ML20009B023

SAR for Armed Forces Radiobiological Research Inst Triga Reactor Facility,License R-84

ML20009B023
Person / Time
Site:	Armed Forces Radiobiology Research Institute
Issue date:	05/12/1981
From:	DAMES & MOORE
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References
NUDOCS 8107140676
Download: ML20009B023 (175)

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RFRRi SAFETY ANALYSIS REPORT FOR THE AFRRI-TRIGA REACTOR l 3 FACILITY LICENSE R-84 l

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PREPARED BY DAMES AND MOORE, ENGINEER CONSULTANTS FOR DEFENSE NUCLEAR AGENCY g ARMED FORCES RADIOBIOLOGY RESEARCH INSTITUTE BETHESDA, MARYLAND 20014 D ADOKObO O K PDP

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AFRRI REACTOR FACILITY SAFETY ANALYSIS REPORT Submitted to:

The Armed Forces Radiobiology Research Institute National Naval Medical Center Bethesda, Maryland

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Submitted by:

DAMES & MOORE 7101 Wisconsin Avenue Suite 700 Bethesda, Maryland 20014 May 12,1981 0

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, The AFRRI-TRIGA Mark-F Reactor Technical Specifications identify the i minimum operable systems for reactor operation; reactor operations will conform to the Technical Specifications. All additional design and operational information

' contained in this Safety Analysis R port (SAR)is for information purposes only.

1 The dimensions, measurements, or other numerical values in this report may i differ from measured values as a result of normal construction and manufacturing tolerances, or from variations in the normal accuracy of instrumentation.

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O b CONTENTS

1.0 INTRODUCTION

AND

SUMMARY

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1.1 INTRODUCTION

1-1 4

1.2 AFRR1 1-1 1.3 AFRRI-TRIGA REACTOR l-1 2.0 SITE 2-1 2.1 LOCATION 2-1 2.2 POPULATION 2-1 2.3 METEOROLOGY 2-12 2.3.1 Regional Climatology 2-12 2.3.2 Long-Term Diffusion Estimates 2-13

2.3.3 Short-Term Diffusion Estimates 2-15 2.3.4 Severe Weather 2-15 2.4 GEOLOGY 2-17 2.5 HYDROLOGY 2-17

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2.6 REFERENCES

2-19 3.0 FACILITY 3-1 3.1 GENERAL 3-1 3.2 REACTOR BUILDING 3-6 3.2.1 Reactor Building Construction 3-6 3.2.2 Reactor Building Ventilation 3-7 l

3.2.3 Reactor Room 3-7 3.2.4 Reactor Room Confinement 3-10 3.3 REACTOR WATER PURIFICATION AND COOLANT SYSTEMS 3-11 3.3.1 Primary Cooling System 3-11 3.3.2 Secondary Cooling System 3-12 3.3.3 Primary Water Purification System 3-12 3.3.4 Primary Water Makeup System 3-13 3.3.5 N 16 Diffuser System 3-13 3.3.6 Water Systems Instrumentation 3-15 O

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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 3-18 3.4.3 Solid Radioactive Wastes 3-24 3.5 SPECIAL NUCLEAR MATERIAL 3 3.6 RADIATION MONITORING SYSTEMS 3-25 3.6.1 Remote A.ea Monitors 3-26 3.6.2 Continu os Air Monitors 3-26 3.6.3 Stack Monitoring Systems 3-31 3.6.3.1 Stack Flow Monitoring System 3-31 3.6.1.2 Stack Particulate Monitoring System 3-31 3.6.3.3 Stack Gas Monitoring System 3-34

() 3.6.4 3.6.5 Perimeter Monitoring Criticality Monitors 3-34 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 Meteorological 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 4-1 4.3 LEAD SHIELDING DOORS 4-4 4.4 SUPPORT STRUCTURES 4-5 4.5 CORE 4-6 4.6 GRID PLATE 4-8 4.7 NEUTRON STARTUP SOURUE AND HOLDER 4-8

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i CONTENTS (cont'd) 4.8 IN-CORE EXPERIMENT TUBE (CET) 4-10 4.9 FUEL ELEMENTS 4-12 4.10 REACTOR CONTROL COMPONENTS 4-15 4.10.1 Standard Control Rods and Guide Tubes 4-15 4.10.2 Standard Control Rod Drives 4-15 4.10.3 Transient Control Rod 4-19 4.10.4 Transient Control Rod Drive 4-19 4.11 REACTOR INSTRUMENTATION 4-21 4.11.1 Multirange Linear Channel 4-22 4.11.2 Wide-Range Log Channel 4-22 4.11.3 High Flux Safety Channels One and Two 4-23 4.11.4 Fuct Temperature Safety Channels 4-24 4.12 ROD WITHDRAWAL PREVENT INTERLOCKS 4-24 4.13 FACILITY INTERLOCK SYSTEM 4-25 O 4.14 SCa^M tOGiC CiRCUiTRv 4-27 4-29 4.15 FUEL HANDLING EQUIPMENT 4.16 REACTOR PARAMETERS 4-30 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 5.0 EXPERIMENTAL FACILITIES 5-1

5.1 GENERAL 5-1 5.2 EXPOSURE ROOMS #1 and #2 5-1 l

5.2.1 Exposure Room Construction 5-1 5.2.2 Exmsure Room Access 5-4 5.2.3 Exposure Room Monitoring 5-5 5.2.'- Patch Panel System 5-8 5.3 PNEUMATIC TRANSFER SYSTEM 5-8 5.4 POOL IRRADIATION 5-14 l

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J CONTENTS (cont'd) 5.5 ARGON ACTIVATION 3-14 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

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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 Gases in the Reactor Coolant 6-4 6.2.5 Experi.nents 6-8 6.3 NONROUTINE OPERATIONAL RISKS 6-9 O 6.3.1 Radioactiva Contamination of Reactor U Shielding hcer 6-9 6.3.2 Fuel Element Cladding Failure 6-10 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 6-15 6.3.4.1 Fuel Element Cladding Failure Accident 6-17

6.4 REFERENCES

6-19 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 7-3 7.4 SCIENTIFIC SUPPORT DEPARTMENT 7-5 7.4.1 The Reactor Branch 7- 6 7.5 RADIATION SAFETY DEPARTMENT 7-6 7.6 NATIONAL NAVAL MEDICAL CENTER 7-8 iv

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

O CONTENTS (cont'd) 4 i APPENDIX A: JOINT WIND FREQUENCY DISTRIBUTION FOR 1960-64 AND 1964 INCLUSIVE A-1 APPENDIX B: ANNUAL AVERAGED RELATIVE CONCENTRATIONS ()(/Q) B-1 APPENDIX C: ANALYSIS OF LOSS-OF-COOLANT ACCIDENT FOR AFRRI-TRIGA REACTOR C-1

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LIST OF TABLES Table Page 2-1 National Naval Medical Center Building Index 2-7 2-2 National Naval Medical Center Population Estimates 2-10 3-1 Reactor Remote Area Monitors 3-27 3-2 Reactor Room Continuous Air Monitors 3-32 _

3-3 Stack Monitoring Systems 3-35 5-1 Prep Area Remote Area Monitors 5-9

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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 AFRR1 Complex 2-4 2-4 Aerial View of the AFRR1 Complex Looking South 2-5 2-3 Site Plan--Annuli and Sectors with Respect to the AFRR1 Stack 2-6 3-1 AFRRI Floor Plan 1st Level . 3-2 3-2 AFRRI Floor Plan 2nd Level 3-3 3-3 AFRRI Floor Plan 3rd Level 3-4

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m) 3-4 AFRRI Floor Plan 4th Level 3-5 3-5 Reactor Pullding Ventilation System 3-8 3-6 Water Purification and Ccoling 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 3-9 Liquid Radioactive Waste Drains Located on the Third Level 3-22 3-10 Liquid Radioactive Waste Drains Located on the Fourth Level 3-23 3-11 AFRRI Radiation Monitors Associated with AFRRI-TRIGA Reactor--Ist Level 3-28

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

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LIST OF FIGURES (cont'd)

Figurs 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 l

4-2 AFRRI-ThtGA Core Support St ucture 4-7 4-3 Upper and Lower Grid Plates '4-9 4-4 Neutron Source and Guide Tube 4-11 4-5 Standard Fuel Element 4-13 4-6 Instrumen'ed 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 -

1 5-6 Pneumatic Transfer System 5-12 7-1 AFRRI Organizational Chart 7-2 7-2 Reactor Branch Chain of Command 7-7 I

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1.0 INTRODUCTION

AND

SUMMARY

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1.1 INTRODUCTION

This Safety Analysis Report is an update of the Fina! Safeguards Report (dated March 1962, with subsequent revisiors). This report is submitted as part of the license renewal process for Facility Li ense No. R-84 f or the Armed Forces Radiobiology Research Institute (AFRRI) TRIGA Mark-F rtactor in accordance with 10 CFR Part 50.

1.2 AFRRI

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The Armed Forces Radiobiology Research Institute is located on the grounds of the National Naval Medical Center, Bethesda, Maryland. AFRRI is a tri-service military organization under the Defense Nuclear Agency. The mission of AFRRIis to conduct scientific research in the field of radiobiology and related matters essential to the medical support of the Department of Defense. To carry out this mission. AFRR1 supports and utilizes the TRIGA Mark-F reactor, research labora-tories, a 50-MeV electron linear accelerator, a hot cell, office space, an animal research facility, a 500-KCi Cobalt-60 facility, and a Theratron unit, all of which O ere ce#1 ime8 1 18e ^ ea ni ce- 91ex.

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) or pulse mode with a step reactivity insertion of up to 2.3 %

Ak/k (technical specifications limits) and utilizes standard-design General Atomic fuel elements. The AFRRI-TRIGA Mark-F reactor 9 as the unique capability of a horizontally movable core. The movable core allows utilization of a variety of .

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

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over 15 years. The first TRIG A 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 staf f or the general public.

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O V 2.0 SITE 2.1 LOCATION The site of '.ne Armed Forces Radiobiology Research Institute (AFRRI),

which houses the TRIGA Mark-F reactor, is located on the grounds of the National Naval Medical Center (.NNMC), Bethesda, Maryland. The coordinates of AFRRI are 39 00'10" N latitude and 77 05'21" W longitude. The National Naval Medical Center is approximately 3 miles north of the Washington, D.C.-Maryland line. The AFRRI site is located on a moderate slope, declining northward toward a narrow

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creek valley. The terrain surrounding AFRR1 is rolling with elevations above sea level ranging from 230 to 320 feet.

The location of AFRRI within the National Naval Medical Center is shown on the Site Plan, Figure 2-1. The uses of each building at NNMC are given on Table 2-1. Aerial photographs of AFRF.I are also provided in Figures 2-2,2-3, and 2-4. The location of buildings at the NNMC complex, with respect to the AFRRI exhaust stack, is shown on Figure 2-5.

2.2 POPULATION The National Naval Medical Center has a peak daytime population of approximately 6,095 persons which includes an estimated 1,600 visitors. Tne 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 1,600 200-400 3,870 400-700 625 Total 6,095 l

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TABLE 2-1 N ATIONAL NAVAL MEDICAL CENTER BUILDING INDEX Building Number Building Use 1 Hospital, Offices 2 Administration 3 Hospital, Of fices 4 Hospital, Of fices 5 Hospital, Of fices -

6 Hospital, Of fices 7 Hospital, Offices S 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 n 17 Research (NMRI)*

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19 20 Fire Station 21 Animal Facilities (NMRI) 22 Flammable Storage 23 Recreation Building 25 Greenhouse 26 Greenhouse 27 Oxygen Storage 23 Research (NMRI) 29 Storage (NMRI) 30 Flagpole 34 Surgeon Gene al's Quarters 35 Of ficers' Quarters 36 Of ficers' Quarters i 37 Of ficers' Quarters l

38 Of ficers' Quarters of ficers' Quarters 39 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)

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TABLE 2-1 (cont'd)

Building Number Building Use 50 Enlisted Males' Quarters 51 Gasoline Service Station 52 Guest Lodge 53 Research (NMRI) 54 Staff Parking Garage, Offices 55 Outpatient Parking Garage 56 Bowling Alley 57 Navy Exchange _

70 Uniformed Services University of the Health Sciences 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, Offices 125 Enlisted Females' Quarters 139 Machine Shop (NNMC), Research (NMRI)

,, 141 School

( ) 142 Research (NMRI), Research and Development Command Of fices 143 Storage 144 Eye Clinic '

146 Dog Run Area (NMRI) 147 Storage 148 Storage 149 Storage 150 Demolished 151 Demolished 152 Storage 153 Storage 154 Storage 155 Automotive Maintenance 156 Demolished 158 Storage (NMRI) 159 'I 'nnis Courts 160 Tennis Courts 161 Baseball Field 165 Lunchroom (NMRI) 174 Demolished 176 Storage (NMRI) - Structure Partially Demolished 177 Underground Tunnel t_j 2-8

TABLE 2-1 (cont'd)

Building Number Building Use ISO Vehicle Bridge 181 Pedestrian Bridge 184 Reactor Cooling Towers (AFRRI) 138 Underground Shelter 189 Vehicle Bridge 190 Pedestrian Bridge 200 Helicopter Pad 202 Switching Station _

203 Storage (AFRRI) 204 Pumphouse (AFRRI) 205 Warm Water Waste Tanks (AFRRI) 207 Gas Tanks 208 Gas Tanks 218 Demolished 219 Dog Shelter (NMRI) 222 Switching Station (AFRRI) 223 Animal Feed Storage (AFRRI) 224 Flammable Storage 225' Storage

(' 239 Storage 240 Urology Clinic 241 Offices 242 Special Services 247 Bus Stop Shelter 248 Bus Stop Shelter 251 Environmental Health Lab i 252 Cooling Tower 253 Oxygen Storage i 255 Meter House 256 Storage

. 257 Recreation Pavilion 258 Recreation Pavilion 259 Rest Rooms i

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I O' TABLE 2-2 NATIONAL NAVit MEDICAL CENTER POPULATION ESTIMATES Building Population C. 200 yards from AFRRI stack Armed Forces Radiobiology Research 240 Institute Complex (Bldgs. 42 47)

Naval Medical Research Institute Complex - -

(Bldgs.17,18, 21,119,139,146,165) 300 Uniformed Services University for the Health Sciencet (Bldgs. 7',-73) 500 i 15 15 16 25 23 510 39 10 i

I TOTAL 1,600.

200-400 yards from AFRRI stack National Naval Medical Center .

Hospital Core (Bldgs. 1-10, 240) 2,80G 11 110 13 25 14 190 20 5 28 20 50 25 52 25 53 100 54 85 56 20 l

125 175 141 130 142 50 -

144 95 155 5 251 10 TOTAL 3,870 2-10

TABLE 2-2 (cont'd)

Building Population 400-700 yards from AFRRI stack Officer Quarters' complex (Bldgs. 34-38) 30 12 85 51 15 -

57 310 122 95 123 85 241 5 TOTAL 625 O

I 2-11

2.3 METEOROLOGY 2.3.1 Regional Climatology The Armed Forces Radiobiology Research Institute lies at the western edge of the Middle Atlantic coastal plain, approximately 30 miles east of the Blue Ridge Mountains 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 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 i ir of Canadian origin is the predominant air mass throughout the winter. However, the Appa-lachian Mountains to the west, which act as a partial barrier to delay the advance of the cold air, and the relatively warm Atlantic Ocean to the east help to moderate the winter temperatures in the Washington area.

The coldest weath2r occurs in January, when the mean temperature is 35.6 F and the mean minimum daily umperature is 27.7 F. The lowest temperature ever recorded in the metrcpolitan a rea 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 Washington area, the semipermanent Atlantic High moves northward and eastward and dominates the circWation of air over the Eastern United States. Consequently, southerly winds prevail for much of the summer and transfer moist, tropical air from 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 highest temperature ever recorded in the metropolitan area was 106 F, v.hich occurred in July 1930. The temperature reaches or exceeds 90 F on an average of 37 days each year.I Precipitation is spread out evenly throughout the year, averaging 38.9 sches of water equivalent. Precipitation in the summer mainly results from thunder-storms, and in the 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 2-12

m

( )

V Washington area, the maximum precipitation to occur in 1 month was 17.45 inches during September 1934, and in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> was 7.31 inches during August 1928. Sno.v, 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 '%. 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 snowfails 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.1 -

Relative humidity normally averages 64 percent annually, and fog (visibility less than E mile) occurs on an average of 13 days each year.

joint wind frequency distributions presented in Appendix A were prepared using hourly National Weather Service observations 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 wir.d 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 Pasquill type D stability, which represents neutral conditions.

2.3.2 Long-Term Diffusion Estimates National Airport wind data from the calendar yea. of 1964 (Table 2, Appenoix A) were used to calculate annual average relative concentrations (Table 1, Appendix B) using procedures outlined in Regulatory Guide 1.111. '"

These relative concentrations are required to 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 O- complex are located in the same general vicinity and it would be expected that 2-13

4 O

V 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 n.shttime observations. Since reactor operations at AFRRI primarily occur during daytime luurs, 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 2.032 ,I d (X/Q)D = NXU;I z:,

=1 where p E z.

• (Uz

+0.5H7,pi2 y . 1 except when E >

z, 6 z., in which case 1 1 E

• z.

i "z.'

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)E' 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; p

d N is the total number of hours of valid hourly data (all wind directions);

i l

i 2-14 5

X is the distance in meters downwind of the source; U is the measured windspeed in meters per second during hour, i; I

7 is the vertical plume spread with the volumetric correction 8

factor for a release within the building wake cavity, at a distance, X, for the stability class which occur' 'd during hour, i; 0

7 is the vertical standard deviation of the materials in the 8

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 AFRR1 stack and is 1.6 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 AFRRI

/ 1 (j complex.

2.3.3 Short-Term Ci fusion 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 vorestricted area. The 5 percent overall site yfQ 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 7

Severe weather in the Washington area is mainly associated with thunder-

> storms, tornadoes, winter ;,owstorms, and hurricanes.

2-15

Thunderstorms, which occur on an average of 29 days per year, of ten 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 reports of hail 3/4 of an inch in diameter or larger within the 1-degree latitude-Icagitude squate containing AFRRI during the period 1955 to 1967.0 Three rather destructive tornadoes have been recorded in the Washington 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 repcrted 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 AFRRI.6,7 Tropical disturbances, during their northward passage, occasionally influence

~

(' Washington's weather, mainly with high winds and heavy rainfall; extensive damage U 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 disturbances, only three made landfall at the Middle Atlantic coast.8 With the passage of Hurricane Hazel on October 15, 1954, the peak wind gust reached 98 mph, but or.ly 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 58 mph. During June ?.1-22,1972, 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.I The " fastest mile" wind speed recorded at National Airport was 73 mph from the southeast at a ll3-foot elevation on October 15, 1954 during the passage of Hurricane Hazel.I From 1955 to 1567 there were 29 reports of windstorms with wind speeds of 50 knots (38 mph) or greater within the 1-degree latitute-longitude square containing AFRRI.6 g

i u.1 2-16

4 (3

V 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 AFRR1 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 mottled light gray _

,and brown or dark green appearance and extends to a depth of 35 to 4.1 feet.

Unweathered bedrock varies from lenses and pods in the saprolite 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 Navdocks 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 nearest AFRRIis approximately 19 miles away and extends from 30 07' N latitude, 77 26' W longitude in Montgomery County to 39 20' N latitude, 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 nas 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 P)

( stream. The flood p!ain encroaches on no buildings at NNMC, but does cover

! portions of two parking lots.U 2-17

O O 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 cMnings 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 ta51e near the AFRRI reactor site is a subdued replica of the surface topography. The derth of the water table measured by test drillings varied O <

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(3 y/ 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 search.

An updated Physical Security Plan for the AFRRI-TRIGA Reactor Facility, 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 associated equipment. As outlined in 10 CFR

~

Part 2.790, the updated security plan contains classified information--DOD Instruction $210.67 (Security Classification For Special Nuclear Material Informa-tion)--and is therefore deemed exempt from public disclosure.

3.2 REACTOR BUILDING The reactor and its associated 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 leaching 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 t

The lines of contact between the poured light concrete roof and the reinforced concrett. walls are further 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.

1 3.2.2 Reactor Building Ventilation The reactor building ventilation system, shown schematically in Figure 3-5, is independent of the ventilation systems servicing the rest of the AFRR1 complex. _

Thus, only the reactor building ventilation system is significant in the discussion of airborne radiation resulting from reactor operations.

Air enters the reactor building primarily through two air supply fans, AC-1

(#15,000 cfm) and MV-1 (#8,500 cfm). The air then circulates through the reactor building as shown on Figure 3-5. The reactor building ventilation system utilizes very slight negative differential pressures so that the air flow within the building will always be from clean areas (i.e., offices, etc.) into potentially contaminated areas (i.e., reactor exposure rooms, hot cell, etc.). Air exhausted by the reactor ventilation system to the AFRRI exhaust stack is passed through a roughing filter i and an absolute filter. These filters remove particulate matter greater than 0.3 microns in diameter, and are maintained at a minimum efficiency of 99.9 percent.

The differential pressure across each filter is measured to determine when the filters should be changed according to manufacturer's specifications. The air is then discharged through air exhaust fan EF-1 (#12,000 cfm) to the AFRRI stack.

Failure of air exhaust fan EF-1 actuates an audible and visual alarm in the reactor control room.

3.2.3 Reactor Room The reactor room (Room 3161) contains approximately 1,800 square feet of floor space and has a ceiling height of approximately 18 feet and contains the open surface of the T'RIGA Mark-F reactor pool. Adjacent to the reactor room is the reactor administration / control area. The reactor administration / control area contains the reactor control' room (Room 3160), five offices (Rooms 3155 through 3159), and Hall 3106. The only access to these rooms is from Hall 3106 in the reactor administration / control area. Large, sealed windows provide visual observa-

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Et tCT esc AL -O- ATuotPMie:C stLif F SYSTEu twommaLLT CLolf Db 4+64444*e ad e ST lT E M  % P0llf tvt CLOlauG DamPE eltase OPEN SPeteeG CLOSEDs 4 af.0$Pues.C Rever vaLvr <=0E=att, Ct0$e0 FIGURE 3-5 REACTOR BUILDING VENTILATION SYSTEM

O tion of operations within the reactor room from the reactor control room and V

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.

4 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 AFRRI front desk to alert the Securityj#a:ch 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.

The reactor room rcof hatch is sealed with a compressible rubber gasket and is secured from within during reactor operation. The reactor room floor hatch is 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.

The double doors from the corridor beside the third floor equipment room to Hall 3106 in the reactor administration / control area serve as the nt rmal entrance to the controlled access area and are sealed with a compres=!ble rubber gasket.

These doors are not lef t open during reactor operation and are automatically locked with a card-key system. To a degree, these doors serve as backup 3-9

p protection in confining the reactor room air in the event that the other double V 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 momentarily 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 appro.cimately 1,800 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 N reactor room is made of ordinary concrete poured over a corrugated steel form (D 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 penetrations in the walls and floor for conduits and ' piping are also sealed.

The reactor building ventilation system contains three air supply dampers (S1, S2, and S3) and one exhaust damper (EI) in the reactor room and the reactor equipment room. These fail-safe dampers are pneumatically operated and require pressurized air to remain open. When the pressurized air, provided by one of the pressurized air supply systems, is removed, the dampers automatically close. The pressurized air is supplied to each damper through a solenoid valve.

A l

l 3-10 1

,-~,

in the event of the release of airborne radioactivity within the reactor room, radioactivity will be detected by the reactor room primary continuous air monitor (CAM). The high radiation alarm set point of the CAM will initiate a signal that causes the solenoid valves to vent and release the pressurized air from the dampers, causing them to close automatically. The reactor room is then confined.

The four dampers, upon closing, actuate microswitches which complete light circuits for four indicator lights in the reactor control room. The dampers may also be closed or reset (opened) manually 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 or fall in pressure sufficient to cause struc.tural 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 stack.

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) dif fuser.

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

1. 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 gallons per minute (gpm). The water is then returned to the reactor pool through the return line, located in the reactor pool approximately 3-11

3 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 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

& 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 o cooling tower. The secondary water is chemically treated to reduce losses via O 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 v.ater purification system removes particulates and ions from the primary cooling water. This minimizes corrosion of all reactor components, reduces activation in tne 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 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 prefilter, 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

\

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%_,i gauges, measuring the pressure drop across the prefilter, determine when the 3-12 i

t

filters need to be changed. From the prefilter, the primary water enters two parallel, mixed-bed demineralizers. The demineralizers maintain the water conductivity leve'. at better than 1 micromho per centimeter. Based on experience, activaticri 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 etch demineralizer uth only local readout provided.

The water is then passed to the primary cooling system return line. A schematic of the primary and secondary csoling 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 erater C)- 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 pool, the primary makeup "ater 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 surf ace of the pool.

3.3.5 N 16 Diffuser System 16 The N diffuser system reduces the gamma radiation level due to N 16 at the 6

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),

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3-14

b Q The N 16 diffuser system consists of a pump and motor mounted on the core 2 dolly structure and two sections of th-inch aluminum piping. Power for the motor is controlled from the reactor control room. The two sections of aluminum piping, located inside the cylindrical core support structure, serve as inlet and outlet lines to the pump. The diffuser pump discharges water through the outlet line at approximately 70 gpm. The outlet line is designed to discharge water in a tangential direction inside the cylindrical core support structure. This 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. The N 16 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 impor-tant to the safe operation of the reactor and the associated cooling system.

The level of the reactor tank water is monitored by a float-activated switch which causes an automatic reactor scram if =6 or more inches of water are lost below the normal pool full water level. When activated, the switch also causes a visual alarm on the reactor console and an audible and visual alarm on the annunciator panel in Hallway 3101. The annunicator panel informs the Security Watch (during off-duty hours) that an unusual reactor situation is present so that appropriate action may be taken.

The primary cooling water temperature is measured at three locations: #3-4 feet above the reactor core inside the core shroud (outlet), #3-4 feet below the nool surface inside the cylindrical core support shroud (bulk), and in the water monitor box of the primary water purification system (inlet). These temperatures are measured by high-temperature-coef ficient 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 60 C (Technical Specifications 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. This rod withdrawal prevent interlock does not prevent control rod insertion or reactor scram, however.

i 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 month. _

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 the 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 radicactive wastes generated as a result of normal reactor operation.

The handling of radioactive waste materials is conducted under the supervi-sion of the Radiation Safety Department to ensure compliance with NRC regula-

tions, s.v

cifically 10 CFR Part 20. Utilizing procedurss as set forth in the AFRRI Rad' mon Safety Department's Health Physics Procedures, the handling and dis-post 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 99.9 percent for 0.3-micron or larger c'immeter 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 partic-ulate and gas effluents which are released to the environment to ensure that the p release of Ar"I and other radioactive particulates and gases meet the requirements of 10 CFR Part 20.106.

3-16

Gaseous radioactive wastes associated with AFRRI reactor operation are generally limited to N and A "I. The N 16 is generated by fast neutron activation of oxygen (via Ol6(n.p)N16) in the coolant water in the vicinity of the -

core. The Ar"I is generated by neutron activation of naturally-occuring Ar40, a constituent of air, dissolved in the coolant water and in the air of the exposure rooms and the pneumatic transfer system. Other gaseous wastes exhausted from the AFRRI complex but not generated by the TRIGA reactor are O IS and N I3 3.4.1.1 Nitrogen-16. The amount of gamma radiation caused by N 16 released to the environment is insignificant due to the rapid decay (7.14 second half-life) of 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 pc>ol surface from N during large steady state power levels, access to the chained area around the reactor pool is limited during operations atove 100 kW(t). The amount of time necessary for the N 16 to be exhausted from the reactor room. into the stack is sufficiently long in comparison with the 7.14 second N 16 half-life to ensure that 1

N gamma radiation levels will be insignificant at the release point from the (U

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"I 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 ArU (released during 1979) was larger than the total Ar release during any year since 1972. This relative maximum in the annual Ar"I release totals is due to the number of reactor operations and the types of experiments performed during 1979. Based on the present experimental facility 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 curies).

To calculate the annual gamma air dose (due to the release of Ar4I) to a individual in an unrestricted area outside the AFRR1 complex, the following equation is utilized:I

/^' DY = 3.17 x 10"()(/Q)(Q) (DFY) where:

V) 3-17

p (j DT = annual gamma air dose at a point outside the AFRR1 complex due to Ar"I (mrad /yr) 3.17 x 10" = number of pCi per Ci divided by the number of seconds per year (XTC) = annual average relative concentration 2 (sec/m3)

Q = annual release rate of Ar"I (Ci/yr)

DFT = gamma air dose factor for a uniform semi-infinite cloud of Ar"I (mrad-m3/pCi-yr).

The total activity released is conservatively assumed to be 40 curies during the

/pCi-yrl . The

~

3 year and the gamma air dose factor for Ar"I is 9.30 x 10-3 mrad-m maximum annual average relative concentration is calculated in Section 2.3.2 to be 1.6 x 10-" seconds per cubic meter at 100 yards 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"I, which makes up the majority of AFRR1 gaseous radioactive wastes, is a factor of five smaller than the design objective of 10 millirads for gamma radiation given in Appendix 1 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 ALAR A concept is being implemented ef fectively.

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, v primarily to provide a means of controlling the release of radioactive liquid wastes 3-18

f- from the AFRRI complex to the public sanitary sewer system. Before being released to the sewer system, the liquid wastes are sampled and analyzed to ensure compliance with 10 CFR Part 20, Section 20.303. 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 co!!ects 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 waste 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 affluent's entry into the waste tanks, the effluent passes through grir.ders 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 ti.e 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.

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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 AFRR1 to support 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 I 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 AFRR1 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 l Material are detailed in the updated Physical Security Plan for the AFRRI-TRIGA Reactor Facility, on file with the NRC.

7y

( ,/

)

3-24

Monitoring of the Special Nuclear Material in accordance 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, v hich satisfies the requirements under 10 CFR /0, 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 AFRRI 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 O reector e gereter resgoeses te redietie , eier m s, eed erece d eres ievolvie s m oniter 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 pane' 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 l

for the safe cperation 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 AFRRI 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 wt in the reactor is i secured. When reacter personnel are present in the reactor control area, the audible alarm on the annunciator panel in Hallway 3101 is turned off.

3-25

3.6.1 Remote Area Monitors 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 in various areas of the reactor building where potential radiation hazards may exist due to reactor operation.

The monitors utilize scintillation detectors which measure gamma radiation with energies greater than 20 kev. The units have a range of 1 mR/hr to 6

105 mR/hr (with the exception of R-1 which has a range of 1 mR/hr to 10 mR/hr) and a nominal accuracy of 20 percent at all levels. The units have a time

~

constant of 2 seconds and a meter and alarm response time of less than 1 second.

The monitors activate radiation alarms, at various locations within AFRRI, at an adjustable radiation level. 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 AFRRI internal documents."

O 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 cfm) through a shielded filter

! assembly, which traps any particulate matte, greater than 0.3 microns in diameter.

l 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 1 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 k reactor room and is visible from Room 3156. The air sampled by this CAM is taken 1

3-26 t . _ _ _ _ .

' TABLE 3-1 REACTOR REMOTE AREA MONITORS RAM Location Readout Radiation Alarm R-1 Approximately 2 feet Meter in reactor Activates audible and visual above the reactor control room and alarm in the reactor room, pool surface on the Security Office in the reactor control room, east side of the (Room 3112) and in Health Physics Div.-

reactor pool sion (Room 1420); activates visual alarm in the Security Office (Room 3112); -

activates audible and visual alarm on annunciator panel in Hallway 3101 R-2 Approximately 7 feet Same as R-1 Activates visual alarm in above the floor on reactor control room and the reactor room in Security Office (Room west wall 3112); activates audible and visual alarm in Health Physics Division (Room 1420)

) E-3 6 feet above the floor Same as R-1 Same as R-2.-In addition, on the West Wall of the 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 floor on the West local visual alarm in the Wall of the prep prep area is located near area opposite ER #2 ER #2 i STACK Approximately 6 feet Same as R-1 Same as R-2.

above the floor of f Room 3152 on the out-l side of the air exhaust stack i

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i 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-II+. The alarm setpoints can be found in the appropriate AFRRI internal documents.

3.6.3 Stack Monitoring Systems fh l / 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 meas =ments 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 AFRRI 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 particulate 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 q assembly where radioactive particulates greater than 0.3 microns in diameter are U trapped. A G-M detector measures the gross beta-gamma activity emitted by the 3-31

O O O TABLE 3-2 REACTOR ROOM CONTINUOUS AIR MONITORS l CAM Location of Air intake Readouts High-Level Alarm Low-Level Alarm I 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 j core carriage itself itself

Strip chart recorder Activates audible and t

located on the unit visual alarm on reactor itself control room annun-clator panel Activates visual i

alarm in reactor y room 4

O 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 roem Alternate Directly above the warm Strip chart recorder Activates audible and Activates visual (Backup) drain along the east side located on the unit visual aiarm on the alarm on the unit of reactor tank itself unit itself itself 1

If the low level alarm is being utilized.

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3-33

(O g radioactive particulates trapped by the filter. The detector has nominal beta efficiencies of 19 percent and 26 percent for Cs I37 and Sr90, respectively. The systern has a range of 10 cpm to 105 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 exhatsted through the stack. The principal radioactive gases released from AFRRI through the stack are Ar"I,015, _

and N . The AFRRI-TRIGA reactor is the primary source of Ar"I, while the AFRR1 linear accelerator is the primary source of 0 15 and N I . 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 102 cpm to 10 6cpm. Normal background is nominally about 600 cpm. The air O maiee 8x t81 ex tem is reterned 1e 18e steck.

The counting gas utilized by the detector is propane, which a 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 l allows approximately 3 months' service before the propane cylinders are depleted, 1

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 4

[ internal documents.

3.6.4 Perimeter Monitoring A perimeter monitoring program is conducted by AFRR1 primarily to measure A

l V environmental doses received from radionuclides produced by the AFRRI-TRIGA

3-34 l

TABLE 3-3 STACK MONITORING SYSTEMS System Readout Radiation Alarm i

Stack Flow Monitoring Strip chart recorder in (Not applicable).

System reactor control room; However, EF-1 failure 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)

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O 3-35

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 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.

The locations of the perimeter monitoring system are as follows:

e At least 50 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.

A detailed description of the perimeter monitoring system can be found in C]

/

HPP 2-1, Perimeter Monitoring Program.3 3.6.5 Criticality Monitors Criticality monitors are used to detect and measure acute radiation dose 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.

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 room criticality monitor has a range of 1 mR/hr to 10 3mR/hr, and a nominal accuracy of i 20 percent at all levels. The unit is designed to activate a radiation 4

d alarm, both audible and visual, at an adjustable, preset level. The unit's readout 4

3-36

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l 3-37

l g meter and radiation alarms are located in the reactor control room. The reactor V 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 AFRRI internal documents" and conforms to 10 CFR 70.24.

The second type of criticality monitor at AFRRI 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

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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 C e Ceiling of the reactor room o East wall of the reactor control room o West wall of Hallway 3106 o East wall of Hallway 3105 o Beside door to Warm Storage Room (Room 1120).

3.7 SUPPORT FACILITIES The support facilities at AFRRI 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 AFRF.I from the National Naval Medical Center Power System. The power is supplied through the AFRRI substation (Structure #222 on Figure 2-1 and Table 2-1) locattJ behind the AFRRI complex.

The electrical power is supplied to the reactor building through two transformers located in the reactor building.

l 3-38

O b' One transformer supplies the low voltage requirements of the reactor building, 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 fan 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 tentilation systems in separate AFRRI buildings are also exhausted through the stack. The stack release point is a minimum of 18 feet above the highest building in the AFRF! complex. The total flow rate through the AFRRI stack is generally about 40,000 cfm during reactor operations.

3.7.3 Meteorological Monitoring m A Bendix-Freiz aerovane measures wind speed and direction at the AFRRI s' complex. The wind sensor is mounted on a tower located on the roof of Building #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 Systems There are two separate air supply systems that are important to the i operation of the AFRRI-TRIGA Mark-F reactor. Both systems basically consist of a compressor, an air drying tower, pressure regulators, and their associated piping, l valves, and fittings. The air drying tower in each system 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 first air supply system provides pressurized air (#9 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 l that the bearing seals rupture or leak. The first air supply system also supplies air

(#80 psi) to the transient rod system.

l The second air supply system supplies air (#20 psi) through a surge tans and A) t

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solenoid valve to the reactor room ventilation system dampers. To remain open, 3-39

i 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 l automatically.

i 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.

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3.3 REFERENCES

1. U.S. Nuclear Regulatory Commission, Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluat-ing Compliance with 10 CFR Part 50, Appendix 1, 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 (HPP's), Radiation Safety Departrnent.

4. Armed Forces Radiobiology Research Institute, Radiation Sources Division Instructions (RSD Instructions), Scientific Support Department.

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F ly/~'% 4.0 REACTOR 4.1 GENERA _L The AFRRI-TRIGA Mark-F reactor was designed and manufactured by Gen-eral Atomic and installed at AFRRIin 1962. The reactor is a tank-type light water reactor, with a horizontally movable core. The reactor tank is embedded in ordinary concrete. The cylindrical core consists of up to 37 standard TRIGA stainless steel clad cylindrical fuel elements with U-ZrH 3,7 as the fuel matrix material (3.5 weight percent uranium) enriched to a nominal 20 percent in Uranium-235; four aluminum-clad borated graphite control rods with solid -

aluminum followers; a startup neutron source and guide tube; and a removable in-core experiment tube (CET). The core moderator consists of the zirconium hydride in the fuel and water. The reactor core is reflected on the top and bottom by graphite end plugs in each fuel element, and at the periphery by water. The cylindrical fuel elements, the control rods, and the CET are positioned in the core in five concentric rings surrounding the centrally located transient control rod.

The fuel elements and control rod guide tubes are held in place with top and O 8ette m 8<ien iete - 18e 8<ie gi te surrounds the core.

<e 11 chee te e criieerice> s*<eee -81c8 There are four principle experimental facilities associated with the AFRRI-TRIGA reactor. These are Exposure Room #1, Exposure Room #2, the Pneumatic Transfer Systems, and the CET. 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.

The AFRRI-TRIGA reactor operates in two basic modes: 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 23 megawatt-seconds (intedrated power).

4.2 REACTOR TANK The reactor core is positioned in tne reactor tank under approximately 16 feet of light, demineralized water. The reactor tank water serves as radiation shielding, neutron moderator and reflector, and reactor coolant. The water purification and coolant systems which service the reactor pool water are discussed

( in Section 3.3.

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() The AFRRI-TRIG A 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

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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 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 19Y2 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 % 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 5 inch. The exposure rooms are described in Section 5.1.

Since the core can be positioned at numerous horizontal locations within the C 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):

o Effectively infinite thickress of water and 3/16 of an inch of aluminum 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 j 5.2.1), and from 0 to 6 inches of lead for 180

! e Effectively infinite water and 3/16 of an inch aluminum for 180 ; 1 inch of water, 3/16 of an inch of aluminum,1 inch of water, and 1/4 of an inch of aluminum for 180 o Effectively infinite thickness of water and 3/16 of an inch of aluminum for 360 .

The axial reflector, both above and below the active core, remains unchanged l for each of the above cases. It consists of the g.aphite end plugs in the fuel elements and the pool water above and below the core region.

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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 e Audible (during nonduty hours) and visual alarm on the annunciator panel in Hallway 3101 (Section 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 LEAD SHIELDING DOORS Two radiation shielding doors, shown in Figure 4-!, 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 n >

exposures while experiments are taking place at the other exposure room. The U interlocking doors are constructed of h-inch aluminum plate and S-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 18,000 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.

Pressurized air is supplied to the shield door bearings to minimize leakage of water into the shield door bearing housing.

The shielding doors may be rotated 90 to permit the core support carriage to move from one end of the reactor tank to the other. Rotation of the doors is accomplished using a fractional horsepower drive motor with slip clutch located in 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 p the lead shielding doors.

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4-4

( Limit switches are used to indicate the fully opened or closed positions of the 3 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 from one operating position within the tank to another. In addition to supporting the core, the carriage also serves as a support for the four control rod drives, the N 16 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 IV: 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 l Interlock System (Section 4.13). As a safety measure, mechanical stops are (3 mounted at both endt of the carriage track to prevent carriage overtravel at the U

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extreme limits. The carriage position is indicated on the reactor console. Precise

(,/ repositioning of the carriage may require the use of reference marks on the carriage tracks. The 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 distar.:e of the core dolly carriage at ER #1.

Region 2 - The range of positions between Region 1 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 maximum ~

travel distance of the core dolly carriage at ER #2.

Movement of the core carriage is initiated from the reactor console. Travel time 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 trail cable and flexible hose. The cable and hose are attached to the q

k' carriage control rod 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 continuous air monitor (CAM) inlet hose and the reactor room criticality monitor.

The core support structure consists of an aluminum cylinder approximately 36 inches in diameter and 12 feet high, and an aluminum adapter 5 fec* high and 19h inches in diameter (Figure 4-2). Both the cylinder and adapter are formed from l 5/16-inch thick aluminum plate. The cylinder connects the adapter to the core carriage. A vertical slot 16 inches wide extends the full height of the aluminum cylinder at its north side. This slot provides access to the inside of the support structure, permitting the installation and removal of core components without raising them above the pool water level.

! 4.5 CORE l The core forms a right circular 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, a startup neutron source and guide tube, and the CET. The core is enclosed in a 3/16-inch thick aluminum shroud attached O

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to the bottom of the core support adapter. Grid plates bolted to the top and bottom of the shroud hold the fuel elements, the control rod guides, the neutron source guide tube, and the CET in place. The active (i.e., fueled) reactor core region forms an approximate 17-inch diameter by 15-inch high cylinder. The core is positioned within the shroud so that its horizontal center line is approximately 29 inches above the bottom of the reactor tank. Serial numbers inscribed on the core components identif y individual fuel elements and control rods.

4.6 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) Ih-inch diameter holes. Feur of these holes receive the guide tubes for the three 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 rings are lettered A through F radially from the center. Spaces in each p ring are identified numerically in a clockwise direction from a reference radial line U 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)

K-inch diameter holes between the grid spaces are used for small in-core experiments and dosimetry. The upper grid plate also contains one hole near the edge of the grid array (between F-22 and F-23) for receiving the neutron source guide tube.

The lower grid plate is 16-5/8 inches in diameter and is goM anodized to reduce grid plate wear and to aid light reflection in the core. The lower grid plate contains eighty-seven (87) K-inch diameter holes to accept the fuel element bottom end fixtures, four (4) In-inch diameter holes to accept the control rod guide tubes, and thirty (30) 5/8-inch diameter holes to permit flow of cooling water. Both grid plates are 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.

4.7 NEUTRON STARTUP SOURCE AND HOI. DER An Americium-Beryllium (Arn-Be) neutron startup source (=3 curies) is' used a

j in the AFRRI-TRIGA reactor. The Am-Be source produces neutrons by the (o, n) reaction:

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The Am-Be neutron startup source utilized in the AFRRI reactor yields approx-imately 2.1 x 106neutrons per :econd per curie and consists of pelletized intimate mixtures of AmO2Be in a weight ratio of 1 to 10 and is doubly encapsuled in type 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 grid 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 .ests 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.

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4-11

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4.9 FUEL ELEMENTS The AFRRI-TRIGA Mark-F reactor utilizes standard TRIGA stainless steel clad, cylindrical fuel elements (Figure 4-5) in which the zirconium bydride moderator is homogeneous!y mixed with the enriched uranium fuel. '."he active part of each fuel element consists of a cylindrical rod of uranium-zirconium hydride containing 8.5 weight-percent uranium with a nominal 20 percent U 235 enrichment. The hydrogen-to-zirconium atom ratio of the fuel-meierator material is approximately 1.7 to 1. The nc,minal weight of U 235 in each standard fuel element is 38 grams. The uranium-zirconium hydride section is approximately 15 inches in length and 1.43 inches in diameter. A solid zirconium rod 0.18 inch in -

diameter and 15 inches in length is cer.cerad in the fuel region of each fuel element to provide structural support. Graphite end plugs, 3.44 inches in length, are located abov i and below the fuel-moderator section as top anj bottom axial reflectors.

Burnable poison (samarium) is included in each fuel element to minimize reactivity changes resulting from fission-product buildup and feel burnup. The bumable poison is mixed with aluminum to form wafers approximately 0.015-inch thick. These wafers are located between the fuel moderator section and the graphite reflector end plugs.

The 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. The 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 tM 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 chromei-alumel thermocouples embedded in the fuel-moderator section at different axial locations. The tips of the thermo-couples are located near the vertical axis of the fuel section. The 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. The

'hermocouple leadout wires pass through a sof t solder seal contained in the 1/2-inch outer diameter stainless steel tube welded to the element's top end fixture.

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4-12

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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 estrol rods anJ one transient control rod, all four of which contain neutron-alscrbing material. Control _ rod movement within the core is accom-plished using rack and pinion electromechanical drives for each standard control rod and a pneumatic-electromechanical drive for the transient control rod.

4.10.1 Standard Control Rods and Guide Tubes _

The shim rod (SHIM), safe rod (SAFE), and regula*ing rod (REC' 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 aluminum 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 a 15-1/2 inch long by 1-1/8 inch diameter solid aluminum rod. The control rod guide t'abes 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.

i j' 4.10.2 Standard Contro' Rod Drives l 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, j safety, and regwating 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 electro-i magnet. The magnet engage: an iron armature attached to the end of a long f connecting rod. The connecting rod is attached at its lower end to the control rod.

l 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 portion of the connecting rod just below the armature. Since k the upper portion of the barrelis well ventilated by slotted vents, the piston moves 4-15

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v 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 graded 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 gravitational force, 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 rigid adjustable foot plate. The foot plate, which projects 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 V engages the actuating lever of the ROD DOWN microswitch.

A second microswitch, the MAGNET UP, 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 g microswitch.

A third microswitch, the MAGNET DOWN micreswitch, is actuated when the electromagnet (drive) is at its lower limit. When the magnet reaches its lower j 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 inter'xk system prevents the simultaneous manual withdrawal of two or more stanoard control rods during steady state modes of operation and prevents the withdrawal of any standard control rod during pulse operation.

{.)\

4-18 l

4.10.3 Transient 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. The upper 15-1/4 inches of the tube are filled with a compacted borated graphite rod (25 percent free boron or boron compounds). The lower portion of the tube contains a machined solid aluminum follower. The transient rod operates in a guide tube identical to those used for the standard 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 Transient Control Rod Drive The AFRRI-TRIGA Mark-F is equipped with a pneumatic-electromechanical drive system for the transient control rod (Figure 4-9). The pneumatic-electro-mechanical drive, referred to as the transient rod drive,is basically a single-acting pneumatic cylinder. A piston within the cylinder is attached to the transient

^

() control rod by means of a connecting rod. The piston rod passes through an air seal e

V, at the lower end of the cylinder. For pneumatic operation, compressed air, admitted at the lowt end of the cylinder, is used to drive the piston upward. As the piston rn.es, 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 decelerated at a controlled rate during its final inch of travel. This action minimizes rod vibration when the piston l reaches its upper limit stop. Adjustment of the anvil's position, i.e., the volume of l the cylinder, controls the piston's stroke length and hence the amount of W"ity j

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 the l

pneumatic cylinder. De-energizing the solenoid valve vents the compressed 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.

(' ; This design ensures that the transient rod is fully inserted into the core except V

h-19

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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 l'D 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-switch, is actuated when the piston reaches its lower/ limit of travel. The transient rod anvil position, measured by a potentiomet'e r on the drive motor, is displayed on the reactor console.

4.11 REACTOR INSTRUMENTATION ( .

l 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 undependent

! " channels" which monitor the power level and fuel temperature of the core. A "ciannel" 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 l

'g reactor scram logic circuitry, and the system of rod withdrawal prevent interlocks.

(G l 4-21

c t 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 o.mpbelling 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 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 I 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 charc recorder between 0 and 100 percent of the power indicated by the power range select switch on the censole.

O T8e strig c8 art receres this eetget fer ali stee8x state medes ef egeratiee bet net during pulse operation.

4.11.2 Wide-Range Log Channe!

The wide-range log channel also measures reactor power from source level

(=10~3 thermal watts) to full steady state power (1 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 te 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.

O V

4-22

The v de-range log channel forms part of the rod withdrawal prevent b

v interlock system. The channel activates variable set point bistable trips in the rod withdrawal prevent interlock system (Sect on 4.12) if source level neutrons (=10-3 thermal watts (t)) are not present, if the reactor power level is above i 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 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 cnannels one and two are independent of one another but 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 pcwer 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 fj 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 ay 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.

The hign flux safety channels form part of the scram logic circuitry. When the steady state reactor power level, as measured by either high flux safety 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 power level during pulse operation, as measured by high flux safety channel two, reaches the maximum pulse power level specified in the technical specificat. 3, a bistable trip circuit is activated which causes an immediate reactor scram. A trip test knob for each O) 4-23

safetj 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

4. . ,

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 a' 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 PREVfNT INTERLOCKS A Rod Withdrawal Prevent (RWP) interlock stops any upward motion of the standard control -ads and prevents air from being supplied to the transient control rod unless specific operating conditions are met. An RWP interlock, however, does N) '

4-24

n

>g not prevent a control rod from being lowered or scrammed. Therefore, any RWP interlock prevents any further positive reactivity from being inserted qto the core until specific conditions are satisfied.

The system of RWP interlocks prevents control rod withdrawals under the following circumstances:

e RWP prevents air from being applied to the transient rod unless the reactor power level is under 1 kilowatt (t) e RWP prevents any control rod withdrawal unless, as a minimum, source level neutrons (=10~3 thermal watts) are present -

e RWP prevents any further control rod withdrawal unless the power lesel is changing on a 3-second or longer period as measured by the wide-range log channel during certain steady state operations e RWP prevents any control rod withdrawal unless high voltage is being supplied to the fission detector for the multirange thear and wide-range log channels e RWP prevents any control rod withdrawal if the wide range log channel is in any mode (i.e., position) other than OPERATE e RWP prevents any control rod withdrawal unless the bulk pool water tempeuture is less than 60 C (Technical Specification 1.imit).

4.13 FACILITY 11JTERLOCK 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 l reactor tank lead shielding doors and the reactor core shroud. These interlocks i

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 1

I positions are classified into three regions:

l l Region 1 - The range of positions within 12 inches of the maximum travel D_

g distance of the core dolly carriage at ER #1 1

l 4-25 i

( .- +

bl gj Region 2 - The range of positions between Region 1 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 maximum travel distance of the core cally carriage at ER #2.

The 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 reacter 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 O condition exists that is hazardous to personnel

~

e The reactor cannot be operated unless the reactor tank shielding doors are either fully opened 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 or, if the reactor tank shielding doors are fully opened, then both exposure room plug doors must be fully closed a The reactor cannot be operated within Region 2 unless the phig doors to both exposure rooms are fully closed and the reactor tank shielding doors are fully opened e The reactor cannot be operated within Region 3 unless the plug door to Exposure Room #2 is fully closed and the reactor tank shielding doors are fully closed or, if the reactor tank shielding doors are full / opened, then both exposure room plug doors must be fully closed e The reactor cannot be moved into Region 2 unless the reactor tank (9

%)

shielding doors are fully open.

4-26

Both exposure rooms are equipped with lighted " emergency stop" buttons. An emergency stop butter. is also located on the reactor console. Pressing any of 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 recctor console key which, when performed, automatically initiates a 3-minute delay with horns sounding in both exposure rooms. It also reinitiates the requirements fo:

opening 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 provid-ing a positive indication to the reactor operator that someone could be trapped in an exposure room.

4.14 SCRAM LOGIC CIRCUITRY The scram logic circuitry assures that a set of reactor core and operational conditions must be satisfied for reactor operation to occur or continue in p -

accordance with the technical specifications. The scram logic circuitry involves a V set of open-on-failure logic relay switches in series: any scram signal or l

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, result'ing in a reactor scram. The time oetween 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 ar.ticipated transients for the AFRRI-TRIGA reactor. The scram logic circuitry causes an automatic reactor scram under the icllowing circumstances:

e 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 e The pulse mode timer causes a reactor scram af ter a given elapsed time, as set on the timer, during pulse operations e The manual scram bar, located on the reactor console, allows the Reactor Operator to manually scram the reactor r3 e Movement of the console key to the OFF position causes a reactor kJ scram 4-27

O e The reactor tank shielding doors in any position other than fully open or fully closed will cause a reactor scram e Any movement of the core dolly carriage or the positioning of the core  ;

dolly direction switch, located on the reactor console, to any position other than the neutral or OFF position causes a reactor scram o Activation of any of the emergency stop buttons in the exposure rooms and on the reactor console causes a reactor scram o Any of the safety channels (fuel temperature safety channels and high flux safety channels) in any position other than OPERATE (i.e.,

cal.lBRATE or ZERO) causes a reactor scram o 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 e 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 e A loss of high voltage to any of the detectors for high flux safety channels one and two causes a reactor scram o Fuel temperature safety channels one and two will each initiate a t

reactor scram if the fuel temperature, as measured independently by either channel, reaches 600 C (technical specifications limit). This assures that the AFRR1 safety limit (core temperature) of 1,000 C for AFRRI stainless steel clad cylindrical TRIGA fuel elements, as stated in the AFRRI technical specifications and testing conducted by General comics, 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 600 C e A loss of reactor rx>ol water which leaves less than or equal to 14 feet of pool water abc,ve the core (technical specifications limit) causes a reactor scram. The actual operational limits for the pool water level O may be set more conservatively than the technical specifications limit.

4-28

l 4

i h 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 cabic, and the control handle. A fuel element is engaged by lowering the tool 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 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 b (bcw) and a device with a dial indicator for measuring fuel element length. The l

l tool is held in place and positioned in the reactor tank by a supporting aluminum i 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-l tions2 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.

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4-29

4.16 REACTOR PARAMETERS 4.16.1 TRIGA Fuel Elements The AFRRI-TRIGA Mark-F reactor core can accommodate up to 87 standard design TRIGA stainless steel clad, cylindrical fuel elements. The 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. The 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 homo-geneously combined with 91.5 wt-percent zirconium-hydride Hydrogen to zirconium atom ratio Nominal 1.7 H to 1.0 Zr 235 Uranium enrichment Nominal 20% U 235 Approximately 38 grams U 1oading per fuel element n Fuel element cladding 304 stainless steel, nominally bl 0.020 inches thick Burnable poison Samarium-aluminum wafers which form an integral part of the manu-f actured fuel element Axial reflector Graphite end slugs which form an integral part of the manufac-tured fuel element.

4.16.2 Excess Reactivity The core excess reactivity is measured each day a reactor operation involving the withdrawal of a control rod is planned. The 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. The sum of the absolute reactivity worths of all experiments in the reactor and the associated experiment and exposure facilities is limited by the technical specifications to 2.1 % Ak/k ($3.00).

,f~] The net change in reactivity caused by the maximum operational withdrawal

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of a standard control rod is approximately 1.4 % Ak/k ($2.00). This amount of 4-30

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reactivity is due to the complete withdrawal of the borated graphite section of the standard control rod from the core and the consequent insertion of the aluminum foilower into the core.

The net change in reactivity caused by the maximum operational withdrawal of the transient controi rod is approximately 2.8 % Ak/k ($4.00). This amount of reactivity is due to the complete withdrawai of the borated graphite section of the transient controi 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 I determined at periodic intervals specified in the technical specifications.

1 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.

O seceese ei the ierse gremet nesetive temgerature ceeificient, eey ierse <>$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-maticaliy 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,

(

thus limiting the reactor steady state power level.

I These intrinsic negative temperature coefficients of reactivity (i.e., prompt and steady state) are primarily due to the hardening of the neutron spectrum l 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 capture before fission can occur increases, which introduces negative reactivity

{ and, thus, limits the reactor power level actually achieved. Hence, any introduc-tion of excess reactivity which increases the reactor power level and fuel temperature will be immediately and automatically compensated for.

4-31

i

)

A second important but smaller contribution to the negative temperature l 238 coefficients of reactivity is due to Doppler broadening 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.

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()\ 5.0 EXPERIMENTAL FACILITIES 5.1 GENERAL The AFRRI-TRIGA Mark-F reactor serves as a source of mixed ionizing and neutron radiation for research and radioisotope production. The unique flexibility of the AFRRI-TRIGA reactor is achieved by the horizontally movable core which can traverse from one irradiation position to another. The major irradiation facilities in which experiments can be carried out using the eactor are:

e Exposure Room #1 o Exposure Room #2 e Pneumatic Transfer System o Pool Irradiation e In-Core Experiment Tube.

The In-Core Experiment Tube 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 The 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) 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 The physical dimensions of ER #1 are approximately 20 feet by 20 feet by 10h feet high. The 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 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 6h feet of concrete separate the exposure rooms

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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 surf aces 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 f ast 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.

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thick masonite panels palated with gadolinium-oxide paint (#11 mg/cm2). 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 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

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v' proj ection. 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 stacked 1-inch lead sheets mounted in aluminum frames. The front curtain i= 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

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consists of an aluminum frame about 30 inches square by 12 inches thick that can 5-3

) 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 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 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 y 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 (el 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 l

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 the microswitches.

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 engaging the coupler to the manual drive. The handwheelis then turned to close or l p open the plug door. Manual operation of the exposure room plug doors overrides v

5-4

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 1 with the reactor tank lead shield doors closed or while the lead shield doors are open with the 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 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.

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5.2.3 Exposure Room Monitoring 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 monitorting (gross beta-gamma activity) of airberne radioactive particulate matter present in ER #1 and ER #2. A general description of the CAM's is p ren 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 certain preset airborne activity levels provide personnel in the prep area with information concerning airborne activity levels in the exposure rooms.

Before opening either exposure room plug deor, the exposure room airborne activity levels, as measured by the appropriate exposure room CAM, must be below

- a certain level, determined by the Chief of the Radiation Safety Department. If 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 area past the opened plug door and in the exposure room itself. If high radiation levels exist, the plug l

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 expos ye rooms will have appropriate personnel radiation monitors and be assigned limited stay times, if significant radiation levels exist in either area, to assure compliance with both NRC (10 CFR 20) and internal AFRRI occupational dose limits.

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

The prep area CAM alarms, both audible and visual, are local (as with the exposure room CAM's). Remote Area Monitors 'iRAM'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 Sectie. 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 20 percent at all levels. The E-4 and E-3 RAM _

readouts are provided locally on the units themselves. R AM's E-4 and E-5 activate radittion alarms within the AFRRI complex at an adjustable, preset radiation level. RAM's E-4 and E-3 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 no the de' ctor or a loss of signal to the remote readout occurs '5"~ e alarm). The RAM's are calibrated at regular intervals using a radiation ' nown intensity. The location of the prep area RAM's, the

( radiation readouts and ,sarms are shown on Table 5-1.

O] 5.2.4 Patch Panel System A patch panel system exists within the AFRRI reactor building so that e'ctrical 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 Panel in CR #1 is connected to a Patch Panel in Hal! 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 TR?.NSFER SYSTEM The Pneumatic Transfer System associated with the AFRRI-TRIGA reactor permits samples, contained in sealed containers called " rabbits," 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 Pneumatic Transfer System is divided into two separate tube banks, I designated as A-Tube Bank and B-Tube Bank (Figures 5-5 and 5-6). The A-Tube 5-8

i O O O TABLE 5-1 PREP AREA REMOTE AREA MONITORS Remote Area Monitor Location Readout - Radiation Alarms E-3 Approximately 6 feet above Meter in the reactor Activates audible and visual

the floor on the prep area control room and in alarm in Health Physics i

west wall opposite ER #1 Room 3112 Division (Room 1420); acti-

plug door vates visual alarm in reactor control room and in~ Room 3112; activates local visual alarm on the prep area east wall near ER #1 E-6 Approximately 6 feet above Same as E-3 Same as E-3 (except that the floor on the prep area the visual alarm activated west wall opposite ER #2 in the prep area is located Y' on the east wall near ER #2) w E-4 Approximately 6 feet above Meter on unit itself Activates audible and visual the floor on the prep area and meter in Room 3112 alarm on unit itself; acti-

} east wall north of ER #1 vates visual alarm in Room J plug door 3112; activates audible

! and visual alarm in Health i Physics Division (Room 1420)

E-5 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 ER's #1 and #2 1

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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 Radiochemisry 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 i 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 fiorn 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 i is drawn into the system through the common air tube. When the UP-DOWN valve i

as placed in the DOWN position, .the partial vacuum is applied to the termini 1

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.

I The relief volves 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 l 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 wall in a small adjoining room north of the Radiochemistry Laboratory.

O.

v i

5-13

c l-( .

The types of experiments conducted with the Pneumatic Transfer System i primarily involve radioisotope production and activation of small samples for ,

i subsequent neutron activation analysis. AFRRI's health physics procedures and internal RSD instructions administratively govern the use of the pneumatic transfer system IAW the technical specific.ations.

5.4 POOL IRR ADIATION

The pool-type tank of . the AFRRI-TRIGA Mark-F reactor permits the

- irradiation 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 3-inserted into the core for irradiation. The decision to perform experiments in the L: reactor pocl 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 l the technical specifications. . At no time shall experiments be performed in the i reactor pool which would introduce more excess reactivity into the core than the i technical specifications allow. The use of the reactor pool as art experimental facility also includes the use of portable beam tubes placed in the reactor pool.

O- The types of experiments performed in the reactor pool are similar to those  ;

i 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 reactor core into these exposure facilities. Ar"I is produced by thermal neutron activation of naturally occurring Ar 40 in the air of these experimental facilities (note that some air is always dissolved in the pool water). Since Ar"I is a noble 4

gas, it is not retained in the human body and radiation exposures are a result of direct exposure (immersion) to the Ar gamma rays given off during its natural

decay.

i j ,

The exposure rooms and Pneumatic Transfer System are designed to minimize I Ar"I- exposures to the AFRRI staff. The exposure rooms contain design features, described in Section 5.2.1, which minimize the production of Ar"I. Both exposure rooms and the Pneumatic Transfer System air is exhausted to the AFRRI stack by

way of the reactor building ventilation system. This limits egress of Ar"I to

{ occupiable areas inside the AFRRI complex. The stack monitoring systems provide r

4 I

5-14 i

r,e ,- - ,- ~ e n ~ v .,m-w,-w,--n-mw,m.-,.,,-nw-,e.,n-,,_,- +

9 q a means of ensuring that releases of Ar" to unrestricted areas meet the U requirements of 10 CFR Part 20.106.

Typical Ar"I production rates in these experimental facilities, based on measurements of Ar"I releases through the AFRRI stack during reactor operations, are:

e Exposure Room #1 with reactor core at extreme position closest to ER #1 0.58 mci /kW-hr o Exposure Room #2 with reactor core at extreme position closest to ER #2 2.10 mci /kW-hr e Pneumatic Transfer System with reactor core -

at position adjacent to pneumatic tubes 0.47 mci /kW-hr.

5.6 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 Specifications and a set of internal written instructions, Of v RSD Instructions 5-X series (approved by the Chief of the Radiation Sources Division (SSRS) and reviewed by the Reactor and Radiation Facility Safety Committee (RRFSC)).

Before performing any proposed experiment (with the exception of routine operations as outlined in the Technical Specifications and RSD instructions, 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 (RRFSC) and other persons or groups as specified in writing by the Director of AFRRI. A Reacter Use Request (RUR) must also be completed, approved, and signed by the AFRRI Reactor Physicist-in-Charge (PIC) before an experiment can be performed. Safety evaluations for proposed experiments conducted without prior NRC approval under provisions of Section 50.59,10 CFR Part 50, are reviewed by RRFSC and 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 tra:ned in radiological safety to consider and recommend alternative operational profiles, such as different core O

V 5-15

positions, power levels, and irradiation times, etc., which might reduce staff exposures, the release of radioactive materials to the environment, or both.

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 fc,:

the AFRRI-TRIGA reactor state that:

e The sum of the absolute reactivity worths of all experiments in the -

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.

e Each fueled experiment shall be limited so that the total inventory of n iodine isotopes 131 through 135 in the experiment is not greater than e

C' 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 reactra in quantities 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 0.5 % 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 l

experiment.

,7

< Typical experiments that have been performed at AFRRI 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:

l l

5-16

j e Irradiation of up to 500 pounds of animals for up to 1 Mw-br with or without shielding e Irradiation of various materials in the pneumatic tube system or CET.

These materials included inorganic compounds of 70 different elements limited on me 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 e" m . eristics.

e Preparation of Argon t+1 calibration standards in the exposure rooms and irradation of up to 5 grams of sodium carbonate in the pneumatic _

transfer system o Irradiation of old paintings for radiographic analysis of age and authen-ticity e Irradiation of electronic components in the exposure rooms and the pneumatic transfer system. These components incitute integrated circuits.

O I

9 4

1 3-17

6.0 SAFETY ANALYSIS 6.1 POTENTIAL CONSEQUENCES ASSOCIATED WITH TRIGA REACTOR OPERATIONS 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-tion, 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 Stata Operation at power levels not to exceed 1.1 megawatt I

. (MWI ) for testing operational purposes and 1.0 megawatt (MW ) for l 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 l September 1980. The total reactor operation from Jariuary 1980 to l

September 1980 has been less than 26 MW-hr.

l 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 limited by the technical specifications to 2.8% A k/k reactivity in the puise mode.I The p average " maximum-pulse" used d'. ring AFRRI's TRIGA pulse operation has been 28 MW-sec. For calcalations, the assumption of a maximum 6-1

= _ - _

(]j excursion pulse of 40 MW-sec was utilized as an upper limit for pulse operations or an inadvertant transient.

e The . reactor operations will be supervised by individuals who are qualified operators trained in the detection and evaluation of radio-logical consequences.

In summary, the analysis indicated that the consequences from the Design Basis Accidents of a fuel element drop accident or a fuel element clad failure accident were insignificant. The 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 AFRR1 facility.

Doses from these postulated accidents to individuals beyond the boundary of the National Naval Medical Center site would be sadnificantly less than 1 mrad.

Therefore, it is concluded that the operation of the AFRR1 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 o'. these materials. Reactor operations must be supervised by responsible individua'.s who are trained in the detection and evaluation of radiological consegences. Administrative, operational, 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.

The radiological consequences associated with fuel elements are of the same nature as those associated with isotope production. Because cf their high radiation levels, fuel elements should be kept under water for shielding purposes. Adminis-trative policy requires that a fuel element not be removed from the 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 cask 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 which a cladding failure occurs in air.

u 6-2

When proper administrative, operating, and health physics procedures are utilized, the handling of radioactive materials 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 element:. vith 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 nvt (neutrons /cm2),

The ATPR fuel elements we e subjeaed 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 h temperature coefficient has been measured to be approximately $0.017 reactivity loss per 1 C rise in fuel temperature.

Th6 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 by 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 n 2.8% A k/k (AFRRI-TRIGA technical specification limit) results in a maximum b) average fuel temperature of less than 550 C.

6-3

Therefore, based on operating experience of the ATPR and the Fuchs-Nordheim mathematical model, it can be concluded that the rapid insertion of the

, total excess reactivity of 3.5% A k/k would not represent an undue risk to the operating personnel cr the general public.

6.2.3 Improcer Fuel Loading 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 fuelloading 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/k would not present a danger of damaging the reactor or fuel. The reactor would undergo 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 i 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 Gases in the Reactor Coolant l

l 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 i half-life of 1.83 hours9.606481e-4 days <br />0.0231 hours <br />1.372354e-4 weeks <br />3.15815e-5 months <br />." 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 Law. At a water temperature of 70 F (21 C), the corresponding water vapor pressure is 26 mm Hg.

l 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-tion of argon in water of 1.367 E-8 gram mole of Ar-40 per I cm H 2O is obtained.

6-4 l

2 Using a microscopic thermal neutron absorption cross section of 6.lE-25 cm for 3

Ar-40 , the macroscopic absorption cross section of argon in water becomes SE-9 cm .

Activity production, assuming a saturation condition and an irradiation time, t, can be given as:

A = On (1-e- At p ) I, t (g)

P 3.7 E+4 where:

A p

= activity in reactor water (pCi/cm ) _

2 Gn = thermal neutron flux (n/cm -sec) t = core circulation time (sec)

E = macroscopic absorption cross section (cm-I) a 3.7 E+4 = constant to convert disintegrations per second to microcuries A = decay constant (sec-I)

(~N t = irradiation time (sec).

O The average thermal neutron flux (Gn ) in the reactor core is estimated to be 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 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" yields:

i Ap= 5.7E-4 p Ci/cm 3, and a total core activity of:

Q = 5.1 p Ci/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 l

decay to be negligible, the maximum activity reaching the surface is 5.1 pCi/sec.

t

! 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 pCi/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

g- vapor pressure will increase and the amount of dissolved Ar-40 will decrease, V] resulting in a lower generation of Ar-41 escaping to the reactor room.

The radioactive argon (Ar-41) will escape from the reactor pool, dissipate in the reactor room air, and recirculate through the ventilation system. It is 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 share (8.24E+7 cm3 / min (2910 ft3/ min)) of the 1.133E+9 cm3 / min (40,000 ft3/ min)

.through the stack.5 The added exhaust is from the other AFRR! 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:

5.1 u Ci/sec = 3.7 E-6 p Ci/cm 3

,q 3 V 1.37 E+6 cm /sec The concentration in the stack is 2.7 E-7 pCi/cm 3and 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/cm3 by assuming submersion in a spherical source equivalent to the reactor room volume.

The exposure rate can be given as:

S D= K (1-e ) (2) where: D = exposure rate (R/hr) 3 Sy = volume source (Mev/sec-cm )

l p = attenuation coefficient for air (cm-I)

K = flux to dose conversion factor (Mev/cm2-sec per R/hr) r = radius of equivalent sphere (cm).

The volume of the reactor room is approximately 9.2 E+8cm3 , which is equivalent to a sphere with a radius (r) of 600 cm.5 An Ar-41 concentration of 3.7

! E-6 pCi/cm 3has a gamma source strength (Sv ) equal to 0.08 Mev/cm3 -sec. For (3

%)

l 6-6 l

\ . , . _

g the Ar-41 gamma photon of 1.29 Mev, the linear attenuation coefficient for air (p)

W 2 is 7.2 E-5 cm-I and the flux to dose conversion factor (K) is 5.44 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 uccurs 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 oxygen in water is approximately 0.89 gram mole of 0-16 per 1 cm 3 H.,0. Using a microscopic thermal neutron absorption cross section of 1.85 E-29 cm I for 0-166, the macroscopic absorption cross section of oxygen in water becomes 9.9E-6 cm -I .

Substituting the appropriate values in Equation 1 yields:

Ap = 8.4 E+2 g Ci/cm ,

o 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 Ci/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 + Scm ) in which the N-16 activity per second is distributed was assumed to be 100-cm high for negligible decay. The 3

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 radios of 136 cm results in an exposure rate from the calculated N-16 concentration of 350 mR/hr immediately abave the reactor pool. The N-16 activity that is being evaluated by this analysis is the amount that may bc released from the pool. It should be noted that due to the shon half life of the N-16,

( ) essentially no N-16 is released from the stack.

6-7

v 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 approximately 0.5 pCi/sec (September 22, 1980, Ar"I Report) as compared to the calculated value of 5.1 pCi/sec, a reduction by a f actor of 10. A typical concentration at the stack has been approximately 3 E-8 pCi/cm 3as compared to the calculated value of 2.7 E-7 pCi/cm ,3 a f actor of 10 reduction.

Gamma radiation :evels 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 Safety 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 pever 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 f ail and release all of the Ar-41

[m~u) activity, the resulting total whole body dose would be less than 2.7 mrad to an 6-8

i ,

p individual more than 25 meters from the AFRRI facility (Equation 3, d 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-TRIG ~ reactor. Such a whole body exposure would not rep.esent 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 fuel-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. Manufacturing inspection and quality controls assure that the possibility of cladding failure is minimal.

2 Experim'ents conducted over a period of 11 years by General Atomic on3.5 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 little significance to the fission-product release fraction considered in this system. The U-ZrH icel has a relatively

/

V low reactivity in water, steam, and air at temperatures up to about 600o2 C.

6-9

hq 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 hydrogen is due to oxide film formation which inhibits hydrogen loss.

6.3.2 Fuel Element Cladding Failure 6.3.2.' Summary of Previous Experience. 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 isotope release from a full-size TRIGA fuel element during irradiation e 1966 - The measurement of the fractional release of several isotopes from small specimens of TRIGA fuel material during and after irradiation- 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 dering irradiation in a duplication of the 1960 experiment.

In addition, qiiench tests were performed on TRIGA low-enriched uranium (L'lU) fuel samples from temperatures ranging from 300 to 1200 C. The subsequent .'uel 0

samples quenched from 800 ,1000 ,1100 , and 1200 C showed remarkably beaign response to the test conditions.3 In the course of developing a reactor for pulsed operation, General Atomic l

I lias experienced three fuel element cladding failures in their Torrey Pines TRIG A reactor. These cladding f ailures have been of two types:

l e Two cladding failures caused by thermal ratcheting of previously used l p aluminum cladding during the cycling l d l

l i 6-10 l

e A cladding f ailure 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 f ailures. The cladding f ailure of an instrumented element was apparently the result of internal pressure buildup, water seepage, or possibly the pressure of volatile materialinside the element.

The consequences of the cladding f ailures were all minor and in the worst case experience may be summarized as follows:

e The activity in the reactor water reached a maximum of 0.2 pCi/cm3.

It decayed very rapidly and was measured 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />" af ter the cladding failure to be 5x10-5 pCi/cm ,3 a 4,000-fold reductic a e The activity in the air of the reactor room reached about 10 times the 7., maximum permissible concentration for fission products and then i \

(j decayed rapidly. The faulty fuel element was removed and experiments were resumed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> af ter the activity release. The maximum integrated exposure to operating personnel resulting from this release was 1 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 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 public.

6.3.2.2 Calculation of Maximum Fission Product Release After a Fuel Element 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.

A

! v 6-11

/

t,]s The calculations are based on the fact that as the reactor operates, the fission products will build in the uranium-zirconium fue! mixture until 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 inadvertent transient following steady state operation.

Calculated Fission Product Inventory. To determine the various inventories

~

of the fission products prouuced 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 iodines 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 kr;,. cons, xenons, and iodines as calculated are shown in Table 2, Appendix C.

O- 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 , f less than 0.01 percent, the theoretical limit of 0.1 percent gap activity for fission O

v 6-12

product gases of noble gases and iodines, as stated in Reference 2, will be used in the consequence analysis for the Design Basis Accidents (Section 6.3.4).

Air and Water Activity Following Cladding Failure. As shown 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 -

35 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 during pulse operation following steady state operation.

V As given in Reference 5, the volume of water in the TRIGA reactor tank is approximately 5.7 E+7 cm 3 and the volume of air in the reactor room is approximately 9.2E+8 cm3 . For 1he 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 falls. Due to the low pressure in the gap and the small gap size in a TRIGA fuel element, the iodine released from a cladding f ailure 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 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

' ahumptions are used to determine the radiological consequences due to the Design Baqia. Accident of a Fuel Element Cladding Failure (Section 6.3.4.2). Of the 7 curie *, that could be released in the cladding failure of an average fuel element, 1-ss,:than 3 curies would be radioiodine while the remaining 4 curies would be the l

.- nehle gases, krypton, and xenon. Therefore, the concentration in the water would l

6-13 I

be less than 0.05 pCi/cm 3, while the air concentration would be less than 0.005 pCi/cm 3.

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 calculate the exposure rate, Equation 2 for submersion exposure from a spherical source is used. The 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. The 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/sec 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/sec of kryptons and xenons into the reactor room volume. The air concentration of 0.005 pCi/cm previously calculated is equivalent to a gamma volume source strength (Sy ) of 1.6 3

E+2 Mev/cm -sec.

O V For the noble gas mixture, the linear attenuation coefficient for air (p)is 9.7 E-5 cm-I and the flux-to-dose conversion factor (K) is 5x105 Mev/cm2-sec 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 i 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 fcitowing steady-state operation, a person could remain in the reactor room for mcre than 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> before exceeding the allowable quarterly occupational exposure lim?t 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 Physicist-in-Charge (PIC) 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 1

l l

l l

l 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 doo s 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.

h) v Predicted whole body dose rates to individuals within the AFRR1 f acility, 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 1MW steady state operation (100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> to reach saturated fission product inventory) has been allowed to decay after 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, or 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 i a s > substitution of stainless steel cladding for the former aluminum cladding on the 6-15

- fuel elements. The fission products released from 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 Pasquill type F stability condition with a 1 m/sec wind speed and a cross sectional area for the AFRRI facility of 450 m 2. 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 cicud 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 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 folicwing equation was utilized:II D =

0.25 (E )(Q)(X/Q)(F) (3) where:

Dy = integrated whole body dose (rad)

E = gamma energy released per disintegration (Mev/ dis) l Q = total activity released (Ci) 3 X/Q = diffusion coefficient (sec/m )

O l 6-16 l

O F = ratio of gamma dose from finite cloud to infinite cloud with same centerline concentration 0.25 = constant for semi-infinite cloud dose analysis.

To calculate the thyroid dose to an individual outside the AFRRI facility, the following equation was utilized:II D th =

(BR)(x/Q)(Sth} (")

where:

D th = integrated thyroid dose (rad)

BR = individual breathing rate (m3/sec)-3.47x10~" m3/sec 3

X/Q = diffusion coefficient (sec/m )

S = (DCF) (Q), iodine inhalation source term (rad) th DCF = dose conversion factor (rad /Ci)

Q = iodine activity released (Ci).

The whole body and thyroid doses calculated for individuals downwind from v the AFRRI facility in the event of a fuel element drop accident are given in Table 5, Appendix C. The conservative assumption is made that the radioactive gases relen ' in the AFRR1 reactor room will be released directly to the atmosphere without significant holdup within the facility. The 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.

The calculated whole body dose is insignificant at all distances downwind from the AFRRI facility. The maximum calculated thyroid dose of 45 mrad is more than a factor of 100 below the projected thyroid dose of 5 rad, below which no protective action is reccmmended.12 The calculated thyroid dose for individuals beyond the boundary of the National Naval Medical Center site within which the AFRR1 facility is located would be significantly less than 1 mrad.

6.3.4.2 Fuel Element Cladding Failure Accident. Another Design Basis Accident for the AFRRI reactor is postulated to be the cladding failure of a fuel element

, during a pulse operation or inadvertent transient following steady state operation

{} at 1 MW. The assumed fission product inventories in the core are those given i

1 6-17

previously in Table 1, Appendix C (1-MW operation) and Table 2, Appendix C (40-MW-sec pulse operation), and analyzed in Section 6.3.2. As discussed for the f ailure 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. The 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 f acility.

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. The resulting whole body and thyroid doses calculated for _

individuals downwind from the AFRRI facility are given in Table 7, Appendix C.

The 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.

The calculated whole body dose is insignificant at all distances downwind from the AFRRI facility. The maximum calculated whole body dose is less than 2 mrad or a factor of 500 below the projected whole body dose of I rad, below which

[ no protective action is recommended.12 The 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 recommended.I2 The calculated thyroid dose for individuals beyond the boundary of the NNMC site would be significant.y less than I mrad.

%,j 6-18 t . _

6.4 REFERENCES

1. Amendment #17 to the Technical Specification, Facility License No. R-84 for AFP RI-TRICA 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-En-!ched Uranium Fuel Quench Test, GA Report No. A15384 (San Diego, California: July 1980.

4. Lederer, C. M., 3. M. Hollander, and 1. Perlman, Table of Isotopes, Sixth Editior. (New York: John Wiley & Sons, Inc.,1967). _

5. Final Safeguards Repcrt: 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, 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 l Energies, APED-5398- A (Revised) (October 1,1968).

8. Bolles, R. C., and N. E. Ballou, Calculated Activities and Abundancies of f'] U-235 Fission droducts, USNRDL-456 (August 30, 1956).

1 G

9. U.S. Nuclear Regulatory Commission, 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--1968, Air Resources Environmental Laboratories, Environmental Science Services Admin-istration, U.S. Departn.ent 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 ? (June 1974).

12. Manual of Protective Action Guides apd Protective Actions for Nuclear Incidents, EPA 420/1-75-00 (Septemoer 1973).

em

/

6 '.^

(~h V 7.0 ADMINISTRATION 7.1 GENERAL 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.

The mission of AFRRI is to conduct scientific research in the field of radiobiology and r.uated 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 DF *. as Chairman, and the Surgeons General of the Army, Navy, and Air Force. The Board of Governors ensures equal support and participatim by the three e rvices in Institute activities and a balanced response by the Institute to the needs of each service.

The supporting staff of AFRRI consists of both military and civilian person-nei. 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 AFRRI is shown on Figure 7-1.

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O 7.2 THE 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 (RRFSC) and the Radionuclide and X-Ray Safety Committee (RXSC), report directly to the Director concerning safety matters at the Institute. The RRFSC is responsible for advising the Director in matters concerning the safe operation of the AFRRI-TRIGA reactor and shall be discussed later. The RXSC exercises no control over the reactor and its operations and, therefore, will not be discussed further.

The Director of AFRRI must be a medical service officer of either the Army, Navy, or Air Force, who is technically qualified in the fields 9f medicine, radiobiology, and radiation effects research. He is appointed by the Director of 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 direct overall responsibility for the safe operation of the AFRRI-TRIGA reactor, he is authorized to communicate directly with the Nuclear

( Regulatory Commission without prior knowledge or approval of any other office, including DNA headquarters. The Deputy Director assumes the responsibilities of the Director in his absence.

The Scientific Advisor to the Directorate of AFRRI must be technically and administratively qualified in the fields of radiation, radiobiology, and radiation effects research. He must have recognized capabilities, such as demonstrated experience in managing research programs, high professional stature as evidenced by publications in technical journals and appointments or membership in profes-sional societies, and additional education above the Bachelor's degree level. 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 advising the Director on technical matters, providing technical direction to the diverse professional staff at AFRRI, and acting as liaison between AFRRI and other agencies.

7.3 REACTOR AND R ADIATION FACII.ITY SAFETY COMMITTEE The Reactor and Radiation Facility Safety Committee (RRFSC) is directly g responsible to the Director of AFRRI. The committee oversees the safety aspects of the following facilities at the Institute:

7-3

e AFRRI-TRIGA Reactor e LIN AC Linear Accelera+or e Cobalt Facility e Theratron Facility e X-Ray Facility e Any other machine capable of producing radiation in excess of I rad /hr in an area which can be occupied at any time.

The safety aspects of these facilities include the physical facilities, the planned -

operations, 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.

The voting membership of the RRFSC, as appointed by the Director of AFRRI, shall be composed of:

e At least two of the following, one of whom must be a licensed NRC U Senior Reactor Operator: Chairman, Scientific Support Department; Chief, Radiation Sources Division; Physicist-in-Charge of the AFRRI-TRIGA reactor; Head of other large radiation sources at AFRRI e Chairman, Radiological Safety Department e A representative of the AFRRI Directorate e Radiological Safety Officers or Source Facility Operators from at least three outside facilities; one of the facilities must be a non-Department of Defense facility and at least one of these individuals must be a reactor physicist or a reactor health physicist e A representative of the AFRRI scientific staff e Any other qualified individuals as designated by the Director of AFRRI.

The RRFSC or a subcommittee thereof meets at least four times each calendar year. The full RRFSC meets at least semiannually. A quorura of the RRFSC for review consists of the Chairman of RRFSC (or his designated alternate) m and at least four other voting members or alternates. A majority of those present b must be regular members.

7-4

All RRFSC decisions are binding upon beint, endorsed by the Director of AFRRI.

7.4 SCIENTIFIC SUPPORT DEPARTMENT The Scientific Support Department (SSD) supports the AFRRI scientific research program by providing, operating, and maintaining radiation sources, which include the AFRRI-TRIGA reactor. The Chairman of the Scientific Support Department must be technically and academically qualified in fields relating to the operation of radiation sources, dosimetry instrumentation, and scientific research equipment assigned to the various divisions of SSD. The Chairman of SSD may also _

serve as a permanent voting member of the Reactor and Radiation Facility Safety Committee at the discretion of the Directo.

The Scientific Support Department is composed of three divisions:

e The Radiation Sources Division (SSRS) e The Radiological Physics Division (SSRP) e The Nuclear Sciences Division (SSNS).

The Radiation Sources Division has sole responsibility for the operation and maintenance of the AFRRI-TRIGA reactor and other assigned radiation sources.

The Radiological Physics Division provides dosimetry techniques and maintains dosimetry instrumentation for conducting measurements and dosimetry for all radiation sources and experiments utilizing AFRRI radiation sources. The Nuclear Sciences Division primarily performs radiation research and also operates and maintains certain scientific equipment such as the electron microscope and the mass spectrometer.

The Chief of the Radiation Sources Division must be academically trained and technically qualified in fields relating to the operation of the AFRRI-TRIG A reactor and other sources assigned to SSRS. He is responsible to the Chairman of the Scientific Support Department for coordinating and being cognizant of all matters relating to the safe operation of the AFRRI-TRIGA reactor. As Chief of SSRS, he may also serve as a permanent voting member on the Reactor and Radiation Facility Safety Committee at the discretion of the Director.

The Radiation Sources Division is composed of various branches, each branch

(~N being responsible for a radiation source assigned to SSRS. The Reactor Branch is V

7-5

() the branch within SSkS responsible for the operation and maintenance of the AFRRI-TRIG A 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 Physicist-in-Charge, the Reactor Operation Supervisor, and the Reactor Operators. The Chain of Command of the Reactor Branch is given ir. Figure 7-2.

The Physicist-in-Charge (PIC), as principal administrative officer for the -

Reactor Branch, is directly responsible to the Chief of the Radiation Sources Division for operational, technical, administrative, 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 and AFRRI instructions. The PIC shall postess an NRC Senior Reactor Operator's License for the AFRRI-TRIGA Mark F :L; tor.

The Reactor Operation Supervisor (ROS) is responsible to the Physicist in-Charge for the efficient and safe operation of the AFRRI-TRIG A 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 tafe and legal operation of the AFRRI-TRIGA reactor in compliance with NRC licenses and regulations, and internal RSD and AFRRI 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 RADI ATION SAFETY DEPARTMENT The Radiation Safety Department (SAF) is responsible for developing and maintaining a comprehensive Health Physics Program, encompassing all sources of radiation within AFRRI. The Chairman of the Radiation Safety Department must be academically and technically qualified in the field of radiological safety and health physics as they relate to reactor operations. He is directly responsible to the AFRR1 Director for the activities of the Radiation Safety Department. If possible, he should a a Board Certified Health ohysicist of the American Board of l

7-6 I .

0 OIRECTOR REACTOR AND RADIATION

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7-7

Health Physics. As Chairman of SAF, he serves as a permanent voting member of the Reactor and Radiation Facility Safety Committee.

The Radiation Safety Department is responsible for developing and maintain-ing an adequate radiation protection program for the Institute. This includes maintaining complete personnel radiation exposure files, implementing and main-taining an ALARA program, i.e., reducing radiation exposures to the staff and the public to levels as low as is reasonably achievable (ALARA), supervising adequate environmental monitoring programs, and ensuring that all applicable radiological safety regulations and directives of the Department of Defense, the Nuclear Regulatory Commission, and other agencies are followed. To ensure continuity in the area of radiation protection at AFRRI, particular attention is given to centering responsibility for radiological safety in experienced and highly qualified civilian personnel where possible.

7.6 NATIONAL NAVAL MEDICAL CENTER The National Naval Medical Center (NNMC) is one of the major medical research centers in the greater Washington area. The close proximity of NNMC, the National Institutes of Health, and the National Library of Medicine to AFRRI

[ ')

gives the AFRRI staff the advantage of having ready access to a well-established and extensive research community.

The National Naval Medical Center hosts several medical commands and organizations in the military. These may or may not be under the direct military command of NNMC. The separate military commands and units at NNMC are as follows:

e Armed Forces Radiobiology Research Institute (AFRRI) e National Naval Dental Center (NNDC) e Naval Health Sciences Education & Training Command (NHSETC) e Naval Medical Data Services Center (NMDSC) e Naval Medical Research & Development Command (NMRDC) e Naval Medical Research Institute (NMRI) e Naval School of Health Sciences (NSHS) e Personnel Support Detachment (PSD)

(7 U Uniformed Services University of the Health Sciences (USUHS).

e 7-3

The National Naval Medical Center itself falls under the command and coordina-tion control of the Commandant, Naval District, Washington, D.C.

AFRR' is houseo 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 recreatio6. facilities. NNMC specifically supplies to AFRRI certain

~

services such as road, parking, and grot..d maintenance; trash removal; mainten-ance of AFRRI facilities; utility services such as electricity, sewage, and water; security protection hr AFRRI's perimeter and security personnel in response to emergency security situations at AFRRI; and fire protection which includes fire 1 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.

o l

7-9

O - -.

' Ee:?j9

May 11,1972 MY ~ ' ,

NUMBER 5105.33

~ .{ AMP c -

j ASD(C)

\ w_ -J /

Department of Defense Directive

~

SU3!ECT Ar:ned Forces Radiobiology Research Institute (AFRRI)

R eis. : (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, j (hereby cancelled)

I. GENERAL l A. Established pursuant to the authority vested in the J Secretary of Defense, the Armed Forces Radiobiology l

Research Institute (AFRRI) is designated a subordinate unit of the Defense Nuclear Agency (DNA).

3. AFRRI shall serve as a radiobiology research l laboratory for the Department of Defense. Other federal and civilian agencies and institutions may ut:li=e 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.

O 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 l

l 7-10 l

c.cns ult r. tic.n w ith .h ' Doard of Gm ; rn.:rs.

Normally, the position of Director will rotate amonc the Sciviecs. His tour of duty shall be consistent with professionel and military require-ments and applicable DoD Directives and Instructions.

2 The Deputy Director will be an officer in Grade 0-o or 0-5 from one of the Military Serdces not represented by the Director.

3. The staff shall' be constituted of military and civilian personnel authorized by a Joint Table of Distribu: ion (JTD), developed by the Director, DNA, and approved by the Joint Chiefs of Staff. Insofar as possible, the milita.ry 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 rnatters by a Board of Governors.

1. The AFRRI Board of Governors shall consist of the Director, DNA, as Chairman, and the Chief Surgeons g General of the Army, Navy, and Air Force, b 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 participate in meetings of the Board of Gove:nors when matters of mutual interest are being considered.

IV. FUNC TIONS Under the direction, control and authority cf the Director, DNA, and in accordance with appropriate DoD Directives and Instructions, 1 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 biolocical effects of ionizing radiation to meet the internal requirement of the Institute, and those which may be des eloped and requested by other DoD Components.

2 7-11

May 11, 72 5105. 33 8 C. Operate facilities for the production of short-lived ra .'.cisotopes 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 A.FRRI and fulfining the functional responsibilities implicit in this Directive.

VI. ADMINISTRATION .

A.

The Military Departments will assign military personnel to AFRRIin accordance with approved authorisations.

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 AFRR1 from participating in reimbursable activities subject to concurrence by Director, DNA.

C.

The pay, allowances (including subsistenes) and permanent change of station travel costs of military personnel assigned to AFRRI will be borne by the Military Department from which assigned.

VH. EFFECTIVE DATE AND CA.NCELLATION l

This Directive is effective immediately. Reference (b) is hereby superseded and cancelled.

J

} A puty Secretary of Defense n

1 3

7-12

l 8

APPENDIX A Joint Wind Frequency Distribution for 1960-64 _

and 1964 Inclusive l

O l

O A-1

- . - .- -+

7 TABLE 1 ij Joint Wind Frequency Distribution for 1960-1964 Inclusive JOINT w!NO FWEtJUENCY Ol%Td!HUf!ON NY STAHILITY CLASS DATA PE4100t 1960 TM40 19e4 gTARILITY CLA558 pa%QUILj A AFFMi-T4[GA erACToa wie 42"" 3 ' kot7sM?"

TAeLELNkRai 10/09/80.i.s;20.02.07.

1. ie@""

c E:P '"-

DAMES ANO M000E J0d NO 11985-001-27 stNO w!NO SPEED CATEG0w!EstMETE45 PEG SECONU) MEAN SECTOR 0.0-1 5 15-30 3 0-5.0 5.0-7.5 7.5-10.0 310.0 TOTAL SPEED NNE 1 1 0 0 0 0 12 2 22

.30 3. a 0 00 0.00 0.00 0.00 3 65

.00 .g 0 00 0.00 r.00 00 .

1.52 5.47 0.00 0.00 0.00 0.00 e.99

.0{ .g 0 00 0.00 0.0 0 00 0 1 52 6.08 0.00 0.00 0.00 0.00 7.60

.0{ .05 0.0g 0.00 0.00 0 00 .g 1.22 4 6 0.00 0.00 0.00 0 00 5.47

.03 0.00 0.00 0 00 0.00 .04

.0{ '

9 *:N 8:8! 8:8l0 8:80 8:30 ':588I lo d' SE 1 52 4.46 16 0 00 0

0.00 0.00 0

0.00 0

6.38 2 17

.04 0.0g 0.00 00g 0 00 .

.0g 3 2 25

.30 3. n.00 0.00 0.00 0.00 3. S

.0g .g 0.00 0 00 0.00 0 00 .

1.52 7.29 0.00 0.00 0.00 0.00 8.61

.01 .05 0.00 0.00 0.00 0 00 .07 SSW 0 9 0 0 0 0 9 2.38 0.00 2.76 0.00 0.00 0.00 0.00 2.74 0.00 .02 0 00 0.00 0 00 0 00 .02 Sw 1 5 0 0 0 0 6 2.25

.30 1.52 0.00 0.00 0.00 0.00 1.82

.00 .01 0.00 0.00 0.0g 00g .0g 1 22 1 22 0.00 0.00 0.00 0.00 2.43 0.00 0.00 0.00 0.00 .0g

. 0 {* .0{

.30 4 6 0.00 0.00 0.00 0.00 4.k6

.00 .03 0.00 0.00 0.00 0.00 .03 15 24 0.00 0.08 0.00 0.00 3.kd

.0 .0 0.00 0.00 0 00 0 00 .03 Nw 3 10 0 0 0 0 13 2 23

31 8:88 8:8" 8:88 8:88 3:33 NNW 1.22 4 ':8'f 3 34 I

0.00 0

0.00 0.0t 0

0.00 0

4.56 15 2 11

.0{ .g 0 00 0.00 0.00 0 00 .0q 1.52 6.0A 0.00 0.00 0.00 0.00 7.60

.01 .05 0 00 0.00 0 00 0 00 .06 CALM 58 id CALM TOTAL 27:i3 110 2 0 0 0 0

":t3 329 1 80

":N **:h9 8:88 8:88 8:88 8:88 ' :M MEY xxx NUM6ER OF OCCURRENCES

!!sPkhk bbbbARkNbkaA CkShkS n

U A-2

TABLE 1 (cont'd) b k,/

e JotNT w!NO FwEQUENCv ul%TW!buTION my STAd!LITY class DATA PERICO: 1960 TH90 19 %

gTARILITY CLA%53 94500[Lge AFFdl.TWIGA WEACTOW c

wils i M 85' E7sM ? U M ITE F TABLE GENERATED: 10/09/s0. 20.02 07.

1. !P' *-

DAMES AND M00dC JOH Not 11965-001-27 wlNO w!ND SPEED CATEG0kIES PETEWS PER SECONDI MEAN '

SECT 00 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 35 41 72 0 0 0 1*e 2.Tl 1 28 1 90 2.66 0.00 0.no 0 00 5.42 Ne 28 1.61 88 3 2.78 u 0.00 8 a8 0.00 0.00 8 113 6.52 2 60 2.I3 ENE li 2 26 6

2 20 2 09 U '-

0.00 8

0.00 8 *8 0.00 til 6.52 2.37 E At 1 63 At 2.24 1 21 H 0.00 8 a8 0.00 0.00 8

13}

4.8 2 33 09 .14 .08 0.00 0 00 0.00 .30 E9E 53 50 3 ', 0 0 0 162 2 26 1 94 1.83 1 63 u.00 0.00 0.00 5.20 O SE

.12 36

.11 48 1 76

.09 1 87 51 0.00 0.00 0

0.00 0.00 0

0.00 0.00 0

.32 133 4.87 2 53 ssE  !

1.2!

1.58 11 2.24 tis 4.58 0.00 8

0.00 8

0.00 8 d8 8.39 2 80 s

2.82 9 At 3 59 si' 13.6 0.00 8

0.00 8 oa 0.00 di 20.08 3 07

.1d .22 .45 0.00 0.00 0.00 1 25

.8 .b8 0.00 0.08 008 3.b

s. 11 i? 157 1 36 a8 a8 38  !! 2 46 1.lg

.62 o-0.00 0 00 0.00 3.gg

.s. .8 53 o8 oo8 a8 0.00 102 27,

. 0.00 0.00 3.7*

2. I 3-

.0 ot iS 48 2.66 6 l o8 0.00

• "8 0.00 o8 0 00 48 3.63 2.99

.N. it

.66 2

8 2.*6

!?

0.00 8 o8 0.00 0.00 8 t'3

3. 9 3 07 l .06 .05 .15 0.00 0.00 0 00 .25 i NW 91 0 0 0 136 3 08

}9 i

l NNW

02 16

g6 2 0

3:p 8:88 0

8:88 0

8:88 0

12 3 03

.59 1.lg 2.82 0.00 0 00 0.00 4.51-

, n 81 2e b* **8 *"8 **8 138 2.en c^'" :ji 2:It :N 8:28 8:88 8:88 6:18 c^'"

4!J 4!!f TOTAL 6kk 679 'g3g 0 0 0 2fbh 2 65 "f:11 "t:li 3:ii 8:88 8:88 8:88 28:33 NEY Txx H M8Em 0F OCCUAWENCES x!!PERhk Ch80 Nk& C Shk5 n

v l

A-3 L

TABLE 1 (cont'd)

(3 U t JOINT Wlay 6wEqu(NCy Ul%1wlauflou My STamlLITY CLA5s OATA PE9!00: 1960 TMwo lume ST A4!L ITY CLAS%8 AAW) HILL L AFFWI.TwlGA scACTod DATA souwCE: NATIONAL AlppQQf. .ASM., 0.C WTM4UA. ad.

w!No SENSO# MFIGNT! 7.no wFTFws AFF*l -

TARLE GENEwAf(Ds 1n/09/=o. /6.02 07 DA"tS ANO "00df JO4 Nut ll4MS-001-27 wlND w!NO % PLEO CAftuoplESintTFws E4 %FCONU) WEAN SECT 00 00-15 1 5-3.0 1.0-3.0 5.n-7.5 7.%-10.0 >10.0 TOTAL SvtED NNE 12 14 104 IM n n 192 J.73

.23 7* /.0/ .74 0.00 0.00 3.11

.nl .0w- .de 09 n.ou n.00 .*4 NE 1J 72 100 15 1 0 201 1.30

.25 1.40 1.44 29 07 0.an 1.90

.03 .16 .24 03 .no 4.00 .=e ENE 17 68 43 /6 o 0 19* 3 26

.33 1 17 1.nl .;n 0.00 0.00 1.77

.04 .16 .tw .06 0.00 0.00 4.

E lA 44 60 IS 0 0 132 3 25

, .31 .85 1.!? .79 n.00 0200 2.Ac

.0* .10 .14 . n .1 0.no 0.00 . 11 E%E 7 35 4. 1 0 0 v7 3 37

.A5 n.o" d

p) r\s SE

.14

.o2 24 47

.63

.nM Al 41

.10 1 11 S/

./

.0

.17 9

4.00

.o*

/

0.00 0.00 0.nn 0

1.MM

./2 til 7.19 3.24

.05 .ns 02 00 0.00 .26 SSE 7 67 l*11n  % 0 0 24M 3 65

.14 1.30 2. 77 .a6 0.00 0.00 4.42

.02 1% . 12 .04 0.00 0.00 .57 .

S 43 210 105 752 12 0 1273 3 98

.n3 6.09 13.71 4.uw .2) 0.no 23.fd

.10 44 1.61 40 .05 0 00  ?.7w S$w 2/ 77 20^ 74 IJ I 393 3.99 43 1.50 *.00 1.44 .25 .02 7.63

.n5 14 ..? .17 .03 .no 40 SW 25 *5 127 40 3 1 241 3.o5 49 .97 P.47 7H .na 02 *.94

.06 .In 29 .ng .at .On s3 w9w 23 17 115 61 0 0 Ela 3.75

.*5 72 /.23 .o1 n.On 0.00 4.73

.05 .oA .26 .10 0.n0 0.nn .io j w 1 37 40 1 2 279 4.10

.2 77 2.111 *.27 19 95 .16 .04

.0 .n4 .24 .11 07 .00 .40 wNw 7 10 14% M7 to 0 3 4 4.39 f

.14 .SA 3.59 1.e4 .19 0.no S. 9

.02 .07 47 .>0 0/ 0.00 . 3 Nw 11 45 /75 139 /7 3 est 4.63

.21 .47 4.39 2.70 .*2 .06 3.76

.03 .10 .57 . 12 06 .01 1.nJ NNw 12 14 260 99 11 I =21 4 27

.23 74 5 05 1.92 ./l .02 4.17

.03 69 .59 ./1 .03 .00 9%

N 11 SS 217 91 5 0 394 4.07

.21 1.07 6 .53 40 1.77 .10 0 00 7.6%

.03 .13 ./1 .01 0.00 90 CALM 90 90 CALM 1 75 1 75

. 1 .21 TOTAL 3 919 7734 InP2 92 o S150 1.d8

6. 17.86 5J.53 19.** 1 79 .16 100.00

.8 2 10 9.29 2. 13 .Pl 62 11 75 mEy axx NuMHED OF OCCHWwtNCEs xx WERCENT OC"UwkENCts Tw15 CLASS axx PEWCCNT OCdudoENCES ALL CLASSES t

O a V

A4

TABLE 1 (cont'd) 8 JotNT e tw FwEoWNCY 91stul* orlon w ifadILITy r.LA%s nata pep!no: Iv6n T-wu 1966 STa4tLITY CLA%5: satoulLL 0 AF6wi-fulNa WCaCind Data 500WCE: Nai!ONAL alnw04f. 4asm.. 0.C dff"Esua. ao.

w!NU %tNSnp mtlGaft f.00 deTtw5 AFF s t -

TAHLE GENtwATED: 10/09/60 an.02.07. OAMtS AND M0dwE JnH NO: 1148%-001-27 wlNo *INO APEEi) CATLbO-ItStutTFus sEs (ECONun wtaN SECTna n.0-1.5 15-30 J.4-9.0 5.o-/.5 7.5-10.0 >10.0 tofat sptio NNE 2 175 13n *es la H 10*e *.d9

.21 .Al 1.71 7.im 1M .1- 5 11 10 .74 .ao 1 07 .1H .o/ 7.39 Nt 54 195 *6* 517 al 12 1323 4.70

.26 9% /.27 2.%) .e0 .nh 6.*1

.12 6% 1.ld .1% .nl 1.n<

ENE 75 21d 1.a3 52 47 77 71 ljes 4.S*

.31 1 07 2.%d 2.19 .1m 10 6.66

.17 .sn 1 26 1.n> .id 9% 3.12 E 66 153 217 1*2 70 1 621 3 51

.32 .75 1 16 .m9 .10 .nl 3.0*

.15 .3% .5* . 12 15 .01 1 42 EsE 65 123 17" 44 am 2 496 3.17 r- .32 .60 .d7 .-6 .14 .01 2 4n

/

i

' .15 .PH .*l .71 .nm .on 1.12

', SE 66 129 705 143 11 1 521 1.ed

.21 6) 1.On .35 .ns .nn 2. e

,15 .29 47 ./6 .n3 .an 1. 9 SSE 31 l*g 2m2 131 30 9 6 9 4 13

.la .6v 1 29 .65 .15 .n* 2 91

.0A .37 .6n . In .nr .n2 1.39 S 95 368 123H 1026 142 13 2442 *.56 46 1.Hn 6.06 6.07 64 .06 14 10

.22 .d4 7.41 2.14 . 32 .13 6 4s 95w 41 166 S57 669 141 79 164/ 9 17

.20 .Al e.70 3.27 41 .14 d.06

.n9 .38 1 26 1 53 .-- .06 3.76 sw 46 I?3 292 307 90 do 466 . 50

.23 .50 1 36 1.40 .** .10 *.?s

.10 .21 64 .79 ./l .6% 1,9M WSw 43 117 149 l== 1* 1* s*1 4.33

.21 .51 .97 .70 .17 .n7 2.e5

. n .27 4 'i . 33 .na .n3 1.23 w e 67 16n 231 99 24 599 5 39

. 9 .31 .6a 1 11 .-8 .'2 2.v3

.09 .19 .37 .53 .23 .qs 1 37 1NW 7 45 74% 786 *n? 141 1701 6 67

. 3 47 1 20 1.d5 1.wv 7% M.32

. 6 14 .wa ./4 91 .1* 3.14 Nw 28 76 354 157 630 299 245* 6.75

.16 .37 1.71 66 3.oe 1 02 12.00

.no .17 .nl 2.6 1 4* 4A 5.60 NNw 27 41 347 900 391 74 141* 6 22

.ll .*n 1 67 s.-n 1 41 .1x A.67

.0S .ld 78 2.09 .=4 14 4.14 N 76 17 429 al6 242 >7 1711 5.41

.37 72 2.10 3.99 1.ld .13 d.70

. l .34 .9e 1.d6 95 .06 1.46 CAtu 223 220 CALM 1 04 1 06

.50 .,o total 1034 231s 34v. 1977 2341 n/n 20**e s.20

%.06 1 [ .1/ 29.32 34.18 17.=e 3.n3 100.00 2.36 3 28 13.*a 11.09 ,.d2 1.42 46.ha MEy tru NiiwH G OF OCCoW*EaCF%

aus DtWCENT OCCoww&,NCt5 Tals CL4ss

/~,

xxx GtaCENT QCCowwtNCES ALL CLA59Et A-5

TABLE 1 (cont'd)

JotNT WING f df.'JuFNC Y 0!%TRIauf'!ON 11 Nfan!LITY CLA%%

Data otwino: 1960 Tnwu 14A*

I sTaHILITY Cla%c Jasqu[L > atF-1.fWIGa wraCTOW Oaf a %00sCE : Nal[ONAL 4 wwOWT. .aSM.. U.C a f f *t. %H a . *9. .

.INO % N90W mtlGMf 7.00 atTLws afFs[

TaALE E Nt' H a it o s 10/09/*0. an.02.07 nauts aNO MOU4t Jud NO: 11995-001-27 vlN0 m!ND cotto C4fEOvw!E%teEftJS WED % ECONO) 46. a N SECT 04 0 0-1.5 1 5-1.0 1.0-5.0 5.0-7.5 7.5-10.0 310.0 TOTAL SvELO NNE O A3 16o JT 0 0 710 3 00 0 0 99 /.70 6> 0.0u 0.00 3.

G. 0 .16 .1) 0% 0.00 0.00 .

NF 0 69 151 >H 3 0 2* 3 50 0 00 1.0M 2.67 66 0.00 0.00 3 99 0.00 .14 .16 06 0.00 0.00 .gn ENE O 76 160 26 o 0 262 3.*5 0.00 1.19  ?.20 .61 0.00 0.00 3.40 0.00 .17 .32 1% 0.00 0.00 .SS E O 51 *k

• o o liv J.In 0.00 80 .9f .ou 0.00 0.00 1 47 0.00 .12 .14 01 0.00 0.00 .2/

E%E o *? ml 11 0 0 136 3 65 0.00 .*M l.27 .20 0.00 0.00 2.16

.31 C) SF 0.00 0.00 0

.10 79 1./6

.l*

OH i To 01 l1

.<0 0.00 0.00 0

0.00 0.00 0 200 1.16 3.2%

0 00 .14 .7% .nl 4.00 0.00

• ssE o 96 136 1+ 0 0 $*17 3.*t 0 00 1.01 P.1o 10 0.00 0.00 3 61 1

0.00 15 .5 .06 0.00 0.00 .3Q S 0 232 $4 0 0 1053 3 67 0.00 3 64 10.4 7 0.00 0.00 16.57 0.00 .91 1 54 40 0.00 0.00 2. I sew 0 169 5 7 d/ 0 0 7*H 1.e7

. 20 2 14 a. 2 19 0.00 0.00 1 75 0.s ) .34 1. .a . 9 0.00 0.00 71 Sw 0 13e dal /4 0 0 343 3 20 0.00 2.*5 3.19 . la 0.00 n.00 6.02 0.00 . le .** .05 0.00 0.00 .47 w%w 0 1 31 Ids 11 0 0 267 3 09

! 0.00 2 06 1 96 .tr 0.00 0.00 6.19 O.00 . in 79 .o3 0.00 0.00 4 0 ga g3r /g o 0 gy

}61 3 57 0.00 41 2.1% . ld 0.00 0.no 1.66 0.00 .11 .11 .0% 0.00 0.00 .%0 WNW 0 %I 316 100 0 0 667 *.01 O.00 .A .96 1.57 0.00 0.00 7.J3 f 0 00 1 72 .7) 0.00 0.00 1 07

( Nw 0 $0 */* 147 0 0 6dl *.02 i 0.00 1 74 1.6* 1. 49 0.00 0.00 10.70 l 0.00 .14 1 04 .79 0.00 0.00 1.35

! NNw 0 56 681 1 0 0 62% *.05

( 0.00 .M5 7.26 1.1174 0.00 0.00 4.ds 0.00 .12 1 04 25 0.00 0 00 1 43 N O v5 115 93 0 0 $23 3.e3 0.00 1.*9 5 26 1.se 0.00 0.00 4.21 0.00 .2? .7m .21 0.00 0.00 1 19 CALu o O CAL" O.00 0.00

[ 0.00 0.00 i total 0 640 6041 9 16 0 0 6367 3 6d 0.00 2 .77 e6.10 13.11 0.00 0.00 100.00

, 0.00 .31 v.31 1.91 0.00 0.00 16.53

! MEY us* NuuMEQ OF OCCuvWP:NCES i

tax e aC NT OC uwwENCts fals Clas AXA G QC NT OC upoENCis aLL CLASS S f'"N c

l I A-6 i

t

t p TABLE 1 (cont'd)

(s / -

J0!NT m !NO F wt'Jhe.NCY 91%fQl10T!bN My $fAolLITY CLA%S DATA u t p 10f)

• 1960 TMwte 14m*

CTAHILITY CLA%% WA40d!LL > AFFwl-Twl'ia dvaCTOW DATA SOuwCF : NATIO.4AL AlkwowT. =A%a.. 0.C dFT-e904. *u. ~

WINO 9tNSOW MttbMT! 1.00 WP. T 6. = % AFFst TAHLE GENtRATEHI in/0e/do. /n.07.07. HAMts A'40 #100d t Jod NO: 11965-001-27 WINO dlND APLEO CartGO*ltStwtTEW% WEW %ECUNUI *EAN SECTOR 0.0-1 5 1 5-1.0 3.n s.0 5.0-7.5 1.5-10.0 310.0 TOTAL SPEED NNF 43 109 4* 0 0 0 238 2 10 96 1 24 .50 0.00 0.00 0.00  ?.64

.19 .2% .10 0.00 0.00 0 00 .54 NF A9 121 %4 0 u n 265 2 14 1 01 1.*0 60 n.00 0.no 0.00 3.0L

.20 .74 .1/ 0.00 0.00 0.00 .60 ENE I 1 116 4e n 0 0 771 2 06 16 1. 10 52 0.00 n.00 0.00 3.09

.5 ./m . O n.00 0.00 0.no 62 F 63 %A 1 0 0 0  ! 1 87

.72 .A* . 5 0.00 0.00 0.00 1.)2 50

.16 .13 .03 0.00 0.00 0.00 .30 E%E 96 ad 24 0 0 0 1 39 1 92 CT 7% .5% .27 0.00 0.00 0.no 1.51 i .31

\s) SE

.15 91 1.n3

.11 97 1 0s

.n4

.3s 31 0.00 n.no 0

0.0n n

0.nn 0.0n 0

n.no 21*

d.*3 1 99

.21 .21 .07 0.00 0.00 0.00 49 59E 61 79 43 0 0 0 ld) 2 19

.69 40 64 J.oo 0.00 0.00 2 09

.l* .14 .10 0.00 0.00 0 00 42 S 350 474 d64 0 0 0 1096 2.21 3 94 5.*3 3.0% 0.00 n.00 0.n0 12.*6

.80 1.n9 .ol 0.00 0.00 0.00 2.50 99d 199 453 di3 0 ft 0 895 2 36 2.15 5.1% d.4e 0.00 0.00 0.00 10.11 63 1 01 .5a 0.00 0.00 0.00 2.0*

9W 366 533 147 0 0 0 1029 2.08 3.96 6.n* 1.67 0.00 0.00 0.00 1 .66 79 1 72 . 34 0.00 0.00 0.00 .

05}S w%w 431 530 v5 0 0 0 6 1.v7 4.90 6.02 1.0H n.no n.00 0.00 12.00

.94 1.d1 .22 0.00 0.00 0.00 2.6L w 19T 216 av 0 0 0 4dd 1 99 I 2 24 2.*^ .78 n.00 0.00 0.on 5.en 65 .-9 .16 0.na n.00 0.00 1 10 wNw 1o 19e LOA 0 0 0 *22 2 24

1. 6 2.23 1.pn ', . n o n.00 0.no 6.40

.T 45 .24 '.00 0.n0 0 00 96 NW 63 190 141 0 0 0 39* 2.*T

.72 2.16 1.60 0.00 0.no 0.no 4.48 14 43 .37 0.00 0.00 0.00 40 NNW 94 207 131 0 0 a 436 2.35 1.1 2.35 l-49 0.JO 0.00 0.no 4.96

.2 .e7 in 0.00 0.00 0.00 1 00 N 18 333 M 0 0 0 670 2 23 2.15 3.7H 1.a4 0.00 0.00 0.00 7.62 43 76 .34 0.00 0.00 0.00 1 53 1Sh 10NO TOTAL 34 31s7 tei? O 0 0  !>94 t.45 3A.97 42.70 1H.32 0.00 0.00 0.00 100.00 7.83 9.57 1.63 0.00 0.00 0.00 20.0M KET axx NUMBER OF OCCodkENCE%

xxx AERCENT OCCudwENCES TMIS CLAS9

10.0 TOTAL SPEED NNE 173 397 710 510 78 8 1866 4.07

.39 .88 1 62 1.16 .13 .42 4.26 NE 205 515 950 560 42 12 224* 3 94 47 1 22 1 94 1.28 .19 .03 5 12 ENE 269 556 353 499 77 21 2275 3 82

.61 1 27 1 95 1.14 .18 .05 5 19 E 188 379 405 163 20 3 115e 3.26 43 .87 .92 .37 .05 .41 2 6*

ESE 194 321 3A6 118 28 2 1029 3.19 44 .73 .d4 .27 .06 .00 2 35

(']

\m) SE 21R 3R5 452 135 13 1 1204 3.11

.50 .88 1 03 .31 .03 .00 2.75 SSE 145 425 704 186 10 9 1499 3.*9

< .33 .97 1.61 42 .07 .02 3 42 5 570 1410 3276 1410 154 13 6833 3.81 1 30 3 22 7.48 3.22 .35 .03 15.60 SSW 276 878 1575 925 204 ?9 3787 4.02

.63 2.00 3.59 1.88 .*7 .07 8.64 Sw 450 879 79e 371 93 21 2610 3.35 1.03 2.01 1 82 .85 .21 .05 5.96 wSw 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 1634 3 80

.61 92 1 20 .69 .24 .06 3.73 wNw 177 394 919 973 417 151 3031 5.25 40 .90 2 10 2.22 45 .34 6.92 Nw 124 *27 1286 1423 657 212 4129 5 53

.28 .97 2 94 3.25 1.50 48 9.42 NNw 152 421 1271 1110 402 19 34J5 4.96

.35 96 2.90 2.53 92 18 7.84 N 312 698 1235 1000 247 27 3519 4 27

.71 1 59 2 82 2.28 .56 .06 8.03 CALM 1370 1370 CALM 3 13 3.13 TOTAL 5614 9339 15606 97a5 2643 629 41815 3 95 12.81 21 31 36 07 22 33 4.03 1.63 100.00 NouAEW OF ALIO Od5Edv4T!0Ns 43815 99.92 PCT.

NUMAER TOTAL Nupe OF NvA 10OBSEdvaT!nNS e o(r OASEdvATIONS 43844 31 .0A ACT.

100.00 DCT.

s key xxx NuMUEA 0F OCCowdLNCES l xxx PERCENT OCCueoENCES A-3

TABLE 2 Joint Wind Frequency Distribution for 1964 s i sd !L ; T Y CLAS9: P450 VILL A AFF4[-THIGA WFACT0d DATA SouwcE: NAi!ONAL AlWPORT. WASH.. 0.C GETMESDA. MO. _

w!NO 9ENSOP MEIGHT: 7.00 METFRS AFFat .

TA9tE GENERATED: 11/19/80. 20.00.32. DAMES AND M00wE J06 NO: 11985-001-27 wlNO WINO 9 PEED CATEGORIES (METERS PER SECONO) dtAN SLCTok 0.0-1 5 1 5-3.0 3.0-5.0 5.0-7.5 7.5-10.0 >10.0 TOTAL SPLEO NNE 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 NE 1 5 0 0 0 0 6 2.33 1.e4 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 4 2 07 1.6* *.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 6 2.-7 0.00 6.56 0 00 0.00 0.00 0.00 6.96 0.00 .05 0.00 0.00 0.00 0.00 .05 ESE O 4 0 0 0 0 4 2.35 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

. SE O 3 0 0 0 0 3 2 43 O.00 6.92 0.00 0.00 0.00 0.00 6.92 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 0.00 .06 0 00 0.00 0.00 0.00 .06 S 1 0 0 0 0 4 2.20 1 6* 4.4 0.00 0.00 0.00 0.00 6.56

.01 .0 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 SW 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 WSW l 1 0 0 0 0 2 2 05 1 6* 1 64 0.00 0.00 0.00 0.00 3.29

.01 .01 0.00 0.00 0.00 0.00 .02 W 0 2 0 0 0 0 2 2.35 0.00 3 28 0.00 0.00 0.00 0.00 3.29 0.00 .02 0 00 0.00 0 00 0 00 .02 wNW 1 2 0 0 0 0 3 2.07 1.64 3.28 0.00 0.00 0.00 0.00 4.92

.01 .02 0.00 0.00 0.00 0.00 .03 Nw 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 3.28 0.00 0.00 0.00 0.00 3.29 0 00 .02 0 00 .0.00 0.00 0.00 .02 N O 7 0 0 0 0 7 2.46 0.00 11 48 0.00 0.G0 0.00 0.00 11.4e 0.00 .08 0.00 0.00 0.00 0 00 .08 CALM 11 11 CALM 18.0]

• 19.0]

TOTAL 6 45 0 0 0 0 61 1 90 26.23 73.77 0.00 0.00 0.00 0.00 100.00

.18 .51 0.00 0.00 0.00 0.00 .69 MEy xxx N M4ER OF OCCURRENCES s xxx A ACENT OCCUooENCES THIS CLASS

} xxx PERCENT OCCURRENCES ALL CLASSES A-9

TABLE 2 (Cont'd) m STAHILITY CLASS 8 PASQUIL o AFFEl-TRIGA WFACTOW "t3 W" CPM "AM?"N.%"

TABLEGEkPAT

• liesS *S"*

II/l9/80. 20.00 '~.

c WW '"-

DAMES AND M00aE JOH Not 11965-001-27 d i 40 wlNO 9 PEE 0 CATEGORIES (M u 45 DER SECONO) MEAN SECT 09 0 0-1 5 1 5-3 0 3.0-3.0 2.0-7.5 7.5-10.0 310.0 TOTAL SPEED NNE 2 6 15 0 0 0 21 3.11 40 3.03 0.00 0.00 0 00 4.65

.02 1.2}

0.0g 0.00 0.00 .g 2.18 2.8 3. 4 0.00 0.00 aa 0.00 7.64 we n 2:y as c#8 ac8 u 2.,6 2:f5 p :f! 8:88 8:88 8:28 *:32 1.

2.4h 1.4 o

0.00 0.08 0.00 5.k

• eSE 4 13 3 oa8 aa8 oo8 38 2.00

/_)

Q

2. 3

.lg 2 63

.lg

.6

.g 0.00

- 0.00 0.00 0.00 0 00 0 00 6.06

.j4 1 01 1.82 2.2 0.00 0.00 0.00 5.0k

.06 3 0.00 0 00 0.00 .28

.g0 .

2.1$ 2.4b 6.6) 0.00 0.08 0.08 11.k2 S 't li 38

• 8 oo2 a8 35 3.27 SSw 5:8A 3

2:t!9 22:13 to 8:88 0

8:88 0

8:38 0

:n22 2.84

.61 1.42 2.02 0.00 0.00 0.00 4.66

.0g .lg .lg 0.00 0 00 0.00 .

1.82 40 1.62 0.00 0.00 0.00 3.d4

.I .02 .09 0.00 0.0g 00g .22 4

!i4 8:88 6

2:8I16 8:88 0

8:88 0

8:88 0

':f!26 2 99

.81 3.23 0.00 0.00 0.00 5.25

.0g 1.2}

.0 .8 0.00 0.0g 0.00 .g Nw 8:8" 3

!)3 2:] 8:88 0

8:88 0

8:88 0  :'10l 3.24 60 .41 0.00 0.00 0.00 4.04

.0g 3.0)

. 0.00 0.00 0.00 .23

.0)

.61 1.4g 2. .h 0.00 0.00 0.00 4 4

.03 .0 . 3 0.00 0 00 0.00 .2g

.ag 2. 0.00 0.00 0.00 3. 3

.g]0

. .0 . 0.00 0.00 0.00 6.46

.]

6.46 TOTAL I2 250 0 0 0 I. 5 2 67 24.6b 24!$k 50.51 0.00 0.00 0.00 100 00 1 39 1.*0 2.85 0.00 0.00 0.00 5.6*

MEy XXX N M8 R OF OCCURRENCES

v  !!!A RC OC8bARkCkk C Skk5 l

A- 10

TABLE 2 (cont'd)

/l O

staulLITY CLAst MaioulLL C AFFQ -TWIGA WEACTOW UATA SOudCE NATIONAL A]WPQWT. WASH., Q,C *H SOA. MO.

• DIN 0 '; ens 0R MFI6nT: 7.00 ME TERS 8(EQ AF TAeLE GENERATfD8 11/19/60. 20.00.32. OAME ANO wo0AE JOB N08 11985-001-27 d!NO .!NO SPEED CATEGORIE$f=ETERS PER SECONO) 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 o 9 12 8 0 0 29 3 81 0.00 .89 1.}9 .80 0.00 0.00 2.8d 0.00 . 0 .je .09 0.00 0.0g .g 0.00 .99 2.0k 40 0.00 0.00 3.44 ENE 8 21 ft i a8 a8 3:ji 3.57 8:88 :88 9

t1 :32 6

8:88 0

8:88 0

31 28 3 64 E 2 11

.89 1.09 6g 0.00 0.00 2.78

.2) 2 8 8 fi EsE 0 19 i 3.06 T 0.00 70 40 .10 0.00 0.00 1.19

[_/ 0.0g .0g .

.0j

.20 0.0g 0.0 00

.j4 2.19 97

.20 .30 12 .20

! - 3 13 8

! ao8 L5 4.06 ssE 0.00 1.I 2.88 1.k9 0.00 0.00 6.)6 s '8 la 133 .2g 0,0g 0.0g 33 ,,,,

.20 3 48 e.76 0.00 0.0g 2306 ss.

.30

$ '8 1.h9 19.6l 5

5.67 I'

1.69 7 o8 0.00 a

0.00 c

9.24 S3 3. 2

.0g . 2 0, .

j .19 0.00 0 00 1.g6

.20 1.f9 3. 8 .70 0.00 0.00 5.2

.02 .14 .6 .08 0.00 0.00 .60 4.06 wSw 2 7 3 10 0 0 42

.20 .70 2. 9 99 0.00 0.00 4.17

.02 .08 6 .11 0.00 0 00 4e w I 7 29 23 0 0 59 *.34

.go .70 2 78 0.00 0.00 5.86

.~.

0.00 a ot

.90 3!

3.38 2.I9 1 19 l o3.20 oo8 0.00 5.21 ti . 39 0.00 .06 .34 .14 .02 0.00 .60 Nw is 6 38 26 2 0 72 4.71 i

0.00 .60 3.79 2.58 .20 0.00 7.le N~. o3

.20 9

.50 43 5.27 38

1. 9 of

.20 o9 0.00 f

8.0k 4.34

.02 .60 .02 0 00 .

.0g .2{

.1 hh 1.f9 g00 0.00 7.55

.g

.8

.0g 4 57

. .19 00 0 00 .y 2.68 2.64 TOTAL 165 545 244 8 0 1006 3.97 4.37 16.*0 54 17 26.25 80 0.00 100.00

.50 1.88 6.21 2.78 .09 0 00 11 45 MEy 4XX R OF OCC09RENCES p XXX P A NT NifM NTOCCURRgNCESALLOCCuowNCESTHISCLASgS xXX P.R CLASS A-li

TABLE 2 (cont'd) o U

Arr g T At41L I T Y { Lass pas 30lt 0 RF. ACTOR

'" ' c '**- '*-

die TABLEqN""8u'bOnM?" 1.W"lia5 Hida.TRIGA -

E9Af 11/19/e0. 20.00.32. DAMES AND M00WE JO8 NO: 11985-001-27 w!NO WIND SPEED CATEGORIES METERS RER SECONDI MFAN SECTod 0.0-1 5 15-10 3.0-5.0 5.0-7.5 7.5-10.0 >10.0 TOTAL SPEED NNE 7 20 54 117 14 0 212 5.13

. 47 1.26 .33 0 00 4.94 g6 2.7]

~E

.09 8

j80 104 2.33

'ik 17 12

.61 0.00 8 231 5

6.89 5.06

.05 1.14 3

1. 55 .30 0 00 3 37

.}4 I *

.0 2fkI fk3 .hk .0

.03 6

.6b

.3 5

1.21 7

f.29 46

.16 4

.05 0

3.li 152 4.12 E

E9E

li6

)

':2g 2b

':Pl$ :83 6

8:28 0

?:13 64 4.07

['N .14 40 .5 .30 .14 0.00 1.49 O se *l

.09 . I*t i.00 0 .

s l! *l

.09 0.00

• 8 2.ie la 4.ne

• 28 5 SsE

- 5 t! a8 i!! 3.88

A3 -}8

3 ':29  :!! :8$ 8:28 f,!!

5 8 7 30 303 28 0 715 4.71

.g9 1.7 7,0 7.06 .65 0.00 16.66 55w 5

1 I8 jak *$h *b I. 8 5.10

.12 .77 4.22 4.33 .91 .n9 10 . a.4

.06 .38 .05 g g

.09 .13 2.$6

1. 7 2.gl 1.9 .g*9 4 .i4 SI0 4

.05 .15 6

.67 97 .20 9

.07 9

2.1J99 5.13 wSe 3 37 5

.07 7 .86 8

• .2g {.3g w 9 -

1 il -

g .2}

- 2 4 iL 5.17

13 :t85  :*3 :18 8 :85 f:13 wNw 3 34 103  :'b6 7 188 6.23

.07 2 79 2.40 .84 16 4.38

~w

- ' .g?

H 217 13 22 2d' 7.12

.0 .16 o*

.86 58k 3. 0 h' 1.02 10.9 NN.

3} .Al i.j4i i 28%

3 4.73 t

2!16 58

.29 5:3! 6.79

.0 i03i N n

.0c .J>

i .J6 2m >e i.23 42 4e)

,.9, c,, .gg 23 .72 4.Al

2. .e0 .2A

.i 3.

7.,gg c,,

t:19 211 3 35 1 64 8Ib 2k 14.61

• b kh 21.25 Ik.kh84 2k 1 42 100 0 48.88

• .12

/% lxx bhht Ochu kHISCLAS xxx PERCENT OCCURRENCES ALL CLASS 5

'% )

A-12

IABLE 2 (Cont'd)

Star $1L i T v CL Ass: pa%u!LL t AFFwl-Tw!GA WP: AC TOW UATA 50uwCE: NATIONAL A!WPOWT. WASM.. D.C BETHE90A. wo.

stND iENSOR HF. ! GM T : 7.00 wETEWS AFFwl -

TAdLE GENERATE 0 11/19/80. 20.00.32. DAMES AND MOORE JOB NO: 11985-001-27 wtNO WIND SPEED CATEG0p!EstwETERS PER SECON01 wFAN SECTOW 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 O 9 30 9 0 0 46 3.80 0.00 .69 29 .69 0.00 0.00 3.66 w

0.00 8 's

.23 2.h 2.06

't

.38 o8 0.00 a8 0.00 H

2.67 3 90 0.00 .03 .31 .06 0.00 0.00 40 ENE O 13 27 3 0 0 43 3.31 0.00 99 2.06 .23 0.00 0 00 3.28 0.00 .15 .31 .03 0.00 0.00 49 E o 14 .5 1 0 0 30 3.12 0.00 1.07 1. ,4 .08 0.00 0.00 2. 9 0.00 .16 . 7 .0 0.00 0.00 . 4 ESE O 9 ,3 0 0 25 3.36 e -N 0.00 .69 99 .23 0.00 0.00 1.91 I  ! J.00 . 0 9

. .03 0.00 0.00 0

.2$

1~/ SE o 4 0

• 3.29 0.00 1. 5 1. .31 0.00 0.00 3.*2 3 0.00 .22 .25 .05 0.00 0.00 .51 SSE O 6 39 2 0 0 57 3.*6 0 00 1. 2 2.97 .15 0.00 0.00 4.3 0.00 .18 .-4 .0 0.00 0.00 .6 5 0 31 164 4 0 0 242 3.93 0.00 2.36 12.51 3.59 0.00 0.00 1A.=6 0.00 .35 1.87 .54 0.00 0.00 2.76 SSw 0 29 147 29 0 0 204 3.83 0.00 2.21 1 .21 2.14 0.00 0.00 15.56 0.00 .33 .67 .32 0.00 0.00 2.32 sw 0 25 41 7 0 0 /3 3 47 0.00 1.91 3.13 .53 0.00 0.00 5.57 0.00 .28 41 .08 0.00 0 00 .63 wSw 0 23 21 3 0 0 47 3.12 0.00 1 75 1.60 .23 0.00 0.00 3.s*

59 0 00 .26 .24 .03 0.00 0.00 w 0 12 22 4 0 0 38 3.52 0 00 42 1.68 .31 0.00 0.00 2.90 0.00 .14 .25 .05 0.00 0 00 .*3

• Nw 0 2 41 20 0 0 63 4.23 0.00 15 3.13 1.53 0.00 0.00 4.8 0.00 .02 47 .23 0.00 0.00 7 Nw 0 6 80 29 0 0 11, 4 20 0.00 46 6 10 2.21 0.00 0.00 9.77 0.00 .07 91 . 3 0 00 0.00 1. 1 NNW 0 8 107 6 0 0 11 4.22 0.00 .61 8.16 2. 5 0.00 0.00 11. 2 0.00 .09 1.22 41 0 00 0.00 1.72 N 0 12 57 26 0 0 95 4 04 0.00 92 4.35 1.98 0.00 0.00 7.25 0.00 .l* .65 .30 0.00 0.00 1.08 CALM 0 0 CALM 0 00 0.00 0.00 0 00 TO T At. 0 231 853 227 0 0 1311 3.d3

?.00 17.62 65.06 17.32 0.00 0.00 100.00 0.00 2.63 9.71 2.58 0.00 0 00 14.93 mEy Exx N R OF OCCUAAENCES

<^ xxx P NT OCCUAA 5 THIS CLAS 4xx PERC NT OCCUAR S ALL CLASS S A-13

TABLE 2 (Cont'd)

/,

%TLo!LITY ClaS; RaSOUll F AFFR REACTop c ' " -

Wd-TRIGA -

2te 42rsS'Me'bsMi"

TABLEGENkRATfs 11/19/80. 20.00.32. N.% "hEniS" -

0AMES AND =00RE JOB NO: 11985-001-27 WINO WIND $ PEED CATEGORIESIMETERS PER SECONO) MEAN SECTod 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 8 11 10 0 0 0 29 2 31 49 .6o .62 0.00 0.00 0.00

.0j .gg .lg 0.00 0.00 0.00

1. ]79 43 1.05 49 0.00 0.00 0.00 1.91 E,

p: n p a: oo8 oe8 a ,. 0

i :38 :11 8:88 8:88 8:28 ':22 E 6 21 1 0 0 0 28 2.22

.37 .06 0.00 0.00 0.00 1.73 9 130e oA oa8 oo8 oa8 33 2 37 EsE

[N 43 .49 49 0.00 0.00 0.00 1.42

.08 .09

.p9 0 00 0.00 0.00 .g

.25 1. .6 0.00 0.00 0.00 2.41

.05 .28 .

g. 0.00 0 00 0.00 .g 2.68

.00 0.00 0.00 7 0.g0 s

23 12 3.65 At 8 "- 8 a8 0.00 1.g4 1

4.65 2.55 1 42 4.58 0.00 0.00 ss.

1 61 18 5.57 u 3.59 ti oa8 0.00 0.00 8 a8 0.00

't?!

10.76 2.50 1 98

.g 1 02 .66 0.00 0.00 0.00 0.08 13kh

3. 5 8. k 2.38 0.00 0.00

.8 1 43 0.00 0.00 0.00 2.52 4.58 7.92 0.00 C.00 0.00 13bb

.84 1.46 56 1.f.8 3

. 0.00 0

0.00 0

0.00 0

2.53 102 2.21 w 28 73 3.46 1. .! 0.00 0.00 0.00 6. 1

1. 32 .64 .do 0.00 0.00 0.00 1. 6 3 2 58 wNw 7 26 20 0 0 0 43 1 61 1.24 0.00 0.00 0.00 3.29

.08 .g . 0.00 0.00 0.00 .

3.46 56 2.57

. 1.24 1.61 0.00 0.00 0.01 64 0.00 0.00 0.00

.14 .g3 .

2.54

.56 2.4h 1.4]2 0.00 0.00 0.00 4.9k 0, . .5 0.00 0 00 00g .jk 2 54

':38 5D CALM }! I 2:28 8:88 8:88 8:88

}41

't:?2 CALM

';$28 ?2 TOTAL 723 366 0 0 0 1617 2 02 32.65 4*.71 22.63 0.00 0.00 0.00 100.00 6.01 8.23 4.17 0.00 0.00 0.00 18.41 NEy xxx N g R OF OCCueRENCES O  !!!PERCE OCCubkNCks Cb55kS (O

A- n

TABLE 2 (Cont'd)

V ALL WIN 05 AFFRI-TRIGA REACTOR s OATA 900RCEI NATIONAL AIRPORT. WASH., D.C 8ETHESDA. wo.

w!ND sENSQR H Gr1T 7.00 METEPS AFFRI -

TABLE GENtRAT 11/19/80. 20.00.32. DAMES AND MOORE JOH NO 11985-001-27 djNO WIND SPEED CATEG091EstMETERS PER SECON01 MEAN StCTOR 0.0-1 5 1 5-1.0 3.0-5.0 5.0-7.5 7.5-10.0 >10.0 TOTAL SPEED 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 *42 4.41

.25 .85 1.98 1.65 .30 0 00 5.03 ENE 27 84 172 121 12 4 420 4.13

.31 .96 1.96 1.38 .14 .05 4.78 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 fG .31 .66 .57 .19 .07 0.00 1.80 L 7 99 3 0 226 3 50 SE

.h .8h 1 13 .3k .0h 0.00 2.57 SSE 23 91 186 53 2 0 355 3.57

.26 1.04 2.12 .60 .02 e.so 4.04 5 41 230 716 418 28 0 1433 4 16 47 2.62 8.15 4.76 .32 0 00 16.31 SSW 37 179 451 231 39 4 941 4.16 42 2.04 5.13 2.63 44 .05 10.71 SW 66 184 17A 99 18 6 551 3 60

.75 2.09 2.03 1.13 .20 .07 6.27 wsW 81 175 109 38 9 9 421 3 15

.92 1.99 1.24 43 .10 10 4.79 d 40 100 122 78 17 4 361 3.85 46 1.14 1.39 .89 .19 .05 4.11 WNW 11 46 140 135 38 7 377 4.96

.13 .52 1 59 1.54 43 .08 4.29 NW 14 42 196 306 133 44 775 5.97

.16 48 2 23 3.48 1 51 .50 8.37 NNW 17 73 267 258 116 38 769 5 47

.19 .83 3.04 2.94 1.32 .43 8.75 N 17 93 212 220 53 9 604 . 81

.19 1.06 2.41 2.50 .60 .10 6.88 CALM 379 379 CALM 4.31 4.31 TOTAL 854 1649 3298 2337 ' l '. 125 8784 4.18 9.72 18.77 37.55 26.61 *

• i 1.42 100.00 NUMB R 0 ALIO 085C VATION 8784 100.00 M

• NU'4H R 0 NVA 10 08 ERVATI NS 0 0.09 **

TOTAL NUM8 R 0 085E VA!!ONS 8784 100.'J C s/ PEY xxx N R 0F OCCtjRRENCES Axx P NT OCCURRENCES A-15

O APPENDIX B Annual Averaged Relative Concentrations (X/Q)

O ,

4 O

B-1

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y _

8 APPENDIX C ANALYSIS OF LOSS-OF-COOLANT ACCIDENT FOR AFRRI-TRIGA REACTOR O

1 O

C-1

APPENDIX C D

c k> 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 tne 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

6. Distance from top of core 13 top of reactor tank = 505 cm 3

7. Distance from reactor room floor level to ceiling = 550 cm

8. Reactor tank diameter = 366 cm 3

9. Ceiling thickness = 10 cm of ordinary concrete

10. Wall thickness = 25 cm of ordinary concrete 3 i

11. Total core activity = 4 x 10 6 Curies "

I ,6

12. Total gamma power = 1.75 x 10 Mev/sec

13. Density of fuel element = 7.9 gm/cm

14. Volume of fuel elements = 2.5 x 10" cm3 ,

O C-2

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 cylinder.

D = Sv Y {-E 2ID s*}

2pK_ _

s where:

D = source gamma dose rate (R/hr)

Sv = volume source strength (Mev/sec-cm3) p attenuation coefficient for source (cm-I)

=

s K = flux to dose rate conversion factor (Mev/cm2-sec per R/hr) t = thic! of source (cm) .

E 2

= exponential function.

The source strength (Sv) for the reactor core is determined to be 7 x 10 12 Mev/sec-cm3. 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 s

d flux to dose conversion factor (K) is 5.2 x 10 5 Mev/cm2 -sec per R/b , he thickness of the source is 87 cm. The calculated gamma dose rate (D) at the top 7

surface (1 cm) of the core is 1.5 x 10 R/hr. The measured gamma dose rate at the surface of TRIGA fuel operated at 250 Kwt af ter 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> decay is 10 3 R/hr.2 5nce the dose reduction factor for I hour decay is 4 5 ,6, and the power level is a factor of 4 lower, the expected surface dose rate of an AFRRI fuel element weuld be 1.6 x 10" R/hr. For an 85 fuel element core, the maximum expected gamma dose 6

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 meast. red 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 scstter must be determined.

p From Reference 8, the reduction factors for a narrow beam at various distances O for i Mev photons are 1.4 x 10-8 at 10 feet,3.8 x 10-9 at 30 feet, and 1.5 x 10-' at C-3

(O g 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:

Dg = (D )(RF)(A) where:

DA = air-scattered gamma dose rate (R/hr)

D = source gamma dose rate at I cm (R/hr)

RF = reduction factor for air scatter A = solid angle of escaping beam (steradian).

With a source gamma dose rate of 1.5 x 107 R/hr, the air scatter gamma dose rates at the rector room floor level are calculated to be 3 x 10-I R/hr at 10 feet, 8 x 10-2 J/hr at 30 feet, and 3.2 x 10-2 R/hr at C0 feet from the reactor tank.

The scattering from the ceiling can be determined by using the method given in Reference 8 for an extended slab. The i'linwing equatior' can be used to d determine the contribution from the ceiling scata. -f ie escap.ag beam (reactor core):

Ds = (D )(r 3 -2)(r;-2)(RF,)

where:

Ds = slab scattered gamma dose rate (R/hr)

D = source gamma dose rate at I cm (R/hr)

-2 inverse square factor for incidence beam r; =

r = inverse square factor for scattered beam s

RF 2

= reflection factor for extended slab.

The inverse square factor for the incidence beam is 9 x 10-7 (1055 cm) and for the scattered beam is 3.3 x 10-6 (550 cm). The reflection factor at normal 7

incidence is 7.8 x 10-". With a source gamma c'a a rate of 1.5 x 10 R/hr, the ceiling scattered gamma dose rate is negligible (3.5 x 10-7 R/hr) and would not

., contribute to the gamma dose rate to an individual in the reactor room or outside

~ the AFRRI facility.

C-4

The gamma dose rate on the roof of the AFRRI facility over the reactor room can be determined by the following equation:8 Dd = D Yi where:

Dd = direct gamma dose rate (R/hr)

D = source gamma dose rate (R/hr)

Y

-2 r; = inverse square factor for inc'.o'ance beam p = attenuation coefficient for snield (cm-I) t = thickness of shield (cm).

For the ordinary concrete ceiling, the thickness is 10 cm and the s ;tenuation coefficie t for 1 Mev photons is 0.145 cm -I The inverse square factor for the 7

incidence beam is 9 x 10-7 With a source gamma dose rate of 1.5 x 10

. R/hr,the 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 gamn:a dose rate of 3.2 R/hr should be reduced by at least the inverse square factor of the scattered beam, r which would be 3.3 x 10 (550 cm).

y To determine the maximum gamma dose rate to an individual outside the AFRRI facility, the calculated air-scatter gamma dose rate at 10 feet (nearest outaide 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, th gamma dose rate would be lese than 1 mR/hr.

All of the calculated gamma dose rates assume a steady state condition of 1 MW operation (at least 1 year operation) which probably t esults in a source term that is conservative by a factor of 10. These calculated gamma dose rates would be 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 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> would be 0.35 of the calculated hourly dose and the integrated gamma dose for i day would be only 3 times the calculated hourly dose.

l C-5 l

REFERENCES

1. Amendment #17 to the Technical Specifications, Facility License No. R-84 for AFRRI-TRIG A Mark-F Reactor, Docket No. 50-170.

2. Personal Communications with AFRR1 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, J. 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, g E., and Fe. nam, G. E., The Activity of the Fission Products of U , APEX-443 (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-503 (June 1955).

8. Wende, C. W. J., The Computation of Radiation Hazards, AECD-3661, January 11,1944.

L O-

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TABLE 1 GASEOUS FISSION PRODUCT INVENTORY 1 MW STEADY STATE OPERATION Saturation Gamma Sogree Half-Life *7 Fission Yield 7 Activity Strength Isotope (Th) (%) (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 m 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 i

O- 135 m 135 15.6 m 9.2 h 1.8 6.3 1.6 E+4 5.5 E+4 3.1 E+14 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+13

, lodine (I) 131 8.05 d 2.91 2.5 E+4 3.4 E+14 132 2.28 h 4.26 3.7 E+4 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.

O C-7

TABLE 2 GASEOUS FISSION PRODUCT INVENTORY FOR 40 MW-SEC PULSE OPERATION 7 One-Minute gecay Gamma Sogree Half-Life *7 Fission Yield Activity ' Strength Isotope (Th) (%) (Ci) (Mev/Sec)

Krypton (Kr) 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 2.8 h 91.0 6.0 E+12 88 3.56

-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

O 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 n1 7.80 165.0 1.2 E+13 1 135 6.7 h 6.17 57.0 3.7 E+12 TOTAL 24.92 2.3 E+2 1.6 E+13 r

G ASEOUS TOTAL 56.9 1.0 E+4 4.2 E+14 l

d = day, h = hour, m = minute.

i O

e

! C-8

,av. , , . . . . , . . . . , - n--- , - - , - ,- - . , -

, , . - .. , .r - ., - . , . . . ,. , . - - , , . . . _ . . _, ,. . , . . - . . . .,

TABLE 3 DIFFUSION FACTOR AND FINITE CLOUD CORRECTION FACTOR USED FOR DOSE CALCULATIONS 9

Distance X/Q (na (sec/m3) F

• 10 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-3 8.0 E-2 200 2.2 E-3 1.0 E-1 m

(v]

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.

m)

C-9

TABLE 4 i

GASEOUS FISSION PRODUCT INVENTORY FOR 1 MW STEADY STATE OPERATION WITH 2-WEEK DECAY f

i Gamma Source Thyroid Source Activity Strength Strength Isotope (Ci) (Mev/sec) (rad)

Krypton (Kr) -

4 85 2.4 E+3 1.8 E+11 --

. Xenon (Xe) 131 m 67.0 3.9 E+11 --

133 ni 21.0 1.8 E+10 --

133 9.2 E+3 2.9 E+13 --

TOTAL 9.3 E+3 2.9 E+13 --

Iodir.e (1) 131 7.5 E+3 1.0 E+14 1.1 E+10 i 132 5.2 E+2 4.5 E+13 2.8 E+7 i 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 O

C- 10 i

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

TABLE 5 CALCULATED DOSES TO INDIVIDUALS STANDING DOVNWIND OF AFRR1 FACILITY FOLLOWING FUEL ELEMENT DROF ACCIDENT 4

j Whole Body Thyroid

Distance Dose Dose I. _

(m) (mrads) (mrads) 25 0.0004 45 50 -* 12 1

75 -- 6 100 -- 3 150 -- 2 i

j 200 -- 1 4

4

.O 1

i a

N Insignificant.

i l

O C-11

TABLE 6 GASEOUS FISSION PRODUCT INVENTORY RELEASED TO ATMOSPHERE FOLLOWING FUEL ELEMENT CLADDING FAILURE ACCIDENT 4

i r

Gamma Source Thyroid Source

, Activity Strength Strength

! Isotope (Ci) (Mev/sec) (rad) i 1.4 8.6 E+10

Krypton --

Xenon 2.9 4.5 E+10 --

lodine 0.06 3.1 E+8 1.7 E+3 ,

i TOTAL 4.4 1.3 E+11 1.7 E+3 i

O i

l l-C-12 l

TABLE 7 CALCULATED DOSES TO INDIVIDUALS STANDING DOWNWIND OF AFRRI FACILITY FOLLOWING FUEL ELEMENT CLADDING FAILURE ACCIDENT Whole Body Thyroid Distance Dose Dose (m) (mrads) (mrads) 25 1.3 57 50 0.6 16 75 0.5 7 100 0.3 5 150 0.2 3 200 0.15 2 O

1 C-13 l