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Environ Assessment for Mcclellan Air Force Stationary Radiography Facility
	ML20212C806
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Site:	University of California-Davis
Issue date:	01/31/1985
From:	; AIR FORCE, DEPT. OF
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ML20212C798	List: ML20212C795; ML20212C806; ML20212C819;
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	NUDOCS 9710300190
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{{#Wiki_filter:1 1 t i , ENVIRONMENTAL ASSESSMENT FOR THE McCLELLAN AIR FORCE STATIONARY RADIOGRAPHY FACILITY l -i JANUARY 1985 9710300190 971007 e {DR ADOCK 05000607 PDR

21 TABLE OF CONTENTS Pag EXECUTIVE.

SUMMARY

SECTION A -' DESCRIPTION OF PROPOSED ACTION AND ALTERNATIVES 1. PROPOSED ACTION... I 2. DESCRIPTION OF-THE PROPOSED SYSTEM 1 - 2.1 General-I 2.2 SNRS I 2.2.1 Enclosure 1 2.2,2 Reactor 2 2.2.3 Cooling System. 3 2.2.3.1 P r i ma ry Sys tem............ 3 2.2.3.2 Secondary System........... 4 2.2.4-Confinement and Emergency Ventilation 4 i Features 2.3 Radiography Bay and Equipment 5-3. DESIGN ENVELOPE 6- -3.1 Facility Design 6 3.2 Environmental and Personnel Safety............ 7 SECTION B - ENVIRONMENTAL CONSEQUENCES 1. SITE DESCRIPTION 9 1.1 SNRS Site Boundaries and Associated' Dose Limits 11 2. ENVIRONMENTAL EFFECT OF REACTOR AND ENCLOSURE 12 INSTALLATION 3. ENVIRONMENTAL EFFECT OF OPERATION,.............. 12 3.1 Thermal Impact... 12 3.2 Radiological. Impact 13 3.2.1 Normal Operation................ 13 3.2.1.1 Solid Waste.............. 16

TABLE OF CONTENTS (contd). .Pa21 3.2.1.2 Liquid Waste'...........-.. 18

3. 2.1. 3 - Other. Waste 18 3.2.1.4' Monitoring..............

19 -3.2.1.5 Spent Fuel..........,... 19 3.2.2 Abnormal Operations 20 3.2.2.1 One-Element Accident..-. 20 =(Maximum Credible Accident) 3.2.2.2 Multiple Element Failures 21 3.2.2.3 Other Accidents............ 23 (Less Severe Accidents) 3.3 Other Impacts (Non-Reactor) 24 4. RESOURCE USE 25 5. LONG-TERM ASPECTS'....................... 25 6. COST-BENEFIT 25 . 7. CONCLUSIONS 25

SECTION C - 0FFICES, AGENCIES, AND PERSONS-CONTACTED........

27 1 -: APPENDIX A - RADIOLOGICAL SOURCE TERMS (Normal Operations) -1.0 Introduction 2.0 Airborne Concentration of Ar-41 In SNRS Enclosure due to Normal Radiography Operations 2.1 Assumptions 2.2 Ar-41 Calculation 3.0 Release of Ar-41 from Reactor Water 4.0 ' Total Ar-41 Release-during Normal Operations 5.0- Nitrogen-16 Activity in the SNRS Enclosure m

4 1 - APPENDIX'8

ANALYSIS OF EXPECTED RADIOLOGICAL DOSES FROM NORMAL-AND ABNORMAL OPERATION 1.0- Introduction

2. 0 : _ Normal Operations

- 3,0 Abnormal Operations'(Accidents) 3.1 -One-Element Accident (Maximum Credible Accident)- APPENDIX C - LIST OF PREPARERS REFERENCES-1 i 1

EXECUTIVE

SUMMARY

This Environmental Assessment was done for the proposed Stationary Neutron Radiography System (SNRS) at McClellan Air Force Base, Sacramento, CA. The system will be operated for the Directorate of Maintenance. The system to be installed will be a small, below grade research reactor that will be operated at a maximum power of 1 MW. The reactor will be cooled by light water, and will contain fuel elements enriched to 20% in Jranium-235, mixed with zirconium hydride. There are over 50 reactors of this type operating world-wide with excellent performance and safety records. Tne reactor will support four beam tubes to perform the necessary neutron radiography operations. Each beam tube will consist of two helium-filled sections. The first section will be in the reactor tank; the second section will be outside the reactor tank and extend into the radiography bay. The beam tubes will not penetrate the tank wall, thereby eliminating the need to disrupt the tank integrity. The radiography bays will be equipped with manipulators, real-time imaging systems, and operator control booths. The doors to the radiography bays will be sized for the largest I' collimator components, part-manipulator components, or aircraft components to enter-the area. The SNRS enclosure will be designed to confine any radioactive material released in an accident so that it may be exhausted in a controlled manner through an emergency cleanup and exhaust system, which can filter out particulate and radioactive materials before the air is exhausted to the ~ environment. The environmental effects of construction will be the same as those associated with other small construction projects and will not adversely affect the surrounding developed area or site resources to any measurable degree. The thermal releases from reactor operation will be comparable to those of an air-conditioning unit for a small office building. Heat dis-sipation will be to the atmosphere by means of a cooling tower. __J

The radiological impact has been evaluated for normal operations and abnormal operations (maximum credible accident condition). For normal operations, the effects of solid, liquid, and gaseous effluents have been addressed. Solid waste typically consists of ion-exchange resini, water partic-ulate filters from the reactor water systems, and various cleaning materials. The volume of solid waste materials per year is estimated to be of the order of five regular 55 gal (208-L) drums. These drums will be shipped once a year to a licensed disposal site in accordance with DOE /NRC/00T regulations. McClellan Air Force Base has agreements in place to handle solid waste. During normal operations the SNRS will generate no radioactive liquid waste. Any suspect liquid waste that is produced will be stored in storage tanks and sampled for possible radioactivity. Should any radioactivity be detected, it will be recycled back into the reactor water system or solidi-fied and disposed of as solid waste. Gaseous waste will consist of two main sources: The first is nitrogen-16 (N-16), produced by the interaction of neutrons with the oxygen in water. The analysis in Appendix A shows that the potential dose rate to SNRS personnel is very low. Because of the short half-life of N-16 (7 s), it presents no impact to the environment beyond the SNRS enclosure. The second source is argon-41 (Ar-41), produced by neutron interaction with the argon in air. The analysis in Appendix A shows that the total Ar-41 released to the unrestricted area (i.e., area where the general public or nonreactor workers may be located) will be below 10 CFR Part 20 limits. Appendix B shows that the reactor-related dose rate from the Ar-41 to reactor workers will be less than 50 mrem /yr and to the general public less than 0.1 mrem /yr (calculated at the closest site boundary). For comparison, the individual dose rate from natural sources of radioactivity is about 100 mrem /yr.

-Since.the siting criteria in-the American Nuclear Society Standard for Research: Reactor _ Site Evaluation. (ANS 15.7) are~ more conservative-than NRC siting regulations and more applicable to this proposed project, the ANS-15.7 Standard criteria w~ere used to evaluate calculated doses to the on-and off-site population. For abnormal operations, the typical TRIGA design bn.is accident (max-imum credible accident) was analyzed. This is the_ fuel-nandling accident, which assumes the loss of cladding in air of a single fuel element. This accident assumes continuous 1-MW' operation for one year, therefore giving a very conservatively large ,sion product inventory.- The model employed _in predicting-the atmospheric dispersion of radionuclides from the accident is 4 the PAVAN computer code. This model is used regularly by the U.S. Nuclear Regulatory _ Commission to estimate down-wind ground-level air concentrations for potential accidental releases of. radioactive material from nuclear power and research reactors. In the event of accidental release from one fuel element in air, (maximum credible accident), the maximum dose to the general public would be- 0.002 mrem to the whole body and 0.18 mrem _ to the thyroid. Even in the noncredible case of all fuel elements being destroyed,-the maximum dose to the general public would be 0.30 mrem to the whole body and 27 mrem to the thyroid,- Both of these dose levels are far below the criteria -in ANS 15.7 of 500 mrem whole body and 1500 mrem to 'other organs. In conclusion, the nonradiological impacts would be no greater than those normally associated with construction of small industrial facilities. During normal operation, exposures of workers and the-general public would be far below the 10 CFR Part 20 limits. For abnormal operations in not only the single element accident but the noncredible case of all fuel elements being destroyed, the maximum dose to the general public would be far below-the criteria in ANS 15.7, and even far below the 10 CFR Part 20 limits for normal operations.

~_. fhe NRC has endorsed a generic statement for research rea: tors below 2 MW,' stating that-such-reactors have no significant environmental impact. Rather than use this generic statement, McClellan Air Force Base has pro-duced:a" site-specific environmental assessment. This assessment shows that - the operation of the. proposed facility meets the criteria of the generic statement and will not pose'a threat to'the health and welfare of the on-site personnel or the general public off site near McClellan Air Force 3ase. i-i E d 4 4 r i 1 i 4 i s

I l SECTION A I . DESCRIPTION OF PROPOSED _ ACTION AND ALTERNATIVES 11. PROPOSED ACTION.

This environmental assessment addresses the installation'and operation-of a Stationary Neutron Radiography System (SNRS) at McClellan Air Force Base, Sacramento, CA (Fig. 1).

~ l-The SNRS will consist of a small TRIGA-type research' reactor 26-ft C (8 m).or more below ground level, shielded' radiography bays capable _of I h1 . inspecting up to 500 ft2 (46.5 m ) of aircraft panels per day, parts-- 2 1 manipulation facilities, component preparation areas, and associated radi- } ography facilities. o 2. DESCRIPTION OF THE PROPOSED SYSTEM ' 2.1 General: -Over 50 reactors throughout the world, including 28 in the United [ States, use the TRIGA-type uranium fuel with zirconium hydride _ moderator. Many of;these reactors have over 20= years of operational experience Their operational and safety records are excellent. ~Not all of:these TRIGA-reactors use ~ identical, core geometry. reflector? a'rrangements.- experimental l facilities,;or control systems. However, the operational and safety charac-teristics of the.SNRS willibeLtypical of most of the world's TRIGA reactors. 4 2.2 SNRS

2. 2.1 -

-Enclosure 6 The SNRS enclosure-will, in general, be configured as shown in Fig. 2.- The' enclosure will consist of radiography and reactor-associated rooms.- The enclosure will be designed to confine the results of i any credible reactor accident;;and in addition to housing the reactor, it-will provide space for all radiography operations and parts handling. The SNRS enclosure will also house the reactor; core, beam tubes, heat exchanger, = domineralizer resin tanks, and both the primary and secondary coolant - p pumps. The enclosure will be designed such that the ventilation systems i l } er . +

for the reactor-and radiography-bay areas can be isolated'from those for the' reactor control room and office area. There will be no exterior conduits, pipelines, or electrical _or mechanical structures such as transmission lines attached to or adjacent to the enclosure, other than utility service facilities similar to those required for other maintenance buildings on the base. The' reactor control room will' be located in the SNRS enclosure. All the-process and reactor instrumentation will be read out in the control room so that all aspects of reactor and radiography operation can be monitored'from this single location. 2.2.2 Reactor The SNRS reactor will be a TRIGA-type steady-state heterogeneous tank-type reactor-(see Fig. 3), operable at a power of 1 MW or less. The reactor core will be cooled by_ light water, will be moderated .by zirconium hydride and water, and will be reflected by water and graphite. The basic components of the SNRS reactor will include the fuel, control rods, control-rod drives, instrumentation, and control console. The reactor core will consist of 130 standard TRIGA elements. This is the maximum number of locations in the core grid plate. The actual number of fuel elements will be-less than this, since the core will consist of control rods-and reflecto*s as well as fuel elements. Each fuel element-will contain a homogeneous mixture of uranium and rirconium hydride. The fuel-elements will be 12 wt % uranium,'which will be enriched in uranium-235 (U-235) to less than 20%. Each element will be approximately 28 in. (710 mm) .long and 1-5 in -(30 mm) in diameter, and will be clad with 0.020 in. -(0.51 mm) of stainless steel (Fig. 4). This type of reactor fuel is class-ified as low-enriched uranium fuel (LEU) of low strategic significance; -therefore only minimal security and safeguards will be required. -The instrumentation and control systems will contain the necessary equipment to operate and protect the reactor and monitor the radiography operation. The SNRS control console will be a modern solid-state TRIGA design whose reliability has been developed over 20-30 years of safe

I'- 3 operation. The basic functions provided by this system are reactivity regulation, neutron monitoring, instrumentation display, startup operation, steady-state operation, and reactor shutdown. Reactivity of the reactor core will be controlled by changing the position of the boron control reds. The SNRS will have three to four control rods that are approximately the same size and shape as a fuel element, and that are suspended from fail-safe electromagnets located on the support bridge. The ionization and fission chambers used for sensing neutron flux densities will be suspended above the core. The control console will be located in the reactor control room adjacent to the reactor room. The reactor core will be located at the bottom of an l aluminum tank, which will be 8 ft (2.4 m) in diameter and approximately 26 ft (8 m) deep. The reactor tank will be below ground level and will contain approximately 10,000 gal (40,000 L) of demineralized water. The reactor tank will provide 16 ft (5 m) of water over the top of the reacter core for radiation shielding at the reactor top. The reactor tank will be surrounded by a concrete and soil biological shield. The concrete and soil will provide the lateral shielding r,ecessary for reactor and radiography operetions. There will be no penetrations of the reactor-tank walls. 2.2.3 Coolina Systems 2.2.3.1 Primary System The primary cooling system will consist of a primary pump, demineralizer pump, resin beds, particulate filter. heat exchanger, make-up water systen, and alu.ninum piping. The demineralized water will be pumped from the top of the reactor tank to the tube side of a tube-and-shell heat exchanger and back into the reactor tank. The primary system will be a low-temperature, low pressure, and completely closed system. - The water purification system will consist of a pool skimmer on the water surface, a demineralizer pump which will pump the water through the demineralizer resin bed, and a particulate filter. This system will act to keep the primary water clean and free of activation products. The i

4 primary system will have a siphon break to prevent water being pumped or siphoned out of the reactor tank in case of a piping break. Make-up water for this cooling system will be readily available and will be supplied from the local water %pply. This water will go through a demineralizer system before being injected into the reactor tank. 2.2.3.2 Secondary System The secondary cooling system will consist of a secondary cooling pump, an automatic flow-control valve, and a cooling tower. The secondary-water flow rate will be automatically controlled to keep the primary water returning to the reactor tank at a prescribed temper-ature. The secondary water will flow through the shell side of the heat exchanger, remove the primary water heat, g.- discharge the heat in the cooling tower. The secondary-system pressure will be higher than the primary-system pressure. This will ensure that any leak in the heat exchanger will be from the secondary side to the primary side, elimin-ating the possibility of primary water getting out of the SNRS enclosure. 2.2.4 Confinement and Emergency Ventilation Features The SNRS enclosure will be designed to confine any radioactive material released in an accident so that it may be exhausted in a controlled manner through an emergency cleanup and exhaust system. This system will filter out the particulate and radioactive materials before the air is exhausted to the environment. The filt. ration system will have, as a minimum, an in place efficiency of 99.95% for 9.3-micron sized particles. The SNRS enclosure ventilation system will be auto-matica11y shut down and the stack exhaust damper closed, thus isolating the reactor room area, if a high-radiation alarm sounds from the stack partic-ulate or gaseous monitor. The ventilation system for the reactor control room and office area will be separate from the reactor ventilation system. j

'4 Although the maximum credible accident analyzed in this document assumes that all enclosure integrity has been lost, the SNRS enclosure will be designed to preclude the direct release of radioactive materials or CJses to the outside environment even in such an accident. Instead, such materials will be exhausted in a controlled manner through the SNRS exhaust stack. 2.3 Radioaraphy Bay and Equipment The SNR$ reactor will be used for neutron radiography of aircraft components. To provide this capability, four beam tubes will be provided. The beam tubes will approach the lower edge of the core tangen-tially, pass upward at some angle from the vertical, and terminate at floor level, such that the required scan area for large parts is available in the l-radiography bay. Each beam tube will consist of two helium-filled sections. The first section will be in the reactor tank; the second section will be l outside the reactor tank and will extend into the radiography bay. The beam tubes will be constructed of aluminum, and will be capable of supporting collars, collimator inserts, and external water pressure. Each beam tube will be constructed such that a collimated beam of neutrons of sufficient area can pass through the aircraft components and then be stopped by a beam stop. The beam tubes will be helium-filled to eliminate Ar-41 production close to the reactor core. The area that immediately surrounds the reactor will be divided into four shielded radiography bays to allow radiography on each of the four beams. The radiography bays will be equipped with manipulators, real-time imaging systems, and operator control booths. The wall shielding and beam stops of each area will be such that with any one beam shut off, operators may enter and set up parts in that area while the other three beams are operational. l Each radiography bay will have its associated radiography control Doors to the radiography bays will be interlocked with the beam room. shutters. The radiography doors will have redundant interlocks, and entry

6 J Linto_the bays will be_ administrative 1y controlled. The doors will be sized for the;1argest collimator components, manipulator components, or aircraf t components-to enter the area. 3.. DESIGN ENVELOPE This environmental assessment is beins issued before the final design has been completed. Therefore, the analytical approach has been to establish a design envelope. The final design parameters must fit within this envelope to meet the criteria set forth in this document. The critical design _ items are divided below into two sections:

first, those items affecting facility design, Section 3.1; second, those item; cffecting the environment or personnel safety, Section 3.2.

All items shall be addressed in the SNRS Safety Analysis Report in appropriate detail. The final design parameters must, in all cases, be as conservative as those used in the environmental assessment. 3.1 Facility Design a. The SNR5 snel,.e a TRIGA-type nonpulsing steady-state heterogeneous tank-type reactor, operable at a power of_1 MW or less. b. The SNRS reactor grid plate shall consist of no more than 130 element locations, including control-rod, graphite, and fuel-element.

locations.

The SNRS reactor fuel shall se the TRIGA standard fuel type, c. less than 20% enriched'in U-235, d. The SNRS reactor tank.shall be below grade, constructtd of aluminum, approximately 8 ft (2.4 m) in diameter and 26 ft (8 m) deep. There shall be no penetrations of the reactor-tank walls, e. The in-tank fuel-storage rack shall hold a maximum of 20 TRIGA-type elements.

~ -. 7-o f._ The SNRS Control System shall be of modern solid-state TRIGA design. g. The primary system shall--have a siphon break to prevent water from being pumped cr siphoned out of the reactor tank. The secondary i pressure shall be higher than the primary system pressure, h. There shall be four helium-filled beam tubes. 1. There shall be no exterior conduits, pipelines, or electrical or mechanical structures such as transmission lines attached to or adjacent to the enclosure, other than utility service facilities siiailar to those required for other maintenance buildings on the base. J. The SNRS facility shall be designed such that the risk'of an aircraft accident affecting the reactor core can be shown to be below a frequency of-10'7/yr. Thereby meeting the requirements established in NUREG 0800 and BNL Design Guide 50831-III. 3.2 Environmental and Personnel Safety a. The SNRS enclosure shall be capable of confining the result' of any credible reacter accident. b. The enclosure ventilation system for the reactor and radiography-bay areas shall be isolated from those for the reactor control room and office areas. c. The enclosure shall consist of an emergency cleanup and exhaust system. The filtra'. ion system shall ' ave as a minimum en in place efficiency of 99.95% for C.3-micron sized particles. The ventilation system shall be automatically shut down and the stack damper closed if a high radiation alarm sounds from the stack particulate or gaseous monitor, d. The SNRS enclosure shall be designed to preclude the direct release of radioactive materials or gases to the outside environment;

. ~. . -instead, these materials shall be exhausted in a controlled manner through the SNRS filtration and exhaust stack. e. Each radiography bay door shall be interlocked with redundant systems such'that the doors shall not open unless the reactor beam shutter is closed, f. Each radiography bay'shall be designed such that the operator heay entar the area with the other radiography bays in operation. Entry shall i made with the beam shutter closed, and radiation levels must be such that ALARA and the 10 CFR Part 20 requirements are met for occupational

workers, g.

The boundaries and dose commitments described in Section B Part 1.1 shall be established and described fully in the SAR. h. The SNRS shall be designed to recycle any radioactive liquid waste that may oe-found in the liquid waste storage tanks. 1. The SNRS solid waste shall be handled in accordance with -existing McClellan Air Fnrce Base policy and procedures. j. The final' design of the handling system for gaseous effluents must result in N-16 and Ar-41 releases and doses during normal operations lower than or equal to those listed in Appendix A. For N-16 this means .less than 1 mrem /h over the reactor top, and Ar-41 release to the unrestricted ~0 3 area of no more than 2 x 10 pCi/cm, 1

~..- SECTION B ENVIRONMENTAL CONSEQUENCES 1. SITE DESCRIPTION PicClellan Air Force Base is situated approximately 8 miles (13 km) north-by-northeast of downtown Sacramento, California (Fig. 1)..Within this base, the site of the radiography system would be about 3000 ft (0.6 mi or 0.9 km) west of Watt Avenue, the nearest eastern border of the base, and 4500 ft (0.9 mi or 1.4 km) south of E Street (Fig. 5). Sacramento.is situated in California's Central Valley between the Sierra Nevada and Coastal Range. The area is characterized by hot summers (July mean maxisium temperature 105*F) and cold winters (January mean minimum temperature 28'F) (Ref 7). As in most of California, the majority of the annual average precipitation, about 17 in. (40 cm) falls in the winter months as rain. The prevailing winds in the area are from the south to south-by-southeast (Fig. 6). McClellan AFB is located between the communities of North Highlands-Foothill Farms, Arden-Arcade, Rio Linda-Elverta, and North Sacramen u. Land-use patterns around McClellan AF8 are mixed, with vacant industrial land to the west; a mixture of agricultural-residential and light industrial land uses to the northwest and north; ttrip commercial and residential land uses to the east; light industrial, warehouse, and open-space land uses to the south; and residential land use to the southwest (Ref. 8). Over the last decade the surrounding area has become more urbanized as the local population has grown. Residential usage of surrounding lands has increased sharply. Metropolitan Sacramento has a population of about 1,000,000 (1980 census).an increase of about 26% since 1970 (Ref. 9). The major popula-tion center lies south-by-southwest of the base (Fig. 7). The city of Sacramento itself has a population of about 276,000, all within 10 mi

. (16 km) of the. air-base. Approximately 700,000 people reside within 10 mi (16 km) of the base (Fig. 7). As suggested-by the land-use patterns,. population densities north and west of-the site tend to be low. The base and adjacent lands are located in th9 Great Valley subdivi-sion of the Pacific Border Physiographic Province (Ref. 10). The area is -situated on the alluvial plains of the Sacramento River and its tributaries -(Ref. 11). LYlia land is relatively flat, ranging in elevation from 50 to 75.ft (15 to 23 m) above mean sea level. . Soil. cover of about 4 ft (1.2 m) consists of a sandy-loam (Ref. 12). The ' surface soil is moderately permeable but the 'suosoil has low perme- -cbility..The soils have moderate water-holding capacity and slight erosion hazard. Natural scrface drainage has been altered by construction of a series of storm drains (Ref. 11). Drainage is directed into Magpie or Arcade Creeks and flows westward off base, ultimately reaching the Sacramento River. The nearest 100 yr floodplain is about 3400 ft (1037 m) from the site of the SNRS. The groundwater aquifer lies approximately 80 to 100 f t (24-to 30 m) below the ground surface (Ref. 29). The area is located in Seismic Zone 3 of-the Uniform Building Code. In general, seismic activity is not as great in the area as it is in the coastal areas (Refs. 13, 14, 15, and 16). Based on a review of historical' records, the maximum-intensity earthquake in Sacramento in historical times has been about VII on the Modified Mercalli scale (Refs. 15 and 16). This intensity was the result of earthquakes centered about 20 mi (32 km) west of Sacramento with an estimated magnitude of 6.0.to 6.5 on the Richter i -scale.. Earthquakes of the intensity of VII are characterized by collapse of weak chimneys, moderate-damage to masonry walls, fall of cornices from high buildings, and fall of some nonstructural, unreinforced brick walls (Refs. 15 and 16). However, earthquakes of higher intensity could have occurred prior to the coverage of the historical record, and higher-intensity earthquakes are possible in the future.

3 4 1.1 SNRS Site Boundaries and Associated Dose Limits The site boundaries and associated dose commitments for the SNRS have been taken from the American Nuclear Society Sta.idard for Research-Reactor Site Evaluation-1977, ANS 15.7 (Ref. 30). The areas for siting the SNRS have been separated into three categories. The first area is the SNR$ operations area. This is the area inside tne perimeter fence (operations boundary) surrounding the SNRS enclosure. This area is denoted by (A) in Fig. 8. The second area is the site area. This is the area outside the SNRS perimeter fence (operations boundary) but inside the perieter fence surrounding McClellan Air Force Base (site boundary). This area is denoted by (B) in Fig. 8. The third and final area is the urban area, This is the area outside the perimeter fence surrounding McClellan Air Force Base (site boundary). This area is denoted'by (C) in Fig. 8. The SNRS operations boundary is defined by the perimeter fence surrounding the enclosure. The reactor administrator has direct authority over all activities in this area, and there shall be prearranged evacuation procedures known to those who frequent the area. In the event of a design-basis, one-element accident (maximum credible accident), the planning assumption shall be that all persons within the operations area are evac-uated in sufficient time so-that the dose commitment does not exceed 25,000 mrem to the "whole body" or 75,000 mrem to any "other organs." The SNRS site boundary is defined by the perimeter fence sur-rounding McClellan Air Force Base. The nearest site boundary will thus be at Watt Avenue, 3000 ft east of the enclosure. The reactor administrator may directly initiate emergency activities within the site area; this area may be frequented by people unacquainted with reactor operations. In the event of a design-basis, one-element accident, the dose commitment for people within the site area shall not exceed 5000 mrem to the "whole body" or 15,000 mrem to any "other organs" over a 2-h exposure period.

The SNRS urban boundary is defined by the area outside the per-imter fence surrounding McClellan Air Force Base. The urban boundary means the nearest boundary of a densely populated area or neighborhood, where a complete rapid evacuation would be difficult or could not be accomplished within 2 h using available resources. The dose commitment associated with the design-basis accident for persons at or beyond the urban boundary shall not exceed 500 mrem to the "whole body" or 1500 mrem to any "other organs" over a 24-h exposure period. 2. ENVIRONMENTAL EFFECT OF REACTOR AND SNRS ENCLOSURE INSTALLATION Installation of the facility will include excavation of a pit, instal-latica of a tank, and erection of the walls and roof of the SNRS enclosure. Standard construction procedures will be followed. Releases to the environ-ment during construction will include dust from earth-moving activities, releases from internal-combustion engines, accelerated erosion of soil materials _due to ground-surface disturbance, and noise due to equipment operation and construction activity. These emissions will not exceed levels normally associated with general construction. Th1y will not adversely affect the surrounding developed area to any measurable degree, and in fact will be similar to the levels shown in Ref. 17, Sec. B. Because this action will entail only a moderate level of construction activity in an area already developed for industrial uses, it is expected that the action will have no-significant imoact on the local aquatic and terrestrial biotic systems, including threatened or endangered species; additionally, no impact to cultural or historical resources is anticipated (Ref. 29). Reactor-fuel transport and loading will follow all applicable NRC/ DOE / 00T requirements, 10 CFR Parts 72 and 73. Transport and loading procedures are well established, and experience has shown that no threat to the public health and welfare will accrue from these activities. '3. ENVIRONMENTAL EFFECTS OF OPERATION 3.1 Thermal Effects The 1 MW or less of heat generated by reactor operation will be dissipated to the atmosphere by means of a cooling tower. This rate of l i

= 4 . 4 heat dissipation is comparable to that associated with light industrial operation and will have no affect on the local environment. Some minor ground fogging may occur in the immediate vicinity, but this will not af fect other maintenance operations-or aircraf t runways in the area. 3.2 Radiological Impact There are over 50 similar TRIGA reactors in operation wor'dwide, including 28 in the U.S. Many of them have been in operation 50 densely 4 populated areas for periods of over 20 years. All operating TRIGA_ reactors have excellent records of safe operation with insignificant radio'iogical impact..The proposed SNRS operating schedule of two shifts, five days a veek, is more than most typical TRIGA facilities operate, but as shown in Appendixes A and B, leads to no radiological problems. The potential for radiological impacts or, the environment is assessed in the following three circumstances: normal operation, abnormal operation (maximum credible accident), and abnormal operation (other accidents).- The abnormal-operation (maximum credible accident) category consists of the accident that leads to the " worst" radiological consequences. This accident it the. fuel-handling accident in air, which is also termed the design basis accident (Ref 4). This accident is postulated using very -conservative assumptions, therefore enveloping the highest release of radioactivity. The abnormal-operation (other accident) category comprises accidents that have been-previously analyzed for TRIGA reactors (Ref. 4) and found to be of. lesser consequence than the maximum credible accident. 3.2.1 Norual Operation During normal operation, the only sources of radio-activity outside of the fuel cladding are caused by neutron activation of reactor str uctural components, ion-exchange-column resins, water, and air filter,. Two sources of radioactive gas are caused by activation of the i

E s l argon in the air and the oxygen in the water.. Additionally, the SNRS could I - cause low-level activation of parts being inspected. Each of these items wi*' be covered below. The reactor structural components that are in or near the reactor core, and therefore will be exposed to intense neutron irradi-ation, are the reactor grid plates, control-rod guides, control rods and in-core instrument housings. To the maximum extent possible, these compon-ents will be made of aluminum or other materials with activation products that have relatively short half-lives for radioactive decay. Furthermore. -these activation products will be distributed throughout and maintained within the volumes of these solid materials, so they are not readily releas-able to tLe environment. Exposures from contact-with in-core structural components will be minimited because maintenance will be performed through 16 ft-(5 m) of water. All exposures to personnel will be minimized through an ALARA (As Low As Reasonably Achievable) program and will be far below the. limits of 10 CFR Part 20. Some low-levels of radioactivity-could be induced in components being radiographed. Experience at similar radiography facilities (Aerotest, GE, NRAD) has shown that induced activity of this type is very low. All items exposed to.the beam will be be monitored before leaving the facility, there will be no significant radiological impact on the environ-ment. There are two main radioactive gases produced during normal reactor operations. The first is N-16, produced _by the interaction of neutrons with the oxygen in water. The second is Ar-41, produced by neutron interaction with the argon in air. The dose due to N-16 to SNRS personnel is analyzed in Appendix A. It is shown in this analysis that the potential dose rate from N-16' directly over the reactor core is less than 1 mrem /h. The N-16, because of its short half-life, presents no impact to the environment beyond the SNRS enclosure. 1 I l

= -. - 15 - The relationships between Ar-41 concentration, enclos-ure volume, exhaust rate-from the SNRS enclosure, and Ar-41 production rate within the SNRS enc!osure assuming that the decay constant (A) can be neglected are generally as-fellows. Where: 3 C = Concentr; tion of Ar-41 (pC1/cm ) V = SNRS enclosure volume i 3 E = Exhaust rate from the SNRS enclosure (cm /s) P = Production rate o" Ar-41 (pCi/s) A = Total Ar 41 activity within the enclosure C=handC=h Then: As these equations show, the concentration of Ar-41 can be controlled by adjusting the exhaust rate (E), the volume (V), or the production rate (P). The Appendix A calculation for Ar-41 concentration during normal operation makes a number of very conservative assumptions. in order to establish an envelope for a release concentration for the SNRS, the assumptions include a fixed enclosure volume, an exhaust-rate near zero, and a constant production rate, resulting in an equilibri

concentra-tion. As the equations above and Appendix A show, the same conantration can be arrived at by leaving enclosure volume unspecified and choosing an 3

6 3 exhaust rate of'60 ft /s (1.68 x 10 cm /s). The important result of these calculations is that a release concentration of.2 x 10-8 3 Ci/cm is attain-able, whether done by fixing building volume or exhaust rate or a combin-ation of the two. As stated in Section 3.2 j, the final design must result in no more than this concentration regardless of.how the requirement is met. t

! [ t i Appentlix A shows that the Ar-41 concentration in the restr!cted area (area with the operations boundary) will be below 10 CFR [ -6 3 Part 20 limits (2 x 10 pCi/cm ). Furthermot., the Ar-41 released to the i unrestricted area (area outside the operations boundary) will not exceed -8 3 2 x 10 uC1/tm which is below the 10 CFR ? art 20 limit of 4 x 10*b 3 pCi/cm. Appendix B shows that the expected occupational doses will be far below current regulatory limits. In addition, the reactor-related dose rate at the r:es/Jst si'e boundary, 3000 f t east of the SNRS enclosure at Watt Avenue, will be much less than 0.1 mrem /yr. Nonradiation workers on the air base would be expected to r ceive dose rates from below 0.1 mrem /yr to { a maxirhue of 100 mrem /yr. Emissions from the SNRS enclosure will be rapidly dispersed, and only workers within a few hundred feet of the enclosure could possibly receive the higher doses. All doses will be Hell belaw current regulatory li.alts (Table 1). For comparison, the individual dose rate from natutL1 sources of radioactivity is about 100 mrem /yr. Personnel working in the vicinity of the SNRS enclosure 4r will also receive a dose from radioactive emissions from nngoing X-ray and nanouverable neutron radiography systems in the NO! facility (Ref. 17). The maximum dose rate received by a worker near_these two systems would not exceed 2 n'lem/h (Ref. 17). Based on a 4000-h work year, the SNRS would add no more than 0.025 mrem /h to the dose rate received by a nonreactor worker outside the SNRS enclosure. Given that this maximum dose rate (100 mrem /yr) would only oe received by individualc immediately (djacent to the radiography enclosures for 4000 h/yr, the nonreactor workers on base will not receive [ cumulative doses in excess of 10 CFR Part 20 limits for aonoccupational conditions and realistically, they will receive well below these limits. -i 3.2.1.1 Solid Waste The lon-exchange resin and water particulate . filters..in the reactor-water purification system will be routinely changed before high levels of radinactive materials have accumulated (normally once a year), thereby minimizing personnel exposure. The irradiated resins and r L

=. l l TABLE 1. Estimated Annual Doses and Annual Dose Limits for Normal Operatio't of $NRS Estimated b e Maximum Dose Dose limit Receptor" (Whole Body) (Whole Body) Occupational 50 mrem 5000 mrem Personnel (Within the Operations Boundary) Other On-base < 100 mrem 500 mrem Personnel (Within the Site Boundary) - General Public 0.1 mrem 500 mrem -(At the Urban Boundary)

Cccupational personnel are individuals directly involved with operation of the reactor.

Other on-base persoiinel are assumed to be nonradiation workers; hence dose limit is the same as for the general public. b Calculated as discussed in Appendix 8. C From 10 CFR Part 20.

4 . l air filters will be packaged and shipped according to 10 CFR Parts 71, 49, and 179, and Air Force Technical Orders 00-110N-2 and 00-110N-3. The volume of such wastes is not expected to be more than five regular SS gal '(208-L) drums per year. No more than one shipment per year to a licensed disposal site is anticipated. This amount of waste will not ontributa cignificantly to the volume of waste at the licensed disposal site. Agree- .nents are in place at McClellan Air Force Base to dispose of solid waste. 3.2.1.2 tiquid Waste During normal operation the SNR$ will generate ~ no radioactive liquid waste. Any suspect liquid waste in the facility will be stored in storage tanks and sampled for possible radioactivity. Should any radioactivity be found, the liquid waste will be either: (a) returned to the reactor water demineralizer system, or (b) solidified and the solid waste disposed of as described in Section 3.2.1.1. Small amounts of nonradioactive solid-content water may be released from the facility through the sanitary sewer during periodic blowdown of the cooling tower. 3.2.1.3 Other Waste Other potential sources of airborne radio-nuclides are neutron-activated impurities in the reactor-tank water, and activated airtaene particulates (dust) from the radiography bays. These~ sources are very small in comparison to those above for the following reasons: The reactor-tank water will be circulated through a filter and demineralizer syste.? that will prevent the build-up of any significant concentrations of impurities that could be activated as the water passes through the core. Any material activated will be trapped in the coolant purifier system and not released to the environment through any credible pathway.

19 - o Any radioactive dust that might be airborne would probably be produced in the radiography bays. The materials in the radiography bays will be chosen for their lack of activation from inter-action with the neutron beam. Furthermore, the air exhaust system will be filtered. Doses to workers will be controlled and limited to levels well below those specified by NRC regulations for occupa-tional doses (i.e., whole-body doses of 3 Rem / quarter or 5 Rem /yr). Workers will be trained with regard to radiation riskt and proper health physics procedures, to ensure that occupational doses are within the limits given in 10 CFR Part 20 and in compliance with ALARA goals.

3. 2.1. 4 Monitoring An environmental-monitoring system will be initiated and maintained once the SNRS construction begins.

Theobjective of this system will be to ensure that SNRS air exhausted to unrestricted areas does not exceed applicable NRC guidelines or regulations (10 CFR Part 20). Based on the operai.Ing schedule of 16 h/ day, five days a week, the $NRS will remain well below the NRC release limits for normal operations. Even an increase of several hundred hours in steady-state operations per year "ould not exceed NRC release limits.

3. 2.1. 5 Spent Fuel Operation of the SNRS reactor will consume very little fissile uranium in the fuel elements, based on the consumption rate of other similar TRIGA reactors.

Typically one new fuel element a year may be added to the core, with the spent fuel element put in the in-tank fuel-storage rack. The in-tank fuel-storage rack will hold a maximum of 20 fuel elements. It is antic.ipated that spent fuel shipments would be required once in 10-15 yr. These shipments will be done according to current DOE /NRC and 00T requirements. Since the fuel will be DOE owned, the spent fuel will be returned to DOE custody. Based on other TRIGA facility experience, the shipments will result in no environmental problems or personnel hazards.

. 3.2.2 Abnormal Operations Since TRIGA-type reactors have been operating for over 20 yr, there have been many studies of possible accidents for this type of reactor (Ref. 4). For the majority of TRIGA reactors, the fuel-handling accident (design-basis accident) consisting of a one-element failure in air, has been identified as the " maximum credible accident." Therefore, the fuel-handling accident or the one-element accident has been analyzed for the SNRS facility. The one element accident scenario is not analyzed here as to "how it occurs." The position taken is that it has happened and the results are analyzed in Appendix B. Since the off-site accident release from a small research reactor would be several orders of magnitude less severe than the criteria developed for power reactors, the more conservative dose limits established in the American Nuclear Society Standard for Research Reactor Site Evaluation (ANS 15.7) will be used as the criteria for the SNRS siting. These dose limits are given in Section B Part 1.1. 3.2.2.1 One-Element Accident (Maximum Credible Accident) At some point in the lifetime of a TRIGA reactor, used fuel within the core may be moved to new positions or removed from the core. Fuel elements are never moved unless the reactor is shut down. The most serious one-element accident involves spent or used fuel that has been removed from the core and dropped or otherwise damaged, causing a release of fission products in air. The accident analyzed in this document is conservative in that assumptions of an extremely unfavor-able meteorology and a large fission product inventory in the damaged fuel are made. Thus, the projected offsite doses represent an upper limit. Even with these very conservative assumptions, radiation doses to the general public, as a result of the one-element accident, will be small. The most stable atmospheric conditions (least dispersive) occur when the prevailing winds are from the south to southe6st. Thus, maximum exposure

. levels would be expected to occur northward from SNRS. The estimated maximum thyroid dose at a distance of 4500 f t (1400 m) is 0.18 mrem, to the north of the SNRS enclosure (Table 2). The thyroid dose results from inhalation of radioactive iodine isotopes. The estimated maximum whole-body dose will be.002 mrem at the same location. 1 The nearest point of public access will be i-3000 f t (910 m) east of tii %NRS. Doses received by members of the general public at this nearest patat *r. shown in Table 3. It can be seen that the doses are far below the limits zur the urban area in ANS 15.7 of 500 mrem whole body and 1500 mrem other organs. These doses were calculated using conserva-tive assumptions and do not account for radioactive decay, containment of releases within 24 h, or evacuation of the exposed individual before receiv-irg the full released dose; hence, the actual doses are expected to be much lower if such an accident occurred. 3.2.2.2 Multiple Element Failures If the noncredible assumption is made that all the elements in the core (maximum 130) and all the elements in the storage rack (maximum 20) are destroyed, this would result in the destruc-tion of 150 elements. If a further extremely conservative assumption of linear extrapolation from the one-element highest power density fission-product-inventory case is made, the maximum resulting dose at the 4500-ft nearest site boundary to the north is 27 mrem to the thyroid and.30 mrem to the whole body. The dose to the general public 3000 ft to the east (Watt Avenue) would be less than this because of the prevailing atmospheric conditions (Fig. 6). Even in this noncredible situation the dose to the general public would be far below the limits of ANS 15.7 (500 mrem whole body and 1500 mrem to other organs) and even far below the 10 CFR Part 20 limits for normal operations (500 mrem whole body). I

E TABLE 2. Whole-Body and Thyroid Doses From a One41ement Accident (in the Direction of the Maximum Exposures, i.e. North of SNRS Enclosure) Maximum Credible Accident (One-Element Accident) j Whole-Body Thyroid Distance Dose Dose t (North) (mrem) (mrem) 330 ft 1.0 x 10'I 10 (100 m) 4500 ft 1.8 x 10'3 0.18 (1400 m) 1 mi 1.6 x 10'3 0.15 l (1610 m) 10 mi 1.6 x 10'4 0.015 (16,100 m) ' Siting Criteria 500 mrem 1500 mrem (ANS 15.7 Urban Area) t ^ TABLE 3. Whole-Body and Thyroid Doses From a One-Element Accident (in the Direction of Closest Base Boundary, i.e. East of SNRS Enclosure) Maximum Credible Accident (One-Element Accident) Whole-Body Thyroid Distance Dose Dose (East) (mrem) (mrem) -2 330 ft 5'.6 x 10

5. 5 (100 m) 1300 ft 5.3 x 10'3 0.52 (400 m)

-3 3000 ft 1.4 x 10 0.14 (910 m) 1 mi 6.7 x 10'4 6.5 x 10-2 (1610 m) 10 mi 4.1 x 10 4.0 x 10'3 -5 (16,100 m) Siting Criteria 500 mrem 1500 mrem (ANS 15.7 Urban Area)

23 - 3.2.2.3 Other Accidents (Less Severe Accidents) Credible accidents for TRIGA-fueled reactors were evaluated in Ref. 4. The reactors were evaluated in the light of contemporary knowledge and the long operating history of this class of reactor. Seven categories of accidents were analyzed: e Excess reactivity insertions e Metal-water reactions e Lost, misplaced, or inadvertent experiment o Mechanical rearrangement of the core e Loss-of-coolant accident Changes in fuel morphology and ZrH, e composition e Fuel handling Each of these seven accidents has been analyzed in Ref. 4. Each accident is evaluated in terms of its off-site radiological consequentes. The " maximum credible accident" is found to be the fuel-handling or one-element accident, which has been addressed. In addition to these seven accidents, two other accidents are addressed here. The first is the seismic breaching of the reactor tank, an'd the second is the complete loss of water from the reactor tank and its impact on the local groundwater. In general, seismic activity in Sacramento is low relative to other areas of California (Refs. 13, 14, 15, and 16). It is possible that a major seismic event could breach the reactor tank, resulting in a release of radioactive materials. However, this situation would not exceed the releases analyzed in Sections 3.2.2.1 and 3.2.2.2. As a result of activation of impurities in the primary cooling water, the water will contain small amounts of radio-nuclid'es. 'Dyring normal operation only two radionuclides are normally present in measurable amounts: Ar-41 (2 x 10'4 pCi/cm ) and Hn-56 (1.5 x 3 -5 3 10 Ci/cm ) (Ref. 28). The activity concentration of manganese was found 4

to be ten times lower-than the limit given in 10 CFR Part 20 (1 x 10'4 s 3 pCi/cm ) for exposure to the general public. However, if primary cooling water were released from containment, 10 CFR Part 20 limits would be reached in about 24 h through nuclear decay. Analysis of the potential for radio-activity to reach the groundwater from such an event was based upon several conservative assumptions: s (1) Groundwater depth of 80 ft (24 m) (Ref. 29) (2) No change in hydraulic head (3) A hydraulic gradient (1) equal to 1 (4) A hydraulic conductivity (k) of 4.57 x 10'4 ft/s (Ref. 29) The relationship to determine the time (t) for water to move from the bottom of the reactor tank a distance D to groundwater is: t = D/(k x i). s Based upon the SNRS site conditions beneath the reactor tank, it would require over 36 h for water to reach a ground-water aquifer, if all reactor-tank containment were removed. Argon-41 -10 3 activity would fall to about 2 x 10 pCi/cm during these 36 h. Because of their low solubility in water, argon and the other noble gases have no limiting water concentrations under 10 CFR 20. However this concentration level is well below the air-release limits to unrestricted areas (10 CFR Part 20), and water concentration limits are generally higher than air concentration limits; therefore the Ar-41 will present no environmental problems to the groundwater. 3.3 Other Impacts Operation of SNRS will result in no significant increase in traffic, noise, solid or liquid wastes, or air emissions. It has been shown and accepted by the NRC (Ref. 19) that reactors of this size do not in general have significant effects on the environment of the areas they are-in. 1

. 4. RESOURCE USE The land is the property of the Federal government and is already dedicated to aircraft-maintenance activities. There will be a small increase in fuel consumption due to construction of the SNRS facility. The uranium fuel currently exists in manufactured form and represents a minute fraction of available fuel. Only a small amount of water will be consumed for cooling and film processing purposes, and water consumption will be a minor fraction (< 1%) of base-wide water use. Personnel requirements for construc-tion and operation will be minor compared to the size of the available work force in metrtpolitan Sacramento. Thus, the project's requirements will have insignficant impacts on resource availability. 5. LONG-TERM ASPECTS The expected useful life of this system is in excess of 20 years. The oldest TRIGA reactor has been operating 25 years, and numerous others have over 20 years of use without signs of age limitation. Decommissioning of sites for such small reactors has been accomplished successfully on several recent occasions (Ref. 18). Companies such as Rockwell and the Institute of Resource Management Inc. have extensive experience in decommissioning such small research reactors at costs of about 10% of those for initial installation (constant dollars). 6. COST-BENEFIT The benefits of the proposed SNRS facility to the U.S. Government include significantly reducing risk of crashed aircraft and significant financial savings in the aircraft-maintenance project itself. The benefits specific to the local public should include, in the long term, improved employment opportunities due to maintained or improved workload at McClellan Air Force Base, and improved safety of military aircraft. A corollary benefit should be improved safety of civilian aircraft and ground popula-tion. 7, CONCLUSIONS In conclusion, the nonradiological impacts would be no greater than those normally associated with construction of small industrial facilities.

. During normal operation, exposures to reactor workers and the general public would be less than 50 mrem /yr and 0.1 mrem /yr, respectively. These exposures are far below the 10 CFR Part 20 limits of 5000 mrem /yr for workers and 500 mrem /yr for the general public. In the event of an accidental release from one element in air (design-basis and maximum credible accident), the maximu'a dose to the general public would be 0.002 mrem to the whole body and 0.18 mrem to the thyroid. Even in the noncredible case of all elements being destroyed, the maximum dose to the general public would be only 0.30 mrem to the whole body and 27 mrem to the thyroid. Both of these dose levels are far below the siting criteria in ANS 15.7 (500 mrem "whole body" and 1500 mrem "other organs"), and even far below the limits in 10 CFR Part 20 for normal operations. The NRC has endorsed a generic statement for research reactors below 2 MW, which states that for such reactors there is no significant environ-mental impact. Rather than using this generic statement, McClellan Air Force Base has produced a site-specific environmental assessment which shows that the operation of the proposed facility will not pose a thrcat to the health and welfare of the on-site personnel or the general public off-site in the vicinity of McClellan Air Force Base.

4 27 - SECTION C 0FFICES, AGENCIES, AND PERSONS CONSULTED Person Position Office / Agency / Telephone Doug Froom NDI Support Section Chief SM-ALC/MANPJ/643-4418 Robert Sarte11 NDI Engineer SM-ALC/MANPJ/643-4418 l John Barton Contractor N-RAY Er.gineering/ l (610) 459-7604 l I Dale Schulze Civil Engineer SM-ALC/ENV and Contract Planning /643-3336 Millard Wohl NRC Expert on Modeling NRC/(301) 492-7065 Radionuclide Releases Jim Hawxhurst NRC Expert on the Use NRC/(215) 337-5135 of PAVAN l Pat Easly NRC Expert on Dosage NRC/(301) 492-8375 Modeling from PAVAN Bill Snell NRC Expert on Use of NRC/388-5513 PAVAN

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- OUTSIDE SAC RAMENTO SMS A - NO DATA Fig. 7

I /,- % N / N / \\ [k 7s[ ' 4. \\ Nj/\\ \\/ l \\f\\ , /\\/ s \\ OPERATION AREA ; I ' l I i / AND BOUNDRY N$k$h ; l i I _lL \\ SITE AREA, l i l AND BO N s%_ _4D8y y _. _LF .R. URBAN AREA P SNRS BOUNDRIES FIGURE J._

APPENDIX A RADIOLOGICAL SOURCE TERMS (Normal Operations)

APPENDIX A RADIOLOGICAL SOURCE TERMS 1. INTRODUCTION This appendix analyzes the concentrations of Ar-41 and N-16 gas that l will be produced during normal SNRS reactor operations. Radioactive gaseous Ar-41 is formed in and near the reactor core. The source is neutron irradiation of natural argon, a minor constituent (about 1%) in normal air. The two principal locations where Ar-41 can be formed are: (1) The air in the radiography bays. (2) The small concentration of air that is normally dissolved in the reactor-tank water. Some of the Ar-41 formed in the reactor-tank water will escape into the SNRS enclosure and then be exhausted to the environment through the SNRS stack. Most of the Ar-41 formed in the radiography bays eventually will be swept out of the bays by the ventilation system and exhausted through the stack. Therefore, Ar-41, with a radioactive half-life of 1.8 h, has the potential of impacting the general public. All TRIGA reactors operating in the U.S. have addressed the Ar-41 release question. Without exception these facilities have shown the Ar-41 release to be of no environmental concern or impact. The majority of these facilities have experimental facilities (i.e., air-filled tuia). hat 0 into the core area; therefore, air is exposed to intense neutro; Dombara ment, resulting in high Ar-41 production. In contrast, the SNRS beam tubes will be helium-filled, and there will be no air-filled b6am tubes in the core ana. The International Commission on Radiological Protection (ICRP) has been a principal organization studying the effects of ionizing radiation for many years. In 1959, Committee r of the ICRP published recommen-

Appendix A Page 2 1 i dations for maximum permissible concentrations (MPC) of radionuclides in air and water. These recommendations became the technical bases for 10 CFR Part 20. More recently, Committee II has reviewed the current state of knowledge and published updated recommendations in Publication 39 (1978/79), f which supersedes Publication 2. In these publications, the ICRP recommends that the MPC in air of radioactive noble gases be based only on the whole-body dose computed for a person immersed in a large cloud of gamma-ray emitters. This guidance is justified in Publication 30, where it is shown that the internal and skin doses from the beta rays wot.it add less than 1% i to the total dose. In a cloud of finite size, as is more applicable to the SNRS, the percentage contribution of the beta rays is larger, but the total dose at MPC is correspondingly lower. Thus, because Ar-41 is a noble radioactive gas, computations are based on the ICRP recommendations. 2. AIRBORNE CONCENTRATION OF Ar-41 IN SNRS ENCLOSURE DUE TO NORMAL RADIOGRAPHY OPERATIONS The Ar-41 calculations done in this appendix represent one way, in -8 which the SNRS can meet the requirements of releasing no more than 2 x 10 3 pCi/cm to the unrestricted area. A set of assumptions has been established and the basis for each assumption has been briefly given. This set of assumptions should not be interpreted as the only set that will accomplish this goal. In fact the final design should strive to make the release as -8 low as reasonably achievable (ALARA). The concentration number of 2 x 10 3 pCi/cm should be interpreted as the maximum that will be allowed to be released to the unrestricted area. 2.1 Assumptions (1) Reactor power 1 MW The SNRS will perform most of its radiography functions at a power of 250 kW. (2) All four beam tubes in operation This is the maximum number of SNRS beam tubes. It is very unlikely that all four will be operating at one time. .a

Appendix A o Page 3 (3) Neutron flux at the plane of radioaraphy 2 x 107 2 n/cm.s (collimator ratio = 50:1) When all the beam fitters are in the beam tube, the neutror. flux will probably be of the order of 5 x 105 2 n/cm,, (4) Dimensions of the plane of radioaraphy 35.6 x 43.2 cm (14 x 17 in.) for the real-time work, which is the majority of the work, a circle of 9-in diameter is the biggest area that can be achieved. (5) Volume of SNRS enclosure approximately 2 x 109 3 cm (30 x 40 x 60 ft) There are many ways to arrive at the Ar-41 concentration calculated in this document. The design will allow for meeting this Ar-41 concentration by fixing facility parameters, such as stack dilution factors, stack exhaust rate, and air recycle rate, in lieu of setting a fixed enclo-sure volume. The ratio of exhaust rate to enclosure volume will be fixed in the final design to give the calculated or lower Ar-41 concentrations. (6) Air path length 20 ft (6.1 m) (based on diameter of biggest part to be radioaraphed) i In actual radiography operation a beam stop will be placed as close as possible to the object being radiographed, since scattered neutrons degrade radiographic quality. These beam stops are composed of moderating and highly absorbing materials. They are typically capable of reducing beam intensity by a factor of 10~4 -6 to 10 The beam stops are usually 2-4 ft behind the object being radiographed. Therefore, the assumed 7 2 beam intensity of 2 x 10 n/cm s with cross sectional dimensions of 14 x 17 in, and path of 20 ft in air to produce Ar-41 in the radiography bay is very conservative, as the air path will probably be no more than 2-4 ft before the beam stop is reached. The beam stop will be designed to reduce scattered neutrons. There is also no credit taken here for sweeping the Ar-41 out of the radiography bays. The final design will provide for diluting and sweeping this air out of the radiography bays, which will further reduce the Ar-41 concentration.

Appendix A Page 4 l (7) Equilibrium Concentration and No Dilution Equilibrium concentration and no dilution of the gas inside l the enclosure is assumed. This leads to very conservative release results. Another set of assumptions could lead to higher or lower Ar-41 coecentrations. The assumptions used here will provide the designer the maxinum envelope allowed to meet the calculated concentrations. The final design will be analyzed in the system's SAR, and it must be shown in the SAR that this design goal has been met. 2.2 Ar-41 Calculation The neutron beam intensity (J) at the plane of radiography is 7 2 assumed to be 2 x 10 n/cm s. The total number of neutrons per second (I) passing the plane of radiography is then I = JR where 3 c,2 (14 x 17 in.) R = area of the plane of radiography = 1.5 x 10 Therefore: 10 I = 3.0 x 10 n/s Becauss the beam tubes are helium-filled and sealed, there is no air to. produce Ar-41 within the beam tubes. The total number of neutrons will travel the length of the radiography bay (20 ft) and will interact with the air to produce Ar-41 in the radiography bay. The number of interactions over the beam path length is I, - I where I, = the number of ne'itrons passing through the air

Appendix A Paga 5 Therefore: -Itf I=I,e where: I = Ar-40 macroscopic cross section for neutron interactions -5 (1.7 x 10 /cm) t = path length (20 ft or 609.6 cm) f = fraction of argon in the air (9.4 x 10-3) then itf 6 I, - I = I (e - 1) = 2.9 x 10 interactions /s This is equivalent to the disintegrations per second at equilibrium. The production rate, (I - I) A -3 P =- = 8.4 x 10 pCi/s c where 4 c = 3.7 x 10 dis /s per pCi ~4 A = decay constant of Ar-41 (1.05 x 10 /s) The activity at equilibrium is ~ 6 2.9 x 10 dis /s 4 = 85.3 pCi 3.7 x 10 dis /s pCi Assuming that all four beam tubes are producing this amount, the total Ar-41 production is: Total Ar-41 = (4 x 85.3) = 341.2 pCi

Appendix A Page 6 Assuming that this activity is evenly distributed over the SNRS enclosure 9 3 volume, 2 x 10 cm, the equilibrium concentration is 341.2 901 3 = 1.7 x 10 pCUcm 3 2 x 10 cm We could, at this point, instead of assuming a constant volume and calculating 3 C(pCi/cm ), use C(pCi/cm)=f 3 where: P = Production rate of Ar-41 (pC1/s) 3 E = Exhaust rate of Ar-41 from the SNRS enclosure (cm /s) -8 3 To achieve a concentration of 2 x 10 pCi/cm with the same production rate, we would need a minimum exhaust rate of d E=f=(8.4xlo pCi/s)(4) -8 3 2.0 x 10 pCi/cm 6 3 E = 1.68 x 10 cm /s or 3530 cfm f-The Radiation Concentration Guide (RCG) limit for Ar-41 in this restric- -6 ted area is 2 x 10 pCi/cm3 (10 CFR Part 20); therefore, the radiography operation will not produce excessive levels of Ar-41 in the SNRS enclosure. The-equilibrium concentration will be discharged from the enclosure during normal operation. Alarms will be provided if a beam-tube seal should fail, at which time radiography on that beam tube will be stopped. Also, access to radiography bays is prevented by interlocks and administrative controls when the beam shutters are open, eliminating the possibility of excessive personnel exposure with the beam-tube shutter open.

Appendix A Page 7 3.0 Release of Ar-41 from Reactor Water The Ar-41 activity in the reactor tank water results from irradiation of the air dissolved in the water. The following calculations were per-formed to evaluate the rate at which Ar-41 escapes from the reactor tank water into the reactor enclosure. The calculations show that the Ar-41 deaays while in the water, and most of the radiation is safely absorbed ir the water. Each of the design parameters used were taken from facilities similar to the SNRS. The changes in Ar-41 concentration in the core region, in the tank water external to the reactor, and in the air of the reactor enclosure are calculated using the variables as defined below: 3 V = Volume of region (cm ) 3 N = Atomic' density (atoms /cm ) A = Decay constant (sec~1) 2 o = Absorption cross section (cm ) 3 q = Volume flow rate from SNRS enclosure exhaust (cm /s) w = Mass flow rate (gm/sec) 3 p = Density (gm/cm ) 3 v = Volume flow rate through the core (cm /s) y 2 $ = Average thermal neutron flux in the core (n/cm,3) A = Fuel element length (heated) c A = Flow area (standard TRIGA fuel element) f The volume flow rate through the core is 3 v = * = 7.0 x 10 a/sec = 7.0 x 103 3 cm /s y P 3 1 gm/cm where a typical mass flow rate w has been used. The exposure time in the core is t=V/vy=Af c A /V = 2.5 see 1*7 y 3 It remains to find the atom density N for dissolved argon in the reactor tank water.

-Appendix A Page 8 According to Henry't. Law for_ gases in contact with liquids, the equi-librium concentration _in the liquid is proportional to the partial pressure of the gas.- The saturated concentration of argon in water at one atmosphere' of argon is-(Ref. 20).

C cm /1000 g H o-g O

52.4 25 30.8 50 22.5 i-Since the argon content of air is 0.94% by volume, the partial pressure of argon above_the water is: 0.0094 (760 mm - 23 mm, vapor pres, ore of eater) = 6.9 mm-(Hg) i The argon concentration (N) in the water at the core inlet temperature (32*C) is-3 3 19 3 N = (0.28 cm Largon /cm HO)k9""(2.7x10 atoms /cm ) 2 15 3 = 6.9 x 10 atom /cm l-l~ At 67'C, the core exit water temperature, the solubility of e gon in H O is 2 15 3 0.020 and the--concentration is 4.9 x 10 atoms /cm, The Ar-41 density-(at equilibrium) at the exit from the core is given by A, = A e-At + Na$ (1-e-At) g and at the entrance A, = A,e where t is the exposure time in-the core (2.5 s). y

r-y

Appendix A Page 9 The average out of core cycle time T is given by 7 3 T=v = 3.7 x 10 cm 3 = 5.29 x 10 sec 3 3 1 7.0 x 10 cm /s where V is the tank volume (8-ft diameter and 26-ft deep) and v is, 2 y again, the volume flow rate through the core. The solution to this set of equations is -At A, = b 1~I 2 , A(t6) Substituting the values from above, one obtains: A,= (6.9 x 1010) (0.61 x 10~24) (1.0 x 1013) x 1 - exp i-(1.06 x 10'4) (2.5): 1 - exp [-(1.06 x 10'4)(5.29 x 10 )] 3 3 3 A,= 25.99 dps/cm s 26 dps/cm and 43 = 1.06 x 10 4 = 2.5 x 105' atoms /cm, 3 N One source of Ar-41 in the room results from the reduced solubility of argon in water as the temperature increases. Considering the expected temperature rise of the water passing through the core, an immediate release of about 29% of the Ar-41 made could be expected during passage. Some of this Ar-41 might be redissolved as it is transported into cooler water, but-since the cooler water is in equilibrium with the air above, it is nearly saturated with argon and wil not absorb all of the argon released. Measure-ments of Ar-41 in the water as a function of height above the core indicate that approximately 60% of the released Ar-41 is reabsorbed.

Appendix A Page 10 Assuming that the 12% (29% x [1 - 0.6]) of the Ar-41 comes out of solution, remains undissolved after leaving the core, and escapes to the air, this source would be: Sy = 0.12 N43 3 = 0.12 x 2.5 x 105 (atoms) x 7.0 x 103 (]3c) v Cm '= 2.1 x 108( ) The tendency of the balance of the argon act e:ty in the tank to escape to the air owing to its proximity to the water-air boundary will constitute the additional source of Ar-41 at the water surface. An estimate of the fraction of argon atoms in the tank water that escape each second can be obtained by considering the relatien between the average distance traversed by a diffusing particle and the diffusion coefric-ient obtained from the random walk treatment.- The relation is (Ref. 21). AX2 = 2Dt where AX is the mean square displacement in time t and D is the diffusion coefficient. We then have for the mean displacement AX-= (2Dt)1/2 The diffusion coefficient for argon atoms in water (at 23'C) has been -5 2 measured to be 1.1 x 10 cm /sec-(Ref. 31) so that in one second the average displacement is (2.2 x 10-5)1/2 or 4.7 x 10 cm. The average dif- -3 -3 fusion velocity is approximately 4.7 x 10 cm/sec and thus about half of -3 the atoms within 4.7 x 10 cm of the surface in any one second will reach \\ it and escape. The diffusion of argon across the water-air interface would constitute a source of

Appendix A Page 11 2=0.88(AXfat)N 5 S A x 2.5 x 10 47 g = 0.88 x 2 x 4.67 x 104 = 0.24 x 100 atoms /s where A is the surface area of the tank. The total source is 3 S=S3+S2 = (2.1 + 0.24) x 100 = 2.34 x 100 atoms /s = 1.17 pCi/s This is consistent with measurements of Ar-41 release rates at similar TRILA facilities. The Ar-41 concentration in the reactor room and building exhaust air is given by SIY 8 N _ R S 2.34 x 10 R ~ A + q/V AYR*9 1.04 x 10 x 2 x 109 + 1,68 x 106 ~ R ~4 3 3 = 123.9 atoms /cm s 124 atoms /cm 9 3 where the room volume is 2 x 10 cm and the room air exhaust rate is 6 3 1.68 x 10 cm /sec. This corresponds to an activity concentration of A=$=1.06x10 x m = 3.55 x 10 pCi/cm. 3 4 3.7 x 10 The MPC for Ar-41 in air is 2 x 10-6 3 pCi/cm in the restricted areas. Therefore, the calculated concentration is below the MPC limit. Using design features of other operating TRIGA facilities such as, stack dilution factor, different q/v rates and sweeping air across the tank surface, the final design can achieve the goals established in Section A, Part 3.2. This calculation results in reactor room concentrations thet are a factor of 10-100 too high. This is based on the many years of operating experience at a number of operating TRIGA facilities and their officially reported annual dose to reactor workers.

Appendix A Page 12 The actual effect of Ar-41 releases from the reactor tank would be substantially less than those estimated due to the conservative assumptions. The major conservative assumption being th transfer amounts of argon from the tank surface and the release rates and volumes. 4.0 Total Ar-41 Release from Normal Operations Assuming an Ar-41 concentration of 3.6 x 10~7 3 pC1/cm and an activity ~7 3 6 3 discharge rate Aq = (3.6 x 10 pCi/cm ) (1.68 x 10 cm /s) =.605 pC!/s, we can show that the SNRS goal of releasing no more than 2 x 10-8 3 pCi/cm to the unrestricted area is achievable. Using the analysis in Ref. 33 and the following assumption, the table shows-the maximum release, q = 1.68 x 106 cm /s 3 Q =.605 pCi/s DA = Atmospheric dilution factor h = Initial height of rise of the cloud (stack height of

60 ft)

X = Maximum concentration point downwind from the release X = Maximum ground concentration Atmospheric conditions are classified as average, stable, and unstable. Parameter Average Stable Unstable X (Max) Meters 208 4076 75 h (m) 19.11 19.89 18.68 DA (m /s) 6082 3378 24432 3 -11 -10 X (pCi/cm ) 9.95 x 10 1.79 x 10 2.47 x 10'11 -These concentracion values agree well with the analysis done in Ref. 33. 5.0 Nitrogen-16 Activity In The SNRS Enclosure The cross-section threshold for oxygen-16 (n.p) nitrogen-16 (N-16) reactions is 9.4 MeV; however, the minimum energy of the incident neutrons must be about 10 MeV because of center of mass corrections. This high

AppenJix A Page 13 threshold limits the production of nitrogen-16 since only about 0.1% of all fission neutrons have an energy in excess of 10 MeV. Moreover, a single hydrogen scattering event will reduce the energy of these high-energy neutrons to below the threshold.- The effective cross section for oxygen-16 (n.p) nitrogen-16-reaction averaged over the TRIGA spectrum is 2.1 x 10-29 2 This value agrees well with the value obtained from integrating the cm. effective cross section over the fission spectrum. 3 The concentration of N-16 atoms per cm of water as it leaves the reactor core is given by 0 &N, N" = (1 - e **) V 3 where 3 N = Nitrogen-16 atoms per cm of water 13 2 $y = Neutron flux (0.6-15 MeV) - 1.0 x 10 n/cm -s at 1000 kW O = 0xygen atoms per cm3 22 3 N of water = 3.3 x 10 atoms /cm a = (n,p) cross section of oxygen = 2.1 x 10'29 2 (averaged cm over 0.6-15 MeV) -2 -1 A ' Nitrogen-16 decay constant = 9.35 x 10 3 t = Average time of exposure in reactor v = Flow velocity (15.2 cm/s) NN = 0.74 x 108 (y,,-(9.25 x 10 x 2.5) = 0.15 x 108 3 atoms /cm as the density of N-16 in the-water leaving the core. .If it is assumed that the water continues to flow at the same velocity to-the surface, the transit time from core to surface is 16 x 30.5 cm/ft = 32 s trise = 15.2 cm/s 4 ?

Appendix A Page 14 This assumption is'quite conservative as energy losses from the fluid stream resulting from turbulent mixing will reduce the velocity signifi-cantly. Furthermore, delays in transit time resiting from operation of the diffuser pump are sizeable. Measurements made of the dose rates at the tank surface of several TRIGA reactors show that the operation of_the diffuser pump reduces the N-16 contribution to the surface dose rate by an order of magnitude or more depending on the size of the tank. -9.35 x 10 x 32 = 0.050 times the value In 32 s the N-16 decays to-e of the activity leaving the core. Thus the concentration of N-16 atoms that reach the region near the surface of the tank is no greater than about 5 3 [ 7.5 x 10 atoms /cm. Only a small proportion of the N-16 atoms present near the-tank surface are trouferred into the air of the reactor room. When a N-16 atom is formed, it appears as a recoil atom with various degrees of ionization. For high purity water (* 2 pmho) practically all of the N-16 combines with oxygen and hydrogen atoms of the water. Most of it combines in an anion form, which has a tendency to remain in the water (Ref. 22). It is-assumed that at least one-half of all ions formed are anions. Because of its 7.4 s half-life, the N-16 decays before reaching a uniform concentration in the tank water. The activity will be dispersed over the surface area of the tank and much of it will decay during the lateral movement. For the purpose of the analysis, ite is postulated that the water-bearing N-16 rises from the core to the surface an$ then spreads across a 4 2 disk source with a radius of 122 cm s (4 ft), and area A = 4.67 x 10 cm, s For a constant velocity of 15.2 cm/see the cycle time for distributing the N-16 over the tank surface would be: t = 122 cm/15.2 cm/s = 8 s s

Appendix A Page 15 The average concentration during this time is ts ~ 5 S= [N"e'At dt) /t* = At -2 (1 - eat,) 7.5 x 10 s 9.35 x 10 x8 -9.35 x 10 (1 - e x 8) = 5.2 x 105 3 atom /cm. l The thickness of the layer of N-16-bearing water is: 3 I h= 5 = L x 10 x 10 = 1.49 cm s 4.67 x 10 The dose rate at the tank surface arising from the N-16 near surface is D= 1 - E2 (ph) Where p is the attenuation coefficient for 6 MeV photons in water with a 4 value of (0.0275 cm'1), K is the flux-to-dose-rate conversion (1.6 x 105 photons j rad) 2 cm sec and E is the second exponential function. This yields p D = 248 mr/hr This'is a value one would predict from extrapolation'of measurements.made on other.TRIGA reactors. The interest from the point of safety is then the number of N-16 atoms escaping into the air from the diffusing surface source above the core. The number escaping to the air would be about 5 3 ~2 3 2 (5.2 x 10 atoms /cm ) (0.9 x 10 cm/s) = 4.7 x 10 cm -s

Appendix A Page 16 -2 where the escape velocity, 0.9 x 10 cm/s, is from Ref. 20. In the room, the activity is affected by dilution, ventilation, and decay. Thus the rate of accumulation of N-16 in the enclosure as a whole is given by d(VN16), 16 , qf ) y Where: S = Number of N-16 atoms entering the enclosure from the tank per 3 2 4 2 8 s'I second (4.7 x 10 /cm s) (4.67 x 10 cm ) = 2.2 x 10 9 V = Volume of the reactor enclosure = 2 x 10 cm 6 c,373, q = Volume flow rate from the reactor enclosure exhaust 1.68 x 10 For saturation conditions, 8 VN16, S 2.2 x 10 9 = 2.3 x 10 nuclei. A + q/V, 9.35 x 10-2 + 8.4 x 10~4 -6 3 This corresponds to an activity concentration of 5.7 x 10 pCi/cm. The gamma dose rate from this concentration of N-16 in the air is 4 tm -6 (pCi/cm ) x 0.78 x 103 3 3.7 x 10 x 5.7 x 10 cm 2 x 1.6 x 105 (photons /s cm / rad /hr) 2 -3 = 0.51 x 10 rad /hr = 0.51 mr/hr when the offective radius of the enclosure, taken to be a hemisphere with a 9 3 volume of 2 x 10 cm, Again, with the diffuser operating, the dose rate from N-16 in the enclosure would be even smaller. Exposure to the public is negligible because of the rapid decay of N-16.

e-l _ APPENDIX B q ANALYSIS OF EXPECTED RADIOLOGICAL DOSES FROM NORMAL AND ABNORMAL OPERATION

APPENDIX B ANALYSIS OF EXPECTED RADIOLOGICAL DOSES FROM NORMAL AND ABNORMAL OPERATION 1.0, INTRODUCTION Over 50 reactors throughout the world, including 28 in the U,S., use the TRIGA-type uranium fuel with' zirconium hydride moderator. Many of them I have been in operation in densely populated areas for periods of over 20 years. All have an excellant safety rscord with insignificant radic- . logical impact. Potential radiation doses for this environmental assessment are based on operating experience of similar reactors. The potential radiation doses in this appendix have been evaluated using very conserva-tive assumptions. These assumptioni are made to allow the envelope for the final SNRS design parameters to be as flexible as possible while maintaining i l a high margin of safety for the operation. A detailed evaluation, utilizing design-specific parameters, will be performed in the safety analysis report (SAR) to be prepared for this project. Potential radiation doses can arise through two means: (1) from normal operation, and (2) from abnormal operation, i.e., accidents. These two means are addressed in the following subsections.

1. 0 NORMAL OPERATION The equilibrium concentration of Ar-41 released from the reactor tank l

and radiography bays has been estimated to be about 2 x 10-8 pCi/cm3 (see Appendix A), This estimate may be a factor of ten high based on measure-ments made at operating TRIGA facilities and their officially reported . annual radiation dose to reactor workers. An SNRS worker submerged inside the enclosure'in a cloud of Ar-41 at concentration of 2.0 x 10-2 pCi/m3 (2 x 10'0 3-pCi/cm ) would be expected to receive about 50 mrem (whole body) during the year (see Table 1 Section B, page 17). This estimate is based upon the following assumptions:

Appendix B" Page 2 (1) Exposure occurs for 2000 h per year. (2) A dose conversion factor for Ar-41 of 8.84 x 10'3 mrem m /pCi yr 3 (Ref. 23). (3) Concentrations of 2 x 10 2 x 10'9 pCi/cm are typical room measured concentrations. b Workers outside the SNRS enclosure would receive lower doses because: (1) Ar-41 releases would be from a position many feet above ground, reducing ground-level concentrations near the enclosure. (2) Air turbulen',e would disperse the Ar-41 rapidly as it moved from the SNRS c.1 closure. (3) Nonradiation workers would spend only a portion of the 2000-h work year near areas of highest concentration. Assuming 365-day persistence of the worst-case, 24-h meteorological condi-tions used below for the accident scenarios, offsite (Watt Avenue) indi-viduals would receive no more than 0.1 mrem per year from normal operation of SNRS. 3.0 ABNORMAL OPERATION - ACCIDENTS Numerous safety committees that have reviewed TRIGA reactor operations have considered potential accidents including rapid insertion of reactivity, loss of coolant, metal-water reactions, rearrangement of fuel, fuel aging, and handling of irradiated fuel. The safety analysis report (SAR) for the proposed SNRS will include analysis of all such credible accidents. In this environmental assessment, the maximum credible accident is addressed! (1) a fuel-handling or one element accident -- the typical design basis accident for a TRIGA reactor. This accident is analyzed to examine potential radiation doses to the general public from a very severe accident. Less severe accidents will result in much lower doses. 3.1 One-Element Accident (Maximum Credible Accident) At some point in the lifetime of the SNRS reactor, used fuel within the core may be moved to new positions or removed from the core.

l Appendix B Page 3 = The fuel elements are only moved during periods when the reactor is shut do'.n. - The most serious fuel-handling accidents involve spent or used fuel that has been removed from the core and dropped or otherwise damaged, causing a release of fission products. The accident described herein assumes that the SNRS reactor was oper-ated continuously at 1 MW for 1 yr prior to shutdown, or 365 MW-days. This operating history, although possible, is extraordinary in view of the proposed facility mission and resources. It is estimated (Ref. 4) that this assumption introduces conservatism of one to two orders of magnitude. For this accident, the fuel element is assumed to be damaged during handling l in air and to lose the entire cladding, therefore releasing a fraction of the total fission product inventory as shown in Table B.1. This fissien-product inventory uses the Blomeke and Toda r.ata (Ref. 5 for gaseous fission products, and Ref. 4 for the solid fission product inventory). Both refer-ences assume infinite operation at 1 MW. Table B.1 is a compilation of the volatile gamma emitters in a fuel element run to saturation with the highest core power density. 4 All of the radioactivity would not be released from the element. The fuel matrix acts to strongly retain fission products. Foushee and Peters (Ref 6) have collected and summarized empirical data regarding release ' actions. Even with unclad, heated, irradiated fuel, the fraction of -5 [ gaseous activity released was only about 1.5 x 10 Release of nongaseous fission products is typically one or more orders of magnitude lower (Ref. 6), and thus exposures from this source are secondary to exposure from gaseous radionuclides. Further, it is assumed that the release is a ground release 1_ with no credit for containment delay. The model employed in predicting the atmospheric dispersion of radio-nuclides from both accidents is PAVAN (Ref. 24). This model is used reg-ularly by the NRC to estimate downwind ground-level air concentrations for potential accdental releases of radioactive material from nuclear power and _research reactors. The computer model implements the guidance provided in Regulatory Guide 1.145, " Atmospheric Dispersion Models for Potential Acci-C

.~ Appendix B' Page 4 j dental Consequence Assessments at Nuclear Power Plants" (Ref. 25), The technical basis for' Regulatory Guide 1.145 is' presented in Ref. 26, PAVAN computes X/Q at any offsite~ location where X is the ground-level 3 airborne concentration of'a specific radionuclide in Ci/m and Q is the i radionuclide release rate in Ci/s. X/Q calculations are made based on the theory that materi31 released to the atmosphere from the accident will be distributed in a Gaussian pattern along the plume centerline. PAVAN assumes no radioactive decay nor deposition from the plume; these assumptions are very conservative. For calculating the dispersion of radionuclides, the following meteorological data are entered into PAVAN: (1) Hourly average wind direction over the period of record (2) Hourly average wind speed (3) Hourly average vertical atmospheric stability. 4 For the present application, these meteorological data were obtained from j-the National Weather Service Station at Sacramento City Executive Airport, approximately 13 mi (20.9 km) southeast of McClellan Air Force Base. Model calculations were made using the following options within PAVAN: (a) The radionuclide releases from either accident would occur-over a 24-h period (the model assumes that plume passage by any fixed ground location would be over a 24-h period as well, and includes the puff release at the very beginning of the accident). (b) The releases are at ground level with no building wake effects that might improve plume mixing. 4 -(c) There is no variation in terrain elevation in the region of interest.

Appendix B Page 5 The X/Q value at any ground-level location, R (defined by distance and direction from source) is computed within PAVAN using the following three steps: (1) An annual average X/Q is computed at R, assuming continuous release over a 1 yr period. (2) A short term X/Q value at R is computed for a 0 to 2-h interial. (3) The ainual average X/Q value is then used with the 0 to 2-h x/Q value in a logarithmic interpolation scheme to determine the X/Q value representative of the 24-h release period. Model predictions of X/Q for each accident were prepared for 16 directions (N, NNE, NE, ENE, E, etc.) for each of several distances: 330 ft, 1300 ft, 3000 ft, 4500 ft, 1 mi, 10 mi (100 m, 400 m, 910 m, 1372 m, 1610 m, 16,100 m). The x/Q values represent the 99.5 perce cile l value -- i.e., it is 99. 'Y probable that the actual X/Q associatea with an accident values will be less than this value. Estimated 99.5 percentile x/Q values are presented in Tables B.2 through B.7. Activity concentrai. ions of dispersed radionuclides were calcu-lated for each X/Q value, assuming a continuous 24-h release of the radio-nuclide fraction freed from the fuel element. Concentrations at each location were obtained by multiplying X/Q by the release rate. From these concentrations, whole-body and adult thyroid doses were calculated using the dose-conversion factors for radionuclides given in Regulatory Guide 1.109 (Ref. 23). These dosage factors are reproduced in Table B.8. The inhalation rate of a resting adult human being was used in calculations for all radionuclides except the noble gases (krypton and xenon radionuclides). No inhalation rate is needed for noble gas contributions to whole-body (or thyroid) dose. The model assumes that the concentration at each location remains constant aver the 24-h period of release, and that the receptor receives exposure for the entire 24-h period. Thus, doses are calculated 1 .J

Appendix B' Page 6 on the assumption that the receptor is exposed to the entire fraction of the released activity that can be expected to pass through a given location over the period of radionuclide release. The results of the atmospheric transport and dose predictions appear in Tr. oles B.2 through B.7 for doses received by the whole body and thyroid. The doses presented are based on those 99.5 percentile values occurring over a 24-h period at any location. The radiation doses to the general > oblic as a result of the one-element accident are shown to be small. The estimated maximum thyroid dose of 0.18 mrem is received at a distance of 4500 ft (1400 m) to the north of the reactor (Table 2, Section 3.2.2.1). The thyroid dose results from inhalation of radioactive iodine isotopes. This dose is far lower than the limit in ANS 15.7 of 1500 mrem thyroid. The estimated maximum whole-body dose is 0.002 mrem at the same location (Table 2, Section 3.2.2.1). This is far lower than the ANS 15.7 limits of 500 mrem whole body. Because these doses were calculated using conservative assumptions, the actual doses would be lower i f such an accident occurred. ._________.__.____.__m_

_, ~ ' Appendix B-Page 7 Table B.1. Source' Terms for One-Element Accident" ' Group I-Group II Group III

Activity, Activity,

Activity, Nuclide

-(C1) Nuclide (C1) Nuclide (C1) . B r-83 76.71 Kr-83M 76.73 Sr-89 555.55 Br-84 -176.18 Kr-85M 238.55 Sr-90 17.22 Br-84M : 3.63-Kr-85 47.94 Sr-91 717.22 Br-85 238.58 Kr-87 433.37 Sr-92 812.22 -Br 429.00-Kr-88 593.42 Sr-93 921.11-I-129 159.86 Kr 736.94 -Cs-134M 1.00 I-131 463.63-Kr-90 828.67 Cs-134-1.67 I-132 702.94 -Kr-91 493.27 Cs-136 14.44 ' I-133.. 1035.84 Xe-131M~ 4.63-Cs-137 275.55 i- -I-134 1213,99 Xe-133M 24.95 Cs-138 1144,44 I-135 944.11 Xe-133 1036.78 I-136 495.77 Xe-135M 283.26 Xe-135 765.34 Xe-137 94.41 Xe-138 876.72 Xe-139 906.98 Xe-140 952.22

Tota _1 5940.24 8394.18 4460.98-

". Release-fractions that were used in the accident scenario are 1.5 x 10 -5 -6 for Groups I-and II, and 1 x 10 for Group III. l m

C Appendix B' Page 8 . Table'B.2. Estimated Maximum Whole-Body'and Thyroid Doses et a Distance of 330 ft (100 m)-from the SNRS Reactor for a Single fuel-Element Failure i Direction. x/Q Whole-Body Dose Thyroid Dose From Site (s/m )* (mrem)* (mrem)* 8 j S 2.05E-03 7.60E-02 7.39 SSW

1. 06E-03 3.93E-02 3.82 SW 1.11E-03

~4.11E-02 4.00 q_ WSW 1.64E-03 6.08E-02 _5.91 W 2.50E-03 9.27E-02 9.02 WNW 2.64E-03 9.78E-02 9.52 NW 2.75E-03 1.02E-01 9.92-NNW 2.57E-03 9.53E-02 9.27 N 2.82E-03 1.04E-01 10.17 NNE 2.40E-03 8.90E-02 8.66 7 i NE 2.27E-03 8.41E-02 8.19 ENE-1.57E-03 5.82E-02 5.66 E 1.52E-03 5.63E-02 5.48-ESE 1.42E-03 5.26E 5.12 SE 1.81E-03 6.71E-02 6.53 c SSE -2.20E-03 8.15E-02 7.93- -3

1.00E-03 is equivalent to 1.00 x 10

__._._______________m_ ._________m_.

Appendix B Page 9 Table B.3. Estimated Maximum Whole-Body and Thyroid 00ses at a Distance of 1300-ft (0.40 km) from the SNRS Reactor for a Single Fuel-Element Failure l Direction X/Q Whole-Body Oose Thyroid Oose From Site (s/m3)* (mrem)* (mrem)* 4 S 1.96E-04 7.27E-03 7.07E-01 SSW -1.00E-04 3.71E-03 3.61E-01 SW 1.06E-04 3.93E-03 3.82E-01 WSW 1.60E-04 5.93E-03 5.77E-01 W 2.44E-04 9.05E-03 8.80E-01 WNW 2.57E-04 9.53E-03. 9.27E-01 l NW 2.67E-04 9.90E-03 9.63E-01 NNW 2.49E-04 9.23E-03 8.98E-01 N 2.72E-04 1.01E-02 9.81E-01 NNE 2.30E-04 8.53E-03 8.30E-01 NE 2.18E-04 8.08E-03 7.86E-01 ENE 1.48E-04 5.49E-03 5.34E-01 E 1.43E-04 5.30E-03 5.16E-01 4 ESE 1.34E-04 4.97E-03 4.83E-01 SE 1.71E-04 6.34E-03 6.17E-01 SSE 2.10E-04 7.79E-03 7.58E-01 -3

1.00E-03 is equivalent to 1.00 x 10 4

App 3ndix B - i Page 10 Table B.4. Estimated Maximum Whole-Body and Thyroid Doses at a Distance of 3000 ft (910 m) from the SNRS Reactor for a Single Fuel-Element Failure Direction X/Q Whole-Body Dose Thyroid Dose 3 From Site (s/m )*' (mrem)* (mrem)* S 5.33E-05 1.98E-03 1.92E-01 SSW 2.63E-05 9.75E-04 0.49E-02 SW 2.78E-05 1.03E-03 1.00E-01 WSW 4.34E-05 1.61E-03 1.57E-01 W 6.83E-05 2.53E-03 2.46E-01 WNW 7.19E-05 2.67E-03 2.59E-01 f NW 7.42E-05 2.75E-03 2.68E-01 NNW. 6.88E-05 2.55E-03 2.48E-01 N -- 7.50E-05 2.78E-03 2.71E-01 NNE 6,23E-05 2.31E-03 2.25E-01 NE 5.91E 2.19E-03 2.13E-01 ENE 3.92E-05 1.45E-03 1.41E-01 E 3.76E-1.39E-03 1.36E-01 ESE 3.51E-05 1.30E-03 1.27E-01 SE 4.58E-05 1.70E-03 1.65E-01 SSE 5.69E-05 2.11E-03 2.05E-01 -3 1;00E-03 is equivalent to 1.00 x 10

App:ndix B-Page 11 Table B.S. Estimated Maximum Whole-Body and Thyroid Doses at a Distance of 4500 ft (1372 m) from the SNRS Reactor for a Single Fuel-Element Failure l Direction X/Q Whole-Body Dose Thyroid Dose From Sito (s/m )* (mrem)* (mrem)* a S 3.37E-05 1.25E-03 1.22E-01 SSW 1.47E-05 5.45E-04 5.30E-02 SW 2.57E-05 5.82E-04 5.66E-02 WSW 2.70E-05 1.00E-03 9.73E-02 W 4.69E-05 1.74E-03 1.69E-01 WNW 4.85E-05 1.80E-03 1.75E-01 NW 4.92E-05 1.82E-03 1.77E-01 NNW 4.50E-05 1.67E-03 1.62E-01 N 4.91E-05 1.82E-03 1.77E-01 NNE 3.96E-05 1.47E-03 1.43E-01 NE 3.74E-05 1.37E-03 1.33E-01 ENE 2.32E-05 8.60E-04 8.37E-02 E 2.21E-05 8.19E-04 7.97E-02 ESE

2. 0'JE-05 7.52E-04 7.32E-02 SE 1.03E<03 1.00E-01 J. ash-f6 1.13E-03 1.29E-01

, - - _ _ _ _. _ _ ~ _.. _... ~. _. 8 N A 4 t c

Appendix B-PagaL12 Table B.6. Estimated Maximum Whole-Body and Thyroid Doses at a Distance of 1 mi (1.6 km) from the SNRS - Reactor for a Single Fuel-Element failure Direction X/Q Whole-Body Dose Thyroid Dose From Site (s/m )* (mrem)* (mrem)* 3 S 2.84E-05 1.05E-03 1.02E-01 SSW 1.17E-05

4. 34E-04 4.22E-02 SW 1.27E-05 4.71E-04 4.58E-02 WSW 2.26E-05 8.38E-04 8.15E-02 W

4.09E-05 1.52E-03 1.48E-01 WNW -4.20E-05 1.56E-03 1.52E-01 NW 4.23E-05 1.57E-03 1.53E-01 NNW 3.85E-05 1.43E-03 1.39E-01 N 4.20E-05 1.56E-03 1.52E-01 NNE 3.34E-05 1.24E 1.20E-01 NE 3.16E-05 1.17E-03 1.14E-01 ENE 1.90E-05 -7.04E-04 6.85E-02 E 1.81E-05 6.71E 6.53E-02 .ESE 1.65E-05 6.12E-04 5.95E-02 SE-2.31E-05 8.56E-04 8.33E-02 SSE 3.02E-05 1.12E-03 1.09E-01 3

1.00E-03 is equivalent to 1.00 x 10 m.__--mmm__-__

a.__-__ _m.-__e__.m---.___________.o:.____.m__ _a_-.___

~. -. Appendix B ze. Page-13 i Table.B 7. Estimated Maximum Whole-Body and Thyroid Doses at a Distance of 10 mi (16 km) from the SNRS Reactor for a Single Fuel-Element Failure Direction-X/Q Whole-Body Dose Thyroid Dose 'From Site (s/m )* (mrem)* (mrem)* 3 S '2.42E-06 8.97E-05 8.73E-03 SSW 5.87E-07 2.18E-05 3.12E-03 SW

6. 66E-07 2.47E-05 2.40E-03 WSW 1.80E-06 6.67E-05 6.49E-03 W

5.11E-06 1.89E-04 1.84E-02 WNW 4.86E-06 1.80E-04 1.75E-02 NW 4.51E-06 1.67E-04 1.63E-02 NNW 3.84E-06 1.42E-04 1.39E-02 4 N 4.20E 1.56E-04 1.52E-02 NNE 2.88E-06 1.07E-04 1.04E-02 NE 2.7CE-06 1.00E-04 9.74E-03 .ENE 1.21E-06 4.49E-05 4.37E-03 E 1.11E-06 4.11E-05 4.00E-03 ESE 9.40E-07 3.48E-05 3.39E-03 SE 1.65E-06 6.12E-05 5.95E-03 SSE 2.51E-06 9.30E-05 9.05E-03 -3

1.00E-03 is equivalent to 1.00 x 10 1

4

Appendix B* Pag 2 14 I t f Table B.8. Dose-Conversion Factors for Radionuclides 3 A. Dose-Conversion Factors (mrem m /pCi yr) for Noble Gases and Daughters (Whole Body Only) Radionuclide Dose Factor -8 Kr-83M 7.56 x 10 -3 Kr-85M 1.17 x 10 -5 Kr-85 1.61 x 10 -3 Kr-87 5.92 x 10 -2

r-88 1.47 x 10

-2 Kr-89 1.66 x 10 -2 Kr-90 1.56 x 10 -5 'e-131M 9.15 x 10 ~4 Xe 133M 2.51 x 10 ~4 Xe-133 2.94 x 10 -3 Xe-135M 3.12 x 10 -3 Xe-135 1,81 x 10 -3 Xe-137 1.42 x 10 Xe-138 8.83 x 10'3

Appendix B Page 15 Table B.9. -Dose-Conversion Factors for Radionuclides (contd) B. Inhalation-Dose Conversion Factors for Adults (mrem pCi inhaled)" "1 Radionuclide Whole-Body Dose Factor Thyroid Dose FactorD -8 - Br-83 3.01 x 10 N/A -8 Br-84 3.91 x 10 N/A ~9-Br-85 1.60 x 10 N/A -6 ~3 1-131 -2.56 x 10 1.49 x 10 ~7 -5 I-132 1.45 x 10 1.43 x 10 ~7 ~4 I-133 5.65 x 10 2.69 x 10 -8 -6 I-134 - 7.69-x 10 3.73 x 10 ~7 -5 I-135 3.21 x 10 5.60 x 10 -6 Sr 1.09 x 10 N/A ~4 Sr-90 7.62 x 10 N/A -10 St-91 3.13 x 10 N/A ~4 Sr-92 3.64 x 10 N/A -5 Cs-134 9.10 x 10 t;/A -5 Cs-136 1.38 x 10 N/A- -5 .Cs-137 5.35 x 10 N/A- -8 Cs-138-4.05 x 10 N/A a Human breathing rate taken to be 2.32 x 10~4 3 m /s ib N/A_ denotes not applicable

9 4 APPENDIX C LIST OF PREPARERS \\

Appendin C' Page 1 APPENDIX C. LIST OF PREPARERS 1. Wade J. Richards (B.S., M.S. Nuclear Physics). Twenty years of exper-ience in research reactor operations and analysis. Eight years exper-ence with TRIGA reactor design. 2. Lars F. Soholt (B.S., Ph.D. Biology). Fifteen years of research experience in wildlife ecology and environmental physiology, six years in assessment of impacts to terrestrial ecosystems. 3. William E. Stephens (B.S., M.E.). Twenty-six years of experience in design and operation of nuclear systems. 4. Anthony J. Policastro (B.S., M.S., Ph.D. Civil Engineering). Nine years of experience in meterological research and environmental impact assessment. 5. John M. Peterson (B.S., M.S. Nuclear Engineering). Three years of experience in nuclear reactor analysis, six years in environmental impact assessments of nuclear programs. 6. Steven Y. H. Tsai (B.S., M.S., Ph.D., Civil Engineering). Seven years experience in hydrologic analysis and environmental assessment. 7. John P. Barton (Ph.D, Nuclear Engineering). Twenty years experience in neutron radiography. Presently serving as a consultant to McClellan AFB. d

l D REFERENCES f i 1. Stationary Neutron Radiography System (SNRS-2): " Source Selection," Sacramento Air Logistics Center, Directorate of Maintenance, United States Air Force, December 1982. 4. Credible Accident Analyses for TRIGA and TRIGA-Fueled Reactors, HUREG/ CR-2387, Pacific Northwest Laboratory,1982. 5. J. O. Blomeke and Mary F. Todd, " Uranium-235 Fission Product Produc-tion as a Function of Thermal Neutron Flux Irradiation Time, and Decay Time," ORNL-2127, August 1957-November 1958. 6. F. C. Foushee and R. H. Peters, " Summary of TRIGA Fuel Fission Product Release Experiments," Gulf-EES-A10801, 1971. 7. Nati)nal Oceanic and Atmospheric Administration, " Climates of the States," 2 Vols, Second Edition, Gale Research Co., Detroit, 1980. l 8. U. S. Geological Survey, Sacramento, California, 1:250,000 Scale. Land Use and Land Cover and Associated Maps, Washington, D.C., 1979 9. U. S. Department of Commerce, " Statistical Abstract of the United States 1984," 104th Ed. 10. C. B. Hurt " Natural Regions of the United States and Canada," W. H. Freeman and Co., 1967. 11. "McClellan AFB Compatible Land Use Report," June 1983. 12. U. S. Soil Conservation Se.vice, " Soil Survey of Sacramento Area," Washington, D. C., 1954 13. U. S. Geological Survey, " National Atlas of the United States of America," Washington, D. C.,1970, p. 66, Major Recorded Earthquakes. 14. J.-H. Bennett, " Foothills Fault Systems and the Auburn Dam," Calif. Geology, August 1978. 15. Tousson R. Toppazada, at el., " Annual Technical Report - Fiscal Year 1980-1981, Preparation of Museismic Maps and Summaries of Reported Effects for Pre-1900 Calif. Earthquakes," September 1981. 16. Tousson R. Toppazada, " Annual Technical Report - Fiscal Year 1981-1982, Areas Damaged by California Earthquakes." 17. " Environmental Assessment for the Nondestructive Inspection Facility McClellan AFB," February 1984, i l

\\ 18. " Technology, Safety and Cost of Decommissioning icTerence Nuclear nesearch and Test Reactors," NUREG/CR-1756. 19. Staff Policy & Practice Statement, " Environmental Considerations Regarding the Renewal of Licenses for Research Reactor," Dec. 1980, 20. Dorsey, N. E., " Properties or Ordinary Water-Sebstances," Reinhold Publ. Co., New York, New York, p. 537-544. 21. Moore, W. J., Physical Chemistry, 3rd Edition, Prentice-Hall, New Jersey, 1962, p. 341. 22. R. L. Mitt 1, and M. H. Theys, "N-16 Concentrations in EBWR," Nucleonics, March 1961, p. 81. 23. U.S. Nuclear Regulatory Commission, " Regulatory Guide 1.109, Caclula-tion of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 20 CFR Part 50, Appendix I." NRC Office of Standards Development, Revision 1, October 1977. 24. Bander, T. J. " PAVAN: An Atmospheric Dispersion Program for Evaluat-ing Design Basis Accidental Releases o' Radioactive Materials from Nuclear Power Stations," NUREG/CR-2858 Pacific Northwest Laboratory, November 1982. U.S. Nuclear Regulatory Commission, " Regulatory Guide 1.145, Atmos-pheric Dispersion Models for Potential Accident Consequence Assess-ments at Nuclear Power Plants," Issued for Comment, August 1979. 26. Snell, W. G. and R. W. Jubach, " Technical Basis for Regulatory Guide 1.145, Atmospheric Dispersioi1 Models for Potential Accident Consequence Assessments at Nuclear Power Plants," NUREG/CR-2260, NUS Corporation, October 1981. D. U.S. Nuclear Regulatory Commission Title 10 Code of Federal Regulations (CFR) Parts 20 and 100, 28. Private Communications taken from NRAD Rea'.d.or Operating Data, July 1984. 29. Engineering Science, 1983, " Final Report Installation Restoration Program, Phase II,.... Confirmation," Vol. 1., June, Prepared for the U. S. Air Force, McClellan Air Force Base, Sacramento, CA, 10. American Nuclear Society Standards, "Research Reactor Site Evaluation," 1977. 31. 80erboom, A. J. H. and G. Kleyn, " Diffusion Coefficients of Noble Gases in Water," J. Chem Phys., V 50, No. 3, 1, Feb. 1969. 32. U. S. NRC Regulatory Guide 1.111, Revision 1, July 1977. 33. Oregon State University, Safety Analysis Report, Section 4.6.5, Aug. 1968. l}}

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