sub .ergence of the core is small, and baffles abose the core are so arranged that any water splash is directed to the outside so that it cannot retum to the core. Consequently, even a relatively mild power excursion (e.g., one hav-ing an exponential period of from 20 to 30 millisec) in the ULTil should result in pent a-nent self-induced shutdown of the reactor. Ify these same design features, the possibility of large successive po*er excursions, such as those studied in the SPEllT project, resu l t -

ing from the ramp addition of excess reactivity is eliminated. It can be anticipated that the UClit will be safe against quite large ramp additions (larger than 2.3% k,gt) provided only that the ramp rate is not so rapid as to add an excess rest tivity of more than 2.3'?

k,rg before the reactor power reaches a high level. To exceed this limit the rany rate would need to be of the 1-der of 1.(G kef f per second or larger.

_B ea m Tube Henetivity I'I t ce t s The UCTil has two 6-inch diameter beam tubes which extend to within 11 inches of the fuel-graphite interfaces. The maximum change on the core reactivity which can be effected by these two beam-tube facilities was calculated to be 0.lfM AK/K or 0.09; AK/K per beam tube. 'lhe calculation is based upon the effect of a black absorber six inches in diameter placed in the same position as the beam tubes. lhe reduction in the reflector savings due to the black absorber was calculated using the following equation.

(reflector savings) = N ore) L (reflector). tan h T( reflector thickness)

D(reflector) L(reflector) lle fe rence: Elements of Nuclear Reactor Theory, Glasstone and Edlund.

The reflector savings for the 49.5 cm and 28.0 cm of graphite were calculated to be 7.83 cn, and 5.26 cir, respectively. 1he area of the black absorber is 134 of the adjacent core face area. Using the reflector savings given above and the area weighting factors, the reflector savings with and without the six-inch diameter black absorber were calculated to be 7.33 an and 7.83 cm respectively.

1he reactivity effect of the single six-inch diameter black absorber was then deter-mined calculating the critical buckling with and without the black absorber.

i Using the value of -0.0% AK/K , for a single beam tube the shortest period which the reactor could go on, due to the sudden withdro al of a black absorber from the six-inch beam tube, would be approximately f;0 seconds. Ther efore, the reactivity change wl.ich can be ef fected by the beam tubes does not represent a hazard to reactor operation.

In addition to the two 6-inch beam tubes which penetrate the outer reflector, there are four 4-inch beam tules which ter.minate out side of the reflector. No calculations were made for the 4-inch tubes since their ef fect on reactivity will be much smaller than that of the 6-inch tules.

III/A-6

ilEFEHl:hCES

1. Dietrich, J. H. and D. C. 1.a y m a n , Transnent and Steady State Claracterts-i tics of a Boiling Reactor. The Borax Expernments, 1953, AECD-3840, Argonne Nat ion al 1. abo ratory, February, 1954.

2. Dietrich, J. H., Expernmental Deterannatnons of the Self-flegulation and Safety of Operating Hater-Moderated lleactors, A/ Conf. 8/Y/43), Interna-tional Con fe renc - on the Yeaceful Uses of Atomic Energy, June 30, 1955.

3. Dietrich, J. H. , Experimen tal inves t nga t nan of the Self-Lims ta tion of Power During iteac t iv n ty Transien t s in a Subcooled, Kater-Moderated Reac tor. Borax-I Experiments, 1954, AECD-3668, Argonne National Labora-tory, August 17, 1955.

4. Lennox, D. ll. und C. N. Nelber, Summary Repor t on the Hazards of the Argonou r Reac tor, ANI.- 5 tit 7, De c e mbe r 1956.

III/A-7

APPENDIX III ARGONAUT SAFETY ANALYSIS REPORT (ASAR)

Attachment B Radiation Doses Resulting From Release of Fission Products into Atmosphere from UCLA TRAINING REACTOR HAZARDS ANALYSIS, Final Report.

R.D. MacLain, UCLA Department of Engineering Report #60-18, UCLA-NEL 2 March,1960 (there titled Appendix C)

III/B

RADIATION DOSES RESULTING FR05)

RLLEASE OF FISSION PRODUCTS Ih10 ATMOSPilERE Estin.ates have been rnade of the radiation dosares which would be received by persons outside the reactor building should there be a release of reactor fission products into the reactor luilding and leakage of the buildinF air to the outside. Eie radiation exposures considered here are those which would result from the passage of the air-borne cloud of radioactive contaminants over the ground. These include the external beta and Famma radiation exposures and the internal exposure of critical body organs resulting from inha-lation of the air-borne contaminants. The most irnportant of the internal exposures are the iodine dose to the thyroid and the strontium dose to the bones.

The radiation exposure received by a person standing at a given distance from the reactor building obviously depends on such f actors as (a) curies of fission products stored within the core at the time of release, (b) fraction of the core fission product s escaping into the building air, (c) building out-leakage rate, and (d) atmospheric dis-persive properties, llence , in t he analysis, certain basic asstrnptions are required as to the circumstances surrounding the release of the fission products, as to atrnospheric ron-ditions and, as to the tightness of the building at the time of release. The results ob-tained here are based on assumptions which, excelt for the arbitrary one that a relea.se has occurred, are considered reasonable for the reactor and building design. The calcula-tion method is described and illustrated in suf ficient detail that additional calculations based on other assumptions can be made i f desired.

The material presented here is divided into three sections. Tie first section de-scribes the model assuned for the release and spread of radioactivity and Fi ves the nec-essary re ferences and formulae used in calculating the radiation doses. The second section illustrates the calculation procecure. The third section present s the results obtained for the radiation eximure hazards with the assurred model.

Me t hod and Assuiryt ions used in Hose Calculations Al t hough such an esent is not considered even plausible because of the limitations on available excess reactivit y and because of the inherent sel f-l imi t ing c ha ract erist i cs of the reactor, it as postulated that an accident has occurred in whic h the reactor po er level has risen to the extent that local melting of the fuel plates has occurred. 1he reactor is assumed to have been operated continuously at t hc 10 L* power lesel lora enourb to have attained esjuilibrium concentrations of the relatisely short-lised fission produc t s, i.e., the iodine, brmine, and krypton isotopes. The incident is assumed to result in tia t rans fer of 1(f" of the volat ile fission products from the reactor fuel plates to the luild-ing air. It is assumed further that none of the nonvolatile fission product s are t ran-ferred to the bus]dany air although they may he released to the reactor coolant water and retained within the reactor building.

The foreFoing set of circumstances is consistent with the reasonable assurrption made here that the ancident is not violent enough to blow off the top ara side 1ioloFi cal shields so as to cause an intense spray of water-steam-radioactivit y mixture into the building air.

1he release of 105 of the volatile fission products is probably t oo high for the assumed incident but is used to gise an upper limit to the radiation exposure insolsel. lhe sula-tile fission product s are hrmine, krypton, iodine, and xenon, llence, the fission product chains which rmst be considered are of atomic masses 82 to 90 and 131 to 135. Iteierence 1 presents tables for each chain giving the equilibrium activity of each of the fission product III/B-1

isotopes in the chain and e t 3- v (or buildup) f ollo*ing react or s!.ut -down. 1he in fonna-tion presented in lie fere .c c 1 is used here for t he fi ssion product act ivit y r elease into the building air foll<, wine the assu w d incid-nt.

'lhe n.ost likely point at which radioactive contamination of the room air would be ,

detected is in the reactor rou, exhaust duct, since the air is pulled from the reactor room, and exhaust ed through the fan torna atop the buildinr. lhe air would not be considered con-taminated until ti.e ac tivity exc eeds that associated wit h ti e A4I nont. ally being discharged.

Upon detection of radioactivit y the air conditioning unit will be shut of f, and t he daapers in the inlet and outlet ventilation will le closed. (1he major avenues of leakage of the volatile fission products and daughter nongaseous products) froin the reactor room are the two high bay room ent rance ik> ors from the control room and the t hree emergency exit s in the reactor roon.. Access to the control roau is by way of an electrically controlled door f run the reception area in Engineering Unit 111. All access doors will be weather-st ripped and emergency doors leading directly to the out side, caulked and sealed f or n,inimum leakage.

1hese doors will be closed at all tines during reactor operation and any breeching will be indicated in the cont rol roon, on an audiovisual alarm system.

To obtain an upper bound for the radiat ion doses, the outleakage, I., in curies per hour of fission product activit y, is obtained f rom

1. i = - .I - . C ,m.x (g)

Q.

E and is assumed constant during the exposure t in e. In istuation (1), the tuilding

{g-- ,

leakage rate, has teen taken as 20"l of the r eac t or s uom s olunc, I!, per hour for a 30 MPil wind. 'lh i s leak-rate salue is assumed to he direc tly proportional to wind velocit y. 1he quantity C i ,m.x is ' he maximun actis v in curies of isotope ,i, prescnt outside the fuel plates following the assuned relet 10: of the solatile fission products. For most of the isotopes in the volatile f i .s s i o, c roduct chains, Ci ,mn is the activity of the isotope at the time of release from the fue plates. 'lhe impor tant exceptions are Sr-89 and Sr-90 and are forn.ed out side the react or and reach a maxinum art is i t y out side the reactor at some time after the fission product release. 1he tables in Re ference 1 permit easy calcu-lation of the activities of St-84 and Sr-90 at t ributable only to the decay of the isolated parent products.

The concent ration of fission product act isit y in the atne phere out side t he reactor building and the resultant radiation exposure will depend on the wind direction and velocity and the degree o f at niospl.er ic tur bulence. 'lhe highest dow rate is obtained when the person exposed is directly downwind f rom the leak. The comput at ion method is l ase d on O. G. but ton's fort:ula und utiliics erp.at ions and cur ves given in lle ferenc e 3.

For calculation of t he external beta dose and inhalation doses from the radioactive iodines an I st ront iun , the concent ration of act isit y in the at nu sph-r i c air was calculat ed by t he for mu l a in Heferera c 3,1%g 153 for ground-lesel continuou- emission of radio-act ivity, this formula reduces to l'

X - (2) 3600 nC2 u x2-n

where, A cont ent rat ion o f ac tivi ty, curies per cubic met er of air

1. :. continuous source strength, i.e., building out -leakage rate in curies per hr III/B-2

x = distance down.ind frorn source, n.eters u = rnean wind speed, n.eters per second C = gercralized di f fusion coe f ficient , neters n/2 n : dimensionless paranieter associated with atmospheric stability

'lhe following represent at ive salues of the di f fusion parawters f or two dif ferent at-nospheric conditions aie umed t o calculat e the concentration of activity, X, for a spets-fied leak rate and atmospheric condit ion in t he outside air for various distances, x, from the leakage sour ce.

Atenospheric Ccndition n C2 u severe inversion 0.5 .008 i mild lapse 0.25 .024 3

'lhe external heta dose rate during passage of the cloud of radioactive fission products is obtained from the following equation given in Beference 3, page 100.

Ir Dg ( 0. 5) (0.61) X. (3)

E' n.R x 1010 h3 m

where Xg is the concentration of A-enervy in Mev per ser per cubic r eter of air and Dj is the external heta dose rate in roentgens pe r set. 'The relation between Xg and X of Ihuation (2) is Ap : 3.7 x '910 AE (4) wl.ere E is the ef fective l eta energy in Mev per disint egrat ion.

The activity A deposited per seconil in (Le r rit ical organs is given by A - JFXo (5) where A - ac t ivity deposit ed in or g an, n.illiture /ser J : inhalat ion rat e, 17/601it ers per sec F, = inhaled f ra< t ion of art ivit y ret ained in critical orran The corresponding initial int ernal dose rat e for a per3on standity in the itssion product strean, is given by the expre.ssior.

D : A 3n00 tE ((, )

where, D : init ial internal cost rate rep / day t t irn. o f exposur e, l.r

% = weight of crit i< al organ, kg III/B-3

%e total integrated dose to the critical organ is related to the initial internal dose rate by the equation

'lli) = D 1.44 T (7)

where, TID = total int egrated dose, reps T = effective half-life of tie radioisotope, days.

He values of F,,, E, u, and T app, aring in Equations (5), (6), and (7) may be obtained from Iteference 4 for the various radioisotopes and c ritical organs insolved. Ik ference 5 gives additional in formation on t he various iodine isotopes.

For calculation of t he external gamma dose rate, t he J. Z. llolland non ogram given as Figure fi.3, lieferente 3 is used. 'lhe nonogram gises the gama dosage resulting f rom sudden discharge into the atmospheie of the content s of a nuclear reactor which has t een operat ing at a steady power le vel . The dosage read f rom the nomogram first nust be corrected to account for the fact that none of the nonvolat ile and only a fraction of t he volatile fission products are assumed to escape frce tiie reactor for the case being considered. Also, since the act ivity is not imediat ely released into t he at em phe re , but leaks out of tic luilding at a finite rat e, the dosage obt ained by use of the n(spogra1 nAlst be Converted to dose rate.

Le cc,rrections (r. tually scaling) applied to the salues obtained f rm the nomogram were calculated as follows:

The gants actit'ity of the vc l1t i le f i c t i cr. frodt. cts assumed to be esc 3 fir.g fr om the reactcs ua dete nir d b) uss cl t >. c cu*ter giten ir. Feference 6 for 20 min time after s t. u t - d c u:n . E: attan (s.1) c' Esference 3 is 1. sed to calcu-late the ganna actitity of all t h+ fission tsclucts for the same *ise after shut-Jcun ani t? c sanc rca:tcr t o :. <

leve l (1L: k). Fr c= t ?. i s t>.e fraction of t F.e total fission f sciuc t gamm1 activity attrib:.ted to the assumed 101 escafe of it.e volatile fission f r odt.c t s i.. dettesitri ar.1 t .c f a r *. 2 Jose re 2d from the ncsc-F gran is s cale d do.v. by t>.it f

a c t i:.n. To ot tain ti.e dosc v1tc resulting fr cs the finite rate of ratic2ctivity lesk into t hr aten hcrc. the ganna dose scaled frcs the nosagram st.lt it lied by t he quantity f/V.. t IIlustratIve Calculations Problem 1: Calculate the 1-131 dose to the thyroid of a person standine at a distance o f ol met ers do*nwinil o f t he leak for engl.t hours. 1)uring the exposure a severe insersion rondit ion exist s i n t he ut nepl e r e.

Solutlon: I rort lie f e rence 1, ti.e equilibrium curies of I-131 in the reactor fuel plates f ollowing 10 km operat ion is 4.307 x 10 I8 x M lO = 94. curie.,

1.3n x 10 I4 For a 30 MPil wind, li/\ r = 0.20 For a severe inversion u = 1 meter /sec o0 (2.25 MPil) so that lifl g = 0.20 x - '5 - .01 5. For 10N release of 3D l - 131 f r om t lie r e a t t o ,

L = .015 x (217 x .10) = 0.37 curiesda III/C-4

From Equation (2) y 2 x 0.37

= 1.7 x 10-5 curies 3600 n x .008 x 1 x 61 5 1 n,e t e r d

'lhe thyroid dose from I-131 is calculated from Equations (5), (6), and (7)

A = 17/60 x .15 x 1.7 x 10 5 a 7.2 x 10 7 millicurie/sec D =

7.2 x 10-7 x 3600 x 8 x 62 x .22 = 14 rep / day

. 02 0 TID = 14 x 1.44 x 7.7 = 155 rep The values of Fo, E, %, and T used in the calculations were obtained from fleference 4.

Problem 2: Calculate the external 4. dose from the I-131 isotope for a person standing at a distance of 61 noters downwind of tle leak for eight hours.

solu _t ion: For 1 = 1.7 x 10- 5 curies /m3 obtained in Problem 1, Xg : 3.7 x 1010 x 1,7 x 19-5 x 0.22 = 1.4 x 105 % v/sec m3 .

From Equation (3)

Dg = 0. 5 x 0.64 x I 4

195 6.8 x 1010 For eight-hour exposure, Tl D: 6.6 x 10-7 x 3600 x 8 = .019 r = 19 mr attributed to only the 1-131 isotope.

'Io oLt ain the total ex ternal 4-dose, t he sane procedure must be follcmed for all of the fission products assumed to le escaping from the reactor. For t he condi tions of t his prchl en, the air concentration of all of the fission products assu:wd to be escuping frun the reactor is 1.2 x 107 E v/see m3 , in which case the total external beta dose for eight-hour exposure is 1.6. Ik cause the decay of the fission products in the building and en-route t o t he person outside the building was neglected, the dose value calculated is higher than the actual value whic h would be obtained for the assumed conditions.

Proble g Calculate the total gama dose to a person standing 61 meters down*ind of the leak for eight hou r.s .

Fotution: birect use of t he nonogram (ikference 3) gives 12r for the total ga,a d ose caused by sudden release of tle total contents of the core into the utnosphere. Fron lieferenc e 6, the ganna activit y of 10% of t h<

solatile fission prr, luc ts is 6.9 x 1012 hv/sec at 20 minutes af ter shu t -dow:. . For all of the fission products at 20 minut es af t er shut-dcun, Equation (8.5) of tieference 3 gives a total gama activity of 5.2 x 16 M Ws hec . Ile nce , on the average, 10% of the volatile proiucts gives a garm.a activity equal to .01324 of the actisity of all the fis -

sion products. 'lhus, tic dose due to sudden release of 10" of the III/B-5

volatile products into the atnw> sphere i .s 12 x .01325 : . lf It>0 n :

Inawuch as the contaminant s leak out froin the I;uilding at a finite rat e (II/V, = .015 br-1),

the ganena dose rat.- is sinply 160 x .015 :

2.4 n1r/hr anil for an eight -hou r exposu re, t he. accunulat ed gamna dose is 2.4 2 8 19.2 rrr gnul ts o f Itadia t ion Esposure Criculations The results for eight-hour eximsure at four different distances downaind of the point of release under two different atnospheric conditions, calculated an illustrated alove, are taln-lated belo*. In all cases, l!G of only the solatile fission products are asswned to he re-leased fron. the react or fnel plat e . L building leakage rates an it/V : .015 br - 3 for the severe-inversion condi t ion and it/V, .015 hr-1 for the mild-lapse # condition (caused by di f ference in wind speeds assumed to l>e pr evailing for t he di f ferent atmospheric conditions).

TOTAL INTEGit ATED DOSI: (rer) litt0 M AN EIGilT-il0111t EXPost!ItE A T V Alt I DE.S D IST ANCE:S DouNutSD 1:11041 Iti: AC10lt ilUILDING Lt.AK Severe Intersion x, meters Eate rmi Gamma Thyroid Bone Beta Dose Do se Oore ' Dose 15 61 14.0 .080 1800 .006 152 1.6 .019 220 .0007 305 0.4 .010 59 .0002 0.15 .005 20 ---

Hild Lapse 15 2.2 .040 290 .001 61 0.19 .007 26 .0001 152 0.04 .004 , 6 ---

305 0.012 .002 2 ---

III/D-6

HEFERENCES

1. Faller, 1. L., 7. S. Chapman a nd J . M. Me s t , Calculations on U-235 Fassion Product Decay Chains, ANI.-4807, Argonne National 1.aboratory.

2. llea t ing Ven t i la t t ng A i r Condi t ion n ny Gu ide , American Society of lleating and Ventilating Engineers, 1958.

3. Meteorology and Atomic Energy, U. S. Depa r t men t o f Commerce, %enther ilureau, July, 1955

4. Massaun Permissible ,;nounts of Radnoisotopes nn the Human Body and Maximur.

Permissible Concentratnons in Air and Hater, Na t ional 13u reau o f St anda rds, llandbook 52.

5. Dunning, G. M., Thyroid Dose from Radioindsne in Fallout, Nucleonses, Vol. 14, p. 40, February, 1956.

6. Cl a rk, F.11. , Decay of Fa s s ion Produc t Gammas, NDA-27-39, Nuclear Develop-ment Associates, Inc., December, 1954.

"I/8-7

APPENDIX IV EMERGENCY RESPONSE PL A'l FOR THE UflIVERSITY OF CALIFORNIA AT LOS AfiGELES_

TRAINING REACTOR LICENSE f;0. R-71 DOCKET NO. 50-142_

February 1980

R R APPENDIX IV EMERGENCY RESPONSE PLAN _

CONTENTS Chapter 1 Emergency Response and Accident Assessment Organization and Procedures. . . . . . . . . . . . . . IV/1-1 IV/l-1 1.1 Types of Emergencies 1.2 Procedures During On-Duty (Working) Hours . . . . IV/1-2 1.3 Emergencies During Off-Duty Has . . . . . . . . IV/1-7 Chapter 2 Emergency Equipment. . . . . . . . . . . . . . . . . . IV/2-1 2.1 Radiological and Emergency Equipment Available for use on the UCLA Campus. . . . . . . . . . . . IV/2-1 2.2 Equipment Available . . . . . . . . . . . . . . . IV/2-1 2.2.1 Non-Portable Equipment . . . . . . . . . . IV/2-1 2.2.1 Portable Equipment . . . . . . . . . . . . IV/2-2 2.2.3 Respiratory Equipment. . . . . . . . . . . IV/2-2 2.2.4 Protective Clothing. . . . . . . . . . . . IV/2-2 2.2.5 Miscellaneous. . . . . . . . . . . . . . . IV/2-2 Chapter 3 Notification Methods . . . . . . . . . . . . . . . . . IV/3-1 Chapter 4 Notification Information . . . . < . . . . . . . . . . IV/4-1 Chapter 5 Emergency Response Training and Planning . . . . . . . IV/5-1 .

Chapter 6 Review, Updating, and Distribution of Emergency 'V/6-1 Response Pl an . . . . . . . . . . . . . . . . . . . . .

Chapter 7 Implementation of Emergency Response Plan. . . . . . . IV/7-1 Attachment A Emergsncy Procedure. . . . . . . . . . . . . . . . . IV/A-1 Attachment B Chart of Organizational Relations. . . . . . . . . . IV/B-1 Reactor Emergency Call List. . . . . . . . . . . . . IV/B-2 Attachment C Radiation Accident Procedure: UCLA Emergency Medicine Center. . . . . . . . . . . . . . . . . . . IV/C-1 C.1 Rackground. . . . . . . . . . . . . . . . . . IV/C-1 C.2 Definition of Radiation Radiation Accident Exposure . . . . Cases.

. . . . . IV/C-1. . . IV/C-1 4/C.2.1 4/C.2.2 Internal Contamination . . . . . . . IV/C-1 4/C.2.3 External Contamination . . . . . . . IV/C-1 4/C.2.4 Contaminated Wounds. . . . . . . . . IV/C-2 O IV/i

C.3 Referral of Radiation Accident Cases. . . . ..I /C-2 C4 Notification of Hospital Personnel. . . . . . IV/C-3 C.5 Transport of Radiation Accident Victims . . . IV/C-3 C.6 Management of the Contaminated Patient. . . . IV/C-3 C.7 Radiation Injuries not Involving Contamination . . . . . . . . . . . . . . . . IV/C-5 C.8 Specific Therapy for Internal Contamination . IV/C-5 Attachment D D.1 General Procedures in Radiation Accidents . . IV/D-1 0.2 Contaminated and/or Injured Personnel in Radiation Areas Without Assistance. . . . . . IV/D-1 IV/ii

APPENDIX IV EMERGENCY RESPONSE PLAN LIST OF FIGURES Figure IV/8-1 Chart of Organizational Relations . . . . . . . . . IV/B-1 LIST OF TABLES Table IV/D-1 Time Dose Table. . . . . . . . . . . . . . . . . . . IV/D-2 Table IV/D-2 Distance (d) versus Dose Table . . . . . . . . . . . IV/D-3 IV/iii

1.0 EMERGENCY RESPONSE AND ACCIDENT ASSESSJENT ORGANIZATION AND PROCEDURES The Nuclear Energy Laboratory (NEL) reactor emergency response team is composed of the following personnel in order of authority: Emergency Response director, Emergency response director alternate, NEL Director, NEL Laboratory Manager, Reactor Supervisor, Security Officer, Resident Health Physicist, Reactor Operators, and the NEL Technical Staff. Members of the Radiation Use Committee and the Campus Radiation Safety Committee are also available for consultation and review of energency action plans.

The UCLA Research and Occupational Safety Office provides a radiological backup (called the radiological monitoring team) for the NEL and the Campus Community Safety Police provide a security backup for NEL.

The "on-duty" operator is the reactor operator who was the last to sign the reactor operating log book. If the reactor is shut down, the "on-duty" operator will automatically be designated as one of the following personnel in line of authority and presence at the NEL facility: the reactor super-visor, the health physicist, the security officer, the laboratory mana-ger, the director, the designated licensed senior or reactor operator, the designated technical staff or radiation qualified backup staff member.

The duties and responsibilities of the organization personnel during declared or undeclared emergencies are specified in Chapter 1, paragraphs 2 and 3 of this emergency Plan.

1.1 TYPES OF EMERGENCIES 1.1.1 Any abnormal reactor operating situation as listed (but not limited to) in part VIII, section L of the USNRC-R-71 License Technical Specifications; or any situation which is an immediate threat to the safe operation of the reactor; or any fuel handling operations that have created a hazardous condition.

1.1.2 The presence of an increase in radiation levels or radioactive con-tamination from any source which is immediately threatening to the safety of the personnel occupying the reactor facility, the NEL, or the School of Engineering, or of immediate threat to personnel or the environment in the unrestricted area.

1.1.3 Fire or explosion.

1.1.4 Flood, windstorm, earthquake or other natural calamity.

1.1.5 Bomb threats, terrorist threats, riots, war, or other man-created dangerous situations.

1.1.6 Emergencies may also be declared at the discretion of the reactor supervisor, the health physicist, or any responsible staff member should any situation immediately hazardous to human safety develop.

IV/1-1

1.2 PROCEDURES DURING ON-DUTY (WORKING) HOURS 1.2.1 "The "on-duty" reactor operator shall manually SCRA't the reactor, if operating.

1.2.2 The "on-duty" reactor shall "03 SERVE" AND " EVALUATE" the situation and decide whether to call for assistance TCampuTphone dial 35) or to evacuate first and then call (35) for assistance (or not to call if the situation can be handled safely in-house). If an evacuation is called it is necessary to call for assistance (35). (I.E., 3 or more of 8 radiation alarms are sounding but the cause is unknown to the "on-duty" operator. Caller must give his name, location, phone and +"ne of emergency.

1.2.3 If necessary, based on the "on-duty" operators observations, he shall set off the evacuation alarm (identified switch at the center of the reactor control console).

1.2.4 The "on-duty" operator shall shut down the ventilation system which also automatically isolates the reactor building including all ducts if evacuating or if required by the emergency situation (i.e., high airborne radiation levels). (The emergency stop switch is located on the air control panel located at the far right of the reactor control console.)

1.2.5 The "on-duty" operator shall shut down the secondary cooling water systems if evacuating or required by the incident. (Switch marked

" secondary water" located on the reactor control console annunciator panel . )

1.2.6 The "on-duty" operator shall take the reactor key, the " health physics suitcase" (located on top of storage cabinets behind reactor control console), and any readily available radiation survey instru-ments with him to the assembly area (2000 level Boelter Hall patio, Palm Court) if evacuating. He should see that the facility doors are locked and secured behind him as he proceeds to the assembly area.

1.2.7 The "on-duty" reactor operator shall remain in charge until relieved by the reactor supervisor, the health physicist or a superior from the "on-duty" list (Chaoter 1.0). The "on-duty" operator shall not allow anyone to leave the assembly area until they have been checked for radioactive contamination. The "on-duty" operator shall assist arriving emergency teams and move non-occupational personnel out of the control area.

1.2.8 The acting reactor supervisor shall make a quick survey of the facility prior to proceeding to the evacuation area to ensure that the facility is cleared of personnel, that no one is lef t injured, and that the facility is secure from further entry.

1.2.9 Upcn arrival at the assembly area, the reactor supervisor and/or the health physicist shall immediately measure the radiation level IV/1-2

at the assembly area. If the area radiation level is greater than 2 mrem / hour, the personnel will be ordered to proceed to assembly area #2 (located at the northeast entrance to parking structure #9 near the public telephone booth). This same 2 mrem / hour radia-tion level at assembly area #1 will necessitate the evacuation of Boelter Hall and the Math Science Addition. Evacuation of these buildings will be initiated and carried out by the Campus Community Safety Police, under the guidance of the health physicist.

1.2.10 The health physicist shall check the personnel gathered at the assembly area for radioactive contamination, injuries, or other related problems. Any victims will be segregated and assigned a qualified monitor to assist with preliminary aid (decontami-nation, first aid, etc. ). The health physicist can now release non-contaminated and non-essential personnel .

1.2.11 If additional assistance is required the health physicist will dispatch a messenger. This messenger will either use the on-site public phone (change is located in the " health physics suitcase")

or go directly to the UCLA Campus Community Police Emergency Center (northwest corner of Westwood Blvd. and Circle Drive South) approxi-mately 1000 feet southwest of either assembly area.

1.2.12 The Campus Conmunity Safety Dispatch person receiving the "35" emergency call shall immediately respond by sending the required police, fire, paramedics and/or radiological assistance. The dispatcher shall use the Reactor Emergency Call List from the top down, and contact n_o non-University individuals until so

, instructed by the CCS police.

1.2.13 The necessity for evacuation of other campus buildings will be determined by the radiological monitoring team based on the ETA guidelines of potential 1 Rem total body dose and/or 5 Rem thyroid dose. Evacuation orders will be cleared with the Vice Chancellor's office (Administrative) prior to execution. Such evacuation will be carried out by the Campus Community Safety Police (thc CCS Police will seek outside assistance as they deem necessary based on the magnitude of the situation). The Radiological Monitoring team working directly with the police will advise them of unsafe areas and meterologic conditions which may affect the evacuation.

1.2.14 If, as determined by measurements, the incident appears to have the potential to extend beyond campus boundaries, the health physicist in charge will tell the police to notify the UCLA Police Watch Commander (at the Police Station), who in turn will notify as necess a ry :

A. The West Los Angeles Police Division Watch Commander, Purdue Avenue, W.L.A.

B. The Los Angeles County Sheriff Department.

C. The Los Angeles County Public Health Department, Radiological

Phone numbers on Reactor Emergency Call List.

IV/1-3

D. The Los Angeles Office of Civil Defense.

E. The Statewide Office of Emergency Services These agencies will assist in protection and/or evacuation of the surrounding population as necessary depending on the magnitude of the incident. Once the incident spreads beyond the campus boundary, it is out of the juris-diction of University of California personnel and subsequently they can act only as. technical advisors to the Local or State authorities. The decision to effect downwind evacuation of the population will be based on the lowest of the following ' EPA Guidelines: 1 rem integrated total body dose, or 5 rem thyroid dose, for directly exposed persons.

1.2.15 The reactor supervisor shall (as the emergency dictates) contact immediately USNRC, DOE, and the FBI (if required by 10 CFR, tech-nical specifications, blackmail, or bomb threat).

1.2.16 Injured personnel (not radioactively contaminated) will be trans-ported to the UCLA Emergency Hospital via UCLA emergency vehicles or local outside ambulances arriving at the assembly area after being dispatched by the Police Emergency Center. The health physi-cist will instruct the driver to deliver the patient to the Main Emergency hospital entrance and to relay the extent of the patient's radiation exposure (if known) to the doctor in charge.

l.2.17 Radioactively contaminated patients shall be tr ansported by the UCLA emergency vehicles or local outside ambulances to the UCLA emergency hospital special entrance for contaminated patients and autopsy rooms (located on Circle Drive South, east of School of Public Health entrance driveway, all the way back and up the ramp to the left). The Police Dispatch person shall inform the hospi-tal of the pntential arrival of radioactively contaminated persons as soon as the first "35" call for assistance is received.

NOTE: The UCLA Emergency Hospital is a Class "A" emergency hospital for the West Los Angeles area and is manned 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months a day, 7 days per week, and has approved procedures for dealing with radioactively contaminated victims (Appendix C).

1.2.18 The reactor supervisor will assign a technical staff member the task of keeping a written chronological history of the events that take place during and after the incident.

1.2.19 As soon as all critical aspects of the emergency are under control, the health physicist or his designee will collect the personnel monitoring devices (film badges, area badges, dosimeters, etc.)

and assign a technical staff member to deliver them to the Campus Radiation Safety Office for immediate processing. The Radiation Safety Office will transmit dosimetry data back to the health physi-cist as soon as possible, to allow the health physicist to evaluate the consequences of each individual exposure.

1.2.20 The health physicist in charge nr his designee (either NEL staff IV/1-4

or Radiation Safety Office Staff) will determine more precise radia-tion and radioactive contamination boundaries by radiation survey measurements in and around the Nuclear Energy Laboratory, Boelter Hall / Math Science Complex and the UCLA Campus. The health physi-cist will inform the police of those areas to which limited access is required. The boundaries will be established, as nearly as pos-sible, by the radiation values set forth in Title 10 Code of Federal Regulations, Part 20, and California Radiction Control Regulations, Title 17. (Radiation areas in which the dose could exceed 2 mrem in any one hour, or 100 mrem in any seven consecutive days will be declared " radiation" areas. Radiation areas in which the dose could exceed 100 mrem in any one hour will be declared "high radiation" areas. Areas in which removable radioactive material exceeds 1000 d/m per 100 cm 2 or 1 mrem / hour will be declared contaminated.

The following procedures will be used to guide the evaluation of the radiation conditions and will be carried out by a technical staff assigned by the health physicist:

A. The health physicist will specify the required staff, moni-toring, protective clothing and respiratory protection for the personnel involved in determining radiation conditions.

B. ' _diation surveys will be done by the appointed staff to find the existing levels of surface contamination and the existing intensities of radiation fields. These surveys will be conducted inside NEL, Boelter Hall, Math Science (as possible) and adjacent areas extending as far as required by the existing conditions.

C. Access to hazardous or poter.tially hazardous areas will be cordoned off ,*!th appropriate warning signs and barriers.

D. Data from on-site continuous monitors will be collected and evaluated.

E. Meteorological data will be collected from the UCLA Depart-ment of Atmospheric Sciences (7127 Math Science) or the National Weather Bureau, West Los Angeles, Office. These data will be used to determine the dispersion, and down-wind direction for further monitoring of any airborne effluent releases.

F. On-site and off-site swipes , soil samples , vegetation samples and water samples will be collected and analyzed, as needed, to aid in determining the nature and extent of the occur-rence. The collections will be made by the assigned staff during their surveys of the involved areas.

G. Approximately 100 area film packs (15% include neutron film), located inside and outside of NEL and Math Science will be collected by the health physicist and processed in-house. Data from these films will be utilized in the analysis of the history of the event.

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H. The health physicist or his designee will supervise any necessary decontamination activities and approve reentry and radiological clearance at the conclusion of such activities.

1.2.22 In an emergency life-saving situation, the health physicist may authorize a dose equivalent of 100 rem if the Campus Radiation Safety Officer cannot be reached. Otherwise, the Campus Radiation Safety Officer, Director of Research and Occupational Safety or the Chairman of the Radiation Safety Committee may authorize a life-saving dose equivalent to a maximum of 100 rem. For purposes of preventing the spread of the incident, a maximum dose equiva-lent of 25 rem may be authorized. Both dose equivalents to be given on a one-time per person lifetime basis only (no one will be allowed to voluntarily receive more than 100 rem under any cir-cumstances and volunteers over the age of 45 will be chosen if possible).

1.2.23 At the earliest possible time after the emergency the following personnel will meet to evaluate the cause and consequences, and to critique the handling of the entire situation. This group will then report to the Reactor Use Comnittee:

A. NEL Director B. NEL Laboratory Manager C. Reactor Supervisor D. NEL Security Officer E. Resident Health Physicist F. NEL technical staff involved G. Director of Research and Occupational Safety H. UCLA Radiological Safety Officer I. UCLA Campus Health Physicist The Radiation Use Committee will review and critique the events prior to submitting a written report to the UCLA Campus Radiation Safety Committee for final review.

1.2.24 A comprehensive written report of the emergency and subsequent action, including the Committee reviews, shall be prepared by the above-named personnel. The draft will be subject to the approval of the Radiation Use Committee prior to release to the Campus Radiation Safety Committee. Upon approval by the Campus Radiation Safety Committee, the report will be distributed to the proper regulatory agencies within the required time limits (refer to Chapter 3, Notification Methods, and Chapter 5, Notification Information in this Emergency Plan).

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1.3 EMERGENCIES DURING 0FF-DUTY HOURS 1.3.1 Emergencies of the types listed in Chapter 1, paragraph 1.1 of this plan will be detected by the Campus Community Safety Police Dispatch person either by phone or in person, or by one of the facility remote alarms (security, fire, radiation) located at the CCS dis-patch control center.

1.3.2 The dispatch person shall respond by contacting the proper team or teams (Police, Fire, Paramedics, Radiation Safety, etc.) as deter-mined from the incoming information.

1.3.3 The dispatch person shall then begin calling individuals (from the top down) on the updated Reactor Emergency Call List. The dis-patch person shall repeat the list until at least two University employees are contacted. The dispatch person shall NOT contact any non-University groups or persons listed until instructed to do so by a member of the emergency response team or the CCS police department.

1.3.4 As soon as a responsible person from the Reactor Emergency Call List arrives, the emergency will be handled as detailed in Chapter 1, paragraph 1.2 (Procedures During on-Duty Working Hours).

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2.0 EMERGENCY EQUIPMENT 2.1 RADIOLOGICAL AND EMERGENCY EQUIPMENT AVAILABLE FOR USE ON THE UCLA CAMPUS This equipment is stored in the following locations and is available on a reciprocal agreement basis for emergency situations.

A. The Nuclear Energy Laboratory and Fusion Engineering Project (2567 Boelter Hall, School of Engineering and Applied Science)

Phone 825-2040 B. Radiation Safety Office

( A6-060J Center for Health Sciences)

Phone 825-7147 C. Campus Comnunity Safety Police Department (112 Physical Plant Bldg. , Corner of Westwood Blvd. & Circle Drive South)

Phone 824-6413 D. Bio-Medical Cyclotron Facility (Center for Health Sciences)

Phone 825-6231 E. Laboratory of Nuclear Medicine and Radiation Biology--Health Physics and Safety (900 Veteran Avenue, West Los Angeles CA 90024)

Phone 825-8797 F. Radiological Sciences / Medical Physics

( AR-259 Center for Health Sciences)

Phone 825-7811 G. Physics Department (30171 Knudsen Hall)

Business Manager Phone 825-3224 Health Physics Phone 825-1515 2.2 EQUIPMENT AVAILABLE 2.2.1 NON-PORTABLE EQUIPMENT 2.2.1.1 Nal and GeLi multichannel gamma spectrometers 2.2.1.2 Liquid Scintillation automatic counting systems 2.2.1.3 Low background automatic windowless gas flow proportional analysis 2.2.1.4 Total body counter 2.2.1.5 Flow through radioactive gas analysis systems 2.2.1.6 Radiation film badge processing and readout capability IV/2-1

2.2.1.7 Multiple GM, proportional and scintillation counting systems for alpha, beta, gamma , and neutron analysis 2.2.1.8 Hand and Foot counters 2.2.2 PORTABLE EQUIPMENT 2.2.2.1 High range Beta, Gamma Survey instruments (10,000 R/hr) 2.2.2.2 Medium range Beta , Gamma Survey instruments (50 R/hr) 2.2.2.3 Low range Beta , Gamm3 Survey instruments (< .1 R/hr) 2.2.2.4 Neutron REM counters (Energy independent spheres) 2.2.2.5 Alpha Survey instruments 2.2.2.6 Pocket dosimeters (neutron and gamma) 2.2.2.7 Personnel film badge monitors (Beta , gamma , neutron) 2.2.2.8 Portable gaseous concentration monitors 2.2.3 RESPIRATORY EQUIPMENT 2 . 2 . 3 .1 Self contained air packs 2.2.3.2 Half face respirators with particulate canisters 2.2.3.3 Surgical masks 2 . 2. 3.4 Continuous air supplied i .astic suits (from building supplied air) 2.2.4 PROTECTIVE CLOTHING 2 . 2 . 4 .1 Coveralls (cloth and disposable) 2 . 2. 4. 2 Caps (and hoods) 2 . 2. 4. 3 Shoe covers and booties 2.2.4.4 Gloves (plastic, rubber and canvas) 2 . 2. 4. 5 Lab coats 2.2.5 MISCELLANE0US 2 . 2. 5.1 Personnel decon station.;

2 . 2.5. 2 High volume air samplers (particulate) 2 . 2.5. 3 First aid supplies IV/2-2

2. 2. 5. 4 On-site paramedic vehicle (UCLA Campus)

2. 2. 5. 5 Portable generator and lights 2.2.5.6 Pad waste drums (DOT approved) 2.2.5.7 Radiation filtered vacuum cleaners 2.2.5.8 Radiation approved hoods and glove boxes 2.2.5.9 Decontamination supplies (chemicals, rags, etc.)

2.2.5.10 Radiation Barriers (stantions , ropes , si gns , etc.)

Radiological protective and measuring equipment will be used when-ever its use will reasonably contribute to a reduction in an existing or projected radiological hazard. This equipment will be used by the radio-logical monitoring teams and can be found at the UCLA sites listed in Section 1 of Chapter 2.

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3.0 NOTIFICATION METHODS 3.1 Notification methods and procedures to be followed in the event of an emergency at the UCLA Nuclear Energy Laboratory (N.E.L.) Reactor are as follows" A. Title 10, Code of Federal Regulations, Part 20.403, " Notification of Incidents". and Sections V. E; and VIII, L2, M1, M2, M3 of the N.E.L. Reactor Technical Specifications and subsequent amendments outline requirements for reporting emergencies to the U.S. Nuclear Regulatory Commission. Notification procedures (e.g. telephone, letter, etc.) will be implemented as required in these documents.

The name, address, and telephone number for the NRC office to be contacted are as follows:

U.S. Nuclear Regulatory Commission Region V Phone: (415) 932-8300 Office of Inspection and Enforcement 1990 N. California Blvd.

Walnut Creek Plaza. Suite 202, W/l nut Creek, CA 94696 B. All communications to the USNRC or the State of California shall proceed through the UCLA Office of Research and Occupational Safety except for requests for immediate radiological emergency response assistance (Title 10, Code of Federal Regulations), requests for security, assistance, and requests for interpretation of N.E.L. Reactor License requi remen ts . In all cases copies or notification of such conmunication shall be furnished to the Radiation Safety office in a timely manner.

UCLA Office of Research and Occupational Safety UCLA Campus Community Safety Police Station U.C.L. A. , Los Angeles, CA 90024 Phone: AC (213) 825-7141 (days)

AC (213) 825-1491,2,3, (24 hrs)

C. Notification of Activation of the NEL Emergency Response Plan along with notification information (Chapter 4, this plan) will be fur-nished to the Nuclear Energy Liability-Property Insurance Association (ANI) in a timely manner dictated by the urgency of the incident at hand. The name, address, and phone numbers are as follows:

The Exchange AC (203) 677-7715 (days)

Farmington i se AC (203) 677-7305 (24 hrs)

Farmington, Cc aecticut 06032 D. The Emergency Response Director or his alternate will, at his dis-cretion, release through the UCLA Public Information Office the events and their associated details to the news media. Those indi-viduals directly involved will confer only with the Director about the details to be released and by no means release unsanctioned IV/3-1

data to the news media themselves.

3.2 The UCLA-NEL Reactor Administrative Procedures incorporate a 24-hour hour day (7 days) Reactor Emergency Call List of key personnel. In the event of an NEL emergency during off-duty hours, the UCLA Campus Community Safety Information Center will begin to call individuals on this list until at least one is contacted. The dispatcher will then infonn this person of the nature of the alarm or observed abnormal condition associated with the NEL facility. The alarm systems (security, fire, radiation, etc.) are oper-able 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months per day and are under continuous surveillance at the manned CCS Information Center during off-duty hours. In addition there are periodic Security foot patrol checks of the site and surrounding area on a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months basis. Upon notification of an individual on the Reactor Emergency Call List *, the situation will be promptly evaluated by the responding personnel and appropriate emergency notification will be carried out in accordance with Chapter 1 of this plan.

All phone numbers appear in Attachment B.

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4.0 NOTIFICATION INFORMATION 4.1 Upon calling the local emergency number 35, request the necessary emergency response organization (police, fire, etc.) and give the following information as appropriate and available:

A. Your name, your phone number, and time of occurrence of the problem.

B. A statement of the problem, incident or accident along with a descrip-tion and location on campus.

C. Personnel status with respect to injury, radiation and contami-nation hazard, toxic, explosive, etc.

D. The exact location at which you will meet the incoming assistance team.

E. Estimate of quantity of radioactive material released, or being released and height of release, if possible.

F. Estimate of chemical and physical form of the release, if possible.

G. Estimate of prevailing weat' ar (wind velocity, direction, tempera-ture, atmospheric stability, etc., or call UCLA Weather 825-1217 or 825-1653 (Math Science Complex--Room 7127MS) or call the National Weather Bureau W.L. A. 554-1212, if possible.

H. Estimate (if applicable) of projected dose at the possible points of maximum concentration in unrestricted areas, and potential con-trolled area boundaries.

I. Estimate of radiation and radioactive contamination in restricted and unrestricted areas, if possible.

J. Emergency response procedures currently in effect, if any.

K. Further emergency actions, or assistance needed or recommended including:

1. Fire, L. A. City Fire Department or L. A. County Fire Dept.

2. Paramedics and ambulance, same as Fire Department

3. UCLA Med Center physicians if a large number of multiple injuries have occurred.

4. UCLA police for communication assistance and/or assistance in securing and controlling the area.

5. The UCLA Radiation Safety Officer, Health Physicists and Radiological Monitors

6. Office of Research and Occupational Safety for fire, toxic exposure evaluation, explosive assistance IV/4-1

7. UCLA facilities for emergency maintenance (plumbers, elec-tricians,etc.)

8. Los Angeles Police for large-scale backup

9. Los Angeles County Sheriff if more assistance is needed

10. Los Angeles County Health Dept. Radiological Division 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months radiological assistance

11. DOE

12. California Statewide Radiological Assistance

13. NRC Regio'i V if Special Nuclear Material is missing

14. FBI if Special Nuclear Material is missing, or security violation.

15. American Nuclear Insurers IV/4-2

5. 0 EMERGENCY RESPONSE TRAINING AND PLANNING 5.1 Appropriate emergency training for the radiological operations and assessment teams will be conducted annually during the reactor operator requalification program. New team members will be trained during their initial radiological qualification. All Police and Security Officers attend a retraining program annually.

5. 2 Evacuation Drills will be conducted semiannually in accordance with the Technical Specifications.

5. 3 On an annual basis the Reactor Supervisor, Security Officer, or the Health Physicist will obtain confirmation from the outside emergency response agencies (fire, police, paramedics, hospital, etc.) that they are up to date in their ability to deal with radiation, radioactive con-tamination and injury involving radiation and radioactive contamination either on or off-site.

IV/5-1

6.0 REVIEW, UPDATING, AND DISTRIBUTION OF EMERGENCY RESPONSE PLAN 6.1 The NEL Emergency Response Plan will be reviewed and updated annually by the Reactor Use Conmittee.

6,2 The NEL Emergency Response Plan and all revisions will be distri-buted to pertinent federal, state and local regulatory and emergency response organizations as required.

IV/6-1

7.0 IllPLEMENTATION Of EMERGENCY RESPONSE PLAN 7.1 The UCLA Nuclear Energy Laboratory Staff will cooperate with all federal, state and local regulatory and emergency response organizations for the purpose of implementation of this emergency plan.

7.2 Implementation will incorporate appropriate training and drills as listed in Chapter 5 of the plan and the associated technical specifi-cations for the reactor.

7.3 The UCLA Nuclear Energy Staff will provide annual training or familiari-zation to federal, state or local emergency response teams as required with regard to plant physical characteristics, plant operations, radio-logical procedures, normal release of radioactive materials and environ-mental controls. This training will be provided for the following groups:

A. In-house employees , faculty and students.

B. Campus Community Safety / Police personnel.

C. Local fire and paramedics as required.

D. Other involved agencies (Sheriffs, etc.), as required.

G IV/7-1

ATTACHMENT 4/A EMERGENCY PROCEDURE 4/A.1 Upon observance of danger from flood, earthquake, riot, terrorist threat, enemy attack, release of toxic substance, police problem, power outage or other similar calamity immediately do the following:

A. shut down reactor B. secure facility C. assemble staff in control room, if possible D. call "35" and advise of situation E. follow the emergency plan 4/A.2 Upon observance of danger from nuclear hazard *, fire, or explosion, immediately proceed to IV. A.5.

4/A.3 Upon notification of a bomb threat to either the reactor, the laboratory, or the School of Engineering, immediately evacuate the building, call "35" and notify UCLA police.

4/A.4 If alarm in 1540 Boelter Hall (subcritical and radioactive storage) is sounding, evacuate area and proceed directly to 4/A5.

4/ A. 5 Proceed to:

A. Manually scram the reactor, if operating B. observe and evaluate the situation, call "35" if necessary C. set off evacuation alarm and announce, if appropriate D. shut off ventilation and water systems, if appropriate E. take reactor key, health physics suitcase and any immediately available monitoring instruments with you, if appropriate F. proceed directly to Boelter Hall Patio (2000 Level Palm Court),

" securing" all doors behind you G. af ter gross personnel contamination check , follow the emer-gency response plan (in suitcase)

H. call official personnel (or have "35" call) listed on Reactor Emergency

Nuclear hazard is defined as any abnormal reactor operating situation, 3 out of 8 radiation monitors are off-scale high from an undetermined cause (4 area monitors, A-41, Effluent, High Stack Radiation monitor, Hand &

Shoe monitor), or deemed necessary by the person in charge.

IV/A-1

Call List and remain in area until relieved by staff personnel I. assist emergency response teams (fire, police, paramedics) where possible J. If the radiation at the 2000 level Palm Court assembly and control point is >2 mr/hr, move assembly area and control point to northeast corner of parking structure #9 at the phone booth. This will now be considered the command post.

K. If the control point is moved, notify CCS police to evacuate the Boelter Hall and Math Science Addition structures to parking structure #9, 4th level.

L. If determined by health physics monitoring teams appointed by the health physicist in charge, the accident is likely to spread beyond the campus boundary (downwind), the health physicist in charge should inform the CCS police watch com-mander who in turn will contact (as determined by the mag-nitude of the accident and the judgment of the Emergency Response Director, with input from health physicist in charge):

1. the local fire department

2. the local Los Angeles police department

3. the Los Angeles County Health Department, Radiological Division

4. the Los Angeles County Sheriffs Department

5. the Los Angeles County Civil Defense

6. the California Office of Emergency Services, Sacramento.

IV/A-2

BOARD OF REGENTS (LICENSEE)

U UCLA CHANCELLOR EXECUTIVE ADMIT 4ISTRATIVE ASSOCIATE VICC CHANCELLOR VICE CH44CELLOR VICE CtWJCELLCR (ACADEMIC) RESPONSIBLE OFFICER (RESEARCH)

I l V V oc^", sc = ' or LEVEL 2 ENGINEERING 4 0

^ss i s t"4' VICE CHANCELLOR U

APPLIED SCIENCE COrtvilTY StJETY REACTOR USE DIRECTOR ff) CLEAR ~

CO'fil TTEE ENERGYLkBORATORY SAFETf CortilTTEE

- EMERGENCY RESPONSE AUON D W_ W F#4AGER, NUCLEAR S FETf - EMERGENCY RESP 0r:SE ENERGY LABORATORY OFF1FFR DIRECTOR, ALTER! ATE

,I ,1 REACTOR

" "T H OPERATIONS " " " " "

LEVELS 3

&4 Fl/JCTIOr:AL RESPONSIBILITIES EXPERIMENT /i USE APPROVAL STAFFING OVEPfiL SAFETY BUDGET NRC CO'itJN! CATIONS TECHNICAL CHN6ES UNIVERSITY PUBLIC PREPARE LICENSE A'C D CNTS ORGANIZATIONAL RELATIONS FIGURE IV/B-1 IV/B-1

ATTACHMENT B REACTOR EMERGENCY CALL LIST NAME EMERGENCY "35" PHONE University

1. Chuck Ashbaugh Home 348-6559

2. Jack Hornor 839-2130

3. Maurice Thelia 837-2340

4. Neill Ostrander Office f2 342-8969

5. Ron Bolek 785-2620

6. Ivan Catton 887-7295

7. Tony Zane 1-714-549-549-3695 Radiation Safety Office

8. John Kaufmann Office 825-6900 Home 645-1849

9. John Evraets 825-7147 475-1358

10. Walter Wegst 824-6413 355-3089

11. UCLA CCS Police Dept. 825-1491 35 Emergency

12. John Barber 825-1633 822-0715

13. James W. Hobson 825-2411 475-5195

14. Charles E. Young 825-2151 825-3353 Non-University

1. L.A. Fire Dept. 473-1155 35 Emergency

2. Ambulance 825-2111 35 Emergency

3. Emergency Medical Center 825-2111 35 Emergency

4. UCLA Physical Plant 825-1391 35 Emergency

5. L.A. Police Dept. (WLA) 478-0841 35 Emergency

6. L. A. Sheriff Dept. (WLA) 456-6652

7. NRC Region V l-415-932-8300 Security--LeRoy R. Norderhaug Operations--Robert Engleken

8. DOE (ERDA) Safety & SNM Div.--Manley Wu l-415-932-8300

9. Calif. State Radiological Asst. Pgrm. 24 hr 1-916-391-7716

10. Calif. Office of Emergency Services 1-916-421-4990 Duty Officer--24 hrs

11. L.A. County Health, Radiation Div.--24 hrs 974-1234 Joe Karbus, Radiation Chief

12. California State Police day 620-5607 24 hr 620-4700

13. FBI /WLA 477-6565

14. NEL-PIA Nuclear Liability / Property Insurance Association, the Exchange, Bldg. 3, Loft A, Suite 323, Farmington, CT 06032 1-203-677-7715

15. Weather report UCLA local campus 825-1217, 825-1954

16. Weather report National Weather Bureau WLA 554-1212

17. WLA Police Dept. Watch Commander 478-0731 IV/B-2

ATTACHMENT C RADIATION ACCIDENT PROCEDURE: UCLA EMERGENCY MEDICINE CENTER C.l BACKGROUND Though remote, the likelihood of a radiation accident has increased sharply in recent years because of the expansion of the nuclear energy industry, and the now widespread use of radionucleides in basic and life sciences research and in medical diagnostics. The UCLA Emergency Medicine Center is a designated provider of emergency medical services for individuals involved in radiation accidents Two general classes of such accidents can be defined:

A. those in which contamination of the individual results in poten-tial hazard to those who transport him to the Emergency Medicine Center and those who provide his subsequent care B. those individuals who have received radiation injury, but who are not contaminated and pose no radiation threat to those who care for them.

These two general classes of radiation accident victims require different management; those who are contaminated must first be decontaminated in an area physically separate from the Emergency Medicine Center. The following is an outline of the procedures to be followed by Emergency Medicine Center personnel in the event of a radiation accident.

C.2 DEFINITION OF RADI ATION ACCIDENT CASES There are four types of radiation accident patients:

C.2.1 Radiation Exposure The individual who has received whole or partial body radiation and may have received a lethal dose of radiation, but is no hazard to attendants, other patients or the environment. He is no different than the radiation therapy or diagnostic x-ray patient.

C.2.2 Internal Contamination Such contamination results from inhalation or ingestion of radioactive material. Such a person is no hazard to attendants, other patients. or the environment. Following cleansing of minor amounts of contaminated material deposited on the body from airborne exposure, this person should be handled similarly to a case involving exposure to a chemical poison such as lead. His body wastes should be cellected and saved in order that measurements of amount of radioactive materials present can be made to assist in determining appropriate therapy.

C.2.3 External Contamination External contamination of body surface and/or clothing by radioactive

Adopted from: Emergency Handling of Radiation Accident Cases. Uss, Atomic Energy Commission and American Medical Association, '969.

IV/C-1

liquids or by dirt particles presents a type of case with protlems simi-lar to an infectious disease. Isolation techniques to protect other patients and the hospital environment must be employed in order to confine and remove any potential hazard.

C.2.4 Contaminated Wounds When external contamination is complicated by a wound, care must be taken not to cross-contaminate surrounding surfaces from the wound and vice versa.

The wound and surrounding surfaces are cleansed separately and sealed off when clean.

These radiation and/or contamination problems may or may not be complicated by serious injury.

C.3 REFERRAL OF RADI ATION ACCIDENT CASES As soon as a suspected radiation accident has occurred, the Emergency Medicine Center should be notified by telephone. The individual who receives the call must obtain as much information as possible about the nature of the accident and make a decision as to whether the patient is to be seen in the Emergency Medicine Center or in a designated decontamination site with-in the UCLA Center for the Health Sciences. Ideally, all patients should be reported by telephone prior to their arrival at the Emergency Medicine Center. This will minimize the risk of contamination of the Emergency Medicine Center, which would result in its being closed. The information to be obtained by telephone should include the following:

A. identification and affiliation of person making the call, and tele-phone number B. number of persons to be admitted and suspected of having:

1. injury but no radiation exposure or contamination

2. radiation exposure

3. internal contamination

4. external contamination

5. contaminated wounds C. identification of patient (s), if known D. nature of accident, radiation, or radioactivity source, if known E. location, name, type of facility at which accident occurred F. persons in charge of radiation evaluation IV/C-2

G. whether or not the patients will be:

1. surveyed for contaaination ii. decontaminated before arrival at hospital H. expected time of arrival at hospital C.4 NOTIFICATION OF HOSPITAL PERSnNNEL In the event of a reported radiation accident, the following hospital per-sonnel should be immediately notified:

A. the emergency medicine attending physician on call B. the health physicist on call C. the hospital administrator on call C.5 TRANSPORT OF RADIATION ACCIDENT VICTIMS Prior to transport of the radiation accident victims to UCLA Hospital, the Emergency Medicine Attending Physician in consultation with the Health Physicist on call should determine if the accident involves possible con-tamination. If there is no contamination of the accident victim, he shall be transported directly to the Emergency Medicine Center. If the poten-tial of contamination exists, he should be decontaminated at the site of the exposure prior to transport, if possible. This decontamination should include the following:

C.5.1 Removal of any possible contaminated clothing and placing of this clothing in a plastic bag to be brought with the patient to the decontami-nation area within the UCLA Medical Center C.S.2 Thorough washing of the contaminated area with a mild soap and water and rinsing of this into a large basin if possible, and if not, into a large holding tank or into city or county water drainage system. The victim should then be reclothed in uncontaminated apparel and be transported by ambulance, in the case of possible significant contamination, to the decontamination area. All those who assist in this transport should use isolation technique, including mask, gloves and protective clothing so that they do not contaminate themselves. The Emergency Physician coordinating the radiation accident will designate the decontamination area to which the patient is to be transported. In the event of serious, life-threatening injury in addition to possible contamination, the patient will be trans-ported immediately to the decontamination area without prior decontamination.

C.6 MANAGEMENT OF THE CONTAMINATED PATIENT The possibly contaminated patient will be directed to a decontamination area within the medical center. Prior to his arrival the area will be closed to hospital personnel not directly involved in the decontamination procedure. Absorbent paper shall be taped to the floor from the site of the patient's entrance into the hospital to the decontamination area. In IV/C-3

addition, in cases that may involve contamination with radioactive dust, the ventilation system of the area will be closed. The decontamination area must contain a table which lends itself to washing with water, a drain in the floor, and if possible, wall and floor coverings that are easily decontaminated. Prior to the patient's arrival, a survey meter should be obtained to determine the extent of contamination. Upon the patient's arrival, the following steps should be taken:

A. The patient should be checked for contamination by use of a survey meter.

B. If seriously injured, immediate emergency life-saving assistance must be provided. This will be directed by the Emergency Physician in charge.

C. All personnel involved in patient care should be attired as they would be for a surgical procedure. They should wear gowns, gloves, cap, mask, shoe coverings, etc.

D. If possible external contamination is involved, all clothing of the victim and those in contact with him and bedding from the ambu-lance should be saved. In addition, all blood, urine, stool, vomi-tus and all metal objects in contact with the patient should be saved. These should be labeled with the patient's name, body loca-tion, time, and date. They should be saved in an appropriate con-tainer; each container should be marked clearly " Radioactive:

Do Not Discard". All clothing is to be saved in plastic bags.

E. Decontamination should start, if medical status permits, with cleansing and scrubbing of the area of highest contamination first. If an extremity alone is involved, clothing may serve as an effective barrier a.d the affected area alone may be scrubbed and cleansed.

Initial cleansing should be done with soap and water. Wash water waste, unless markedly radioactive, may be flushed into community sewage systems where dilution will obviate any hazard effect.

Special care must be paid to hair parts body orifices, and body fold eas. Remeasure and record measurements of radioactivity a f te ach washing. If a wound is involved, prepare and cover the wound with self-adhering disposable surgical drapes. Remove wound covering and irrigate wound with sterile water, catching the the irrigating fluid in a basin or can and mark and handle as described in (D) above. Each step in the decontamination should be preceded and followed by monitoring and recording of the extent of the contamination.

F. Clothing of all physicians, nurses and attendants should be saved as described for the patients. Nurses, doctors and attendants must follow the same monitoring and decontamination routine as patients.

G. Debridement for gross contamination: The physician in the Emergency Room confronted with a grossly contaminated wound with dirt particles IV/C-4

and crushed tissue should be prepared to do a preliminary wet debridement. An emergency minor surgical set should be used.

Further measurements may necessitate a sophisticated wound counting detection instrument supplied by the consultant who will advise if further definitive debridement is necessary.

H. Internal contamination: If internal contamination is suspected after external decontamination, request a whole body count as soon as the patient's condition permits.

I. Upon completion of decontamination, the patient should be handled according to the following guidelines :

1. decontaminated and no injuries requiring hospitalization--

discharge

2. decontaminated and injured--admit to the appropriate hospital service

3. irradiated--admit to Intensive Care

4. serious radiation exposure, serious internal contamination and/or external and wound contamination not responsive to decontamination--admit to a nursing floor with special con-tamination control procedures C .7 RADIATION INJURIES NOT INVOLVING CONTAMINATION In the event nf a radution injury not involving contamination, it is neces-sary to estimate the amount of radiation the victim has sustained. The attending emergency physician and health safety physicist will estimate this and determine what further steps may be necessary. Partial body ex-posure, if significant, will result in a clinical picture which consists of erythema, edema, vesiculation and, with high doses, tissue necrosis.

Symptems secondary to whole body exposure with x or gamma radiation are dependent on the dose. With low doses, i.e. less than 100 rads, these signs will be minimal . With very high doses, i .e. over 1,000 rads, the clini-cal picture is entirely determined by extensive anatomical and functional damage of the gastrointestinal tract and hematopoietic system. Systems will include severe and practically irreversible disturbances of the electro-lyte equilibrium in the body, generalized weakness, decreased resistance to infection, and increased bleeding tendencies. Patients who may have received high doses of radiation should be admitted to an isolation unit.

Initial laboratory studies should include a complete blood count with a platelet count and serum electrolytes. IV fluid therapy should be instituted.

C.8 SPECIFIC THERAPY FOR INTERNAL CONTAMINATION It is important to recognize that some forms of internal contamination (e.g., plutonium) should be treated with specific pharmacologic agents to minimize the deposition of radioactive material in bone. It is imperative to contact the Health Physicist and a physician trained in radiation acci-dents in the event of these and all other radiation accidents immediately.

IV/C-5

In any case, it is to be understood that this Radiation Accident Procedure is a guideline that is to be used on conjunction with the expertise pro-vided by such personnel.

ADDEflDUM Radiation Safety Office personnel to be contacted in the event of a radiation accident procedure: during working hours, call 57147; after hours call:

A. Mr. John Evraets, 475-1358 B. Mr. John Kaufman, 645-1849 C. Mr. Jack Hornor, 839-2130 D. Dr. Walter Wegst, 355-3089 IV/C-6

ATTACHMENT D D.1 GEN 5RAL PROCEDURES IN RADIATION ACCIDENTS E - evacuate local area immediately M - monitor personnel and areas, if possible F - evaluate situation R - restrict and/or secure areas G - get assistance (as soon as possible)

E - employ routine radiation safety procedures N - notify authorities C - clean up and shield where possible Y - ye'.1 for hel p. DON'T PANIC OR LEAVE BEFORE HELP ARRIVES D.2 CONTAMINATED AND/OR INJURED PERSONNEL IN RADIATION AREAS WITHOUT ASSISTANCE A. after a prompt situation evaluation (i.e. , extent of injury, radia-tion rates, time before discovery, and extent of contamination problems)

B. drag on blanket or support away from field C. strip down and isolate clothes D. administer serious first aid problems E. decon personnel and self, monitor F. isolate area and seek help (call "35")

G, follow the emergency plan IV/D-1

Table IV/D-1 TIME DOSE TABLE Dose Rate Time to Culumative Dose Of R/hr 25R 50R 100R 200R 450r 25 1 hr 2 hrs 4 hrs 8 hrs 18 hrs 50 30 min 1 hr 2 hrs 4 hrs 9 hr, 100 15 min 30 min 1 hr 2 hrs 4.5 hrs 500 3 nin 6 min 12 min 24 min 54 min 1,000 1.5 min 3 min 6 min 12 min 27 min 5,000 18 sec 36 sec 72 sec 2.4 min 5.4 min 10,000 9 sec 18 sec 36 sec 72 sec 2.7 min (1.2 min)

Max dose Max dose Radiation LD 50/30 other than life or sickness Po s ble life and death or '

probable d death ,

si tua tion 50% exposed si tua tion personnel DOSE RATE AT THE MONITOR IV/D-2

Table IV/D-2 DISTANCE (d) VERSUS DOSE TABLE d If You Read at d Feet Fraction R/hr 2 .25 25 125 250 1250 5 .04 4 20 40 200 10 .01 1 5 10 50 20 .0025 .25 1.25 2.5 12.5 50 .0004 .04 .2 .4 2 100 .0001 .01 .05 .1 .5 Dose Rate at 1 Foot is 100R/hr 500R/hr 1000R/hr 5000R/hr

)

< v a

< d )

monitor source DOSE RATE AT THE OBJECTIVE IV/D-3

/ TNDIX V TECHNICAL SPECIFICATIONS FOR THE UNIVERSITY OF CALIFORNIA AT LOS ANGELES T_ RAINING REACTOR LICENSE NO. R-71 DOCKET NO. 50-142 February 1980

APPENDIX V TECHNICAL SPECIFICATIONS FORWARD The Technical Specifications contained in this appendix, embody the earlier Technical Specifications (of 1971 as amended in 1976), in revised format and expanded content. With four exceptions noted below, no attempt has been made to alter the content and provisions of the earlier Technical Specifications, and any other discrepancies should be interpreted as typographical errors or editorial deficiencies.

The exceptions are:

(a) the name of the Radiation Use Committee is changed to Reactor Use Committee; (b) the concept of Inhibit is broadened by definition; (c) Safety System Margin, not previously defined, is now defined; and (d) the existing Technical Specifications requires the reporting of uncontrolled or unanticipated changes in reactivity. We propose to limit this reporting requirement to uncontrolled positive reactivity changes that lead the system to, or beyond, a safety system trip point.

Exception (a) is designed to conform with a commonly used, but heretofore erroneous, name. There has been some confusion in referencing the Radiation Use Committee versus the Radiation Safety Committee. The latter is of campus wide representation, and the name Reactor Use Committee is more appropriate to a committee having review and audit duties that are specifically reactor related.

Exception (b) arises from an unreviewed and umimplemented plan to alter the system response to " Inhibit". The Reactor Use Committee has recommended the investigation of ways to improve the inhibit response, a plan exists, but review and approval are unlikely within the filing time of the present application.

Safety System Margins, representing the distinction between safety limits and trip-points, have existed for some system parameters, for the life of the UCLA Reactor. No margins were ever alloted to the power controls, and margins are not explicitly stated in the existing Technical Specifications. The specifications that follow, attempt to clarify this area, and to add a margin above the over-power trip point (Section 2.1.2.c[2]). Reactor period is not recorded, and on period scrams, the actual period is unknown. Thus a Period Safety Margin is not meaningful, the trip point and the Safe'.y Limit are treated as identical.

V/i

Some of the design bases for set-points and trip-points are historical at UCLA, and not easily rationalized. For example, there is no clear reason why the primary coolant (water) flow rate should equal or exceed 10 gpm at all power levels from one watt to 100 kilowatts. Operating experience shows that this flow rate is adequate, in that fuel surface temperature will not exceed 200 F.

This condition could be realized with other flow rates, particularly at reduced power level. Thus, although there is a temptation to liberalize this specification on rational grounds, we have chosen to continue the same specification without knowing the reasons for that particular choice. In the text, unknown and non-rationalized bases will be identified by the words " operating experience".

Exception D concerns the existing Specification L.3.6 that presents, as an example of an abnormal occurence, "an uncontrolled or unanticipatcd reactivity change". Anticipation is undefined. The operator does not anticipate each and every correction made by the auto-controller, nor does he anticipate an imminent power failure.

Thus reactivity changes occur and are controlled with expectation but without anticipation. The phrase is replaced in the proposed Technical Specifications, 6.5.2.1.B.2, with "an uncontroiled positive reactivity change that leads the system to, or beyond, a safety system trip point". This definition accepts as normal:

(a) rod-drops with loss of magnet or system power; (b) the reactivity changes that are made by the controller; (c) the possible inhibit response suggested in Exception B; and (d) the controlling function of an operator who can perceive and correct the situation prior to passage to the trip point.

V/ii

APPENDIX Y TECHNICAL SPECIFICATIONS CONTENTS Fo rwa rd . . . . . . . . . . . . . . . . . . . . . . . . . . V/ i Chapter 1 Definitions . . . . . . . . . . . . . . . . . .

1.1 Sa fety Channel . . . . . . . . . . . . . . V/1-1 1.2 Reactor Safety System. . . . . . . . . . . V/1-1 1.3 Operabl e . . . . . . . . . . . . . . . . . V/1 -1 1.4 Channel Check. . . . . . . . . . . . . . . V/1-1 1.5 Channel Tes t . . . . . . . . . . . . . . . V/1 -1 1.6 Channel Calibra tion. . . . . . . . . . . . V/1-1 1.7 Unscheduled Shutdon . . . . . . . . . . . V/1-1 1.8 Reactor Shutdown . . . . . . . . . . . . . V/1-1 1.9 Reactor Operating. . . . . . . . . . . . . V/l-2 1.10 Reactor Secured. . . . . . . . . . . . . . V/1-2 1.11 Measuring Channel . . . . . . . . . . . . . V/1-2 1.12 Reportable Occurrence. . . . . . . . . . . V/1-2 1.13 An Experiment. . . . . . . . . . . . . . . V/1-2 1.14 Experiment Facilities. . . . . . . . . . . V/1-3 1.15 Control Rod. . . . . . . . . . . . . . . . V/1-3 1.16 Readily Available on Call . . . . . . . . . V/1-3 1.17 Rod Drop Time . . . . . . . . . . . . . . V/l-3 1.18 Drop-Rod Scram . . . . . . . . . . . . . . V/1-3 1.19 Full Scram . . . . . . . . . . . . . . . . V/1-4 1.20 Inh 1M t. . . . . . . . . . . . . . . . . . V/1-4 1.21 Sa fety L !mi ts. . . . . . . . . . . . . . . V/1 -4 1.22 Safety System Margin . . . . . . . . . . . V/1-4 Chapter 2 Safety Limits and Limiting Safety System Settings 2.1 Safety Limits of Reactor Operation . . . . V/2-1 2.1.1 Applicability . . . . . . . . . . V/2-1 2.1.2 Objective. . . . . . . . . . . . . V/2-1 2.1.3 Specifications . . . . . . . . . . V/2-1 2.1.4 Bases. . . . . . . . . . . . . . . V/2-1 2.2 Limiting Safety System Settings. . . . . . V/2-2 2.2.1 Safety Channel Set Points. . . . . V/2-2 2.2.1.1 Applicability . . . . . . V/2-2

2. 2.1. 2 Objective . . . . . . . . V/2-2 2.2.1.3 Specification . . . . . . V/2-2 2.2.1.4 Bases . . . . . . . . . . V/2-2 V/iii

e t

Chapter 3 Limiting Conditions for Operation . . . . . .

3.1 Reactivity Limitations . . . . . . . . . V/3-1 3.1.1 Shutdown Margin. . . . . . . . . V/3-1 3.1.2 Excess Reactivity. . . . . . . . V/3-1 3.1.3 Experiments. . . . . . . . . . . V/3-1 3.1.4 Control Rods . . . . . . . . . . V/3-1 .

3.2 Control and Safety Systems . . . . . . . V/3-2 3.2.1 Scram Time . . . . . . . . . . . V/3-2 3.2.2 Measuring Channels . . . . . . . V/3-2 3.2.2.1 Bases. . . . . . . . . V/3-2 3.2.3 Safety Channels. . . . . . . . . V/3-3 l

3.2.3.1 Bases. . . . . . . . . V/3-3 P 3.3 Radiation Monitoring Systems . . . . . . V/3-4 3.4 Engineered Safety Features . . . . . . . V/3-5 3.4.1 Safety High Level Radiation g

Monitor . . . . . . . . . . . . V/3-5 3.4.1.1 Specification. . . . . V/3-5

3. 4.1. 2 Basis. . . . . . . . . V/3-5 3.4.2 Containment . . . . . . . . . . V/3-5 3.4.2.1 Specification. . . . . V/3-5 3.4.2.2 Bases. . . . . . . . . V/3-5 1.5 Limitations on Experiments . . . . . . . V/3-6

3. 5.1 Experiments . . . . . . . . . V/3-6

3. 5.1.1 Applicability. . . . . V/3-6 3.5.1.2 Objective. . . . . . . V/3-6 3.5.1.3 Specification. . . . . V/3-6 3.5.1.4 Bases . . . . . . . . V/3-6 .

3.6 Fu el . . . . . . . . . . . . . . . . . . V/ 3-8

3. 6.1 Applicability. . . . . . . . . . V/3-8 3.6.2 Objective. . . . . . . . . . . . V/3-8 3.6.3 Specifications . . . . . . . . . V/3-8 V/iv O

3.6.4 B a s e s . . . . . . . . . . . . . . . V/ 3 - 8 3.7 Primary Water Quality. . . . . . . . . . . V/3-9 3.7.1 Applicability . . . . . . . . . . . V/3-9 3.7.2 Objective . . . . . . . . . . . . . V/3-9 3.7.3 Speci fications. . . . . . . . . . . V/3-9 3.7.4 Bases . . . . . . . . . . . . . . . V/3-9 3.8 Radioactive Releases . . . . . . . . . . . V/3-10 3.8.1 Airborne Stack Release Limit. . . . V/3-10 3.8.2 Dose in Unrestricted Areas. . . . . V/3-10 3.8.3 Liquid Effluent Releases. . . . . . V/3-ll 3.9 Radiological Environmental Monitoring. . . V/3-12 3.10 Bases for Environmental Specifications . . V/3-14 Chapter 4 Surveillance Requirements . . . . . . . . . . . V/4-1 4.1 General. . . . . . . . . . . . . . . . . . V/4-1 4.2 Safety Channel Calibration . . . . . . . . V/4-1 4.3 Reactivity Surveillance. . . . . . . . . . V/4-1 4.4 Control and Safety Systeni Surveillance . . V/4-1 4.5 Radiation Monitoring System. . . . . . . . V/4-1 4.6 Engineered Safety Features . . . . . . . . V/4-2 4.6.1 Safety High Level Stack . . . . . . V/4-2 4.6.2 Containment . . . . . . . . . . . . V/4-2 4.7 Rea ctor Fuel . . . . . . . . . . . . . . . V/4-2 4.8 Primary Water. . . . . . . . . . . . . . . V/4-2 Chapter 5 De s i gn Fea tu re s . . . . . . . . . . . . . . . . V/5- 1 5.1 Rea c to r Fuel . . . . . . . . . . . . . . . V/ 5- 1 5.2 Control and Safety Systems . . . . . . . . V/5-2 5.2.1 Power Level (normal channels) . . . V/5-2 5.2.2 Long Power Level Channel . . . . . . V/5-2 5.2.3 Count Rate (start-up channel) . . . V-5-2 5.2.4 Neutron Source. . . . . . . . . . . V-5/2 V/v

5.3 Rod Control System . . . . . . . . . . . . V/5-3 5.3.1 Shim (control) Rods . . . . . . . . V/5-3 5.3.2 Regulating Rod . . . . . . . . . . . V/5-3 5.4 Cooling System . . . . . . . . . . . . . . V/E-4 5.4.1 Primary Coolant System. . . . . . . V/5-4 5.4.2 Secondary Cooling System. . . . . . V/5-4 5.5 Containment System . . . . . . . . . . . . V/5-5 5.5.1 Physical Fea tures . . . . . . . . . V/5-5 5.5.2 Emergency Sequence. . . . . . . . . V/5-5 5.5.3 Exhaust Duct Monitor (" stack monitor). . . . . . . . . . . . . . V-5-5 5.6 Fuel S tora ge . . . . . . . . . . . . . . . V/5-6 5.6.1 New Fuel. . . . . . . . . . . . . . V/5-6 5.6.2 Irradiated Fuel . . . . . . . . . . V/5-6 Chapter 6 Administrative Controls . . . . . . . . . . . . V/6-1 6.1 Organi za tion . . . . . . . . . . . . . . . V/6-1 6.1.1 Structure . . . . . . . . . . . . . V/6-1 6.1.2 Res pons i bi l i ty. . . . . . . . . . . V/6-1 6.1.3 S ta f fi n g . . . . . . . . . . . . . . V/ 6- 1 6.1.4 Selection and Training of personnstl V/6-2 6.1.5 Review and Audit. . . . . . . . . . V/6-2 6.1.5.1 Composition and Qualifications . . . . . . V/6-2 6.1.5.2 Charter and Rules. . . . . V/6-2 6.1.5.3 Review Function. . . . . . V/6-2 6.1.5.4 Audit Function . . . . . . V/6-3 6.2 P rocedures . . . . . . . . . . . . . . . . V/6-5 6.3 Experiment Review and Approval . . . . . . V/6-6 6.4 Required Actions . . . . . . . . . . . . . V/6-7 6.4.1 Action to be Taken in Case of a Reportable Occurrence . . . . . . . V/6-7 6.5 Reports. . . . . . . . . . . . . . . . . . V/6-8 6.5.1 Operating Reports . . . . . . . . . V/6-8 6.5.2 Special Reports (reportable occurrences). . . . . . . . . . . . V/6-8 V/vi

_- . A

6.6 Reco rds . . . . . . . . . . . . . . . . . . V/6-10 6.6.1 Records to be Retained for a Period of at least Five Years. . . . . . . V/6-10 6.6.2 Records to be Retained for at least One Requalification Cycle or for the Length of Employment of the Individual, Whichever is Smaller. . V/6-10 Chapter 7 ALARA (10 CFR 50.36a) . . . . . . . . . . . . . V/7-1 Chapter 8 Re fe rence s . . . . . . . . . . . . . . . . . . . V/8-1 V/vif

APPENDIX V TECHNICAL SPECIFICATIONS LIST OF FIGURES PAGE Figure V/6-1 Organizational Relations . . . . . . . . V/6-2 LIST OF TABLES Table V/3-1 Radiological Environmental Monitoring System . . . . . . . . . . . . . . . . . V/3-13 V/viii

1.0 DEFINITIONS The terms Safety Limit (SL), Limiting Safety System Setting (LSSS),

and Limiting Condition of Operation (LCO) are as defined in 50.36 of 10 CFR Part 50.

1.1 SAFETY CHAN"El A Safety Channel is a measuring or protective channel in the reactor safety system.

l.2 REACTOR SAFETY SYSTEM The Reactor Safety System is a combination of safety channels and associated circuitry which forms the automatic protective system for the reactor, or provides information which requires the initiation of manual pro-tective action.

1.3 OPERABLE Operable means a component or system is capable of performing its intended function in its required manner.

1.4 CHANNEL CHECK A Channel Check is a qualitative verification of acceptable performance by observation of channel behavior.

1.5 CHANNEL TEST A Channel Test is the introduction ' ~ a calibration or test signal into the channel to verify that it r.sponds in the specific manner.

1.6 CHANNEL CALIBRATIO" A Channel Calibration is an adjustment of the channel components such that its output responds, within specified range and accuracy, to known values of the parameter which the channel measures. Calibration shall encompass the entire channel, including readouts, alarm, or trip.

1.7 II'mennLED SHUIN2l An Unscheduled Shutdown is any unplanned shutdown of the reactor, after startup has been initiated.

1.8 RE ACTOR SHUTD0'!N The reactor is shut down when the negative reactivity of the cold, clean core including the reactivity worths of all experiments is equal to or greater than the shutdown margin.

V/1-1

1.9 REACTOR OPERATING The reactor is considered to be operating whenever it is not secured nor shut down.

1.10 REACTOR SECURED The reactor is secured when:

A. The core contains insufficient fuel to attain criticality under optil.um conditions of moderation and reflection, or B. The moierator has been removed, or C. 1. All control rods fully inserted as required by ~echnical Specifications, and

2. The console key switch is in the off position and the key is removed from the lock, and

3. No work is in progress involving core fuel, core struc-ture, installed control rods or control rod drives unless they are physically decoupled from the control rods, and

4. No in-core experiments are being moved or serviced with a reactivity worth exceeding the maximum value allowed for a single experiment or one dollar, whichever is smaller.

l.11 MEASURING CHANNEL A Measuring Channel is the combination of sensor, lines, amplifiers, and output devices which are connected for the purpose of measuring the value of a parameter.

1 12 REPORTABLE OCCURRENCE A Reportable Occurrence is any of those conditions described in Section 6.5.3 of this specification.

1.13 AN EXPERIMENT An Experiment is an apparatus, device or material, placed in the reactor core, in an experiment facility, or in line with a beam of radiation amanating from the reactor, excluding devices designed to measure reactor characteristics such as detectors and foils.

V/1-2

A. Secured Experiment. Any experiment, experiment facility, or component of an experiment is deemed to be secured, or in a secured position, if it is held in a stationary position relative to the reactor core.

B. Movable Experiment. A movable experiment is one which may be inserted, removed, or manipulated while the reactor is critical.

C. Untried Experiment. An Untried Experiment is a single experi-ment or class of experiments that has not been previously evalua-ted and approved by the Reactor Use Committee.

1.14 EXPERIMENT FACILITIES An Experiment Facility is any structure, device or pipe system which is intended to guide, orient, position, manipulate, control the environ-ment or otherwise facilitate a multiplicity of experiments of simi-lar character.

1 .15 CONTROL R0D A Control Rod is a semephore-type blade fabricated with cadmium as the neutron absorbing material which is used to compensate for fuel burnup, temperature, and poison effects. A control rod is magneti-cally coupled to its drive unit allowing it to perform the safety function when the magnet is de-energized.

1.16 READILY AVAILABLE ON CALL Readily Available on Call means an individual who A. has been specifically designated and the designation known to the operatur on duty, B. keeps the operator on duty informed of where he may be rapidly contacted (e.g. by phone, etc.)

C. is capable of getting to the reactor facility within a reason-able time under normal conditions (e.g.,1 hr. or within a 30 mile radius).

1.17 ROD DROP TIME Rod Drop Time is the elapsed time between the instant a liniting safety system set point is reached and the instant that the rod is fully inserted.

1.13 DROP-ROD SCRAft All four control rods fall by gravity into the core. Cooling water circulation continues.

V/1-3

1.19 FULL SCRA" The cooling water moderator is dumped in addition to the Drop-Rod Scram.

1.20 IrmIBIT Inhibit prevents the withdrawal of control rods under a potentially unsafe condition. When under auto-control, inhibit is ef fected by decoupling the rod drive from the auto-controller. The decoupling may initiate a drive-reg-rod-down sequence.

1.21 SAFETY LIMITS Safety Limits are limits on important process variables which are found to be necessary to reasonably protect the integrity of certain physical barriers which guard against release of radioactivity. The principal physical barrier is the fuel cladding.

1.22 SAFETY SYSTEM MARGIN The difference between a trip point setting and the corresponding Safety Limit that is known, estimated or dictated. Trip points are set at, or below the safety limit.

V/1-4

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIf1ITS OF REACTOR OPERATION 2.1.1 APPLICABILITY This specification applies to the variables that affect thermal and hydraulic performance of the core. They are:

A. Power in KW B. Flow in GPM.

C. Maximum coolant temperature in F.

2.1.2 OBJECTIVE To assure fuel cladding integrity 2.1.3 SPECIFICATIONS A. The maximum steady power level under various flow conditions shall not exceed 100 kw. The maxinum transient power, as shown on the linear recorder, shall not exceed 150 kw.

B. The coolant flow rate shall be at least 10 gpm at all power levels greater than one watt.

C. The core water outlet temperature shall not exceed 200 F.

D. The specific resistivity of the prinary water shall not be less than 0.5 megohm centimeters.

2.1.4 BASES Operating experience shows that specifications (A), (B), and (D) suffice to maintain cladding temperatures below 2120F. Specification (B) is known to be conservatively safe from tests conducted under a temporary amendment that permitted brief periods of operation at 500 kw. Although steady state operation at 150 kw and 10 gpm might produce boiling (depending upon seasonal variations in secondary water temperature), the fast transients envisioned here, will be temperature-damped by the system thermal inertia. The transient specification is introduced here to provide a Safety System Margin (Definition 1.2?).

V/2-1

2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 SAFETY CHANNEL SET-POINTS 2.2.1.1 Applicability This specification applies to the set-points of the safety channels.

2.2.1.2 Objective To ~nsure that automatic action is initiated that will prevent a safety limi from being exceeded.

2.2.1.3 _Speci fi cation The limiting safety system settings are:

A. Power level at any flow rate shall not exceed 125 kw.

B. The primary coolant flow rate shall be greater than 10 gpm .

C. The average primary coolant outlet temperature shall not exceed 180 F.

D. The primary coolant shall be demineralized light water with a specific resistivity not less than 1 megohm centimetors.

2.2.1.4 Bases These limits arise from operating experience. The trip-points A and B initiate automatic scram. Trip-ooint C is signalled by a horn and light, trip-point D by a light. Points C and D can be approached and exceeded only slowly and timeng :ill permit operator intervention.

At power levels below one watt, the primary coolant finw trip may be bypassed and the primary circulation cut off. This wou.o normally occur during fuel transfer operations, or when trying to isolate a breach of cladding in a fuel bundle.

V/2-2

3.0 LIMITING CONDITIONS FOR OPERATION 3.1 REACTIVITY LIMITATIONS 3.1.1 SHUTD0'.!N MARGIN The minimum shutdown margin provided by control rods in the cold, xenon-free condition with the highetc-worth rod fully withdrawn shall not be less than $2.77.

This specification ensures that the reactor can be shut down from any operating condition and remain shut down after cool-down even if the highest-worth control rod is stuck in its fully withdrawn condition.

3.1.2 EXCESS REACTIVITY The core excess reactivity at cold critical, without xenon poisoning, shall not exceed $3.54.

3.1.3 EXPERIMENTS Reactivity limits on experiments are specified in 3.5 below.

3.1. 4 CONTROL RODS The reactivity insertion rate for a single control blade shall not exceed $0.077/sec.

V/3-1

3.2 CONTROL AND SAFETY SYSTEMS 3.2.1 SCRAM tit:E The scram time shall not exceed 1 second (roa drop time).

3.2.2 MEASURING CHANNELS The minimum number and type of measuring channels operable and pro-viding information to the control room operator required for reactor operation are given as follows:

Channel No. Operable Safety Amplifier 2 Linear w/ auto controller 1 Log N and Period Channel 1 Start up la Rod Position 4 Coolant Flow 1 Coolant Temperature 2 (primary)

Core Level 1 NOTE: a. Operable below 0.02U 3.2.2.1 Bases The normal power level instruments (" Level Safeties") provide redundant information on reactor power in the range 5% - 150% of the normal operating power level of 100 kw. The linear power channel presents the reactor power in a linear manner by decades.

The linear recorder also serves as part of the feedback network for the auto controller servo.

The log power instrument (" Log N") provides usable reactor power information in the logarithmic range of 0.01 W to 1 ftet. By differentiating the log of the power, the reactor period infor-mation is also displayed.

The count rate channel covers the neutron flux range fron the source level (% 1 cps) to 10" cps on a digital scale. It enables the operator to start the reactor safely from a shutdown condition, and to bring the power to a level that can be measured by the Log N instrument.

Coolant flow rate and temperature instruments allow the operator to calculate reactor power and calibrate the neutron flux channels in terms of power. The primary water outlet temperature must be monitored.

Rod position indicators show the operator the relative positions of control rods, and enable rod reactivity calibrations to be made.

V/3-2

3.2.3 SAFETY CHANNELS The minimum number and type of channels providing automatic action that are required for reactor operation are as follows:

Channel No. Operable Function Power Level (Safety) 2 Scram 0 > 125 kw Power Level (Log N & Peri.od) 1 Scram 0 < 3 sec. period Inhibit 0 < 6 sec. period a

Count Rate l Inhibit 0 < 2 cps Core Water Level 1 Scram 0 < 45 in.

b Primary Coolant Flow l Scram 0 < 10 gpm Manual Button 1 Full Scran Keyswitch 1 Scram Notes: a. Operable below 0.02 W and bypassed above

b. May be bypassed at power levels below 1. att.

3.2.3.1 Bases The power level scram provides redundant automatic protective action to prevent exceeding 125% of the license limit on reactor power.

T he period scram, assisted by the intermediate rod inhibit, limits the rate of increase in reactor power to va ues that are controllable l

without reaching excessive power levels or temperature. These functions are not limiting safety system settings.

The inhibit on the count rate channel prevents inadvertent criticality during cold startup that could arise from lack of source neutrons and the neutron instrument response.

The scram on core level is redundant, full scram automatically damps the core water (moderator). The scram acts as an inhibit during start-up until the minimum core level is reached.

The coolant flow scram ensures adequate coolant flow to prevent boiling in the core.

The keyswitch scram prevents unauthorized operation of the reactor.

Bypass is permitted on non-power parameters for experimental and test purposes only.

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3.3 RADI ATION MONITORING SYSTEMS The minimum acceptable radiation monitoring instrumentation required for reactor operation is as follows:

No. Operable Max. Alarm Tyge No. Operable Setpoint Function Safety high level 1 25 mR/hr Detect high radiation radiation and/or high stack gas concentration Alarm and isolate at > 25 mR or 8x10 s u Ci/ml Exhaust Duct Monitor 1 1.8x10 Detect Ar41 and other

(" Stack Monitor") pCi/ml* radioactive gases via ion chamber Alarms with displays in the control room

In the event that the limits for Argon 41 contained in 10 CFR Part 20, Appendix B. Table II, with a reduction factor of 460 are exceeded in the stack, the ventilation fans shall be shut down and the automatic damper system closed to limit natural circulation from the reactor room to the external environment.

Fixed Area Monitors 2 5 mR/hr Detect radiation (y) in key locations; alarm in Control room Evacuation Switch 1 --

Alarm to initiate evacua-tion sequence.

(Manual).

NOTE: For maintenance or repair, required radiation monitors may be replaced by portable or substitute instruments for periods up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months provided the function will still be accomplished. Inter-ruption for brief periods to permit checking or calibration is permissible.

V/3-4

3.4 EflGiflEERED SAFETY FEATURES These specifications apply to required equipment for the confinement of activity through controlled release of reactor building air to the atmosphere.

3.4.1 SAFETY HIGH LEVEL RADIATI0fl MONITOR 3.4.1.1 Specification See 3.3.

3.4.1.2 Basis This monitor senses excessive radiation in the reactor room and/or high concentrations of radioactive gases in the exhaust stack and automatically initiates the reactor room ventilation shutdown sequence, which consists of the shutdown of the reactor room supply and exhaust fans, closure cf dampers in the building ventilation system and causing a drop rod scram if the reactor is in operation.

3.4.2 CONTAlfiMEf4T 3.4.2.1 Specification A. The exhaust fan shall be capable of sustaining a negative pressure within the reactor building at an exhaust flow rate of approximately 14,000 cfm.

B. The high bay ventilation exhaust and intake fans are inter-locked to shut off simultaneouslj when the ven'ilation system is shut down.

C. Spring loaded, air operated damper motors automatically close the intake and exhaust dampers.

D. All doors to the reactor high bay shall be normally closed while the reactor is operating. Transit is not prohibited, and a door may remain open if guarded.

3.4.2.2 Bases To effect controlled release under normal conditions of gaseous activity present in the building atmosphere, a negative pressure is required so that any building leakage will be inward. Under emergency condi-tions, the reactor room ventila'. ors will be shut down and the dampers closed, thus effectively conta ling the radioactivity in the high bay.

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3.5 LIMITATIONS ON EXPERIPENTS 3.5.1 EXPERIf1ENTS 3.5.1.1 Applicability, This specification applies to those experiments installed in the reactor and its experimental facilities.

3.5.1.2 Objective The objective is to prevent damage to the reactor, excessive release of radioactive material in the event of experiment failure, and to prevent the safety limits from being exceeded.

3.5.1.3 Speci fi ca tion Experiments installed in the reactor shall meet the following conditions:

A. The reactivity worth of any single secured experimed shall not exceed $.92.

B. The sum of the absolute reactivity worths of experiments shall nc exceed $3.54.

C. An experiment shall not be inserted or removed unless all the control blades are fully inserted or its absolute reactivity worth is less than that which would cause a 20-second posi-tive stable period ($.28).

D. Experimental apparatus, material, or equipment to be inserted in the reactor shall be reviewed to insure non-interference with the safe operation of the reactor.

E. Each type of experiment utilizing the reactor must have been previously reviewed and approved by the Reactor Use Committee and/or the Supervisor and Health Physicist for routine experiments.

F. No explosive materials shall be irradiated.

3.5.1.4 Bases A. Specification A prevents prompt criticality.

B. SPERT and BOPfJ tests showed that plate type fuel elements survived step reactivity insertions of $3.54.

C. The insertion of all rods assures that the reactor will not go critical if a sample inadvertently adds reactivity upon sample insertion or removal.

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D. Ensures that no physical or nuclear interference with the safe operation of the reactor will occur.

E. Ensures that all experiments are evaluated by an independent group knowledgeable in the appropriate fields.

F. These Technical Specifications prohibit the irradiation of explosives.

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3.6 FUEL 3.6.1 APPLICABILITY These specifications apply to the number and condition of the fuel elements present in the core.

3.6.2 OBJECTIVE To geometrically constrain excess reactivity, and to avoid excessive release of fission products.

3.6.3 SPECIFICATIONS 3.6.3.1 The maximum fuel loading shall consist of 24 bundles, each containing 11 plates. The plates consist of enriched uranium-aluminum sandwiched between high purity aluminum clad.

1 6.3.2 Fuel element loading and distribution in the core shall comply with the fuel handling procedures.

3.6.3.3 Fuel elements exhibiting release of fission products due to cladding rupture shall, upon positive identification, be removed from the core. Fission proiuct contamination of the primary water shall be treated as evidence of fuel element failure, and positive identi-fication shall be determined by y-ray spectroscopy.

3.6.4 CASES 3.6.4.1 The original and present configurations.

3.6.4.2 There are written procedures for fuel handling and loading.

3.6.4.3 Migration of fission products within the fuel meat will be strongly inhibited by the large volume fraction of aluminum (nearly 98% aluminum) in the alloy. Thus a relatively small defect cannot release large quantities of fission products originating at locations remote from the defect. Small releases will be largely captured by the primary water system. (See 3.7 Primary Water Quality.) A major failure would be readily evident via one or more of the inhouse radia-tion monitors.

3.6.4.4 ANSI N398-1974.

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3.7 PRIMARY WATER QUALITY 3.7.1 APPLICABILITY This specification applies to primary cooling system water in contact with fuel elements.

3.7.2 03JECTIVE To minimize corrosion of the aluminum cladding of fuel plates and activation of dissolved materials.

3.7.3 SPECIFICATIONS 3/7.3.1 Primary water temperature will not exceed 200 F.

3.7.3.2 Primary water specific resistance is to be not less than 500,000 chm-cm.

3.7.3.3 Primary water shall be sampled, evaporatively concentrated, and the grnu radioactivity of the residue measured with a thin window GM counter. Radioisotope identification shall be by y-ray spectro-scopy. This specification and procedure shall prevail:

A. upon the appearance of any unusual radioactivity in the pri-mary water or the primary water demineralizers and B, prior to the release of any primary water from the site.

3.7.4 BASES 3.7.4.1 Specification A is designed to protect the fuel element integ-rity and is based upon operating experience.

3.7.4.2 At this quality, the activation products (of trace minerals) do not exceed acceptable limits.

3.7.4.3 This specification is designed to:

A. detect and identify fission products resulting from fuel failure and

8. to fulfill reportability requirements pertaining to liquid wastes.

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3.8 RADIOACTIVE RELEASES 3.8.1 AIRBORNE STACK RELEASE LIMIT

3. 8.1.1 The concentration of Argon 41 released to the atmosphere shall not exceed the limits of 10 CFR Part 20, Appendix B, Table II, Column 1 with a reduction factor of 460 defined as the product of A. a reactor use factor B. an occupancy factor, and C. a dilution factor.

The Comission shall be notified if, over any one year period, the reactor use factor, the occupancy factor, or the dilution factor change so as to increase the effective reduction factor.

3.8.1.2 In the event that the limits for Argon 41 contained in 10 CFR Part 20, Appendix B, Table II, with a reduction factor of 460 are exceeded in the stack, the ventilation system shall be secured and shall cause the automatic damper system to isolate the reactor room, and the reactor shall be shut down.

3.8.2 DOSE IN UNRESTRICTED AREAS Roof top users in the prevailing downwind direction from the stack (NE of the stack) are subject to exposure to argon-41 and its radia-tions. Reference 1, page 8, indicates that the plume dilution factor, under the prevailing wind condition, and at the point of closest approach (in the NE direction) is(5.05 1.68) x 10 3 For conservatism, taking the factor as 7 x 10-3, the concentration at that point is estimated to be 8.4 x 10-epci/ml if the reactor is operating at 100 kw. By Amendment 10 the reactor is authorized to operate no more nan 18.8%

of the 45 hour5.208333e-4 days 0.0125 hours 7.440476e-5 weeks 1.71225e-5 months work week or 5% of the year (effectively 44,000 kwh per year). The annual average argon-41 concentration at the point under consideration is limited to an estimated 0.05 x 8.4 x 10-8

= 4.2 x 10-9pci/ml. This assumes 100% occupancy, that the prevailing wind is realized 100% of the time that the reactor operates, and that the reactor is operated at maximum authorized level.

A concentration of 4.2 x 10~9pci/ml is 10.5% of the MPC for argon-41 and, if treated as a semi-infinite cloud, could be expected to yield a radiation exposure equal to 10.5% of 500 mrem /yr. or 52.5 mRen/yr.

A thermoluminescent Dosimeter program undertaken over a two year period in 1976 and 1977, indicated y-dose levels of 18 to 50 mrem /yr. on the roof areas. The higher ranges were closely related to concrete proxi-mity and unrelated to direction or distance from the reactor room venti-lation stack. For dosimeters more isolated from concrete, values near the stack ranged from 18 to 24 mrem /yr. Assuming that these are reflecting the y radiation from argon-41, and adjusting for the ratio of maximum authorized usage to average actual usage in 1976-7, V/3-10

it is estimated that roof top exposures on the order of 90 mrem /yr.

(at 100% occupancy) might be expected at the full authorized operating level.

Reference 1 (pages 9 and 10) treats the potential exposure of occu-pants within the Mathematical Sciences building and concludes inferen-tially that individual exposures are not likely to be more thar; 0.5 mrem /yr.

In recognition of ALARA, UCLA is committed to the installation of hold-up/ decay tanks. An amendment will be prepared and submitted prior to September 1,1980 to authorize such operation. The amend-ment will provide estimates of the releases expected with the hold-up/

decay system.

3.8.3 LIQUID EFFLUENT RELEASES 3.8.3.1. Before release from the holding tank or sump, the liquid waste shall be sampled and the activity level measured.

3.8.3.2 Liquid waste siiall not be released from the site to the uncon-trolled storm sewer unless its activity concentration, including dilution with non-radioactive waste water, is below that specified in 10 CFR, Part 20, Appendix B. Table II, Column 2. For releases to the controlled sanitary sewer, the limits shall not exceed those of Appendix B, Table I, Column 2.

3.8.3.3 Records of and reports on liquid radioactive effluent releases shall be as specified in Section 6 of these Technical Specifications.

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3.9 RADIOLOGICAL ENVIRONMENTAL MONITORING The radiological environmental monitoring program shall be conducted as specified in Table V/3-1. The results of analyses performed on the radiological environmental monitoring samples shall be summarized in the Annual Report to the Nuclear Regulatory Commission.

V/3-12

Table V/3-1 RADIOLOGICAL ENVIRONMErlTAL f10NITORIflG PROGRAM Exposure Pathway Number of Samples Sampling and Type and Frequency and/or Sample and Sample locations Collection Frequency of Analysis

1. Airborne

a. Gaseous 1 sample from exhaust Continuous during Flow-through ion chamber stack operations and core al tera tions

b. Particulates 1 sample from exhaust '

Continuous ~ operation of Particulate sampl er.

stack ,

sampler with sample Analyze for gross beta p rom collection as required radioactivity following N

u reac or b 1 by dust loading but at filter change. Perform intake air least once per 7 days gamma isotopic analysis on h; each sample when gross beta activity is > 10 times the mean of control samples for any medium.

2. Direct Radiation Annual survey of all Annually Gamma and neutron, with internal areas and portable instruments extending beyond the controlled perimeter of the facility

3. Other Area Wipes Weekly Gross al pha-beta-gamma 14 within the controlled area 4 external to the controlled area

3.10 BASES FOR ENVIRONMENTAL SPECIFICATIONS 3.10.1 Specification 3.8.1 is provided to ensure that the dose at the uncontrolled area boundary from gaseous effluents from the site will be within the annual dose limits of 10 CFR Part 20 for unrestricted areas. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an indi-vidual at or beyond the restricted area boundary to 5 500 mrem /yr.

to the total body.

3.10.2. Specification 3.8.2 is an interim measure, and a statement that the conditions imposed by Amendment 10 (1976) will continue until amended.

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4.0 SURVEILLANCE REQUIREMENTS 4.1 GENERAL The requirements listed below generally prescribe tests or inspections to verify periodically that the performance of required systems is in accordance with specifications given above in Sections 2 and 3. In all instances where the specified frequency is annual, the interval between tests is not to exceed 14 months; and when semiannual, the interval should not exceed 7 months.

4.2 SAFETY CHANNEL CALIBRATION A channel calibration of each safety channel shall be performed annually (see Section 3.2.3).

4.3 REACTIVITY SURVEILLANCE 4.3.1 The reactivity worth of each control rod (including the regulatina rod) and the shut-down margin shall be determined whenever oparation requires a reevaluation of core physics parameters, or annually, whichever occurs first. The rod worth will be determined by observing the rod position changes necessary to compensate for samples of known worth.

4.3.2 The reactivity worth of an experiment shall be measured at low power, before conducting the experiment. A reactivity measurement on a rabbit experiment may be performed at high power if the experiment has been performed previously and there is goed reason to believe the reactivity worth is less than 28c.

4.4 CONTROL AND SAFETY SYSTEM SURVEILLANCE 4.4.1 The rod drop times shall be measured annually. The reactivity worth of each control blade, reactivity insertion rate of each control, blade, and shutdown margin shall be determined annually.

4.4.2 A channel test of each measuring channel in the reactor safety system shall be performed following related maintenance.

4.4.3 A channel check of each measuring channel in the reactor safety system shall be performed prior to each startup. It is not performed for startup from temporary shutdown.

4.5 RADIATION MONITORING SYSTEM 4.5.1 The safety high level monitor, stack monitor, and area monitors shall be calibrated semi-annually.

4.5.2 The area monitors shall receive a channel test prior to each startup.

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4.6 ENGINEERED SAFETY FEATURES 4 . 6.1 SAFETY HIGH LEVEL STACK MONITOR The safety high level stack monitor will shut down the reactor room ventilation system and dampers will isolate the room from the surrounding environment if the trip point is exceeded.

4.6.2 CONTAINMENT The operability of the evacuation alarm and containment isolation system (fan and damper shutdown) shall be tested and verified semiannually.

4.7 REACTOR FUEL 4 . 7 .1 Upon receipt from the fuel vendor, all fuel elements shall be visually inspected and the accompanying quality control documents checked for compliance with specifications.

4.7.2 Each new fuel element will be inspected for damage and flow obstructions prior to insertion into the core.

4.8 PRIMARY WATER 4.8.1 The specific resistance of the pool water shall be measured prior to each start-up.

4.8.2 An analysis of the primary water for radioactive material shall be done annually, and prior to the release of any primary water from the site. An analysis shall also be made upon the appearance of any unusual activity in the water or the demineralizer.

V/4-2

5.0 DESIGN FEATURES Those design features relevant to operational safety and to limits that have been previously specified are described below. These fea-tures shall not be changed without appropriate review.

5.1 REACTOR FUEL Fuel elements shall be of the general MTR type with thin plates con-taining uranium fuel enriched to about 93% 2350 and clad with alumi-num. The fuel matrix may be fabricated by alloying high purity aluminum-uranium or by the powder metallurgy method where the starting ingredients (uranium-aluminum) are in fine powder form. The fuel matrix may also be fabricated from uranium oxide-aluminum (U38 0 -A1) using the powder metallurgy process. Elements shall conform to these nomi-nal specifications:

A. Overall size (bundle): 2.845 in. x 2.14 in. x 25.625 in.

B. Clad thickness: 0.015 in.

C. Plate thickness: 0.070 in.

D. Water channel width: 0.137 in.

E. No. of plates : standard element - 11 fueled plates F. Nate attachment: bolted with spacers G. Fuel content per plate: 149 U nominal H. Fuel bundles: 24 bundles, each containing 11 plates V/5-1

5.2 CONTROL AND SAFETY SYSTEMS Design features of the components of this system (3.2.2, 3.2.3) that are important to safety are given below.

5.2.1 POWER LFVEL (NORMAL CHANNELS)

For this function two independent measuring channels are provided, both are required for the reactor to be operable. Each channel covers reliably the range from about 5% to 150% (of 100 kw). Each channel comprises an uncompensated boron-coated ion chambe- 7eeding an ampli-fier that controls electronic switches in the DC current that flows through each control rod electromagnetic clutch. Each channel con-trols and drops all control rods. Each channel is fail-safe. Scram from each channel is accomplished through interruption of power to the electromagaetic clutches that couple the rod drives to the rods.

Each channel indicates power level on a panel meter. Each chamber can be changed in position, over a limited range, so as to allow the channel reading to be standardized against reactor thermal power.

5.2.2 LOG POWER LEVEL CHANNEL For this function a single channel is provided, covering reliably the range 10-2W to 1.0 Mw with a logarithmic output indication on both a panel meter and a chart recorder. To cover the rar.ge under all core conditions a camma-compensated boron-ion chamber is used to supply a logarithmic amplifier. The chamber can be changed in position, over a limited range, so as to allow the channel reading to be standardized against reactor thermal power. Rate of change of power information is also derived, in the form of a period, that can initiate full scram. This channel also provides control and inhibit actions, viz. bypassing of the start-up channel functions, bypassing of " closures" scram below 1.0 W, and inhibit of control rod.

5.2.3 COUNT RATE (START-UP CHANNEL)

A BF3 proportional chamber is used to supply pulses to a scaler and digital count rate circuitry. Pulse height discrimination selects pulse amplitudes that correspond to neutron events and rejects those from ganna events. Count rate on a digital counter is displayed on a panel meter. The channel covers a range of 1 - 10' cps, corresponding roughly from source to .02 W. To prevent control-rod withdrawal when the neutron count rate information may not be reliably indicated, an inhibit is provided when the count rate is less than 2 per second.

This inhibit is bypassed at a power of > .02 W. The scaler can also be used for obtaining accurate values at low count rates if needed (e.g., approach to critical with new fuel or for instructional purposes).

5.2.4 NEUTRON SOURCE For obtaining reliable neutron information necessary to startup from a cold shut-down condition, a radium-beryllium neutron source is per-manently installed.

V/5-2

5.3 R00 CONTROL SYSTEM

5. 3.1 SHIM (CONTROL) RODS Three control rods are provided for the control of core reactivity.

These rods are cadmium-tipped magnesium (see 4.3.1). Individual inte-gral worths vary from about $2.40 - $2.70, depending on position and individual characteristics. The rods are coupled to drive shafts through electromagnetic clutches that allow release of the rods within 12 ms after receiving a scram signal. Position indicators on the control console show the extent of withdrawal for each rod. To limit the rate of reactivity increase upon startup, the rod drive speeds are limited to 7.7C/sec. and only one rod can be withdrawn at a time.

These rods are not otherwise automatically controlled, but are used to compensate for seasonal and long-term reactivity changes.

5.3.2 REGULATING R00 One regulating rod is provided to aid in fine control and maintenance of constant reactor power for long periods. The rod is limited to a total worth of about $1.80 and can be either manually or servo-controlled.

The drive speed is approximately 1% (of fuli range) per second. In the commonly used range, this amounts to approximately $0.02/second.

V/5-3

5.4 COOLING SYSTEM 5.4.1 PRIMARY COOLANT SYSTEM In normai operation, the primary water is pumped from the dump tank to the bottom of the fuel boxes, upward past the fuel plates, to over-flow pipe weirs above the fuel plates, and returned to the dump tank.

The dump tank is fitted with tube bank through which secondary water is circulated to effect heat removal . There is a bypass from the fuel box fill line to the dump tank, that is controlled by the dump valve. Under the full scram condition, the dump valve is opened and drains the core water, by gravity, to the dump tank. The removal of the core water moderator introduces approximately $25,00 of nega-tive reactivity into the core. The dump valve is spring loaded open, closed by air pressure, and will open upon failure of the air supply.

Primary water quality is maintained by the circulation of a slip-stream (30 gph) from the pump discharge through cartridge filters and demineralizers to the dump tank. The primary water flow rate (10 gpm minimum, normally 16 gpm) is automatically controlled and indicated at the control console. Core inlet and outlet water temperatures are monitored. The combination of temperature rise and flow rate yield the heat removal rate that defines the reactor thermal power.

5.4.2 SECONDARY COOLING SYSTEM Reactor power transferred through the heat exchanger is dissipated to the storm drain by once-through cooling water. To minimize cor-rosion, the exchanger has stainless steel tubes. To prevent water from entering the secondary system should a tube leak occur, the static pressure in the secondary is made higher than that of the primary through the relative elevations of the two systems and the back pres-sure of the city water system.

V/5-4

5.6 FUEL STORAGE 5.6.1 NEW FUEL Unirradiated new fuel elements are stored in a vault-type room security area equipped with intrusion alarms in accordance with the Security Pl an . Elements are stored in a steel, fireproof safe in which a cad-mium plate separates each layer of bundles. With such an arrangement, subcriticality is assured.

5.6.2 IRRADIATED FUEL Irradiated fuel is stored upright in a dry storage pit within the reactor building in criticality-safe holes. Holes are located with center-to- center square spacing of 13 inches, each can accomodate 1 to 3 elements.

V/5-6

5.5 CONTAINMENT SYSTEM

5. 5.1 PHYSICAL FEATURES The containment structure consists of the reactor building, with a free air volume of about 1600m3 . This building houses the reactor, the primary cooling system including the dump tank / heat exchanger and the hold-up tanks. Personnel access is via specially keyed doors.

Ventilation access to the building is through pneumatically operated damper valves that can be used to prevent natural through-flow when the fans are turned off. These dampers are fail safe and close upon loss of air pressure.

5.5.2 EMERGENCY SEQUl:NCE The emergency sequence is initiated either automatically by the safety high level radiation stack monitor or manually by the console operator.

The sequence is that the reactor room ventilation supply and exhaust fans are shut down and the dampers are closed.

5.5.3 EXHAUST DUCT MONITOR (" STACK MONITOR")

Air in the exhaust duct is continuously sampled for particulate activity.

During reactor operations or core alterations, exhaust air drawn from the reactor room is continuously monitored for gross concentrations of radioactive gases.

V/5-5

6.0 ADMINISTRATIVE CONTROLS 6.1 ORGANIZATION 6 .1.1 Structure The organization structure for the management and c;;cratica of the reactor facility is shown in Figure V/6-1. Job titles are shown for illustra-tion and may vary. Four levels of authority are provided, as follows:

Level 1: Individual responsible for the facility license and site administration.

Level 2: Individual responsible for the reactor facility operation and management.

Level 3: Individual responsible for daily reactor operations.

Level 4: Reactor operating staff.

The Reactor Use Committee is appointed by, and shall report to, the Dean of the School of Engineering and Applied Science. Radiation safety personnel shall report to Level 2 or higher through a separ-ate and distinct organizational channel .

6 .1. 2 Responsibility Responsibility for the safe ope:ation of the reactor facility shall be within the chain of command shown in FigureV/6-1. Management levels in addition to having responsibility for the policies and operation of the reactor facility shall be responsible for safeguarding the public and facility personnel from undue radiation exposures and for adhering to all requirements of the operating license and technical specifications. In all instances responsibilities of one level may be assumed by designated alternates or by higher levels, conditional upon appropriate qualifications.

6.1. 3 Staffing 6 .1. 3.1 The minimum staffing when the reactor is not secured shall be:

A. A licensed Reactor Operator in the control room, and a health physics qualified individual in the facility.

B. A licensed Senior Reactor Operator shall be readily avail-able on call .

6.1.3.2 Events requiring the presence of a Senior Operator:

A. All fuel-element or control-rod alterations within the reactor core region.

B. Recovery from unplanned or unschedulec shutdowns.

V/6-1

BOMD OF REGENTS l .g (LICENSEE)

L. B UCLA CHN4CELLOP

/ 4 %

ASSOCD,TE EXECUTIVE ADMINISTRATIVE VICE C%NCELLOR VICE C MNCELLOR VICE CHNJCELLOR RESPONSIBLE OFFICER (RESEARCH)

(ACADEMIC)

LEVEL 2 DEN 4, SCHOOL OF ASSISTANT 4- -

ENGINEERING NO VICE CtW4CELLOR V

APPLIED SCIENCE COffU11Tf SAFETY REACTOR USL ' ~ ~

DIRECTOR, NUCLEAR ~ - '

NO OCCU'PATIOfAL SAFETY COFf41TTEE ENERGY LABORATORY SAFETY COH'.I TTEE

^"

tW4AGER, f JUCLEAR AFE ENERGY LABORATORY OFF M R I

k k REACTOR REACTOR HEALTH OPERATIONS gg g 37 LEVELS 3&4 FUFCTIONAL RESPOfiSlBILITIES EXPERIMENTAL USE APPROVAL STAFFING OVERALL SAFETY B'JDGET NRC COFfiUNICATIONS TECH 11 CAL CHN EES t#4IVERSITY PJBLIC PREPARE LICENSE

1Ef0MENTS ORGANIZATIONAL RELATIONS FIGURE V/6-1 V/6-2

6.1.4 Selection and Training of Personnel The selection, training, and requalification of personnel shall meet or exceed the requirements of ANS-15.4/N 380 and Appendix A of 10 CFR Part 55 and be in accordance with the requalification plan approved by the Commission.

6 .1. 5 Review and Audit There is a Reactor Use Committee which reviews and approves new experi-ments and proposed alterations to the reactor. The Committee shall review and audit reactor operations for safety.

6.1. 5.1 Composition and Qualifications. This committee shall be composed of the reactor supervisor and radiation health physicist, both ex officio (voting), and 3 other members having exnertise in reactor tech-nol ogy. Committee members shall be appointed by the Dean of the School of Engineering and Applied Science.

6.1.5.2 Charter and Rules. The committee shall function under the following operating rules:

A. The Reactor Use Committee shall mcet at least semiannually and shall keep written records of its meetings. The Committee shall report directly to the Dean of the School of Engineering and Applied Science.

B. A quorum shall be three members.

C. Any action recommended by the Reactor Use Committee, which may affect the operation and/or safety of the University com-munity beyond the Nuclear Energy Laboratory facility, shall be brought to the attention of the Campus Radiation Safety Committee which shall have veto power to such a recommendation.

D. The Committee may appoint one or more qualified individuals to perform the Audit Function.

6.1.5.3 Review Function. The following items shall be reviewed by the review group or a subgroup thereof:

A. Determinations that proposed changes in equipment, systems, tests, experiments, or procedures do not involve an unreviewed safety question.

B. All new procedures and major revisions thereto having safety sigaificance, proposed changes in reactor facility equipment, or systems having safety significance.

C. Tests and experiments in accordance with section 6.3.

D. Proposed changes in technical specifications, license, or charter.

E. Violations of technical specifications, license, or charter.

Violations of internal procedures or instructions having safety significance.

V/6-3

F. Operating abnormalities having safety significance, and audit repo rts .

G. Reportable occurrences listed in section 6.5.3.

6.1. 5.4 Audit Function. The audit function shall include selective (but comprehensive) examination of operating records, logs, and other documents. Where necessary, discussions with responsible personnel shall take place. In no case shall the individual or individuals conducting the audit be immediately reponsible for the area being audited. The following items shall be audited:

A. Review facility operations, procedures and records for safety considerations and recomend improvements where appropriate.

In addition to a continuing review of these matters, an inten-sive in-depth review of facility operations shall be made at least annually.

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6.2 PROCEDURES The facility shall be operated and maintained in accordance with approved written procedures. All procedures and major changes thereto shall be reviewed and approved by the Director of the Nuclear Energy Labora-tory prior to being effective. Changes which do not change the origi-nal intent of a procedure may be approved in writing by the reactor supervisor. Such changes shall be recorded and submitted to the Director for routinc review. The following types of written procedures shall be maintained:

A. Normal startup, operation and shutdown procedures for the reactor. These procedures shall include applicable checkoff lists and instructions.

B. Procedures which delineate the operator action required in the event of specific malfunctions and emergencies.

C. Radiological control procedures for all facility personnel.

D. A laboratory emergency procedure to guide the behavior and action of all personnel in the event of an emergency condition.

E. Procedures for the installation, operation and removal of experiments where reactor safety is concerned.

F. Procedures for handling irradiated and unirradiated fuel elements.

G. Procedures for operation of the pneumatic Sample Transfer System.

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6.3 EXPERIMENT REVIEW AND APPROVAL 6 . 3.1 All new experiments or classes of experiments that could affect reactivity or result in release of radioactive materials shall be reviewed by the Reactor Use Committee. This review shall assure that compliance with the requirements of the license, technical specifi-cations, and applicable regulations has been satisfied, and shall be documented.

6.3.2 The reactor supervisor and the resident health physicist shall review and approve in writing all proposed experiments prior to their performance.

6.3.3 The following conditions shall govern the performance of experiments:

A. The reactivity worth of any single secured experiment shall not exceed $0.92.

B. An experiment shall not be inserted or removed unless all the control blades are fully inserted or its absolute reactivity worth is less than that which would cause a 20-second posi-tive stable period ($0.28).

C. No explosive materials shall be irradiated.

D. The sum of the absolute reactivity worths of experiments shall not exceed $3.54.

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6.4 REQUIRED ACTIONS 6.4.1 ACTION TO BE TAKEN IN CASE OF A REPORTABLE OCCURRENCE Reportable Occurrences are described in 6.5.3.

A. Corrective action shall be taken to return conditions to normal; otherwise, the reactor shall be shut down and reactor opera-tion shall not be resumed unless authorized by the Level 2 authority or designated alternates.

B. All such occurrences shall be promptly reported to the Level 2 authority or designated alternates.

C. All such occurrences where applicable shall be reported to the Commission in accordance with section 6.5.3.

D. All such occurrences including action taken to prevent or reduce the probability of a recurrence shall be reviewed by the Reactor Use Committee.

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6.5 REPORTS In addition to the requirements of applicable regulations, reports shall be made to the Commission as follows:

6.5.1 OPERATING REPORTS A routine report (in writing to the Director, Division of Operating Reactors, USNRC, Washington, D.C. 20555) at the end of each 12-month period providing the following information:

A. A narrative summary of reactor operating experience, including the energy generated by the reactor (in megawatt-hours).

B. A discussion of unscheduled shutdowns, including the corrective action taken to preclude recurrence.

C. A summary of the preventive and corrective maintenance operations performed having safety significance.

D. A discussion of the changes in the facility and procedures, and the tests and experiments, carried out, without prior approval by the U.S. Nuclear Regulatory Commission pursuant to 10 CFR Part 50, Section 50.59.

E. A summary of the nature and amount of radioactive material released to the environs.

F. The results of any environmental surveys performed outside the facility.

G. A summary of significant (above 500 mrem) radiation exposures received by facility personnel and visitors in any one year, including the dates and times of significant exposures.

6.5.2 SPECIAL REPORTS (REPORTABLE OCCURRENCES) 6.5.2.1 There shall be a report not later than the following working day (by telephone or telegraph to the Director, NRC Region V Inspection

& Enforcement Office) and a report within 10 working days (in writing to the Director, Division of Operating Reactors, USNRC, Washington, D.C. 20555) of:

A. Release of radioactivity from the reactor above allowed limits, as provided by section 3.8.1 of this specification.

B. Any of the following:

B.1 a violation of the Technical Specifications or the facility license; B.2 an uncontrolled positive reactivity change that leads the system to, or beyond, a trip point.

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B.3 an uncontrolled or unanticipated release of radioactivity from the site; B .4 a safety system component malfunction or other system or component malfunction which renders or threatens to render the safety system incapable of performing its intended safety function; B.5 an observed inadequacy in the implementation of either administrative or procedural controls, such that the inadequacy causes or could have caused the existence or development of an basafe condition with regard to reactor operation; and B.6 abnormal degradation of reactor fuel as revealed by periodic inspection.

6.5.2.2 A written report within 30 days to the Commission of:

A. Permanent changes in the facility organization structure.

B. Significant changes in the transient or accident analysis as described in the Safety Analysis Report.

C. Substantial variances of safety related operating characteris-tics from previously predicted or measured values.

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6.6 RECORDS Records of the following activities shall be maintained and retained for the periods specified below. The records may be in the form of logs, data sheets, or other suitable forms. The required information may be contained in single, or multiple records, or a combination thereo f. Recorder charts showing operating parameters of the reactor (i.e., power level, temperature, etc.) for unscheduled shutdown and significant unplanned transients shall be maintained for a minimum period of two years.

6 . 6.1 RECORDS TO BE RETAINED FOR A PERIOD 0F AT LEAST FIVE YEARS A. Reactor operations, including unscheduled shutdowns and tests and experiments performed. Note: Supporting documents such as checklists, log sheets, chart recordings, etc. shall be maintained for a period of at least two years.

B. Principal maintenance operations.

C. Reportable occurrences.

D. Reviews of changes made to the facility or procedures and reviews of tests and experiments performed without prior approvals by the U.S. Nuclear Regulatory Commission pursuant to 10 CFR 50.59.

E. Shipments of radioactive materials.

F. Releases of gaseous and liquid wastes to the environs.

G. Facility radiation and containment surveys.

H. Fuel inventories and fuel transfers.

6.6.2 RECORDS TO Pt RETAINED FOR AT LEAST ONE REQUALIFICATION CYCLE OR FOR THE LEhuTH OF EMPLOYMENT OF THE INDIVIDUAL, WHICHEVER IS SMALLER:

A. Retraining and requalification of licensed operations personnel .

However, records of the most recent complete cycle shall be maintained at all times the individual is employed.

6.6.3 RECORDS TO BE RETAINED FOR THE LIFETIME OF THE REACTOR FACILITY (NOTE: ANNUAL REPORTS MAY BE USED WHERE APPLICABLE AS RECORDS IN THIS SECTION.)

A. Radiation exposure for all personnel monitored.

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7.0 ALARA (10 CFR 50.36a)

The principal routine emission from the UCLA Nuclear Reactor is the argon-41 in the reactor room ventilation exhaust. There is no known biological uptake of argon-41 and exposure limits are based upon external, total body irradiation.

The concentration of argon-41 in the stack effluent is continuously monitored when the reactor is operating, and is normally less than lx10 5 pCi/ml after several hours of full power operation. The annual release is approximately related to the number of equivalent hours of 100 kw operation (kwh per year).

Operations are limited by prior agreement, and by these Technical Specifications, to the equivalent of 5% of the total hours per year at 100 kw. Thus the annual average concentration at the stack release point is no more than 5x10 7 pCi/ml.

The meterological conditions attendant to the stack release are such as to deny the accumulation of a semi-infinite cloud containing 5x10 ' pCi/ml that would exist for any significant period of time. This was confirmed by the TLD program undertaken in 1976 and 1977 (Reference 3) and the plume tracing studies of Mark Phillip Rubin (Reference 4). These matters have been elsewhere summarized (Reference 2) with the conclusion that exposures to individuals were well under the 500 mrem /yr.fcJeral limit.

During the TLD test years (1976 and 1977), utilization averaged 14.5 Wh per year, or very nearly 1/3 of the license agreement limit (effectively 44.0 Nh/ year). In any interpretation of the TLD results, exposure of the general public, due to the argon-41 in the reactor ventilation exhaust and at the maximum agreed operating capacity, would remain well below 500 mrem /yr.

Historically, the ALARA program at UCLA minimizes unnecessary reactor operations by consolidating multiple users into common runs whenever possible. Reactor utilization has been increasing for the last several years, and the opportunity to simultaneously accomodate multiple users has been exploited.

In addition, UCLA plans to install argon-41 holdup-decay tanks to further reduce argon-41 emissions. Preparation of a License Amendment to permit this facility change, will commence upon receipt of information confirming the acceptability of the present application. The Amendment will, among other things, alter the contents of this section of the Technical Specifications.

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8.0 REFERENCES

1. "The UCLA Reactor Is Safe", a statement transmitted to Mr. John F. Ahearne, Chairman, U.S. Nuclear Regulatory Commission by R. R. O'Neill (12-20-79) with one letter by W. Wegst, Director, Office of Research and Occupational Safety (1-30-80).

2. R. D. McLain "UCLA Training Reactor Hazards Analysis",

UCLA Report No. 60-18, March 1960.

3. UCLA Annual Report to the U.S. Nuclear Regulatory Comission for 19/8.

4. M. P. Rubin " Atmospheric Dispersion of Argon-41 from the UCLA Nuclear Reactor", a thesis submitted in partial satisfaction of the requirements for the degree Master of Science in Engineering,1976.

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APPENDIX VI OPERATOR REQUALIFICATION PROGRAM FOR THE UNIVERSITY OF CALIFORNIA AT LOS ANGELES TRAINING REACTOR LICENSE NO. R-71 DOCKET N0. 50-142 February 1980

APPENDIX VI OPERATOR REQUALIFICATION PROGRAM CONTENTS Chapter 1 Exi s ting Program . . . . . . . . . . . . . . . . . . . VI/1-1 Chapter 2 Complementary Programs and Alternatives. . . . . . . . VI/2-1 Attachment A Content and Scheduling . . . . . . . . . . . . . . . VI/A

1.0 EXISTING PROGRAM The existing program is described in the attached letter of August 13, 1975 (Hicks to Rusche).

VI/1-i

2.0 COMPLEMENTARY PROGRAMS AND ALTERNATIVES Monthly operator meetings have been initiated for the principal purpose of critiquing the operating 109 Based upon entries, and the prior months experiences, it is believed that training weaknesses can be identified and corrected via these meetings. The possibility of extending this program into a year-around requalification program is under review.

VI/2-1

APPENDIX VI OPERATOR REQUALIFICATION PROGRAM Attachment A Content and Scheduling from Thomas E. Hicks Director Nuclear Energy Laboratory VI/A

Appendix 6 Operator Requalification Program 50.34(b)(8)

The attached letter specifies the content and scheduling of the UCLA Operator Requalification Program.

February 1980

Augur,t 13, 1975 Bernard Ruseho Director of Nuclear Reactor Regulatient U. S. Nuclear Regulatory Com=ission Washington, D. C. 20555

Dear Sir:

In accordance with Title 10, Code of Federal Regulations, Parts 50 and 55 and following the incorporation of certain suggestions made through personal communications with the NRC staff, we are submitting our revised requalification program for reactor operators and senior reactor operators.

It is our intention to establish a six week lecture serien to augment on-the-job reactor operations and training films which also covers facility design, license and procedural changes. The lecture >

series will be given on an annual basis during the summer when the laboratory work load is at a minimum. The schedule is composed of twelv e , two-hour training periods given twice a week, and will be attended by all reactor operators, senior reactor operators, and trainees. The lecture schedule assigned is as follove.

1. Reactor Physics: nuclear definitions and terminology, radiation, radioactive decay, and mass deiect.

2. Reactor Physics: Fission process, cross sections, far,t and thermal neutron distributions and activation analysir.

3. Reactor Physics: neutron slowing down and diffusic; theory, effective neutron multiplication factor, buckling and reflectors.

4 Reactor Kinetics: reactivity equations, cor; reactivity coefficients (temperature, void, and mass) and control rod theory.

5. UCLA R-1 Reactor Characteristics: technical specificationc and core design.

6. Reactor Plumbing and Ventilation Systet..

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7. Reactor Instrumentation and Control Syste,n

8. Reactor Instrumentation and Control Systems.

9. Health Physics: radiation effects, control, safety and Title 10, Code of Federal Regulations.

10. Procedures: reactor operations, emergency, sample loadirg, rabbit, and refueling.

11. Problem Solving: activation, criticality, sample worth, reactivity, period and control rod worth.

12. Qualification Examination: written and oral walk through exams.

The lecture series and examinations will be conducted by the reactor supervisor. He will delegate specific lectures to other senior reactor operators to be given in their area of expertise.

The examinations will be given and graded by the reactor supervisor and hence he vill be exempt from taking the exams. In the event that a licensed reactor operator scores less than 75* on the qua-lification examinations, he will be removed from all licensed duties for a period of one month. During this period, he will be given the time to study and will be given an expanded qualification examination in the area of his deficiency. Rather than giving a pre-examination, the areas which have the weakest scores on the qualification exams will be stressed more heavily in the next lecture series.

Records will be maintained to document the perfornnnec of each reactor operator and senior reactor operator. The record will include attendance, reactor startups, reactivity control manipu-lations, and the written and oral exams. This will ensure that each reactor operator and senior reactor operator performs at least 10 reactivity manipulations during their license period. Documentation of any additional training and testing for any individual exhibiting unsatisfactory performance will likewise be recorded.

We hope that this program meets with your approval.

Sincerely, Thomas E. Hicks Director Nuclear Energy Laboratory TEH:CEA:v1 Vl/A-2

ML19290E213

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8.0 REFERENCES

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