ML20209C832
| ML20209C832 | |
| Person / Time | |
|---|---|
| Site: | University of California-Davis |
| Issue date: | 03/31/1999 |
| From: | AIR FORCE, DEPT. OF |
| To: | |
| Shared Package | |
| ML20209C830 | List: |
| References | |
| NUDOCS 9907120233 | |
| Download: ML20209C832 (49) | |
Text
{{#Wiki_filter:_ McClellan Nuclear Radiation Center Reactor facility Sa)kty Analysis Report Rev.3 j i if Preparedby Research Reactor Safety Analysis Services Kennewick, WA McClellan Air Force Base Sacramento, California hrch1999 g f[>07*$0$$$$867
CHAPTER 10 EXPERIMENTAL FACILITIES AND UTILIZATION I
Chapter 10 - Valid Pages Rev.3 03/15/99 i Rev. 3 03/15/99 ii Rev. 3 03/15/99 10-1 Rev.2 04/03/98 10-2 Rev. 2 04/03/98 10-3 Rev. 2 04/03/98 10-4 Rev. 2 04/03/98 10-5 Rev. 2 04/03/98 10-6 Rev. 2 04/03/98 10-7 Rev. 2 04/03/98 10-8 Rev. 2 04/03/98 10-9 Rev. 2 04/03/98 10-10 Rev. 2 04/03/98 10-11 Rev. 3 03/15/99 10-12 Rev. 2 04/03/99 10-13 Rev. 2 04/03/99 10-14 Rev. 3 03/15/99 5' 10-15 Rev. 3 03/15/99 ~ 10-16 Rev. 3 03/15/99 10-17 Rev. 3 03/15/99 10-18 Rev. 3 03/15/99 10-19 Rev. 3 03/15/99 10-20 Rev. 3 03/15/99 10-21 Rev. 3 03/15/99 10-22 Rev. 3 03/15/99 10-23 Rev. 3 03/15/99 10-24 Rev. 3 03/15/99 10-25 Rev. 3 03/15/99 10-26 Rev,3 03/15/99 10-27 Rev. 3 03/15/99 10-28 Rev. 3 03/15/99 10-29 Rev. 3 03/15/99 10-30 Rev. 3 03/15/99 10-31 Rev. 3 03/15/99
Rev. 3 03/15/99 l TABLE OF CONTENTS 10.0 EXPERIMENTAL FACILITIES AND UTILIZATION . . . .. . 10-1 10.1 Summary Description . . . . . . .. .. .. .... .. . . ... 10-1 10.2 Beam Tubes and Beam Tube Shutter / Bulk Shield . .. ....... . 10-1 10.2.1 Beam ..... . . .. . . .. .. .... .. 10-1 10.2.1.1 Design Basis . ..... ... . .. ... . .. ... 10-1 10.2.1.2 Descdption . . .. .... . ... .... ... .. 10-1 10.2.1.3 Evaluation . . ... . . . .. . 10-6 10.2.2 Beam Tube Shutter / Bulk Shield .. .. 10-6' 10.2.2.1 Design Basis . . ... ... .... ... . 10-6 10.2.2.2 Description .. .. . .... . .. .. .10-7 10.2.2.3 Evaluation . . . . . . . . 10-7 10.3 Component Positioning Equipment . . ... . .. .... 10-7 10.3.1 Bay 1 Component Handling System . . .. . . 10-8 10.3.2 Bay 2 Component Handling System . .. . ..... .10-8 10.3.3 Bay 3 Component Handling System . . . . ... . .... ... 10-8 10.4 In-core Facilities .. .. . . . ..... . . ....'10-11 10.4.1 Centrallrradiation Facility . . ... . ... ... 10-11 10.4.1.1 Central Thimble . . .. .. .. . . .. 10-11 10.4.1.2 Graphite Thimble Plug . . . .. . ... .10-11 10.4.1.3 Aluminum Thimble Plug .. . . . . 10-11 10.4.1.4 Automated Central Sample Insertion Facility . . . 10-14 l 10.4.2 Experimental Tube Locations .. . . . 10-17 l 10.4.3 Pneumatic Transfer System . . .. . 10-17 l 10.4.4 Individual Fuel Element Positions . .10-20 l 10.5 Ex-core In-tank Facilities . .
.. . 10-20 l 10.5.1 Neutron Irradiator 10-20 l 10.5.1.1 Conditioning Well . .10-24 l 10.5.1.2 Exposure Vessel . . . 10-24 l 10.5.2 Silicon Doping Facility . . . .. . 10-24 l 10.5.3 Argon-41 Production Facility .. . .. 10-25 l 10.6 Experiment Review . . . ... 10-28 l 10.6.1 MNRC Experiment Coordinator (EC) . . . . 10-28 l 10.6.2 Office of Nuclear Licensing and Operations (TI-1) . . . 10-30 l 10.6.3 MNRC Experiment Review Board (ERB) . . . . . 10-30 l 10.6.4 Nuclear Safety Committee (NSC) . . . 10-30 l 10.6.5 Licensee (Responsible Commander, (SM-ALC/CC)) . . ... 10-30 l 10.6.6 Nuclear Regulatory Commission, Directorate of Non-power Reactors . . . . 10-31 l 10.6.7 Authorized Experiment .. . 10-31 l 10.6.8 Experiment Implementation . .. . 10-31 l
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'Rev. 3 03/15/99 l' LIST OF FIGURES 10.1 Reflector Beam Tube Location . . . . . . . . . . ........................ 10-2 10.2 In-Tank Beam Tube . . ........ .........:... ....... .. .... 10-3 10.3 MNRC Beam Tube and Biological Shield . . . . . . . . . . . . . . . . . . . . . . . . . . 10-4 10.4 Bays 1 and 2 Component Positioning System . ........ .............. 10-9 10.5 Bay 3 Component Positioning System .........................10-10 10.6 MNRC Typical In-Core Facilities . . . . . . . . . . . . . . . . . . . . . ...... 10-12 10.7 MNRC Central In-Core Irradiation Facility . . . . . . . . . . . . . . . . . . . . . 10-13. -l 10.8 Automated Central Sample Insertion Facility ..... ...... ......... 10 l 10.9 MNRC Pneumatic Transfer System ........................... .. 10-18 l - 10.10 MNRC Ex-Core In-Tank Facilities . . . . . . . . . ..... .., .. ...... 10-21 l ' 10.11 Neutron Irradiator - Plan View . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 0-2 2 l 10.12 Neutron Irradiato' r - Vertical View . . . . . . . . . . . ................. .. 10-23 l 10.13 Silicon Irradiation Facility - Vertical View . ........ . ... . . . . 10-26 l 10.14 Floor Lay-out for Argon-41 Production Facility . . . . . . . . . . . . . . . . . . . . 10 "
l 10.15. Experiment Review and Approval Process . . . .. ............... . 10 .
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10 1 Rev. 2 04/03/98 10.0 EXPERIMENTAL FACILITIES AND UTILIZATION 10.1 Summary Description The MNRC provides a broad range of radiographic and irradiation senices to the military and non-military sector. The facility presently provides four radiography bays and consequently four beams ofneutrons for radiography purposes. In addition to the radiography bays, the MNRC reactor core and associated experiment facilities are completely accessible for the irradiation of material. These irradiation senices include, silicon doping, isotope production (both medical and industrial) and neutron activation analysis (e.g., g - 31ogical samples). Although all four radiography bays are capable of using radiography film techniques, Bays 1,2, and 3 are equipped with, and will normally use, electronic imaging depces. All bays contain the equipment required to position parts for inspection as well as the radiography equipment. To meet facility use requirements, the reactor system and associated experiment facilities are designed to operate three shifts per day. 10.2 Beam Tubes and Beam Tube Shutter / Bulk Shield 10.2.1 Beam Tubes 10.2.1.1 Design Basis The design basis for the beam tubes is to provide a path for primary neutrons with minimum scattering and attenuation between the reflector and the radiography bays. 10.2.1.2 Description Four beam tubes spaced at 90* intervals around the base of the reactor tank penetrate the reactor graphite reflector and provide a direct path for neutrons to each of the radiography bays. The beam tubes are positioned tangentially with respect to the reactor core and are inclined (20* and 30*) with respect to the horizontal plane (Figures 10.1,10.2 and 10.3). Each of the four beam tubes is made up of three major sections: the in-tank section, the tank wall section, and the reactor bulk shielding section. The in-tank section of the beam tube (a replaceable aperture, made from neutron absorbing material and graphite housed in an air-tight aluminum container) is shown in Figure 10.3. This section is the most imponant part of the beam tube since it is part of the reactor core reflector, provides a source ofneutrons, and purifies and shapes the beam. It consists of a
-large graphite block with a 6 in. diameter hole bored along the beam centerline. The key elements within the bore are a graphite end plug which serves as a source of neutrons, a bismuth crystal which attenuates gamma rays and a boron-carbide aperture which shapes the beam. An aluminum spacer and lead-cadmium sleeve and shield are also located in the bore.
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10-5 Rev. 2 04/03/98 The sleeve and shield serve as neutron and gamma ray shie!ds. All of these components are contained in an aluminum housing that transitions into a 12-1/2 in. diameter circular cross section with a bellows assembly and flange with a bolt-on faceplate. A lead coated metal 0-ring forms the seal between the flange and the faceplate. The faceplate and the in-tank assembly have two tube fittings that connect to a helium supply and vacuum system. The entire unit is watenight and can be remotely removed and replaced from the tank top. The assembly mates with the tank-wall section of the beam tube to provide a water free path within the reactor tank for the neutron beam. Removal and replacement of the in-tank section of the beam tube has a small effect on the reactor core reactivity. Although the entire in-tank section of the beam tube is watertight, none of the components will react with water nor will they degrade should water enter the assembly. The tank-wall section of the beam tube consists of a 12-1/2 in. diameter pipe welded to the tank wall and a special flange welded to the core end (Figure 10.3). An aluminum container filled with 50 vol % boron frit and 50 vol % #9 lead shot by volume is located within the pipe section. The internal surface of the aluminum container is coated with gadolinium. The tank , wall section does not penetrate the tank wall and serves as a watenight container when i assembled as well as both a neutron and gamma shield. The gadolinium helps prevent scat- ; tered neutrons from reentering the beam. This section contains two tube fittings that are attached to a helium supply and evacuation system. The in-tank and tank-wall section flanges are held together by a two-piece bolted clamp. The clamp bolts can be remotely removed and replaced from the tank top. The bulk shielding section of the beam tube extends from the outside of the tank wall to the radiography bays (Figure 10.3). The housing for this section is a 20 in. diameter steel pipe and bellows assembly imbedded into the concrete. The bellows assembly provides flexibility for expansion and contraction. The pipe is in close proximity, but is not physically attached to the tank wall. Within the housing are a number of annular shaped aluminum containers filled with 50 vol % boron frit and 50 vol % #9 lead shot. The primary function of these ! materials and their design is to provide neutron and gamma ray shielding, help shape the beam, and prevent scattered neutrons from reentering the beam. The annular section next to the tank wallis permanently installed. The remaining baron / lead filler sections can be removed and replaced with units of different internal diameters if the beam size (cross section) needs to be changed. The two annular containers at the exit of the beam tube into the radiography bay contain 100% boron frit and a Research Chemicals MHL-277, respectively. These elements are the final beam shapers and both are excellent neutron shields. Both assemblies can be replaced from the radiography bay. The inner surfaces of all containers in this section are also coated with gadolinium. The ends of the beam tubes are ciosed with aluminum plates. These plates are 0.60 inches for the beam tubes in radiography Bays 1,2, and 4, and 0.75 inches for the beam tube in radiography Bay 3,
i i Rev. 2 04/03/98 10-6 The amount of explosive material allowed in radiography Bays 1,2,3, and 4 is 3 pounds of TNT equivalent per bay. This explosives limit is supported by safq analyses performed by Southwest Research Institute (Reference 13.11) and by the MNRL (Reference 13.10). Actual results of these analyses show that the four radiography bays could each safely contain op to 6 pounds of TNT equivalent, provided the door tracks and suspension on Bays a 2 and 3 were strengthened. However, by establishing a limit of 3 pounds of TNT equivalent for each bay, only the beam tube cover plates specified in the previous paragraph are required. All three sections of the beam tube are equipped with two gas lines. These lines are attached to the helium supply and evacuation system and can be used to either evacuate or fill the tubes with helium to prevent degradation of the neutron beam. The helium supply and vacuum system has venting and/cr pressure controls to prevent over-prossurization of the beam tube (Section 9.2). There is very little, if any, Ar-41 formed in these beam tubes because of the absence of air. 10.2.1.3 Evaluation The beam tubes, by use of shaped rings and being sealed and void of air, provide a neutron path with minimum neutron scattering. The beam tubes do not penetrate the reactor tank wall, and therefore, do not increase the probability of tank leakage. The beam tube cover plates on the ends of the beam tubes, where they exit into the bays, provide closure and prevent pressure waves frem reaching the reactor core and damaging the fuel rhould the maximum allowable amount of explosives being radiographed detonate. It should be noted that supplemental shieldin3has been placed in the teactor bulk shield to compensate for the void volumes created by the beam tubes. 10 2.2 Beam Tube Shutter / Bulk Shield 10.2.2.I Design Basis The design basis for the beam tube shutter /shic!d is: (a) To attenuate the neutron radiation beam at the location where it exits into the radiography bay such that radiation levels in the radiography bay are as-low-as-reasonably-achievable; (b) . To proMde a fast-acting t.hermal neutron shutter so that radiography film exposure times and real-time imaging can be controlled. k 14 i s'
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10-7 Rev. 2 04/03/98 10.2.2.2- Description Each of the beam tubes has a bulk shield and shutter. These units are located adjacent to the radiography bay end of the beam tubes as shown in Figure 10.3, and serve two basic functions. First, they provide the biological shielding from reactor core neutrons and gamma rays when the beam is not being used and the radiography bays are occupied. Second, they . provide a means to start and stop the flow of thermal neutrons during radiography operations. The shield / shutter unit is moter-driven and can be positioned so that the bulk shield covers the beam tube or so that only the thermal neutron shutter is in the beam path. The bulk shield is a massive composite stmeture containing materials to thermalize fast neutrons, capture thermal neutrons, and shield against both direct and capture gamma rays. The bulk shield has an average density of 4.7 gm/cm8 and is made up of cement, boron carbide, limonite, and steel shot. Boron frit, approximately 1 in. thick contained in aluminum, is placed in front of the composite shield to attenuate thermal neutrons. This shield has been designed so that the surface radiation level on the radiography bay side where personnel will be working during reactor operation at 1 MW will be less than 1 mr/hr. The motor driye on , { the shield is controlled from the radiography control room or in the radiography bay. -
- Indicator lights in the radiography and reactor control room show the shutter position. There is an interlock system that prevents the shield from being moved from the closed position any time the radiography bay door is opened and the reactor is operating. Sections 9.6 and 11.1.5.1 contain a complete description of the shield, shield controls and interlocks.
The thermal neutron shutter is a rectangular aluminum can approximately 1 in. thick filled with boron frit. The shutter is air actuator-driven, and remotely controlled from the radi-ography control room. As far as radiation protection is concemed, it is not considered an integral part of the bulk shield. - 10.2 2.3 Evaluation The beam tube shutter / shield provides the necessary bic!ogical shielding to protect personnel working in the radiography bay from the intense source ofneutrons in the radiography beams. These shields limit the radiation levels, within the radiography bays, to less than 1 mr/hr at 1 MW are expected to do the same at 2 MW (Chapter 11). The boron frit shutter provides an effective means of controlling the ' low of thermal neutrons. 10.3 .Cannonentfositionine Equipment i The MNRC has three automated component positioning systems. The automated systems are located in Bays 1,2 and 3. Bay 4 is provided v.ith an inspection table and fixtures. l l
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Following are specific design features which have been included in the component handling system of Bay 1. Figure 10.4 shoves an elevation layout of the component handling system. This system is used to position large components. The maximum size component which can be inspected measures 32.5 ft long x 12.5 ft high and weighs 3800 lbs. The system consists of one can with fixtures to hold the components. The can is latched to the positioning system which provides five axes of motion. Large components are held with soecial fixtures which provide positive location of the component on the can. This fixtudng has been designed to hold the components at each end to eliminate suppon structures at the center of the component which would interfere with the radiograph. 10.3.2 Bay 2 Component Handling System Following are specific design features which have been included in the component handling system of Bay 2. Figure 10.4 shows an elevation layout of this system. The system is sized to handle parts weighing up to 1500 lbs and measuring up to 18 x 9 ft. The system in Bay 2 includes two carts which hold the components to be inspected. The cans have been designed to accept large part fixturing which is used on the Bay I cart. The carts are also equipped with adjustable fixturing to hold smaller pans. This fixture can accommodate four components at one time. This system provides the same degrees of freedom as provided in Bay 1. 10.3.3 Bay 3 Component Handling System following are specific design features which are included in the component handling system I of Bay 3. Figure 10.5 shows an elevation layout of this system. This system is sized to handle small pans up to 5 n x 5 ft and curved parts with curvatures up to 160 deg. For inspecting curved parts, more yaw motion is required in this bay than in the other bays. For this reason, this positioner is designed differently than the ones in Bays 1 and 2. This system does not use a cart. Instead, operators load the components onto the positioner ' in the inspection bay. To facilitate this, an adjustable frame has been provided which can be adjusted to support small parts. After the fixture is loaded, the system will position the component in the beam path. _l
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4 i 10 11 Rev. 3 . 03/15/99 l 10.4 in-core Facilitig The MNRC has been designed with multiple in-core irradiation facilities to facilitate a broad range of potential experimental activities. These facilities are depicted in Figure 10.6 and , consist of a central cavity, four experimental tube locations, the pneumatic transfer tube, and individual fuel element locations. Each of these facilities are individually described below. 10.4.1 CentralIrradiation Facility The central irradiation facility, located in the center of the reactor core, may contain either a 1 plug usembly (as described in Sections 10.4.1.1 through 10.4.1.3 and Figure 10.7) or a l movable sample can system (as described in Section 10.4.1.4). All parts are removable from l the reactor using undenvater tools. I 10.4.1.1 Central Thimble . The central thimble once installed in the reactor core will not be removed froin the reactor core unless it is to be replaced with another facility of similar dimensions and has been t I analyzed to show how it affects the overall operation of the reactor. The central thimble is ~i approximately 55 inches in length and 4.22 inches in diameter with an inside dimension of approximately 4.0 inches. The central thimble once in place passes through the upper grid plate, the lower grid plate and the safety plate. The bottom of the central thimble sits on the bottom of the reactor tank. An aluminum ring located approximately 24.5 inches from the bottom of the central thimble aligns with the bottom grid plate and prevents samples or the graphite thimble plug and aluminum thimble plug from dropping below the lower grid plate. There is a 1.5 inch hole in the center of the aluminum ring and twenty-four 1.0 inch holes in the lower 24 inches to allow cooling flow through out the central thimble. Aluminum shims have been added to the outer periphery of the central thimble in the fuel region. These shims align the central thimble and displace the water from the scallops of the fuel element locations in the B hex ring 4.25 inch hole. 10.4.1.2 Graphite Thimble Plug The graphite thimble plug is a graphite-filled sealed aluminum can having dimensions of 26.88 inches long and 3.95 inches in diameter with a central through hole of 2.25 inches. A 6 inch long aluminum pipe welded to the top of the graphite thimble plug allows the removal or installation into the central thimble. 10.4.1.3 Aluminum Thimble Plug The aluminum thimble plug is a 2 inch aluminum bar approximately 29 inches in length. The upper end of the plug has been machined so the fuel element handling tool can be used to insert or remove the plug.
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l Rev. 3 03/15/99 10-14 In addition, the four upper grid plate triangular shaped cut-out positions, described in Section 4.2.3, can be used for irradiation of samples up to 2.4 in. in diameter. l '10.4.1.4 Automated Central Sample Insertion Facility l l The automated central sample insertion facility, shown in Figure 10.8, allows samples to be l inserted into the reactor core (i.e. central facility) while operating the reactor at power. The l reactor operator controls the insertion and removal of samples from the central facility l through the use of a drive mechanism similar to the control rods. l l The central thimble is approximately 52 inches in length and 4.22 inches outer diameter with l an inside dimension of approximately 4.0 inches. The central thimble, once in place, passes l through the upper grid plate, the lower grid plate and the safety plate. Aluminum shims have l been added to the outer periphery of the central thimble in the fuel region. These shims align l the central thimble and displace the water from the scallops of the fuel element locations in l the B hex ring 4.25-inch hole. Two captive bolts attach the central thimble to the upper grid I plate. These bolts prevent the accidental removal of the facility when removing samples from l the central thimble. l l An 1100 aluminum slug located inside the cemral thimble is normally positioned in the l reactor core. The aluminum slug is 4 inches in diameter and 24.75 inches in length. This l voids the water from the central thimble when the sample can is removed from the thimble. i l l An orifice plate is located on the bottom of the central thimble. In the event the aluminum l slug releases from the locating holes and falls to the bottom of the central thimble, the rate of l descent will be less than the normal control rod drive speed. l l The sample can is approximately 30.5 inches long with an outside diameter of 3.99 inches l and an inside diameter of 3.75 inches. The can could be free flooding or dry and is used to 1 position samples for i; radiation in the reactor core. The positioning of samples can be l accomplished during full power reactor operations (i.e. 2 MW). During insenion into the l reactor core, and while in the reactor core, the assembly has the capability of being rotated. I l The drive mechanism has the same type of drive motor as the control rod drives, except the l model selected will have more torque. All other aspects of the motor and controller are l identical. l l There are two sets of controls, one in the reactor room and the other in the control room. l Normal operational control is from the reactor console where the reactor operators will treat I the insertion and removal of the sample can as ifit were a control rod. The reactor room l controls can only be enabled from the reactor console. The normal indicators are as follows: I
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Limit switches on the rack are used in the logic design to deterumn .nd of travel indications, l.
- stop driving limits and start /stop rotation of the carrier. =l 10.4.2 Experimental Tube Locations Four cutout sections have been provided in the upper grid plate to allow for future removal of groupings of thrr- Sei elements and the insertion of tubes for experiment placement.
Prior to installation or any tubes in these regions, appropriate analysis and experiment review will be required. 10.4.3 Pneumatic Transfer System + f , The MNRC Pneumatic Transfer System, shown in Figure 10.9, is designed to quickly l transfer individual specimens into and out of the reactor core. The specimens are placed in a small polyethylene holder, " rabbit," which in turn is placed into the receiver. The rabbit travels through aluminum rubing to the terminus at reactor core centerline, then returns along the same path to the receiver. Directional air flow moves the rabbit between receiver and terminus. A blower assembly supplies air to the system, and a solenoid valve directs air flow. Controls to operate the blower and so!enoid valve are wall-mounted adjacent to a hood which contains the receiver. The air flow design is such that the rabbit is never pushed but rather pulled from place to place, minimizing the possibility of fragments from a shattered rabbit becoming trapped in the terminus. The key system elements and their functions are described below. The" rabbit"is an enclosed polyethylene holder. Experiments are inserted into the rabbit and contained by a screw cap on one end. Available space inside the rabbit is approximately 0.625 in. in diameter and 4.5 in in length. The receiver positions the rabbit for transfer to the terminus and receives the rabbit after irradiation. An aluminum door retains the rabbit in the receiver during transfer operations. Two transfer lines connect the receiver to the terminus: one allows the rabbit to travel ' between the receiver and tenninuc, the other controls the air flow direction. The receiver is located in a stainless steel hood which encloses the area around the receiver - and prevents uncontrolled release of airborne radioactivity. The hood's exhaust fan maintains f r i h .-
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10-19 Rev. 3 03/15/99 l the hood at a negative pressure with respect to the surrounding room and maintains a hood face air velocity of approximately 150 ft/ min when the sash is open. The air to the fan passes through a prefilter, an absolute filter and exhausts to the facility stack. The hood is located in the preparation area and provides working space around the receiver for handling rabbits before and after irradiation: The terminus consists of two concentric tubes which extend into the reactor core. The inner tube is perforated with holes (which are smaller than the rabbit diameter). The bottom of the inner tube contains an aluminum spring snock absorber to lessen the impact of the rabbit when it reaches this end of the transfer line, which is approximately at the mid-plane of the core. When air flows to the tenninus, the capsule rests in the bottom of the inner tube; when air flows to the receiver, the capsule moves out of the inner tube by air flowing through the tube's holes. The outer tube supports the inner tube and prosides a path for the air to flow through. The outer tube bottom support is shaped like the bottom of a fuel element and can fit into any fuel location in the core lattice. Both tubes, which extend to the top of the reactor tank, are offset to reduce radiation streaming. A weight has been installed to counteract the buoyancy of the air-filled tubes and keep the terminus firmly positioned in the core. The 5 terminus can be removed from the core by releasing two couplings. Two 1.25 in. aluminum transfer lines form a loop with receiver and terminus. The " rabbit" transfer line provides a path for rabbit travel between the receiver and terminus while the
" air" transfer line directs air flow between receiver and terminus. Tubing bends are a minimum 2 fl radius, allowing clearance for the rabbit.
A solenoid valve directs flow through the transfer-line-loop sending the rabbit either to the terminus or to the receiver depending on valve position. When the solenoid valve is deenergized, rabbit transfer line air flows from terminus to receiver; when the solenoid valve is energized, rabbit transfer line air flows from receiver to terminus. Solenoid status (energized or deenergized) is indicated by red markings on the solenoid alignment rod. A two horsepower blower circulates air through the transfer lines. The blower draws filtered room air through the solenoid valve, transfer lines, and a High Efficiency Paniculate Air (HEPA) filter. The blower outlet goes to the facility exhaust system. The transfer systems' controls allow operations in either manual or automatic modes. In manual mode, the s'olenoid valve is activated by the operator; in the automatic mode, the solenoid valve is activated by the timer mechanism, sending the rabbit into the core when the timer starts and retrieving the rabbit after a predetermined time period. The blower is manually operated in either mode. The controls for the system are located in a box next to the hood.
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An interlock switch in the reactor control room provides the reactor operator with overall control of operation. The switch is interlocked to the power supply for the blower such that-the switch must be "ON" for the blower to operate.
- 10.4.4 Individual Fuel Element Positions Reactor grid positions vacant of fiael elements may be utilized for the irradiation of materials.
These in-core irradiation facilities or the positioning of a single experiment in a fuel element grid position shall meet the requirements of the Technical Specifications for design, safetyL evaluation, restrictions and approvals. 10.5 Ex-Core In-Tank Facilities l Ex-core in-tank facilities have been established as shown in Figure 10.10. These facilities include the neutron irradiator facility and multiple silicon doping fixtures. 10.5.1 NeutronIrradiator The Neutron Irradiator is used to expose experiments to a high energy neutron environment l with minimal thermal neutron and gamma radiation (Figures 10.11 and 10.12).' The Neutron Irradiator has four main components: a Conditioning Well, an Exposure Vessel, a Motor Drive Unit, and a Computer. The Conditioning Well is installed inside the reactor tank adjacent to the reflector and consists of boron nitride and lead (for shielding thermal neutrons and gammas respectively) encased in aluminum. The Exposure Vessel (EV) is lowered into the Conditioning Well for irradiation. The EV houses the experiment (s) and contains temperature probes for monitoring the EV internal temperature during irradiation. ' A 5-piece lead and boron nitride shield assembly placed on top of an assembled EV completes the shielding around the experiment (s). The Motor Drive Unit is mounted at the top of the reactor tank and rotates the exposure vessel to provide a uniform neutron flux distribution. The Computer is connected to the EV and the Motor Drive Unit to monitor temperature and control rotation respectively. The Conditioning Well and Exposure Vessel are described in funher detail below. 1 3
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-The Conditioning Well is installed adjacent to the core's graphite reflector in the reactor tank :
and is held vertically in place by a three wheeled stand which rests on the bottom of the tank. It is held laterally by the levering action of two arms with steel rollers lightly pressing against the reactor tank wall. No fasteners, nuts, or bolts are required to secure the well in place. The inner sleeve is approximately 9.5 in. in diameter and 12 in. deep and both the inner sleeve and outer casing are made of aluminum. The lead and boron nitride are completely enclosed between the inner sleeve and the outer casing. 10.5.1.2 Exposure Vessel The Exposure Vessel (EV) consists of three major componentsi the Main Body, the
- Cylindrical Cup, and tl'e 5-piece Detachable Upper Shield. The Main Body consists of a titanium top plate welded to a 48 in. titanium tube with a multi-pin electrical connector at the top. Attached to the bottom of the top plate are six titanium plates arranged in a hexagon; each plate is approximately 4 in. high and 3.5 in. wide with threaded holes for attaching experiments using aluminum screws and straps. Four Temperature Probes (Thermisters) with aluminum protective shells are attached to the interior of the hexagon. The probes are factory calibrated and are accurate in a range of 20* to 200* C to *l
- and have a life expectancy of 20 years with a 1/2" variance during that time. The probes are designed to either work or fail and therefore do not require re-calibration in their expected lifetime. The Cylindrical Cup is constructed of aluminum and covers the hexagonal plates enclosing and sealing the experiments. The cup is approximately 9 in. in diameter and 10 in. high. The cup's inside surface is lined with a gadolinium coating to absorb thermal neutrons. A thin -
sheet of aluminum protects the gadolinium coating and shields secondary radiation resulting from the neutron absorption in the gadolinium. The 5-piece Detachable Upper Shield is constructed using lead and boron nitride for shielding *and is completely encased in aluminum. The shield assembly is placed on top of an assembled EV ensuring the seams are overlapping by at least 45* and then anchored in place by a co!!ar to completely enclose the EV. 10.5.2 Silicon Doping Facility
.A typical silicon doping facility consists of 5 individual motor drive assemblies mounted as a group to the tank top and positioned over an assembly at the bottom of the tank that -
l positions irradiation canisters in locations adjacent to the reflector (Figure 10.10). The irradiation canisters containing silicon ingots have a recessed bottom section that fit over bayonets for positioning and have drive shans extending vertically to the motor drives. The: shan assembly has a cross pin that is positioned in a yoke attached to the motor shan. The weight of the drive shan and irradiation carister is carried by the yoke assembly. Each gear reduction motor and drive shaft assembly rotates the irradiation canister at a slow
' rotational speed for uniform irradiation of the silicon ingot. The motor drive and shan
10-25 Rev. 3 03/15/99 l assembly is protected from damage by a clutch mechanism in the event the shaft or irradiation canister binds or locks in position. A vertical view of the silicon irradiation facility installation is shown in Figure 10.13. l The silicon irradiation canisters provide a water environment for the silicon ingot and are designed to accomodate the removal underwater of an irradiated ingot. This handling procedure reduces irradiation exposures to individuals handling the ingots in the interest of the ALARA program. An underwater table having an adjustable work platform for venical positioning in the tank is also utilized in the handling and loading of the irradiation canisters. 10.5 3 Argon-41 Production Facility l l The Argon-41 Production Facility will produce 1-2 curies of"Ar for research and , l commercial use. The "Ar will be produced by introducing argon gas into a stainless steel l container located in one of the silicon irradiation positions (adjacent to the graphite reflector l and external to the reactor core - Figure 10.14). All the components containing activated l "Ar are located in the reactor room. l
, i I Argon gas from a commercial argon gas cylinder will supply the irradiation containerJ Afler l the irradiation container is pressurized (approximately 500 psig) to the desired level, the gas l cylinder will be isolated from the irradiation container. To produce the desired activity level l of"Ar the sample will be irradiated for appror.imately 24 hours. l l
After irradiation, liquid nitrogen is added to a Dewar. A remotely operated solenoid valve is l opened to pressurize the cooling coih above the liquid nitrogen bath. The Dewar is then l raised to cover the cooling coils and "Ar is cryogenically extricated from the irradiation l container. After extrication is completed, the solenoid valve from the irradiation container is l shut and another remotely operated solenoid valve is opened. This allows diffusion of"Ar l gas to the sample containers. The liquid nitrogen Dewar is lowered thus exposing the l cooling coils to room temperature. When that portion of the system between the cooling l coils and the sample containers has reached equilibrium, the sample containers will be l isolated and removed from the system. The coil is surrounded with a lead shield to minimize j the radiation exposure to personnel. l l A catch tank surrounds the Dewar to contain any liquid nitrogen escaping from the Dewar or l in the unlikely event of a total failure of the Dewar. l l Over pressure protection of the overall system is provided by several relief valves that vent to l an over pressure tank. The over pressure tank is protected by its own reliefvalve which l vents to the reactor room. The tank is located as high as possible in the reactor room. l l All piping and valves in the system are stainless steel.. Compression fittings or double-ended l shut-ofrquick connectors are used for all connections normally in contact with the "Ar. l t
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l Rev. 3 03/15/99 10-28 l The Argon-41 Production Facility consists of several different components with the major l components listed below: l l COMPONENT MATERIAL DESCRIPTION Irradiation Chamber 304 stainless steel The irradiation container is a 1000-ml sample cylinder with a working pressure of 1800 psig and a burst pressure of 6000 psig. It conforms to the " Shipping Container Specifications" from the U.S. Code of Federal Regulations, Title 49 or Bureau of Explosives Tariff No. BOE 6000 Overpressure Relief 304 stainless steel The adjustable proportional pressure relief l'a/res valves have a working pressure up to 6000 psig. When upstream pressure overcomes the force exerted by the spring, the popper opens, allowing flow through the valve. As the upstream pressure increases, flow through the valveincreasesproportionately. Cracking pressure is only sensitive to inlet pressure and is not affected by outlet pressure Over Pressure Relief Carbon steel 30 gallon tank Tank l l'alves 304 Stainicss steel Bellows scaled valves l Tubing 304 Stainless steci %-inch and %-inch i 10.6 Exneriment Review The MNRC experiment review and authorization process is described in MNRC-0027-DOC,
" Utilization of Technology and Industrial Support Nuclear Facilities," and in more detail in MNRC-0033-DOC, " Technology and Industrial Support Directorate Nuclear Facilities Experiment Review and Authorization Process." This process requires that any individual wishing to utilize the MNRC reactor experimental facilities submit an Experimenter Approval Request Form to the Office of Nuclear Licensing and Operations (SM-ALC/TI-1). Once submitted, the request is coordinated through the MNRC Experiment Review and Approval l Process as depicted in Figure 10.15.
10.6.1 MNRC Experiment Coordinator (EC) The MNRC Experiment Coordinator (EC) is the primary point of contact between the experimenter and the use of the MNRC experimental facilities. The EC reviews all forms submitted and ensures that all required information has been supplied and validated. The EC then forwards completed requests to the Office of Nuclear Licensing and Operations (TI-1).
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l' Rev. 3 03/15/99 10-30 10.6.2 Office of Nuclear Licensing and Operations (TI-1) The Office of Nuclear Licensing and Operations (TI-1) reviews submitted experimental requests and takes one of the following actions:
- If the proposed experiment poses no unreviewed safety questions based on facility experience or similar experiments, and falls within currently authorized experiments, approves the proposed experiment; or If the proposed experiment requires safety review, TI-l forwards the experiment package to the MNRC Experiment Review Board (ERB). The ERB reviews the experiment package and recommends either approval or disapproval of the experiment. Depending upon the ERB recommendation, TI-l either prepares the
. Nuclear Safety Committee Facility Use Authorization Form for review and approval, or returns the experiment package to the EC for resolution.
10.6.3 MNRC Experiment Review Board (ERB) 1 The Chairman of the MNRC Experiment Review Board (ERB) (Reactor Operations Supervisor) conducts the ERB in accordance with a written charter. The ERB is assembled as a working group that performs a technical evaluation of proposed MNRC experiments sent to them by TI-1. As a result of their technical evaluation, the ERB makes a recommendation to TI-l concerning the approval or disapproval of the experiment. 10.6.4 Nuclear Safety Committee (NSC) The Nuclear Safety Committee (NSC) is responsible to the licensee (SM-ALC/CC) for oversight of radiation safety and nuclear operations of the MNRC and operates in accordance with a written charter. Once TI-l submits Facility Use Authorization Form to the NSC for approval, the NSC reviews the experiment package and recommends approval to the licensee or sends the experiment package back to TI-l for resolution. The NSC will advise the licensee of any special nuclear safety concerns that the committee feels needs to be brought to that level for action or resolution. The NSC will also advise the licensee as to whether the experi nent falls within the existing MNRC Technical Specifications. 10.6.5 Licensee (Responsible Commander, (SM-ALC/CC)) Once the licensee receives the request, it can be processed in one of two ways: If the experiment falls within the existing MNRC Technical Specifications, the licensee authorizes the experiment via signature, and returns the NSC Facility Use Authorization Form to TI-1;
10-31 Rev. 3 03/15/99 l
- If the experiment falls outside the existing MNRC Technical Specifications, the licensee authorizes the experiment via signature signifying that the propored Technical Specification change (s) are to be sent to the Nuclear Regula:ory Commission, Directorate of Non-power Reactors, for final approval.
10.6.6 Nuclear Regulatory Commiss ca, Directorate of Non-power Reactors The Nuclear Regulatory Comndssion, Directorate of Non-power Reactors, has final authorization concerning changes to the existing MNRC Technical Specifications. Once the NRC has authorized the requested technical specification changes, it shall inform the licensee ofits action. Upon receipt of the NRC's written approval, the licensee returns the NSC Facility Use Authorization Form to TI-1. 10.6.7 Authorized Experiment Upon TI-l receiving the experiment authorization from the licensee, TI-l informs the Experiment Coordinator (EC) that experiments using the specific NSC Facility Use Authorization Form at the MNRC are approved to run. 10.6.8 Experiment Implementation The EC and experimenter work together from that point on to implement the actual performance of the MNRC approved experiment in accordance with the approved NSC Facility Use Authorization Form.
CHAPTER 11 RADIATION PROTECTION , AND WASTE MANAGEMENT PROGRAM l 4 i I l 1 l l
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Rev. 3 03/l'/99 i TABLE OF CONTENTS 11.0 RADIATION PROTECTION AND WASTE MANAGEMENT PROGRAM . Il-1 11.1 Radiation Protection . . ....... .. Il-1 11.1.1 Radiation Sources . ... .... .. . . . 11-1 11.1.1.1 Airborne Radiation Sources . . . .. . . Il-2 11.1.1.1.1 Argon-41 in the Radiography Bays ,... .. Il-2. I1.1.1.1.2 Production and Evaluation of Ar 41 in the Reactor Room . . . ... . . 11-4 i 1 11.1.1.1.3 Ar-41 from the Pneumatic Transfer System . . Il 11.1.1.1.4 Ar-41 Release to the Unrestricted Area . . 11-5 11.1.1.1.5 Production and Evolution ofN-16 in the j Reactor Room . .... .. , 11-7 11.1.1.1.6 Ar-41 from the Ar-41 Production Facility . 11-7 l 11.1.1.2 Liquid Radioactive Sources ... . .. .. ..... 11-8 ; 11.1.1.2.1 Radioactivity in the Primary Coolant . . ... I l-8 11.1.1.2.2 N-16 Radiation Dose Rates from Primary Cooling System Components . . .. .. . 11-10 11.1.1.3 Solid Radioactive Sources . . . ... . Il-10 11.1.1.3.1 Shielding Logic . . . . ... . Il-13 j 11.1.2 Radiation Protection Program . . . .. .. . ... 11-13 6 11.1.2.1 Organization of the Health Physics Branch . . . . . Il-13 ; 11.1.2.2 Working Interface Patween Heak.h Physics and React r Operations .... . . . . Il-17 11.1.2 3 Health Physics Pre adures and Document Control . . . . Il-18 11.'.2.4 Radiation Protec an Training . . Il-19 11.1.2.5 Audits of the Health Physics Program . .. . I1.20 11.1.2.6 Health Physics Records and Record Keeping . ., 11-21 j 11.1.3 ALARA Program . . . . Il-22 j 11.1.4 Radiation Monitoring and Surveying . . . . Il-23 l 11.1.4.1 Monitoring for Radiation Levels and Contamination . Il-23 1 11.1.4.2 Radiation Monitoring Equipment . .. ... . . Il-25 11.1.4.3 Instrument Calibration , . . 11-29 11.1.5 Radiation Exposure Control and Dosimetry .. 11-29 11.1.5.1 Shielding . . . . .. 11-30 11.1.5.2 Ventilation System . . . . . . . .. .. . .. 11-42 1 11.1.5.3 Containment . . . . .. Il-45 l 11.1.5.4 Entry Control - Radiography Bays and Demineralizer j Cubicle .. . . Il-45 11.1.5.4.1 Entry Control for Radiography Bays . . Il-45 i I L1.5.4.2 Entry Control for the Demineralizer Cubicle 11-46
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11-7 Rev. 2 54/03/98 1 l 11.1.1.1.5 Production and Evolution of N-16 in the Reactor Room i In addition to Ar-41, the other source of airborne radioactivity during normal operation of the MNRC reactor is Nitrogen-16 (N-16). N-16 is generated by the reaction of fast i neutrons with Oxygen-16 (0-16)in water passing through the core. The amount of oxygen ~ pres (nt in air, either in a beam path or entrained in the water near the reactor cow, is i insignificant compared to the amount of oxygen in the water molecule in the liquid state. Production of N-16 resulting from neutron interactions with the oxygen in air and air . entrained in the cooling water can therefore be neglected. - The cross-section energy threshold for the 0-16 (n.p) N-16 reaction is 9.4 MeV; however, the minimum energy of the incident neutrons must be about 10 MeV because ofcenter or 1
-{
mass corrections. This high energy threshold limits the production of N-16, since only about 0.1% of all f1 scion neutrons have ar. energy in excess of 10 MeV. Moreover, a single hydrogen scattering event will reduce the energy of these high-energy neutrons to below the necessary threshold. ( After N-16 is produced in the core regiot, it rises to the tank surface and forms a disc source which creates a direct radiation field near the top of the tank. Some of the N-16 is . j subsequently released into the reactor room. Calculations for the production and mixing of N-16 in the primary coolant and for the evolution of N-16 from the reactor tank into the reactor room air are presented in Appendix A. Radiation levels associated with the N-16 in the tank and in the reactor room air are also addressed as part of Appendix A. Without exception, the calculated N-16 concentrations and dose rates are very conserwtive because they do not assume use of the conventionalin-tank N-16 dihser system, which is present in the MNRC primary water circulation system. Since this ditruser system is used during all normal operation of the reactor, and is designed to signiGcantly delay the N-16 transit time to the upper rp3 ions of the tank, the 7.14 second N-16 half-life brings about considerable decay and a corresponding reduction in N-16 radiation levels at the tank surface and in the t reactor room itself. l Recognizing the conservatisms involved in the N-16 calculations, and assuming the di%ser I system is otT, it is possible to predict the dose rate from N-16 at the tank water surface at a I 2 MW power level. Using the technique shown in Appendix A. this value turns out to be ) approximately 1350 millirem per hour. This value agrees very well with the " diffuser off' l
'N-16 dose rate measurd at the surface of the tank water at the MNRC reactor and with N. '
16 dose rates measured at several other comparable 1 MW 11UGA@ reactors, after the 1 MW values were extrapolated to 2 MW (Section 11.1.5.1, Tab!e 11-7). , 1 Since operation with the diffuser offis not a normal mode ofoperation, it is more realistic to l estimate N-16 dose rates over the reactor tank with tLe difruser on. Calculation of these j dose rates would be difYicd!t without knowing the actual effect of the di&ser in the I presence of the new 2 MW coolant flow rates. Therefore, esdmates of N-16 dose rates are { based on extrapolations of actual dose rate measurements at about I foot and 3 feet over the q s ! l e. i/ 1 ii
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11 9 Rev. 3 3/15/99 j 11.1.1.2.1 Radioactivity in the Primary Coolant As mentioned above, the only significant liquid radioactive source at the MNRC is the reactor primary coolant. Radioactivity in this liquid source occurs due to neutron activation of Argon-40 in entrained air (creating Ar-41); neutron interactions with oxygen in the water > molecule (creating N-16); and neutron interactions with tank and structural components with subsequent transfer of the radioactivity into the primary coolant. Radionuclides such as Manganese-56 and Sodium-24 are common examples of waterborne radioactivity created in this manner. Tritium is also present in the primary coolant due to activation of D:0 and other mechanisms. As noted, other sources ofliquid radioactivity are not currently projected for the MNRC reactor system, and no radioactive liquid efUuents and no liquid wastes have been generated as part of the current 1 MW operatior < is anticipated that this situation will continue during the operation at 2 MW. Radionuclides and their concentrations in the primary coolant vary depending on reactor power, reactor operating time and time since reactor shutdown, assuming that other variables (e.g., the effectiveness of the water purification system) remain constant. To characterize the radioactivity expected to be present in the MNRC primary coolant at 2 MW, measured concentrations for the predominant radionuclides at 1 MW were adjusted to reficc: estimated equilibrium concentrations at 2 MW. These values are shown in Table 11-4. I I I Radionuclide . Halflife Projected Equilibritmi Concentration at 2 MW (uci/ml) Aluminum-28 '2.3 min 6.0 x 10~3 Argon-41 1.8 hr 3.0x1f Hydrogen-3 12 yr 6.5 x 10-3
- Magnesium-l' 9.46 min 4.0 x 10" J Manganese-56 2.58 hr 4.7 x 10" Nitrogen 16 7.14 sec 131**
Sodium-14 14.96 hr 2.6 x 10-3 i Table Il-4 Predominant Radionuclides and Their Projected Equilibrium Concentrations in the MNRC Reactor Primary Coolant at 2 MW (* maximum buildup in 20 years without any water audition; *
- calculated approximation based on water leaving the core - not a uniform concentration)
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ACCIDENT ANALYSIS
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Chapter 13 - Valid Pages Rev. 3 03/15/99 i Rev.2 04/03/98 13-38 Rev.2 04/03/98 i1 ' Rev. 3. 03/15/99 13-39' Rev.2 04/03/98 lii Rev. 2 04/03/98 13-40 Rev. 2 04/03/98 iv Rev. 2 04/03/98 13-41 Rev.2 04/03/98 v Rev. 3 03/15/99 13-42 Rev. 3 03/15/99 13-1 Rev. 2 04/03/98 13-43 Rev. 3 03/15/99. 13-2 Rev. 2 04/03/98 13-44 Rev.3 03/15/99 13-3 Rev.2 04/03/98 13-45 Rev. 3 03/15/99 ~ 13-4 Rev.2 04/03/98 13-46 Rev. 3 03/15/99 - 13-5 Rev. 2 04/03/98 13-47 Rev.3 03/15/99 13-6 Rev.2 04/03/98 13-48 Rev. 3 = 03/15/99 13-7 Rev. 2 04/03/98 13-8 Rev. 2 04/03/98 13-9 Rev. 2 04/03/98 13-10 Rev. 2 04/03/98 13-11 Rev. 2 04/03/98 - 13-12 Rev. 2 . 04/03/98 13-13 Rev.2 04/03/98 13-14 Rev.3 03/15/99 13-15 Rev. 2 04/03/98 - --. Rev. 2 04/03/98 13-16 .- 13-17 Rev. 2 04/03/98 13-18 Rev. 2 04/03/98 13-19 Rev. 2 04/03/98 13-20 Rev. 2 04/03/98 13-21 Rev. 2 04/03/98 , 13-22 Rev. 2 04/03/98 13-23 Rev,2 04/03/98 13-24 Rev. 2 04/03/98 13-25 Rev. 2 04/03/98 13-26 Rev.2 04/03/98 13-27 Rev.2 04/03/98 13-28 Rev. 2 04/03/98 13-29 Rev. 2 04/03/98 13-30 Rev.2 04/03/98 13-31. Rev. 2 04/03/98 32- ' Rev. 2 04/03/98 13 Rev.'2 04/03/98 13-34 Rev. 2 ' 04/03/98
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13-37 ~ .. Rev. 2 04/03/98
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) Rev. 2 04/03/98 TABLE OF CONTENTS 13.0 ACCIDENT ANALYSIS . .. ...... . . .. . .. . .. . 13-1 13.1 Introduction . . ... . . ........... ....... . .. .... . 13-1 13.2 Accident Initiating Events and Scenarios, Accident Analysis, and Determination of Consequences . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13-2 13.2.1 Maximum Hypothetical Accident . . .. ..... .. ... . 13-2 13.2.1.1 Accident Initiating Events and Scenario . . . . . ... . 13-2 13.2.1.2 Accident Analysis and Determination of Consequences . 13 13.2.2 Insertion of Excess Reactivity . . . . . . . . . . . . . . . . .. . .. 13-5 13.2.2.1 Accident Initiating Events and Scenario . ..... . . 13-5 13.2.2.2 Accident Analysis and Determination of Consequences . 13-6 13.2.2.2.1 Maximum ReactivityInsertion . ... .. . 13-6 13.2.2.2.2 Uncontrolled Withdrawal of a Control Rod . 13-9 13.2.2.2.3 Uncontrolled Withdrawal of All Control Rods 13-10 13.2.2.2.4 Beam Tube Flooding or Removal .. . 13-11 13.2.2.2.5 Metal-water Reactions .. .. 13-11 13.2.3 Loss of Coolant Accident (LOCA) . . . . . ... . .. 13-12 13.2.3.1 Accident Initiating Events and Scenario . ..... 13-12 13.2.3.2 Accident Analysis and Determination of Consequences .13-12 13.2.3.2.1 Pumping of Water from the Reactor Tank . 13-12 13.2.3.2.2 Reactor Tank Failure . ..... .. . . 13-12 13.2.3.2.2.1 Description ofECCS and Assumptions . . . 13-14 13.2.3.2.2.2 Spray Cooling . . .. 13-15 13.2.3.2.2.3 Air ~ Cooling . . . . .. 13 15 13.2.3.2.2.4 Assumptions Made for ECCS Operation . . . . ... . . 13-16 13.2.3.2.2.5 Performance of the ECCS . 13-16 13.2.3.2.2.6 Thermal Model for Natural Convection Air Cooling . 13-17 13.2.3.2.2.7 Reactor Core for LOCA . . . 13-18 13.2.3.2.2.8 Mixed Air Temperature in the Reactor Room . .. ... 13.l8 13.2.3.2.2.9 Results of the ECCS Calculations . .. .. . . .. 13-20 13.2.3.2.2.10 Cladding Stress Analysis . 13-24 13.2.3.2.2.11 Ground Water Contamination 13-24 13.2.3.2.2.12 Radiation Levels from the ; Uncovered Core . ... . 13-29 f 1
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Rev. 3 03/15/99 l 13.2.4 Loss of Coolant Flow . . ... .. ., . .. . . ?3-31 13.2.4.1 Accident Initiating Events and Scenario . . . . . . . . 13-31 13.2.4.2 Accident Analysis and Determination of Consequences . 13-32 13.2.4.2.1 Loss of Coolant Flow Without Immediate Operater Action . . . . .... .. . 13-32 13.2.5 Mishandling or Malfunction of Fuel . . . . . . . . . . . . . . . 13-32 13.2.5.1 Accident Initiating Events and Scenario .. .. . . 13-32 13.2.5.2 Accident Analysis and Determination of Consequences . 13-33 13.2.5.2.1 Single Element Fa: lure in Water . . . . . . 13-3 3 13.2.5.2.2 Fuel Loading Error . . . . . . 13-36 13.2.6 Experiment Malfunction . ... ..... ..... .. 13-37 13.2.6.1 Ac.,ident Initiating Events and Scenario . ..... . 13-37 13.2.6.2 Accident Analysis and Deterrr.inati6n of Consequences . 13-37 l 13.2.7 Loss of Normal Electrical Power . . ... .... .. . . . . 13-45 l 13.2.7.1 Accident Initiating Events and Scenario . . . . . 13-45 l 13.2.7.2 Accident Analysis and Determination of Consequences . 13-46 l 13.2.8 External Events . .. .. .... . ... . . 13-46 l 13.2.8.1 Accident Initiating Events and Scenario . . . 13-46 l 13.2.8.2 Accident Analysis and Determination of Consequences . 13-46 l 13.2.9 Mishandling or Malfunction of Equipment . ... . . . . 13-47 l 13.2.9.1 Accident Initiating Events and Scenarios . . . . 13-47 l 13.3 Summary and Conclusions . . .... . . . . .. . 13-47 LlST OF FIGURES 13.1 Prompt Negative Temperature Coefficient for TRIGA* Fuels . . 13-8 13.2 Axial Power Distribution for Fuel in Core 20E' . . . 13-19 13.3 Maximum and Average Fuel Temperature During Air Cooling Cycle for Various Spray Cooling Times ... . 13-21 13.4 Maximum and Average Fuel Temperatures for the Hottest Fuel Element as Funcuons of Time after End of Spray Cooling for Three Hours 13-22 13.5 Clad Strength and Applied Stress Resulting from Equilibrium Hydrogen Dissociation Pressure as a Function of Temperature . .... .. 13-25 13.6 Maximum and Average Fuel Temperatures for the Hottest Fuel Element as Functions ofTime after End of Spray Cooling for 3.7 Hours . . . . ... 13-26
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- Rev. 3 03/15/99 l 13.14 Engineering Science,1983, " Final Report Installation Restoration Program, Phase II;
...Confumation," Vol.1, June, Prepared for the U.S.' Air Force, McClellan Air Force Base, Sacramento, CA.
13.15 Safety Analysis Repon, TRIGA* Reactor Facility, University of Texas at Austin, May 1991. 13.16 10 CFR 50.59 Safety Analysis ofExplosive Limits for Radiography Bays 1,2,3, and 4, McClellan Nuclear Radiation Center, Sacrunento, CA.,1996. 13.17 Southwest Research Institute, " Safety Analysis To Determine Limiting Criteria for Explosives in Bay 3 of the McClellan Nuclear Radiation Center," September 1995. 13.18_ Liu, H. Ben, " Safety Analysis for the Central Irradiation Facility (CIF) at the MNRC", l Memorandum to Wade J. Richards, September 22,1998. l 13.19 Strohmayer, W.H., and M.G. Stamate!atos,"Probabilistic Assessment of the Airplane l
. ;^ Crash Risk for the McClellan Air Force Base TRIGA* Reactor," G.A. Technologies,-
March 1986. 9 5
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13-13 Rev. 2 04/03/98 Provisions to monitor for and co!!ect tank leakage have been incorporated into the facility design. Firtt, the tank is surrounded by cormgated metal. The cormgations provide a path to the bottom of the tank for any watu leakage from the walls. Second, a drain, see Chapter 5, within the bulk shield surrounds the bottom of the tank. This drain will collect any water that may leak from the tank walls or bottom. Third, a duct leads from the drain to Radiography Bay 1 and the exit of this duct is periodically monitored for water leakage. Ifleakage is detected, the water could be easily co!!ected at this point or diverted to the liquid holdup tank outside the building. Consequences of a slow tank leak would be minimal and would require collection and containment of the water which leaked from the tank. This would be easily accomplished by using the existing liquid effluent control system described above. Small tank leaks due to corrosion are normally repairable using conventional techniques for patching aluminum, and thus it is expected that a leak could be located and fixed before there would be any significant loss of water from the tank. An earthquake of much greater intensity than the Uniform Building Code Zone 3 earthquake appears to be the only credible mechanism for causing a large rupture in the tank, since the tank when supponed by its associated biological shield structure was designed (with an importance factor of 1.5) to withstand this magnitude of earthquake. Even if such an event is , assumed to cau's e very rapid loss of water while the reactor is operating at peak power; a reactor shutdown would be caused by voiding of water from the core, even if there were no sc ram. I A large mpture of the tank would obviously result in a more rapid loss of water than a leak i due to corrosion or a minor mechanical failure in the tank wall. The MNRC reactor tank has no breaks in its structural integrity (i.e., there are no beam tube protrusions or other { l discontinuities in the reactor tank surface). In' addition, the reactor core is below ground level. Thus the potential for most types ofleaks is minimized. Part of the 2 MW upgrade to the reactor included a new cavity (Bay 5) cut into the biological shield. This cut exposes the reactor tank wall below the reactor core level, and this introduces ; an increased possibility of draining water from the core area. While steps have been taken to ' minimize the probability of a rank mpture in this location, and it is believed that the likelihood of such a rupture is very low, an unplanned occurrence could nevenheless initiate such a i event. Therefore, an Emergency Cere Cooling System (ECCS) has been installed to cool the core until the fuel has decayed to a level where air cooling is adequate to maintain fuel
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temperatures below the design basis limit (see Chapter 6 for details of the ECCS design and ; operation). An analysis detailing the cooling capabilities of the ECCS is described in the sections which follow. This analysis does not postulate the occurrence of a particular initiating sequence of events leading to all finel eleinems in the core being uncovered Instead, it simply assumes that . the tank has ruptured and all the water is lost. Such an event has several different l 1 d
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l Rev. 3 03/15/99 13-14 consequences. First there is the possibility of fuel clad rupture should the fuel temperature exceed design basis values. This event is covered in the analysis that follows, and focuses on the action of the ECCS to prevent fuel temperatures from reaching safety limits. Second, there is a possibility of personnel exposure to radiation from the uncovered reactor core due to the direct beam from the core or from radiation scattered from the reactor room walls and ceding. Finally, there is a chance that the lost water could cause ground water contamination. Both of these latter events are also analyzed as part of the LOCA evaluation. 13.2.3.2.2.1 Description ofECCS and Assumptions A loss-of-coolant accident (LOCA) is postulated for the MNRC in which the reactor pool is rapidly drained of water during operation at 2 MW (it is assumed that the reactor has been running at 2 MW for an infinitely long time). Because the LOCA uncovers the core quickly, the fuel clad temperature in some of the centrally located fuel elements could exceed the l design basis temperature limit of 930 C after a period of at least 20 minutes. When the reactor tank water level drops below the normal operating range (typically a loss of
, approximately six (6) inches of water) a tank low-level alarm sounc's. This alens the operator - that action must be taken. Depending upon the rate of water loss, the suspected cause of the loss, and other considerations, several different actions may be taken by the operator in response to a reduction in the tank water level. One such action could be activation of the ECCS.
Upon activation of the ECCS, cooling water from the domestic water supply will be introduced into the reactor tank and maintained until the fuel no longer contains sufficient decay heat to present a threat to the fuel cladding or water is restored to a level above the core. If the tank water level has dropped to less than about two (2) feet above the core, water from the ECCS will be sprayed onto the top of the r,emaining water column above the core; however, if the tank water has dropped below or partially below core level, the ECCS water will be sprayed directly onto the core. During this time, the decay heat will be removed by the remaining tank water or by the water spray and the maximum fuel temperature will be reduced rapidly from an elevated operating temperature down to about 200*C and then gradually to 100*C with continued spray cooling. At the end of spray cooling, natural air convection will be established in the core. During this cooling phase, the temperature of the fuel will rise slowly over several hours to a maximum and then decrcase with continued air cooling. The maximum fuel and cladding temperature is controlled by the length of spray cooling and by the natural air cooling. Under the preceding conditions, no fuel cladding will be ruptured. The detailed components of the emergency core cooling system to be used to maintain fuel temperatures below the design basis limit are described in Chapter 6. Basically the system consists of a quick connect system for coupling to the domestic water supply, sensing devices to indicate the need to initiate emergency cooling water flow, a nozzle to distribute the coolant flow over the core, a chimney mounted above the core structure to provide a
13-41 Rev. 2 d4/03/98 As a result of the preceding analysis, a limit of 25 milligrams of TNT equivalent explosives is deemed to be a safe limitation on explosives which may be irradiated in facilities located inside the reactor tank. Irradiation oflarger quantities of explosives in the reactor tank is not allowed. However, safety analyses have been performed which show that three pounds of TNT equivalent explosives may be safely irradiated in radiography Bays 1,2, 3 and 4, provided the beam tube cover plates are at least 0.5 inch thick (Reference 13.16). Southwest Research Institute (SRI) completed a safety analysis to determine the maximum amount of TNT equivalent explosive allowable in radiography Bay 3, (i.e., the amount that will not cause failure of the beam tube cover plate and will cause only repairable strr ~al damage to the bay)(Reference 13.17). Bay 3 is the smallest in volume of all the rt graphy bays at the MNRC. The study concluded that Bay 3 can withstand a detonation of 6 pounds of TNT equivalent explosive with certain modifications. The study performed by SRI concluded that the Bay 3 door track must be strengthened. The recommended strengthening consists of welding three additional anchor bolt plates to the door track and bolting these
) plates into the wall with additional drilled anchor bolts. This strengthening assures that the : ~
door will respond in a ductile manner to an unexpected high blast load, absorbing the additional load with larger deflections, rather than responding in a brittle failure mode. The MNRC completed a similar study to deterraine the maximum amount of TNT equivalent explosives allowable in all radiography bays (Refencnce 13.16). This study concluded that Bays 1,2 and 4 can withstand a detonation of 6 pounds of TNT equivalent explosives without any damage provided the criteria in Table 13-14 are implemented in each bay, However, to meet category 1 protection requirements for 6 pounds of explosives, the west door of Bay 2 also requires modification by rneans of an additional wheel and post assembly. The analysis performed by the MNRC demonstrates that for 3 pounds of TNT equivalent explosives, no modifications are necessary to the radiography bay doors for Bays 1,2 or 4. These doors will also respond in a ductile manner. As a result of the above studies, it is concluded that installation of beam tube cover plates with the thicknesses shown in Table 13-14 and implementing an explosives limitation of 3 pounds of TNT equivalent for each of the four radiography bays will satisfy the safety limitations established by the two previous safety analyses. b
l Rev. 3 03/15/99 13-42
. BAU - f Coser Plate ~ Explosive Explosive Deflection 1 - Resistance i Ultimate?
_: 3 OThickhess : Location distance re' f(in)? d(psi); x A Resistance!.
?" ? ~ (nf : @ 0'2 , -
~W 7 ' MpsiP 1 0.60 UD = 100 13.00 0.294 43.6 108 2 0.60 UD = 100 10.40
- 0.353 52.3 108 3 0.75 UD = 100 8.80 0.433 125 168 4 0.60 L/D= 100 13.10 0.248 37 108 Table 13-14 Changes to Beam Tube Cover Plates (* Minimum distance from the beam tube cover plate to the explosive.)
l The Argon-41 Production Facility (see Chapter 10) can produce argon-41 in excess of the l amounts analyzed in Appendix A. However, if the system releases argon-41, the gas will be l contained in the reactor room and the existing reactor room ventilation system will be used in l recirculation mode to prevent the release of argon-41 to the environment by recirculating the l gas until it decays. The existing stack continuous air monitor will also be used to verify that l none has been released outside the MNRC boundary. l l If the system had a catastrophic failure and 4 Curies of argon-41 were released to the volume l of the reactor room, the argon-41 concentration in the reactor room would be 2 x 10-2 , l Ci/ml and the gamma dose rate in the reactor room would be approximately 22 R/hr (based l on a semi-infinite cloud, see following calculation). Personnel would be evacuated from the l reactor room and access would be restricted. The reactor room ventilation system (as l described in Chapter 9) would be operated in the recirculation mode for approximately one l day before the dose rate from argon-41 decays to less than 1 mR/hr. Therefore, the l argon-41 discharge limit defined in the MNRC Technical Specifications will not be exceeded l due to the recirculation mode of the reactor room ventilation system. l l Another potential accidents include failure of the irradiation canister due to over-l pressurization from the argon gas supply cylinder, since a new argon supply cylinder is l typically delivered at a pressure of 2200 psi and the canister is rated for 1800 psi. However, l this requires multiple failures and is considered non-credible: a) the operator would have to i violate an operational procedure, b) the regulator would have to fail, and c) at the same time l the pressure relief valve would have to fail. Also, another potential accident is that liquid j nitrogen could spill into the reactor tank, causing expansion of the water and expelling a i portion of tank water. To prevent this, a catch basin sunounds the Cold Trap, and the liquid l nitrogen is supplied through a pipe in the reactor room wall connecting the trap to a supply l container in the equipment room. A third accident could result if the pressure relief valve l became choked with supersonic flow; however, the flow rates are estimated to be less than l sonic as shown in the following calculation. - ^
13-43 Rev. 3 03/15/99 l I I-ARGON-41 CONCENTRATION IN, REACTOR ROOM Given- ,
- 1. Reactor room volume = 7.39 x 10' A8 (1)
- 2. 4 Ci Argon-41 in argon production rystem
- 3. D(y).3 = 0.25E,x (2) where D(y)s. = gamma dose rate from a semi-infinite cloud (rad /sec)
E, = average gamma energy per disintegration (Mev/ dis)
= 1.2936 MeV/ dis for Argon-41 (3) x ' concentration of gamma emitting isotope in the cloud (Ci/m')
Therefore:
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8 x = (4Ci)/(7.39 x 10 ft'X1 m'/35.314 n') = 1.91 x 10-8 Cihn' D(y).3 = 0.25E,x j
= (0.25X1.2936 Mer/dirX1,91 = 10-8 Ci/m') ?
= (0.0062 rads /secX3600 sec/hr)
= 22.24 rads /hr Since D=Df -
1 t = -(l/A)ln(D/D.) {
= -(TJln2)ln(D/D ) 1 1
Then fcr: 1 D = 1 mrad /hr i i l t = 41.8 hr/In2)1n(1 mrad /hr/22.240 mrad /hr) l = 26 hr I ( (1) See Figure 9.11 (2) Shleien, B., L Slaback, and B. Birky, Handbook of Health Physics and Radiological Health, Williams A [ Wilkins, Jamsary 1997, p. 439. (3) Nuclides and hotopes,14* cdition, Chart of the Nuclides, GE Nuclear Energy, p. 22. I l f
- j. ~ Rev. 3 03/15/99 13-44 I
i. I SONIC FLOW l Assume: Perfect Gas l
! Constants: Property Value Units !
l R 208 N-m/kg4egK l K(cp/cv) 1.67 . dimensionless Problem: Determine if the pressure relief valve will experience choking due to supersonic flow. I l Solution-First calculate the speed of sound in argon at 40 *C and -200 *C given that c = speed of sound in a l med;um = (IGTg,)* l-j e = [1.67 x 208(N-m/kg-degK) x (40+273)K w 1(kg-m/N-s:)]'
= 329.7327 m/s at 40 'C
,,.. l c = [1.67 x 208(N-m/kgdegK) x (-200+273)K x 1(kg nt'N-s2 )]*
= 159.2397 m/s at -200 *C l
Next calculate the velocity of the argon in the tubing at the pressure relief valve given volumetric flow l rate V = velocity = area. l From technical data on valve, assume V = 1 fP/ min, based on air and relief at 1125 psi. l l V = 1 fP/ min x (12 in111)' = (2.54 cm/in.)' x 1 min /60 sec = 471.9474 cm'/sec l area = nr8 = 3.14 x (0.18 inJ2)2 = 0.025434 in.' (based on % in. tubing) l
= 0.16409 cm:
I l velocity = V/ area = 2876.15 cm/sec = 28.7615 m/s , I mach number = velocity /c = 0.180618 at -200 'C l = 0.087227 at 40 'C l -- .
Conclusion:
Oas velocity at the relief valve is less than the speed of sound in argon and therefore should not l ' experience choking at the valve l Reference Zucher, Robert D., fundamentals of Gas Dynamics, Weber Systems, Incorporated,1977, pp. 89,' [ l30 133,375. I l .; Another potential accident involves the automated central sample insertion facility (see
' l Chapter 10) since it may be considered'simihr to a control rod / Therefore, consider three
- l.-potential scenarios for an uncontrolled reactivity insertion analogous to the Uncontrolled j 1 Withdrawal of a Control Rod (see Section 13.2.2.2.2). First, if the material in the' sample can
; { ' were of sufYiciently different worth than the aluminum cylinder, the sample can would cause 4 l reactivity changes in the ;same fashion as a control rod, and either operator error or lo mechanical failure could cause an uncontr'olled reactivity insertion.1 Secondlif the aluminum 4
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13-45 Rev. 3 03/15/99 l cylinder f ailed to engage upon the sample can's insertion, a water void would be created in j the central facility as the aluminum cylinder descended ahead of the sample can. Similarly, if l the aluminum cylinder failed to replace the can upon removal from the central facility a water l void would result. l l All three of the above scenarios can be bounded by the Uncontrolled Withdrawal of a l Control Rod analysis (Section 13.2.2.2.2). Specifically, the automated central sample l insertion facility must have less reactivity and must drive slower than the control rod l analyzed ($3.50 and 42 inches / minute, respectively). To that end, the reactivity of any l material in the sample can shall be measured at low power to verify its worth is not only less l than $3.50, but also less than 51.75, the reactivity limit for the Central Irradiation Facility l (based on the Technical Specification limit of $1.75 for the pulsed reactivity insertion). For l example, the worth of a silicon ingot in the previous 1 MW in-core ex'periment facility was l measured at $0.73 positive (vs. water, reference exp. #96-01,1/30/96, reactor run #2411). l The wonh of an aluminum cylinder vs. void and vs. water has been analyzed by computer l simulation (Reference 13.18). The most positive reactivity effect in the computer simulation l q is from Case 3 to Case 9, where the voided sample can is lowered 18 inches, resulting in an l l
- g l
increase of about 50.06. The most negative effect is from Case 3 to Case 12, where in an accident the sample can not only floods but'also the aluminum cylinder drops, resultini; in a l l
] !
decrease of about $1.76. Thus, the wonh of the sample can or the aluminum cylinder vs. l l water is less than $3.50, and also less than the most reactive control rod (for example, a l typical regulating rod worth is S2.57, measured 6/98). With respect to the drive mechanism, I the maximum drive speed is identical to the rod speed analyzed in Section 13.2.2.2.2. l Furthermore, in the event of failure of the aluminum cylinder to engage upon installation of l l l the sample can, the base of the Central Thimble is designed (by sizing the hole in the base) to l allow the aluminum cylinder to descend at no more than the analyzed 42 inches / minute. j Therefore, the accident analysis for Uncontrol, led Withdrawal of a Control Rod (Section l 13.2.2.2.2)is sufficient to bound any accident associated with the automated central sample l insenion facility since: a) the material in the sample can shall be measured and verified to be l 1 l less than $1.75 (half of the analyzed S3.50); b) the drive speed cannot exceed the analv J 42 l inches / minute; and c) the aluminum cylinder cannot fall uncontrolled faster ti.tn th . ,alyzed l l 42 inches / minute. I I Finally, physical impact on the fuel is considered non-credible since the sample can is always l ; contained in a guide tube or attached to a drive mechanism such that it is unlikely to drop l j onto the core (see description in Section 10.4.1.4). I 13.2.7 Loss of Normal Electrical Power I 13.2.7.1 Accident Initiating Events and Scenarios Loss of electrical power to the MNRC could occur due to many events and scenarios which routinely affect commercial power.
l Rev. 3 03/15/99 13-46 13.2.7.2 Accident Analysis and Determination ofConsequences Since the MNRC does not require emergency backup systems to safely maintain core cooling, there are no credible reactor accidents associated with the loss of electrical power. A backup power system is present at the MNRC which mainly provides conditioned power to the reactor console and control instrumentation. Therefore, the reactor will not automatically scram when there is a loss c,f normal electrical power. In fact, the backup power system is capable of providing electrical power for reactor control and various operational measurements for a period of time after loss of normal electrical power and until ; its battery power supply is exhausted. Loss of normal electrical power during reactor operations is addressed in the reactor operating procedures, which require that upon loss of normal power an orderly shutdown is , to be initiated by the operator on duty. The battery backup power will allow monitoring of I the orderly shutdown of the reactor and verification of the reactor's shutdown condition. l 13.2.8 External Events 13.2.8.1 Accident Initiating Events and Scenanos Hurricanes, tornadoes and floods are virtually nonexistent in the area around the MNRC d reactor. Therefore, these events are not considered to be viable causes of accidents for the reactor facility. In addition, seismic activity in Sacramento is low relative to other areas of California (Chapter 2). Seismic activity has already been mentioned in connection with postulated reactor tank damage in Section 13.2.3. The MNRC facility is surrounded by a security fenc9 and a physical security plan is continuously in force for personnel and activities inside the fence. The reactor site is located on a U.S. Air Force Base where base access and overall security is far stricter than the l civilian business and residential areas surrounding the base. Therefore, accidents caused by ! human controlled eveats which would damage the reactor, such as explosions or other l unusual actions, are considered to be of very low probability. Since the MNRC reactor is located at the edge of the runway at McClellan AFB, airplane crashes iavolving the reactor may potentially cause reactor damage. 13.2.8.2 Accident Analysis and Determination of Consequences A study of the probability of aircraft crashes which could cause reactor damage at the MhEC was conducted by GA Technologies as a part of the original Stationary Neutron Radiography l System Proposal (Reference 13.19). The conclusions show that the calculated reactor damage probability due to aircraft accidents is 5 x 10 per reactor year. This value was obtained using conservative assumptions and the "best estimate" value is expected to be considerably lower than 5 x 10 Safety analyses ofnuclear power reactors have generally
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13-47 Rev. 3 03/15/99 l concluded that a :eactor damage probability due to an aircraft accident which is less than 1 x 4 10 per year does not represent a significant contribution to the overall reactor risk. l, Therefore, it is concluded that no specific aircraft accident and no radiological consequences ! need to be considered for the MisiiC reactor. 13.2.9 Mishandling or Malfunction of Equipment 13.2.9.1 Accident Initiating Events and Scenarios No credible accident initiating events were identified for this accident class. Situations involving an operator error at the reactor controls, a malfunction oz loss cf safety related ! instruments or controls and an electrical fault in the control rod system were anticipated at ) the reactor design stage. As a result, many safety features, such as control system interlocks and automatic reactor shutdown circuits, were designed into the overall TRIGA* Control System (Chapter 7). TRIGA* fuel also incorporates a number of safety features (Chapter 4) which together with the features designed into the control system assured safe reactor response, including in some cases reactor shutdown. Malfunction of confmement or containment systems would have the greatest impact during the maximum hypotheticci accident (MHA), if they were used to lessen the impact ofsuch an 3 accident. However, as shown in Section 13.2.1, no credit is taken for confmement or ' containment systems in the analysis of the MHA for the MNRC reactor. Furthermore, no safety considerations at the MNRC depend on confinement or containment systems, although simple confinement devices like a fume hood might be used as part of normal operations. Rapid leaks ofliquids have been previously addressed in Section 13.2.3. Although no damage to the reactor occurs as a result of these leaks, the details of the analyses provide a more comprehensive explanation. l 13.3 Summary and Conclusions Chapter 13 of the Safety Analysis Report contains a conservative analysis ofmany different types of hypothetical accidents as they relate to the MNRC reactor and the surrounding 1 l environment. Beginning with the maximum hypothetical accident and continuing on through an entire array of other accidents, it has been shown that the consequences of such accidents will not result in occupational radiation exposure of the MNRC staff or radiation exposure of the general public in excess of applicable NRC limits in 10 CFR Part 20. Furthermore, there is no projected significant damage to the reactor as an outcome of the accidents evaluated, except the damage or malfunction assumed as part of the different accident scenarios analyzed. Details of the assumptions used for each accident scenario and the specific consequences of each accident are presented in the text of this Chapter. , I i
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