ML20100D052

Revised SAR for Armed Forces Radiobiology Research Inst Triga Mark-F Reactor

ML20100D052
Person / Time
Site:	Armed Forces Radiobiology Research Institute
Issue date:	09/30/1984
From:	Dougherty M, Hodgdon K, Maria Moore ARMED FORCES RADIOBIOLOGICAL RESEARCH INSTITUTE
To:
Shared Package
ML20100D012	List: ML20100D010 ML20100D030 ML20100D052
References
NUDOCS 8504010062
Download: ML20100D052 (200)

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{{#Wiki_filter:. F 3 lit 10001 Safety Analysis Report for AFRRI TRIGA Mark-F .gg Reactor si DEFENSE NUCLEAR AGENCY ARMED FORCES RADIOBIOLOGY RESEARCH INSTITUTE en-k,5.f, BETHESDA, MARYLAND 20814-5145 q..p,...., 40 g

==onvFn FOR PUBLIC RELEASE: DISTRIBUTION UNLIMITED

a = 7 9 n m = 4 This Safety Analysis Report (SAR) incorporates changes to the original SAR -( submitted 12 May 1981 and approved with the AFRRI relicensing action of 1 August 1984. 'Ihe document was originally prepa:ed by 2_ Dames & Moore ') 7101 Wisconsin Avenue, Suite 700 Bethesda, Maryland The changes reflect AFRRI facility updates made under 10 CFR 50.59 submitted with facility annual reports from 1981 through 1984. 'Ihe September 1984 changes are indicated by a single line in the margin. a r 'Ihis current SAR was prepared by ] M. L. Moore 5 M. N. Dougherty K. M. Hodgdon ]s e i 5 1 ria 5 ill ? J =

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== .wm' 1 A e 1 The AFRRI-TRIG A Mark-F Reactor Technical Specifications identify the i minimum operable systems for reactor operation; reactor operations will conform i -a to the Technical Specifications. All additional design and operational information j me contained in this Safety Analysis Report (SAR)is for information purposes only, g The dimensions, measurements, or other numerical values in this report may differ from measured values as a result of normal construction and manufacturing ) tolerances, or from variations in the normal accuracy of instrumentation. _s n i w a 4 m a ]= C a A

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CONTENTS 2

1.0 INTRODUCTION

AND

SUMMARY

l-1

1.1 INTRODUCTION

1-1 1.2 AFRRI l-1 l.3 AFRRI-TRIGA REACTOR l-1 j 2.0 SITE 2-1 J 2.1 LOCATION 2-1 2.2 POPULATION 2-1 i = 2.3 METEOROLOGY 2-12 d 2.3.1 Regional Climatology 2-12 2.3.2 Long-Term Diffusion Estimates 2-13 l 2.3.3 Short-Term Diffusion Estimates 2-15 1 2.3.4 Severe Weather 2-15 2.4 GEOLOGY 2-17 $= 2.5 HYDROLOGY 2-17 3

2.6 REFERENCES

2-19 3.0 FACILITY 3-1 5 3.1 GENERAL 3-1 3.2 REACTOR BUILDING 3-6 1 3.2.1 Reactor Building Construction 3-6 ~ 3.2.2 Reactor Lui! ding Ventilation 3-7 1 3.2.3 Reactor Room 3-7b } 3.2.4 Reactor Room Confinement 3-10 3.3 REACTOR WATER PURIFICATION AND COOLANT SYSTEMS 3-11 4 3.3.1 Primary Cooling System 3-11 3 } 3.3.2 Secondary Cooling System 3-12 3.3.3 Primary Water Purification System 3-12 i_ 3.3.4 Primary Water Makeup System 3-13 j 3.3.5 N16 Diffuser System 3-13 k 4 3.3.6 Water Systems Instrumentation 3-15 q $r h i Z i 1

CONTENTS (cont'd) 3.4 RADIOACTIVE WASTES 3-16 3.4.1 Gaseous Radioactive Wastes 3-16 3.4.1.1 Nitrogen-16 3-17 3.4.1.2 Argon-41 3-17 3.4.2 Liquid Radioactive Wastes 3-13 3.4.3 Solid Radioactive Wastes 3-24 3.5 SPECIAL NUCLEAR MATERIAL 3-24 3.6 RADIATION MONITORING SYSTEMS 3-25 3.6.1 Remote Area Monitors 3-26 3.6.2 Continuous Air Monitors 3-26 3.6.3 Stack Monitoring Systems 3-31 3.6.3.1 Stack Flow Monitoring System 3-31 3.6.3.2 Stack Particulate Monitoring System 3-31 3.6.3.3 Stack Gas Monitoring System 3-34 3.6.4 Perimeter Monitoring 3-34 3.6.5 Criticality Monitors 3-36 3.7 SUPPORT FACILITIES 3-38 3.7.1 Electrical Power 3-38 3.7.2 AFRRI Stack 3-39 3.7.3 Me' orological Monitoring 3-39 3.7.4 Pressurized Air Supply Systems 3-39 3.7.5 Water, Sewer, and Telephone 3-40

3.8 REFERENCES

3-41 4.0 REACTOR 4-1 4.1 GENERAL 4-1 4.2 REACTOR TANK 41 4.3 LEAD SHIELDING DOORS 4-4 4.4 SUPPORT STRUCTURES 4-5 4.5 CORE 4-6 4.6 GRID PLATE 4-3 4.7 NEUTRON STARTUP SOURCE AND HOLDER 4-8 ii

__5 '55 45 l EE i 5 CONTENTS (cont'd) ^"I i i a s 4.8 IN-CORE EXPERIMENT TUBE (CET) 4-10 L 4-12 [ 4.9 FUEL ELEMENTS 2 4.10 REACTOR CONTROL COMPONENTS 4-15 y i 4.10.1 Standard Control Rods and Guide Tubes 4-15 a 4.10.2 Standard Control Rod Drives 4-15 j r = 4.10.3 Transient Control Rod 4-19 j ~ 4.10.4 Transient Control Rod Drive 4-19 )= a -1 4.11 REACTOR INSTRUMENTATION 4-21 4-22 5 4.11.1 Multirange Linear Channel m 4-22 -S i 4.11.2 Wide-Range Log Channel S i 4-23 4.11.3 High Flux Safety Channels One and Two S 4-24 [ 4.11.4 Fuel Temperature Safety Channels 4.12 ROD WITHDRAWAL PREVENT INTERLOCKS 4-24 4.13 FACILITY INTERLOCK SYSTEM 4-25 5 t J 4.14 SCRAM LOGIC CIRCUITRY 4-27 .-m 4.15 FUEL HANDLING EQUIPMENT 4-29 q 5 4.16 REACTOR PARAMETERS 4-30 !h 4.16.1 TRIGA Fuel Elements 4-30 4.16.2 Excess Reactivity 4-30 4.16.3 Negative Temperature Coefficients of Reactivity 4-31 i ) 4.17 REFERENCES 4-32 j 5.0 EXPERIMENTAL FACILITIES 5-1 5-1 5.1 GENERAL } 5.2 EXPOSURE ROOMS #1 AND #2 5-1 P g = 5-1 5.2.1 Exposure Room Construction 5-4 y 5.2.2 Exposure Room Access L 3.2.3 Exposure Room Monitoring 5-5 q 3 j-5-8 ~1 5.2.4 Patch Panel System o 5.3 PNEUMATIC TRANSFER SYSTEM 5-8 l-5.4 POOL 1RRADIATION 5-14 5.5 ARGON ACTIVATION 5-14 T j iii y

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t s 1 CONTENTS (cont'd) 5.6 PROCEDURES 5-15 5.7 EXPERIMENTS 5-16 6.0 SAFETY ANALYSIS 6-1 6.1 POTENTIAL CONSEQUENCES ASSOCIATED WITH TRIGA REACTOR OPERATIONS 6-1 6.2 ROUTINE OPERATIONAL RISKS 6-2 6.2.1 Handling Radioactive Material 6-2 6.2.2 Reactor Power Transients 6-3 6.2.3 Improper Fuel Loading 6-4 6.2.4 Production of Radioactive Cases in the Reactor Coolant 6-4 6.2.5 Experiments 6-8 [ 6.3 NONROUTINE OPERATIONAL RISKS 6-9 c 6.3.1 Radioactive Contamination of Reactor Shielding Water 6-9 m 6.3.2 Fuel Element Cladding Failure 6-10 r 6.3.2.1 Summary of Previous Experience 6-10 6.3.2.2 Calculation of Maximum Fission Product Release Af ter a Fuel Element ^ Cladding Failure 6-11 6.3.3 Loss of Shielding and Cooling Water 6-14 6.3.4 Design Basis Accidents 6-1) 6.3.4.1 Fuel Element Cladding Failure Accident 6-17 [.

6.4 REFERENCES

6-19 7 7.0 ADMINISTRATION 7-1 ^ 7.1 GENERAL 7-1 7.2 THE OFFICE OF THE DIRECTOR 7-3 7.3 REACTOR AND RADIATION FACILITY SAFETY COMMITTEE 7-4 ? - 7.4 RADIATION SCIENCES DEPARTMENT 7-5 s 7.4.1 The Reactor Branch 7-5 f 7.5 RADIATION SAFETY DEPARTMENT 7-6 7.6 NAVAL MEDICAL COMMAND, NATIONAL CAPITAL REGION 7-6 ~~

CONTENTS (cont'd) APPENDIX A: JOINT WIND FREQUENCY DISTRIBUTION FOR 1960-64 AND 1964 INCLUSIVE A-1 APPENDIX B: ANNUAL AVERAGED RELATIVE CONCENTRATIONS (y/Q) B-1 APPENDIX C: ANALYSIS OF LOSS-OF-COOLANT ACCIDENT FOR AFRRI-TRIGA REACTOR C-1 'm' E M W-- m S 'm a s i

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LIST OF TABLES = c Table Page 2-1 National Naval Medical Center Building Index 2-7 j 2-2 National Naval Medical Center Population Estimates 2-10 3-1 Reactor Remote Area Monitors 3-27 l 3-2 Reactor Room Continuous Air Monitor. 3-32 3-3 Stack Monitoring Systems 3-35 5-1 Prep Area Remote Area Monitors 5-9 I" E E t l L o t s fi n h i i e b It [ vi n

u LIST OF FIGURES Figure Page 1-1 Cut-Away View of the AFRRI-TRIGA Reactor 1-2 2-1 Site Plan 2-2 2-2 Aerial View of the AFRRI Complex Looking Due West 2-3 2-3 Aerial View of the AFRRI Complex 2-4 2-4 Aerial View of the AFRRI Complex Looking South 2-5 2-5 Site Plan--Annuli and Sectors with Respect to the AFRRI Stack 2-6 3-1 AFRR1 Floor Plan 1st Level 3-2 3-2 AFRRI Floor Plan 2nd Level 3-3 3-3 AFRRI Floor Plan 3rd Level 3-4 3-4 AFRRI Floor Plan 4th Level 3-5 3-5 Reactor Building Ventilation System 3-8 3-6 Water Purification and Cooling Systems 3-14 3-7 Liquid Radioactive Waste Drains Located on First Level 3-20 3-8 Liquid Radioactive Waste Drains Located on the Second Level 3-21 l 3-9 Liquid Radioactive Waste Drains Located on the Third Level 3-22 2 3-10 Liquid Radioactive Waste Drains Located on the Fourth Level 3-23 1 3-11 AFRRI Radiation Monitors Associated with AFRRI-TRIGA Reactor--Ist Level 3-28 j 3-12 AFRRI Radiation Monitors Associated with i AFRRI-TRIGA Reactor--2nd Level 3-29 3-13 AFRRI Radiation Monitors Associated with AFRRI-TRIGA Reactor--3rd Level 3-30 < ;$u

LIST OF FIGURES (cont'd) Figure Page 3-14 Perimeter Monitoring System on NNMC Grounds 3-33 3-15 Perimeter Monitoring System Near AFRRI Complex 3-37 4-1 AFRRI-TRIGA Reactor Tank Plan 4-3 4-2 AFRRI-TRIG A Core Support Structure 4-7 4-3 Upper and Lower Grid Plates 4-9 4-4 Neutton Source and Guide Tube 4-11 4-5 Standard Fuel Element 4-13 4-6 Instrumented Fuel Element 4-14 4-7 Transient and Standard Control Rod 4-16 4-8 Standard Control Rod Drive for Safety and Shim Rods 4-17 4-9 Transient Rod Drive Mechanism 4-20 5-1 Exposure Rooms 5-2 5-2 Plug Door to ER No.1 in Closed Position 5-6 5-3 Plug Door to ER No.1 in Opened Position 5-7 5-4 Patch Panel Schematic 5-10 5-5 Pneumatic Transfer System Routing Diagram 5-11 56 Pneumatic Transfer System 5-12 7-1 AFRRI Organizational Chart 7-2 7-2 Reactor Branch Chain of Command 7-7 vill

a c

1.0 INTRODUCTION

AND

1.1 INTRODUCTION

This Safety Analysis Report is an update of the Final Safeguards Report (dated March 1962, with subsequent revisions). This report is submitted as part of the license a renewal process for Facility License No. R-34 for the Armed Forces Radiobiology Research Institute (AFRRI) TRIGA Mark-F reactor in accordance with 10 CFR Part 50. 1.2 AFRRI The Armed Forces Radiobiology Research Institute is located on the grounds of the Naval Medical Command, National Capitol Region (NMC, NCR), Bethesda, Maryland. AFRRI is a tri-service military organization under the Defense Nuclear Agency. The mission of AFRRI is to conduct scientific research in the field of radiobiology and related matters essential to the medical support of the Department of Defense. To cary out this mission, AFRR1 supports and utilizes the TRIGA Mark-F reactor, research laboratories, a 50-MeV electron linear accelerator, a hot cell, office space, an animal research f acility, a 500-KCi Cobalt-60 facility, and a Theratron unit, all of which are contained in the AFRRI complex. t 1.3 AFRRI-TRIGA REACTOR The AFRRI-TRIGA Mark-F reactor was originally developed and installed by the General Atomics Division of the General Dynamics Corporation. The AFRRI-TRIGA [ reactor first achieved criticality in 1962. The reactor is an open pool-type light water reactor which can operate in either the steady state mode up to 1 megawatt (thermal) 5: or pulse mode with a step reactivity insertion of up to 2.So6 A k/k (technical specifications limits) and utilizes standard-design General Atomic fuel elements. The

t AFRRI-TRIG A Mark-F reactor has the unique capability of a horizontally movable core. The movable core allows utilization of a variety of experimental f acilities, such as two separate exposure rooms and the pneumatic transfer system. The reactor and associated experimental facilities and equipment are contained in the Reactor Building located in the AFRRI complex. A cutaway view of the reactor is given in Figure 1-1.

TRIGA reactor designed by General Atomic provide facilities for training, 'i research, and isotope production to universities and research institutions. Many TRIGA reactors, including the AFRRI-TRIGA reactor, have been in operation for 1-1 1

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?.,:. a 'M;- ti.- 4,.. : : : (: ',9 :, q, l.:,.s 1 ~."; Y p /,[a,1. ' { : s.,. -;,. }.[A ^ +. a, FIGURE 11 },.f.,,. $ CUT AWAY VIEW OF THE AFRRl TRIGA REACTOR . i ' r-k'I-' ?,Q.4 +,,;. y.. 4, j .4 l 1-2

over 15 years. The first TRIGA reactor was installed in 1958. The ongoing testing program of standard TRIGA fuel elements by General Atomic and the long experience with a large number of TRIGA reactors have demonstrated the inherent safety of the family of TRIGA reactors. On the basis of the Safety Analysis (Section 6.0) presented in this Safety Analysis Report, it can be concluded that there is reasonable assurance that the AFRRI-TRIGA Mark F reactor can be operated at its present location without undue risk to the health and safety of the AFRRI staff or the general public. 1-3

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i 2 s 4 2.0 SITE = 2.1 LOCATION g The site of the Armed Forces Radiobiology Research Institute (AFRRI), which ] houses the TRIGA Mark-F reactor, is located on the grounds of the Naval Medical Z Command, National Capital Region (NMC, NCR), Bethesda, Maryland. The coordinates of AFRRI are 39000'10" N latitude and 77005'21" W longitude. The Naval Medical Command, National Capital Region is approximately 3 miles north of the ~ Washington, D.C.-Maryland line. The AFRRI site is located on a moderate slope, a declining northward toward a narrow creek valley. The terrain surrounding AFRRI is a rolling with elevations above sea level ranging from 230 to 320 feet. i The location of AFRRI within the Naval Medical Command, National Capital Region is shown on the Site Plan, Figure 2-1. The uses of each building at the NMC, 7 3 NCR are given on Table 2-1. Aerial photographs of AFFRI are also provided in Figures 2-2, 2-3, and 2-4. The location of buildings at the NMC, NCR complex, with respect 7 to the AFFRI exhaust stack, is shown on Figure 2-5. 2.2 POPULATION The Naval Medical Command, National Capital Region has a peak daytime population of approximately 6,095 persons which includes an estimated 1,600 visitors. l The distribution of persons, within the center is shown on Table 2-2, and is summarized ] below: POPULATION DISTRIBUTION

SUMMARY

Distance from Institute (yds) Population 0-200 1,600 j 200-400 3,870 400-700 625 4 Total 6,095 2-1

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TABLE 2-1 NAVAL MEDICAL COMMAND, NA110NAL CAPITAL REGION BUILDING INDEX Building Number Building Use 1 Hospital, Of fices 2 Administration 3 Hospital, Of fices 4 Hospital, Of fices 5 Hospital, Of fices 6 Hospital, Of fices 7 Hospital, Offices 8 Hospital, Offices 9 Outpatient Treatment 10 Nursing Tower 11 Offices 12 Enlisted Males' Quarters 13 Laundry 14 Public Works 15 Maintenance Shops 16 Power Plant 17 Research (NMRl*) 13 Research (NMRI) 19 Greenhouse 20 Fire Station 21 Animal Facilities (NMRI) 22 Flammable Storage 23 Recreation Building 25 Greenhouse 26 Greenhouse 27 Oxygen Storage 28 Research (NMRI) 29 Storage (NMRI) 30 Flagpole 34 Surgeon General's Quarters 35 Of ficers' Quarters 36 Of ficers' Quarters 37 Of ficers' Quarters 38 Of ficers' Quarters = 39 Of ficers' Quarters 42 Reactor, Research (AFRRI) 43 Animal Research Facility (AFRRI) 44 LINAC Modulator (AFRRI) 45 Research, Technical Support (AFRRI) 46 Research, Technical Support (AFRRI) 47 Animal Building (AFRRI)

• Naval Medical Research Institute 2-7

i l TABLE 2-1 (cont'd) i s Building Number Building Use = l 50 Enlisted Males' Quarters $1 Gasoline Service Station l 52 Guest Lodge 53 Research (NMRI) 54 Staff Parking Garage, Offices l 55 Outpatient Parking Garage ~_ 56 Bowling Alley 57 Navy Exchange T 70 Uniformed Services University of the Health Sciences i 71 Uniformed Services University of the Health Sciences 72 Uniformed Services University of the Health Sciences 73 Uniformed Services University of the Health Sciences 119 Research (NMRI) 121 Greenhouse 122 Dental School 123 Dental School, Of ficas 125 Enlisted Females' Quarters 3 139 Machine Shop (NNMC), Research (NMRI) 141 School +_ 142 Research (NMRI), Research and Development Command Offices 143 Storage 144 Eye Clinic 146 Dog Run Area (NMRI) 147 Storage 148 Storage 149 Storage 150 Demolished 151 Demolished 152 Storage 153 Storage t 154 Storage 155 Automotive Maintenance = 156 Demolished 158 Storage (NMRI) f 159 Tennis Courts 160 Tennis Courts 161 Baseball Field 165 Lunchroom (NMRI) 174 Demolished 176 Stotage (NMRI) - Structure Partially Demolished 2 177 Unoerground Tunnel 1 a w 2-8 ie

~' l TABLE 2-1 (cont'd) l Building Number Building Use 180 Vehicle Bridge 181 Pedestrian Bridge = 184 Reactor Cooling Towers (AFRRI) 188. Underground Shelter 189 Vehicle Bridge F 190 Pedestrian Bridge i 200 Helicopter Pad i 202 Switching Station 203 Storage (AFRRI) l 204 Pumphouse (AFRRI) i 205 Warm Water Waste Tanks (AFRRI) E 207 Gas Tanks 208 Gas Tanks l 218 Demolished l 219 Dog Shelter (NMRI) h 222 Switching Station (AFRRI) b 223 Animal Feed Storage (AFRRI) [ 224 Flammable Storage 225 Storage [ 239 Storage i 240 Urology Clinic h 241 Offices 242 Special Services t 247 Bus Stop Shelter 248 Bus Stop Shelter I 251 Environmental Health Lab 252 Cooling Tower g 253 Oxygen Storage e w 255 Meter House h 256 Storage y 257 Recreation Pavilion E 258 Recreation Pavilion 259 Rest Rooms 9-5e ia N L M-

4 Table 2-2 . NAVAL MEDICAL COMMAND,' NATIONAL CAPITAL REGION POPULATION ESTIMATES Building. Population 0-200 yards from AFRRI stack "!p Armed Forces. Radiobiology Research p Institute Complex (Bldgs. 42-47) 240 - Naval Medical Research Institute Complex (. Bldgs.17,13, 21,119,139,146,165)- 300 o d, . Uniformed Services University for the Health Sciences (Bldgs. 70-73) 500 - 15 15 16 - 25 .23- -510 39 10 - g- ^ TOTAL 1,600

200-400 yards from AFRRI stack Naval Medical Command, National Capital Region Hospital Core (Bldgs.. 1-10, 240) 2,800

.11 110 13 25 14 190 ~ 20' 5 !28. 20 -'50 25 52 25 53 100 -54 SS 56i 20 125 175 c,- '141. 130 -142 -50 144: 95 155 5 251-~ 10 . TOTAL 3,870 f 2-10 2

TABLE 2'-2 (cont'd)-

Building Population 400-700 yards from'AFRRI stack Officer Quarters' complex (Bldgs. 34-38)' 30 .12 85 51-15 -37 310- . 122 95 ' 123 85 241 =5 TOTAL 625 k f k.,, 4 E 2-11

(b 2.3 METEOROLOGY j 2.3.l~ Reasonal Climatology ~ The Armed Forces Radiobiology Research Institute lies at the western edge i ~ of the Middle Atlantic coastal plain, approximately 50 miles east of the Blue Ridge JMountains and 35 miles west of-the Chesapeake Bay. The site has a continental

type.of climate, moderated by the proximity 'of the Atlantic Ocean. Summers are p

warm'and humid, and winters are relatively mild. During the winter and spring, the ~ - site lies ~near the principal track of storms that originate over the Gulf of Mexico ^

~and move northeast through the region. From October through June, the prevailing

. winds are from the northwest due to the preponderance 'of anticyclonic circulation over the northern portion of the country. Hence, continental polar air of Canadian origin is the' predominant air mass throughoet. the winter. However, the Appa-lachian -Mountains to the west, which act as a partial barrier to delay the advance 1 of : the cold air,'and the relatively warm Atlantic Ocean to the east help to moderate the winter temperatures in the Washington area. 1 The coldest weather occurs in January, when the mean temperature is 35.6 F

and the. mean minimum daily temperature is 27.7 F. The lowest temperature ever g

recorded in. the metropolitan-area was -15 F, which occurred in February 1899.

Temperatures of 32 F or lower normally occur on an average of 75 days annually, while temperatures.of O F or below usually do not occur.I

' During,the ' summer, as the mean storm track moves far north of the EWashington area, the semipermanent Atlantic High moves ' northward and eastward -

and dominates the circulation of air over the Eastern United States. Consequently, southerly winds prevail ~for much of the summer and transfer moist, tropical air Lfrom the Gulf'of Mexico into the region.

The.. warmest weather occurs in July, when the mean temperature is 78.7 F and the mean maximum daily temperature is 88.2 F. The high'est temperature ever recorded in the metropolitan area'was 106 F, which occurred in July 1930. The j temperature reaches or exceeds 90 F on an average of 37 days each year. ~ Precipitation'is spread out evenly throughout the year, averaging 38.9 inches

of water equivalent. Precipitation in the summer mainly results from thunder-

' storms, and inithe-winter from rain or snow caused by the nearby passage of winter ? storm systems. Annual precipitation rates (in water equivalent) as low as 26.87 . inches (1941) and as high as 57.54 inches (1948) have been recorded. In the entire i 2-12 7J k -pg

• eif'y t-

~ 4-g. -= e-weg. Fw*w,---- 9M-9 -r-i- -w wWwer-1-i.m -w t---rwe ,e v ww wwwrwMv--*

Washington area, the maximum precipitation to occur in 1 month was 17.45 inches during September 1934, and in 24 hours was 7.31 inches during August 1928. Snow, which averages 16.4 inches a year, has occurred in measurable amounts as early as October (1940) and as late as April (1940,1957,1964, and 1972). Annual snowfalls as low as 0.1 inch (1972-1973) and as high as 40.4 inches (1957-1958) have been recorded. In the entire Washington area, the maximum monthly and 24-hour snowfalls recorded _ were 35.2 inches during February 1899 and 25 inches during January _1922, respectively. Snowfalls of 1 inch or more occur (on the average) only 5 days each year, and significant accumulations (10 inches or more) are relatively rare due to rapid melting.l ~ Relative humidity normally averages 64 percent annually, and fog (visibility less than % mile) occurs on an average of 13 days each year.l The joint wind frequency distributions presented in Appendix A were prepared using hourly National Weather Se, We oburvations taken at National Airport, Washington,' D.C., for the period from January 1960 through December 1964. This period was utilized because 1964 was the last year that hourly observations were ~ recorded and archived by the National Weather Service. The ' wind speed and direction were measured at a height of 7.0 meters. The stability is based on observations of wind speed, insolation, and cloud cover.2

The wind data show that the predominant wind direction is from the south with a frequency of 15.6 percent; winds from the northwest occur 9.4 percent of the time. This reflects the seasonal prevailing wind patterns described earlier.

The most. frequent stability is Pasquilt type D stability, which represents neutral conditions. ~ 2.3.2 Long-Term Diffusion Estimates National Airport wind data from the calendar year of 1964 (Table 2, Appendix A) were used to calculate annual average relative concentrations ' (Table 1, Appendix B) using procedures outlined in Regulatory Guide 1.111.3'" These relative concentrations are required t'o assess possible doses that could be received by a hypothetical individual in an unrestricted area, primarily from Argon-41, due to routine radioactive releases. Although the meteorological data were not measured onsite, it represents a conservative assessment of AFRRI diffusion conditions for the following reasons: National Airport and the AFRRI complex are located in the same general vicinity and it would be expected that 2-13 l

meteorological conditions measured at one location would be representative of the other; and the relative concentrations calculated are based on a full year's worth of data, which includes nighttime observations. Since reactor operations at AFRRI primarily occur during daytime hours, the dif fusion of radioactive gaseous effluents ' from AFRRI would be considerably better than the relative concentrations presented. The relative concentrations are calculated assuming a ground level release with a building wake factor given by the 16.3-meter height of the tallest facility at the AFRRI complex. The equation used to determine the annual average relative concentration caused by a ground level source is: n (x/Q)D = 2.032,I NXU;I -l z t=1 i . where 2 IU + 0.5 H 7y)5 E z; z except when E "z., in which case z. 1 1 E "z.' z. 1 1 and the variables are defined as follows: (X/Q)D is the average effluent concentration, x, normalized by the source strength, Q, at distance, X, in a given downwind direction, D; 2.032 is (2/n)N divided by the width in radians of a 22.5 sector; n is the number of hours of valid hourly data in which the wind direction occurs in the opposite sector' from the sector containing the given downwind direction, D; N is the total number of hours of valid hourly de ta (all wind directions); 2-14 s

^X-is the distance in meters downwind of the source; U is :the measured windspeed in meters per second during g hour, i; I . is the vertical plume spread with the volumetric correction z. 8 factor for a release within the building wake cavity, at a distance, -X, for the stability class which occurred during hour, i; o is the -vertical standard deviation of the materials in the z. plume at the distance, X, for the stability class which occurred during hour, i; and H. is the maximum adjacent building height in meters either upwind or downwind from the release point. The maximum annual average relative concentration occurs 100 yards north of the AFRRI stack and is 'l.6 X 10~" seconds per cubic meter. The minimum distance considered in the annual average relative concentration is 100 yards since ~ this is.the distance to the closest building (#158) which is not a part of the 5FRRI complex. 2.3.3 Short-Term Diffusion Estimates The same set of wind data used for the annual averaged relative concentra-tions was used to calculate relative concentrations for potential accident con-sequences, using procedures outlined in Regulatory Guide 1.145, regulatory po-sition' 3.5 These relative concentrations are required to assess possible doses, . caused by Design Basis Accidents (such as fuel element cladding failures), that could be received by a hypothetical individual in an unrestricted area. The 5 percent overall site y/Q relative concentration for a 25 meter circular exclusion area boundary is equivalent to dispersion conditions produced by Pasquilt type F stability with a wind speed.of 1.2 meters per second. Thus, the short-term atmospheric dif fusion estimates used in Section 6.3.4, " Design Basis Accidents," represents a conservative value appropriate for evaluating dose consequences due to a potential accident at the AFRRI facility. 2.3.4 Severe Weather Severe weather in the Washington area is mainly associated with thunder-storms, tornadoes, winter snowstorms, and hurricanes. 2-15

Thunderstorms, which occur on an average of 29 days per year, often bring sudden,' heavy rains and may be accompanied by damaging winds, hail, or lightning. . On June 9,1929, a violent local thunderstorm occurred in the area with wind gusts ~ up to 100 mph..Two hailstorms (April 1938 and May 1953) have occurred in the Washington area which have caused more than-$100,000 damage.I There were 15 m reports of hail 3/4 of an inch in diameter or larger within the 1-degree latitude-j longitude square containing AFRRI during the period 1955 to 1967.5 Three rather destructive tornadoes have been recorded in the Washington 1 area. Two tornadoes (April 1923 and November 1927) caused more than $100,000 in damage, and 'the third (April 1973) caused an estimated $15,000,000 in damage in the Fairfax, Virginia, area.I Tornadoes are relatively rare. Eleven tornadoes were reported within 1-degree latitude-longitude square containing the AFRRI site

during the period 1955 to 1967, giving a mean annual tornado frequency of 0.85 and a recurrence interval. of 1,500. years for a tornado at the specific location of AFR RI.6,7 Tropical disturbances, during their northward passage, occasionally influence Washington's weather, mainly with high winds and heavy rainfall; extensive damage

' from wind and tidal flooding is rare. During the period from 1901 to 1963, 83 tropical storms, hurricanes, and depressions have passed close enough to the Middle Atlantic States (New York to Virginia) to affect various sections of the coast with strong winds, heavy rainfall, or high storm tides. However, of these 83 tropical Ldisturbances, only three made landfall at the Middle Atlantic coast. With the } passage of Hurricane Hazel on October 15, 1954, the. peak wind gust reached ~ ' 98 riph,' but only _1.73. inches of rainfall was recorded. Hurricane Connie, August. 12-13,'1955, produced 6.60 inches of rainfall but the peak wind gust' was only. j

58 mph. - During June 21-22, 1972, Hurricane Agnes produced 7.52 inches of rain,

~ causing 16 deaths in the metropolitan area and $200,000,000 in damage in Virginia, Maryland, Delaware, and the District of Columbia. The." fastest mile" wind speed recorded at National Airport was 78 mph from the southeast at a ll3-foot, elevation on,0ctober.15,1954 during the passage of Hurricane Hazel.I From 1955 to 1967 there were.29 reports of windstorms with . wind speeds of 50 knots (58 mph) or greater within the 1-degree latitute-longitude square contgining AFRRI.' 2-16 I . _,.. - ~ -

2.4 GEOLOGY-- [' The AFRRI reactor site is " situated in the. Piedmont Physiographic Province - [ and is underlain by the Lower Pelitic schist of,the Wissahickon Formation."9 The Wissahickon schist is composed of quartz-mica schists, phyllites, and quartzites.

Quartz forms the bulk of the schists, 30 to 60 percent, while sercite, biotite, and

chlorite are also major constituents.

Test drillings at the AFRRI reactor site and the National Naval Medical

- Center indicate.that the saprofite consists of micaceous silt, clay, and fine-grained ' sand with dense to very dense consistency.' The saprolite has a mottied light gray ~ and brown or dark green appearance and extends. to a depth of 35 to 45 feet. Unweathered bedrock varies from lenses and pods in the-saprofite to a massive. gneissic bedrock at about 45 feet. ' The sorption of soil depends primarily on a large surface area of soil particles per unit volume. Since the percentage of small-sized particles (i.e., clays) is low in the weathered metamorphic rocks in the vicinity of Washington, the ion exchange capacity'is also low (about 25 milliequivalents or less for 100 grams).

In..accordance with Navd'ocks DM-2 (December 1967) and the Uniform -

- Building : Code (1967),' AFRRI is situated in an earthquake zone where the probability for seismic activity is very remote.10 The fault line known to exist I ' nearest AFRRIis approximately 19 miles away and extends from 30 07' N latitude, 77'26' W longitude in Montgomery County'to 39 20' N iatitude, 77 12' W longitude -in Howard County. 2.5 HYDROLOGY ' The main surface water feature'at the National Naval Medical Center is an unnamed second order stream which serves as storm drainage for NNMC. This stream transverses the complex from SW to NE and flows directly into Rock Creek. II The watershed which drains into the unnamed' stream has an area-of 652 acres 'and includes most of the NNMC grounds and nearly all its buildings. About three-

fourths of.the watershed is upstream of NNMC and includes commercial areas and suburban residential neighborhoods in Bethesda, Maryland.

The 100-year flood plain onsite, calculated for the level of urbanization in the area during 1975, is small due to the high slopes in close proximity to the stream.. The flood plain ~ encroaches on no buildings -at NNMC, but does cover

portions;of two parking lots.II 2-17

Groundwater occurs in both the weathered and unweathered rock. In weathered rock it occurs in the interstices between the soil or rock particles, and also in the-openings along joints. Groundwater in the unweathered rock occurs in openings along joints or thinly laminated planes of schistosity. The complex contains no major aquifer.II The silts and clays of the weathered rock have low permeability. However, openings along relief features in the weathered rock increase the permeability locally. The high permeability of the weathered rock near the water table is indicated by rock samples and loss of drilling water in this zone during test drilling. ~ Permeability of the bedrock is low because of smaller openings along joints and lack of openings between the individual grains of the rock. The recharge of the groundwater in the vicinity of the AFRRI reactor site is from local precipitation, which averages 39 inches a year,I but only a small fraction of the annual precipitation enters the zone of saturation. The water table near the AFRRI reactor site is a subdued replica of the surface topography. The depth of the water table measured by test drillings varied from 38.8 to 41.4 feet. Seasonal variation of the water table in the Washington area is typically 5 to 7 feet.9 Shallower water tables are associated with the channels of the onsite stream. The movement of groundwater above the water table is through pores and openings along partings in the weathered rock and is probably in a near-vertical direction. After entering the zone of saturation, most of the groundwater probably moves along a permeable zone in the saprofite near the top of the water table, and is discharged to the onsite stream near the foot of the slope or to small wet weather seeps to the east or west of the site. Because the movement of L groundwater in the bedrock is confined to fracture zones, the. direction of L groundwater movement may vary from the direction normal to the water table contours. l No determination of rate of movement of groundwater has been made. In general, the velocity of groundwater movement through fine-grained materials, such as clay or silt, is very low (on the order of a few feet per year). The velocity of groundwater movement in the saturation zones near the water table, where velocities are at a maximum, may range from less than a foot to several feet per day. l 2-18

2.6 REFERENCES

1.' Environmental-. Data Service, Local Climatological Data, Annual Summary with Comparative Data--Washington, D.C., National Airport, U.S. De-partment of Commerce (published annually through 1979). T ~ 2. - Slade, David H.,' ed., Meteorology and Atomic Energy--1968, Air Resources Environmental Laboratories, Environmental Science Services Admin- ' istration, U.S. Department of Commerce; prepared for U.S. Atomic - Energy Commission (July 1963). 3. - U.S.. Nuclear. Regulatory Commission, Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors, Regulatory Guide 1.111, Revision 1 (July

1977).

4. Sagendorf, J. F., and J. T. Goll, XOQDOQ - Program for the Meteorological ' Evaluation of Routine Effluent Releases at Nuclear Power Stations, U.S. Nuclear Regulatory Commission (September 1977). -5. U.S. - Nuclear Regulatory Commission, Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants, ei - Regulatory Guide 1.143, (For Comment), (August 1979). 6. National Severe Storms Forecast Center, Severe Local Storm Occurrences, 1955-1967, ESSA Technical Memo-WBTM FCST 12, SEIS Unit Staff, . Office of Meteorological Operations (1969). 7. Thom, H. C. S.'" Tornado Probabilities," Monthly Weather Review (October- -December 1963). 8. Cry, G. W., Tropical Cyclones of the North Atlantic Ocean, Technical Paper No. 55, U.S. Weather Bureau (1965). 9. Woodward-Gardner & Associates, Inc., Preliminary Geotechnical Investiga-tion - Uniformed Services University of the Health Sciences and i Hospital Additions, National Naval Medical Center, Bethesda, Maryland !~ -(1974). . 10. - Letter from 3. P. Mason, Project Director, Giffels & Rossetti, Inc., Detroit, Michigan, to H. B. Crawford, AFRRI, contained in the August 28, 1980, application for renewal of U.S. Nuclear Regulatory Commission License No.19-08330-03. 11~ - Department of the Navy, Final Environmental Impact Statement for Uni- ' formed Services University of the Health Services, NNMC, Bethesda, I. Maryland (1975). l 2-19 e + y-p-- + - --:r-, .v,-,na-.w-m. e,,-.

r 3.0 FACILITY 3.I' GENERAL The AFRRI-TRIGA Mark-F reactor is used to study the effects of neutron and gamma radiation on living orggisms and instruments, and to produce radio-isotopes..In addition to the reactor, the AFRRI complex also houses extensive - research laboratories,'a hot cell, a 50-MeV electron linear accelerator (LINAC), a 500 KCi Cobalt-60 facility, a Theratron unit, office space, an animal clinical research facility, and related support areas. Aerial views of the AFRRI complex are shown in Figures 2-2 through 2-4. The AFRRI complex inclu(es six separate primary buildings with seven supporting facilities, as follows: Primary Buildings Support Structures 1. Reactor Building (#42) 7. LINAC Modulator Building (#44) 2. . Laboratory Building (#42) 8. Reactor Cooling Tower (#184) 3. Animal Clinical Research Facility (#43) 9. Waste Water Control System (#205) 4. Laboratory and Technical Support Building (#46) '

10. -Pump House (#204)

~ 5. Laboratory and Technical Support Building (#45) 11. Switching Station (#222)

6. -- Animal Building (#47) 12.

Anim'al Feed Storage Building (#223) 13. Trailer. The six. primary buildings are arranged in an interconnected complex. - Building.#42, which appears from the outside to be one building, is actually two buildings: one building houses the TRIGA Mark-F reactor, and the other building contains office and laboratory space. The fact that there are six separate buildings and that the reactor is contained in its own building is important because the -ventilation system in the reactor building is independent of the ventilation systems for the rest of the AFRRI complex. The floor plan of the AFRRI complex, from the. lowest : level to the highest level, is shown in Figures 3-1 through 3-4, respectively. The building number designations are from Figure 2-1 and Table 2-1. - Access to the entire AFRRI complex is controlled and all personnel are required to enter and exit the facility through the front entrance. All of the exterior doors to the facility are locked from the inside and equipped with 3-1 =

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' intrusion-detection ' alarms. A computerized card-key system restricts access through the front entrance and into the reactor area and records the entrance and exit times of personnel. Visitors must be escorted by permanent AFRRI staff members while in the complex and persons carrying packages and briefcases are subject to seard. An updated Physical Security Plan for the AFRRI-TRIGA Reactor Facility, l prepared in accordance with 10 CFR Part 73 and on file with the U.S. Nuclear Regulatory Commission (Docket Number 50-170) as part of the relicensing process, details AFRRI's procedures and security measures for the physical protection of the AFRRI-TRIGA reactor and its associate'd equipment. As outlined in 10 CFR Part 2.790, the updated security plan contains classified information--DOD instruction 5210.67 (Security Classification For Special Nuclear Material Informa-tion)--and is therefore deemed exempt from public disclosure. 3.2-REACTOR BUILDING The reactor and its as wiated equipment are housed in a single building of reinforced concrete. This building is one of two separate buildings that are designated as Building #42. Figures 3-1 through 3-4 illustrate AFRRI's four levels, the respective working areas, and the exposure rooms. 3.2.1 Reactor Building Construction The exterior walls of the reactor building are constructed entirely of reinforced ordinary -(Portland) cement concrete with no voids in the concrete. There are no exterior windows in the reactor building. The interior of most of these walls are plastered and painted. Most interior partition walls in the reactor building are constructed of Portland cement concrete block in Portland cement mortar, and most of the walls are plastered on one_ side and painted. A mat foundation under the building distributes floor and shielding loads and also provides shielding against potential soil activation. A concrete seal slab is provided under and over contained earth areas used for shielding, which might be subjected to possible activation, to prevent teaching and transport of contaminants to unrestricted areas. The building roof construction consists of lightweight concrete poured over a . corrugated steel form supported by steel roof trusses. The roof is sealed and waterproofed by the application of five-ply, hot-application, roofing compound. 3-6

The lines of contact between the poured light concrete roof and the reinforced concrete-walls are furthers sealed by flashing, expansion joint material, and roofing compound. All reactor building exterior doors are well fitted and weatherstripped with

good-quality,~ commercial, interlocking-type stripping at the top and sides. The bottoms of the doors are provided with' interlocking thresholds.

3.2.2.. Reactor Building Ventilation .The AFRR1 ventilation system ' supplies filtered and conditioned air to the laboratories and offices of the Institute. The ventilation system is a standard industrial type that mixes outside (new) air with recirculating building air, which is then conditioned to meet recognized standards for air quality. The air supplied 'to laboratories is approximately 85% outside air; air supplied to offices is approximately 25% outside air. The supply air is filtered through a 25% roughing filter and an 85% ' final-filter. The cleaned air is then heated or cooled to approximately 550F. Then it is passed by air handlers through supply ducts to terminal devices, which are capable of changing the air to the temperature set by a thermostat in the area to be serviced. The' entire system consists of several-branches that supply and exhaust the different sections of AFRRI. - One branch is of special interest in the control of possible airborne radiation hazards that may arise from actions or operations involving the TRIGA reactor. That branch services select areas within the reactor building. Several smaller branches serve fume hoods and other sources of possible airborne radioactive contamination. The reactor branch and other branches are installed in a parallel manner, using the same supply and exhaust. However, the reactor related area is held negative with respect to other areas within the complex. The area of interest is shown in the block diagram (Figure 3-5). . The select portions of the reactor building included in the branch are the reactor room; the. Exposure Rooms 1, 2, and 3; the warm storage and 3-7

_y. ~ s'D r v 2 decontamination room; and the reactor room suite. The branch is subdivided intoi ~ three parallel legs 'that service these areas. These areas'are supplied air from the - ? building' supply air.1 Air is then single-passed through each parallel leg, exhausted ~ . thro. gh a set.of absolute filters, monitored,:and then released via the AFRRI u l - stack.1 Air from[the' building system 'is supplied to the legs of the reactor branch ~ . through a single supply damper. This damper (which allows air movemen') is held t 4

open only ~as long as there 'is airflow.in the exhaust ducts beyond the absolute

~ f filters'.o A flow-sensitive switch that senses a loss of exhaust flow would send a signal t'o'close -the supply damper, thus stopping the supply of air to the entire - branch. A

To provide isolation capabilities in the event of airborne contamination, the M

reactor -room? air ~ ducts (b'oth (supply and exhaust) are provided.with sealing [ Lddmpers.HWhen closed, these dampers prevent air movement into or out of the-f

reactor room via the ventilation' system.

~ These dampers may 'be.; opened or closed fromL three points: a radiation-3 L sensing device to sample air from the' reactor room; a manual operate button in ~ 'the control room;'and each damper manually.~ The dampers are located so that Jaccess-to them can be achieved without entering the reactor room. The dampers- ~ !will be a?"close on fail" type to insure a safe configuration in the event of system L failure. The location of the control room suite and its proximity to the reactor room

requires air control primarily in the event of an airborne release. Since entry into the reactor room would be through.this area and since action stations would be

established in this area,- the area is. held negative with respect to the rest of

. iAFRRI, tbut. positive ;to a the reactor room. - During normal conditions, air is i circulated parallel : to thei reactor > room; however, during airborne radiation i ': conditions,' the exhaust damper closes with the reactor room dampers. This j L 7a }.. ,l>. +.,,. _ ,-,n- -,n a - n

m e L .causes a buildup of pressure relative to the reactor room, causing leakage into that room.either through the installed conduits or through the entrance door if entry becomes necessary.' To allow any small overpressures from the reactor room to vent in a controlled manner after long-term shudown of the system, small vent lines with ~ atmospheric relief dampers communicate from the reactor room to the stack. ~ Another leg of this branch supplies air to the Exposure Rooms (ER1, 2, and 3)~ and the warm storage-decontamination area. The conditioned supply air is carried ' through ducts to these areas. The exhaust air is drawn through the filter bank and, along _with the other legs, is exhausted through an exhaust stack. - Af ter being pulled from the various serviced areas by fan EF1, the exhaust air .is pulled through a set of absolute filters that insure. stoppage of particulate material' O.3 microns or. larger with a 99.97% eificiency. The filter bank will . nominally. consist of.prefilters, roughing filters, and absolute filters. Filters are tested upon installation to insure that no leakage occurs. The filter buildup -is measured by a differential pressure method. They 'are replaced when pressure - Isok'netic sampling is done on all air reaches manufacturer's recommended levels. i . exiting the stack, and radiological monitoring'is _ then performed. Readouts and - alarms for both_ flow and radiological levels are in the control room, and.are set according to the current limits given'in' appropriate instructions. Normal airflow rate through the exhaust stack is about 35,000 CFM. -3.2.3 Reactor Room The reactor room (Room 3161) contains approximately 1,800 square feet of floor space, has a ceiling height of approximately 18 feet and contains the open surface of the TRIGA Mark-F reactor pool.- Adjacent to the reactor room is the reactor administrative / control room (Room 3160), five of fices (Rooms 3155 through 3159),. and Hall 3106. The only access to these rooms is from Hall 3106 in the reactor adminstration / control area. Large, sealed windows provide visual observa-3-7b ~

W OVER PRES $URE R ELIE F STACX " R g ir MAIN EP 1 ( h 0 ll STAND 5Y' ' EF 2 gg g SUPPLY 04 AlA ISOLATION / DAMPERS AIR DAMPER FLOW SWITCH ir ,r 01 A8 SOLUTE DAMPER 3r FILTER 1r if if gg CONTROL ROOM nOUGHING SulTE. OF FICES REACTOR 8tLTER SHOP. ETC-ROOM l l-> t-l l PREFILTER l Q-hN 0 3 04S$$$! 0:2 l OAMPER { CAMPER _n l 'N l I I I T [ REACTOR nEACTOR LINAC WARM OECON A )k ER 1 ER 2 En: Sion t-i t-l (-i -) i. l _1 1 I A l i OTHER OTHER l RADIOLOGICAL A85 FILTER AREAS FIGURE 3-5 REACTOR BUILDING VENTILATION SYSTEM 38 l i - - - - - ~ - - -,.. - - -. -. - - - -. -...

tion of operations within the reactor room from the reactor control room and Rooms 3156 and 3158. These windows are sealed and cannot be opened. The only entrances to the reactor room are a pair of exterior doors in series, doors from Hall 3106 in the reactor administration / control area, a floor hatch, and a roof hatch. The pair of exterior doors in series are sealed with compressible rubber gaskets. The inner door cannot be opened from the outside. At least one of these airlock doors is closed during reactor operation. All ingress and egress through these airlock doors is controlled and actuates an alarm at the AFRR1 front desk to alert the Security Watch on duty. The double doors from the reactor room to the reactor administration / control area are sealed with a compressible rubber gasket. These doors are normally not left open during reactor operation and are key-locked during nonduty hours. The doors are also automatically locked by a local coded security system mounted on the doors. The security system will unlock the doors when the appropriate number code combination is depressed or when the system is momentarily bypassed by a Reactor Operator from the reactor console when the key lock is open. The number code combination is issued only to reactor branch personnel, their superiors, and other personnel designated by the Director of AFRRI. The local security system provides a means of limiting access to the reactor room and the AFRRI-TRIGA reactor to those personnel essential for reactor operations. The security system does not prohibit personnel from leaving the reactor room during normal duty hours when the key lock is open. l- - The reactor room roof hatch is sealed with a compressible rubber gasket and is secured from within during reactor operation. The reactor room floor hatch is l sealed with a compressible rubber gasket and secured from within the reactor room during reactor operation. Both the floor and roof hatches are equipped with intrusion alarms which will actuate an alarm at the AFRRI front desk to alert the Security Watch on duty. l The double doors from the corridor beside the third floor equipment room to Hall 3106 in the reactor administration / control area serve as the normal entrance to the controlled access area and are sealed with a compressible rubber gasket. These doors are not left open during reactor operation and are automatically locked with a card-key system. To a degree, these doors serve as backup 3-9 l-I'

protection in confining the reactor room air in the event that the other double - doors are inadvertently opened. - These doors are automatically locked by a local ~ ' security system mounted near the doors. Additionally, during nonduty hours they . are key-locked. The local security system will unlock the doors when operated by special ID badges or when the system is m'omentarily bypassed by a Reactor Operator. :Special badges are issued only to reactor branch personnel, their superiors, and other personnel designated by the Director of AFRRI. The security system limits access to the reactor administration / control area to those personnel - essential to reactor operations. The local security system does not prohibit personnel from leaving the reactor administration / control area during normal duty hours when not secured by the key lock. 3.2.4. Reactor Room Confinement ^ The reactor room contains approximately 1,300 square feet of floor space, - has a ceiling height of approximately 18 feet, and a volume of about 32,400 cubic feet. The reactor building construction and ventilation system design allow confinement of the reactor room air volume. The external walls of the reactor room are constructed of reinforced ordinary (Portland) concrete. The roof of the reactor room is made of ordinary concrete poured over a corrugated steel form supported by steel roof trusses. The roof is sealed and waterproofed by the application of roofing compound. The contact surfaces between the concrete roof - and the reinforced concrete external walls are further sealed by flashing, .. expansion joint material, and roofing compound. The interior walls of the reactor - room are formed from ordinary concrete blocks in Portland cement mortar and most of the walls are plastered and painted on one side. The doors and hatches leading to the reactor room are kept closed and locked and are sealed with compressible rubber garlute. The windows between the reactor room and the reactor control room and adjacent offices are sealed and cannot be opened. All penetration in the walls and floor for conduits and piping are also sealed. 3-10 L

S % 3 + . In the event of the release of airborne radioactivity within the reactor room, l radioactivity,will be detected by the reactor room primary continuous air monitor '(CA M). The high' radiation. alarm set point of the reactor room primary CAM will ' initiate automatic closure of the reactor room dampers. The' reactor room is then - confined! The four dampers,Lupon closing, actuate microswitches which complete - light circuits for four indicator lights in the reactor control room. The dampers may Talso be closed or reset (opened) ma'nually from the reactor control room. In' order to accommodate significant changes in barometric pressure or a change ' in the reactor room air temperature that could appreciably alter the pressure in the reactor ; room, a'special normally-closed atmospheric relief valve connects the reactor room directly to the atmosphere. A rise f or fall in pressure sufficient to cause structural damage to:the reactor room would open the atmospheric relief valve, allowing air to be expelled from the room to the atmosphere through a pipe ~ ' attached to the sta'ck. 3.3 REACTOR WATER PURIFICATION AND COOLANT SYSTEMS The systems of the AFRRI-TRIGA Mark-F reactor associated with the reactor. coolant ~are the primary cooling system,' the secondary cooling system, the primary water purification system', the makeup water system for the primary coolant, and the reactor pool Nitrogen-16 (N16) diffuser. The open, nonpressurized reactor pool contains approximately 15,000 gallons of . light, demineralized water. Natural convection' of 'the water in the reactor pool disperses the heat generated by the reactor core. The reactor tank water level is . monitored by a float-activated switch. A drop in the reactor tank water level of 6 ' inches causes an immediate reactor scram, and activates several alarms. 3.3.1.

Primary Cooling System The primary cooling system consists of the reactor tank, the primary pump, the tube side of the shell-and-tube heat exchanger, and associated piping, valves, and fittings,' all situated at elevations above the top of the core. The primary pump

-~ draws ' water from the reactor pool through the suction line, located in the reactor pool approximately 4 feet beneath the pool surface. The primary pump passes the water through the tube side of the heat exchanger at a rate of approximately 350-1 gallons per minute.(gpm). The water is then returned to the reactor pool through r the' return

line,

-located -in. the-reactor . pool 'approximately 3-11 n.. m. ,_n.

8 feet beneath the pool surface. Small holes drilled in the suction and return lines about 4 inches beneath the pool surface prevent water from being syphoned out of I the reactor pool and uncovering the core in the event of a primary coolant line leak or rupture. In the event of significant coolant depletion (below core height), due to a reactor tank leak or rupture, an emergency fill line will be connected from an outside, adjacent hydrant main to the primary coolant loop in order to maintain the reactor tank coolant level above the reactor core. 3.3.2 Secondary Cooling System The. secondary cooling system consists of an enclosed forced-airflow wet tower with sump, a secondary pump, the shell side of the six-pass heat exchanger, and associated piping, valves, and fittings. The secondary pump draws chemically treated raw industrial water from the sump of the cooling tower, passes the water through the shell-side of the heat exchanger at a rate of about 700 gpm, and returns the water to the top of the cooling tower, providing the heat sink to cool the primary water. The water cascades down through the tower, where it is cooled by direct contact with the outside air, and returns to the sump at the base of the cooling tower. The secondary water is chemically treated to reduce losses via evaporation, and hence reduces the amount of makeup water that must be added to ~ the secondary cooling system. 3.3.3 Primary Water Purification System The water purification system removes particulates and ions from the primary cooling water. This minimizes corrosion of all reactor components, reduces activation in the primary cooling water, and maintains the optical clarity of the reactor pool. The primary water purification system consists of a purification pump, water monitor box, prefilter, two mixed-bed demineralizers, and associated piping, valves, and fittings. The purification pump draws a small j portion (<20 gpm) of primary water from the return line in the primary cooling system. From the water purification pump, the water passes through the water monitor box, the pre' filter, and the mixed-bed demineralizers. The water monitor box measures the conductivity, water temperature (inlet), and gamma activity of the primary water before it enters the prefilter and demineralizers. Readout of these variables is provided in the reactor control room. The prefilter removes particles larger than 5 microns in diameter from the primary water. Pressure gauges, measuring the pressure drop across the prefilter, determine when the 3-12

filters need to be changed. From the prefilter, the primary water enters two parallel, mixed-bed demineralizers. The demineralizers maintain the water conductivity level at better than 1 micromho per centimeter. Based on experience, activation of the primary water at this conductivity level is not significant. Each demineralizer has a capacity of 12 gpm and contains about 3 cubic feet of nuclear-grade ion exchange resin. The conductivity, measured at the outlet from each demineralizer, is also available in the control room. The flowrate is also measured at the outlet from each demineralizer with only local readout provided. The water is then passed to the primary cooling system return line. A schematic of the primary and secondary cooling systems and the primary water purification system is shown on Figure 3-6. 3.3.4 Primary Water Makeup System The primary water makeup system replaces any primary coolant water lost due to evaporation from the pool surface. The evaporation of 67 gallons of primary water results in an approximate 1-inch decrease in the pool water level. During normal operations, approximately 80 gallons of primary water are lost each week due to evaporation, depending on the season of the year. The primary water makeup system consists of a straight-feed system (with valving and coupling) which contains a mixed-bed demineralizer and filter system. The system is capable of using raw industrial water as feed water. Before being discharged into the reactor l pool, the primary makeup water is checked to ensure a conductivity at or below that of the reactor pool water. In the event of a primary makeup water system f ailure, the replenishing of the evaporated reactor pool water is achieved via an 80-gallon holdup tank and distillation unit (still) connected via valving and piping to the primary water makeup line. The discharge line from the holdup tank is 5 feet above the reactor pool, which allows the makeup water to be gravity-fed to the reactor pool. This makeup water line is located along the east wall of the reactor pool and discharges the makeup water from above the surface of the pool. 3.3.5 'N ' Diff user System I The N diffuser system reduces the gamma radiation level due to N at the I6 16 16 top of the reactor pool immediately above the core. N is a short-lived, gamma-emitting gas (#7.14-second half-life) produced by the fast neutron irradia-tion of oxygen in the pool water (Ol6(n,p)N16), i I i 3-13

See i V 700 GPen see EMERGENCY PILL LINE Psl J COOLING l TOWER let wate 7 -- -- , bir; w socPm sT0aacE r (10L A TION { j vatve w ? PURIPICafl0N tnt 9m. if it e P451 pgup g g 6.............. 4 y g,, 0 44 0-ee PSI DEMINE R alt 2E R Pla CA RT RIDGE S jaarge now (COMO CELL >',g,,,,g,, I tar

• ts r MAK E 4JP T2

_ CITY MM Pi C i WATER sv E e8 ' WATE R SUPPLY k 'N-Veno "CLOSE O" encept Coupline " OPE N" encept during pooa reptoneshing,p u dueens pool water g PmgstLTER resteneeheng l f 8 See Pli C I SEms m^ ws ST Pall aa y y V STILL [ DisseN l l _ OtmiN] l'3 tn. vtNT Is0Lil l Cosso. CELL l h0M_0 CELL} T K StaCTOR TANK h b] ,tT ~ m m l l l FIGURE 34 WATER PURIFICATION AND COOLING SYSTEMS i l r l I l 1 3-14 l

De N16 diffuser system consists of a pump and motor mounted on the core dolly structure and two sections of Il-inch aluminum piping. Power for the motor 'is controlled from the reactor control room. De two sections of aluminum piping, located inside the cylindrical core support structure, serve as inlet and outlet lines to the pump. De diffuser pump discharges water through .the outlet line at approximately 70 gpm. De outlet line is designed to discharge ! water in a tangential direction inside the cylindrical core support structure. Dis -imparts a swirling motion to the water in the reactor pool which breaks the large gas bubbles into smaller ones, thereby decreasing the buoyance vector and the travel time to escape the reactor pool. We N16 diffuser system is not required - and is utilized at the discretion of the Reactor Operator. .3.3.6 Water Systems Instrumentation Instrumentation in the reactor tank, primary cooling water system, and primary water purification system permits the measurement of parameters important to the safe operation of the reactor and the associated cooling system. De level of the reactor tank water is monitored by a float-activated sw'tch which causes an automatic reactor scram if =6 or more inches of water i are lost below the normal pool full water level. When activated, the switch also j causes a visual alarm on the reactor console and an audible and visual alarm on the -annunciator panel in Hallway 3101. We annunciator panel informs the Security Watch (during off-duty hours) that an unusual reactor situation is present so that appropriate action may be taken. - De primary cooling water temperature is measured at three locations: I ~3-4 feet above the reactor core inside the core shroud (outlet), ~3-4 feet below-the pool surface (bulk), and in the. water monitor box of the primary water purification system (inlet). 'Ihese temperatures are measured by high-temperature-coefficient nickel wires in a bridge circuit. A water temperature readout meter with select switch is provided on the reactor console for these temperature probes. When the bulk water temperature reaches or exceeds 600C (Technical !!Decifications Limit), the bulk water rod withdrawal prevent interlock (described _ further in Section 4.12) prevents the further addition of positive reactivity by locking out all control rod withdrawals. His rod withdrawal prevent interlock _ does not prevent control rod insertion or reactor scram, however. ~ 3-15

'l Primary cooling water. conductivities are measured at several points by conductivity cells containing platinum electrodes in resistance bridges with a range of 0.2 megohm-cm to 20 megohm-cm. Water conductivity is measured at three places in the primary water purification system: the water monitor box (upstream from the mixed bed demineralizers), and at the output from each demineralizer. The monitors provide measurements to ensure the efficient operation of the mixed-bed demineralizers,' as well as the overall conductivity of ~ the pool water. Readout for. the conductivity monitors is located in the control. room and in Room 3152. Under the Technical Specifications, the reactor shall 'not' be operated if the pool water. conductivity exceeds 2.0 micrombos - per centimeter, averaged over a week. The gamma activity of the water in the purification system is measured by a G-M detector in the water monitor box. The readout and visual alarm for ths detector are located in the control room. 3.4 RADIOACTIVE WASTES Radioactive wastes generated by normal reactor operation include radio-active particulates, gases, and solid wastes which consist of routine laboratory ~ wastes and animal carcasses. There are no liquid radioactive wastes generated as a result of normal reactor operation. The handling of radioactive waste materials is conducted under the supervision of the Radiation Safety Department to' ensure compliance with NRC . regulations, specifically 10 CFR Part 20. Utilizing procedures as set forth in the AFRRI Radiation Safety Department's' Health Physics Procedures, the handling and disposal of radioactive waste materials is conducted without endangering the health and safety of AFRRI personnel or the general public. 3.4.1 Gaseous Radioactive Wastes ' Gaseous wastes generated by the AFRRI-TRIGA Mark-F reactor are released to the environment from the reactor building ventilation system through the AFRRI stack. The ventilation system (Section 3.2.2) contains a filter system which has an efficiency of 99.97 percent for 0.3-micron or larger diameter particles. - The. filters may be removed and analyzed for release inventory calculations. Stack monitor systems (Section 3.6) measure the stack flow rate and radioactive particulate and gas effluents which are released to the environ-ment to ensure that the release of Ar 1 and other radioactive particulates and 4 gases meet the requirements of 10 CFR Part 20.106. l 3-16 )

Gaseous radioactive wastes associated with AFRRI reactor operation are The N ' is generated by fast I generally limited to N ' aa.d Ar"I. I neutron . activation of oxygen-(via OI6 (n,p) N ') in the coolant water in the vicinity of the I core. KThe Ar"I is generated by neutron activation of naturally-occuring Ar40,, constituent of air, dissolved in the coolant water and in the air of the exposure h . rooms and the pneumatic transfer system. Other gaseous wastes exhausted from 15 I3 the AFRRI complex bui not generated by the TRIGA reactor are O and N 16 L 3.4.1.1 Nitrogen-16. The amount of gamma radiation caused by N released to l. the environment is insignificant due to the rapid decay (7.14 second half-life) of 2 16-N and the time involved from its production to release. ' During 1 MW(t) steady state, operation, the gamma radiation levels directly above the reactor pool and at the reactor pool chain were measured to be 200 mR/hr and 14 mR/hr, respectively. 16 Because of the amount of gamma radiation present near the pool surf ace from N during large steady state power levels, access to the chained area around the reactor pool.is limited during operations above 100 kW(t). The amount of time 16 .necessary for the N to _be exhausted from the reactor room into the stack is 16 -sdificiently long in comparison with the 7.14 second N half-life to ensure that [' N16. gamma radiation levels will be insignificant at' the release point from' the AFRRI stack. 3.4.1.2 Argon-41. The gaseous waste of primary concern generated by the AFRRI- ?TRIGA reactor is Ar"I. The total amount of Ar released through the AFRRI stack to the environment was measured to be slightly less than 19 curies during all of Calendar Year 1979. This amount of Ar"I (released during 1979) was larger than 7 the' total Ar~ 41 release during any year since 1972. This relative maximum in the l

annual Ar"I release totals is due to the number of reactor operations and the types L

l. of experiments performed during 1979. Based on the present experimental facility l- , operational configuration and anticipated reactor use, it is expected that the total amount of Ar"I_ released during any year will be no more than that released during 1979 (#20 curies) and certainly no. more than twice that released during 1979 (#40 E curies). To calculate the ' annual gamma air dose (due to the release of Ar I) to an i. . individual in an unrestricted area outside the AFRRI complex, the following equation is utilized:I DY 3.17 x 10" (FQ) (Q) (DF ) where: Y -= l ,' 17 4 a

Il DT ~ annual gamma air dose at a point outside the AFRRI complex =- due to Ar"I (mrad /yr) ~ 3.17 x 10" . number of pCi per Ci divided by the number of seconds per-year = 3 annual average relative concentration 2 (sec/m ) -(FQ) = annual release rate of Ar"I (Ci/yr) Q. = DFY- ' gamma air dose factor for a uniform semi-infinite cloud of = 3 Ar"I (mrad-m /pCi-yr). The' total activity released is conservat'ively assumed to be 40 curies during the 3 I year and the gamma' air dose factor for' Ar"I is 9.30 x 10-3 mrad-m /pCi-yr. The maximum annual average relative concentration is calculated in Section 2.3.2 to be .l.6 x 10~" seconds per cubic meter at 100 ya'rds north of the stack. The maximum integrated air dose in an unrestricted area is thus calculated to be 1.9 millirads for the year due to Ar"I. This value is conservative because the calculation does not consider: a reduction in the gamma dose for the actual finite cloud size at that distance; the occupancy factor for normal working hours; and the degree of shielding provided for occupants in buildings located at that distance. The maximum allowable annual dose for all radioactive gaseous effluents released to unrestricted areas which could be occupied by individuals is 500 milli-rads for gamma radiation as specified in 10 CFR Part 20. However, every reasonable effort should be made to maintain radiation exposures and releases of radioactive materials in effluents to unrestricted areas to levels that are as low as is reasonably achievable (ALARA). Design objectives for nuclear " power" reactors have been established in 10 CFR Part 50, Appendix 1, to meet ALARA. The calculated maximum annual air dose due to Ar, which makes up the majority of AFRRI gaseous radioactive wastes, is a factor of five smaller than the design objective of-10 millirads for gamma radiation given in Appendix l to 10 CFR ~ Part 50. The conservatisms in the relative concentration calculations (described in Section 2.3.2) and in the gamma dose calculations provide assurance that the ALARA concept is being implemented effectively. 3.4.2 Liquid Radioactive Wastes There are no liquid radioactive wastes generated by normal reactor opera-tions. The liquid waste disposal system is designed for temporary holding, primarily to provide a means of controlling the release of radioactive liquid wastes 3-18 u

. from the AFRRI compi::x'to the' public sanitiry sewer-systrm. Before being released to the sewer system, the liquid wastes are sampled and analyzed to ensure compliance with 10 CFR Part 20, Section 20.301 The liquid radioactive waste is divided ~into three subsystems or categories: "very hot," " hot," and " warm." The -liquid radioactive waste system consists of seven separate underground tanks and associated drains, piping, and valving. ' The very hot-liquid radioactive waste subsystem consists of one 500-gallon - tank which is filled' manually. The radioactive waste placed in this tank is usually too highly radioactive to-be ' diluted and discharged into the sanitary sewer and therefore must be shipped out in a tank truck for disposal with an NRC-authorized ~ recipient. ;This subsystem has rarely been utilized at AFRRI. The hot liquid radioactive waste subsystem consists of one 5,000-gallon tank with associated drains, piping, and valving. This subsystem collects liquid effluents . from areas in AFRRI in which high-level radioactive activation or contamination -may occur. The drains ~within AFRRI from which this subsystem collects effluents are shown in Figures 3-7 through 3-10. Raw industrial water can be added to the 5,000-gallon tank to dilute the effluent. Compressed air can be applied to the tank . either to enhance the mixing of the raw water and the effluent or to stir the tank prior to sampling. The warm liquid radioactive wa:;te subsystem consists of five 5,000-gallon

tanks with associated drains, grinders, piping, and valving. This subsystem collects liquid -' effluents from those areas in AFRRI in ~ which. low-level radioactive activation or contamination may occur. Prior to the effluent's entry into the waste tanks, the effluent passes through grinders to ensure that all of the waste effluent is broken up into small pieces which can be thoroughly mixed with the water in the

-tanks. The drains within AFRRI from which this subsystem collects effluents are shown in Figures 3-7 through 3-10. Raw industrial water can be added to each of the five 5,000-gallon tanks for dilution; compressed air can be applied to the tanks . either to enhance mixing of the raw water and the effluent, or to stir the tank prior to sampling. All of these five tanks are interconnected with overflow lines so

that these tanks communicate with one another, reducing the likelihood of a spill

-condition. J 4 3-19

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i - 3.4.3 Solid Radioactive Wastes. Solid radioactive wastes consist of routine laboratory wastes, such as glassware, paper, plastics, scintillation vials, disposable gloves, and radioactive animal carcasses. As a result of reactor operations, the same type of solid radioactive wastes are generated due to the handling of experiments. The radioactive solid wastes are temporarily stored in clearly labeled containers located throughout AFRRI and are periodically removed to the Warm Storage Room for long-term storage. The Radiation Safety Department is responsible for the preparation of solid radioactive wastes for shipment to an NRC-authorized recipient either directly or through other military installations in accordance with 10 CFR Part 71 and other NRC and DOT regulations. 3.5 -SPECIAl. NUCLEAR MATERIAL All 'Special Nuclear Material utilized by the AFRRI-TRIGA reactor is contained in the fuel elements (Section 4.9) or in small quantities of fission foils. The Special Nuclear Material is used by AFRRI to st7 port the research mission of the AFRRI-TRIGA reactor in accordance with 10 CFR Part 70.23. The majority of the fuel elements for the AFRRI-TRIGA reactor are located in the reactor core. The reactor pool provides a minimum of 14 feet of pool water for shielding these fuel elements. Fuel elements not being utilized in the reactor core are stored in Fuel Element Storage Racks in the reactor pool around the inside tank perimeter. The reactor pool provides a minimum of 9 feet of pool water for shielding the fuel elements in the storage racks. The storage racks permit sufficient natural convective cooling by the pool water and assure that the fuel elements will not achieve criticality. The handling of fuel elements at AFRRI is supervised by a licensed Senior Reactor Operator. The Special Nuclear Material is the responsibility of the Director of AFRRI. Access to the fuel elements is possible only from the reactor room. Access to the reactor room is controlled by the reactor staff and described in Section 3.2.3. Procedures for safeguarding and ensuring the accountability of the Special Nuclear Material are detailed in the updated Physical Security Plan for the AFRRI-TRIGA ' Reactor Facility, on file with the NRC. 3-24

Monitoring cf the Special Nuclear Mtt rini in acc:rdanca with 10 CFR Part 70.24 is provided by two (i.e., one primary and one backup) gamma-sensitive radiation monitors. The reactor room criticality monitor, which satisfies the requirements under 10 CFR 70, serves as the primary special nuclear material criticality monitor, while RAM R-1 serves as a backup. The reactor room criticality monitor is capable, at all times, of detecting an accidental criticality of the fuel elements. During nonduty hours, the R-1 RAM is also capable of detecting an accidental criticality, if 'AC power is available. If an accidental criticality occurs, these monitors will actuate audible and visual alarms at appropriate locations within the AFRRI complex. A detailed description of these radiation monitors is provided in Section 3.6. 3.6 RADIATION MONITORING SYSTEMS The radiation monitoring systems associated with reactor operations at AFRR1 are maintained as a means of ensuring compliance with radiation limits established under 10 CFR Part 20. These systems consist of remote area monitors, continuous air monitors, cooling water monitors, and AFRRI perimeter monitoring. Detailed information (such as alarm setpoints for the various monitors, appropriate reactor operator responses to radiation alarms, and procedures involving monitor data evaluation and archiving) can be found in References 3 and 4. The radiation monitoring systems associated with AFRRI reactor operations provide readouts and radiation alarms at key locations in the AFRRI complex. These locations are: e Reactor room (Room 3161) e Reactor control room (Room 3160) e Security office (Room 3112) e Health Physics Division office (Room 1420) e Annunciator panel in Hallway 3101. The radiation alarms in the reactor room and the radiation alarm readouts in the reactor control room provide the reactor operators with information necessary for the safe operation of the AFRRI-TRIGA reactor. A general audible and visual radiation alarm in the Health Physics Division office alerts the Radiation Safety Department staff of high radiation levels in the AFRR1 complex. The audible and visual alarms on the annunciator panel in Hallway 3101 alert the Security Watch (present during off-duty hours) of unusual reactor conditions when the reactor is secured. When reactor personnel are present in the reactor control area, the audible alarm on the arcunciator panel in Hallway 3101 is turned off. 3-25

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3.6.1 Remote Area'Mor.1+ ors i

'The remote area monitors (RAM's) in the remote area monitoring system of ~ - primary concern to the reactor are R-1, R-2, E-3, and E-6. These units are placed. Ein various' areas of the reactor building where potential radiation hazards may exist .due to' reactor operation. iThe monitors utilize scintillation detectors which measure gamma radiation

with energies greater than 20 kev.

The units have 'a range of 1ImR/hr.to.105

- mR/hr and a nominal accuracy of 120 percent at all levels. The units have a time ! constant o'f 2 seconds. and a meter and alarm response time of less than I second. - i ~ . The' monitors activate radiation alarms at various locations within AFRRl; the ~ l alarm set points are variable. The-monitors also activate visual alarms in the control room and the Security Office (Room 3112). The RAM's are calibrated at ' regular intervals using a radiation source of known intensity. The locations of the RAM's, the readouts, and the audible and visual radiation alarms are given in Table + ' 3-1 and Figures 3-11 through 3-13. The alarm setpoints can be found in the

appropriate AFRRIinternal documents.4

' 3.6.2. Continuous Air Monitors - The continuous air monitors (CAM's) of primary importance to the reactor are two CAM's located in the reactor room. Three additional CAM's, which monitor the' exposure rooms and the prep area, are discussed in Section 5. The CAM's provide continuous air-sampling and '-monitoring.(gross beta-gamma ' activity) primarily of airborne particulate matter. The CAM's draw-air with an air pump ( 7 cim) through a shielded filter ' assembly, which traps any particulate matter greater than 0.3 microns in diameter. A G-M detector measures any radioactive particulates trapped by the filter. The . count rate (counts per minute) is recorded by a three-cycle, logarithmic, strip- ~ chart recorder mounted on the CAM itself. The units have a sensitivity range of 50 3 cpm to 50 x 10 cpm and a nominal accuracy of 110 percent. The units have a time constant which is inversely proportional to the count rate, being 200 seconds at 50 cpm and I second at 50,000 cpm. The units have the capability of actuating alarms at two adjustable radiation levels. The primary reactor room. CAM is located in the southwest corner of the -reactor room'and is visible from Room 3156. The air sampled by this CAM is taken 3-26 o a aF*

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h TABLE 3-1 REACTOR REMOTE AREA MONITORS u-R- . RAM - Location ~ Readout Radiation Alarm R-1

• Approximately 2 feet Meter in reactor Activates audible and visual 1 1 above the reactor control room and alarm in the reactor room 42 pool surface on the

~ Security Office and in the reactor control ( east side 'of the (Room 3112) room; activates visual alarm

reactor pool in the Security Office (Room 3112) and in the Health Physics Division (Room 1420); activates visual and ~

2-optional audible alarm on annunciator panelin Hallway 3101 R-2 Approximately 7 feet Same as R-1 . Activates visual alarm in the above the floor on reactor control room, . the reactor room in the Security Office (Room . west wall 3112) and in the Health Physics Division j. _(Room 1420) ^ E-3 - 6 feet'above the floor 'Same as R-1 - Same'as R-2. In addition, on the West Wall - a visual local alarm exists prep' area opposite in the prep area near ER #1 ER #1 plug door ~E-6 About 6 feet above the Same as R-1 Same as E-3 except the 1 floor on the West - local visual alarm in the

Wall of the prep prep area is located near

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from approximately 18 inches above the reactor pool surface inside the core support structure. The air is passed through a hose to the CAM. The air is exhausted by the CAM back to the reactor room. The primary reactor room CAM forms an integral part of the reactor room containment capability, in that when the primary CAM high-level alarm is activated, the supply and exhaust dampers to the reactor room in the ventilation system are automatically closed to isolate the reactor room air volume. The backup reactor room CAM is located along the west wall of the reactor room and is visible from the control room and Room 3158. The air sampled by this CAM is taken directly from above the warm drain located along the east side of the reactor pool. The air is exhausted by the backup CAM back to the reactor room. A description of the CAMS' alarms, locations, and readouts is given in Table 3-2 and Figures 3-12 through 3-14. The alarm setpoints can be found in the appropriate AFRRI internal documents." 3.6.3 Stack Monitoring Systems The stack monitoring systems consist of the stack flow monitor, the stack particulate monitor, and the stack gas monitor. These systems provide data about the radioactive effluents discharged through the AFRRI stack. The stack flow monitor and stack gas monitor measurements are recorded by strip chart recorders. 3.6.3.1 Stack Flow Monitoring System. - The stack flow monitoring system measures the average flow rate of air exhausted through the AFRR1 stack. The system consists of a pair of pitot tubes and magnehelic pressure gauges which mechanically measure the dynamic pressure in the stack and produce a proportional electrical signal. A strip chart recorder located in the reactor control room records the stack flow. There are no level alarms associated with this system, except when exhaust fan EF-1 fails, in which case an audible and visual alarm is activated in the reactor control room. 3.6.3.2 Stack Particulate Monitoring System. The stack particulste monitoring system measures the beta-gamma activity emitted by radioactive particulates exhausted from the reactor building into the stack. The system draws a small portion of the air exhausted from the reactor building at the stack through a filter assembly where radioactive particulates greater than 0.3 microns in diameter are trapped. A G-M detector measures e gross beta-gamma activity emitted by the 3-31 ~

m TABLE 3-2 REACTOR ROOM CONTINUOUS AIR MONITORS. I CAM Location of Air intake Readouts - High-Level Alarm . Low-Level Alarm Primary Approximately 18 inches Meter in reactor control Activates audible and Activates visual above reactor pool inside room visual alarm on unit - alarm on the unit core carriage itself -itself Strip chart recorder Activates audible and - located on the unit visual alarm on reactor itself control room annun-ciator panel Activates visual alarm in reactor Y room M Activates audible and visual alarm on annunciator panel in Hallway 3101 Causes solenoid valves to vent and close all ventilation dampers to and from the reactor room Alternate Directly above the warm Strip chart recorder Activates audible and Activates visual (Backup) drain along the east side located on the unit visual alarm on the alarm on the unit of reactor tank itself unit itself itself I lf the low level alarm is being utilized.

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radioactive particulates trapped by the filter. The detector has nominal beta efficiencies of 19 percent and 26 percent for CsI37'and Sr90, respectively. Th: 5 system has a range of 10 cpm to 10 cpm. Normal background is typically about 60 cpm. The air is returned to the stack af ter sampling. The system is capable of activating alarms at two radiation levels. The locations of the readouts and alarms are shown in. Table 3-3. The radiation alarm setpoints can be found in the appropriate ' AFRRI internal documents." ~ 3.6.3.3 Stack Gas Monitoring System. The stack gas monitoring system measures the activity of gaseous radioactive nuclides exhausted through the stack. The principal radioactive gases released from AFRRI through the stack are Ar"I, OI S, and N The AFRRI-TRIGA reactor is the primary source of Ar"I, while the I3 I I AFRRI linear accelerator is the primary source of O ' and N The system samples a small portion of the stack air through a multiported sampling probe centered inside the stack near the top. The air is passed through a filter to remove particulates and then is passed through a gas flow proportional detector system, which measures the beta gamma activity present. The system has a range of 2 6 10 cpm to 10 cpm. Normal background is nominally about 600 cpm. The air sampled by this system is returned to the stack. The counting gas utilized by the detector is propane, which is supplied by two pressurized propane cylinders. The propane, used at a rate of about 0.1 cubic foot per hour, is discharged outside the building directly to the atmosphere. A visual alarm in the reactor equipment room is activated when the propane cylinder pressure drops below about 25 pounds per square inch (psi). This alarm set point allows approximately 3 months' service before the propane cylinders are depleted, ensuring essentially continuous uninterrupted operation of the stack gas monitoring system. A visual and audible alarm in the reactor control room is activated when a low flow condition occurs in the stack gas monitor. The stack gas monitoring system is capable of activating radiation alarms at two levels. The location of the system readouts and alarms are given in Table 3-3. The setpoints for the radiation alarms can be found in the appropriate AFRRI internal documents." 3.6.4 Perimeter Monitoring A perimeter monitoring program is conducted by AFRRI primarily to measure environmental doses received from dionuclides produced by the AFRRI-TRIGA { 3-34 l

f TABLE 3-3 STACK MONITORING SYSTEMS System Readout Radiation Alarm _ Stack Flow Monitoring Strip chart recorder in (Not applicable). System reactor control room; However, EF-1 f ailure local readout in reactor 'gives audible and visual equipment room (Room 3152) ' alarm in reactor control room. Stack Gas Monitoring Meter and strip chart Activates visual alarm System _ recorder in reactor - in reactor control control room and Secur-room ity Office (Room 3112); local readout in reactor equipment room (3152) Stack Particulate - Meter in reactor control Same as Stack Gas Monitoring System. room and local readout Monitoring System in reactor equipment room (3152) 3-35

_~ ^ a reactor, particularly ~Ar"I. However, significant amounts of other gamma-emitting - radionuclides generated by all the radiation sources at AFRRI are also measured. The perimeter monitoring system consists of a large number of stations, each i . equipped with a thermoluminescent dosimeter (TLD), which detects X-ray and gamma radiation from 1 mR to 1,000 R. The TLD has a minimum detectable dose generally below that produced by background radiation. Fading of the TLD is not . significant for the exposure periods used. 1 - The lo ations of the perimeter monitoring system are as follows: i o-At least 30 stations located in unrestricted areas around AFRRI on the National Naval Medical Center grounds (Figure 3-14) e At least six stations located on the AFRRI complex roof, which provide backup data for evaluating releases of gamma-emitting radionuclides released through the AFRRI stack (Figure 3-15) e At least five stations located a minimum of 2 miles from AFRRI, which provide background radiation, measurements. l ~ A detailed description of the perimeter monitoring system can be found in f HPP 2-1, Perimeter Monitoring Program. L 3.6.5. Criticality Monitors t . Criticality monitors are used to detect and measure acute radiation dose f levels, in excess of those measurable with personnel monitors, in the event of a [ - criticality accident occuring in an inhabited work area. There are two types of criticality monitors

• used at AFRRI.

,r [ The reactor room criticality monitor is mounted at the end of the trail cable boom directly above the core carriage. There is only one of these units associated with the reactor. This unit monitors gamma radiation emitted by fuel elements ' stored in the reactor pool as well as the reactor core itself when the reactor is secured. The unit contains an auxiliary D.C. power supply which can provide continuous power for #80 hours in the event of an electrical power outage at the facility. This criticality monitor consists of a scintillation detector and is capable of measuring gamma radiation with energies greater than 20 kev. The reactor 3 room criticality monitor has a range of 1 mR/hr to 10 mR/hr, and a nominal i accuracy of i 20 percent at all levels. The unit is designed to activate a radiation l f alarm, both audible and visual, at an adjustable, preset level. The unit's readout h 3-36 i l t e - fr WWyW 7g =r%ry-%ee-gg---r,% g-

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• PEHIMETER MONITOR $ AT GROUND LEVEL i

FIGURE 315 PERIMETER MONITORING SYSTEM NEAR AFRRI COMPLEX l 3-37

i . meter and radiation alarms ara located in the rcactor contrcl room. Th2 rarct:r room criticality monitor also activates a visual alarm in the reactor control room when a loss of high voltage to the detector or a loss of signal occurs. The alarm setpoint.is given in the appropriate AFRR1 internal documents" and conforms to 10 CFR 70.24. The second type of criticality monitor at AFRR1 is the criticality accident dosimeter which contains various metal foils and other materials which are activated when exposed to large neutron and gamma radiation fields. In case of a criticality accident, the activated materials can be analyzed and the total integrated dose calculated. These dosimeters measure neutron exposures above i rem and gamma exposures above 50 rems. The dosimeters are also capable of measuring exposures within discrete energy levels of the neutron spectrum. The foils and materials utilized in these dosimeters are contained in vials and the units - are mounted on red plaques in the following locations: e North wall of the reactor room o East wall of the reactor room (2) e South wall of the reactor room o Ceiling of the reactor room o East wall of the reactor control room o West wall of Hallway 3106 e East wall of Hallway 3105 e Beside door to 4 arm Storage Room (Room 1120). 3.7 SUPPORT FACILITIES The support facilities at AFRR1 consist of those systems which service the AFRRI-TRIGA Mark-F reactor and its associated facilities. 3.7.1 Electrical Power Electrical power is supplied to AFRRI from the National Naval Medical Center Power System. The power is supplied through the AFRRI substation (Structi.re #222 on Figure 2-1 and Table 2-1) located behind the AFRRI complex. i The electrical power is supplied to the reactor building through two transformers located !n the reactor building. 3-38 i L

One transformer sunplies the low voltage requirements of the reactor building, L such as lighting, receptacles, and the reactor console. A voltage regulator prevents transient power surges from overloading the circuits in the reactor console. The second transformer supplies the high voltage requirements of the reactor building, such as pump motors, the cooling tower f an motors, the air compressors (Section 3.7.4), and the ventilation system fans. 3.7.2 AFRRI Stack The AFRRI stack, located on top of Building #45 (Table 2-1), releases the air exhausted from the reactor building ventilation system (Section 3.2.3). Other ventilation systems in separate AFRR1 buildings are also exhausted through the stack. The stack release point is a minimum of 13 feet above the highest building in the AFRRI complex. The total flow rate through the AFRRI stack is generally about 40,000 cfm during reactor operations. 3.7.3 Meteorological Monitoring A Bendix-Freiz aerovane measures wind speed and direction at the AFRRI complex. The wind sensor is mounted on a tower located on the roof of Building

1. 42. The wind direction and speed are recorded on a strip chart located in the Security Office, Room 3112.

3.7.4 Pressurized Air Supply System The air supply system consists of a compressor, an air drying tower, pressure regulators, and associated piping, valves, and fittings. The air drying tower removes moisture in the pressurized air supplied from the compressor. The air is then passed through pressure regulators which reduce the air pressure to the desired levels. The air supply system provides pressurized air (N9 psi) to the reactor tank lead shield door bearings. The bearings are supplied with pressured air to minimize the ingress of water into the shield door bearing housings, in the event that the bearing seals rupture or leak. The air supply system also supplies air (NSO psi) to the transient rod system and air (N20 psi) through a surge tank and solenoid valve to the reactor room ventilation system dampers. To remain open, 3 39

the pneumatically operated dampers need a supply of pressurized air. When the reactor room is confined, either manually by a reactor operator or automatically by a high level primary reactor room CAM alarm, the solenoid valve vents the supply of pressurized air to each damper, all of which then spring closed . automatically. 3.7.5 Water, Sewer, and Telephone Water, sewer, and telephone requirements are supplied to AFRRI by existing systems at the National Naval Medical Center. 3-40

3.8 REFERENCES

1. U.S. Nuclear Regulatory Commission, Calculation of Annual Doses to Man from Routine Releases of Reactor Effhients for the Purpose of Evalua-ting Compliance with 10 CFR Part 50, Appendix I, Regulatory Guide ' 1.109, Revision 1 (October 1977). 2. U.S. Nuclear Regulatory Commission, Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors, Regulatory Guide 1.111, Revision 1 (July 1977). 3. Armed Forces Radiobiology Research Institute, Health Physics Procedures (HPP5), Radiation Safety Department. 4. Armed Forces Radiobiology Research Institute, Radiation Sources Division Procedures (RSD Procedures), Radiation Sciences Department. 3-41

n 4.0 REACTOR ' 4.1 ' GENERAL' De AFRRI-TRIGA -Mark-F reactor was designed and manufactured by General Atomic and installed at APRRIin 1962. De reactor is a tank-type light . water reactor, with a horizontally movable core. De reactor tank is embedded ..w. In ordinary concrete. De cylindrical core consists of up to 87 standard TRIGA stainless steel elad cylindrical fuel elements with U-ZrH.7 as the fuel matrix 1 material (8.5 weight percent uranium) enriched to less than 20 percent in

Uranium-235; four aluminum-clad borated graphite control rods with solid aluminum followers; and a startup neutron source and guide tube. D e core moderator consists of the zirconium hydride in the fuel and water. De reactor core is' reflected on the top and bottom b'y graphite end plugs in each fuel element, and at the periphery by water. De cylindrical fuel elements and the control rods are positioned in the core in five concentric rings surrounding the centrally located transient control rod. He fuel elements and control rod guide tubes are held in place with top and bottom grid plates. He grid plates are attached to a cylindrical shroud which surrounds the core.

Here are three principal and one optional experimental facilities associated with the AFRRI-TRIGA reactor. Dese are Exposure Room #1, Exposure Room #2, the removable in-core experiment tube (CET), and the Pneumatic Transfer Systems (installed as required). Experiments can also be placed either in the reactor tank or in the reactor core between the fuel elements, but these facilities are not normally used. De AFRRI-TRIGA reactor operates in two beste modest steady state power levels up-to 1 megawatt (thermal), and pulsed operation with a step insertion of up to 2.8% ak/k reactivity, resulting in an average maximum pulse of 28 megawatt-seconds (integrated power).I 4.2 REACTOR TANK De reactor core is positioned in the reactor tank under approximately 16 feet of light, domineralised water. De reactor tank water serves as radiation shielding, neutron moderator and reflector, and reactor coolant. De water puriffeation and coolant systems which servlee the reactor pool water are discussed in Section 3.3. 4-1

The AFRRI-TRIGA reactor tank is constructed of aluminum and is embedded in ordinary concrete. The core is shielded in the radial directions by the reactor tank water and a minimum of approximately 9 feet of. ordinary concrete (with the exception of the exposure rooms). The vertical shielding consists of approximately ~ 16 feet of reactor tank water above the core and approximately 8 feet of ordinary concrete below the core separating the reactor tank from the subsoil underlying i the reactor building. Aluminum was selected as the tank material to improve long-term reliability and to minimize problems of corrosion and neutron activation. The reactor tank, which is cloverleaf-shaped, is approximately 19% feet deep with a distance across the tank lobes of approximately 13 feet (Figure 4-1). The basic wall thickness of the tank is 3/8-inch, except for two cloverleaf projections that extend into the exposure rooms; this. wall thickness is K inch.. The clover-leaf projections which extend into the exposure rooms allow the core to ' face the exposure rooms over a 210 arc in the horizontal plane. The tank bottom and projection shelf thickness is n inch. The caposure rooms are described in Section 5.1. Since the core can be positioned at numerous horizontal locations within the - reactor tank, it is possible to create a variety of radial reflector conditions. The radial reflector ' materials, made up of the following combinations of water, aluminum, and lead, with the core adjacent to Exposure Rooms #1 and #2 and at the middle of the reactor tank are (respectively): Effectively infinite thickness of water and 3/16 of an inch of aluminum e for 180 ; 1 inch of water,3/16 of an inch of aluminum,1 inch of water, 1/4 of an inch of aluminum, a cadmium-gadolinium shield (Section 5.2.1), and from 0 to 6 inches of lead for 180 Effectively infinite water and 3/16 of an inch aluminum for 180 ; 1 inch e of water, 3/16 of an inch of aluminum,1 inch of water, and 1/4 of an inch of aluminum for 180 Effectively infinite thickness of water and 3/16 of an inch of aluminum e for 360". The axial reflector, both above and below the active core, remains unchanged for each of the above cases. It consists of the graphite end plugs in the fuel elements and the pool water above and below the core region. 4-2

IRIMMRR ~ COOLING COOLING WATER OUT WATER IN EDGE OF Pli PLEXIGLA55 SNEET / UNDER GRATING ~ /. _ _.. ..N 1 EDGE OF TANK f y ... _9.. / ~ - - ~ l N l l ) ~~ [ I MECHANICAL CARRit.GE TRACK r ll o l. SW5 j Q V Td l N t -1 ig j CARRIAGE ~T P, ~ r- - - - ~ " 1 AGE Rig __.. .n,, =:.::..sc. . ? e- ' LEAD 5HIELD l h..O. N \\ ',,, 'I. o \\ S I i . :. e s l [ ~; ; ) 68N. LEAD / l \\k j w [ 'Y<,.. II ) J 1 s.

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• l LEAD 5HIELD g

g DOOR DRIVE " v ------------- -TUSES . Q s l ~ l IAS REQUIRED'1 fL._ - ~~ L *C' j p o 1 I! _# IJ 8 \\ ig 3 l 3, LIGMT5 FUEL ELEMENT " ~ ~ f' g) G (.f. ' ~~- i STORAGE RACK / c LEADSHIELD l l DOOR 5-CLOSED FIGURE 4-1 AFRRI-TRIGA REACTOR TANK PLAN

h t The reactor tank water level is monitored by a float-activated switch. A drop of approximately 6 inches in the reactor tank water level causes an immediate reactor scram and activates the following alarms: e Visual alarm on the reactor console Audible (during nonduty hours) and visual alarm on the annunciator e panel in Hallway 3101 (Secthn 3.6). The reactor tank is illuminated by waterproof lights positioned along the wall of the reactor tank. Each lamp consists of a sealed-beam light enclosed in a waterproof aluminum housing and tube, suspended from the top edge of the reactor tank. As in the case of reactor piping, the sealed aluminum tubing supporting the underwater lights is bent to prevent radiation streaming from the reactor tank. 4.3 1.EAD SHIELDING DOORS l Two radiation shielding doors, shown in Figure 4-1, are located in the reactor pool and divide the reactor tank into two equal sections. The shielding doors, when fully closed, allow access to one exposure room without significant radiation f exposures while experiments are taking place at the other exposure room. The interlocking doors are constructed of K-inch aluminum plate and 8-inch aluminum Z-sections. The doors are approximately 19 inches thick, 5 feet high, and 6 feet wide. Each watertight door is filled with approximately 13,000 pounds of //6 lead shot and approximately 90 gallons of transformer oil to fill the gaps between the lead shot. The interlocking end pieces (Z-sections) of the shield doors are stepped to prevent radiation streaming between the doors when closed. Each door is [ supported on a low-friction thrust bearing mounted at the bottom of the tank. f Pressurized air is supplied to the shield door bearings to minimize leakage of water into the shield door bearing housing. l l The shielding doors may be rotated 90 to permit the core support carriage to h move from one end of the reactor tank tu the other. Rotation of the doors is { accomplished using a fractional horsepower drive motor with slip clutch located in L a small pit at the top of the reactor tank. Power for door rotation is transmitted [ through a set of reduction gears. Each shielding door is connected to a reduction [ gear mounted on the side of the carriage track by a vertical shaft extending from the. top of each door. Approximately 3 minutes are required to fully open or close the lead shielding doors. E 4-4 E

Limit switches are used to indicate the fully opened or closed positions of the shielding doors. These limit switches, located on top of the reduction gears, are part of the Facility Interlock System (Section 4.13) which prevents movement of the core support carriage into the mid-pool region unless the shielding doors are fully opened and also denies power to the control rod magnets unless the shielding doors are either fully closed or fully opened. The position of the shielding doors is indicated on the reactor console and a ceiling-mounted TV camera in the reactor room provides an overhead view of the reactor tank in the reactor control room. 4.4 SUPPORT STRUCTURES A four-wheeled carriage, traveling on two tracks that span the reactor tank, is used to move the reactor core laterally frorn one operating position within the tank to another. in addition to supporting the core, the carriage also serves as a 16 support for the four control rod drives, the N diffuser system (Section 3.5), and various electronic devices. The carriage consists of a structural steel framework enclosed by removable aluminum covers. The carriage is approximately 64 inches square by 24 inches high. Four control rod drives are attached to a mounting plate elevated above the carriage. Elevation of the control rod drives assures adequate clearance above the pool water surface and permits direct-line access to the core components and fuel elements. The wheels on one side of the carriage are grooved to match a double-beveled track. Engagement of the wheels and track, therefore, restrains any lateral displacement of the carriage. The two wheels on the opposite side of the carriage are flat-faced and roll on a flat track. The carriage is propelled using a two-speed electric motor and a rack and pinion gear system. The gear rack is attached to the inside of the double-beveled track. The carriage is driven at two speeds, nominally th feet per minute and 2K feet per minute. Low speed is used during the first and last foot of travel as the carriage moves from one extreme limit of travel to the other. The intervening distance is traversed at the higher speed.- Microswitches are employed to automatically change the drive motor's speed, depending on the carriage's position on the track, and to stop the drive motor when the carriage has reached either of its two extreme limits of travel. These switches also form part of the Facility Interlock System (Section 4.13). As a safety measure, mechanical stops are mounted at both ends of the ' carriage track to prevent carriage overtravel at the i i 4-5 IL

i r.' lr extreme limits.1The carriage position is indicated on the reactor console. ' Precise ' repositioning of the carriage may require the use of reference marks on the carriage

tracks. DThe1 variety of core carriage positions is broken down into three general regions. These regions are:

Region 1 - The range of positions within 12 inches of the maximum travel distance of the core dolly carriage at ER #1. . Region 2 - .The range of positions between Region I and Region 3 in which interference between the core shroud and the rotating reactor tank shielding doors could occur.

Region 3 -

. The range of positions within 12 inches of the maximu'm travel distance of the core dolly carriage at ER #2. Movement of.the core carriage. is initiated from the reactor console. Travel 't'ime for the carriage' from one extreme position in the reactor tank to the other is : approximately 5 minutes'. , Power, control wiring, and compressed air are supplied to the carriage through a tra'il' cable and flexible hose. The cable and hose are attached to the carriage control ~ Jrod drive motor enclosure. '.The trail cable and hose are supported by a wall-mounted swinging boom; The: swinging boom also supports the reactor room primary ycontinuous air monitor (CAM) inlet hose ared the reactor room criticality monitor. k .Ti$e ~ core support structure consists of an aluminum cylinder approximately '36 . inches in diameter and 12 feet high, and an aluminum adapter 5 feet high and 191/2 ~ . inches in diameter (Figure 4-2).L Both the cylinder and adapter are formed from 5/16- ~ ~ t Linch thick aluminum plate. The cylinder connects the adapter to the core carriage. e

7. A vertical slot 16 inche's wide extends the full height of the aluminum cylinder at its north.' side.

This slot ~ provides access to the inside of the ' support : structure, II

permitting-~the installation andl removal of core components without raising them above the pool water level.

4.5 CORE ~ u. ' The core forms a right cylinder consisting of a relatively compact array of up to-87: standard TRIGA stainless steel-clad : cylindrical fuel elements, four. control rods and control rod guides, and a startup neutron source and guide tube. The core l is ; -enclosed in a-3/16-inch thick aluminum shroud attached 4 ^ ~

FJ .2 STANDARD CONTROL p \\ g ROD DRIVES (3) CONTROL 200 (RACK AND PINION) J MOTOR ENCLO5URE I! PNEUMATIC. CONTROL ROD ELEC1dOMECNAMICAL DRIVE MOTOR SUPPORT x (TRAN5lENT ROD DRIVE) 'lg ' p-DIFFUSER PUMP AND MOTOR CARRIAGE DRIVE I 6 CARRIAGE COVERS 0008LE BEVELED Q CARRIAGE WHEELS (2) t b p CARRIAGE l / FLAT WNEEL5 (2) GUIDE SUPPORT 0I BARREL GUIDE (3) - l' CORE SU5 PEN 5 TON O LEVELING SCR EWS (5) 0 0 0 0 CONTROL ROD CONNECTING ROD 5 (4) g I( TANGENTIAL FLOW ~ DIFFU5ER ADAPTER N, i CONTROL ROD GUIDE TUBE 5 (4) GAMMA (PULSING) ION CHAMBER UPPER GRID PLATE q l' Fl5530N CNAMBER w 0 FUE L-MOD ERATO R T $NROUD lj / ELEMENT 5 g LOWER GRID PLATE b NEUTRON SOURCE AND GUIDE TUBE CONTROL, ROD TURE EXTEN510NS FIGURE 4-2 AFRRI-TRIGA CORE SUPPORT STRUCTURE 4-7

y ,, to the bottom of the core support adapter. Grid plates bolted to the top and bottom Jof the shroud hold the' fuel elements, the control rod guides, and the neutron source. . uide tube in_ place. ;The active (i.e., fueled) reactorf core is positioned'within the g shroud so tha't its horizontal center line is approximately 29 inches above the l>ottom . of : the ' reactor. tank. Serial numbers inscribed on the core components identify.- ~ ' individual fuel elements and control rods.- J " l 4.61 GRID PLATES - The AFRRI-TRIGA fuel elements and other in-core assemblies are held in place in the core by ' grid plates _(Figure 4-3) which also define the core matrix. Each grid ~ ~ ' plate is bolted to the shroud.- . The. upper grid plate is'18-3/4 inches in diameter, and contains ninety-one (91) I - 1/2-inch' diam'eter holes. Four of-these holes receive the guide tubes for the three g standard control.- rods - and the transient rod. The remaining.87 holes accept fuel [

elements and the CET. These holes, or grid spaces, are located in concentric rings.

The rin'gs are lettered A through F radially from the center. Spaces in each ring are identified ~ numerically in a clockwise. direction from a reference radial line which- ' points;in.the direction of Exposure Room.#1./ Thus,,a particular. space in the grid array,is identified by a letter-number combination, e.g., C-10. Eighteen (18) 1/4-inch diameter holes between;the grid spaces are used for small in-core experiments and i dosimetry. The' upper grid plate also contains one hole near the edge of the grid array - (between F-22 and F-23) for rec'eiving the neutron source guide tube. - The lower grid plate is 16-5/8 incl'es in diameter and is gold-anodized to reduce =l grid plate wear and to' aid light reflection in the core. The lower grid plate contains . eighty-seven.l(87).1/4-inch diamter holes to accept the. fuel element bottom end ! fixtures, four'(4)"1 'l/2-inch diameter holes to accept the control rod guide tubes, and, 2 thirty (30) 5/8-inch diameter ' holes to permit flow of cooling water. Both grid plates ~ tare.made~from 3/4-inch thick' aluminum plates. Each plate is bolted to the shroud by four captive screws.. Correct = positioning of the - grid plates is insured by two-positioning dowels on the grid plate support pads. b

4.7 ' NEUTRON STARTUP SOURCE AND HOLDER

~ ~- '- An Americium-Beryllium (Am-Be)' neutron startup source (cf 3 curies) is used in M~

the. AFRRI-TRIGA reactor.

TheIAm-Be source : produces' neutrons by the( m, n) n-E reaction: w 4-8 u

Eh! 2-r Il=I$ i e E o 4 o o o o g o g 8 .o o o o g E dAh 2 o o o .5;adExU$h o [ O O o o o O.O O Oo o o g. o o g'.3 } } ) O o oO O [O oo o o _C ireI6 Oo oO o O g8520 o oO Oo o O o o o o o o o k o o o Y o o o o o a 5 9o oe { Q IE E o o o w n@ !o! ! ! !l!! !!! lE I pe es ~ =r my [ ca != s Oo !B 0 s = s El a.* .0 !!0 0 E.i = = 0 O O O 00 = rO' = O E 'Oo9 0 0Y n! O i ooc 9 t E 6 E l 4-9

~ 241 237 2"4 95^* N + 93 p (5.5 MeV) 9 a" Be + 2 (CI3 * * + + 6 o" Americium-Beryllium neutron sources are used for applications where small size and constant neutron strength over a relatively long useful life are desired. The Am-Be neutron startup source utilized in the AFRRI reactor yields approx-6 imately 2.1 x 10 neutrons per second per curie and consists of pelletized intimate mixtures of AmO Be in a weight ratio of 1 to 10 and is doubly encapsuled in type 2 304L stainless steel. The neutron source is inserted into the reactor core inside a 7/8-inch diameter tubular source guide tube (Figure 4-4). A hole in the upper grid plate (between F-22 and F-23) near the edge of the gri:1 array receives the neutron source guide tube. When bolted to the upper grid plate, the guide tube projects downward alongside and just outside the fuel array. When the neutron source is inserted into the guide tube, the source material is positioned at the horizontal centerline of the core. The neutron source can be withdrawn from the guide tube for tests or storage by means of a wire attached to the top of the source holder. 4.8 IN-CORE EXPERIMENT TUBE (CET) The CET provides -users an exposure facility with a characteristic high thermal neutron flux level. The CET is primarily used for the production of radioisotopes and the activation of samples for subsequent analysis. The CET is an air-filled guide tube of aluminum with a 1-5/16-inch inner diameter, into which sealed containers, called " rabbits," are lowered. The guide tube is formed from aluminum tubing with a nipple sealed to the lower end, which fits into the lower grid plate in a core fuel element location. The tube extends upward through the upper grid plate at the same grid array position and terminates at the core carriage. The CET has an S-bend above the upper grid plate to prevent radiation streaming. The CET may be positioned in any core fuel element location. The actual use of the CET is subject to limitations under the technical specifica-tions. For example, at no time shall experiments be performed in the CET which would introduce more excess reactivity into the core than the technical specifica-tions allow.

• Excited state; i.e., unstable.

4-10

HANDLING WIRE % i n 6 IN. ,m l l inur! TOP GRID PLATE 3# IN-i a bd h SOURCE GUIDE TUBE ( 7/8 IN. P.D. TUBING X.745 IN l.D.) l l 131/2 IN. ADAPTER l l' RETAINING PIN i A x' MATERIAL SOURCE HOLDER # r y' .70 IN. O.D. L-a SUPPORT PIN FIGURE 4 4 NEUTRON SOURCE! AND GUIDE TUBE 4-11

4.9 FUEL ~ ELEMENTS ' De AFRRI-TRIGA Mark-F reactor utilizes standard TRIGA stainless steel elad, cylindrical fuel elements (Figure 4-5) in which the zirconium hydride moderator is homogeneously mixed with the enriched uranium fuel. De active part of _each fuel element consists' of a cylindrical rod of uranium-zirconium hydride containing 8.5 weight-percent uranium with a less than 20 percent U235 enrichment. He - hydrogen-to-zirconium atom ratio of the fuel-moderator 235 n each standard material is approximately 1.7 to 1. 'Ihe nominal weight of U i fuel eiement is 38 grams. 'Ihe-uranium-zirconium hydride section is approxi-mately 15 inches in length and 1.43 inches in diameter. A solid zirconium rod . 0.18 inch in diameter and 15 inches in lingth is centered in the *uel region of each fuel element to provide structural support. Graphite end plugs,3.44 inches t

in~ length, are located above and below the fuel-moderator section as top and bottom axial reflectors.

Burnable poison (samarium) is included in each fuel element to minimize reactivity changes resulting from fission-product buildup and fuel burnup. De burnable poison is mixed with aluminum to form wafers approximately 0.015-inch a thick. Dese wafers are located between the fuel moderator section and the - graphite reflector end plugs. 4 De fuel-moderator section, the two graphite end plugs, and the burnable poison wafers are contained in a Type 304 stainless steel 0.020-inch thick can. 'Ihe rod is' sealed at the top and bottom with' stainless steel end fittings. An AFRRI-TRIGA fuel element is approximately 28.3 Inches in length and weighs nearly. 7.5 pounds. Up to 87 fuel. elements (86 fuel elements if the CET is in place) may be loaded into the core lattice. In order to monitor the fuel temperature, two instrumented fuel elements ' (Figure 4-6) are placed in the core, one' each in the B-and C-rings. An instrumented fuel element has three chromel-alumel thermocouples embedded in the -fuel-moderator.section at different axial locations. De tips of the .thermocouples are located near the vertical axis of the fuel section. 'Ihe center thermocouple is located at the midplane of the fuel section and the other two ' thermocouples are located 1 inch above and below the center thermocouple. 'Ihe ~ thermocouple leadout wires pass through a soft solder seal contained in the 1/2- - inch outer diameter stainless steeel tube welded to the element's top end fixture. .4-12 ..-.-a. a.-.-.-

l l 5TMNLE55 5 TEEL TOP END FIXTURE STAINLESS STEEL i O TRIFLUTE l-SPACER { g GRAPHITE R 1 ( l l l .015 IN. BURN A8LE l )l POISON S AMARIUM. ALUMINUM \\jy WAFER l 7 db . $... [ URANIUM ZlRCONIUM -l-STAlHLE55 -(t.: HYDRIDE 8 5 WT % STEEL

i. hi-

) URANig'M LESS THAN CLADDING TU8E ?. 20% UI THICKNESS lY-ENRICHMENT 0.02IN.

I u

,j 28.31 IN. 0.18 IN. ZlRCONIUM

1 ROD l

15 IN. t 1' 1 l +- 1.43 IN. + l e--1.47IN. O l i .. ~ 015 IN. SURNASLE l ?O: Poison SAMARIUM ALUMINUM i; WAFER Ii a b STAINLESS STEEL BOTTOM END. 3.44 IN. i \\GRAPHIT E FIXTURE u lj ? u 11 ASSEMBLED Dl5A55EMBLED FIGURE 44 STANDARD FUEL ELEMENT 4-13

m-w- SOFT SOLDER o ALU8AINU88 LEAO*0UT i l i TUSE ( I/2 NI. DIA.) 3 lN. 7 k C TR1At4 GULAR SPACER /h kk i (TRIFLUTE) j - ig j% 3 %,\\ END PLUG GRAPHITE END REPLECTOR SURNASLE P0 50N WAFER ZlRCONIUm R00 26.53 LN. PUEL.N00ERATOR IAATERIAL y o 6.5 IN. TestRae0 COUPLES I3I g 7.5 IN. 3.5 IN. I! t 15 IN. L SURNASLE POISON L " ^ " " NU/ t ia STAINLESS STEEL ~ %l . CLADDING 1 ( j GRAPNtTE ENO REFLECTOR t l I l g FIGURE 44 INSTRUMENTED FUEL ELEMENT l 4-14 t. i

The instrumented fuel elements are identical to the standard fuel elements in all other respects. 4.10 REACTOR CONTROL COMPONENTS = Reactor power in the AFRRI-TRIGA reactor is regulated by using three standard-control rods and one transient control rod, all four of which contain neutron-absorbing - material. Control rod movement - within the core is accomplished using rack and pinion -electromechanical drive for the transient control rod.' 4.10.1 Standard Control Rods and Guide Tubes The shim rod (SHIM), safe rod (SAFE), and regulating rod-(REG) constitute the three standard control rods and are located in core positions D-1, D-7, and D-13, respectively (Figure 4-3). A standard control rod (Figure 4-7) consists of a sealed 'alu'minum' tube (0.065-inch thick) approximately 31 inches long and 1-1/4 inches in . diameter (Figure 4-7). The upper 15-1/4 inches of the tube contain a compacted borated graphite rod (25 percent free boron or boron compounds) as the neutron absorber, or poison. The lower end of the tube (the follower) contains either a 15-11/2 inch long by 1-1/8 inch diameter solid aluminum rod or a hollow aluminum air-filled tube known as an air follower. The control rod guide tubes provide space for inserting and withdrawing the control rods and pass through the upper and lower grid plates and are attached to the lower grid plate. 4.10.2 L Standard Control Rod Drives - Rack and pinion electromechanical drive mechanisms (Figure 4-8), referred to as standard control rod drives, are used to change the positions of the shim, safety, .and regulating rods.- The standard drive consists of a two-phase motor, a magnetic coupler, a rack and pinion gear system, and a potentiometer used to provide an indication of rod position, which is displayed on the reactor console. The pinion gear engages a rack attached to a draw tube supporting an electromagnet. The -magnet' engages an iron armature attached to the end of a long connecting rod. The connecting rod is attached at its lower end to the control rod. The magnet, its draw tube, the armature, and the upper portion of the connecting rod are housed in a tubular barrel. The barrel extends below the reactor pool water surface. The . lower end of the barrel acts as a mechanical stop limiting the downward travel of the control rod assembly. A piston.with a teflon seal is attached to the upper L portion of the connecting rod just below the armature. Since the upper portion of .the barrel is well ventilated by slotted

vents, the piston moves 4-15

WELOING VENT (ONLY ON STANDARD CONTROL RODS) THREADED END CAP--- n di I i.' I L I r I !i 1. a 'l:l: 15.25 IN. BORATED GRAPHITE % (POISON SECTION) 1.12 . DIA. i li..

j L jjjr

% lIin i: ALUMINUM SHELL (1-1/4 IN. O.D. TUBING X.065 W.T.) yl,5

h!?

31.13 IN. u ew o !!b:. ,sll; T:, y '. ;.'. ..;j 15.50 IN. 9 SOLID ALUMINUM OR AIR' ll.! FOLLOWER SECTION 1ll:

d. i

1 [j,i M!n v

FIGURE 4 7 TRANSIENT AND STANDARD CON.'ROL ROD 4-16

FLEXISLE WIRE GUIDE % E MTOR MICR05wlTCNES SPRINGLOADED l PULL ROD POTENTIOMETER g e e f I-ARMATURE e d kl gDRAW TURE I TIP OF PUSN ROD FOOT ELECTROMAGNET SARREL PISTON P CONNECTING ROD 'I I d l ALUMINUM CLAD 80 RATED GRAPHITE CONTROL ROD WITH l SOLID ALUMINUM FOLLOWER i 1 i i I I i L FIGURE 4 8 l STANDARD CONTROL ROD DRIVE FOR SAFETY AND SHIM RODS

• 4-17

'fhesteties Red Dehe SimHer Emeset Drive Motor Contains Techometer

freely in this range, but when the piston is within 2 inches of the bottom of its travel, its movement is dampened by a dashpot action of the grac'ad vents in the lower end of the barrel. This dashpot action reduces bottoming impact during a scram. Clockwise rotation of the motor shaf t raises the draw tube assembly. When the electromagnet, attached to the draw tube, is in contact with the armature and energized, the armature and connecting rod rise with the draw tube, withdrawing the control rod from the reactor core. When the reactor is neutronically shut down (scrammed), the electromagnet is de-energized, releasing the armature. The armature, connecting rod, and the control rod then drop by gravita+ianal force, i reinserting the neutron poison (absorber material)into the core. A ROD DOWN microswitch indicates when the control rod is at its lower limit of travel, i.e., fully inserted into the core. When a standard control rod is fully inserted into the reactor core, a stainless steel washer beneath the armature bears against a

id adjustable foot plate The foot plate, which gojects through a slot in the barrel, is attached to a spring-loaded pull rod that extends vertically up through the drive mount. An adjustable fixture on the upper end of the push rod engages the actuating lever of the ROD DOWN microswitch.

l A second microswitch, the MAGNET E microswitch, is used to indicate the full out position and stop the movement of the electromagnet (drive) when it reaches its upper limit of travel. When the magnet draw tube is raised to its upper limit, the upper surface of the magnet contacts a push rod which extends vertically up through the drive mount. An adjustable fixture on the upper end of the push rod engages the actuating lever of the MAGNET UP microswitch. A third microswitch, the MAGNET DOWN microswitch, is actuated when the electromagnet (drive) is at its lower limit. When the magnet reaches its lower limit, a rigid adjustable fixture at the upper end of the draw tube engages the actuating lever of the MAGNET DOWN microswitch which stops the movement of the electromagnet (drive). The reactor interlock system prevents the simultaneous manual withdrawal of - two or. more standard control rods during steady state modes of operation and j prevents the withdrawal of any standard control rod during pulse operation. ) 4-18 d p. ,,---wey- .-%,y_,v--,,<-m--,%- ,--,vm, -,,,. -w--,_mwm.-- x.,--, ~...-,

~_ 4.10.3 'IYansient Control Rod A fourth control rod, the transient rod, consists of a sealed aluminum tube having the same dimensions as a standard control rod. It is located in core position A-1.' he upper 15-1/4 inches of the tube are filled with a compacted borated graphite rod (25 percent free boron or boron compounds). 'Ihe lower portion' of the tube contains a machined solid aluminum or air follower. 'Ihe transient rod operates in a guide tube identical to those used for the standard i control rods. Both the standard control rods and the transient rod have a maximum travel of 15 inches.. When the control rods are at their upper limits of travel, the neutron poison (absorber material) section of each rod is slightly above the fueled or active region of the core. 4.10.4 'lYansient Control Rod Drive He AFRRI-TRIGA Mark-F is equipped with a pneumatic-electromechanical drive system for the transient control rod (Figure 4-9). The pneumatic-electromechanical drive, referred to as the transient rod drive, is basically a single-acting pneumatic cylinder. A piston within the cylinder Lc attached to the transient control rod by means of a connecting rod. 'Ihe piston i rod passes through an air seal at the lower end of the cylinder. For pneumatic operation, comge cd air, admitted at the lower end of the cylinder, is used to

drive the piston upward. As the piston rises, the air being compressed above the piston is forced out through vents at the upper end of the cylinder. At the end of its stroke, the piston strikes the anvil of a shock absorber; the piston is l

decelerated at a controlled rate during its final inch of travel. 'Ihis action minimizes rod vibration when the piston reaches its upper limit stop. Adjust-ment of the anvil's position, i.e., the volume of the cylinder, controls the piston's stroke length and hence the amount of reactivity inserted during a pulse. An accumulator tank mounted on the core support carriage stores compressed air required to operate the pneumatic portion of the transient rod-drive. A solenoid valve, located in the piping between the accumulator tank and the cylinder, acts as an on/off switch controlling whether or not air is supplied to L' 'the pneumatic cylinder. De-energizing the solenoid valve vents the compressed L air supply and relieves the pressure in the cylinder, allowing the piston to drop by gravity to its lower limit thereby fully inserting the transient rod into the core.

'Ihis design ensures that the transient rod is fully inserted into the core except i

I 4-19

SHOCK' ABSORSER RETAINING NUT I SHOCK ABSORSER s SUPPORT ' 9_ SHOCK A850RSER I AIR YENT 4.- SHOCK ASSORSEE l ANVIL Pl570N l I BALL NUT DRIVE MOTOR THRUST BEARING r% BALL NUT - ' f / l 6 CYLINDER SCREW p (g( M BALL NUT ~ DRIVE GEAR .. SALL NUT - _Y' DRIVE WORM ' I POTENTIOMETER + Al DRIVE WORM ANVIL UP MCR0$ WITCH ~% a, POTENTIOMETER i DRIVE GEAR l ,l l CYLINDER SCREW END CAP i - l l I l "O" RING SEAL 8 i i! Alt SUPPLY !l ROO GUIDE i.' ANYllDOWN SUPPORT + - ROD DOWN MICR05 WITCH MCR0$ WITCH ll !I fl -PISTON ROD lit ROD CONNECTOR. ej a DAMPER CONNECTING ROD 4 FIGURE 4-9 TRANSIENT ROD DRIVE MECHANISM 4-20

i-when compressed air is supplied to the cylinder and the anvil is raised above its . lower limit. The electromechanical portion of the transient rod drive consists of an electric motor, a ball-nut drive assembly, and the externally threaded air cylinder. During electromechanical operation of the transient rod, the threaded section of the air. cylinder acts as the screw in the ball-nut drive assembly. These threads engage a series of balls contained in the ball-nut assembly of 'the drive housing. The ball-nut assembly is then connected through a worm gear drive to 'an electric motor. Therefore, the cylinder and the anvil at its upper end may be raised or lowered independently of the piston and the transient control rod by using the electric drive.when compressed air is not supplied to the cylinder. Conversely, ~ when compressed air is supplied and the electromechanical drive is raised from its lower extreme, the transient rod operates like a standard control rod. A system of microswitches is used to indicate the position of the air cylinder (anvil) and the transient rod. Two of these switches, the (anvil) DRIVE UP and (anvil) DRIVE DOWN microswitches, are actuated by a small bar attached to the bottom of the air cylinder. This bar extends through the rod support guide to prevent rotation of the cylinder. A third microswitch, the ROD DOWN micro-7 switch, is actuated when the piston reaches its lower limit of travel. The transient rod anvil position, measured by a potentiometer on the drive motor, is displayed on the reactor console. 4.11 REACTOR INSTRUMENTATION The AFRRI-TRIGA reactor core is monitored by six detectors. One thermo-couple from each of the two instrumented fuel elements comprise two of the six detectors. A fission detector and three ion chambers comprise the remaining reactor detectors. These six detectors are utilized to provide six independent ' channels" which monitor the power level and fuel temperature of the core. A " channel" is the combination of a detector, interconnecting cables or lines, amplifiers, and output device (s) which measure a specific variable. The six channels utilized in the AFRRI-TRIGA reactor include: the multirange linear channel, the wide-range log channel, high flux safety channel one, high flux safety (; channel two, and fuel temperature safety channels one and two. Some of these channels,-in addition to having readouts on the reactor console, form part of the reactor scram logic circuitry, and the system of rod withdrawal prevent interlocks. l 4-21

\\ ^ 4.11.1' Multirange Linear Channel The multirange linear channel reports reactor power from source level (=10-3 thermal watts) to full steady state power (1 MWt). The output of the fission detector, fed through a preamplifier, serves as the channel input. The multirange linear channel consists of two circuits: the count rate circuit, and the campbelling circuit. For power levels less than 1 kilowatt (t), as selected on the power range select switch, the count rate circuit is utilized. The count rate circuit generates I an output voltage proportional to the number of pulses or counts received from the fission detector. Hence, the output is proportional to the neutron population and the reactor power level. For steady state power levels at or above 1 kilowatt (t), as selected on the power range select switch, the campbelling circuit is utilized. The campbelling circuit generates an output voltage proportional to the reactor power level by a verified technique of noise envelope amplitude detection and measure-ment known as campbelling. The output from the appropriate circuit is fed to an amplifier which supplies a signal to the strip chart recorder located on the reactor console. The power level is scaled on the strip chart recorder between 0 and 100 percent of the power indicated by the power range select switch on the console. The strip chart records this output for all steady state modes of operation but not during pulse operation. 4.11.2 Wide-Ranae Log Channel The wide-range log channel also measures reactor power from source level-(=10-3 ' thermal watts) to full steady state power (l'MWt). The circuitry of this channel is very similar to that of the multirange linear channel. The preamplified output of the same fission detector feeding the multirange linear channel is also fed into the count rate and campbelling circuits of the wide-range log channel. The outputs of these two circuits are log amplified and then summed in a summing amplifier. The summing amplifier supplies a signal to the strip chart recorder located on the reactor console. The power level is indicated on a 10-decade log scale (10'3 watts (t) to 1 MW(t)). The strip chart records this output for all steady state modes of operation but not during pulse operation. During certain steady state modes, the wide-range log channel also measures the rate of change of the power level, which is displayed on the period / log meter located on the reactor console. 4-22

1 The wide-rang 2 1:g channel forms part cf the rod withdrcwal przvent

interlock system.- The channel activates variable set point bistable trips in the rod withdrawal prevent interlock system (Section 4.12) if source level neutrons (:r10~3

thermal watts (t)) are not present, if the reactor power level is above 1 kilowatt (t) when switching to the pulse mode, if a steady state power increase has a period of

- 3 seconds or faster during certain steady state modes, or if high voltage is not supplied to the fission detector. 4.11.3 High Flux Safety Channels One and Two o High flux safety channels one and two report the reactor power level as measured by three ion chambers placed above the core in the neutron field. High flux safety channels one and two are independent of one another but f operate'in identical manners during steady state operation. Each channel consists of an ion chamber placed above the core and the associated electronic circuitry. The steady state power level, as measured by the two high flux safety channels, is

displayed on two separate meters located on the reactor console.

During pulse operation, high flux safety channel one is shunted and the sensor for high flux safety channel two is switched to a third, independent ion chamber placed above the core. High flux safety channel two measures the peak power level achieved during the pulse (NV channel) and the total integrated power produced by the pulse (NVT channel). The NV channel output is displayed on the strip chart recorder located on the reactor console. The NVT channel output is displayed on the reactor console NVT meter. Knobs for each channel, located on the reactor console, allow the channels to be checked for calibration. Switching these knobs to any mode from operate (i.e., to the zero or calibrate positions) causes an immediate reactor scram. Loss of high voltage to any of the ion chambers also causes an immediate reactor scram. i i The high flux safety channels form part of the scram logic circuitry. When j. the steady state reactor power level, as measured by either high flux safety l channel, reaches maximum power levels specified in the technical specifications, a bistable trip circuit is activated which breaks the scram logic circuit, causing an immediate reactor scram. Similarly, when the reactor @wer level during pulse operation, as measured by high flux safety channel two, reaches the maximum pulse power level specified in the technical specifications, a bistable trip circuit is l activated which causes an immediate reactor scram. A trip test knob for each 4-23

safety channel, located on the reactor console, adds additional current to the channel and provides a means of testing the bistable trip circuit without actually exceeding 100 percent of authorized power. Checks and calibrations of these channels are performed periodically in accordance with the technical specifications. 4.11.4 Fuel Temperature Safety Channels Fuel temperature safety channels one and two are independent of one another but operate in identical manners. One thermocouple from each of the two instrumented fuel elements, one in the B-ring and one in the C-ring, provide inputs to fuel temperature safety channels one and two, respectively. The two fuel temperature signals are amplified and displayed on two separate meters located on the reactor console. During pulse operation, the output of fuel temperature safety channel one is also recorded on the reactor console strip chart recorder. The fuel temperature safety channels have internal compensation for the chromel-alumel thermocouples and high noise rejection. The channels also have zero and calibrate modes for checking channel. operation. Switching to either of these two modes from the operate mode results in an immediate reactor scram. In addition to providing information to the Reactor Operator on fuel temperature, the fuel temperature safety channels also form part of the scram logic circuitry. When the fuel temperature, as measured by either fuel tempera-ture safety channel, reaches the maximum allowable fuel temperature specified in the technical specifications, a bistable trip circuit is activated which breaks the scram logic circuit, causing an immediate reactor scram. The operational fuel temperature limit is usually set below the technical specifications limit of 600 C to assure an adequate degree of reactor protection. A trip test knob for each fuel temperature safety channel, located on the reactor console, provides a means of testing the scram capability of each channel without having to actually reach or exceed the technical specifications limit on allowable fuel temperatures. Checks and calibrations of these channels are performed periodically in accordance with the technical specifications. 4.12 ROD WITHDRAWAL PREVENT INTERLOCKS A Rod Withdrawal Prevent (RWP) Interlock stops any upward motion of the ~ standard control rods and prevents air from being supplied to the transient control rod unless specific operating conditions are met. An RWP interlock, however, does 4-24

not prevent a control rod from being lowered or scrammed. Therefore, any RWP interlock prevents any further positive reactivity from being inserted into the core until specific conditions are satisfied. The system of RWP interlocks prevents control rod withdrawals under the following circumstances: RWP prevents air from being applied to the transient rod unless the e-reactor power level is under 1 kilowatt (t) RWP prevents any control rod withdrawal unless, as a minimum, source e level neutrons (=10-3 thermal watts) are present e RWP prevents any further control rod withdrawal unless the power level is changing on a 3-second or longer period as measured by the wide-range log channel during certain steady state operations RWP prevents any control rod withdrawat unless high voltage is being e supplied to the fission detector for the multirange linear and wide-range log channels RWP prevents any control rod withdrawal if the wide range log channel e is in any mode (i.e., position) other than OPERATE e RWP prevents any control rod withdrawal unless the bulk pool water temperature is less than 60 C (Technical Specification Limit). 4.13 FACILITY INTERLOCK SYSTEM The Facility Interlock System is designed to eliminate the possibility of accidental radiation exposure of personnel working in the exposure rooms or the preparation area, and to prevent interference (i.e., contact or impact) between the reactor tank lead shielding doors and the reactor core shroud. These interlocks prevent rotation (i.e., opening or closing) of the reactor tank shielding doors and the operation and movement of the reactor core between dif ferent regions unless specific operating conditions are satisfied. The system logic depends on the positions of the reactor core, the reactor tank shielding doors, and the plug doors to the exposure rooms. The reactor core positions are classified into three regions: Region 1 - The range of positions within 12 inches of the maximum travel distance of the core dolly carriage at ER #1 4 25

Region 2 - The range of positions between Region 1 and Region 3, in v+.ich interference between the core shroud and the rotating reactor tank shielding doors could occur l Region 3 - The range of positions within 12 inches of the maximum travel distance of the core dolly carriage at ER #2. Tlie Facility Interlock System prevents the following operations unless the specific conditions given are satisfied: e The reactor tank shielding doors cannot be closed unless the reactor core is in either Region 1 or Region 3 e The reactor tank shielding doors cannot be opened unless the reactor core is in either Region 1 or Region 3 e The reactor tank shielding doors cannot be opened to allow movement of the reactor core through Region 2 to an exposure room unless a warning horn has been sounded in that exposure room or unless two licensed Reactor Operators have visually inspected the exposure room to ensure that, prior to closing the exposure room plug door, no condition exists that is hazardous to personnel The reactor cannot be operated unless the reactor tank shielding doors e are either fully cpened or fully closed e The reactor cannot be operated within Region 1 unless the plug door to Exposure Room #1 is fully closed and the reactor tank shielding doors are fully closed er, if the reactor tank shielding doors are fully opened, then both exposure room plug doors must be fully closed e The reactor cannot be operated within Region 2 unless the plug doors to . both exposure rooms are fully closed and the reactor tank shielding doors are fully opened The reactor,cannot be operated within Region 3 unless the plug door to e Exposure Room #2 is fully closed and the reactor tank shielding doors are fully closed or, if the reactor tank shielding doors are fully opened, then both exposure room plug doors must be fully closed e The reactor cannot be moved into Region 2 unless the reactor tank shielding doors are fully open. 4-26

y 1 s Both' exposure rooms are equipped with lighted " emergency stop" buttons.- An p emergency stop button is also located on the reactor console.~ - Pressing any.of i these emergency stop buttons causes an immediate reactor scram and gives a.. scram indication to the reactor operator at the console. Magnetic power to the

~

' standard control-rods and the air supply to the transient rod cylinder cannot be. obtained for 'an ' emergency -stop scram condition without resetting the reactor console key which, when performed, automatically initiates a time delay with horns sounding in both exposure rooms.. It also reinitiates the requirements for opening l the,-lead shield : doors. The emergency stop circuit, therefore,' provides an - independent means for an individual accidentally trapped in an exposure room to - prevent. an unsafe = condition -Involving operation from occurring, while.also : providing a positive -indication to the reactor operator that someone could be - trapped in an exposure room. -4.14 SCRAM LOGIC CIRCUlTRY The scram logic circuitry assures that a set of reactor core and operational J - conditions.must be' satisfied for reactor operation to. occur or continue !in accordance' with the technical specifications. The scram logic circuitry involves a set.of open-on-failure logic relay switches in series: any scram signal or . component failure in the scram logic, therefore, results in a loss of standard control rod magnet power and a loss of air to the transient rod cylinder, resulting in a reactor scram.' The. time between activation of the scram logic and the total insertion of the control rods is limited by the technical specifications to assure the safety of the reactor and the fuel elements for the range of anticipated transients i = for the AFRRI-TRIGA reactor. The scram logic circuitry causes an automatic l. reactor scram under the following circumstances: o The steady state timer causes a reactor scram af ter a given elapsed time, as set on the timer, when utilized during steady state power operations o The manual scram bar, located on the reactor console, allows the Reactor l Operator to manually scram the reactor -o Movement of the console key to the OFF position causes a reactor scram. l-4-27 .,.,..,.,--~--.._-,._,n,,, .-,-..,,._,.,-n n -,-n.n,,,--m,,,,,--,. w.._,s n.-n.,. -n-,---.,

F .) The reactor tank shielding doors in any position other than fully open or o fully closed will cause a reactor scram g I o. Activation of any of the emergency stop buttons in the exposure rooms and on the reactor console causes a reactor scram , Any of the safety channels (fuel temperature safety channels and high o flux safety channels) in any position other than OPERATE (i.e., ' CALIBRATE or ZERO) causes a reactor scram o l A loss of AC power to the reactor causes a reactor scram o High flux safety channel one causes a reactor scram at a reactor power level specified in the technical specifications for steady state modes of operation o.High flux safety channel two causes a reactor scram at a reactor power level specified in the technical specifications for steady state and pulse - operations A loss of high voltage to any of the detectors for high flux safety channels o one and two causes a reactor scram Fuel temperature safety channels one and two will each initiate a reactor o scram if ' the_ fuel temperature, as measured independently by either channel, reaches 6000C (technical specification limit). This assures that the AFRRI safety limit (core temperature) of l,0000C for AFRRi stainless. steel clad cylindrical TRIGA fuel elements, as stated in the AFRRI technical specifications and testing conducted by General . Atomics, is never approached or exceeded. The actual operational limit for the fuel temperature safety channels may be set lower than the technical specifications limit of 6000C A loss of. reactor pool water which leaves less than or equal to 14 feet of o pool water above the core (technical specifications limit) causes a reactor scram. The actual operational limits for the pool water level may be set more conservatively than the technical specifications limit. 4-28

./ s 4.15 FUEL HANDLING EQUIPMENT The fuel handling tool is used to grasp, move, and position fuel elements . within the reactor tank. The all-stainless steel tool consists of an end grapple mechanism, flexible steel control cable, and the control handle. A fuel element is engaged by lowering the tooi over the top end fixture of a fuel element when the control handle is closed. Four stainless steel balls, protruding through the inside cylinder of the grapple mechanism, slip into a groove on the fuel element top end fixture. The cylinder slides with reference to the rest of the grapple mechanism, holding the balls in place. The fuel element thus becomes locked in the fuel handling tool. To release the fuel element, it is necessary to unlock (i.e., open) the I control handle at the top of the control cable. This action forces the balls back into the grapple mechanism, which in turn releases the fuel element from the end . grapple. The fuel element inspection tool is used to measure overall length and bow of a fuel element. The tool is designed to permit inspection of a fuel element while the element is submerged below 9 feet or more of water. The inspection tool consists of a tubular "go/no-go" gauge for detecting fuel element lateral bending (bow) and a device with a dial indicator for measuring fuel element length. The ~ I' tool is held in place and positioned in the reactor tank by a supporting aluminum structure. The fuel element inspection tool may be placed within the core support structure or at other convenient locations within the reactor tank. Six fuel storage racks permit temporary storage of up to 12 fuel elements per storage rack within the reactor tank, submerged a minimum of 9 feet beneath the ' pool water surface. Each rack is supported by two aluminum rods, which are . attached by brackets to the aluminum channels at the top of the reactor tank. Use of the fuel storage racks allows core fuel loading changes to be made without the use of fuel handling casks. The fuel storage racks assure that stored fuel will not become critical and will not reach unsafe temperatures. Conservative calcula-2 tions show that in the event of a fully loaded storage rack failure where all 12 fuel elements fall to the bottom of the reactor tank in an optimal (i.e., worst case) neutronic geometrical configuration, a criticality excursion would not result. 4-29

4.16 REACTOR PARAMETERS 4.16.1 TRIGA Fuel Elements --De AFRRI-TRIGA Mark-F reactor core can accommodate up to 87 standard 4. design TRIGA stainless steel clad, cylindrical fuel elements. De fuel elements - are positioned in concentric rings to form a cylindrical, active (i.e., fueled) core region approximately 17 inches in diameter and 15 inches high. De TRIGA fuel elements, described in Section 4.9, are summarized below: Maximum number of fuel elements 87' in core Fuel-moderator material 8.5 wt-percent uranium homogenously combined with 91.5 wt-percent zirconium-hydride Hydrogen to zirconium atom ratio Nominal 1.7 H to 1.0 Zr Uranium enrichment ' Less than 20% U235 235 oading per fuel element Approximately 38 grams i U l Fuel element cladding 304 stainless steel, nominally 0.020 inches thick Burnable poison Samarium-aluminum wafers which form an integral part of the manu-factured fuel element Axial reflector Graphite end slugs which form [ an integral part of the manufac-tured fuel element. 4.16.2 Ereess Reactivity i De core excess reactivity is measured each day a reactor operation involving' the withdrawal of a control rod 'is planned. He reactor loading is limited by the technical specifications to a maximum of 3.5% Ak/k ($5.00) excess reactivity above cold critical, with or without all experiments in place. Single fuel element worths are approximately 0.90% Ak/k and 0.20% Ak/k in the B-ring and the F-ring, respectively. De sum of the absolute' reactivity worths of all e experiments in the reactor and the associated experiment and exposure facilities I is limited by the technical specifications to 2.1% Ak/k ($3.00). l-De net change in reactivity caused by the maximum operational withdrawal of a standard control rod is approximately 1.4% Ak/k ($2.00). His amount of l 4-30 l

reactivity is due to the complete withdrawal of the horated graphite section of the standard control rod from' the core and the consequent insertion of the aluminum follower into the core. The net change in reactivity caused by the maximum operational withdrawal of the transient control rod is approximately 2.8 % Ak/k ($4.00). This amount of reactivity is due to the complete withdrawal of the borated graphite section of the transient control rod and the consequent insertion of the aluminum follower into ~ the core. The net change in reactivity caused by the complete physical removal of the transient control rod from the core is approximately 3.0 % ak/k ($4.30). The reactivity worths of each control rod and the shutdown margin are determined at periodic intervals specified in the technical specifications. 4.16.3 Negative Temperature Coefficients of Reactivity ' The safe operation of the AFRRI-TRIGA Mark-F reactor during high power steady state and pulse operations is ensured by the reactor's inherent large steady state and prompt negative temperature coefficients of reactivity, respectively, due to intrinsic characteristics of the uranium zirconium-hydride fuel during heatup. Bec'ause of the large prompt negative temperature coefficient, any large (>$1.00) step insertion of excess reactivity above cold critical will be immediately and automatically compensated for by the U-ZrH, terminating the excursion auto-matically without any dependence on the electronic or mechanical reactor systems or_ the reactor operator. Similarly, because of the large steady state negative ' temperature coefficient of reactivity, any large insertion of excess reactivity during steady state operation will be automatically compensated for by the U-ZrH, ~ hus limiting the reactor steady state power level. t l These intrinsic negative temperature coefficients of reactivity (i.e., prompt and steady state) are primarily due to the hardening of the neutron spectrum caused by heating the U-ZrH fuel. As the power level of the reactor is increased, the fuel temperature also increases. This increases the probability that thermal neutrons will ~ gain energy from excited hydrogen atoms in the ZrH. Thus, the -relative number of neutrons escaping from the fuel or being absorbed by parasitic f ' capture before fission can occur increases, which introduces negative reactivity L and, thus, limits the reactor power level actually achieved. Hence, any introduc-l l tion of excess reactivity which increases the reactor power level and fuel l temperature will be immediately and automatically compensated for. 4-31

---

a

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\\ A second important but smaller contribution to the negative temperature 238 coefficients of reactivity is due to Doppler oroadening of the U neutron resonance absorption peaks at increased fuel temperatures. This mechanism \\ decreases the resonance escape probability, and thus k,ff decreases. The total prompt ~ negative temperature coefficient of reactivity for the AFRRI-TRIGA Mark-F reactor is measured to be approximately -0.0126 % Ak/k per degree C. The total steady state negative temperature coefficient of reactivity for the AFRRI-TRIGA Mark-F reactor is measured to be approximately -0.0051 % Ak/k per degree C. 4.17 REFERENCES 1. Amendment #17 to the Technical Specifications, Facility License No. R-84 for AFRRI-TRIGA Mark-F Reactor, Docket No. 50-170. 2. Sholtis, Joseph A., Jr., Capt., USAF, Nuclear Criticality Safety Analysis of Hypothetical AFRRI-TRIGA Fuel Element Storage Rack Accidents, AFRRl/SSD Memorandum for Record, January 19,1981. I L 4-32

5.0 EXPERIMENTAL FACILITIES 5.1 GENERAL De AFRRI-TRIGA Mark-P reactor serves as a source of mixed ionizing and neutron radiation for research and radioisotope production. De unique flexibility of the AFRRI-TRIGA reactor is achieved by the horizontally movable core which can traverse from one irradiation position to another. We major irradiation facilities in which experiments can be carried out using the reactor are: Exposure Room #1 Exposure Room #2 e Pneumatic Transfer System (Optional) 9

• PoolIrradiation In-Core Experiment Tube (CET).

We In-Core Experiment %be was described in Section 4.8 and thus will not be addressed further here. A description of the other irradiation facilities is presented below. 5.2 EXPOSURE ROOMS #1 AND #2 %e exposure rooms are located on the first level of the AFRRI reactor building at opposite ends of the reactor tank (Figure 5-1). Exposure Rooms #1 (ER

1. 1) and #2 (ER #2) are located at the north and south ends of the reactor tank, respectively. A semi-cylindrical portion of the aluminum reactor tank (1/4 inch thick) extends into each exposure room. When the reactor core is moved to the extreme position adjacent to either exposure room, a water clearance of about 1 inch exists between the core shroud and the inside surface of the reactor tank.

5.2.1 Exposure Room Construction De physical dimersions of ER #1 are approximately 20 feet by 20 feet by 101 feet high. We physical dimensions of ER #2 are approximately 15 feet by 14 feet by 9 feet high. Except for the aluminum reactor tank projection into each a exposure room, all of the walls, the ceiling, and the floor of each exposure room are constructed of ordinary (Portland) concrete which serves as the primary biological shield. Approximately 61 feet of concrete separate the exposure rooms i from the subsoil underlying the building. At least 12 feet of. concrete or concrete 5-1

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and soil separate ER #1 from the occupied areas of AFRRI, and at least 9 feet of concrete or concrete and soil separate ER #2 from the occupied areas of AFRRI. All six surfaces of both exposure rooms are covered with a 1-foot thick wood lining. By thermalizing a portion of the neutron population leaving the exposure room, the wood lining minimizes fast neutron activation of the concrete biological shield and, thereby, reduces the effects of secondary gamma radiation. The surface of the wood lining is covered with an epoxy resin to facilitate decon-tamination of.the rooms, if necessary. The walls and ceilings of the exposure rooms are also covered with 1/8-inch 2 thick masonite panels painted with gadolinium-oxide paint (#11 mg/cm ). The gadolinium, which has a large absorption cross section for thermal neutrons, reduces the thermal neutron population in the exposure rooms. This, in turn, reduces the amount of Ar"I produced in the exposure rooms due to neutron 40 activation of Ar, a normal constituent of air. Exposure Room #1 contains three lead curtains, suspended from the exposure room ceiling by aluminum tie rods, which surround the top of the reactor tank projection. The lead curtains reduce scattered gamma radiation above or not attenuated by the reactor tank shielding doors when the reactor is operating adjacent to ER #2 with the reactor shielding doors closed. The lead curtains are 22 inches high and are constructed of stackeo 1-inch lead sheets mounted in aluminum frames. The front curtain is 4 inches thick and 32 inches wide, and the two side curtains are 2 inches thick and 23 inches wide. Exposure Room #1 has a cadmium-gadolinium (Cd-Gd) shield welded to the tank projection to reduce the thermal neutron leakage from the reactor core into ER #1. The Cd-Gd sheet is approximately 4 feet square and is centered on the reactor tank projection with respect to the core midplane. The shield consists of a 2 0.016-inch thick aluminum sheet sprayed with gadolinium oxide paint (=22 mg/cm ) and covered with a 0.040-inch thick cadmium sheet. By reducing the thermal neutron population in ER #1, the Cd-Gd sheet helps to reduce the production of Ar"I in ER #1. The cadmium sheet also reduces secondary beta radiation emitted from the gadolinium. A portable lead shield is also provided in ER #1 to provide a means of enhancing the neutron-to-gamma ratio for experimental purposes. The shield consists of an aluminum frame about 30 inches square by 12 inches thick that can 53

be filled with lead bricks to obtain a lead shielding thickness of from 1 to #6 inches, as desired. The shield is mounted. on a frame with wheels, which rolls on tracks secured to the floor.and positions the shield approximately at the core j midplane. When not needed, the portable shield can be rolled away from the tank projection against one of the exposure room walls. When positioned in front of the tank projection, the portable lead shield is bolted to tie rods which support the lead curtains to ensure that the shield will not move during an experiment. 5.2.2 Exposure Room Access l Access to each exposure room is gained from the prep area through separate reinforced concrete rolling plug shield doors. Both plug doors are mounted on four - wheels and can be moved, either by an electric motor or manually, along two recessed tracks anchored in the concrete floor. Both doors are stepped on the sides and top to prevent radiation streaming from the exposure rooms and are approx-imately 7 feet wide by 9 feet high. The ER #1 door is about 12 feet thick and weighs approximately 160,000 pounds. The ER #2 door is about 9 feet thick and weighs approximately 106,000 pounds. The Facility Interlock System (Section 4.13) acts to minimize any radiation exposure to personnel either in the prep area or entering either exposure room, while AFRRI's health physics procedures admini-stratively govern the opening of the exposure room plug doors. The plug doors may be opened or closed with electrical motors. The electric motor (#1 horsepower) for each plug door is controlled by a locked control power box in the prep area. The keys to the control power box are under the administrative control of the Reactor Operator. Once the control power box is opened, the plug door can be opened, closed,' or stopped by pushing the appropriate button, given that the Facility Interlock System is satisfied. Microswitches interrupt the electrical power to the motor when the door is either fully opened or fully closed. A circuit breaker for each plug door motor, installed beside each plug door, provides a means of manually interrupting the electric power supply to the plug door motor in the event of a malfunction of the control power box or one of l the microswitches. I i The plug doors may also be opened and closed manually. This is accomplished by disengaging the padlocked coupler from the electric motor drive and then engagmg the coupler to the manual drive. The handwheelis then turned to close or o 1 open the plug door. Manual operation of the exposure room plug doors overrides 5-4.

=, v < one part of the Facility Interlock System but does not affect its remaining logic. . For example, if ER #1 is' opened manually,while the core is in Region I with the

reactor tank lead shield doors closed or while the lead shield doors are open with Cthe core in any region, an immediate reactor scram will occur. The key to the

~ . padlock : which. secures-the coupler is under the ' administrative control of the - 1 Reactor Operator. Pictures of the ER #1 plug door in a closed and open position are provided in - Figures 5-2 and 5-3.. The ER #2 plug door has identical characteristics with respect to the electrical components and the type of drive mechanisms. 15.2.3 - Exposure Room Monitoring l Each exposure room is monitored by a separate continuous air monitor (CAM) . located in the prep area close--to the appropriate-exposure room plug door. The CAM's provide continuous sampling and monitoring (gross beta-gamma activity) of : airborne radioactive particulate matter. present in ER #1 and ER #2.. A general

description of the CAM's is given in Section 3.6.2. The ER #1 CAM samples ' air taken directly from ER- #1-and discharges the air back into ER #1. The ER #2 CAM : samples air taken directly from ER #2 and discharges the air to Warm Storage (Room 1120).; Local readout with audible and visual alarms associated with certainL preset airborne ' activity levels provide personnel in.the prep area with information concerning airborne activity levels in the exposure rooms.

9 Before opening /either exposure room plug door,- the exposure room airborne activity levels,'as measured by the appropriate exposure room CAM, must be below F a certain level.~ ilf the CAM indicates that the plug door may be opened, access to the exposure room is; permitted only af ter extensive radiation surveys are taken

(with portable radiation monitors) of the exposure room, both from the prep arca-

- past the opened plug door land in the exposure room itself. If high radiation levels exist, the plug door must be closed and a certain amount of time allowed to elapse before exposure room entry is again attempted. All personnel in either the prep

area or exposure rooms will have appropriate personnel radiation monitors and be

" assigned limited stay times, if significant radiation levels exist in either area, to assureicompliance with both NRC (10 CFR 20) and internal AFRRI occupational ' dose limits.. 5

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I' In addition to' airborne radiation monitoring of the exposure rooms, radiation monitoring is also provided in the prep area. A CAM located north of the prep area . near ER #1 samples air taken directly from, and discharged back to, the local area.

The prep area CAM ~ alarms,~both audible and visual, are local (as with the exposure ~

~ room ' CAM's). : RemoM Area Monitors (RAM's) are also provided in the prep' area. ~ Remote Area Monitors E-3 and E-6, which provide readouts in the reactor control-room, are discussed in Section' 3.6.1. RAM's E-4 and E-5, also located in.the prep . area, consist of a scintillation detector which measures gamma radiation with' energies greater than 20 kev. The units have a range of 0.05 mR/hr to 50 mR/hr and a nominal accuracy of 120 percent at all levels. The E-4 and E-5 RAM readouts are provided locally on the units themselves. RAM's E-4 and E-5 activate radiation alarms within the AFRR1 complex at an adjustable, preset radiation level. -RAM's E-4 and E-5 also activate visual alarms in the Security Office (Room 3112) and on the units themselves when a high-level alarm occurs or when loss of high voltage' to th'e datector or a loss of signal to the remote readout occurs (failure alarm)..The. RAM's are calibrated at regular intervals using a radiation source of known intensity. The location of the prep area RAM's, the radiation readouts and alarms are shown on Table 5-1. 5.2.4 ~ Patch Panel System - A patch panel system exists within the AFRRI reactor building so.that electrical signals from instruments in either exposure room can be transmitted to other areas inside AFRRI. - A diagram of the patch panel system is given in Figure 5-4. The Patch Panelin ER'#1 is connected to a Patch Panel in Hall 1106, and the patch panel in ER #2 is connected to a Patch Panel in the prep area near ER #2. From here, the signals may be routed through other patch panels, eventually ending at the reactor control room patch panel and other areas, as desired. ~ 5.3 PNEUMATIC TRANSFER SYSTEM The Pneumatic Transfer System (PTS) associated with the AFRRI-TRIGA reactor permits samples to be transported to a position in the reactor pool near the - core shroud when the core is ' moved to the appropriate pool position. The PTS is an optional experimental facility as it is normally not in-place - but can be installed as required.- The Pneumatic Transfer System is divided into two separate tube banks, ' designated as A-Tube Bank' and B-Tube Bank (Figures 5-5 and 5-6). The A-Tube 5-8 L~

1 ~ TABLE 5-1 PREP AREA REMOTE AREA: MONITORS 4 -Remote Area 1 Monitor Imeation Readout Radiation ' Alarms E-3 Approximately 6 feet above Meter in the reactor Activates visual alarm'in' the floor on the prep area control room and in . Health Physics Division (Room 1 . west wallopposite ER #1 Room 3112 1420), in reactor control room i plug door and in Room 3112; activates local i visual alarm on the prep area east wall near ER #1' l E-6 Approximately 6 feet above Same as E-3 Same as E-3 (except that the the floor on the prep area ~ visual alarm activated in the I west wall opposite ER #2 prep area is located on the east I wallnear ER #2) T E-4 Approximately 6 feet above Meter on unit itself Activates audible and visual alarm the floor on the prep area and meter in Room 3112 on unit itself; activates visual east wall north of ER #1 alarm in Room 3112; activates plug door audible and visual alarm in Health l Physics Division (Room 1420) E 'S Approximately 6 feet above Same as E-4 Same as E-4 the floor on the east wall of the prep area between the plug doors to ERS #1 and #2

i a PATCH PANEL PATCH PANEL EA e t ER #2 3 g 4 6 i l I a' PATCH PANEL IN PREP AREA MALLties gm. OUTSIDE ER e 2 I I I l {i $i r PATCH PANEL PATCH PANEL LINAC CONTROL REACTOR CONTROL MOOM ROOM EEY DENOTE POSSISLE PATCH PANEL CONFIGURATION FOR INSTRUMENTATION IN ER e 2 DENOTE POS818LE PATCH PANEL CONFIGURATION FOR INSTRUMENTATION IN ER e 1 b FIGURE 5-4 PATCH PANEL SCHEMATIC i J-5-10 e s -n -~,e ,w--, n- -,-.--- -,, _, - _ -_ ,_y, a-

ese5 RE ACTOR POOL .. TU.E.- A.TUSE SANK fb A ~ ) HOT cec. A.TUSE BANK / \\J l / / / / / / @@@@IAY @@6.' SENDER RECEtVER STATIONS RADIOCHEMISTRY LAB FIGURE 5-5 PNEUMATIC TRANSFER SYSTEM ROUTING DIAGRAM 5.I1

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• • • **'*I SEND t

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• WE MK 10L E hotD Comm0M vaLvtl att LINE 4.waf vaLvt al# >

7 70 ,A ( n07 CELL RECtivt SEND ~a ExMaulf RtLitPVALvt 8tLTER BLO*ER FIGURE 54 PNEUMATIC TRANSFER SYSTEM 5-12

Bank and B-Tube Bank are capable of rapidly transporting samples between the Radiochemistry Laboratory and the reactor pool. The A-Tube Bank has the additional capability to divert the rabbits, containing irradiated samples, from the sender-receiver stations in the Radiochemistry Laboratory to the adjacent Hot Cell. Basically, each tube bank consists of a blower, an absolute filter, four sender-receiver stations, four rabbit tubes, a common air line, four solenoid (rabbit) valves, one four-way control valve, and a terminus. The samples are transported by small polyethylene or aluminum rabbits having nominal internal dimensions of <1 inch in diameter and #4 inches in length. The rabbits are transported within aluminum pneumatic tubes (rabbit tubes) that have a 1-1/4 inch inside diameter and a 0.083-inch wall thickness. The rabbits are transported to and from the pool irradiating position by applying a partial vacuum to the appropriate rabbit tube volume. The partial vacuum is created by a local blower in the Radiochemistry Lab pulling air through the absolute filter from the rabbit tubes. From the absolute filter, the air is exhausted to the AFRRI stack by way of the reactor building ventilation system. The absolute filter prevents radioactive particulate matter in the rabbit tubes from being discharged into the Radiochemistry Laboratory and subsequently into the reactor building ventilation system without first being filtered. The partial vacuum is applied to the UP-DOWN (4-way) valve. When the valve is in the UP position, the partial vacuum is applied to the manifold at the sender-receiver station. When the rabbit valves are opened, rabbits in the pool termini are drawn through the rabbit tubes to the sender-receiver stations and air is drawn into the system through the common air tube. When the UP-DOWN valve is placed in the DOWN position, the partial vacuum is applied to the termini through the common air tube. When the rabbit valves are opened, the rabbits in the sender-receiver stations are drawn through the rabbit tubes into the pool termini and the atmospheric air is drawn into the system through the manifold The relief valves allow air to flow to the blower when the rabbit valves are clo preventing the blower from overheating. The control circuits for the operation of the Pneumatic Transfer System are contained in two control units. One local control unit is in the Radiochemistry Laboratory; the second unit is the remote control unit, which is located on the west I wall in a small adjoining room north of the Radiochemistry Laboratory. j i 5-13

The types of experiments conducted with the Pntumatic Transfer System primarily involve radioisotope production and activation of small cample AFRRI's health physics procedures and subsequent neutron activation analysis. internal RSD procedures administratively govern the use of the pneumatic transfer system 1AW the technical specifications. 5.4 POOL IRRADIATION The pool-type tank of the AFRRI-TRIGA Mark-F reactor permits the irradia-(18) tion of waterproofed experiments submerged in the reactor pool. Eighteen holes in the upper grid plate above the core also allow small samples to be inserted into the core for irradiation. The decision to perform an experiment in the reactor pool as opposed to the Pneumatic Transfer System or the CET is The dictated by specimen size and the required radiation flux and intensity. actual placement of experiments in the reactor core or the reactor pool will be At no time shall experiments be limited by the technical specifications. performed in the reactor pool which would introduce more excess reactivity i l the core than the technical specifications allow. The use of the reactor poo as an experimental facility also includes the use of portable beam tubes placed in => the reactor pool. The types of experiments performed in the reactor pool are similar to those performed in the Pneumatic Transfer System. 5.5 ARGON ACTIVATION _ Argon-41 is produced in the exposure rooms, CET, reactor pool, and the Pneumatic Transfer Systems due to the leakage of thermal neutrons from the 41 s produced by thermal neutron i reactor core into these exposure facilities. Ar 40 n the air of these experimental facilities i activation of naturally occurring Ar (note that some air is always dissolved in the pool water). Since Ar41 s a noble i gas, it is not retained in the human body and radiation exposures are a res direct exposure (immarsion) to the Ar 1 gamma rays given off during its natural 4 decay. and Pneumatic Transfer System are designed to The exposure rooms minimize Ar41 exposures to the AFRRI staff. The exposure rooms contain i-design features, described in Section 5.2.1, which minimize the production of Ar41 Air from Exposure Room 1, Exposure Room 2, and the exhausted to the AFRRI stack by way of Pneumatic Transfer System is This limits egress of Ar 1 to 4 the reactor building ventilation system. occupiable areas inside the AFRRI complex. The stack monitoring systems provide 5-14

R s a means'of ensuring that releases of 'Ar 1 to unrestricted areas meet the requirements of 4 10 CFR Part 20.106.- ! Typical Ar41' production rates -In these experimental facilities, based on ' measurements of Ar41 releases through the AFRRI stack during reactor operations, are o: Exposure Room #1 with reactor core at extreme position closest to ER #1 0.58 mCl/kW-hr i 0-. Exposure Room #2 with reactor core at extreme position closest to ER #2 2.10 MCI /kW-hr 0 Pneumatic Transfer System with reactor core , at position adjacent to pneumatic tubes 0.47 mC1/kW-hr. -3.6 PROCEDURES All. experiments associated with the AFRRI-TRIGA reactor are performed in accordance with the technical specifications and regulations promulgated by 10 CFR Part 50, especially Section 50.59. The administrative procedures practiced at AFRRI provide a means of compliance with the appropriate regulations and are detailed in the Technical ~

5pecifications and a: set of internal written, procedures, RSD Procedures 10 series (approved by the Chief of the Reactor Branch and reviewed by the Reactor and Radiation Facility Safety Committee (RRFSC)).

Before performing any proposed first time experiment (with 'he exception of routine t operations as outlined in the Technical Specifications and RSD procedures,'such as reactor parameter measurements and instrument checks and calibrations), the experiment must be reviewed for radiological safety and approved by. the Reactor and Radiation Facility Safety Committee (RRF5C) and other persons or groups as specified in writing by the Director of AFRRI. A Reactor Use Request (RUR) must also be completed, approved, and signed by the AFRRI Reactor Facility Director (RFD)'before any experiment can be performed. Safety evaluations for proposed experiments conducted without prior NRC ~ . approval under provisions of Section 30.59, 10 CFR Part 50, are reviewed by RRFSC and i- ' the other appropriate groups or individuals to determine whether the proposed experiment constitutes an unreviewed safety question or a change to the technical specifications. The review and approval process allows personnel specifically trained in radiological safety _to consider and recommend alternative operational profiles, such as different core 5-15

positions, power levels, and irradiation times, etc., which might reduce staff exposures, the release of radioactive materials to the environment, or both. l A record of tests and experiments performed, including written safety evaluations, is maintained by AFRRI in accordance with Section 50.59,10 CFR Part 50. 5.7 EXPERIMENTS The limitations on experiments specified in the technical specifications for the AFRRI-TRIGA reactor state that: The sum of the absolute reactivity worths of all experiments in the e reactor and in the associated experimental facilities shall not exceed $3.00 (2.1 % Ak/k). This value includes the total potential reactivity insertion which might result from experiment malfunction, accidental experiment flooding or voiding, and accidental removal or insertion of experiments. Each fueled experiment shall be limited so that the total inventory of e lodine isotopes 131 through 135 in the experiment is not greater than 1.5 curies, and the maximum strontium-90 inventory is not greater than 5 millicuries. e Known explosive materials shall not be irradiated in the reactor in quantitles greater than 25 milligrams. e Samples shall be doubly contained when release of the contained material could cause corrosion of the experimental facility. Before any experiment can be conducted using the reactor or experimental facilities, the experiment must be reviewed by the Reactor and Radiation Facility Safety Committee. A Reactor Use Request (RUR) must also be completed, approved, and signed by the AFRRI Reactor Physicist-in-Charge (PIC) before an experiment can be performed. If it is anticipated that an experiment will cause a reactivity change of approximately 10.5 % Ak/k, k-excess measurements must be made at the core position in which the experiment will be performed, both with and without the experiment inserted, to determine the actual reactivity worth of the experiment. Typical experiments that have been performed at AFRR1 using the AFRRI-TRIGA reactor and its associated experimental facilities, within the conditions stated above and as authorized by Facility License No. R-84, include: 3-16

^ Tha types if sxperiments conducted with the Pneumatic Transf;r System primarily involve radioisotope ' production and activation of small samples for ' subsequent neutron activation analysis. AFRRI's health physics procedures and internal R50 procedures administratively govern the use of the pneumatic transfer system LAW the technical specifications. 5.4 POOL IRRADIATION The pool-type tank of the AFRRI-TRIGA Mark-F reactor permits the irradia-tion of waterproofed experiments submerged in the reactor pool. Eighteen (18) holes in the upper grid plate above the core also allow small samples to be inserted into the core for irradiation. The decision to perform an experiment in the reactor pool as opposed to the Pneumatic Transfer System or the CET is dictated by specimen size and the required radiation flux and intensity. The actual placement of experiments in the reactor core or the reactor pool will be limited by the technical specifications. At no time shall experiments be performed in the reactor pool which would introduce more excess reactivity into the core than the technical specifications allow. The use of the reactor pool as an experimental facility also includes the use of portable beam tubes placed in the reactor pool. The types of experiments performed in the reactor pool are similar to those performed in the Pneumatic Transfer System. 5.5 ARGON ACTIVATION Argon-41 is produced in the exposure rooms, CET, reactor pool, and the Pneumatic Transfer Systems due to the leakage of thermal neutrons from the 4 i reactor core into these exposure facilities. Ar 1 s produced by thermal neutron activation of naturally occurring Ar40 n the air of these experimental facilities i i (note that some air is always dissolved in the pool water). Since Ar41 s a noble i gas, it is not retained in the human body and radiation exposures are a result of direct exposure (immersion) to the Ar41 gamma rays given off during its natural decay. The exposure rooms and Pneumatic Transfer System are designed to l minimize Ar41 exposures to the AFRR1 staff. The exposure rooms contain design features, described in Section 5.2.1, which minimize the production j of Ar%1 Air from Exposure Room 1, Exposure Room 2, and the Pneumatic Transfer System is exhausted to the AFRR1 stack by way of the reactor building ventilation system. This limits egress of Ar41 to occuplable areas inside the AFRRI complex. The stack monitoring l systems provide 4 5-14 ___._..__._._._,,_r,,_.,,,,__ ____.-_,__,,_._,_.,_~-.__s_,

Bank and B-Tube Bank are capable of rapidly transporting samples between the Radiochemistry Laboratory and the reactor pool. The A-Tube Bank has the additional capability to divert the rabbits, containing irradiated samples, from the sender-receiver stations in the Radiochemistry Laboratory to the adjacent Hot Cell. Basically, each tube bank consists of a blower, an absolute filter, four sender-receiver stations, four rabbit tubes, a common air line, four solenoid (rabbit) valves, one four-way control valve, and a terminus. The samples are transported by small polyethylene or aluminum rabbits having nominal internal dimensions of <1 inch in diameter and #4 inches in length. The rabbits are transported within aluminum pneumatic tubes (rabbit tubes) that have a 1-1/4-inch inside diameter and a 0.083-inch wall thickness. The rabbits are transported to and from the pool irradiating position by applying a partial vacuum to the appropriate rabbit tube volume. The partial vacuum is created by a local blower in the Radiochemistry Lab pulling air through the absolute filter from the rabbit tubes. From the absolute filter, the air is exhausted to the AFRRI stack 'oy way of the reactor building ventilation system. The absolute filter prevents radioactive particulate matter in the rabbit tubes from being discharged into the Radiochemistry Laboratory and subsequently into the reactor building ventilation system without first being filtered. The partial vacuum is applied to the UP-DOWN (4-way) valve. When the valve is in the UP position, the partial vacuum is applied to the manifold at the sender-receiver station. When the rabbit valves are opened, rabbits in the pool termini are drawn through the rabbit tubes to the sender-receiver stations, and air is drawn into the system through the common air tube. When the UP-DOWN valve is placed in the DOWN position, the partial vacuum is applied to the termini through the common air tube. When the rabbit valves are opened, the rabbits in the sender-receiver stations are drawn through the rabbit tubes into the pool termini and the atmospheric air is drawn into the system through the manifold. The relief valves allow air to flow to the blower when the rabbit valves are closing, preventing the blower from overheating. The control circuits for the operation of the Pneumatic Transfer System are contained in two control units. One local control unit is in the Radlochemistry Laboratory; the second unit is the remote control unit, which is located on the west wall in a small adjoining room north of the Radiochemistry Laboratory. i 5-13

1 e Irradiation of up to 500 pounds of animals for up to 1 Mw-br with or without shielding Irradiation of various materials in the pneumatic tube system or CET. e These materials included inorganic compounds of 70 different elements limited on the basis of the chemical or nuclear characteristics of the material. The alloys of 10 metals were also included in these experi-ments, limited on the basis of their nuclear characteristics. Preparation of Argon-41 calibration standards in the exposure rooms e and irradiation of up to 5 grams of sodium carbonate in the pneumatic transfer system Irradiation of old paintings for radiographic analysis of age and authen- . e ticity. Irradiation of electronic components in the exposure rooms and the e pneumatic transfer system. These components include integrated circuits. 4 ( 5-17

- 6.0 SAFETY ANALYS15 ' 6.1 POTENTIAL CONSEQUENCES ASSOCIATED WITH TRIGA REACTOR OPERAT10N5 The potential radiological consequences associated with routine and non-routine operation of the AFRRI-TRIGA Mark-F reactor have been reviewed. This section presents the results of this study. . Routine operational risks are present in the manipulation of any radioactive material because of its potential for biological and physical damage. This study investigated potential consequences which may be associated with routine opera- .tlon, including the handling of radioactive materials, reactor power transients, experiments associated with reactor operations, and production of radioactive gases (primarily through activation of argon). Nonroutine operational risks may arise from the improper handling or from - malfunctions of materials and equipment. The specific problems investigated were fuel element cladding failure, radioactive contamination of the reactor pool water, loss of shielding and cooling water, and Design Basis Accidents. In order to evaluate the potential consequences associated with the operation of this reactor the following assumptions have been made: e The reactor will normally have two modes of operation: Steady State Operation at power levels not to exceed 1.1 megawatt I I (MW ) for testing operational purposes and 1.0 megawatt-(MW ) fw normal operations. ' For calculations, the assumption of operation in the steady-state mode at a power level of 1.0 MW was utilized. The total reactor operation has been less than 400 MW-hr from June 1964 to I -September 1980. The total reactor operation from January 1980 to l l September 1980 has been less than 26 MW-hr. Pulsed Operation achieved by intentionally placing the reactor on a prompt critical excursion by making a step insertion of reactivity above critical utilizing the transient rod and appropriate scrams and inter-locks. The maximum step insertion is !!mited by the technical specifications to-2.8% & k/k reactivity in the pulse mode.I The ~ average " maximum-pulse" used during AFRRI's TRIGA pulse operation has been 28 MW-sec. For calculations, the assumption of a maximum l 6-1 l r ~. -.

7 v l excursion pulse of 40 MW eee was utilized as an upper limit for pulse operations or an inadvertent transient. 1he reactor operations will be supervised by individuals who are e qualified operators trained in the detection and evaluation of radio-logical consequences. h summary, the analysis indicated that the consequences from the Design Basis Aeeldents of a fuel element drop accident or a fuel element elad failure accident were insignificant. 1he maximum calculated whole-body dose was less than 2 mrad 'beyond 25 meters from the AFRRI facility. The maximum. calculated thyroid dose was less than 57 mrad beyond 25 meters from the AFRRI facility. Doses from.these postulated accidents to individuals beyond the boundary of the Naval Medical Command site would be significantly less than 1 mead.1herefore, it is concluded that the operation of the AFRRI reactor in the manner authorized by Facility License No. R-84 does not represent an undue risk to the health and safety of the operational personnel or the general public. 6.2 ROUTINE OPERATIONAL RISKS 6.2.1 ~ Handling Radioactive Material Because of the potential biological effects of radioactive material, special precautions should be taken in the handling of these materials. Reactor operations must be supervised by responsible individuals who are trained in the ' detection and evaluation of radiological consequences. Administrative, opera-tional, and health physics procedures will be followed, and special equipment and procedures which are needed to maintain the ALARA (as low as is reasonably achievable) concept of radiation protection will be used. l 1he radiological consequences associated with fuel elements are of the same nature as those assoelated with Isotope production. Because of their high l radiation levels, fuel elements should be kept under water for shielding purposes. Administrative policy requires that a fuel element not be removed from the L reactor pool for at least 2 weeks following its use in the reactor core. When a fuel element is removed from the reactor tank, a conventional fuel element transfer eask may be used to reduce radiation levels to within acceptable limits. A Design Basis Accident (Section 6.3.4) is a fuel element drop during such a transfer in whleh a cladding failure occurs in air. I-6-2 Lt

y When proper administrative, operating, and health physics procedures are utilized, the handling of radioactive ma;erials does not represent a significant risk to the health and safety of operating personnel or the general public. 6.2.2 Reactor Power Transients The following discussion is based on experimentation and testing performed by General Atomic and presented in Reference 2. Theoretical estimates are also reviewed. The U-ZrH fuel elements used in the TRIGA reactor are capable of operating under conditions of transient experiments for delivery of high intensity bursts of neutrons. Fuel elements with 8.45 wt.%U have been pulsed repeatedly in General Atomic's Advanced TRIGA Prototype Reactor (ATPR) to peak power levels of over 8,000 MW providing a neutron fluence per pulse of approximately 1.0E+15 2 nyt (neutrons /cm ), The ATPR fuel elements were subjected to thousands of pulses of 2,000 MW or more. The inherent safety of the fuel element stems from its large prompt negative temperature coefficient of reactivity, which causes the automatic termination of a power excursion before any core damage results. This temperature coefficient has been measured to be approximately $0.017 reactivity 0 loss per 1 C rise in fuel temperature. The reactor loading is limited by the technical specifications to a maximum of 3.5% A k/k ($5.00) excess reactivity above cold critical, with or without all experiments in place.I Thus, the maximum reactivity transient that could possibly occur would be that produced by the rapid insertion of the entire available amount of excess reactivity. In regard to TRIGA fuel performance, experiments at General Atomic's ATPR have been performed for step insertions of up to 3.5% Ak/k ($5.00) reactivity. The fuel elements were subjected to thousands of pulses of 2,000 MW or more and attained temperatures of up to 1000 C. For example, the annular core TRIGA reactor at JAERI (Japan) has operated since 1975, with over 1,000 pulses of all sizes at fuel temperatures up to 1,000 C. From the results of the tests performed b)/ General Dynamics, there was no external evidence of change in any of the five special test elements in the first 200 pulses. Theoretical estimates based on the Fuchs-Nordheim mathematical model of the AFRRI-TRIGA reactor have also been made. It was shown that an insertion of 2.8% a k/k (AFRRI-TRIGA technical specification limit) results in a maximum average fuel temperature of less than 350*C. 6-3

Therefore, based on operating experience of the ATPR and the Fuchs-Noreelm mathematical model, it can be concluded that the rapid insertion of the total excess reactivity of 3.3% A k/k would not represent an undue risk to the operating personnel or the general public. 6.2.3 Improper Fuel Loadina. Fuel loading of. the reactor is always supervised by trained, licensed, supervisory personnel. All reactor monitoring and shutdown devices will be operational during loading. The worst possible case of improper fuel loading would be for the operator to insert a fuel element in a core already slightly critical. In a core near its critical point, all of the inner fuel positions would be occupied so that the extra fuel elements could be added only in a peripheral position, where a fuel element worth is approximately 0.20% A k/k. Since step additions of 2.1% A k/k excess reactivity are made on a routine basis for pulsing the reactor, the addition of 0.20% A k/t would not present a danger of damaging the reactor or fuel. The reactor would undeego a mild transient and then operate at a steady state power level of about 60 kW. Even in the extremely unlikely event that a fuel element in the "B" ring should be improperly handled, its rapid insertion would result in an addition of about 0.90% A k/k. As indicated above, such an addition would not result in any damage to the reactor or the fuel and would not represent an undue risk to the health and safety of the operating personnel or the general public. 6.2.4 Production of Radioactive Cases in the Reactor Coolant The production of radioactive gases by the reactor in its associated facilities originates through neutron activation of elements in the air or water. One of the most important of these activation products is radioactive argon (Ar-41) with a half-life of 1.83 hours." Calculations are based on a temperature of 70*F (21 C), and Ar-40 content of air of 0.94 percent by volume. in the calculation to determine the amount of argon dissolved in the reactor pool water, it is assumed that the argon follows Henry's 1.aw. At a water temperature of 70*F (21*C), the corresponding water vapor pressure is 26 mm Hg. The partial pressure for air would therefore be 760-26 = 734 mm Hg. Using an argon content of air as 0.94 percent by volume, the resulting partial pressure of argon is 734x(9.4E-3) = 7 mm Hg. Applying Henry's Law, a saturation concentra-3 tion of argon in water of 1.367 E-8 gram mole of Ar-40 per 1 cm H O is obtained. 4 2 6-4 s

1 2 Using o microscopic thermal neutron absorption cross section cf 6.lE-25 cm for 3 Ar-40, the macroscopic absorption cross section of argon in water becomes SE-9 -I. cm Activity production, assuming a saturation condition and an irradiation time, t, can be given as: O (1-e~ *p) a* (1) A = n P 3.7 E+4 where: 3 activity in reactor water (uCi/cm ) A = p 2 thermal neutron flux (n/cm -sec) Q = n core circulation time (sec) t = p macroscopic absorption cross section (cm-I) I = a constant to convert disintegrations per second to microcuries 3.7 E+4 = decay constant (sec~I) A = frradiation time (sec). t = The average thermal neutron flux (Q )in the reactor core is estimated to be n 2 1.0 E+13 n/cm -sec at the 1.0 MW power level. The water circulates through the core by natural convection and is estimated to change completely in approximately 4 seconds. The Ar-41 decay constant is 1.05x10~" sec-I. The core of the reactor 3 holds 3.4E+4 cm of water and the rate of flow of water through the core is 0.9E+4 3 cm /sec." Substituting the appropriate values in Equation 1 yleidst 3 A = 5.7E-4 pC1/cm, p and a total core activity of Q = 5.1 pCi/sec. The travel time of Ar-41 from the core to the water surface, a minimum distance of 457.2 cm (15 ft), has been estimated to be 42 seconds. Assuming the decay to be negligible, the maximum activity reaching the surface is 3.1 pCi/sec. Under saturated, steady state conditions (a highly conservative assumption for this facility), the maximum rate at which Ar-41 can escape from the water surface will be 5.1 pCl/sec, and an equivalent amount of Ar-40 will dissolve in the water to replace the Ar-41 depletion. As water temperature increases, the water 6-5

vapor pressure till increase and the amount of dissolved Ar-40 will decrease, resulting in a lower generation of Ar-41 escaping to the reactor room. The radioactive alrgon (Ar-41) will escape from the reactor pool, dissipate in i ~l the reactor room air, and recirculate through the ventilation system. It is l estimated that this system handles 2.80E+9 cm3 (92,000 ft3 of air), and that it 3 3 recirculates 3.34E+8 cm / min (11,800 ft / min) through the system and exhausts its 3 3 3 3 share (8.24E+7 cm / min (2910 ft / min)) of the 1.133E+9 cm / min (40,000 ft / min) through the stack.3 The added exhaust is from the other AFRRI facility ventilation systems. Since the air is recirculated through the system at a high rate, it is assumed that the Ar-41 is distributed throughout the areas handled by the system and that the Ar-41 decay is negligible. If it is assumed that there is continuous reactor operating time, equilibrium conditions can be assumed. Since equilibrium conditions are assumed, the same concentration of Ar-41 will exist in the reactor room as in the exhaust air. The concentration in the reactor room is: 3.1 p Ci/sec 3 = 3.7 E-6 p C1/cm 3 1.37 E+6 cm /sec 3 The concentration in the stack is 2.7 E-7 pCi/cm and is further dispersed in the air before reaching the surrounding population. The gamma dose rate in the reactor room can be estimated for the Ar-41 concentration of 3.7 E-6 pCi/cm by assuming submersion in a spherical source equivalent to the reactor room volume. The exposure rate can be given as: S K h (1-e ") (2) D= where: D = exposure rate (R/hr) 3 S = volume source (Mev/sec-cm ) y =. attenuation coefficient for air (cm'I) p 2 K = flux to dose conversion f.1ctor (Mev/cm -sec par R/hr) ra radius of equivalent sphere (cm). 3 The volume of the reactor room is approximately 9.2 E+8cm, which is equivalent to a sphere with a radius (r) of 600 cm.3 An Ar-41 concentration of 3.7 3 3 E-6 pCl/cm has a gamma source strength (S ) equal to 0.08 Mev/cm -sec. For y l 6-6

[ l the Ar-41 gamma photon of 1.29 M;v, the linetr cttinuation coefficient for cir (p) -I 2 - is 7.2 E-5 cm and the flux to dose conversion factor (K) is 3.44 E+5 Mev/cm -sec per. R/hr. Substituting _these values into Equation 2 results in an exposure rate from the calculated Ar-41 concentration of 0.18 mR/hr. Such an exposure rate is more than an order of magnitude below the equivalent allowable quarterly exposure rate of 2.5 mrem /hr in a restricted area. Another important activation product is radioactive nitrogen (N-16), with a half-life of #7.14 seconds. The activation occurs as a result of the oxygen content in the pool water from the 016 (n,p) N16 production process. Accordingly, the activity production and exposure rate equations were used to calculate the radioactive nitrogen (N-16) releases from the reactor pool. The concentration of 3 oxygen in water is approximately 0.89 gram mole of 0-16 per 1 cm H O. Using a 6 microscopic thermal neutron absorption cross section of 1.85 E-29 cm for 0-16, -I the macroscopic absorption cross section of oxygen in water becomes 9.9E-6 cm Substituting the appropriate values in Equation 1 yields: Ap = 8.4 E+2 pCl/cm, and a total core activity of: Q = 7.6 Ci/sec. Using a travel time for N-16 from the core to the water surface of 42 seconds, the maximum activity reaching the surface is 0.13 Cl/sec. The radioactive nitrogen will escape from the reactor pool, dissipate and decay rapidly (7.14 second half-life) in the reactor room air. The volume of air 2 above the surface of the reactor pool (1.0E + 5cm ) in which the N-16 activity per second is distributed was assumed to be 100-cm high for negligible decay. The concentration of N-16 activity would be approximately 1.2 E-2 pCi/cm with a gamma source strength (Sv) equal to 2.1 E+3 Mev/cm -sec (Ey = 4.6 Mev/disinte-gration). For the N-16 primary gamma photon of 6.13 Mev, the linear attenuation coefficient for air (p)is 3.2 E-5 cm'I and the flux-to-dose conversion factor (K) is 2 8.4 E+5 Mev/cm -sec per R/hr. Substituting these values into Equation 2 for an equivalent sphere radius of 136 cm results in an exposure rate from the calculated N-16 concentration of 350 mR/hr immediately above the reactor pool. The N-16 activity that is being evaluated by this analysis is the amount that may be released from the pool. It should be noted that due to the short half life of the N-16, essentially no N-16 is released from the stack. 6-7

Based upon measurements made during reactor operations at the 1-MW power level, the typical Ar-41 rate release from the pool surface has been approdmately 0.5.pC1/sec (September 22,1980, Ar"I Report) as compared to the calculated value of 5.1 pCi/sec, a reduction by a factor of 10. A typical concentration at the 3 stack has been approximately 3 E-8 pCi/cm as compared to the calculated value 3 of 2.7 E-7 pCi/cm, a factor of 10 reduction. Gamma radiation levels measured over the pool ranged from 75 mR/hr to 200 mR/hr. Gamma radiation levels measured around the reactor pool chain ranged up to 14 mR/hr. It should be noted that all measured activity values of N-16 include N-16 in the pool water. The calculated exposure rate for N-16 of 350 mR/hr . assumes that all N-16 activity is in the air. 6.2.5 Experiments All experiments performed as part of the TRIGA reactor operations are reviewed by the Reactor and Radiation Facility Satety Committee and must be authorized prior to their performance. The technical specifications contain requirements that must be met before such experiments can be performed using the AFRRI-TRIGA reactor. Experiments are always supervised by trained, licensed, supervisory personnel. However, failure of an experiment is possible and worst-case conditions can be calculated to determine the postulated consequences. The two worst-case conditions for failure of an experiment could result in instantaneous insertion of reactivity or the release of radioactive material from an experiment undergoing activation in the reactor. For an experiment failure in which reactivity could be added, the worst possible case would be the prompt addition of less than 0.36% A k/k in either Exposure Room 1 or 2. As discussed for the case of improper fuel loading (Section 6.2.3), the addition of 0.36% A k/k would be within the range of an improper fuel loading condition. Such an addition would not result in any damage to the reactor or the fuel. For an experiment failure -in which radioactive material could be released . from the experiment, i.e., activation products, the worst case would be the prompt release of the radioactive material to the atmosphere. An authorized experiment . involves the irradiation of 20 liters of argon gas for I hour at a power level of 1 MW. The resulting activation would result in a total Ar-41 activity of 5.6 Ci in the sealed container. If the container should fail and release all of the Ar-41 - activity, the resulting total whole body dose would be less than 2.7 mrad to an 6-8

individual more than 25 meters from the AFRRI facility (Equation 3, Section 6.3.4.1). The failure of this authorized experiment represents the worst case for radiological consequences from an experiment failure in the AFRRI-TRIGA reactor. Such a whole body exposure would not represent an undue risk to the health and safety of the general public. 6.3 NONROUTINE OPERATIONAL RISKS 6.3.1 Radioactive Contamination of Reactor Shielding Water Contaminant material susceptible to neutron irradiation in the shield water is maintained at low concentrations by the water purification system and an in-line set of particulate filters which remove particulates of 5 micron or larger. The consequences associated with a failure of the fuei-element cladding and subsequent fission product contamination of the water have been calculated and studied experimentally, as described in Section 6.3.2.1. The results (1962 FSAR)' show that in the improbable event of fuel element cladding failure, the water can be decontaminated by the resins of the purification system. Manuf acturing inspection and quality controls assure that the possibility of cladding failure is minimal. 2 Experiments conducted over a period of 11 years by General Atomic on 8.5 U-ZrH (8.5 weight percent uranium) fuel elements under various conditions have shown that a small fraction of fission products are released from U-ZrH fuel into the gap between fuel and cladding. This release fraction varies from less than 1.5E-5 for an irradiation temperature of 350 C to a theoretical maximum of 1.0E-2 at 800 C. Three mechanisms (recoil, diffusion, and dissolution) have been shown to be involved in fission product migration. The first mechanism is the fission fragment recoil into the gap between the fuel and cladding. This effect predominates at temperatures up to approximately 400 C. In this range, the recoil release rate is dependent on the fuel surface-to-volume ratio but is independent of temperature. This is the most important mechanism in this system and is considered in the calculations dealing with fission product release following fuel element cladding failure (Section 6.3.2). At temperatures above approximately 400 C, the diffusion-like process predominates. The amount released is dependent on the fuel temperature, the fuel surface-to-volume ratio, the length of time of irradiation, and the isotope half-life. This mechanism is of littic significance to the fission-product release fraction considered in this system. The U-ZrH fuel has a relatively 2 low reactivity in water, steam, and air at temperatures up to about 600 C, 6-9

7 Massive zirconium hydride has been heated in air for extended periods of time at temperatures up to 600 C with negligible loss of hydrogen.2 This negligible loss of 0 hydrogen is due to oxide film formation which inhibits hydrogen loss. 6.3.2 Fuel Element Cladding Failure 6.3.2.1 Summary of Previous Experlmce. Since the original development of the TRIGA reactor, General Atomic has maintained a program of research and development oriented toward continuing improvement in the design of the fuel elements so that TRIGA reactors could be operated in both steady-state and pulsing modes without undue concern regarding the integrity of the fuel. Extensive operational testing of TRIGA fuel elements has been conducted by General Atomic in the Torrey Pines TRIGA and the prototype TRIGA Mark-F since the early 1960's. Experiments on fuel with 8.5 wt.%U were conducted over a period of 11 years under a variety of conditions. These experiments included:2 e 1960 - The measurement of the quantity of a single fission product isot9x release from a full-size TRIGA fuel element during irrao2ation e 1966 - The measurement of the fractional release of several isotopes from small specimens of TRIGA fuel material during and after irradjation at temperatures ranging from approximately 25 C to 1100"C e 1971 - The measurement of the quantities of several fission product isotopes released from a full-size TRIGA fuel element during irradiation in a duplication of the 1960 experiment. In addition, quench tests were performed on TRIGA low-enriched uranium (LEU) fuel samples from temperatures ranging from'800 to 1200 C. The subsequent fuel samples quenched from 800,1000,1100, and 1200 C showed remarkably benign response to the test conditions.3 in the course of developing a reactor for pulsed operation, General Atomic has experienced three fuel element cladding failures in their Torrey Pines TRIGA reactor. These cladding failures have been of two types: Two cladding failures caused by thermal ratcheting of previously used e aluminum cladding during the cycling 6-10

e A cladding. failure associated with an element instrumented with ' internal thermocouples. In the original TRIGA fuel element design, aluminum clad ratcheting occurred ~ during pulsed operations of the reactor. Modification of the fuel element essentially eliminated this cause of failure. The present cladding is stainless steel, and has greater durability than the original aluminum which experienced the metal cladding failures. The cladding f ailure of an instrumented element was apparently the result of internal pressure buildup, water seepage, or possibly the pressure of volatile material inside the element. The consequences of the cladding failures were all minor and in the worst case experience may be summarized as follows: 3 e . The activity in the reactor water reached a maximum of 0.2 pCi/cm. It decayed very rapidly and was measured 24 hours after the cladding 3 failure to be 5x10-5 -pCi/cm, a 4,000-fold reduction ^ e The activity in the air of the reactor room reached about 10 times the maximum permissible concentration for ' fission products and then decayed rapidly. The' faulty fuel element was removed and experiments were resumed 2 hours af ter the activity release. The maximum integrated exposure to operating personnel resulting from this release was I mR e The noble gases were not collected on the air monitor filters used, but it may be inferred from the nature of the particulates collected that 4 only noble gas. fission products escaped from the TRIGA pool in significant proportions when the fuel cladding failed. It is concluded from these experiments that a cladding failure or even the simultaneous failure of the cladding of several fuel elements would not constitute an undue risk' to the health and safety of the operating personnel'or the general l ~ public. 6.3.2.2 Calculation of Maximum Fission Product Release After a Fuel Element L Cladding Failure. Calculations; and a related experiment have been made to determine the maximum concentration of fission products that might be present in the reactor room air following a fuel element cladding failure. l. i L 6-11 E

v. The calculations are based on the fact that as the reactor operates, the fission products will build in the uranium-zirconium fuel mixture unti; an equili-brium concentration is reached for each nuclide. The resulting equilibrium nuclide concentration of fission products depends upon the total energy release in the reactor, the decay process for each nuclide, and the yield of the species from fission. Only the kryptons, the xenons, and the iodines will migrate into the gap between the fuel material and fuel element cladding. A Design Basis Accident (Section 6.3.4) is.a fuel element cladding failure during a pulse operation or i inadvertent transient following steady state operation. Calculated Fission Product Inventory. To determine the various inventories of the fission products produced in the core, data were used from Reference 7 for a steady state power level of 1 MW. The inventories for the kryptons, xenons, and lodines were calculated as shown in Table 1, Appendix C. The prompt fission product inventory for a 40 MW-sec pulse operation was calculated using the data from Reference 7 with the buildup and decay of activities from Reference 8 for prompt fission. The prompt inventories for the kryptons, xenons, and iodines as calculated are shown in Table 2, Appendix C. These inventories will be utilized in the calculation of radiological conse-quences due to the various fuel element cladding failures which may result during the operation of the AFRRI-TRIGA reactor. 2 Determination of Fission Products in Gap. In order to determine the actual percentage of fission-product gases that escape from the fuel material and collect in the gap between the cladding and the fuel material, experiments were conducted in the TRIGA reactor at General Atomic. A fuel element was fabricated with a sealed tube that vented the gap to a charcoal-filled cold trap at the surf ace of the reactor tank. All of the fission-product gases that accumulated in the gap were collected in the liquid-air-cooled charcoal trap by purging the system with helium, and the trap was analyzed. This measured amount of radioactive noble gases enabled the determination of the fraction of the fission products that diffused through the uranium-zirconium hydride materialinto the gap. Although the measured amount of radioactive noble gases for the operating conditions in the AFRRI reactor fuel would indicate a gap activity percentage of 'less than 0.01 percent, the theoretical limit of 0.1 percent gap activity for fission 6-12

product gases of noble gases and lodines, as stated in Reference 2', will be used in the consequence analysis for the Design Basis Accidents (Section 6.3.4). Air aa.d Water Activity Following Cladding Failure. As shcwn in Table 1, Appendix C, the total quantity of all gaseous fission products in a TRIGA core for 1-MW operation at equilibrium is 5.8 E+5 curies. For the purposes of this calculation, it is assumed that the fractions of the iodine, krypton, and xenon isotopes produced that collect in the gap between the fuel material and fuel cladding are 0.1 percent.2 Thus, the total core activity in the gap is calculated to be 5.8 E+2 curies. The maximum amount of fission products that could be released in the event of a cladding failure of a single average fuel element for an average 85 fuel element AFRRI-TRIGA core is less than 7 curies during steady state operation. As shown in Table 2, Appendix C, the total quantity of all gaseous fission products from a 40 MW-sec pulse operation is 1 E+4 curies in the core with a gap activity of 10 curies. The additional release in the event of a cladding failure in an average fuel element is less than 0.2 curies. The total release from a cladding failure event is approximately 7 curies from an average fuel element i during pulse operation following steady state operation. As given in Reference 5, the volume of water in the TRIGA reactor tank is approximately 5.7 E+7 ' cm and the volume of air in the reactor room is 3 approximately 9.2E+8 cm. For the purpose of calculation, it is assumed that all the gaseous fission products in the gap are available for release. As concluded from measurements of the worst cladding failure experienced, only noble gas fission products would be expected to escape from the TRIGA pool when the fuel cladding fails. Due to the low pressure in the gap and the small gap size in a TRIGA fuel element, the lodine released from a cladding failure should be completely dissolved in the pool water and only be released to the reactor room through re-evolution from the pool at a very low rate. However, for calculational purposes, it is assumed that only 99.8 percent of the iodine dissolves in the pool j . water with the remaining 0.2 percent of the iodine and all of the noble gases, krypton, and xenon being released into the reactor room atmosphere. The release assumptions are used to determine the radiological consequences due to the Design Basis Accident of a Fuel Element Cladding Failure (Section 6.3.4.2). Of the 7 cunes that could be released in the cladding failure of an average fuel element, f' less than 3 curies would be radioiodine while the remaining 4 curies would be the l noble gases, krypton, and xenon. Therefore, the concentration in the water would 6-13

7 be less than 0.05 pCi/cm3, while the air concentration would be less than 0.005 3 pCi/cm, Calculation of Doses Following Cladding Failure. Since krypton and xenon are inert gases, the exposure due to their presence in air is from submersion. To I calculate the exposure rate, Equation 2 for submersion exposure from a spherical source. is used. 'Ihe gamma source strengths for the kryptons, xenons, and iodines are given in Table 1, Appendix C, for the steady state operation and Table 2, Appendix C, for the pulse operation. As previously determined, the volume of the reactor room is approximately 9.2 E+8 cm3, which is equivalent to a sphere with a radius (R) of 600 cm. 'lhe gamma source strength from kryptons and xenons for steady-state operation is calculated as 1.04 E+16 Mev/sec, while for pulse operation it is calculated as 0.04 E+16 Mev/see or a total of 1.1 E+16 Mev/sec. Using the gap activity as 0.1 percent and an 85 fuel element core, the cladding failure of an average fuel element will release 1.3 E+11 Mev/see of kryptons and xenons into the reactor room volume. 'Ihe air concentration of 0.005 pCi/cm3 previously calculated is equivalent to a gamma volume source strength (Sy) of 1.6 E+2 Mev/cm3-sec. For the noble gas mixture, the linear attenuation for air (p) is 9.7 E-5 cm-1 and the flux-to-dose conversion factor (K) is 5x105 2 Mev/cm -see per R/hr assuming a 0.7 Mev gamma photon. Substituting these values into Equation 2 results in an exposure rate in the reactor room from the calculated noble gas concentration of about 0.2 R/hr. Within an hour, the exposure rate in the reactor room will be reduced to less than 0.1 R/hr. Based upon these conservative calculations for a fuel element cladding failure during pulse operation following steady-state operation, a person could remain in the reactor room for more than 10 hours before exceeding the allowable quarterly occupational exposure limit of 1.25 rem. Standard operating procedures require prompt evacuation of personnel following an indication of excessive airborne radioactivity in the reactor room. Re-entry to the reactor room would be determined by the Reactor Facility 1 Director (RFD) when the airborne concentration was safe for normal operations

to resume.

6.3.3 Loss of Shielding and Cooling Water It is highly improbable that the reactor will lose the pool shielding and cooling water. Loss of this water can occur by only two means: the tank may be 6-14 ~

pumped dry, or a tank rupture may allow the water to escape in a relatively short period of time. However, should such an event occur during the operation of the reactor, the reactor is designed to shutdown if the pool water level is 14 feet or less above the core floor; and in any case, the loss of moderation will stop the fission process. In the event that the only mechanism for heat removal after the loss of coolant is natural convection of air through the core, the fuel element cladding would not fail and the fission products would be retained within the fuel element.2 In the event that a rupture of the reactor tank should occur and repair of the reactor tank wall is required, the core and any stored fuel elements that are present would be moved behind the lead shielding doors, so repairs can be made safely. The predicted gamma dose rates that would result from loss of pool water before the reactor core could be moved behind the lead shielding doors would be approximately 300 mR/hr near the reactor pool (10 feet) due to skyshine (air scatter), and 3.2 R/hr on the roof above the reactor due to direct gamma exposure. Predicted whole body dose rates to individuals within the AFRRI facility, but outside the reactor room, should be less than 10 mR/hr. Whole body dose rates to individuals outside the AFRRI facility beyond 20 meters should be less than 1 mR/hr. These dose rates would be reduced by a factor of 4 within I hour due to ~ decay of the fission products in the reactor core. The assumptions, equations, and attendant calculations are given in Appendix C. 6.3.4 Design Basis Accidents 6.3.4.1 Fuel Element Drop Accident. A Design Basis Accident (DBA) for the AFRRI reactor is postulated to be the occurrence of a cladding failure of a fuel element af ter a 2-week period where the saturated fission product inventory of a 1 MW steady state operation (100 hours to reach saturated fission product inventory) has been allowed to decay af ter being taken out of the operating core and placed in storage. The cladding failure could occur when the fuel element is withdrawn from the reactor pool. While the fuel element is exposed to air, a cladding failure could occur coincidentally, er due to a drop. The probability of such an accident is considered to be extremely remote. The probability of cladding failure has been further reduced under the postulated accident conditions by the j substitution of stainless steel cladding for the former aluminum cladding on the 6-15

p fuel cl;ments. The fission products r:1:ased frcm the gap will depend upon the temperature of the fuel following 2 weeks decay. This temperature is expected to be less than 50"C. The temperature needed to volatilize iodine (183 C) is, therefore, not reached and gaseous iodine should not be released. The kryptons and xenons will normally remain in the gaseous state and would be released (100

percent).

Although iodine will not be volatile under the assumed - accident conditions, a release of 1 percent of the gap activity has been assumed for calculational purposes. The _ diffusion factor (X/Q) and finite cloud correction factors (F) are dependent upon the distance from the source and are presented in Table 3, Appendix C. These parameters were based upon a Pasquilt type F stability condition with a 1 m/sec wind speed and a cross sectional area for the AFRRI 2 facility of 450 m. The methodology described in Reference 9 was used to determine the diffusion factor. The methodology described in Reference 10 for ~ defining the ratio of gamma dose from a finite cloud to an infinite cloud with the same centerline concentration was used to determine the finite cloud correction factor. Table 6-4 presents the source terms of radioactivity and thyroid and gamma source strengths in the total core for 1 MW steady state operation after 2 weeks of decay. These values are to be reduced to 0.1 percent for the gap activity and for the 85 fuel element core to determine the amount of fission products that might be released in an average fuel element drop accident resulting in cladding failure. -The total release is less than 0.15 curies with a gamma source strength of i 3.6 E+8 Mev/sec. The iodine release fraction is assumed to be 1 percent for a thyroid source strength of 1.3 E+3 rads. To calculate the whole body gamma dose to an individual outside the AFRRI facility, the following equation was utilized:II 0.25 (E )(Q)(X/Q)(F) (3) D = where: integrated whole body dose (rad) D = y gamma energy released per disintegration (Mev/ dis) E = total activity released (Ci) Q = 3 X/Q diffusion coefficient (sec/m ) = 1 I 6-16 l

q F' ratio of gamma. dose from finite cloud to infinite cloud with = same centerline concentration - 0.2 5. = constant for semi-infinite cloud dose analysis. To calculate the thyroid dose to an individual outside the AFRRI facility, the following equation was utilized:ll Dth =- (BRXx/QXsth) (4) where: integrated thyroid dose (rad) Dth = 3 / individual breathing rate (m /sec)-3.47x10-4 m3 sec BR = 3 X/Q = ' diffusion coefficient (sec/m ) ^ (DCF) (Q), iodine inhalation source term (rad) Sth = -DCF = dose conversion factor (rad /CI) Q .= iodine activity released (Ci). 'Ihe whole body and thyroid doses calculated for individuals downwind 'from the'AFRRI facility in the event of a fuel element drop accident are given in Table 5, Appendix C.- He conservative assumption is made that the - radioactive gases released in the AFRRI reactor room will be released directly to-the atmosphere without significant holdup within the facility. De current

design of the AFRRI reactor room would cause isolation of the reactor room by -

automatic closure of the ventilation pathway to the stack and would prevent excessive leakage to other parts of the AFRRI facility by the airtight access

doors.-

He calculated whole body dose is insignificant at all distances downwind from the AFRRI facility. He maximum calculated thyroid dose of 45 mrad is more than a factor of 100 below the projected thyroid dose of 5 rad, below which i no ' protective action is recommended.12 %e calculated thyroid dose for . Individuals beyond the boundary of the Naval Medical Command site within 1 which the AFRRI facility is located would be significantly less than 1 mrad.- '6.3.4.2 Fuel Element Cladding Failure Accident. Another Design Basis Accident l for the AFRRI reactor is postulated to be the cladding failure of a fuel' element I during a pulse operation or inadvertent transient following stead,y state operation at 1 MW. He assumed fission product inventories in the core are those given 6-17

previously in Table 1, Appendix C (1-MW operation) and Table 2, Appendix C (40-MWee pulse operation), and analyzed in Section 6.3.2. As discussed for the - failure of cladding in a fuel element drop accident, the likelihood of occurrence - of a fuel element cladding failure during normal operation or an inadvertent . transient is considered to.be extremely remote. 'Ihe data given In Table 6-3 as well as. Equations 3 and 4 will be used in the evaluation of this DBA to calculate the whole body and thyroid doses to individuals downwind from the AFRRI ~ facility. -Table 6, Appendix. C, presents the source terms of radioactivity and thyroid and gamma source strengths postulated to be released to the atmosphere from.the. AFRRI reactor room. 'Ihe resulting whole body and thyroid doses

calculated for individuals downwir.d from the AFRRI facility are given in Table 7,~ Appendix C.

'Ihe same conservative assumption made for-the fuel element drop' accident regarding-the prompt release of radioactive material to the atmosphere has been made for this DBA. 'Ihe calculated whole body dose is insignificant at all distances downwind from the'AFRRI facility. 'Ihe maximum calculated whole body dose is less than

2. mrad or a factor of.500 below the projected whole body dose of 1 rad, below which no protective action is recommended.12 'Ihe maximum calculated onsite thyroid dose of 57 mrad is nearly a factor of 90 below the projected thyroid dose of 5 rad, below which no protective action-is recc.3 mended.12 'Ihe calculated thyroid dose for individuals beyond the boundary of the NMC, NCR site would be significantly less than 1 mrad.

4 6-18 +

q r~

6.4 REFERENCES

1. Amendment iP17 to the Technical Specification, Facility License No. R-84 for AFRRI-TRIGA Mark-F Reactor, Docket No. 50-170. 2.. General Atomic Company, The U-ZrHx Alloy: Its Properties and Use in 'TRIGA Fuel--MT Simnad, GA Report No. 4314 (San Diego, California: February 1980). 3. General Atomic Company, TRIGA Low-Enriched Uranium Fuel Quench Test, GA Report No. A15384 (San Diego, California: July 1980. 4. Lederer, C. M., 3. M. Hollander, and I. Perlman, Table of Isotopes, Sixth Edition (New York: John Wiley & Sons, Inc.,1967). 5. Final Safeguards Report: AFRRI, Revised Edition, Docket No. 50-170 (General Dynamics and Holmes & Narver, Inc., March 1962). 6. Nuclear Science and Engineering, Vol. 3, 145-150 (1958). ~ 7. ' Meek,L M. E., and B. F. Rider, Summary of Fission Yields for U-235, U-238, Pu-239,~ Pu-241 at Thermal, Fission Spectrum, and 14 MeV Neutron Energies, APED-5398-A (Revised) (October 1,1968). 8. Bolles, R. C., and N. E. Ballou, Calculated Activities and Abundancies of j U-235 Fission Products, USNRDL-436 (August 30, 1956). 9. U.S. Nuclear Regulatory Comrnission, Atmospheric Dispersion Models-for Potential Accident Consequence Assessments at Nuclear Power Plants, Regulatory Guide 1.145 (for comment), (August 1979). 10. Slade, David H., ed., Meteorology and Atomic Energy--1963, Air'kesources

Environmental Laboratories, Environmental Science. Services Admin-

. istration, U.S. Department of Commerce; prepared for-U.S. Atomic Energy Commission (July 1968).' 11. U.S. Nuclear Regulatory Commission, Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors, Regulatory Guide 1.3, Revision 2 (June 1974). 12. Manual of Protective Action Guides and Protective Actions for Nuclear Incidents, EPA 420/1-75-00 (September 1973). a 6-19 1

-p 7.0 ADMINISTRATION - 7.l' GENERA 1, According to Department of Defense (DOD) Directive 5105.33 (dated May 11, 1972, and included herewith), the Armed Forces Radiobiology Research Institute (AFRRI)is designated as a subordinate unit of the Defense Nuclear Agency (DNA). . The Defense Nuclear Agency is a designated agency of the Department of Defense and, as such,'is under the direction,' authority, and control of the Secretary of Defense. DNA is responsible for management of DOD's nuclear weapons testing programs, nuclear weapons' effects research programs, and nuclear weapons stock-pile. DNA' provides advice and assistance on nuclear weapons matters to the Secretary of Defense, the Military Departments, the Joint Chiefs of Staff, other - DOD components,- and other government agencies, as - appropriate and when requested. j - The mission of AFRRI is-to con' duct scientific research ~ in the field of' radiobiology and related matters essential to the medical support of the Depart-ment of Defense. ' The. main organization of ~ AFRRI consists of the Director, the Deputy ~

Director, and the' supporting staff. A Board of Governors is also established to advise the Director of AFRRI in matters of professional policy direction and related issues.

The Board of Governors consists of the Director of DNA as -Chairman, and the Surgeons General of the Army, Navy, and Air Force. The Board of Governors ensures equal support and participation by the three services in 4 Institute activities and a. balanced response by the Institute to the needs of each b ' service. ' The supporting staff of AFRRI consists of both military and civilian person-nel. In accordance with the triservice nature of the Institute, military personnel are provided on an approximately equal basis by each military department. The cost of maintaining and operating AFRRIis borne by DNA. The size of the staff is - determined by a Joint Table of Distribution, which is developed by the Director of -DNA and approved by the Joint Chiefs of Staff. The present internal structure of AFRRIis shown on Figure 7-1. l L 7-1

U q Defense Nuclear Agency De,ecio,, DN A Armed Forces Radiobiology Research Institute { 31 August 1984 OHece of the Director AFRRt Die n tor Scienrefic Advesor '"H'U""# S*IY and Health Ofice E mecutive Of fecer / Special Staf f Civilian Personne: Office Education Proipasn Courdanaton I I I I I Physiology Department dehavioral Sciences Biocheemstry Department F aperimental Hematology Radiation Sciences Y General Physiology Depanment Physiulogic.I Chennste y PJ Division E xperimernal Psychotoy Division Inununology Diviseun Ha<hation Sources 0"""" Neurophysiology Molecutar Biology Cellular Rashobiology Division Physiological Psychology Div sion Devision Radiological Physics Cellular Physiology Imenunological Hematology Devision ' Division Chermsar y Nuclear Sceences Division Division. Biological Spectroscopy Divis+on g I I I I I Admemstraten Computer & Electronics Comptroller Of fice Logistes & Enguieering Radiahon Safety Veterinary Medicine Department Depar tment Depar tment Department Department F unans's and Accounung Ademnastrative Services Mainienarse Division Design and Fahncateal Radiation Heaith Laboratory Animal wiss n Phy sics Division Resources Dwision Hardware Divesson 'd**" Ys Comparative Pathology 50f t*d'e Devision vis n as usinmuy Dwison Dwision Supply Dwision Technical Graphecs and Publicat.ons Acieuisitim Management Division Division Figure 7-1 AFRRl Organizational Chart

y q,, m. y L{ l + 4 I '?.' ~ 7.2 LTHE' OFFICE OF THE DIRECTOR. The' Office of the! Director of AFRRI consists of the Director, the Deputy

Director, and the -Scientific _ Advisor -to. the Directorate.- Two committees, the

Reactor and;; Radiation Facility Safety Committee (R'RFSC) and the Radionuclide

,: and X-Ray: Safety-Com'mittee (RXSC), report directly to.the Director concerning. e(' safety -matters 'at: the Institute. The RRFSC is responsible for advising the 5 Director:in matters concerning the ' safe operation of the AFRRI-TRIGA reactor (and shalll be' discussed later. The=RXSC exercises no control over the reactor and its operations and, therefore, will not be discussed further. ThS Director of AFRRI must be a medical department officer of either the ~ rm', slavy,.or. Air Force,1who.is technically qualified in the ' fields of medicine, m y radiobiology,"and radiation effects research.L ~He is appointed by the Director of i DNA af ter consultation with the Board of Governors. He is directly responsible to ! the Director of DNA for the overall administration and supervision of the Institute. , Since the Director has overall responsibility' for the safe operation ~ of. the AFRRI-j ; lTRIGN! reactor,1 he. is ' authorized to. communicate directly; with-the Nuclear- ~ y RegulatorylCommission without priorL knowledge or approval of 'any other office, - including: DNA-headquarters. The Deputy. Director assumes the responsibilities of : 4 the Director in his absence. The Scientific' Advisor to the Directorate of AFRRI must be technically and t administratively qualified inithe fields of radiation, radiobiology, and radiation . effects research. He~ must have recognized capabilities, such as demonstrated ~ M fexperience in managing research programs, high professional stature as evidenced ,; F . byj publications'~ in ; technical journals. -and - appointments or membership in i professional societies,~and additional education above the Bachelor's degree level. v i He is'directly responsible to the Director of AFRRI for the overall development and ~ coordination of the multidisciplined research programs conducted at AFRRI'. These (,

responsibilities include Ladvising the; Director on, technical matters, providing.

s L. technical direction to the diverse professional staff at AFRRI, and acting as liaison )between 'AFRRI and other agencies. m 7-3 g y e,.. m.

p. [7.3 ' REACTOR AND RADIATION FACILITY SAFETY COMMITTEE ~ The Reactor and Radiation Facility Safety Committee (RRFSC) is _ directly i responsible to.the Director of AFRRI. The committee oversees the safety aspects of the following facilities at the In'stitute: o ' AFRRI-TRIGA Reactor o LINAC Linear Accelerator 0 Cobalt Facility 0 Theratron Facility - o X-Ray Facility Any other machine capable of producing radiation in excess of I rad /hr in o an area which can be occupied at any time. - The safety' aspects of these facilities include the physical facilities, the planned ope ations, 'and the' qualifications of. supervisory and operating personnel which - relate to the safety of the _ Institute, its staff, the public, and the environment. . With respect -to the AFRRI-TRIGA reactor, the RRFSC reviews all radiological . health and safety matters concerning the reactor and its associated equipment. Regular (permanent) RRFSC membership shall be composed of: 0 Chairman,'as appointed by the AFRRI Director b 'o- ~ Head, Radiation Safety Department, AFRRI o Reactor Facility Director, AFRRI o One to three non-AFRRI members, as appointed by the AFRRI Director, who are knowledgeable in fields related to reactor safety. At least one - shall be a Reactor Operator or a Health Physics Specialist. Special (temporary) RRFSC members may include: 0 Other knowledgeable persons, as appointed by the AFRRI Director, to -serve as alternates for the one to three non-AFRRI regular members listed above. Voting ad hoc members, invited by the Director of AFRRI, to assist in 0 review of a particular problem. The RRFSC or a subcommittee thereof meets at least four times each ~ calendar; year. The full .RFSC meets at least semiannually. A quorum of the RRFSC for review consists of the Chairman of RRFSC (or his designated alternate) and two other members (or alternate members), one of which must be a non-AFRRI. member.. A majority of those present must be regular members. 7-4 ~'

7 All RRFSC decisions are binding upon being endorsed by the Director of AFRRI.~ 7.4 RADIATION SCIENCES DEPARTMENT The Radiation' Sciences Department. (RSD) supports the AFRRI scientific research program by providing, operating, and maintaining radiation sources, which include the AFRRI-TRIGA reactor. ' The o Radiation Sources Division (RSDS), within the Radiation Sciences Department, has sole responsibility for the operation and maintenance of the AFRRI-TRIGA reactor and other assigned radiation sources. RSDS is composed of various branches, each branch being responsible for a radiation source. The - Reactor Branch is the branch within RSDS responsible for the operation and maintenance of the AFRRI-TRIGA reactor. 7.4.1 The Reactor Branch The Reactor Branch of the Radiation Sources Division operates, calibrates, and maintains the AFRRI-TRIGA reactor and= associated systems in compliance with appropriate regulations. The Reactor Branch consists of the Reactor Facility Director, the Reactor Operations Supervisor, and the Reactor Staff. The Chain of Command of the Reactor Branch is given in Figure 7-2. - The Reactor Facility Director (RFD) is directly responsible to the AFRRI Director for operational, technical, and safety matters pertaining to the _ utilization of the AFRRI-TRIGA reactor, and for ensuring compliance with NRC licenses and regulations as well as internal RSD procedures and AFRRI instructions. The Reactor Facility Director is responsible to the Chief of the Radiation Sources ( Division for administrative matters. The RFD shall possess an NRC Senior Reactor Operator's License for the AFRRI-TRIGA Mark F reactor. The Reactor Operations Supervisor (ROS) is responsible to the RFD for the L efficient and safe operation of the AFRRI-TRIGA reactor on a daily, routine basis. The ROS shall possess an NRC Senior Reactor Operator's License for the AFRRI-TRIGA' reactor. . Reactor Operators (RO's) are directly responsible to the Reactor Operation Supervisor for the safe and legal operation of the AFRRI-TRIGA reactor in compliance with NRC licenses and regulations, and internal RSD and AFRRI I instructions. RO's shall possess an NRC Senior Reactor Operator's License or an NRC Reactor Operator's License for the AFRRI-TRIGA reactor. 7-5 E_

[$, p [ 7 yw

w 3 i

' m-fS s w.:. ~ , ~ jj i < w,, t I 2 ~ q 1 1 ~ I + ^4 ? 7.f RADIATION SAFETY DEPARTMENT (The1 Radiation { Safety - Department.(SAF) is. responsib!e for developing and - ~ ~

maintaining a comprehensiveLHealth Physics Program, encompassing all sources of

~ m , 3,' ,fradiation'within AFRRI. :The Chairman of the_ Radiation Safety' Department must 1 i be~ academically a$d technically qualified in-the field of r'adiological safety and! ^ * ~ ' ~

health physics as they relate t'o reactor operations. He is directly responsible to l'the (AFRRI Director. for :the activities of the Radiation ~ Safety Department. If y_-

, ' 4possible,Lhe should _be alBoard Certified Health Physicist of thS A'merican Board of. ~ NHealtii Physics.'iAs Chairman of SAF, he serves as a permanent voting member of [the Reactor and Radiation Facility Safety Committee. o >Tlie LRadiatiori-l Safety c Department. _ is responsible for - developing and main ltaining an adequate ' radiation ~ protection' program-for. the Institute. This -. s [inclu' des'rnaintaining complete, personnel radiation exposure file's, implementing and

maintaining ~an ALARA program, i.e./ reducin'g radiation' exposures to'the staf f and

. the public to _lefels as low as reasonably' achievable (ALARA), supervising adequate - R Iienvironmental m'onit'oring pirograms, and ensuring tha' all applicable radiological-t Tsafety : regulations ' an' fdirective's of !the ? Department of / Defense,.- the Nuclear Mi d

• Regulatory bommission, and other a' 'encies'are fdllowed.' To ensure continuity in

~ g sthedreahof) radiatioin. protection at.-5FRRI,;particularfattention 'is :given -to i e 4... -- " centering responsibility for radiological safety in experienced and highly qualified civilian ~ personnel where possible. 's n W, (7.6' NAVAL MEDICAUCOMMAND,5 NATIONAL CAPIT5L-REGION ~. 5' l The Nhval _ Medical Command, National Capital Region (NMC, NCR)-is one,of a ' the major medical research centers in the greater Washington area. The close 5 8 ~.. ~

~

proximity of J NMC,- NCR,. the National Institutes of Health, and the National c

Q L Library of Medicine to' AFRRI gives the AFRRI staff the' advantage of having ready ~ ~ taccess to a well-established and extensive research community. 4 u C + 7 s a v I-4 .-7-6 . : s 7 .Ig9

r,- . Director, AFRRI

Chairman,

_ Administrative Operation, Radiation Control Safety, and Reactor & Radiation ' Safety-Emergency - Facility Safety Committee Department - Control 1 I -l l l I e Advisory

Chairman, Advisory I

Radiation I i Sciences Department I l-1 I I l-1 I - Chief, Radiation I l I ^ . Sources Division 1 l I l-1 l E------ Reactor Facility Director

---

I Reactor Operations Supervisor Reactor Staff

• FIGURE 7 2 REACTOR BRANCH CHAIN OF COMMAND 7-7

.~. p. l The Naval Medical Command, National Capital Region hosts several medical I These may or may not be under the commands and organizations in' the military. direct military command of NMC, NCR. The separate military commands and units at NMC, NCR are as follows: o' Armed Forces Radiobiology Research Institute (AFRRI) o' - National Naval Dental Center (NNDC) o' Naval Health Sciences Education and Training Command (NHSETC) o Naval Medical Data Services Center (NMDSC) o Naval Medical Research & Development Command (NMRDC) o. Naval Medical Research Institute (NMRI) o. . Naval School of Health Sciences (NSHS) o Personnel Support Detachment (PSD) o Uniformed Services University of the Health Sciences - The Naval Medical Command, National Capital Region itself falls under the command and = coordination control of the Commandant, Naval District, Washington, D.C. AFRRI is housed at NMC, -NCR under the conditions of a host-tenant agreement. This agreement specifies that, although AFRRI is under the military ~ command of the Defense Nuclear Agency, NMC, NCR provides certain logistical support to AFRR1 on either a reimburseable or nonreimbursable basis. NMC, NCR supplies to the military personnel of AFRRI certain services such as officer and - enlisted quarters as available; medical and dental treatment; and access to the Center's stores, services, and recreation facilities. NMC, NCR specifically supplies to AFRRI certain services such as road, parking, and ground maintenance; trash removal; maintenance of AFRR1 facilities; utility services such as electricity, sewage, and water; security protection for AFRRI's perimeter and security personnel in response to emergency security situations at AFRRl; and fire protection which includes fire safety inspections, training of AFRRI personnel in - firefighting techniques, routine' maintenance of AFRRI fire extinguishers, and firefighting equipment and personnelin response to a fire at AFRRI. 7-8

q The National Naval Medical Center itself falls under the command and coordina-tion control of the Commandant, Naval District, Washington, D.C. AFRRI is housed at NNMC under the conditions of a host-tenant agreement. This agreement specifies that, although AFRRI is under the military command of the Defense Nuclear Agency, NNMC provides certain logistical support to AFRRI on either a reimbursable or nonreimbursable basis. NNMC supplies to the military personnel of AFRRI certain services such as officer and enlisted quarters as available; medical and dental treatment; and access to the Center's stores, services, and recreation facilities. NNMC specifically supplies to AFRRI certain services such as road, parking, and ground maintenance; trash removal; mainten - ance of AFRRI facilities; utility services such as electricity, sewage, and water; security protection for AFRRI's perimeter and security personnel in response to emergency security situations at AFRRI; and fire protection which includes fire safety inspections, training of AFRRI personnel in firefighting techniques, routine maintenance of AFRRI fire extinguishers, and firefighting equipment and personnel in response to a fire at AFRRI. i L 7-9

n .g 99 's uay 11,1972 'i NUMBER 5105,33 %~ %, ASD(C) Department of Defense Directive SUBJECT Armed F~orces Radiobiology Research Institute (AFRRI) R efs. : (a) DoD Directive 5105.31, " Defense Nuclear Agency (DNA). " November 3,1971 (b) DoD Directive 5105.33, " Armed Forces Radiobiology Research Institute (AFRRI), " November 20, 1964, (hereby cancelled) I. GENERAL A. Established pursuant to the authority vested in the Secretary of Defense, the Armed Forces Radiobiology Research Institute (AFRRI) is designated a subordinate unit of the Defense Nuclear Agency (DNA). B. AFRRI shall serve as a radiobiology research laboratory for the Department of Defense. Other federal and civilian agencies and institutions may utilize AFRRI services as agreed upon by the Secretary of Defense or his designee. II. MISSION The mission of AFRRI shall be to conduct scientific research in the field of radiobiology and related matters that are essential to the medical support of the Department of Defense. III. ORGANIZATION A. AFRRI shall consist of a Director, a Deputy Director, and a supporting staff. l 1 The Director, AFRRI will be a medical department officer in Grade 0-6 nominated by the appropriate Service and appointed by the Director, DNA after 7-10

{ ~ ~ 7 c.casultatie.n with (he Board.of Gm rnars. Normally. the position of Director will rotate among the Sc2 viecs. His tour of duty 'shall be consistent with professional and military require-ments and applicable DoD Directives and Instructions. 2 The Deputy Director will be an officer in Grnde 0-o , or 0-5 from one of the Military Services not represented by the Director. - 3. - The staff shall' be constituted of military and civilian personnel authorised by a Joint Table of Distribu: ion (JTD), developed by the Director, DNA, and approved by the Joint Chiefs of Staff. Insofar as possibic, the milita y members of.the supporting staff shall be provided on an equal basis by the Military Departments. B.. . In exercising command of AFRRI, the Director, DNA, will be advised in matters of professional policy direction and related matters by a Board of Governors. 1 The AFRRI Board of Governors shall consist of the Director, DNA, as Chairman, and the Chief Surgeons General of the Army, Navy, and Air Force. 2 The Director, National Institutes of Health, and the Director, Division of Biology and Medicine, Atomic Energy Commission, or their designees, may be invited to pasticipate in meetings of the Board of Governors when matters of mutual interest are being considered. IV. FUNCTIONS Under the direction, control and authority of the Director,. DNA, and in accordance with appropriate DoD Directives and Instructions, AFRRI shall: 'A. Operate facilities for conducting research on the biological effects of ionizing radiation and disseminate the results. B.' Conduct advanced training in the field of biological effects of ionizing radiation to meet the internal requirement ci the ~ Institute, and those u hich may be developed and requested by other DoD Components. 2 7-11

Y,. May 11, 72 5105. 33 C.. Operate facilities for the production of short-lived radioisotopes for use in direct or collaborative research performed by the Institute. D. Perform such other functions as may be assigned. V.. AUTHORITY-The appropriate provisions 'of reference (a) apply to the Director, . DNA, in exercising command of AFRRI and fulfilling the functional responsibilities implicit in this Directive. VI. ADMINISTRATION A. The Military Departments will assign military personnel to AFRRIin accordance with approved authorizations. Procedures for such assignments will be as agreed upon between the Director, DNA, and the individual Military

Departments..

B. Maintenance and operation of AFRRI normally will be programmed, budgeted and financed by DNA. This provision does not preclude AFRRI from participating in reimbursable activities subject to concurrence by Director, DNA.' C. The pay, allowances (including subsistence) and permanent change of station travel costs of military personnel assigned to AFRRI will be borne by the Military Department from which assigned. VII. EFFECTIVE DATE AND CANCELLATION . This Directive is effective immediately. Reference (b) is hereby superseded and cancelled. .d i A puty Secretary of Defense 3 7.

7 APPENDIX A Joint Wind Frequency Distribution for 1960-64 and 1964 inclusive A-1

x TABLE 1 Joint Wind Frequency Distribution for 1960-1964 Inclusive p -JOINT WINO FWEQUENCy DI57wlHUTION MY STAHILITY CLASS DATA PERIOD: 1960 THRu 1966-arru[-TRIGA RFACTOR class: 9450Vilya c WM5 " - gTARILITY .fle tWa85' M!l?" ^.88 MiEstS"" s ' TABLE GENERA Os 10/09/80. 20.02.07. DAMES ANO 900RE Joe Not 11985-001-27 wlNO WINO SPEED CATEGORIES (METERS RER SECONU) MEAN SECTOR 0.0-1 5 1 5-3.0-3.0-5 0 5 0-7.5 7.5-10.0 >10.0 TOTAL SPEED

d

':N 8:d 8:2! 8:8! 8:8! ': N NE -5 - 14 0 0 0 0 23 2.08 5.47 0.00 0.00 0.00 0.00 e.99 1.j{2 .g 0.00 0.00 0.0 0.00 .0 1 52 6.08 0.00 .0.00 0.00 0.00 7.60.g .0{ .05 0.00 0.00 0.00 0.00 1 27 4. 6 0.00 0.00 0.00 0.00 5.47 .01 .03 0.00 0.00 0.00 0.00 .04 0.08 0.0 7. -888 h08 9 6. 0.00 0.00 .06 .0 05 SE 5 16 0 0 0 0 21 2 17 -1 53 4.46 0.00 0.00 0.00 .0.00 6.3d 0.0g 3.kg 0 00 0.00 .0 .0 .04 0.00 .3A. 3.kh 5 0.00 0.00 0.00 0.00 .g 0.00 0 00 0.00 0.00 .0g 1 33 1 52 7.29 0.00 0.00 0.00 0.00 8.81 0.00 .0 00g 0.0g .0 .0 0.00 8188 2:31 8:88 8:28 8:38 8:38

n Sw 1

5 0 0 0 0 6 2 25 .30 1 52 0.00 0.00 0.00 0.00 1.82 .0g 0.00 0.00 0.0 0 00 .0{ .00 0.00 0.00 0 00 0.00 2.63 l.22 12{ 0.00 0.00 0.00 0.00- ,.0g . 0 g* .{ .30 4 26 0 00-0.00 0.00~ 0.00 4 6 .00 .03 0 00 0 00 0 00 0 00 .03 3.kh ~.08 0.00 0.00 g.0k 0 15 2.4 .00 0.00 0 00 n.0 36 0.00 0.00 0.00 0.00 3.95 .9l .{ 0.0g 0.0g .pg 0.00 0.00 .0 1 2T!

3. 4 0 00 0.00 0.00 3.00 6.56

.g 0 00 0 00 0 00 .g 0.0g .0 g 1 52 ~ 6.04 0.00 0.00 0.00 0.00 7.60.g .05 0 00 0.00 0 00 0.00 ":t3 ":t3 { TOTAL 110 219 0 0 0 0 329 1 80 ":14 -

• • 13 8:88 8:88 8:88 8:88

M KEY XAx M8ER Or OCCURRENCES

!!!PERbkN UR C ShkS A-2

r q TABLE 1(cont'd) JotNT w!ND FWEQUENCY ul%TW!bOTION Hy STAd!LITY CLASS DATA PERIO0t 1960 THRu 1964 p AFFWl-iWlGA WEACTop CLASSI P4500ll 6. .c gTARILITY .fleiM85'47sM?"t.%9hE"tS"- E P a ""-

TABLE GENERAftot 10/09/eu. 20.02 07.

n4MES AND 400WE JOB NOI 11985-001-27 WINO SPEED CATEG0kIES(*ETERS PEW SECONO MEAN WINO -0 0-1 5 't.5-3 0 3 0-5.0 5 0-7.5 7.5-10.0 310.0 TOTAL SPLED SCCTOR-i.ll i.4 2.!! 0.08 0.08 0.08 5!'i ~~~' 8; n

• • 8

"*8

• • 8 th 2 60 Ne.

2.73 0.00 0.00 0.00 6.52 1 61 2.l3 li U 8 "'8 8 til 2 37 s Ew 2 24 2.20 2.09 0.00~ 0 00 '0 00 6.54 P, At H oa8 oa8 oog ty 2 33 .E 4.8 1 63 2.24 1 21 0.00 0.00 0.00 si 3* a*! a8 ili 2 26 5 EsE ':ii 2:t? 2:33 8:88 8:88 8:88 533 SE 36 48 51 0 0 0 '133-2 53 1 2% 1 76 1.87 0.00 0.00 0 00 4.87 8.$% .t. oo8 a8 oa8 at .ki 2.2.1 1 3 2 80 SSE 9 0.00 0 00 0.00 i.St -)? At 511 o8 o8 o08 di 3 07 5 22 3.59 13.67 0.00 0 00 0 00 20.08 .d .22 .45 0.00 0.00 0 00 1.25 0.00 0'.00 008 3. .N 1.II oa8 o o8 oo8 -. 8 a! t?

2. 6

':$9

8!

2:4 8:88 8:88 8:88 ':it 5.. ' WSW 24 20 58 0 0 0 102 2 79 0.00 0 00 0.00 3 74 3 i3 2.b3 o8 "8 a8 63 2.,9 .7g .88 0 64 2.46 0.00 0 00 0 00 3.63 11 aa8 a8 o8 til 3 07 -lg- .N. 2.46 0.00 0 00 0 00 3 49 .6 j;2 36 li aa8 6a8 oo! til 3 08 N.

0

'l

':ll 8:88 8:88 8:28 ':3* 0 0 0 12 3 03 -NNW 16 0 .59 1.1g 2 82 0.00 0 00 0.00 6.51 oo8

• a8 oa8 ti!

2 8i 4*} St 28 N

3 8:88 8:88 8:88 6:$3 2:b1 2:It c^'"

..!.f c" 2fb9 2 65 $$k 679 'g 0 0 0 TOTAL 'f:lf 't:lt 3.ti 8:88 8:88 8:88 t"8:33 KEY xXA N R OF OCCURWENCES x55$ERkENTOkbuRRNkk C SS A-3

'A TABLE.1(cont'd) 4 ' JOINT Wlan FwEquCNCY UlsiwlmuTION Hy STAn!LITY CL Aw ~ DATA PERIOut 1960 THdu 19%d. .'sT A4It. f f t CLA%%t D A%'Mlt L L AFF41-Tw[Ga wcACTod DATA SouwCE a AT !0N AL a [wco r.. As.e.. 0.C

• TW SUA.
  • 0.

e WING %ENSow MFIGNT 7.00 ET9s aFFwi T art.E GENEwAf fo t = 10/n9/No. "20.<12 07 nAuts ANO *00dE JO4 Not 119M5-001-27 wlNo wlND CPEEn C A f tnor>IES PtL TNs *Ed SFCorm

• EAN SECT 00 0 0-1 5 1.5-3.0 1.0-3.0 s.n-7.s 7.s-10.0

>10.0 TOTAL SetE0 l - NNE' ,12 '14 104 IM 0 0 192 3.13 . 23 .. 7 * /.0/ . t r. 0.00 0.00 1.71 .01- .09 .4% .09 u.00 0.00 44 .NE 33-72 100 IS 1 O 201 1 30 '. 25 1.40 't.44 39 07 0.00 1.90 . 03 .I6 .?) .o) 00 '0.00 .*6 -ENE' 17 68 - 41 .33-1 17. 1.61 2,6 0 0 194 3 26 0 0.n0 0.00 1.71 .04 .16 E 16' ~ 46 '.19 .06 0.00 0.00 44 ^Q IS 0 . 0 132 3 25 .31 .8% 1.tr .79 n.00 0.00 2.A2- .06 .10 .14 .01 0.00 0.00 .11 E%E. 7 -- 15 6e t 0.00 0.00 1.6N 0- - 0 v7 3.31 .14- '.68 .A5 2 02 .0M- '.10 .0J 0.n0 4.00 ./2 - SE ' 24 el st-9 2 0 11l 3.24 .47 41 1.11- .t7 44 0.00' 7.19 .0% .Os .I1-07 00 0.00 .26 %%E 1 n7 10 is 0 0 243 3.65 .14 1 30 ~ 2.17 .A6 0.00 0.00 4.42 .02 .19 . 12 .04 0.00 0.00 .51 .5 43. 210 106 752 12 -0 122J -3 98 .N3 4.09 11.71 4. '* w 7.3 u.00 23.f5 .10 44 1 61 .M . 0.1 0 00-- 2.14 - $$W. 2/ 77 20h - 14 43 1 50 4.00

1. ' 4

. 25) - 1 -393 3.99 1 .02 F.6J 05 .14 47 .tr - SW 25 65 1213 40 '.03 .00 40 3 1~ 261 3.65 . 49 - .87 7. 4. 7 - 1H .nh .a2 4.no .06 .10 .29 .09 01 .00 t:9W ~' 2 3 17 115 61 o 0 Ele 3 75 . 65 .72 /.23 .o] 4.00 .0.00 4 23 .05 .04 .26 .to 0.00 0.00 50 49 1 2 E70 4 10 e 1 17 2.111 .2 77 19 45 .16 04 4.27' -. 0.1 04 .24 .I1 .u? .00 ..so WNW - 7 10 19% M7 19 0 3; 4 4.39 44 .98 3.59 1.e9 .19 0.00

e. L9

.02 .0F 42 ./0 .nt 0.00 .J3 .NW 11 45-27% 139 /7 1 451 4.63 .2 .97 4 3v 2.To .%2 '.an 9.76 .0 .10 .57 . 12 .06 .01 10J NNW-1 16 2%0 99 11 1 621 4.27 .23 74 5.05 1.92 ./t .02 8 11 .01 09 .59 ./1 03 .00 .96 N-1 - l.07 a. 'g o

1. F [r.

55 217 9 5 0 39* 4.07 .2 .10 0.00 7.63 .0I 13 .53 ./l- .41 0.00 90 - CALM 90 90 CALM 1.75 1./5 1 .21 . TOTAL. 3 919 7 73 R 1072 92 4 5150 1.d8 e .6.

17. rts 5J.55 19.44 1.79

.16 1u0.00 2 10 9.29

2. 43

.21 .42 !!.75 KEY - xxx NH*lsER OF OCCoWwtNCE% mas PERCENT OCCON9ENCES THIS CLASS

*AA PERCENT OCCtJWAENCES ALL CLA5SES i

A-4 F ++r-a y wie O e w W

y p. TABLE 1 (cont'd); s JulNT w!NU FwE9'ffNCY 8)! % f 41mJ T ilsN My STAHILITY CLA%S . DATA pep 100 ' 1960 inwie 1 * * %TARILITY CLatt:-wasoulLL o AFF wl-Tu lt:A REACfod DATA %00wCE! NATIONAL.AldwQfT. 4ASM.. O.C MfTHESUA. 40 4 . wlNO St;NSop MElGHT8 F.04 MrTF.*$ - AFFdi .TAHLE GENEWATEDt. 10/09/80. 2n.02.07. oApe,s An0 uoowE acq wo: I l9 sts-001 -27 PINO - w [ND % PEED CA tt' GO-IP.5 t ut, TF us uEd CECONup => AN SECTOR- ~0 0-1 5 1 5-3.0 .J.0-5.0 5.u-I.5 7.5-10.0 310.0 TOTAL SPEE0 'NNE: 42 175' 15 0 44% /* A 1064 4.d9 .21 61 1.71 7.14 .18 44 51I .30 .?9 .mo-1 07 .19 .07 2.39 NC -- '.54 199' 464 517 al 12 132J 4.10 .26 95- /.27 2.%1 40 0% 6.47 .12 45' l.06 1.14 19 . 0.1 3.04 .ENE ' 75 2td 577 447 77 71 1305 4.54 .31 1 07 2.%H 2.19 .1% 10 6.6d . 17 90-1.70 1.n7 .id .95 1.12 E 66 153 237 142 30 1 621 38s1 .32- .75 1.16 69 .10 .01 3.0* .15 .3% .S. .12 - .ns .01 1.42 'ESE 65 173 174 94 ?9 2. 490 3.77 .32 .60 .d7 .*6 .14 .01 2.40 SE. '-.15. ' .?M .41 .71 .nm .00 1 12 64 129 ?im 113 11 1 523 1 08 1 63 1.00 ..is 0% .00 26 ..5 .29 67 ./6 03 .00

1. 4 SSE 1

142 762' 131 30 9 6 9 4 13 -.In .69 1 28 .65 .15 .0* 2.94 .04-37 .60 . 10 .01 .02 1.39 5 95 368 1238 1026 142 13 2992

• .56 46 1 80 h.06 5.07.

.69 .06 14 10 .22' . R t. ^7.93 2.14 . J2 ' .03 6.56 SSw 41 166 SSP 669 191 29 164f S.17 .20 .Al 2.70 3.27 9) .14 d.06 09 _.18 1.26 1.93 44 .06 3.76 Sw - 46 !?) 282 - 307 90 >0 966 4.90 .23 .60 1 38 1.i0 .*6 .10 4.25 .In .29- .e4 70 .21 .0% 1199 = tSW 33 117 1M9 _1** 14 14 561 4.33 .21 .5T 97 .70 .11 .07. 2.65 0 .27 4I . 13 0A 03 1 23 t) ' - 6 67 140 231 99 76 $99 5.39 . 9 .31 .6M l.11 .*4 .12 2.9J .09 .1% .37 .93 .23 .05 1.31 94W - 7 A5 74% 766 407 til 1701 6 67 .3 42 1 20 1.M5 1.99 76 H.32 . 6 19 96 19 9) .14 3 14 NW-28 76 354 157 630 dw 245* 6.75 .14 .17 1.71 66 3.06 1 02 12.00 .06 .17 .M1 2.64 1 4= .uA 5.60 NNW- ~.' 2 2 41 34) 900 191 74 1814 6.22 14 40 1.67

• .*o 1.91

.1H 8.67 05 18 7A 2.0% .w4 14 4.14 N 76-la7 429 M16 2*2 2T 1711 5.41 .37 .72 2.10 3.99 1 18 .11 4.50 17: .34 46 1.d6 .%S 06

1. 46

. CALM 270 220 CALM 1 04 1 04 .50 .wo TOTAL-- 103% 2Jl9 399* I477 2w41 n>0 20**e s.20 %.06 11.1/ 29 32 34.18 12.So 3.03 1u0.00 2 36 S.28

13. Aa 19.09 5.42 1.*2 46.6h MEy era NowHEW OF HCCowwEnCF%

f 4xs p(WCENT OfCuwWkNCts T"l% CLA%5 AAA Pt.RCENT OdCowwt,.NCE S Al.L CLASSES I A-5

E K TABLE 1 (cont'd) 1 b 2 y t- _p L-u P JOIN T w !NO f df.'h FNCY 0!%f 4!ai1T inn of NTANILITY CLA%%

_

DATA PEaloot 1960 f M+i 14A* i .sfadlLITY class: JA%fsOILL P M fwl-IWIGA WFACTOW OATA source

• NAll0NAL ale 447.

4SM.. U.C MFi*t%DA. "O. ett40 sENsow ntlGHit 7.un W f t..e s AFFst T art.E GENf H Af t0 8 10/09/40. / 4. ilt. 0 7. DAut$ A40 MotAt Jod NO: 1199S-001-27 = N vlND w!ND Cpttu C ATEb@ lEs taE raws 40 ;ECONIO '#. a N E SECTOW 0 0-1 5 1 5-1.0 1.0-S.O 7.0-7.i 1.5-10.0 >10.0 TOTAL SetLO [ NNE O-63 149 >F 0 0 /30 3 60 0.00 99 d.79 47 0.00 0.00 1.61 0.00 .l* .1/ 0% 0.00 0.00 .%2 NF 0 64 1%f /H 0 0 2%* 3.50 0.00 1.0M 2.47 .44 0.00 0.00 3.w

e 0.00

.16 .16 n6 c.00 0.00 .5a E ENE o 76 14n 26 n 0 242 1.*S A 0.00 1 49 2 20 41 0.00 0.00 3.o0 0 00 .17 .32 3% 0.00 0.00 55 E O 31 na o O 114 3.In 0.00 .A0 .97 .ow 0.00 0.00 1.47 0 00 . 12 .14 ul 0.00 0.00 .27 - E<E o 4P nl 11 0 0 136' 3 45 0.00 a* 1.21 20 0.00 0.00 2.14 0 00 .10 .l* 41 0.00 0.00 .31 e SF 0 Fw 104 11 0 0 200 1.2* 56_ 0.00

1. / 4.

1.10 20 0.00 0.00 1.14 0.00 14 .7% n3 0.09 0.00 Ewh J.44 SSE O 94 134 19 0 - 0 li 0.00 1.01 7.19 10 0.00 0.00 J.41 0.00 15 11

JF 04 0.00 0.00 7Q S

0 232 641 112 0 0 1057 3 67 ~ 0.00 3. M. 10. J45 7.07 0.00 0.00 16.57 0.00 .91 1.54 to 0.04 0.no

2. I 55w 0

144 5 7 4/ 0 0 743 3.07 0.00

2. %

a. 2 1.29 o. is u 0.00 11.75 0.00 3 '. 1. 4 .19 0.00 0.00 1.71 Sw 0 136 201 /6 o 0 313 3.26 0.00 2.49 3.19 . IR 0.00 0.00 6.04 0.00 . 16 66 05 0.00 0.00 .dr wSW 0 1 31 12% 11 0 0 267 3.04 0.00 2.06 1.w6 .11 0.00 0.00 4.19 0.00 . 10 .2w 9) 0.00 0.00 .61 4 0 5a 137 /4 0 0 219 3.S7 0.00 91 2.1% . la 0.00 1.00 1.44 0.00 .11 .11 .oS 0.00 0.00 .%0 WNw 0 %I 316 100 o 0 en? . 01 0.00 30 m.96 1.97 0.00 0.00 7.33 0.00 .l? 72 7) 0.no 0.00 1 01 NW 0 d0 4 14 127 0 9 641 4.02 0.00 1.7A 7.44

1. 44 n.00 0.00 10.70 0.00

.1A 1 0H .29 0.00 0.00 1.55 NNw 0 %4 4*1 111 o o %de 4 05 0.00 .d5 1.74 1.74 0.00 0.00 4.41 0.00 .12 10% 25 0.00 0 00 1:43 - N-0

• 5 115 9) 0 0

323 3.o3 0.00

1. <. 4 5 26 1.46 0.00 0.00 d.21 0 00

.27 .76- .dl 0.00 0.00 1 14 CALM ~0 0 CALM 0.00 0.00 0.00 0.00 TOTAL 0 1440 6041 N 16 0 0 6367 3 66 0.00 22.77 e4.10 13.11 0.00 0.00 100.00 0 00 3 11 v.31 1.91 0.00 0.00 14.53 KEY-axA NUwHER OF OCCodWEr4Cd5 xxx pEGCKNT OCCUwwENCts TMts CLASS AAA PERCENT OCCliAoENCts Alt. CLASSES h I A-6 w sm-um-- m = m

7 y E-p; e TABLE-1 (cont'd) JolNT ' WINN F WtfJoetNCY UliTDI sOTION MY Miao!LITY CLASS' Def a Ptp!Of): 196o.TNwil l 'a

• 4FFwl-TWif4 #ACTOW

%TAHILITY Cl. A%C t H a %0ill t.L t HFTM Sua. 40. 0474 SOUWCF t NATIO4AL Alle)wT. e4Sd...0.C w!No StiNS0W MelbHT:

1. 0 n wt T *. %

AFFa! f ast.E GENtp ATEnt 10/O udo. /n.n?.0 7. nAst% a40 #100wt JOH Not 1194s-001-27 t!]NO - WIND % PLEO C4TtGowltis htTEus *EW SECONot mEAN . St CTOR. ..0.0-1 5 1 5-1.0 3.n-9.0 5.07.S 1.v-lo.0 210.0 TOTAL SPEED NNF~ - 4 3 '. - 109 O tl 0 23M 2 10 '.94 '1.24 .50 0.00-0.00 0.00 ?.$4 .19 .2% .30 0.00 0.00 0.0n .74 NE R 4 -. 121 54 0 u n 265 2 14 01 1.40 60 0.00 n.no 0.00 3.0L

• 1 1 20

./4 .12 0.00 0.00 0.00 .60 -ENE 111 'lle 4e n 0 0 271 .2.06 1 26 1 10 .52 0.00 n.00 0.00 3.01 . /. ./% .30 n.00 0.00 n.no 62 .F e %4 11 0 0 0 132 1 87 .7 '. m * .lS 0.00 0.00 0.00 1.50 .14 .13 .01 n.00 0.00 0.00 .JG E st.- . 66 '

• M 24 0

0 0 139 1 92 7% .5% '. 21 0.00 0.00 0.00 1.51 .1%

.11-

.n; 4.00 0.0n 0.0n .31 SE 4 . 42' 31 0 0 0 21* 1 99 1.0 't.05 .1% 0.00 0.nn 0.00 2 61 .2 .21 .07 n.00 0.0n 0.00 49 iSSE 6' 79 .43 0 0 0 Id3 2 19 '. 6 9 90 . 49 J.00 0.00 0.00 2 06 .ld .10 0.00 0.00 0 00 42 350 479 264 0 0 0 1096 2 21 51 - 3.94 5.*3 3 05 0.n0 0.n0' O.no !?.46 .80 1.n9 . ri l 0.00 0.00 0.00 2.50 %9W- ~199 443 253 0 4 0 693 . 2 36- -2 15 - 5.1% 2.As 4.u0 0.no 0.00 10.11 -.41 1 01 .54 0.00 0.00 0.00' 2 0* -. %W 34d 533 147 0 0 0 102d 2 08 3 96 6.n6 1.67 0.00 0.00 0.00 1 .69 . 79 '1 72- . 34 0.00 0.n0 0.00 .35 'ESW 431 S10 9% 0' o 0 056 1 97 4.90 6.02. 1.OR 0.00 n.00 0.00 12.00 .94 1.21 .27 0.00 0.00 0.00 2.41 0 191 216 69 0 0 0 4d2 1 99 2 24 2.46 .74 n.00 0.00 0.00 5.44 45 .*9 .16 0.00 0.00 0 00 1.10 WNW I0 196 106 0 0 0 422 2 24

1. i.

2.21 1.20 0.00 n.00 0.00 4.40 . r 45 .24 '0.00 0.00. 0.00 96 -NW 63 '190 141 0 0 0 394 2.*T .72 2 16 1 60 0.00 0.no 0.nn 4.4%

.14 41

.32 0.00 0.00 0.00 40 NNW 94 207 131 0 0 0 436 2.35 11 - 2.35 1.49 0.00 0.00 0.00 4.96 -.2 47 .30 0.00 0.no 0.00 1 00 33 143 0 0 0 670 2 23

N 18 2.15~

3.)18 1.A4 u.00 n.00 0.00 7.62 Ig.60 0.00 0.00 0.00 1.93 43- .76 .34 ~ CALM u M60 CALM 00 10.00 '3A.h '42.70 lM.32 0.00 0.00 0 00 100 00 3Ie 3757 1612 0 0 0 1 95 TOTat. - 97 7.83 8 57 3.6M 0.00 0.00 0 00 20.08 MEy sxa wtsER OF OCCOWErENCES !!!PRbkN ' C S A-7 b

TABLE 1 (cont'd) ' JOINT wl*40 FRE00ENCY UlsTWlwoT10N My STABILITY CLAS% DATA PERI 008 1960 TNdu 1964 aLLCL45%E!' arrwr-TwtoA uEACTop Sie 5N""RANb"!M?" t!n%3tS" c E!P' ~ - TABLE t 10/09/80. 20.02 07. DAMES AND M009E JOB Not 11985-001-27 w!NO SPEED CATEGORIEstMETERS PER SECOND) MEAN WING - 0 0=1.5 1 5-3 0 3 0-5 0 5.0-7.5 7 5-10.0 >10.0 TOTAL SPEED SECTOR NNE. 17J 387 710 510 78 8 1866 4 07 .39 .88 1 62 1.16 .18 .02 4.26 NE 205' 515 850 560 R2 12 2244 3 94 47 1 22 1 94 - 1.28 .19 03 5.12 ENE 269 556 853 499 17 21 2275 3.82 .61 1 27 1 95 1.14 .18 .05 5.19 E 188 379 405 163 20 3 Ilse 3 26 43 .87. .92 .37 .05 .01 2.64 Est 194 321 366 118 28 2 1029 3.19 44- .73 .84 .27 .06 .00 2.35 SE 21A 3AS 452 135 13 1 1204 3.11 . 50 .88 1.03 .31 .03 .00 2.75 SSE 145 425 704 186 10 3.42 1499 3 49 .33 .97 1 61 42 .07 .02 iS8 l'Al Ela l'13 !n .!d it*'d 5 S$w 276 878 1575 825 204 29 3787 4.02' .63 2.00 3.59 1.88 47 .07 8 66 SW 450 879 796 371 93 21 2610 3.35 1 03 2.01 1 82 .85 .21 .05 5.96 w5w '525 839 582 198 34 14 2192 2 91 1 20 1.91 1.33 45 .08 .03 5.00 W 266 405 526 304 107 26 3.34 16 3 80 .61 .92 1 20 .69 .24 .06 73 wNW. 177 394 919 973 417 151 3031 5.25 40 .90 2 10 2.22 .95 .34 6.92 NW 124 ' 427 1286 1623 657 212 6129 5.53 .28 .97 2.94 3.25 1 50 .48 9.42 NNw 152 421 1271 1110 402 79 36J5 4.96 .35 .96 2.90 2.53 92 18 7.34 N 312 698 1235 1000 247 27 3519 4.27 .71 1 59 2.82 2 28 .56 .06 8 03 CALM CALM TOTAL" 5614 9339 15806 97A5 2643 629 43815 3.98 12 81 21 31 J6.07 22 31 A.03 1 63 100 00 NOMREw 0F vaLIO OBSEdv&TIONs 43815 99.92 PCT. NOMMER OF INVALID OASEdvATIONS 31 .04 ACT. TOTAL Num8ER OF. OBSERVATIONS 43848 100.00 PCT. KEY xxx NumeER OF OCCudf.rTES xxx PERCENT OCCuppENCES A-8

3 TABLE 2 Joint Wind frequency Distribution for 1964 STAdILITY class: P450ulLL A AFF41-THIGA kFACTOR 0ATA souwCEI NATIONAL AINPORT. WASH.. 0.C OETHE50A. MO. ' WIND SENSOR MEIGHT: T.00 METFRS 4FFui TA9tE GENERATfut -11/19/80. 20.00.32. DAMES'AND MOORE J06 NOt 11985-001-27 - WINO WINO SPEED CATEGORIESIMETERS PER SECONDI MEAN SECTOR :0 0-1 5 1 5-3.0 3.0-5.0 5.0-7.5 7.5-10.0 >10.0 TOTAL SPEED NNE-0 4 0 0-0 0 4' 2.47 -0.00 6.96 0.00 0.00 0.00 0.00 6.56 0.00 .05 -0.00 0.00 0.00 0.00 .05 NE 1 5 0 0 0 0 6 2 33

1.64-8.20 0.00 0.00 0.00 0.00 9.84

.01 .06 0 00 0 00 0 00 0.00 .07 ENE. 1 3 0 0 0 0 6 2 07 1.64 4.92 0.00 0.00 0.00 0.00 6.56 .01 .03 0.00-0.00 0.00 0.00 .05 E: .0 4 0 0 0 0 4 2.47 -0.00 6.56. 0.00-0.00-0.00 0 00 6.56 0.00 .05 0.00 0.00 0.00 0.00 .05 0 4 0 0 0 0 4 2 35 ESE . 0.00 6.56 0.00 0.00 0.00 0.00 6.56 0 00 00g 0 00 0 00 .05 -0.0g .0} 0.00 4.92 0.00 0.00 0.00 0.00 4.92 0.00 .03 0.00 0.00 0.00 0.00 .03 SSE o 5 0 0 0 0 5 2 30 0.00 8.20 0.00 0.00 0.00 0.00 8.20 S-0.00 .06 0.00 0.00 0.00 0.00 .06 1 3 0 0 0 0 4 2.20 1.64-4.92 0 00 0.00 0.00 0.00 6.56 .01 .0J 0.00 0.00 0.00 0.00 .05 SSW 0 0 0 0 0 0 0 0.00 -0 00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00-0 00 0.00 0.00 0.00 0.00 Su -0 0 0 - 0 0 0 0 0 00 0.00 0.00 0.00 0.00-0.00 0.00 0.00 0.00 0.00 0 00 0.00 0.00 0 00 0.00 t:Sw 1-1 0-0 0 0 2 2 05 1.64 1 64 0 00 0.00 0.00 0.00 3.28 .01 .0 0 00 0.00 0.00 0.00 .02 ~ 0 0' 0 0 0 0 2 2 35 0.00 3.2 0.00 0.00 0.00 0.00 3.28 4 00g .0g 0.00 0.00 00g 00g .0j l.64' 3.28 0.00 0.00 0.00 0.00 4.92 .01 .02 '0.00 0 00 0.00 0.00 .03 N*J 0 0 0 0 0 0 0 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0 00 0.00 NNW 0 2 0 0 0 0 2 2 35 0.00 3.2( 0.00 0.00 0.00 0.00 3.28 00g .0g 0 00 0.30 0.00 0.0g .0j 0.00 11 48 0.00 0.00 0.00 0.00 11.48 0 00 .08 0 00 0.00 0.00 0.00 .08 18 d]l 19.0g .' CALM l 11 CALM - TOTAL ~ 4 0 0 0 0 100.0b 1 90 26 43 73.7 0.00 0.00 0.00 0.00 l1 .18 .51 0 00 0.00 0.00 0.00 .69 T C HIS CLAS PERC.NTOCCURRENCESALLCLASSkS XXX W-9

y \\ TABLE.2.(cont'd) 1 1 l STAHILITY class's PasougLL ts AFFki-TRIGA WEACTOR "t e ta "8P sk"s M ?" ^!3 # i #s " - WW ^ " - 5 c- -T A8LE GENEP AT 11/19/80.-20.00.32. DAwES ANDJMOORE JOB NO: 11965-001-27 . WINO , WINO SPEED CATEGORIESIwETERS PER SECONO) MEAN SECTOR- '0 0-1 5, .l.5-3.0 3.0-5.0 5.0-7.5 7.5-10.0 310.0 TOTAL SPEED

NNE 2

6 15 0 0 0 23 3.11 .40 1.2) 3.03-0.00-0.00 0 00 4.65 .g .Ij 0.00 0.00 .0 2.02-2.b2 3.k4 0.0g .g 0.00 0.00 0.00 7.e4 .1 .lj .20-0.0g 0.og 0.00 .y 2. 2.47. 1.82 0.00 0.00 0.00 6.46 .14 .10 0.00 0.00 0.00 12} =2.b .g 1.4g 0.00 0.00 0 00 5.05 EsE .f4 h a8 "8-8 38 3 2.00 2.83 2.63 .6 f.00 0.00 0.00 6.06 .lg .lg .g u.00 0.00 0.0g .g 1 01 1.82 2.2 0.00 0.00 0.00 5.05 . 06 ~ .10 .g3 0.00 0.00 0.00 .28 0.08 2. 2. 6.6 0.00 0.0 11. 5 3 3.27 1.41-2.42 12.93 0.00 0.00 0.00 16.77 'SSW '.08 .14 73 0.00 0 00 0.00 95 3 9 10 0 0 0 22 2.84 .61~ 1.82 2.02 0.00 0.00 0.00 4.44 .0g .lg .lg 0.00 0.00 0.0g .] l.82. . 40-1.62 0.00 0.00 0.00 3.d4 .10 ' .02 .09 0.00 0.00 .2b 0.0g .2f 0.00 1.6 0.00 0.00 0.00 1.82 .01 0.00 .0 0.00 0.00 0.00 .10 ~d' 4 6 16 0 0 0 26 2.99 f.0g 12} 3 23 0.00 0.00 0.00 5.25 .8 .g8 .0 '0.00 0.00 0.0g .y -0.00 1.2) 2.2 0.00 0.00 0.00 3.43 0.00, -0 .g 0.00 0.0g 0.0g .g 40 .61 3.03 0.00 0.00 0.00 4.0* .0j .0)- 7 0.00 0.00 .23 .0{- l.41 2.<,h 0.0g .6 0.00 0.00 0.00 4 4 .08; 3 0.00 0.00 00g .2g 3.h3 ..g .0g .8 2. 0.00 0.00 0.00 0.00 0.00 0.00 .g-6.46: 6.46-TOTAL kk 50 0 0 0 /.9 5 2 67 -24.6 24 50 51 0.00 0.00 0.00 100.00 1 39 1.*0 2.85 0.00 0.00 0.00 5.64 MEY I!! $yM8(R OF OCCURRENCE) 'C dSk5 .Axx-N $$ENI ob800kN A t A-10 k j

7 TABLE 2 (cont'd), f f sTAu!LITY class: ' pas 301LL i AFFRI-THIGA REACT 0W 500RCF. 3 - NATIONAL AINPOWT. WASH.. 0.C BEIHEs0A. MO. u.ATAIND SENSOR NEIGHT:7.00 METERS Arr a! TAHLE GENERATED: 11/19/80. 20.00.32. CAMES AND MOOPE JOB NO: 11985-001-27 ulND .IND SPEED CATEGORIESIMETERS PER SECON01 MEAN SECTOR 0.0-1 5 1 5-3.0 3.0-5.0 5.0-7.5 7 5-10.0 >10.0 TOTAL SPEED NNE. 0 '9 12 8 0 0 29 3.81 0.00 .89 1.19 .80-0.00 0.00 2.8A 00g .3g -0.00 4 .09 0.00 0.00 .99 2.0 40 0.00-0.00 3.44 0 00 .11 .24 .05 0.00 0 00 40 ' ENE. 0 8 15 5 0 0 28 3.57 1 1.49 .50 0.00 0.00 2.78 0.00 .8009 .17 .06 0.00 0.00 .32 0.00 E 2 9 - 11 6 0 0 28 3 64 .6 0.00 0 00 2.78 c.20 .89 1.09 . 0g 0.00 0 00 .32

_. 0 2

.lg .13 0.00 70 40 .10 ~0.00 0.00 1.19 0.0g .jk ) 97 g. 00g .og .gg .0j 0.00 2.tk. .20 1.30 1 29 .20 .20 0.00 5 .02 .02. 0.00 .25 . 02 .03 ~ SSE O 16 9 19 0 0 64 4.06 - 0.00 1 59

2. 8 1.49 0.00 0.00 6.36-0.00

.18 .33 .22 0.00 0.00 .73 15 127 68 0 0 232 4.17 4 S 2 . 3.48 2g.06 1{ 5l~ ?T o"8 o8 6 6.76 0.00 0.00 .20-83 "i 3.82 1.h9 -SS. .03 .20 ' 5.47 't.69 0.00 0.00 9.24 .30 .19 0.00 0.00 1 06 Sw-2 12 7 0 0 53 3 63 .20 1.19 .3. . 70 0.00 0.00 5.27 . 6 .08 0.00 0 00 .60 - .02 .14 cS. 2 7 3 10 0 0 62 4.06 .20 .70

2. 9

.99 0.00 0.00 4.17 . 02 .08-6 .11 0.00 0.00 48 1 7 28 23 ~ 0 0 59 4.34 u ' 0.00 0.00 5.86 g0 .7l 3! il 3 o o8 ti 2.78 2.j9 A 4.39 .N.. 0.00 .50 3.38 1.19 .20 0.00 5.27 .6j '. 0 2 0.00 .3g .lg ,0.00: '.06 7 2.k8 .20 0.00 7.16 ' 0.00 .6

3. 8 00g

.09 -.gj .3g .02 00g .g2 .20 .50 .5.27 1.k9 8.0k .20 0.00 .60 .02 0.00 92 .02 .0g .23 h 1.k9 0.00 0.00 7.55 .1 .A 4 '.0g .57 .19 0.00 0.00 .jj .g ~ 2.68 2.66 - TOTAL = 4-165 545 244 8 0 1006 3 97 4.37 16.40 54.17 26.25 .80 0.00 100.00 .50' l.88 6.21 2.78 .09 0.00-11.45 MEY-XXX N R OF OCCURRENCES ' t. X X - R NT OCCuQWENCES THIS CLASS xXX P NT OCCURRENCES ALL CLASSES A-11

r-TABLE 2 (cont'd) h4 'uECEk O.C s0 0 Op{. WASH., AT 0 L H ~ T480E E0A ii/i9/d 0 32. GAMES AND MOOWE JOB NO: ti48s-00i-27 sECTod - WIND cPEED CATEG0piEstMETERs PER SECON01 MFaN wind -0 0-i.s I.5-3 0 3.0-s.0 s.0-7.5 7.s-i0.0 >i0.0 TOTAL .sPEE0 2.h3 .b3 .i6 i. 0.00 4 'l 2 41 i At 8 3 3 s.06 f.h .6i 0.00 6.89 .09-70 2 3 .0} 3g 2.29 2!1g .11 i i- .30 0.0g 3g g 3 .09 e.gs .63 .0 !s ?} t2: 't isi 6.i2 E

At

1j-2: 2

':8' 2b~ i$

82

-8:88 ?:13 6 0 6. 4.07 EsE 6 i .i4 40 .sl .30 .is 0.00 i.49 .}' !3 jj 8 O sE ..is .09 i.00 .09 0.00 2.i4-on

3I-}-

3: e! o o8 ie r 3... ssE

13

':n

l

83 8:88 fM

.A i.Il -7.'O 7'02 008 i6!62 'l 5 'ite 3it! .N o2

• dt t

.9i s.iG ss. .i2 .77 4.22 4.33 .09 10.*4 o: gg zy Ag

2I

47 2:[7

!:37 Sqj s.39 t-a: s.

83 -

is 2:i

' t:sw 3 7 5 9 9 s.i3

8'

g'

g6

2}

f.l f:j9d 1

9 al- -ii. s.i7 4 .8 o-

5{

t8

21 2:1.

'b6

8'

?:l$ s 34 iO3 i8d 6.23 tNw

81.

!{

.f'9

A3 i:j!

.J s!!A 3!L i.n ioM .ie -.0l .11 i.}t 3.93 2!!6 .89 io.4i o o* 2 86 18 5:!! 6.79 NN. i i lj 211 ' dj 'l 'hj -) 1 s.93 N 6

8s

13

':12 1:11 ':!0

fl 1:10-iCALM

-- i.99 i.99 68 68 CALM 3!s.t s't! 2483 d829 i4.42 2!85 iO3!!3 55 5 i.ul

t..

.i2 io. i 2i.2s a.as t r % 8S M atti is CLAs !!! upRCpNT OCCuenENC s_ ALL CLAssks EXA P A-12

TABLE:2.(cont'd) 1 1 E ~ ST ArilLI f f CL ASS:- pawull L AFFwl-TWIGA REACTOR li e tP" T E W !**' t "8 aliE#iS" o '. WWo' "o-

TAeLEGEkRATI II/19/E. 20.00.32.

DAMES AND M000E JOB NO3 11965-001-27 h0.0 >10.0 TOTAL EED OW' 'O.0-1 5 0 5 0 9 30 - 9 0 0 4e 1 80 - NNE .0.00. .69' ' 2.29 .69 0.00 0.00 3.66 .}5 0 00 0.00 .lg .g .19 0 0g ' ~ 0.00 .23 2.06 . 38 0.00 0.00 2.67 ':(h 4 3.31 ' ENE : .8:88

is.

!h 8:88 8:88

':2; l.k7 1. .6. .00 0.00 '0.00 2. ~ 10. 0. .16 .0 '0 00 0.00 .g 0.00 .69 99 .23 0.00 0.00 1.91 100g .j .g 0.08 3.g .03 0 00 0.00 0.00 1.d5 -1.68 .31 0.00 43 .g .05 0.00 0 00 0.0g. .2g

1l-8:88 8:88

':21 L 8:8) 2:ft 2! dji 3.j9 1896 2.y 5 0.00 0 00 i 0.00 8 a8 302

3. 3 2

L8 19 i7 2. -5s.

88 2:(1 l:!h 2:lj 8:88 8:28

'!:13 0.08 0.08 5.0 1.6 3.tl '5" 0.08- .53 oo8 oog S o* ':p$ ':!g.j o'8 3.12 8:88 8:88 ,s, 12 ,':.a .8:88 2 > 52 .3; 0.e8 0.08 008 .a + 1. .0 0.00 00g .] 0.0g- .lg-8:8l-

11

H

':30 8:88 8:88 ':11 0.00- .1 - 6."8 2.11 0.08 0.08 .!W '2 1;.b'. Ig .13 0.00 0 00 0.00 .0j .!6 2.14 0.08 .6p 0.00 0.00 . ;2, .0

i. ;

..i 0.0g 00g <0.0a 8:8:

N

': 1 2:!! 8:88 8:88 I:!! l ' CALM O' O CALM .8:81l 8 88 17kh 65$6 17fN ~ 0.00 0.00 0.00 100 00 0 00 2.63 9.71 2.58 0.00-0 00 14.93 Cbskks A-13

TABLE 2 (cont'd)! l i STAulLITV'Classa PaSQUILL F AFFRI-TRIGA REACTOR 2tle42"*kRAThD4'nis4H"^'^'t;%It#S S" - c W W ^*"- TA9LE.G 11/19/80 20.00.32. DAMES AND MOORE JOB NO 11985-001-27 . WINO WIND SPEED CATEGORIEstMETERS PER SECONO >10.0 TOTAL SPEED MEAN %ECTOR-0 0-1 5 1 5-3.0 3.0-5.0 5.0-7.5 7.5-10.0 NNE 8 11 10 0 0 0 29 2.31 . 49 .68 .62 0.00 0.00 0.00 1.79 9 H 'A 8 o8 a8 33 Ne 2.34 .43 1.05 -.49 0.00 0.00 0.00 1 93 ji 13 3 8 o8 a8 2%

.ENE

.:i: 2:34

N 8:88 8:88 8:88

82 2.u0

'E' .37 1.30 .06 0.00 0.00 0.00 1 73 6 21 1 0 0 0 28 2 22 .07 .2g .0g 0.00 0.0g 00g .g '.4 3 44 49 0.00 0.00 0.00 1.42 .08 .0g .09 0.00 0.00 0 00 1.k5 .k! .g .25 0.00 0.00 0.00 2.41 .05 .28 .I 0.00 0 00 0.00 .*4 SSE 3 11 3 0 0 0 27 2 68 .19 .68 .00 33 12 18 g.g0 0 00 0.g0 1.67 8 8 a8 is! S. 2.55 1.42 4.58 3.65 0.00 0.00 0.00 4.65 55. 12 u 83 8 8 8 'ti! 2.50 1.61~ 5.57 3.59 0.00 0.00-0.00 10.76 .g0 1 02 .66 0.00 0.00 0 00-1 98 3.lk afk 2.3$ 0.00 0.00' O.08 .}8 .1 k 13 g 43 -0.00 0.00 0.00 2.52 4.58 7.k2 1.f.3 0.00 0.00 0.00 13 . 84 0.00 0.00 0.00 2.53 U 1.46 ':j8 3:22- ':b0 8:88 8:88 8:28 t:1 56 8 0 0 0 102 2.21 3 A t:Nw 26 0 0 0 3 2 58 43 1.61 .08 .3 1.24 3.00 0.00 0.00 3.28 .23 0.0g 0.00 0.00 . 56 2.57 .2

1. 4 -

1. 1 0.00 0.00 0.00 3.46 .lg' .23 .{0 0.00 0.00 0.00 .64 .56 2.4h 1.5 0.00 0.00 0.00 4.98 2.54 1

t 3

oo8 oog 8k oo

i 2:28

':d 8:28 8:88 8:88 5$9 2 54 -- CALM-241 E41 CALM 14.90 14.90 2 74 ?.74 -TOTAL $28 723 366 0 0 0 1617 2 02 32.65 44.71 22.63 0.00 0.00 0.00 100.00 l 6.01 8 23 4.17 0.00 0.00 0.00 18.41 iEY - xxx Ni R OF OCCURRENCES xxx Pi NT 0CCURRENCES THIS CLASS xxx P! NT OCCURRENCES ALL CLASSES A-14 P-

+

T BLE 2 (cont'd)

-At,L W N05 AFFRI-TRIGA' REACTOR I Sils ifus's "sW!" i!n "eiE45"" Waf*^ " - c ~Ta9LEGENERATkb.11/19/80. 20.00.32.. 9AME5 AND M000E JOH N08 11985-001-27 . WIND SPEED CATEGOPIES(METERS PER SECOND) MEAN JefNO ..0.0-l.5' l.5-1 0 3.0-5.0 5.0-7.5 7.5-10.0 >10.0 TOTAL SPEED '5fCTOR -NNE 17 54 121 134 14 0 345 4.43 .19 .67 1.38 1.53 .16 0.00 3.93 NE - 22 75 174 145 26 0 442 4.41 .25 .85 1.98 1.65 .30 0 00 5.03 .ti .it .i!?! i!ik .l! 4 12 ' " ' ~ .05 E. ~20 85 105' 53 4 0 267 3.58 .23 .97 1.20 .60 .05 0.00 3.04 ESE 27 58 50 17 6 0 158 3.20 .31 .66' .57 .19 .07 0.00 1.80 2 hk .8 1. .0h 0.00 55' .32 i.!! 2!it .01 0.08 4?!!. 3 57' S 41' 230- '716 418 28 0 1633 4.16 47 2.62 8 15 4.76 .32 0.00 16.31 .553 37 179 451 231 39 4 941 4.16 42 2.04 5.13 2.63 44 .05 10.71 SS 66 184 178 99 18 6 551 3.60 .75 2.09. 2.03 1 13 .20 .07 6.27 ."1 i!?! i!S: .28 .i .i! 4 94 3 25 '5" e 40 100 122 78 17 4 361 3.85 .46 1 14 1.39 .89 .19 .05 4.11 WNW 1[. 46 140 135 38 7 377 4.96 .lJ .52 1.59 1.54 43 .08 4.29 'N3 14 42 196 306 133 64 715 5.97 .16 48 2.23 3.48 1.51 .50 3.37 NNJ 17 73 267 258 116 38 769 5.47 .19 .93 3.04 2.94 1.32 43 8.75 9 10 2kf 2fh8 .ko 6I88 .10 CALM 379 379 CALM 4.31-4.31 IkI) 3hS$ 2!f!f sib !!!) 10 00 9 2 ""212Si l ~b:fALNUM6pROh1883"kvATIONS-l41?%5 8'*t "8:88 8E?: s.. 085 8784 100.00 PCT. !!!'kR C2 SP T ~ A-15

APPENDIX B Annual Averaged Relative Concentrations (X/Q) m B-1

(xc 0

1. 4

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1 APPENDIX C ANALYSIS OF LOSS-OF-COOLANT ACCIDENT FOR AFRRI-TRIGA REACTOR C-1 ' ~ - - - '

P - APPENDIX C ANALYSIS OF LOSS-OF-COOLANT ACCIDENT - FOR AFRRI-TRIGA REACTOR The assumption is made that the water from the reactor tank is completely lost due to a' tank rupture at or below the reactor core level. Rupture of the fuel cladding or fuel melt would not occur even if only convective air cooling was available following loss of the water. - Within an hour, the reactor core would be moved behind the lead shield so that repair of the tank could be made without a radiation' exposure risk to workers and other personnel. If the water shield 'is lost, the possible exposure to AFRRI personnel or the general public would be due to direct or scattered gamma radiation from the exposed reactor. fuel inside the reactor tank. The following assumptions have been made for the irradiated fuel in the analysis of postulated gamma doses within the reactor room and outside the AFRRI facility: 1. ~ Power level = 1 MWt (steady state operation) I 2. Number of fuel elements = 85 3. Mass of each fuel element = 2.3 kg 2 4. Mass of uranium = 0.2 kg 2 5. Mass of zirconium and stainless steel = 2.1 kg 3 - 6. Distance from top of core to top of reactor tank = 505 cm 3 7. Distance from reactor room floor level to ceiling = $50 cm 3 8.- Reactor tank diameter = 366 cm 3 9. Ceiling thickness = 10 cm of ordinary concrete 3 l 10. Wall thickness = 25 cm of ordinary concrete 6 Curies " 11. Total core activity = 4 x 10 I7 5 12. Total gamma power = 1.75 x 10 Mev/sec ,6 3 13. Density of fuel element = 7.9 gm/cm 3 14. Volume of fuel elements = 2.5 x 10" cm. j i l C-2 l

l The gamma' dose rate on the end surface of a right _ cylinder is determined by the following equation.7-The reactor core is assumed to be a right e ylinder. D 5" Y l'- E IU *) 2~ s 2 9,K L - where: source gamma dose rate (R/hr) D =. 3

volume source strength (Mev/sec-cm )

Sv = attenuation coefficient for source (cm-I) p, = 2 flux to dose rate conversion factor (Mev/cm -sec per R/hr) K. = thickness of source (cm) t = exponential function. E = 2 12 The source strength (Sv) for the reactor core is determined to be 7 x 10 3 Mev/sec-cm. The average photon energy is assumed to be 1 Mev. Therefore, the ' attenuation coefficient for the core (p,) is 0.46 cm-I (coefficient of iron) and the 3 2 flux -to dose conversion factor (K) is - 5.2 x 10 Mev/cm -sec per R/hr..The thickness of the source is 87 cm. The calculated gamma dose rate (D) at the top 7 l surf ace (1 cm) of the core is 1.5 x 10 R/hr. The measured gamma dose rate at the 3 R/hr.2 Since surface of TRIGA fuel operated at 250 Kwt after 1 hour decay is 10 . the dose reduction factor for 1 hour decay is 4, and the power level is a factor of 4 lower, the expected surface dose rate'of an AFRRI fuel element would be ~ 1.6 x 10" R/hr. For an 85 fuel element core, the maximum expected gamma dose 6 I rate ~ above the core from the measured dose rate would be 1.4 x 10 R/hr, or a factor of more than 10 less than the calculated value. The actual operating time for the fuel element which was measured would be another factor in reducing the difference between the calculated and measured gamma dose rate. The calculated value assumes an equilibrium or steady state condition. For example, the gamma decay power level for a 1-day operation is only a factor of 1.5 less than the equilibrium value." To determine the gamma dose rate to an individual in the reactor room, the contribution from the air scatter and the ceiling scatter _must be determined. From Reference 8,.the reduction factors for a narrow beam at various distances .for 1 Mev photons are 1.4 x 10-8 at 10 feet,3.8 x 10-9 at 30 feet, and 1.5 x 10-9 at C-3

60 feet. The other factor to.be determined is the solid angle of the escaping cone ' from the reactor core (a point source). The solid angle for the reactor tank diameter opening is approximately 0.45 (40 solid angle). The following equation is used to determine air-scattered gamma dose rate within the reactor room: (D XRFXA) D =- A ~where: D = - air-scattered gamma dose rate (R/hr) g source gamma dose rate at I cm (R/hr) D = RF ' reduction factor for air scatter = A- = . solid angle of escaping beam (steradian). 7 With a source gamma dose rate of 1.5 x 10 R/hr, the air scatter gamma dose rates at the reactor room. floor level are calculated to be 3 x 10-I R/hr at 10 feet, 8 x 10-2 R/hr-at 30 feet, and 3.2 x 10-2 R/hr at 60 feet from the reactor tank. ~ The scattering from the ceiling can be determined b'y using the method given in Reference 8 for ~ an extended slab; The following equation can be used to determine the contribution from the ceiling scatter of the escaping beam (reactor core): 1 D (D )(r -2Xr;-2XRF) = 3 3 s where: D ' slab scattered gamma dose rate (R/hr) = s D source gamma dose rate at I cm (R/hr) = ~2 .= inverse square factor for~ incidence beam r; h r," -inverse square factor for scattered beam = RF2= reflection factor for extended slab. The inverse square factor for the incidence beam is 9 x 10 (1055 cm) and for the scattered beam is 3.3 x 10-6 (550 cm). The reflection factor at normal incidence is 7.8 x 10~". With a source gamma dose rate of 1.5 x 10 R/hr, the 7 ceiling scattered gamma dose rate is negligible (3.5 x 10-7 R/hr) and would not i contribute to the gamma dose rate to an individual in the reactor room or outside i the AFRRI facility. I l C-4 b

m = s The gamm2 dose r;ta on the roof cf the AFRRI ft.cility cver the rsector room lcan'be determined by the following equation:8 D r[2,-ut ' -D !=- d 'where: direct gamma dose rate (R/hr). f LD ; = d D ?= source gamma dose rate (R/hr) 4 i.r[2.1,.- Linverse square factor for incidence beam .p-

= ~

. attenuation coefficient for shield (cm-I) ' thickness ' f shield (cm). 't -=- o I For 'the -ordinary concrete ceiling, the thickness is 10 cm and the attenuation Ecoefficient. for 'l Mev_ photons is.0.145 cm-I. The inverse square factor for the -incidence beam is' 9 x 10-7 o With 'a source.. gamma dose rate of 1.5 x 10 R/hr, the 7 gamma dose rate on the roof above the reactor core is 3.2 R/hr. The air scatter to .the ground;from.this source would be negligible. This gamma dose rate of 3.2 R/hr 1should-be reduced by 'at-.least the inverse square factor of the scattered beam, ..which would'be 3.3'x 10-6 (550 cm). To determine th'e maximum gamma dose rate to an individual outside. the ~ ' AFRRI facility, the calculated air-scatter gamma dose rate at.10- feet (nearest -outside wall of reactor. room) should be reduced by the shielding of the concrete wall.~ The' calculated gamma dose rate at the outside surface of the nearest wall in .the reactor room would be 8 mR/hr. Beyond 20 meters from.the nearest wall, the

gamm'a dose rate would be less than 1 mR/hr.

s. - p

- All of the' calculated gamma dose rates assume a steady state condition of 1 MW operation (at least 1 year operation) which probably results in a source term that is conservative by a factor of 10. These calculated gamma dose rates would s he reduced by;a factor of 4 within I hour and by a factor of 11 within I day.5,6 Therefore,- the integrated gamma dose for 1 hour would be 0.35 of the calculated " hourly'. dose and the-integrated gamma dose for 1 day would be only,3 times the g . calculated hourly dose.- 1 'I C-5

r REFERENCES 1. Amendment #17 to the Technical Specifications, Facility License No. R-84 for AFRRI-TRIGA Mark-F Reactor, Docket No. 50-170. 1 2. Personal Communications with AFRRI staff (Reactor Fuel Records). 3. AFRRI, Final Safeguard Report: Revised Edition, Docket No. 50-170, General Dynamics and Holmes & Narver, Inc. (March 1962). 4. Biomeke, 3. O., and Todd, Mary F., Uranium-235 Fission Product Production as a Function of Thermal Neutron Flux, Irradiation Time, and Decay time, Part 1, Volume 2, ORNL-2127 (December 1958). 5. Knabe,2$$,E., and Putnam, G. E., The Activity of the Fission Products of U APEX-448 (January 1959). 6. Smith, M. R., The Activity of the Fission Products of U235(Program 408), XDC-60-1-137 (December 10, 1959). 7. Foderaro, A., and Obenshain, F., Fluxes from Regular Geometric Sources, WAPD-TN-508 (June 1955). 8. Wende, C. W. J., The Computation of Radiat.on Hazards, AECD-3661, January 11,1944. C-6 i

TABLE 1 GASEOUS FISSION PRODUCT INVENTORY I MW STEADY STATE OPERATION Saturati n 7 Gamma Sogrce l Half-Life *7 Fission Yield Activity Strength -Isotope -(TM) (%) (Ci) (Mev/Sec) Krypton (Kr) .83 1.86 h 0.52 4.5 E+3 7.0 E+12 85 m - 4.40 h 1.30 1.1 E+4 1.9 E+14 85 10.76 y 0.271 2.4 E+3 1.8 E+11 87' 76.0 m 2.53 2.1 E+4 1.1 E+15 88 2.80 h 3.56 3.1 E+4 2.0 E+15 89 3.20 n1 4.59 4.0 E+4 3.4 E+15 TOTAL 12.77 1.1 E+5 6.7 E+15 Xenon (Xe) - 131 m 11.9 d 0.017 1.5 E+2 9.0 E+11 133 m 2.30 d 0.16 1.4 E+3 1.2 E+12 133 5.27 d 6.69 5.8 E+4 1.7 E+14 1 135 m 15.6 m 1.8 1.6 E+4 3.1 E+14 - 135 9.2 h 6.3 5.5 E+4 5.0 E+14 137 3.8 m 6.0 5.2 E+4 2.9 E+14 . 138 14.0 m 5.9 5.1 E+4 2.4 E+15 TOTAL 26.87 2.3 E+5 3.7 E+15 lodine (1) 131 8.05 d 2.91 2.5 E+4 3.4 E+14 132 2.28 h 4.26 3.7 E+4 i 3.3 E+15 133 21.0 h 6.69 5.8 E+4 1.0 E+15 134 52.0 m 7.80 6.7 E+4 4.9 E+15 135 6.7 h 6.17 5.4 E+4 3.6 E+15 TOTAL 27.83 2.4 E+5 1.3 E+16 G ASEOUS TOTAL 67.47 5.8 E+5 2.34 E+16

• y = year, d = day, h = hour, m = minute.

C-7

p,- fi . TABLE 2 L GASEOUS FISSION PRODUCT INVENTORY - FOR 40 MW-SEC PULSE OPERATION s 7 7 'One-Minute igecay Gamma Sogrce l - Half-Life *7 ~ Fission Yield Activity Strength Isotope' -(Th) .(%) (Ci) 1(Mev/Sec) - .- Krypton'(Kr) 2 85 m-- 4.4-h 1.3 2.9

• 5.0 E+10.

87 76.0 m ' -2.53 87.0 - 4.4 E+12 ^ 88: 2.8 h - 3.56 91.0 6.0 E+12 .89. -3.2 m 4.59 3600.0 3.1 E+14 ? TOTAL 11.98- -3.78 E+3 3.2 E+14 Xenon (Xe)' 135 m 15.6 m 1.8-56.0 1.1 E+12 135-9.2 h' 6.3 1.4-1.4 E+10 137 3.8 m 6.0 4600.0 2.6 E+13 138 14.0' m "5.9 1300.0 6.0 E+13 TOTAL 20.0 5.96 E+3 8.7 E+13 Iodine (I) 132 2.28 h ~ 4.26 - 5.7 4.7 E+11 133- .21.0 h~ 6.69-4.3 7.4 E+10 .134 '- 52.0 m 7.80 165.0 1.2 E+13 135 6.7 h 6.17 '57.0 3.7 E+12 TOTAL 24.92 2.3 E+2 1.6 E+13 ' GASEOUS TOTAL- .56.9 1.0 E+4 4.2'E+14 'l

• d a day, h = hour, m = minute.

-C-8 .. + - f

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a.. % ss..

TABLE 3 +- DIFFUSION FACTOR AND - FINITE CLOUD CORRECTION FACTOR USED FOR DOSE CALCULATIONS .: Distance : y/Q' '(m) - (sec/m ) - F

• 10 3

.25~ 1.0 E-1 1.5 E-2 50 2.7 E-2 2.7 E-2 '75 -1.2 E-2 4.2 E-2 100 7.5 E-3 5.5 E-2 150 3.7 E 8.0 E-2 200 2.2 E-3. 1.0 E-1

• F represents a ratio of gamma doses from a finite size cloud to that calculated

- from an infinite cloud with same center-line concentration for Type F diffusion. I 4 L C-9 ~..

[-i TABLE 4 l l ' GASEOUS FISSION PRODUCT INVENTORY [ FOR 1 MW STEADY STATE OPERATION l WITH 2-WEEK DECAY r Gamma Source Thyroid Source L Activity Strength Strength ' isotope (Cl) (Mev/sec) (rad) Krypton (Kr) 85 2.4 E+3 ~ 1.8 E+11 Xenon (Xe). 131 m' 67.0 3.9 E+11 133 m 21.0-1.8 E+10 .133-9.2 - E+3 2.9 E+13 TOTAL 9.3 E+3 2.9 E+13 lodine (1). '131 7.5 E+3 1.0 E+14 1.1 E+10 132 5.2.E+2 .4.'5 E+13 2.8 E+7 ) . TOTAL 8.0 E+3 1.5 E+14 1.1 E+10 GASEOUS TOTAL 2.0 E+4 1.8 E+14 1.1 E+10 l C-10 o _ e

TABLE 5 CALCtlLATED DOSES TO INDIVIDUALS STANDING DOWNWIND OF AFRRI FACILITY FOLLOWING FUEL ELEMENT DROP ACCIDENT Whole Body Thyroid Distance Dose Dose (m)' (mrads) (mrads) 25 0.0004 45 -50 12 6 75 i 3 !~ 100 2 l 150 t 200 1 1 f l

• Insignificant.

1 ('. ~ C-11

p. u TABLE 6 GASEOUS FISSION PRODUCT INVENTORY RELEASED TO ATMOSPHERE FOLLOWING FUEL ELEMENT CLADDING FAILURE ACCIDENT Gamma Source Thyroid Source - Activity Strength Strength Iso' tope (Ci) (Mev/sec) (rad) Krypton 1.4 8.6 E+10 Xenon - 2.9 4.5 E+10 n' O.06 3.1 E+8 1.7 E+3 TOTAL 4.4 1.3 E+11-1.7 E+3 i 4 + 4 i C-12

TABLE 7 1 CALCULATF.D DOSES TO INDIVIDUALS STANDING DOWNWIND OF AFRRI FACILITY FOLLOWING FUEL ELEMENT CLADDING FAILURE ACCIDENT i Whole Body Thyroid Distance Dose Dose (m) (mrads) (mrads) 25 1.3 57 50 0.6 16 75 0.5 7 100 0.3 5 i 150 0.2 3 200 0.15 2 l l C-13 _}}