ML20212C819
| ML20212C819 | |
| Person / Time | |
|---|---|
| Site: | University of California-Davis |
| Issue date: | 07/26/1995 |
| From: | ARGONNE NATIONAL LABORATORY |
| To: | |
| Shared Package | |
| ML20212C798 | List: |
| References | |
| NUDOCS 9710300193 | |
| Download: ML20212C819 (33) | |
Text
{{#Wiki_filter:- -- I I Addendum i Envircomental Assessment for the McClellan Nuclear Radiation Center Reactor Operation at 2 MW1 1 Prepared by the Analysis and Data Management Section Engineering Division Argonne National Laboratory July 26,1995
' Prepared in accordance with the Council on Environmental Quality Reget.tions, 40 CFR.15001508; 43 FR 65978 56007, November 29,1978, amended July 30,1979; and with the Department of the Air Force Regulation for EnvironmentalImpact Anal-ycis Proccu, AFR 1902,10 August 1982 ko 7
[ l l l Addendum 1 Environmental Assessment for the McClellan Nuclear Radiation Center Reactor Operation at 2 MW1 Prepared by the Analysis and Data Management Section Engineering Division Argonne National Laboratory July 20,1995 5 Prepared in accordance with the Council on Environmental Quality Regulations, 40 CFR 15001508; 43 FR 55978-56007, November 29,1978, amended July 30,1979; and with the Department of the Air Force Regulation for EnvironmentalImpact Anal-ysis Process, AFR 1902,10 August 1982
'O I I Contents 1 l l Executive Summary 5 A Description of Proposed Action and Alternatives 7 A .1 Proposed A ction . . . . . . . . . . . . . . . . . . . . . . . . . . 7 j A.2 Description of the Proposed System ................ 7 A.2.1 G eneral . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 A.2.2 h1 N R C . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 A.2.2.1 Enclosur e . . . . . . . . . . . . . . . . . . . . . . 8 A.2.2.2 Reactor . . . . . . . . . . . . . . . . . . . . . . . 8 A.2.2.3 Cooling Systems . . . . . . . . . . . . . . . . . . 9 A.2.2.4 Confinement and Emergency Ventilation Featuren 10 A.2.2.5 Radiography llay and Equipment ........ 10 A.2.3 Design Envelope . . . . . . . . . . . . . . . . . . . . . . . 11 A.2.3.1 Design Requirements Affecting General Safety . 11 A.2.3.2 Design Requirements Affecting Environmental Assessment . . . . . . . . . . . . . . . . . . . . . 11 A.3 Site Description . . . . . . . . . ....,............. 12 AA htNRC Site Boundaries and Associated Dose Limits . . . . . . . 12 A.5 Alternatives to htNRC hiodification . . . ............ 13 D Environmental Consequences 17 D.1 Environmental Effect of hiNRC hiodification ........... 17 U.2 Environmental Effects of Operation . . . . . . . . . . . . . . . . . 17 B.2.1 Thermal Impact ....................... 17 11.2.2 Radiological Impact ..................... 18 U.2.3 Normal Operation . . . . . . . . . . . . . . . . . . . . . . 18 U.2.3.1 Fixed Parts . . . . . . . . . . . . . . . . . . . . . 18 l B.2.3.2 hiovable Parts . . . . . . , . . . . . . . . . . . . 18 B.2.3.3 G aseous Wast e . . . . . . . . . . . . . . . . . . . 19 B.2.3.4 Solid Waste . . . . . . . . . . . . . . .. ... 19 D.2.3.5 Liquid Waste . . . . , , . . . . . . . . . . . . . 20 ! D.2.3.6 Other Waste . . . . . . . . . . . . ....... 20 D.2.3.7 Cumulative Effects of Surrounding Operations . 21 U.2.3.8 Personnel hionitoring . . . . . . . . . . . . . . . 21 1
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e-B.2.3.9 Environmental hionttoring . . . . . . . . . . . . 21 B.2.3.10 Spent Fuel . . . . . . . . . . . . . . . . . . . . . 21 B.2.4 Abnormal Operations . . . . . . . . . . . . . . . . . . . . 22 i B.2.4.1 One. Element Accident (hlaximum Credible Ac. cident ) . . . . . . . - . . . . . . . . . . . . . . . . 22 B.2.4.2 hiultiple Element Failure in Air . . . . . . . . . 23 B.2.4.3 Other Accidents . . . . . . . . . . . . . . . . . . 24 B.2.5 Other impacts . . . . . . . . . . . . . . . . . . . . , , , 25 B.3 Resource Use . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 i B.4 Long. Term Aspects . . . . . . . . . . . . . . . . . . . . . . . . . . 25 B .5 ' Cost. Benefit . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -. 25 i i B .6 Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 i 1 l- C OfRees, Agencies, and Perrons Consulted 27 1 Appendix: List of Preparers (Addendum 1) 28 i l 2
t-( [ List of Tables 1 l B.1 Comparison of Predisted MNRC Maxhnum Annual Exposure . 20 B.2 Whole Body and Thyroid Doses from a One Element Accident (in the Direction of the Maximum Exposures,i.e., North of SNRS ( Enclosure) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23
- B.3 Whole-Body and Thyroid Doses from a One Element Accident (in the Direction of Cksest Base Boundary, i.e., East of SNRS Enclosu re) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 B.4 Doses for the 123 Element Case . . . . . . . . . . . . . . . . . . . 23 B.5 Tank Failure Event (30 Peak Elements) . . . . . . . . . . . . . . 25 l
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List of Figures A.1 Location of McClellan AFD . . . . . . . . . . . . . . . . . . . . . 14 A.2 SNRS Plan View - Main Floor . . . . . . . . . . . . . . . . . . . 15 A.3 Typical Mark-! TRIG A Reactor . . . . . . . . . . . . . . . . . . . 10 1 4
l l Executive Summary The proposed action is to modfly the existing Stationary Neutron Radiography l System (SNRS) to operate up to 2 htW. The facility is now known as the hic-Clellan Nuclear Radiation Center (h!NRC), and it is located within hicClellan Air Force Base 0.9 km west of Watt Avenue and 1.4 km south of E Street. The purpose of hfNRC is to provide the high sensitivity, high throughput inspection of aircraft components necessary to detect corrosion, thereby reducing crash risks, extending aircraft life, and reducing maintenance costs. The hlNRC also provides facilities for development of advanced solid-state materials, enhance-ment of boron neutron capture therapy (medical) research, and production of medical isotopes. The only alternative considered was no action. The facility is built around a TRIG A reactor, and the design is such that the environmental impact in normal operating conditions, or as a consequence of the maximum credible accident, is as acceptable as for the several other TRIG A reactors located in city centers and university campuses throughout the USA. The reactor operates at a nominal steady state power level of 2 htW or less. It also operates in the pulsing and squart-wave mode in accordance with the authorized limits. The fuel is enriched to less than 20% U 235, and the beam tubes do not penetrate the tank. The enclosure includes ventilation and filtered exhaust systems for both normal and worst credible accident situations. There is no additional construction required for this modification. Factors considered and found to have insignificant environmental impact include: water consumption, traffic, noise, chemical waste (solid, liquid, or gaseous), thermal waste, and radiological effects. The radiological effects are described below. Radioactive liquid waste is generated only in very small, low-level quantitles, and is recycled or evaporated such that no radioactive liquid waste will be released to the environment. Solid radioactive waste, typically consisting of low activity filters, is no greater than 1 m 8per year, and is shipped annually, according to existing regulations and agreements. Spent fuel will accumulate at one or two rods per year, and shipments will take place infrequently (e.g., each ten years). The fuel, loaned by the Department of Energy, will be returned to its custody. The principal radioactive gaseous wastes consist of small quantities of ni-trogen (N 16) and Argon (Ar-41) mixed with the air. For h!NRC personnel working inside the enclosure, the maximum exposure due to Ar-41 and N-10 is administratively controlled to be as low as reasonably achievable (ALARA), 5
( [ and well below the 5000 mrem /yr limits of Title 10 Code of Federal Regulations L Part 20. For people immediately outside the controlled h1NRC operations area, the maximum annual exposure would be less than 2 mrem /yr. To the general public outside the air base, the exposure rate is negligible for N 10, and less than ( 1 mrem /yr for Ar-41. The exposures are significantly below the guidelines set in both 10 CFR Part 20 and 40 CFR Part 61. These exposures may be comp ed with the exposure received by the general public from natural sources, which is s about 300 mrctn/yr for the United States population. The Nuclear Regulatory Commission has endorsed a generic statement for research reactors operated at or below 2 A1W in power which concludes that
*6 t iete will be no significant environmental impact. The h!NRC project has been l independently studied, and the findings support the same conclusion.
( l 6
b Section A Description of Proposed Action and Alternatives A.1 Proposed Action This Environmental Assessment, Addendum 1, addresses the modification and operation of the hicClellan Nuclear Radiation Center (hiNRC) reactor formerly known as the Stationary Neutron Radiography System (SNRS) at hicClellan Air Force Ilase, Sacramento, CA (Figure A.1). The hiNRC is needed to provide a high sensitivity inspection capability for detection of early stage corrosion in aluminum alteraft components, thereby re-ducing aircraft crash risk and reducing repair costs. Additionally, the facility provides irradiation services. The htNRC consists of a TRIG A research reactor below ground level with shielded radiography bays, parts-manipulation facilities, component-preparation areas, and associated radiography facilities. The pro-posed modification will consist of changing the reactor grid plates and increasing the maximum thermal output of the reactor to 2 h!W, The heat exchanger will be modified to accommodate the increased thermal lout. e A.2 Description of the Proposed System A.2.1 General The Environmental Assessment of h!NRC is a necessary prerequisite to place-ment of a contract for detailed design of the system modifications, This Environmental Assessment uses a general description of the system and a design envelope within which the detailed system design must fall. The h!NRC is designed such that, from the aspect of safety, it will be within the limits set by the numerous similar TRIGA reactors operating world wide and within the limits set by the Nuclear Regulatory Commission generic finding for environmentalimpact of research reactors at or below 2 h!W in power [19]. 7
I The Environmental Assessment does not attempt to address accident risks in detail, but simply assumes that the accident does take place. It will be a requirement of the design that nteident risks fall within this envelope, and this must be established in a Safety Analysis Report (SAR) prior to operation at 2 h!W after modification. A general description of the system is presented in A.2.2. below. This is followed by a summary of key requirements in A.2.3. l A.2.2 MNRC A.2.2.1 Enclosure l The htNRC enclosure is configured as shown in Figure A.'t Briefly, the en-closure consists of radiography and reactor associated rooms. The enclosure is designed to confine the results of any credible reactor accident; and,in addition to housing the reactor, it provides space for all radiography operations and parts handling. The enclosure is designed such that the ventilation systems for the reactor and radiography bay areas can be isolated f om the ventilation systems for the reactor control room and office area. 7 u. are no exterior conduits, pipelines, or electrical or mechanical structures attached to or adjacent to the enclosure, other than utility service facilities similar to those required for other Directorate of hiaintenance buildings on the base. The reactor control room is located in the hiNRC enclosure. All the process and reactor instrumentation have read-outs in the control room, so all aspects of reactor and radiography operation are monitored from this single location. A.2.2.2 Reactor The htNRC includes a TRIG A reactor (see Figure A.3), operable in the steady-state, square wave and pulse mode at power levels prescribed by the authoriza-tion documentation. The reactor core is cooled by light water, moderated by zirconium hydride and water, and reflected by water and graphite. The ba-sic components of the reactor include the fuel, control rods, control-rod drives, instrumentation, and control console. The core grid plate contains 121 fuel element locattuna. Ilowever, the core will be loaded with approximately 100 fuel elements for 2h1W operation with the remaining locations occupied by control rods, a central experimental facility and graphite dummy elements. Each fuel element contains a homogeneous mixture of uranium and zirconium hydride. The mixture contains 8-20 wt% uranium enriched to about 20E The fuel elements are approximately 710 mm (28 in.) long and 38 mm (1.5 in.) in diameter with 0.51 mm (.020 in.) thick stainless steel clad. This type of reactor fuel is classified as low-enriched uranium fuel (LEU) of low strategic significance; therefore, only minimal security and safeguards are required. The in-tank and out-of-tank fuel storage positions remain unchanged from the description in Ref. [2] The instrumentation and control systems contain the necessary equipment to operate and protect the reactor and monitor the radiography operation. The 8
I s [ hiNRC control console is a modern solid-state TRIGA design whose reliability L has been developed over 20-30 years of safe operation. The basic functions pro-vided by this system are reactivity regulation, neutron monitoring, instrumen-tation display, startup operation, steady-state operation, and scactor shutdown. ( Reactivity of the reactor cose is controlled by changing the pos!.* ... ,f the boron control rods. The MNRC incorporates three to five control rods that are approximately the same size and shape as a fuel element and that are sus- [ pended from fall-safe electromagnets located on the support Widge. A trer. den.t t control rod designal for pulsing is also provided. The reactivity control system is designed to prevent safety limits from being exceeded. The ionization and g fission chambers used for sensing neutron flux densities will be suspended near l the core. The control consolo is located in the reactor control room adjacent to the reactor room. The reactor core is located at the bottom of an aluralnum tank, which is approximately 2.4 m (8 ft) in diameter and 8 m (25 ft) deep. The reactor tank is below ground level and contains approximately 40,000 ( (10,000 gal) of demineralized water. The reactor tank provides about 5m (16 ft) of water over j the top of the reactor core for radiation shielding at the reactor top. The reactor i tank is su rounded b,s a concrete and soll biological shield. The concrete and soll provide the lateral shielding necessary for reactor and radiography operatious. There is no penetration of the reactor tank wall. A.2.2.3 Cooling Systeins Prlinary Systern The primary cooling system consists of a reactor tank, a primary pump, demineralizer pump, resin beds, particulate filter, heat ex-changer, make-up water system, and aluminum piping. The primary water is 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 is a low-temperature, unpressurized self-contained system. The water purification system consists of a pool skimmer on the water surface, a deminerallzer pump, whic. pumps the water through the demineralizer resin bed, and a particulate filter. This syatem acts to keep the primary water clean and free of activa-tion products. The primary system has 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 is readily available and is supplied from the lo-cal water supply. This water goes through a demineralizer system before being injected into the reactor tank. Secondary System The secondary cooling system consists of a secondary
- cooling pump, an auto,aatic flow-control valve, and a cooling tower. The secondary water flow rate is automatically controlled to keep the primary water rnu.ning to the reactor tank at a prescrii>ed temperature. The secondary water flows through the shell side of the heat exchanger, removes the primary water heat, and discharges the heat into the cooling tower.
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i b The secondary system pitssure is maintained higher than the primary system pressure. This will ensure that any leak in the heat exchanger will be hom the secondary side to the primary side, climinating the possibility of primary water getting out of the h1NitC enclosure. A.2.2.4 Confinement and Einergency Ventilation Features The h1NitC enclosure is designed with a ventilation system such that in nortnal operations, all exhaust takes place through a system for filtration and exhaust stack release. In addition, the h1NitC reactor room is designed to confine any tr.dk> active material relemed in an accident so that it may be exhausted in a controlled manner through an timergency cleanup and exhaust system. This system filters out the particulate and radioactive materials before the air is exhausted to the environment. The filtration system has, as a ruinimum, an in-place efficiency of 99.95% for 0.3-micron sized particles. The h1NRC reactor room ventilation system is automatically shut down and the stack exhaust damper closed, thus isolating the reactor room area, if a high-radiation alarm is received from the stack particulate or gaseous inonitor. The ventilation systein for the tractor control room and office area is separate from the reactor ventilation system. A.2.2.5 Radiography Bay and Equipment The hiNRC reactor is primarily used for neutron radiogrsphy of aircraft com-ponents. 'Ib provide this capability, four beam tubes are provided. The beam tubes approach the lower edge of the core tangentially, pass upward at approx. imately 45 degrees from the vertical, and terminate at floor level, such that the required scan area for large parts is available in the radiography bay. Each beam tube consists of two sections. The first section is in the reactor tank. The second section is outside the reactor tank, and it extends into the radiography bay. The beam tubes are constructed of aluminum and are capable of upport-ing collars, collimator inserts, apertures, filters, and beam shutters. Each beam tube is 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 are helium-filled to reduce nieutron attenuation and eliminate Ar-41 production close to the reactor corv. The area that immediately surrounds the reactor is divided into four shielded radiography bays to allow radiography on each of the four beams. The radio-graphy bays are equipped with manipulators, real-time imaging systems, and radiography control booths. The wall shielding and beam stops of each area are such that with any one beam shielded off, operators may enter and set up parts in that area while the other three beams are operational. Entry to the radiography bays is through two sets of doors separated by a staging area for components such that both sets of doors will not be open at the same time during normal operations. Doors to the radiography bays are interlocked with the beam shutters. The radiography doors have redundant 10 11 ~
I l interlocks, and entry into the bays is administratively controlled. The doors are sized for the largest collimator components, manipulator components, or aircraft components to enter the area. A.2.3 Design Envelope A.2 3.1 Design Requirements Affecting General Safety
- 1. The MNRC shall include a tank type TRIG A reactor, operable within the constraints of the authorization basis.
- 2. The hfNRC reactor grid plate shall consist of no more than 121 element locations, including control-rod, graphite, and fuel-element locations.
- 3. The htNRC reactor fuel shall be the TRIGA fuel type (8520%U), less than 20% enriched in U 235.
- 4. The MNRC reactor tank shall be below ground, constructed of aluminum, approximately 2.4 rn (8 ft) in diameter and approximately 8 m (26 ft) deep.
There shall be no penetration of the reactor-tank wall for beam tubes.
- 5. The in tank fuel stmage racks shall hold a maximum of 20 fuel elements each.
G. The MNRC Control System shall be of modern solid state TRIG A Golgn.
- 7. The primary coolant system shall have a siphon break to prevent water from being pumped or siphoned out of the reactor tank. The secondary pressure shall be higher than the primary system pressure.
- 8. There shall be no more than four beam tubes, each helium filled.
- 9. There shall be nc xterior conduits, pipelines, or electrical or mechanical structures, such as it ansmission lines attached to or adjacent to the enclo-sure, other than utility service facilities similar to those required for other i maintenance buildings on the base.
- 10. The MNRC shall be designed such that the risk of an aircraft accident afTecting the reactor core can be shown to be inconsequential or to be below a frequency of 10-7/yr, thereby meeting the requirements established in NUREG 0800 and DNL Design Guide 50331 I11.
A.2.3.2 Design Requirements Affecting Environinental Assessment l 1. The MNRC enclosure shall be designed to preclude the direct telease of I airborne radioactive materials or gases to the outside environment; in-l i stead, these materials shall be exhausted in a controlled manner through the MNRC filtration and exhaust stack. The exhaust shall be discharged at the top of the tallest nearby building, which will be the 18.3 m (60 ft) high roof of the NDI building and will be a restricted area. 11
b
- 2. The enclosure ventilation system for the reactor and radiography bay areas shall be isolated from those for the reactor control room and office arcan.
- 3. The enclosure shall include an emergency cleanup and exhaust system capable of confining the results of any crejilk accLient. The fhtration system shall have, as a minimum, an in place efficiency of 99.95'76 for 0.3-micron sized particles. The ventilation system shall be automatically shut down and the stack drunper closed if isting hicClellan Air Force Base policy and procedures.
- 4. The detailed design of the htNRC shall be st.ch tht radiological doses in normal and abno that may be found in the liquid waste storage tanks.
- 5. The h1NRC solid waste shall be handled in accordance with existing hic-Clellan Air Force Base policy and procedures.
G. The detalled design of the AINRC shall be such that radiological doses in normal and abnormal operations do not exceed those listed in 10 CFR Part 20. A.3 Site Description The site description is essentially unchanged from that give in Ref. [35). A.4 MNRC Site Boundaries and Associated Dose L11 nits Listed below are t he definitions of site boundaries and recommended dose limits taken from ANSI 15.7. Ar stated elsewhere in this report, doae rates will be limited to meet the guidelines of 10 CFR Part 20 and 40 CFR > art 61. In addition to ANSI 15.7. The areas for siting the h!NRC have been separated into three categories. The first area is the hf NRC operations area. This is the area inside the perime-ter fence (operations boundary) surrounding the h1NRC enclosure. The second area is the site area. This is the area outside the htNHC perimeter fence (op-erations boundary) but inside the perimeter fence surrounding hicClellan Air Force Base (site boundary). The third and final area is the urban area. This is the area outside the perimeter fence surrounding hicClellan Air Force Base (site boundary). The reactor administrator has direct authority over all activities in the op-erations area, and there shall be prearranged evacuation procedures known to those who frequent the area. In the event of a design-basis, one-element acci-dent (maximum credible acciderth the planning assumption shall be that all persons %in the operations area are evacuated in suflicient time 60 that the dose cou.: , mt does not exceed 25 rem to the "whole body" or 75 rem to any
^
"other org'-
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l ) The h1NRC site boundary is defined by the perimeter fence surrounding hicClellan Air Force Ilase. The nearest site boundary will thus be at Watt Avenue,915 rn (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 committnent for people within the site area shall not exceed 5 rem to the "whole body" or 15 rem to any "other organs" over a 2-hr exposure period. The h1NRC urban boundary is defined by the perimeter fence surrounding hicClellan Air Force Base. The urhan 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 hr using available rewurces. The dose commitment associated with the design basis accident for persons at or beyond the urban boundary shall not exceed .5 rem to the "whole body" or 1.5 rem to any "other organs" over a 24 hr exposure period. A.r> Alternatives to MNRC Modification A prime mission of hicClellan AFD is the maintenance of aircraft. This requires adequate methods of inspection. A particular problem is the early detection of hidden corrosion. If not detected and repaired, this corrosion can rapidly damage parts in service, causing structural failure, fuel leaks, fires, and loss of aircraft. No practical alternative inspection method is capable of detecting low level corrosion. The problem of corrosion is expected to increase with the age of an aircraft, such as the F111. The only alternative to increasing authorimi power to 2 htW is to continue operation at an authodzed power level of 1 h!W. 13
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Figure A.3: Typical Mark 1 TRIGA Reactor 16 m . .
1 Section B Environmental Consequences B.1 Environmental Effect of MNRC Modifica-tion Modification of the facility will simply consist of replacement of the reactor grid plates and addition of heat exchanger capacity. The action will I've no impact on the local aquatic and terrestrial blotic systems, including three'ened or endangered species; additionally, no impact to cultural or historical resources will arise from this modification [29). Itcactor fuel transport and loading will follow all applicable NRC/ DOE / DOT requirernents,10 CFil Parts 72 and 73, hansport and loading procedures are well established, and experience has shown that no threat to the public health and welfare will result from these activities. B.2 Environmental Effects of Operation D.2.1 Thermal Impact The 2 MW or less of heat generated by reactor operation will be dissipated to the atmosphere by means of a cooling tower. This rate of heat dissipation is comparable to that associated with light industrial operation and will have no affect on the local erwironment. Some rninor ground fogging may occur in the immediate vicinity, but this will not affect other maintenance operations or aircraft runways in the area. To place this level of heat dissipation in con-text, a typical central electricity generating plant rated at 1000 MWe will reject approximately 2000 MW af heat to the environtnent, 17
I B.2.2 lhidlological Impact Their are approxirnately $0 similar TitlGA reactors in operat!nn worldwide, including several nu the U.S. Many of thern have been in operation in deraely populated areas for periods of over 30 years. All operating TRIGA reactors have excellent records of safe operation with insignificant radiological innpact. The proposed MNRC operating schulule of two 8-hr shifts, five days a week, is more than enor.t typical TRIG A facilities operate, but as shown in Ref. [34], this leads to no radiological problem (Ref. [31] upos.tes the radiological analyses i given in Appendices A and 11 of flef. [2]). The potential for radiologicalimpacts on the enetonrnent is assessed in the following three circumstancer normal operation, aanormal operation (inaxl. mum credible accident), and abnormal operation (other accidents). Studies of numerous other TRIGA reactor systems have shown that the maximum credible accident, also termed design basis accident, is the rupture of a fuel element in air causing release of fission productr. immediately after maximum power irradiation [4]. The other accident category carnprises accidents that have been previously analyzed for TRIGA reactor:; results from some of these postualted accidents are included for completeness. B.2.3 Normal Operation D.2.3.1 Fixed Parts The reactor structural components near the reactor core that are exposed to in-tense neutron irradiation, are the reactor grid plates, control rod guides, control rods and in core instrument housings. To the maximum extent possible, these components are made of aluminum or other materials with attivation products that have relatively short. half lives for radioactive decay. Furthermore, these ac-tlvation products are distributed throughout and maintained within the volumes of these colid materials, so they are not readily released to the environment. Ex-posures from contact with in-core structural components are minimized because maintenance is performed through approximately 5 m (10 ft) of water. All expo-sures to personnel are rainimited through an ALARA (As Low As Reasonably Achievable) program and are far below the limits of 10 CFR Part 20. D.2.3.2 Movable Parts Some low levels of radioactivity cou'd be induced int. components being ra-diographed. Experience has shown taat induced artis ty of this type is very low. Allitems exposed to the beam wil be monitored before leaving the facility to ensure there will be no significant radiological impact on personnel or the environment. I 18
1 B.2.3.3 Gaseous Waste Small quantitles of radioactive gas are generated, the dominant one being Ar 41 which has a half life of 1.8 hr. Calculations of the Ar 41 effect are provided in lleference [34) for a power level of 2 hlW. Thne show a production rate of somewhat less than 2 Ci/s due to neutron irradiation of the water in the reactor core with subsequent relene at the water surface, and approximately 0.11 Cl/s due to neutron irradiation of the air in the beam path after emergence from the four helium filled beam tubes. The radioactivity of the atmosphere inside the enclosure will be controlled by the decay constant of Ar 41, the enclosure volurne, and the exhaust rate. This concentration inside the enclosure will be below the 10 CFit Part 20 limits for a restricted area (3 x 10-8 pC1/cm8 ) at all conditions except when the reactor is operated for extended periods at 2 MW; at this level, the average concentration is predicted to reach the 10 CFIt 20 thrnhold. The maximum concentration of Ar 41 at an accessible unrestricted area, outside the operations boundary, will be significantly below the 10 CFIt Part 20 limits or the 40 CFR 61 lhnits, due in part to dilution from the release point at the top of the 18.3 m (60 ft) high nearby building. The only other radioactive gas generated in incasurable quantities is N-
- 10. Calculations of the N 10 production are also revised in [34). With the assumptions of stated MNitC design characteristics and adequate ventilation, the exposure rate is about 1 miem/hr or 2000 mrem /yr (based on 2000 hr).
This is within acceptable limits for the restricted area (see footnote c Table 11.1) llowever, the N 10 presents an exposure risk only in the reactor room, where doses are controlled to be as-low as reasonably achievable. Outside the restricted area, N 10, because of its short half life, makes no impact. The combined maximum exposures for personnel outside the operations boundary are shown to be less than 2 mrem /yr. These are all significantly hss thar 'he 10 CFit Part 20 and 40 CFR Part 61 guidelines (see Table D.1.). An understanding of the unit of radiation exposure, mrem, can be pro-vided by comparison with the natural background radiation, which is about 300 mrem /yr per person averaged for the entire population of the United States. D.2.3.4 Solid Waste The ion-exchange resin and water particulate filters in the reactor water purifi-cation system will be routinely changed before high levels of radioactive mate-rials have accumulated (normally once a year), thereby minimizing personnel exposure. The resins and air filters contaminated with radioactive material will be packaged and shipped according to 10 CFR Parts 49, 71, and 179, and Air Force Technical Orders 00-110N 2 and 00-110N 3. The volume of such wastes is 8 not expected to be more than 1 m (five regular 55-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 contribute significantly to the volume of waste at the licensed disposal site. Potential impacts related to shipments of waste to the licensed disposal site have been considered and no significant impact is 19
l l h!NitC Predicted to CFit Part 20 40 Crit Part 01 hiaximum Dose
- Dose Limit Dose Limit iteceptor 6 (Whole Dody) (Whole Ilody) (Whole llody)
Occupational Personnel ALAllA ' 5000 mrem /yr - (Inside the Operations Doundary) 1 Other On base Personnel <2 mrem /yr 500 mrem /yr 10 mrem /yr (Detween the Operations and Site Iloundary) General Public <1 mrem /yr 500 mrem /yr 10 mrem /yr j (At the Urban Doundary)
- Originally calculated in Appendia A and outmequently revised in for 2 htW operation.
The actual values are expected to be significantly less than thme conservative calculations. 6 Occupational personnel are individuals directly involved with operation of the htNitC. Other on base personnel outside the h1NitC operations boundary are nasumed to be nonradi-allon workers; hence dose limit is the same as for the generai publ3.
'Drme rates due to Ar.41 and N.16 are predicted to average less than 2 mrem /hr assuming that ventilation provides complete mixing of air within the enclosure. In practice the venti-latlon will be such that areas frequented by htNRC workers have much lower exposure rates (typically 0.1 to O 2 mrem /ht). Annual exposures to htNRC personnel will be administratively controlled to be as low as reasonably achievable (ALARA), and well below the 5000 mrem /yr limits of 10 CFit Part 20, j
Table D.1: Comparison of Predicted h!NRC hlaxirnum Annual Exposure anticipated either as a result of normal or accident conditions. D.2.3.5 Liquid Waste The A!NRC will generate no radioactive liquid waste during normal operation. 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) returmd to the reactar water demineralizer system, or (b) solidified and the solid waste disposed of as described in Solid Waste above. Small amounts of nonradioactive solid-content water may be r. ased from the facility through the sanitary sewer during periodic blowdown of the cooling tower, 1s.2.3.0 Other Waste Other potential sources of radionuclides are neutron-activated impurities in the reactor tank water, and acthated airborne particulates (dust) from the radio-graphy bays. These sources are very small in comparison with those above for the following reasons:
. Che reactor-tank water will be circulated through a filter and demineral-izer system that will prevent the build-up of any cignificant r oncentration ofimpurities that could be acthated as the water passes through the core.
20
l Any material acthated will be trapped in the coolant purifier system and not released to the environment through any credible pathway.
. Any radioactive dust that might be altborne would be produced in the radiography bays. Airborne activity in the enclosure will be monitored, and corrective action taken if necessary. Furthermore, the air-exhaust system will be filtered.
The proposed action of increasing authorized reactor power level will not gen-l erate waste that would be subject to the Resource Conservation and Recovery Act. U.2.3.7 Curnulative Effects of Surrounding Operations The MNRC will produce immediately outside its operations boundary only a small fraction of the radiation exposure levels permitted in 10 CFR Part 20 and 40 CFR Part 61. The only other source of radiation near the MNRC is the NDI facility containing small X ray and N ray sources for radiography. These too are restricted areas, and produce significantly less than 10 CFR Part 20 and 40 CFR Part 61 limits on surrounding areas [17]. The denigns ensure that the cumulative effects remain well below acceptable radiation exposure limits. D.2.3.8 Personnel Monitoring Doses to MNRC workers are controlled and limited to levels well below those specUled by NRC regulations for occupattmal doses (i.e., whole body doses of 3 rem / quarter or 5 rem /yr). Workers p trained with regard to radiation risks 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. D.2.3.9 Environinental Monitoring An environmental monitoring system was initiated and maintained at the start of MNRC construction. The objective of this system was to ensure that MNitC air exhausted to unrestricted areas does not exceed applicable NRC guidelines or regulations (10 CFit Part 20 or 40 CFR Part 61). Based on the operating schedule of 16 hr/ day, five days a week at 2 MW, the MNRC will remain well below the NRC telease limits for normal operations. D.2.3.10 Spent Fuel Operation of the MNRC scactor will consume only a modest amount of fissile uranium in the fuel elements. Typically one new fuel element a year may be added to the core, with the spent fuel element put in an in-tank fuel-storage rack. Each in-tank fuel-storage rack will hold a maximum of 20 fuel elements (there is a total of five racks). It is anticipated that spent fuel shipments would 21
1 I be required once in 10-15 yr. These shipments will be made according to current DOE /NRC and DOT requirements. Since the fuel is DOE owned, the spent fuel is returned to DOE custody. B.2.4 Abnormal Operations Since TRIGA reactors have been operating for over 35 years, there have been many studies of possible accidents for this type of reactor [4). For the majority of TRIGA reactors, the fuel handling accident (design-basir accident) consisting of a one-element failure in air, has been identified as the "naximum credible accident." Therefore, the fuel-handling accident or the one-element accident has been analyzed for the h1NRC facility. The risk of accidents has been analyzed in the SAa [2] and revised in Addendurn 11 to the SAR [341. In this environmental assessment it is assumed that the design basis accident does take place. Since the off-site release frcan a small research reactor accident would be several orders of magnitude less severe than the criteria developed for power reactors, the conservative dose limits established in ANSI 15.7 will be used as the criteria for the h!NRC siting. These dose limits are given in Section A, Part 4. D.2.4.1 One-Elernent Accident (Maximurn 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, but fuel elements are never moved unlean tl~ reactor is shut down. The most serious one-element accident involves used fuel that has been removed from the core and dropped or otherwise damaged causing a release of fission products in air. Calculations of the fission product effect resulting from rupture of a fuel element in air are provided in Appendix D of the SAR for the stated design criteria. These reflected the original 1 A1W power limit; the release conditions have been revised to reflect 2 hlW in Addendum 11 to the SAR. The accident analyzed is conservative in that assumptions of an extremely unfavorable me-teorology and a large fission product inventory in the damaged fuel are made. Thus, the projected off site 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 atmosphere conditions (least dispersive) occur when the prevailing winds are from the r,outh to southwest. Thus, maximum exposure levels would be northward from h!NRC. The calculated whole b4 dose and thyroid dose from the maximum credible accident are ghen in Table II.2. l The nearest point of public access will be 910 m (3000 ft) east of the h!NRC. Doses received by members of the general public at this nearest point are shown in Table B.3. It can be seen that the doses are far below the limits for the urban l area in ANSI 15,7 of 500 mrem whole body and 1500 mrem other organs. These doses do not account for radioactive decay, containment of releases, on evacua. tion of the exposed individual before receiving the full released dose; hence, the 22
f* L f L [ hinximum Credible Accident (One-EJement Accident) Distance
- Whole-Dody Dose Thyroid Dose (North) (mrem) (mrem) 100 m (330 ft) 1.05 102
( 1400 m (4500 ft) 0.029 2.9 1010 m (1 mi) 0.024 0.24 10.100 m (10 mi) 2.4E-03 0.24
*The dosent urban boundary to the north is at least 4500 ft away.
Table D.2: Whole Body and Thyroid Doses from a One Element Accident (in the Direction of the Maxirnum Exposures,i.e., North of SNRS Enclosure) Maximum Credible Accident (One Element Accident) [ Distance * (East) Whole-Body Dose (mrem) Thyroid Dose (mrem) L 100 m (330 ft) 0.91 87.2 400 m (1300 ft) 0.084 8.12 910 m (3000 ft) 0.022 2.17 1610 m (1 mi) 0.011 1.05 10,100 rn (10 mi) 7.2E-04 0.005
*The closest urban boundary is at least 3000 ft east of snits.
Table D.3: Whole Body and Thyroid Doses from a One-Element Accident (in the Direction of Closest Base Boundary,i.e., East of SNRS Enclosure) I actual doses are expected to be much lower if such an accident occurrel. B.2.4.2 Multiple Eternent Failure in Air If the noncredible assumption is made that 123 elements in the core and in the tank storage racks are destroyed, this would result in a higher dose rate. A group of 123 elements would be representative of a complete spectrum of ele- { ments across the core. If the release fraction is weighted according to operating [ temperature, it has been shown [34) that peak doses to the North and East of , MNRC are those shown in Table D 4. These doses are relatively modest, well within the ANSI 15.7,10CFR Part 20 and 40 and 40CFR Part 01 limits. Peak Doses-123 Element Distance Whole Body Dose (mrem) Thyroid Dose (rnrem) 910 m 1.55 155 (east of MNRC) 1400 m 1.99 199 (north of MNRC) Table D.4: Doses for the 123 Element Case 23
L D.2.4.3 Other Accidents Credible accidents for TRIGA fueled reactors are evaluated in Ref. (f The rcactors are evaluated in the light of contemporary knowledge and the long - operating history of this class of reactor. Seven categories of accidents are analyzed: (
. Excece reactivity insertions e hietal-water reactions e lost, misplaced, or inadvertent experiment e hiechanical rearrangement of the core l
l e Loss-of-coolant accident ( - e Changes in fuel morphology and Zrilx composition e Fuel handling g Each accident is evaluated in terms of its off site radiological consequences. The " maximum credible accident" is found to be the fuel-handling or one-eternent accident, which has been addressad. In addition to these seven acci-dents, two other accidents are addressed here. The first is the seismic breaching of the reactor tank, and 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 [13)[14) (15)[16). In this Environmental Analysis, no attempt is made to analyze the risk of a breach of the tank. The discussion below simply assumes that a release of all tank water does take place. The event analysis for tank rupture [2] defined the potential effect of ground water contamination by the activation products hin-56 and Ar 41. From rei-crence data and an analysis (for Ar 41), these were estimated to potentially reach the groundwater at concentrations of approximately 1 10 -5 and 210-"Op Ci/cm 8, respectively. Assuming that the neutron flux will double the rate of prodruction of these isotopes (a conservative assumption), the updataed esti-mates of concer.trations are 210 -5 and 410-30p Ci/cm 8for hin-50 and Ar-41. The allowable concentration in water from the most recent version of 10 CFR 20 for hin 56 is 710-5p Ci/cm ;8 the h1NIlO level is within this guideline. No limit for effluent concentration in water is given for Ar 41 but clearly 10-30p 8 C1/cm is an extremely low level and represents negligible risk. Separately, if a tank failure occurred with complete draining of the coolant, it is estimated that up to 30 eternents in the core could experience cladding rupture following extended operation at 2 htW. This event is analyzed in Ref. [34); Table D.5. presents the predicted dose consequences. This event exhibits doses that, while moderate in scope, are still well within the ANSI 15.7 and 10 CFR 20 criteria. The group of elements assumed to be involved are at the highest end of the heat range, where the source terms and fission product release 24
]
4 Distance Whole Body Dose (mrem) Thyroid Dose (mrem) 910 m east 0.95 95 1400 m north 1.22 122 { Table D.5: Tank Failure Event (30 Peak Elements) fraction: are greater than the average for the core. Both of the multiple-element cases described in this section are considered incredible. B.2.5 Other Impacts This modification of h!NRC will cause no significant !ncrease in trafIlc, noise, solid or liquid westes, or air emissions. B.3 Resource Use The land is the pro, arty of the Federal Government and is already dedicated to aircraft maintenance activities. The reactor fuel currently exists in manu-factured form rod represents a minute fraction of available fuel. Only a small amount of water will be consumed for cooling and film processing purposes; water consumption will be a minor fraction (<1%) of base-wide water use. Per-sonnel requirements for modification and operation will be minor compared to the size of the available work force in metropolitan Sacramento. Thus, the project's requirements will have insignificant impacts on resource availability. B.4 Long-Term Aspects The expected useful life of this system is in excess of 20 years. The oldest TRIGA reactor has been operating 35 years, and numerous others have over 30 years of use without signs of age limitation. Decom.nissioning of sites for small reactors has boca accomplished successfully on several recent occasions. Com-panies such as Rockwell and the Institute of Resource hianagement Inc. have extensive experience in decommissioning such small research reactors at costs of about 10% of those for initial installation (constant dollars). The Nuclear Reg-ularity Commission has sponsored extensive studies showing that satisfactory decommissioning tecimologies are available [18). B.5 Cost-Benefit The benefits of the proposed h1NRC modificatica to the U.S. Government in-c'ude reducing risk of crashed aircraft and financial savings in the aircraft-maintenance project itself. The benefits specific to the local public should in-clude improved employment opportunities due to workload at hicClellan Air 25
[1 .- Force Base, and improved safety of niilitary aircraft. A potential benefit is im-proved safety of cleillan aircraft and ground population. Other benefits may include development of advanced not;d-state materials, enhancement of boron neutron capture therapy (medical) rosesrch, and production of medical isotopes. B.6 Conclusions h Nonradiological: There is essentially no nonradiological impact from increan-
~
ing the authorized reactor power level from 1 MW to 2 MW. Radiological: During normal operation, radiation exposures to reactor i h workers are administratively controlled to meet ALARA criteria. Routine ex-posures to the general public would be less than 2 mrem /yr. These exposures l will be far below the 10 CFR Part 20 limits of 5000 mrem /yr for workers and 500 mrem /yr for the general public and would also be below the 40 CFR Part G1 limit of 10 mrem /yr to the general public. Thus, although exposures can be expected to increase with increased power production, the higher values are all within industry rules, regulations, and guidelines. The NRC has endorsed a generic statement for research rcactors at or below
- 2 MW, whidi states that for such reactors there is no significant environmental impact (19). Rather than relying on this generic statement, McClellan Air Force Base has produced a site-specific emironmental assessment which shows that -
the operation of the facility at 2 MW will not pose a threat to the health and welfare of the on-site personnel or the general public off site in the vicinity of McClellan Air Force Isase. [ E 26
O l Section c Offices, Agencies, and Persons Consulted ' Person Position Office / Agency / Telephone Wade Ilichards Sh!.ALC/Till hicClellan AFB /916-643-1024 27 ) 4
O-
)
[ [ 1 Appendix: ' List of Preparers (Addendum 1) { Person Area Contact (_ Itichard M. hyer George R. Imel Accident Analysis ANI,W (208) 533 7008 Reviewer ANI,W (208) 533-7068 Darrell C. Cutforth Reviewer, Editor ANIrW (203) 533-7110 ( { ( { { { 28
p Bibliography [1] Stationary Neutron Radiography System (SNRS-2), " Source Selection," Sacramento Air Logistics Center, Directorate of hiaintenance, United States Air Force, December 1982. [2] Stationary Neutron Radiography System Final Safety Analysis Report, pre-pared by Argonne National Laboratory-West, Jan 1992. [3] Richard bl. Fryer, Gary L. Grasseschi, and George R. Imel, Addendum to the Safety Analysts Report, SNRS, Sept.1994. [4] Credible Accident Analyses for TRIGA and TRIGA-Fue'ed Reactors, NUREG/CR-2387, Pacific Northwest Laboratory,1982. [5] J. O. Blomeke and h1ary 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," Cutf-EES-A10801,1971. [7) National Oceanic and Atmospheric Administration, " Climates of the States," 2 Vols, Second Edition, Gale Research Co., Detroit,1980. [8] U. S. Geological Survey, Sacramento, California, 1:250,000 Scale, Land Use and Land Cover and Associated hiaps, Washington, D. C.,1979. [9] The World Almanac and Book of Facts, World Almanac,1994. [10] C. B. Hurt, " Natural Regions of the United States and Canada," W. H. Freeman and Co.,1967. [11) "hicClellan AFD Compatible Land Use Report," June 1983. [12] U. S. Soil Conservation Service, " Soil Survey of Sacramento Area," Wash-ington, D. C.,1954. [13] U. S. Geological Survey," National Atlas of the United States of America," Washington, D. C.,1970, p. 66, hiajor Recorded Earthquakes. 29 f!
\
s I l [14) J.11. Dennett," Foothills Fault Systems and the Auburn Dam," Calif. Ge-l ology, August 1978. { j l [15) Tousson R. Toppazada, at el., " Annual Technical Report - Fiscal Year 1980-1 1981, Prep . ration of Isoseismic hiaps and Summaries of Reported Effects ! for Pre-1900 Calif. Earthquakes," September 1981. l [16] Tousson R. Toppazada, " Annual Technical Report - Fiscal Year 1981 19S2, Areas Damaged by California Earthquake.s." (17] " Environmental Assessment for the Nondestructhe Inspection Facility hic-Clellan AFB," February 1984. [18] "Tecimology, Safety and Cost of Decommhsloning Reference Nuclear Re-search and Tbst Resctors," NUREG/CR 1750. [19) Staff Policy & Practice Statement,"Emironmental Considerations Regard-ing the Renewal of Licenses for Research Reactor," Dec.1980. [20] F. Daniels, " Outlines of Physical Chemistry." Wiley & Sons, New York, 1948,p.414. [21] W. J. llenderson and P. R. Tunnicliff, "The Production of N-16 and N 17 in the Cooling Water of the NRX Reactor," NSE,1958, pp.145-150. [22] R. L. hiitti and h!. H, Theys,"N 16 Concentrationsin EBWR," Nucleonics, hf arch 1961, p. 81. [23) U. S. Nuclear Regulatory Commission, " Regulatory Guide 1.109, Calcula-tion of Annual Doses to hian from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix 1.," 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 of Radioactive hiaterials from Nuclear Power Stations," NUREG/CR-2858 Pacific Northwest Laboratory, Novem-ber 1982. [25) U. S. Nuclear Regulatory Commission, " Regulatory Guide 1.145, Atmo-spherie Dispersion h!odels for Potential Accident Consequence Assessments 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 Dispersion hiodels for Potential Accident Consequence Assessments at Nuclear Power Plants," NUREG/CR-2260, NUS Corpora-tion, October 1981.
[27] U. S. Nuclear Regulatory Commission Title 10 Code of Federal Regulations (CFR), Parts 20 and 100. 30
e. s) - [28] Private Communications taken from NRAD Reactor Operating Data, July - 1984. 4--
- [29) Engineering Science,1983, " Final Report Instrllation Restoration Pro-gram, Phase II, ... Confirmation," W1.1., Jur.e, Prepared for the U. S.
Air Force, McClellen Air Force Base, Sacramento, CA. [30] ANSI /ANS-15.71977 (N379), Research Reactor Site Evaluation, American Nuclear Society,1978.- [31) Boerboom, 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. [34] Addendum 11 to the SNRS Safety Analysis Report, Operation at Two Megawatts, prepared by Argonne National Laboratory, April 1995. it [35) Environment Assessment for the Stationary Neutron Radiography System s 31 4}}